Academic literature on the topic 'Metallic uranium-plutonium fuel'
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Journal articles on the topic "Metallic uranium-plutonium fuel"
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Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, and Anatoliy Blokhin. "Lead reactor of small power with metallic fuel." Nuclear Energy and Technology 4, no. 2 (2018): 99–102. http://dx.doi.org/10.3897/nucet.4.30527.
Full textAbstract:
The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.
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Khorasanov, Georgiy, Dmitriy Samokhin, Aleksandr Zevyakin, Yevgeniy Zemskov, and Anatoliy Blokhin. "Lead reactor of small power with metallic fuel." Nuclear Energy and Technology 4, no. (2) (2018): 99–102. https://doi.org/10.3897/nucet.4.30527.
Full textAbstract:
The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (<sup>nat</sup>Pb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of <sup>208</sup>Pb, a low neutron-moderating isotope. An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO<sub>2</sub>) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of <sup>241</sup>Am increases from 0.359 to 0.536 barn, while the probability of the <sup>241</sup>Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.
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Sato, I., H. Furuya, T. Arima, K. Idemitsu, and K. Yamamoto. "Behavior of metallic fission products in uranium–plutonium mixed oxide fuel." Journal of Nuclear Materials 273, no. 3 (1999): 239–47. http://dx.doi.org/10.1016/s0022-3115(99)00071-9.
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Moiseenko, V., and S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste." Nuclear and Radiation Safety, no. 1(81) (March 12, 2019): 30–35. http://dx.doi.org/10.32918/nrs.2019.1(81).05.
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A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.
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Prathibha, T., K. Rama Swami, S. Sriram, and K. A. Venkatesan. "Interference of Zr(IV) during the extraction of trivalent Nd(III) from the aqueous waste generated from metallic fuel reprocessing." Radiochimica Acta 108, no. 7 (2020): 543–54. http://dx.doi.org/10.1515/ract-2019-3220.
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AbstractA metallic alloy of uranium–zirconium and uranium–plutonium–zirconium has been proposed as a fuel for fast reactors, owing to the possibility of achieving high breeding ratio in a short span of time. About 6–10 wt.% of zirconium has been added to these actinide fuels to increase the melting temperature and thermal-mechanical stability. Aqueous reprocessing of the spent metallic fuel generates the high-level liquid waste (HLLW) that contains about 60 % of the total zirconium from the fuel. In view of this, the extraction behavior of a trivalent representative ion, Nd(III) in the presence of Zr(IV) was studied from nitric acid medium using the candidate ligands proposed for trivalent actinide separation from HLLW, such as N,N,N′N′-tetraoctyldiglycolamide (TODGA), and N,N-di-octyl-2-hydroxyacetamide (DOHyA). The extraction was studied as a function of nitric acid concentration, zirconium and neodymium concentration and Nd(III) to Zr(IV) ratio. The findings of dynamic light scattering (DLS) and ATR-FTIR spectral techniques were used for understanding the complex chemistry of Zr(IV) extraction under different conditions. Poor extraction of nitric acid, smaller aggregate size, no third phase formation during the extraction of Zr(IV) and Nd(III) and other unique solvent properties favor the DOHyA molecule in n-dodecane as a solvent for partitioning of trivalent actinides from HLLW generated from metallic fuel reprocessing.
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Okunev, Viacheslav. "The concept of a fast reactor with liquid metal fuel in tungsten capsules." E3S Web of Conferences 411 (2023): 01013. http://dx.doi.org/10.1051/e3sconf/202341101013.
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The concept of a dual-purpose high-power nuclear reactor is proposed. One of the goals is the production of electricity, the other is the production of high-potential thermal energy. It is proposed to use liquid fuel based on waste uranium and plutonium extracted from the spent fuel of VVER reactors (purified from the 238Pu isotope). The fuel is in sealed tungsten capsules. Lead extracted from thorium ores is used to cool the reactor. The electrical power of the reactor is 3.3 GW. The layout of the reactor is identical to the BREST-OD-300 reactor under construction. The analysis of emergency modes from among ATWS (anticipated transients without scram) is carried out. The reactor is reliable and safe. The maximum temperature of a high-temperature reactor coolant is close to the boiling point of lead. By the nature of the change in the maximum temperatures of the core components, the reactor occupies an intermediate position between a reactor with solid metallic fuel and a reactor with cermets based on UN-PuN and metal uranium nanopowder.
