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Journal articles on the topic 'Light Enriched Uranium (LEU) Assembly'

Author: Grafiati

Published: 5 June 2025

Last updated: 15 July 2025

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1

Hossain, Md. Imtiaj, Yasmin Akter, Mehraz Zaman Fardin, and Abdus Sattar Mollah. "Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code." Nuclear Energy and Technology 8, no. (1) (2022): 1–11. https://doi.org/10.3897/nucet.8.78447.

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A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study & scrutinize the characteristics of one of the VVER-1000 LEU & MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data library END

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2

Flores y Flores, Alain, Guido Mazzini, and Antonio Dambrosio. "Development of a MELCOR Model for LVR-15 Severe Accidents Assessment." Energies 17, no. 14 (2024): 3384. http://dx.doi.org/10.3390/en17143384.

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LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for mod

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3

Truong, Thinh, Heikki Suikkanen, and Juhani Hyvärinen. "Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER." Journal of Nuclear Engineering 2, no. 2 (2021): 207–14. http://dx.doi.org/10.3390/jne2020019.

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In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sec

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4

Tran, Vinh Thanh, Thanh Mai Vu, Van Khanh Hoang, and Viet Ha Pham Nhu. "Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions." Nuclear Science and Technology 9, no. 4 (2021): 16–26. http://dx.doi.org/10.53747/jnst.v9i4.134.

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The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The com

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5

Wight, Jared, Stéphane Valance, and Stefan Holmström. "Innovation and qualification of LEU research reactor fuels and materials." EPJ Nuclear Sciences & Technologies 9 (2023): 3. http://dx.doi.org/10.1051/epjn/2022051.

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Two projects within the Euratom Research and Training Programmes 2014–2018 and 2019–2020 are focused on the innovation and qualification of novel nuclear fuels for conversion from highly-enriched uranium to low-enriched uranium (LEU) and for securing the supply chain of EU research reactors into the future. The LEU-FOREvER project is drawing to a close and has made significant progress in developing and demonstrating the uranium-molybdenum fuel system, demonstrating the viability of a high-density uranium-silicide fuel for EU high-performance research reactors (BR2, RHF, FRM-II, JHR). This pro

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6

Tran, Vinh Thanh, Hoai-Nam Tran, Huu Tiep Nguyen, Van-Khanh Hoang, and Pham Nhu Viet Ha. "Study on Transmutation of Minor Actinides as Burnable Poison in VVER-1000 Fuel Assembly." Science and Technology of Nuclear Installations 2019 (November 3, 2019): 1–12. http://dx.doi.org/10.1155/2019/5769147.

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Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feasibility study on transmutation of minor actinides as burnable poison in the VVER-1000 LEU (low enriched uranium) fuel assembly. The VVER-1000 LEU fuel assembly was modeled using the SRAC code system, and the SRAC calculation model was verified against the MCNP6 calculations and the available published benchmark data. Two models of minor actinide lo

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7

Govindarajan, Srisharan G., Brian S. Graybill, Philip F. Makarewicz, Zhentao Xie, and Gary L. Solbrekken. "Assembly and Irradiation Modeling of Residual Stresses in Low-Enriched Uranium Foil-Based Annular Targets for Molybdenum-99 Production." Science and Technology of Nuclear Installations 2013 (2013): 1–9. http://dx.doi.org/10.1155/2013/673535.

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This paper considers a composite cylindrical structure, with low-enriched uranium (LEU) foil enclosed between two aluminum 6061-T6 cylinders. A recess is cut all around the outer circumference of the inner tube to accommodate the LEU foil of open-cross section. To obtain perfect contact at the interfaces of the foil and the tubes, an internal pressure is applied to the inner tube, thereby plastically and elastically deforming it. The residual stresses resulting from the assembly process are used along with a thermal stress model to predict the stress margins in the cladding during irradiation.

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8

Koltochnik, S. N., D. S. Sairanbayev, L. V. Chekushina, Sh Kh Gizatulin, and A. A. Shaimerdenov. "COMPARISON OF NEUTRON SPECTRUM IN THE WWR-K REACTOR WITH LEU FUEL AGAINST HEU ONE." NNC RK Bulletin, no. 4 (December 30, 2018): 14–17. http://dx.doi.org/10.52676/1729-7885-2018-4-14-17.

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WWR-K is the research tank-type light-water heterogeneous reactor. Reactor operation started in 1967 with enrichment36 % in uranium-235. In 2016 reactor conversion to low-enriched uranium fuel (19.7 % in uranium-235) was implemented with the VVR-KN-type fuel assemblies (FA). In view of reactor operation, compact configuration of the core was chosen, where, following fuel burning up, side water reflector is gradually changed by beryllium one. Besides, an amount of work elements of the reactor control and protection system is increased in the new reactor core. Following results of measurement of

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9

Khazhidinov, A. S., A. S. Akayev, and D. A. Ganovichev. "COMPUTATION OF A TEMPERATURE FIELD OF THE IVG.1M WCTC-LEU IN PTIMIZED AND ADVANCED MODELS." NNC RK Bulletin, no. 3 (September 30, 2019): 76–80. http://dx.doi.org/10.52676/1729-7885-2019-3-76-80.

