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Takashi Hirosawa - One of the best experts on this subject based on the ideXlab platform.
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melting temperature of mixed oxide fuels for fast reactors
Journal of Nuclear Science and Technology, 2002Co-Authors: Koichi Konno, Takashi HirosawaAbstract:Alternation of the melting temperature of irradiated mixed oxide (MOX) fuel in fast reactors with progress of Burnup was determined in relation to the actinide fractions contained and of the oxygen-to-metal (O/M) ratio. Based on ideal solution models of UO2-PuO2 and UO2-PuO2-Am2O3 systems and on correlations obtained of the measured melting with O/M ratio and with Burnup, an equation was derived expressing the estimated melting temperature T rev (K) as function of the factors mentioned above that affect the temperature: where X 1 is the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), X 3 the Burnup (GWd/t), and X 4 the O/M ratio (-) (not exceeding 2.00). Toward advanced stage of Burnup, the tendency of irradiated fuel melting temperature to lower with Burnup progress presented a tendency to level off its rate of descent.
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melting temperature of simulated high Burnup mixed oxide fuels for fast reactors
Journal of Nuclear Science and Technology, 1999Co-Authors: Koichi Konno, Takashi HirosawaAbstract:The melting (solidus) temperatures were measured for fuel simulating Burnups of 50, 90, 130, 170, 210 and 250 GWd/t (SIMFUEL) which were added non-radioactive soluble fission products (FPs) to unirradiated fast reactor MOX fuel. The melting temperatures for fuels of 250 GWd/t which were blended non-radioactive soluble FPs and irradiated fuels of 110.6 and 119.0 GWd/t were also compared to the SIMFUEL of 250 GWd/t. The melting temperature decrease of the SIMFUEL tended to saturate with increasing Burnup and the melting temperatures appeared virtually constant above 170 GWd/t. An equation for melting temperature was obtained from the proposed equation in the previous report by revising the coefficient of the fifth term: where is the expected melting temperature (K), X 1 the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), and X 3 the Burnup (GWd/t). After the temperature measurement, the radial distribution of eight FP oxide additives in SIMFUEL of 250 GWd/t was measured by X-ray m...
Koichi Konno - One of the best experts on this subject based on the ideXlab platform.
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melting temperature of mixed oxide fuels for fast reactors
Journal of Nuclear Science and Technology, 2002Co-Authors: Koichi Konno, Takashi HirosawaAbstract:Alternation of the melting temperature of irradiated mixed oxide (MOX) fuel in fast reactors with progress of Burnup was determined in relation to the actinide fractions contained and of the oxygen-to-metal (O/M) ratio. Based on ideal solution models of UO2-PuO2 and UO2-PuO2-Am2O3 systems and on correlations obtained of the measured melting with O/M ratio and with Burnup, an equation was derived expressing the estimated melting temperature T rev (K) as function of the factors mentioned above that affect the temperature: where X 1 is the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), X 3 the Burnup (GWd/t), and X 4 the O/M ratio (-) (not exceeding 2.00). Toward advanced stage of Burnup, the tendency of irradiated fuel melting temperature to lower with Burnup progress presented a tendency to level off its rate of descent.
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melting temperature of simulated high Burnup mixed oxide fuels for fast reactors
Journal of Nuclear Science and Technology, 1999Co-Authors: Koichi Konno, Takashi HirosawaAbstract:The melting (solidus) temperatures were measured for fuel simulating Burnups of 50, 90, 130, 170, 210 and 250 GWd/t (SIMFUEL) which were added non-radioactive soluble fission products (FPs) to unirradiated fast reactor MOX fuel. The melting temperatures for fuels of 250 GWd/t which were blended non-radioactive soluble FPs and irradiated fuels of 110.6 and 119.0 GWd/t were also compared to the SIMFUEL of 250 GWd/t. The melting temperature decrease of the SIMFUEL tended to saturate with increasing Burnup and the melting temperatures appeared virtually constant above 170 GWd/t. An equation for melting temperature was obtained from the proposed equation in the previous report by revising the coefficient of the fifth term: where is the expected melting temperature (K), X 1 the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), and X 3 the Burnup (GWd/t). After the temperature measurement, the radial distribution of eight FP oxide additives in SIMFUEL of 250 GWd/t was measured by X-ray m...