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Nikitin, D. I., I. B. Polovov, and O. I. Rebrin. "ELECTROREFINING OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl–UCl<sub>3</sub> MELTS." Расплавы, no. 3 (May 1, 2023): 316–28. http://dx.doi.org/10.31857/s0235010623030052.
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The technology of pyrochemical processing of mixed nitride uranium-plutonium spent fuel, realizable at the experimental and demonstration energy complex of the site of the Siberian Chemical Plant, includes several operations with the ultimate goal of isolating the target fission products. It’s planned to use the electrofining of the products of the previous stage, metallized spent nuclear fuel, аs the penultimate stage of processing. It’s necessary to determine the processes and technological modes of electrolytic refining of alloys modeling the product of this stage of the processing module to implement electrolytic refining. This paper presents the results of electrofining of model alloys (simulating the raw materials of the stage of electrofining processing) on an enlarged laboratory electrolyzer. The initial parameters of uranium refining processes in melts based on 3LiCl–2KCl–UCl3 were determined earlier. The basic parameters of refining were the use of electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) and conducting experiments at 550°C. Uranium alloys containing palladium and neodymium were prepared by direct fusion of uranium metal, PdAP-1 grade palladium metal powders and neodymium metal (99.99%) in a medium of high-purity argon (99.998%). The data obtained showed that at a temperature of 550°C, cathode precipitates are typical dendritic forms of alpha-uranium in rhombic syngony with a tendency to needle formation with an increase in cathode current density. An increase in the company time and cathode current density leads to a decrease in the current output due to short-circuiting of the electrodes with cathode sediment needles or metal shedding from the cathode. The modes of the cathode process have been experimentally refined as a result of electrofining. When electrofining alloys U–Pd(1.59 wt %), U–Pd(1.62 wt %), U–Pd(1.54 wt %), U–Pd(1.58 wt %)–Nd(5.64 wt %), U–Pd(1.84 wt %)–Nd(6.49 wt %), U–Pd(1.79 wt %)–Nd(6.54 wt %), uranium cathode precipitates were obtained, which were subjected to chemical analysis, which showed the high purity of the resulting metallic uranium, as well as the absence of metallic palladium and molybdenum in it. The palladium purification coefficient exceeds 5000, the neodymium purification coefficient exceeds 1000, which meets the requirements for purification from fission products at this stage of pyrochemical processing of spent fuel. Palladium accumulates in anode slime, while the bulk of neodymium passes into the molten electrolyte.
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Kuzina, Yu, D. Klinov, G. Mikhailov, A. Sorokin, and V. Alekseev. "COMPLEX OF EXPERIMENTAL FACILITIES FOR DESIGN AND SAFETY JUSTIFICATION OF FAST REACTORS WITH LIQUID METAL COOLANTS." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 4 (2021): 172–94. http://dx.doi.org/10.55176/2414-1038-2021-4-172-194.
