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Journal articles on the topic 'Low-enriched cermet-based uranium fuel'

Author: Grafiati

Published: 5 June 2025

Last updated: 15 July 2025

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1

Irkimbekov, Ruslan A., Aleksandr D. Vurim, Sergey V. Bedenko, Artur S. Surayev, and Galina A. Vityuk. "Neutron background of composite low-enriched uranium fuel of the IVG.1M research reactor." Nuclear Energy and Technology 8, no. (3) (2022): 167–72. https://doi.org/10.3897/nucet.8.93895.

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IVG.1M is a research pressurized water reactor designed to use high-enriched fuel. As part of the core conversion program, the reactor will be switched to a new low-enriched composite uranium fuel. Further operation of the reactor is determined by the availability of fresh fuel to replace the core after the next campaign and the possibility of ensuring safe storage of irradiated spent nuclear fuel (SNF) unloaded from the core. The SNF storage conditions are assessed in terms of ensuring nuclear and radiation safety. Radiation safety of the research reactor fuel storage is achieved, first of al

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2

Sannen, Leo, Sven van den Berghe, and Ann Leenaers. "Status of the Low Enriched Uranium Fuel Development for High Performance Research Reactors." Advances in Science and Technology 94 (October 2014): 43–54. http://dx.doi.org/10.4028/www.scientific.net/ast.94.43.

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Historically, uranium enriched to >90% 235U has been used for many peaceful applications requiring high fission densities such as driver fuels for research reactors. However, the use of high-enriched uranium or HEU (all enrichments >20% 235U are considered HEU) for civil applications, is considered a proliferation concern. Since the 1970's, efforts are being devoted to the conversion of research reactors operating on HEU to alternative fuels using uranium with enrichment below 20% or LEU. These efforts imply the development of high-density LEU fuels to replace the low volume-density (mos

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3

Naymushin, Artem G., Yuri B. Chertkov, Vasily V. Kurganov, Ivan I. Lebedev, Svetlana A. Mongush, and Natalya V. Daneikina. "Feasibility Study of Using New Fuel Composition in IRT-T Research Reactor." Advanced Materials Research 1084 (January 2015): 306–8. http://dx.doi.org/10.4028/www.scientific.net/amr.1084.306.

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The results of simulation of IRT-T reactor conversion from highly enriched fuel to new perspective low enriched fuel based on uranium-molybdenum alloy are given. Main characteristics of reacting core with the use of highly enriched and low enriched fuel are calculated. It is shown that impact of using new materials in fuel composition remains neutronic and thermal hydraulic characteristics of the core at an acceptable level.

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4

Beyer, Gerd, Bernd Eichler, Teja Reetz, Rudolf Muenze, and Jozef Comor. "New head process for non-HEU 99Mo production based on the oxidation of irradiated UO2-pellets forming soluble U3O8." Nuclear Technology and Radiation Protection 31, no. 1 (2016): 102–8. http://dx.doi.org/10.2298/ntrp1601102b.

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All fission-based 99Mo producers worldwide are required to convert their 99Mo production processes from using highly enriched uranium to low-enriched uranium. At a recent IAEA meeting in Vienna, problems related to bottlenecks and target modification and optimization of low-enriched uranium-based 99Mo production processes were discussed. Ceramic UO2-pellets (as used in fuel) were excluded from the discussion with the argument that this material cannot be dissolved under practically applicable conditions. In this paper, we suggest transforming the non-soluble ceramic UO2 fuel-pellets into the U

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5

Schillo, Kevin J., Akansha Kumar, Kurt E. Harris, Yayu M. Hew, and Steven D. Howe. "Neutronics and thermal hydraulics analysis of a low-enriched uranium cermet fuel core for a Mars surface power reactor." Annals of Nuclear Energy 96 (October 2016): 307–12. http://dx.doi.org/10.1016/j.anucene.2016.05.035.

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6

Valance, Stéphane, Bruno Baumeister, Winfried Petry, and Jan Höglund. "Innovative and safe supply of fuels for reactors." EPJ Nuclear Sciences & Technologies 6 (2020): 40. http://dx.doi.org/10.1051/epjn/2019013.