Hiroshi Sekimoto - One of the best experts on this subject based on the ideXlab platform.
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feasible region of design parameters for water cooled thorium breeder reactor
Journal of Nuclear Science and Technology, 2007Co-Authors: Sidik Permana, Naoyuki Takaki, Hiroshi SekimotoAbstract:The performances of a light water cooled thorium breeder reactor have been investigated. A feasible region of fresh fuel enrichment and moderator to fuel ratio (MFR) is found to satisfy the constrains of criticality, breeding, and negative void coefficient for several Burnups of discharged fuel. The equilibrium fuel cycle Burnup calculation has been performed which is coupled with the cell calculation. The MFR is changed to investigate its effect to the breeding capability and void reactivity coefficient profile for different average discharged Burnups. For moderated cases, the conversion ratio (CR) decreases with increasing Burnup and MFR. The ratio of fissile inventory in equilibrium core to the initial fissile loading (FIR) has the maximum value at certain Burnups depending on the MFR and its value increases with the decreasing MFR. Considering to the breeding capability of the reactor, for Burnups of equal to 30 GWd/t or higher, the MFR ≤ 0.3 is needed. For the larger MFR and lower Burnups, the void r...
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long life small candle htgrs with thorium
Annals of Nuclear Energy, 2007Co-Authors: Yasunori Ohoka, Peng Hong Liem, Hiroshi SekimotoAbstract:Abstract CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) Burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE Burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel ( 233 U, 232 Th)O 2 are investigated for a range of enrichment ⩽15%. Discharged fuel Burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better Burnup performance in terms of higher discharged fuel Burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.
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startup of candle Burnup in fast reactor from enriched uranium core
Energy Conversion and Management, 2006Co-Authors: Hiroshi Sekimoto, Seiichi MiyashitaAbstract:Abstract A new reactor Burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable Burnup reactivity control mechanisms such as control rods are not required. This Burnup strategy can derive many merits. The change of excess reactivity along Burnup is theoretically zero, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed along Burnup. Therefore the operation of the reactor becomes much easier than the conventional reactors especially for high Burnup reactors. The transportation and storage of replacing fuels become easy and safe, since they are free from criticality accidents. Application of this Burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. The initial core must be prepared with easily available materials. Actinides are simulated by enriched uranium with changing enrichment at each position, and fission products by niobium. The maximum value of enrichment is 13% well below 20%. The obtained effective neutron multiplication factor oscillates with Burnup, but the maximum change with time is only 0.0008.
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application of candle Burnup to block type high temperature gas cooled reactor
Nuclear Engineering and Design, 2004Co-Authors: Yasunori Ohoka, Hiroshi SekimotoAbstract:Abstract The CANDLE Burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a Burnup control rod can be eliminated, and several merits are expected. This Burnup may be realized by enriched uranium and burnable poison with large neutron absorption cross-section. With the fuel enrichment of 15%, gadolinium concentration of 3.0%, and fuel cell pitch of 6.6 cm, the CANDLE Burnup is realized with the burning region moving speed of 29 cm/year and the axial half-width of power density distribution of 1.5 m. When the concentration of natural gadolinium is higher, the burning region moving speed becomes slower and the Burnup becomes higher, though the effective neutron multiplication factor becomes smaller. When U-235 enrichment is higher, the effective neutron multiplication factor becomes larger, the speed becomes slower, and the Burnup becomes higher. When the pitch is wider, the effective neutron multiplication factor becomes larger, the speed becomes faster, and the Burnup becomes higher.