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A test object is a water-cooled technological channel with low-enriched uranium fuel (WCTC-LEU # 24) of the IVG.1M reactor. A porous two-dimensional axisymmetric model of IVG.1M WCTC was designed using the Ansys Fluent computation program, available operating model of a fuel assembly (FA) was optimized, which was simplified to a segment of one fuel element (FE). Computation models have been checked through comparison of stationary computation results with experimental data of the P17-08 start-up. Designed models enable performing thermophysical computations in emergency situations for safety s

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10

Hossain, Md Imtiaj, Yasmin Akter, Mehraz Zaman Fardin, and Abdus Sattar Mollah. "Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code." Nuclear Energy and Technology 8, no. 1 (2022): 1–11. http://dx.doi.org/10.3897/nucet.8.78447.

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Abstract:

A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study &amp; scrutinize the characteristics of one of the VVER-1000 LEU &amp; MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data lib

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11

Martynenko, Ye A., D. A. Ganovichev, A. S. Akayev, A. S. Khazhidinov, and L. K. Zhagiparova. "ANALYSIS OF DESIGN ACCIDENT AT THE IVG.1M REACTOR WITH EJECTION OF ACTUATING DEVICE OF CONTROL AND PROTECTION SYSTEM." NNC RK Bulletin, no. 3 (September 30, 2018): 14–17. http://dx.doi.org/10.52676/1729-7885-2018-3-14-17.

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The paper presents analysis of thermal condition of a fuel assembly (FA) with low-enriched uranium (LEU) fuel of the IVG.1M reactor at design accidents due to failure of control and protection system (CPS). The accidents with spontaneous turning of one control dram as well as dram control system with peak efficiency were considered. The issue of research presented in the article was in conduction of nonsteady heat computation of reactor’s FA with double profiling of energy release throughout the height of the assembly and in time. Based on results of the computation research, charts of changes

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12

Nguyenn, Kien Cuong, and Hai Dang Vo Doan. "Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate." Nuclear Science and Technology 6, no. 1 (2021): 14–31. http://dx.doi.org/10.53747/jnst.v6i1.142.

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Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical a

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13

Nguyen, Nhi-Dien, Kien-Cuong Nguyen, Ton-Nghiem Huynh, Doan-Hai-Dang Vo, and Hoai-Nam Tran. "Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel." Science and Technology of Nuclear Installations 2020 (August 25, 2020): 1–11. http://dx.doi.org/10.1155/2020/7972827.

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The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle

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14

Mann, Neal, Zeyun Wu, and Mihai (Mike) G. M. Pop. "Molten Uranium Breeder Reactor (MUBR) and Its Development Steps." Nuclear Science and Technology Open Research 2 (October 1, 2024): 68. http://dx.doi.org/10.12688/nuclscitechnolopenres.17592.1.

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Background The Molten Uranium Breeder Reactor (MUBR) is a radical new mixed-energy spectrum breed and burn reactor concept. The MUBR is fueled with molten uranium metal fuel in large fuel tubes instead of thin fuel rods, and is cooled by circulating the molten fuel through a heat exchanger. The purpose of this research is to find MUBR configurations with a SCALE (RSICC request/license 203869) burnup at least 10 times greater than the initial fissile content. Methods A proprietary computer program uses parameters to generate MCNP (RSICC request/license 176034) or SCALE input files and initiate

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15

Mann, Neal, Zeyun Wu, and Mihai (Mike) G. M. Pop. "Molten Uranium Breeder Reactor (MUBR) and Its Development Steps." Nuclear Science and Technology Open Research 2 (March 5, 2025): 68. https://doi.org/10.12688/nuclscitechnolopenres.17592.2.

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Background The Molten Uranium Breeder Reactor (MUBR) is a radical new mixed-energy spectrum breed and burn reactor concept. The MUBR is fueled with molten uranium metal fuel in large fuel tubes instead of thin fuel rods, and is cooled by circulating the molten fuel through a heat exchanger. The purpose of this research is to evaluate MUBR configuration variations with SCALE (RSICC request/license 203869) to show that the results are robust and that the simulated burnup is at least 10 times greater than the initial fuel fissile content without any refueling and holds up even if some of the assu

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16

Hossain, Md Imtiaj, Abdus Sattar Mollah, Yasmin Akter, and Mehraz Zaman Fardin. "Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code." Nuclear Energy and Technology 9, no. (4) (2023): 215–25. https://doi.org/10.3897/nucet.9.91090.

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The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The k<sub>eff</sub> values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition,

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17

Karpovich, Gleb W., Yuri A. Kazansky, Nikita D. Vasechkin, and Kirill A. Bakhantsov. "Minor actinides transmutation in pressurized water reactors. 2. Using uranium and thorium fuel to burn minor actinides in a system with several VVER reactors." Nuclear Energy and Technology 10, no. 4 (2024): 281–87. https://doi.org/10.3897/nucet.10.144559.