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Within the Euratom research and training program 2014–2018, three projects aiming at securing the fuel supply for European power and research reactors have been funded. Those three projects address the potential weaknesses – supplier diversity, provision of enriched fissile material – associated with the furbishing of nuclear fuels. First, the ESSANUF project, now terminated, resulted in the design and licensing of a fuel element for VVER-440 nuclear power plant manufactured by Westinghouse. The HERACLES-CP project aimed at preparing the conversion of high performance research reactor to low e

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7

DeHart, Mark D., John Darrell Bess, and Germina Ilas. "A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels." Journal of Nuclear Engineering 4, no. 3 (2023): 602–24. http://dx.doi.org/10.3390/jne4030038.

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Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and softwa

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8

Yang, Kun. "Study of Uranium and Thorium Fuels in Breed-and-Burn Mode." Frontiers in Science and Engineering 3, no. 10 (2023): 14–22. http://dx.doi.org/10.54691/fse.v3i10.5663.

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The formation of fissile nuclei through breeding conversion is a hot topic in academic research, as it provides a continuous source of nuclear fuel for nuclear reactors. Fast neutron reactors, which have been extensively studied, use natural uranium or low-enriched uranium as the nuclear fuel, achieving burning after uranium-plutonium conversion. Thorium, as another potential fissile fuel, can theoretically be converted into nuclear reactor fuel through the thorium-uranium cycle. In this study, the physical evolution process of nuclear fuel in a specific core parameter is simulated using the M

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9

Solntsev, Vladimir A., Dmitry M. Soldatkin, and Vladimir N. Nuzhin. "Modernization of uranium-zirconium fuel rod of IVG.1M research reactor." Nuclear Energy and Technology 10, no. 2 (2024): 111–15. http://dx.doi.org/10.3897/nucet.10.130131.

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This paper describes the development of a dispersion-type uranium-zirconium fuel rod. Uranium is distributed in the zirconium matrix material in the form of axis-oriented fibers. The fuel rod is designed for the conversion of the IVG.1M research reactor (Republic of Kazakhstan) from highly enriched uranium (HEU) to low enriched uranium (LEU). The need for the HEU-LEU conversion arose in connection with Kazakhstan joining the program to convert research and test reactors to fuel with reduced enrichment (RERTR 2023). The study solves the problem of deformation of a low-tech U-Zr alloy (located i

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10

Nguyen, T. S., G. B. Wilkin, and J. E. Atfield. "Monte Carlo Calculations Applied to SLOWPOKE Full-Reactor Analysis." AECL Nuclear Review 1, no. 2 (2012): 43–46. http://dx.doi.org/10.12943/anr.2012.00017.

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Monte Carlo simulations are applied to the full-reactor analysis of the SLOWPOKE design. The temperature reactivity feedback calculated by using the MCNP code for either the high enriched uranium (HEU) or low enriched uranium (LEU) core is in good agreement with the experimental data, with a k-eff bias of +3.3 mk for a HEU core and +6 mk for a LEU core. Two methods that are based on existing third-party codes have been developed for use in core following: 1) MCNP (for the transport calculation) in conjunction with WIMS-AECL (for fuel burnup advancement), and 2) SERPENT (that combines both tran

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11

Bukina, O. S., Yu Yu Baklanova, M. N. Azbergenov, and M. А. Kuksa. "METHODOLOGY OF SELECTING CEMENT MATRIX COMPOSITION FOR IMMOBILIZATION OF IRRADIATED URANIUM-GRAPHITE FUEL." NNC RK Bulletin, no. 4 (December 30, 2024): 43–53. https://doi.org/10.52676/1729-7885-2024-4-43-53.

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The Impulse Graphite Reactor (IGR) is a unique nuclear facility in the world. The core of the research reactor is a stack of uranium-graphite blocks (fuel elements) enriched to 90 wt. % in 235U isotope. As part of the project on conversion of the IGR reactor to low-enriched uranium fuel, the specialists of the Institute of Atomic Energy (IAE) studied the possibility of immobilizing the first core in a cement matrix. Research into the immobilization process included both the formation of technical requirements for the uranium-graphite fuel matrix, determined by the conversion conditions, intern

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12

Pinem, Surian, Tagor Sembiring, and Tukiran Surbakti. "Reactivity insertion accident analysis during uranium foil target irradiation in the RSG-GAS reactor core." Nuclear Technology and Radiation Protection 35, no. 3 (2020): 201–7. http://dx.doi.org/10.2298/ntrp2003201p.