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CANDLE: The New Burnup Strategy
Nuclear Science and Engineering, 2001Co-Authors: Hiroshi Sekimoto, Kouichi Ryu, Yoshikane YoshimuraAbstract:The new Burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this Burnup strategy, distributions of fu...
Chris H Rycroft - One of the best experts on this subject based on the ideXlab platform.
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analysis of the pebble Burnup profile in a pebble bed nuclear reactor
Nuclear Engineering and Design, 2019Co-Authors: Bing Xia, Chris H Rycroft, Yushi Tang, Liguo Zhang, Qiuju Guo, Zaizhe Yin, Jianzhu Cao, Jiejuan TongAbstract:Author(s): Tang, Y; Zhang, L; Guo, Q; Xia, B; Yin, Z; Cao, J; Tong, J; Rycroft, CH | Abstract: © 2019 Elsevier B.V. In a pebble bed nuclear reactor, each fuel pebble draining through the core experiences a different amount of Burnup depending on the precise trajectory that it follows. Understanding the Burnup profile of pebbles is essential for reactor safety, as well as for fuel economy. Here, we introduce a method for constructing the Burnup profile based on performing a discrete element simulation of the pebble drainage, followed by a Burnup calculation in each individual pebble. This method is more accurate than previous approaches, and in particular it captures the extremal cases of pebble Burnup. We demonstrate the method using the geometry, neutron flux data, and thermal characteristics from the HTR-10 reactor being developed at Tsinghua University. We examine pebble Burnup during a single drainage cycle, and over multiple drainage cycles characteristic of normal reactor operation. Our results show that the presence of slow-moving boundary layers of pebbles near the reactor wall strongly influences the Burnup profile. We perform a systematic study where the pebble–pebble and pebble–wall friction coefficients are independently varied, and we show that the strength of the boundary layers is a complex interplay of these two parameters.
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analysis of the pebble Burnup profile in a pebble bed nuclear reactor
Nuclear Engineering and Design, 2019Co-Authors: Bing Xia, Chris H Rycroft, Yushi Tang, Liguo Zhang, Qiuju Guo, Zaizhe Yin, Jianzhu Cao, Jiejuan TongAbstract:Author(s): Tang, Y; Zhang, L; Guo, Q; Xia, B; Yin, Z; Cao, J; Tong, J; Rycroft, CH | Abstract: In a pebble bed nuclear reactor, each fuel pebble draining through the core experiences a different amount of Burnup depending on the precise trajectory that it follows. Understanding the Burnup profile of pebbles is essential for reactor safety, as well as for fuel economy. Here, we introduce a method for constructing the Burnup profile based on performing a discrete element simulation of the pebble drainage, followed by a Burnup calculation in each individual pebble. This method is more accurate than previous approaches, and in particular it captures the extremal cases of pebble Burnup. We demonstrate the method using the geometry, neutron flux data, and thermal characteristics from the HTR-10 reactor being developed at Tsinghua University. We examine pebble Burnup during a single drainage cycle, and over multiple drainage cycles characteristic of normal reactor operation. Our results show that the presence of slow-moving boundary layers of pebbles near the reactor wall strongly influences the Burnup profile. We perform a systematic study where the pebble–pebble and pebble–wall friction coefficients are independently varied, and we show that the strength of the boundary layers is a complex interplay of these two parameters.
Tetsuo Matsumura - One of the best experts on this subject based on the ideXlab platform.
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analyses of Burnup at plutonium spots in uranium plutonium mixed oxide fuels in light water reactors by neutron transport and Burnup calculations
Journal of Nuclear Science and Technology, 1997Co-Authors: Takanori Kameyama, Akihiro Sasahara, Tetsuo MatsumuraAbstract:Plutonium concentrations and Burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and Burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and Burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated Burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average Burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for Burnup at Pu spots. In the outer half v...