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There is currently a consensus among the scientific and engineering community regarding the solution to the problem of minor actinides (MAs) formed in the process of nuclear power operation: MAs need to be converted to fission products during burnup in power reactors. Fast neutron reactors (BN, BREST) and molten salt reactors (MSR) are considered largely to this end. Despite the advantages of using fast reactors, there are no currently power unit designs with BN or BREST reactors available for commercial operation. The possibility of using VVER reactors for this purpose is rarely covered in sc

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18

Karpovich, Gleb W., Yuri A. Kazansky, Nikita D. Vasechkin, and Kirill A. Bakhantsov. "Minor actinides transmutation in pressurized water reactors. 2. Using uranium and thorium fuel to burn minor actinides in a system with several VVER reactors." Nuclear Energy and Technology 10, no. (4) (2024): 281–87. https://doi.org/10.3897/nucet.10.144559.

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There is currently a consensus among the scientific and engineering community regarding the solution to the problem of minor actinides (MAs) formed in the process of nuclear power operation: MAs need to be converted to fission products during burnup in power reactors. Fast neutron reactors (BN, BREST) and molten salt reactors (MSR) are considered largely to this end. Despite the advantages of using fast reactors, there are no currently power unit designs with BN or BREST reactors available for commercial operation. The possibility of using VVER reactors for this purpose is rarely covered in sc

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19

Li, Wei, Chaoran Guan, Houde Song, Xiang Chai, and Xiaojing Liu. "Numerical investigation of heat transfer characteristics of moderator assembly employed in a low-enriched uranium nuclear thermal propulsion reactor." Frontiers in Energy Research 10 (September 2, 2022). http://dx.doi.org/10.3389/fenrg.2022.875371.

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The design of a nuclear thermal propulsion (NTP) reactor based on low-enriched uranium (LEU) requires additional moderator elements in the core to physically meet the critical requirements. This design softens the core energy spectrum and can provide more thermal neutrons for the fission reaction, but the heat transfer characteristics between the fuel and moderator assembly are more complex. Aiming at the typical LEU unit design, the heat transfer mathematical model is established using the principle of heat flow diversion and superposition. The model adopts the heat transfer relationship base

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20

Keiser, Dennis, Jan-Fong Jue, Francine Rice, and Eric Woolstenhulme. "Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element." Frontiers in Energy Research 11 (March 17, 2023). http://dx.doi.org/10.3389/fenrg.2023.1106601.

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Low-enriched (LEU) U-ZrH fuel, with a235U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, &lt;20% 235U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography an

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21

Berry, Jessica, Paul Romano, and Andrew Osborne. "Upsampling Monte Carlo reactor simulation tallies in depleted LWR assemblies fueled with LEU and HALEU using a convolutional neural network." AIP Advances 14, no. 1 (2024). http://dx.doi.org/10.1063/5.0169833.

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Simulating nuclear reactor cores at the highest achievable spatial and energy resolution is critical in modeling these systems accurately. Increasing the resolution, however, can dramatically increase the memory and central processing unit time required to run simulations. A convolutional neural network was shown previously to accurately upsample tally results of simulated light water reactor assemblies fueled with fresh, low enriched uranium. Here, we show that a convolutional neural network can be used to upsample tally results in assemblies containing fresh and depleted fuel enriched from 1

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22

Pyeon, Cheol Ho, and Kota Morioka. "Uncertainty quantification of light-water-moderated and light-water-reflected cores with highly-enriched uranium fuel at Kyoto University Critical Assembly." Journal of Nuclear Science and Technology, January 9, 2022, 1–9. http://dx.doi.org/10.1080/00223131.2021.2017371.

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23

Pyeon, Cheol Ho, and Kota Morioka. "Monte Carlo analyses of light-water-moderated and light-water-reflected cores with highly-enriched uranium fuel at Kyoto University Critical Assembly." Journal of Nuclear Science and Technology, August 26, 2021, 1–9. http://dx.doi.org/10.1080/00223131.2021.1961636.

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24

Noah, Olugbenga O., Johan F. Slabber, and Josua P. Meyer. "Introducing Passive Nuclear Safety in Water-Cooled Reactors - Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres." Journal of Nuclear Engineering and Radiation Science, November 14, 2022, 1–43. http://dx.doi.org/10.1115/1.4056239.

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Abstract The development of an accident tolerant nuclear fuel for water-cooled reactors would re-defined the status of these reactors from traditional active safety to passive safety systems. As a possible solution towards enhancing the safety of light-water reactors (LWRs), loose-coated particles of enriched uranium dioxide fuel with the ability to retain gaseous and metallic fission products in the case of a loss of cooling event can be introduced inside Silicon-Carbide cladding tubes of the fuel assembly (see Fig.1a & b). These coated particles are treated as a bed from where heat is tr

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