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Analysis of the steady-state and reactivity insertion accident is very important for the safety of reactor operations. In this study, steady-state and reactivity insertion accident analysis when the low enriched uranium foil target is irradiated in the reactor core has been carried out. The analysis is carried out by the best estimate method by using a coupled neutronic, kinetic, and thermal-hydraulic code, MTR-DYN. The MTR-DYN code is based on the 3-D multigroup neutron diffusion method. The cell calculations for the target are carried out by the WIMSD/5 and MTR-DYN code. After reactivity ins

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13

Solntsev, Vladimir A., Dmitry M. Soldatkin, and Vladimir N. Nuzhin. "Modernization of uranium-zirconium fuel rod of IVG.1M research reactor." Nuclear Energy and Technology 10, no. (2) (2024): 111–15. https://doi.org/10.3897/nucet.10.130131.

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This paper describes the development of a dispersion-type uranium-zirconium fuel rod. Uranium is distributed in the zirconium matrix material in the form of axis-oriented fibers. The fuel rod is designed for the conversion of the IVG.1M research reactor (Republic of Kazakhstan) from highly enriched uranium (HEU) to low enriched uranium (LEU). The need for the HEU-LEU conversion arose in connection with Kazakhstan joining the program to convert research and test reactors to fuel with reduced enrichment (RERTR 2023). The study solves the problem of deformation of a low-tech U-Zr alloy (located i

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14

Muhammad, Atta, Masood Iqbal, and Tayyab Mahmood. "Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core." Nuclear Technology and Radiation Protection 29, no. 4 (2014): 253–58. http://dx.doi.org/10.2298/ntrp1404253m.

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In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium). Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt

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15

Nor Azman, Muhammad ‘Adli, Nur Syazwani Mohd Ali, Muhammad Syahir Sarkawi, Muhammad Arif Sazali, and Nor Afifah Basri. "Nuclear fuel materials and its sustainability for low carbon energy system: A review." IOP Conference Series: Materials Science and Engineering 1231, no. 1 (2022): 012016. http://dx.doi.org/10.1088/1757-899x/1231/1/012016.

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Abstract World energy generation for electricity is still dependent on fossil fuels since it is more reliable and secure than the current intermittent renewable energy systems. Although the integration of renewable energy as an energy mix is in progress, still it could not be able to replace fossil fuels. Dependency on fossil fuels will not only contribute to severe climate change but will also degrade future generation quality of life. Hence, the solution to quandary is by integrating nuclear power plants with those of renewable energy such as solar and wind to meet the energy demand and to e

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16

Ho, Nguyen Thanh Vinh, Vinh Vinh Le, Nhi Dien Nguyen, et al. "Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code." Nuclear Science and Technology 8, no. 1 (2021): 10–16. http://dx.doi.org/10.53747/jnst.v8i1.78.

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VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems

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17

Hummel, Andrew J., and Todd S. Palmer. "Analysis of Multiple TRIGA-Based Molybdenum Production Reactor Cores Using a New Low-Enriched Uranium Target as Fuel." Nuclear Science and Engineering 183, no. 1 (2016): 149–59. http://dx.doi.org/10.13182/nse15-37.

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18

Lapin, Anton S., Aleksandr S. Bobryashov, Victor Yu Blandinsky, and Yevgeny A. Bobrov. "Analysis of system characteristics of a reactor with supercritical coolant parameters." Nuclear Energy and Technology 6, no. 4 (2020): 243–47. http://dx.doi.org/10.3897/nucet.6.60296.

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For 60 years of its existence, nuclear energy has passed the first stage of its development and has proven that it can become a powerful industry, going beyond the 10% level in the global balance of energy production. Despite this, modern nuclear industry is capable of producing economically acceptable energy only from uranium-235 or plutonium, obtained as a by-product of the use of low enriched uranium for energy production or surplus weapons-grade plutonium. In this case, nuclear energy cannot claim to be a technology that can solve the problems of energy security and sustainable development

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19

Lapin, Anton S., Aleksandr S. Bobryashov, Victor Yu. Blandinsky, and Yevgeny A. Bobrov. "Analysis of system characteristics of a reactor with supercritical coolant parameters." Nuclear Energy and Technology 6, no. (4) (2020): 243–47. https://doi.org/10.3897/nucet.6.60296.