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To substantiate the safety and characteristics of fast reactors with liquid metal coolants, a complex of more than 20 stands of various profiles and purposes, well equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological stands, has been created at SSC RF - IPPE. In addition, JSC “SSC RF - IPPE” has a complex of fast physical stands, including two critical stands - BFS-1 and the world's largest physical stand BFS-2. The article presents the characteristics and the possibility of stands designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment elements and assemblies of operating and planned installations with fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space: - Hydrodynamic stands - “SGDI” (air), “V-2” (air), “SGI” (water), “V-200” (water), “GDK” (air). - Thermal-hydraulic liquid metal stands - “6B” (Na, Na-K), “AR-1” (Na, Na-K), “Pluton” (Na), “SPRUT” (Na, Na-K, Pb, Pb-Bi, water). - Technological liquid metal stands - “Protva-1” (Na), “Protva-2” (Na), “PUSHM” (Na), “Armatura” (Na), “IK-MT” (Na), “SID” (Na), “BTS” (Na), “TT-1M” (Pb), “TT-2M” (Pb-Bi), “LIS-M” (Li). A large-scale sodium test stands “SAZ” is under construction, which allows testing full-scale prototypes of equipment and its elements to substantiate existing and future projects of fast sodium reactors. The BFS complex of physical stands is the world's only experimental tool for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). The materials and construction of the stands allow simulating the core, breeding zones, reflectors and in-core shielding, as well as elements of fuel cycles and storage facilities for spent nuclear fuel and radioactive waste. Reactor materials of the stands (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.
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Heidet, F., and J. Roglans-Ribas. "CORE DESIGN ACTIVITIES OF THE VERSATILE TEST REACTOR – CONCEPTUAL PHASE." EPJ Web of Conferences 247 (2021): 01010. http://dx.doi.org/10.1051/epjconf/202124701010.
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The Versatile Test Reactor (VTR) is a new fast spectrum test reactor being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. This includes neutron irradiation capabilities which would support alternate coolants including molten salt, lead/lead-bismuth eutectic mixture, gas, and sodium. The VTR aims at addressing most of the needs of the various stakeholders, which is primarily composed of advanced reactor technologists, developers and vendors, as well as a number of others interested parties. Design activities are underway targeting a first criticality date by 2026, with General Electric recently joining the project to contribute to the VTR plant design. Current efforts are focused on all aspects of the VTR design, with the core design being at the center of the initial steps. The VTR is currently proposed as a 300 MWth sodium-cooled fast reactor able to reach peak fast flux levels in excess of 4.0x1015 n/cm2-s (and total flux level of about 6.0x1015 n/cm2-s). In this configuration, it is using ternary metallic fuel with reactor-grade plutonium and 5% low-enriched uranium.
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Bondarenko, G. G., G. S. Bulatov, K. N. Gedgovd, D. Yu Lyubimov, and M. M. Yakushkin. "Effect of the electron decay of metallic fission products on the chemical and phase compositions of an irradiated uranium-plutonium fuel." Russian Metallurgy (Metally) 2009, no. 5 (2009): 426–30. http://dx.doi.org/10.1134/s0036029509050115.
Full textDissertations / Theses on the topic "Metallic uranium-plutonium fuel"
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Iizuka, Masatoshi. "Studies on electrorefining and electroreduction processes for nuclear fuels in molten chloride systems." 京都大学 (Kyoto University), 2010. http://hdl.handle.net/2433/120864.
Full textConference papers on the topic "Metallic uranium-plutonium fuel"
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Li, Xunzhao, Hongchun Wu, Liangzhi Cao, and Youqi Zheng. "A Neutronics Concept Design of Lead-Bismuth Cooled Accelerator-Driven System for Minor Actinide Transmutation." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16309.
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Pursuing a high minor actinide (MA) transmutation rate, this paper proposes a neutronics concept design of lead-bismuth (LBE) cooled accelerator-driven system (ADS) with burnup reactivity swing less than 1% and proton beam current smaller than 17mA. After a comparison with other types of fuels, Uranium-free metallic dispersion fuel (TRU-10Zr)-Zr* is selected to obtain a harder neutron spectrum to transmute MA. With a MA initial loading, the suitable proportion of initial Plutonium to transuranium element (TRU) is found around 33% to make sure that the burnup reactivity swing is less than 1%. The location of the spallation target is optimized to guarantee high external spallation neutron source efficiency and to lower proton beam current. For the subcritical system, initial effective multiplication factor is 0.97, and the thermal power is 1000 MW. For the accelerator, proton with energy of 1.5GeV and a parabolic spatial profile is provided by proton linac. It is demonstrated by the numerical results that the transmutation rate of MA is about 28% after 600 effective full power days (EFPD) while the support ratio for LWR units with the same power is about 46.