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For 60 years of its existence, nuclear energy has passed the first stage of its development and has proven that it can become a powerful industry, going beyond the 10% level in the global balance of energy production. Despite this, modern nuclear industry is capable of producing economically acceptable energy only from uranium-235 or plutonium, obtained as a by-product of the use of low enriched uranium for energy production or surplus weapons-grade plutonium. In this case, nuclear energy cannot claim to be a technology that can solve the problems of energy security and sustainable development

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20

C., D. Vu, Q. Thien T., V. Doanh H., D. Quyet P., T. Anh T.T., and N. Dien N. "Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel." Nuclear Science and Technology 4, no. 1 (2014): 70–75. http://dx.doi.org/10.53747/jnst.v4i1.216.

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In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75% 235U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron spectrum parameters which are used in k0-NAA such as thermal neutron flux (fth), fast neutron flux (ffast), f factor, alpha factor (a), and neutro

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21

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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22

Ebert, Elena L., Andrey Bukaemskiy, Fabian Sadowski, Steve Lange, Andreas Wilden, and Giuseppe Modolo. "Reprocessability of molybdenum and magnesia based inert matrix fuels." Nukleonika 60, no. 4 (2015): 871–78. http://dx.doi.org/10.1515/nuka-2015-0124.

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Abstract This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA)-oxide (MA = minor actinide) fuel within a metallic 92Mo matrix (CERMET) and a ceramic MgO matrix (CERCER). Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L) and temperature (25-90°C). The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration

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23

Shannon, Colin, Paul Chan, and H. W. Bonin. "CONCEPTUAL DESIGN OF AN ORGANIC-COOLED SMALL NUCLEAR REACTOR TO SUPPORT ENERGY DEMANDS IN REMOTE LOCATIONS IN NORTHERN CANADA." CNL Nuclear Review 9, no. 1 (2020): 39–44. http://dx.doi.org/10.12943/cnr.2019.00002.

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Small nuclear reactors can offer safe, reliable, and long-lasting district heating and electrical power generation to remote locations in northern Canada. A conceptual design of an organic-cooled and moderated reactor based upon the SLOWPOKE-2 research reactor is proposed for potential employment in northern Canada. For viability, this design extends the SLOWPOKE-2’s power to 1 MWth. An added pump circulates the organic coolant, a partially hydrogenated terphenyl mixture known as HB-40, to facilitate greater heat transfer. The reactor incorporates the same low-enriched uranium dioxide fuel as

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24

Merk, Bruno, Anna Detkina, Dzianis Litskevich, Omid Noori-kalkhoran, Lakshay Jain, and Gregory Cartland-Glover. "A HELIOS-Based Dynamic Salt Clean-Up Study Analysing the Effects of a Plutonium-Based Initial Core for iMAGINE." Energies 15, no. 24 (2022): 9638. http://dx.doi.org/10.3390/en15249638.

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Nuclear technologies have strong potential and a unique role to play in delivering reliable low carbon energy to enable a net-zero society for future generations. However, to assure the sustainability required for its long-term success, nuclear will need to deliver innovative solutions as proposed in iMAGINE. One of the most attractive features, but also a key challenge for the envisaged highly integrated nuclear energy system iMAGINE, is the need for a demand driven salt clean-up system based on the principles of reverse reprocessing. The work described provides an insight into the dynamic in

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25

Johnson, L. H., and L. O. Werme. "Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel." MRS Bulletin 19, no. 12 (1994): 24–27. http://dx.doi.org/10.1557/s088376940004865x.

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The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form,

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Merk, Bruno, Anna Detkina, Seddon Atkinson, Dzianis Litskevich, and Gregory Cartland-Glover. "Evaluation of the Breeding Performance of a NaCl-UCl-Based Reactor System." Energies 12, no. 20 (2019): 3853. http://dx.doi.org/10.3390/en12203853.

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The energy trilemma forms the key driver for the future of energy research. In nuclear technologies, molten salt reactors are an upcoming option which offers new approaches. However, the key would be closed fuel cycle operation which requires sufficient breeding for a self-sustained long term operation ideally based on spent fuel. To achieve these attractive goals two challenges have been identified: achieving of sufficient breeding and development of a demand driven salt clean up system. The aim is to follow up on previous work to create an initial approach to achieving sufficient breeding. F

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27

Skakov, M. K., A. D. Vurim, V. S. Gnyrya, et al. "IN-PILE TESTS OF WATER-COOLED TECHNOLOGICAL CHANNELS WITH LOW-ENRICHED URANIUM FUEL WITHIN CONVERSION OF IVG.1M RESEARCH REACTOR." NNC RK Bulletin, no. 3 (September 30, 2018): 18–21. http://dx.doi.org/10.52676/1729-7885-2018-3-18-21.

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IVG.1M research reactor performs tests of WCTCs-LEU in order to determine whether channels correspond to technical requirements or not and to obtain experimental data necessary to take a decision concerning manufacture of a batch of standard channels and their insertion into reactor core based on results of tests and researches. The paper presents results of comparative assessment of technological parameters (flow rate, pressure, temperature) during in-pile testing of WCTCs-LEU and results of U235 burn-out calculation in WCTCs-LEU according to implemented start-ups of IVG.1M research reactor.

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Bourenane, A., L. Hamidatou, M. Dougdag, and M. L. Yahiaoui. "A Neutronic Study of A Low-Enriched Uranium-Fueled Microreactor Cooled with A Sodium Heat Pipe System Using The OpenMC Code." Atom Indonesia 50, no. 2 (2024): 151–57. http://dx.doi.org/10.55981/aij.2024.1318.

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The development of Small Modular Reactors (SMRs) represents a pivotal shift in nuclear technology, emphasizing enhanced safety, efficiency, and adaptability. This study examines Toshiba's MoveluX, an innovative micro-reactor, exemplifying advancements in reactor miniaturization suitable for limited spatial environments and hybridization with other energy sources. In this paper, the performance and safety of the MoveluX are rigorously evaluated using the OpenMC code, with an emphasis on critical parameters such as the effective multiplication coefficient and the reactivity worth of control devi

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29

Gattu, Vineeth Kumar, and William L. Ebert. "Multiphase Alloy Nuclear Waste Forms Developed for Pyrochemical U-10Mo Scrap Recovery Waste Streams." ECS Meeting Abstracts MA2022-02, no. 12 (2022): 772. http://dx.doi.org/10.1149/ma2022-0212772mtgabs.

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The US High Performance Research Reactor (USHPRR) Conversion program is developing an Al-clad Zr-bonded U-10Mo nuclear fuel foil to enable the conversion of research reactors to high-assay low-enriched uranium (HALEU) alloy fuel. Fabrication scrap from the fuel manufacturing will be electrorefined to recover the HALEU, which will generate waste streams consisting of Zr, Mo, and residual U retained in the stainless steel anode basket used in the electrorefiner. Four alloys were formulated with different relative amounts of Zr, Mo, and 316L-SS to represent the expected range of waste composition

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30

Betzler, Benjamin R., David Chandler, Thomas M. Evans, et al. "As-Built Simulation of the High Flux Isotope Reactor." Journal of Nuclear Engineering 2, no. 1 (2021): 28–34. http://dx.doi.org/10.3390/jne2010003.

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The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) is an 85 MWt flux trap-type research reactor that supports key research missions, including isotope production, materials irradiation, and neutron scattering. The core consists of an inner and an outer fuel element containing 171 and 369 involute-shaped plates, respectively. The thin fuel plates consist of a U3O8-Al dispersion fuel (highly enriched), an aluminum-based filler, and aluminum cladding. The fuel meat thickness is varied across the width of the involute plate to reduce thermal flux peaks at the radial edges of the fu

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31

Alhassan, S., S. V. Beliavskii, and V. N. Nesterov. "Investigative study of the radiation damage on fuel clad of miniature neutron source reactor using computational tools." Journal of Physics: Conference Series 2064, no. 1 (2021): 012103. http://dx.doi.org/10.1088/1742-6596/2064/1/012103.

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Abstract Core conversion requires some evaluation of the reactor safety. Changes to the reactivity worth, shutdown margin, power density and material properties are crucial to the proper functioning of the reactor. The focus of this article is to study the neutron flux distribution in the reactor core and radiation damage on candidate clads. The Ghana Research Reactor-1 (GHARR-1) operates at maximum power of 30 kW in order to attain a flux of 1.0× 1012 n·cm–2·s for the high enriched uranium core. Using the GHARR-1 core geometry, considering 348 fuel pins, the multiplication factor (Keff) is ca

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32

Dau, Duc-Tu, Nhi-Dien Nguyen, Kien-Cuong Nguyen, et al. "Kinetic parameters of the Dalat nuclear research reactor with LEU fuel using MCNP6 and JENDL-5 library." Nuclear Technology and Radiation Protection 40, no. 1 (2025): 1–9. https://doi.org/10.2298/ntrp2501001d.

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Kinetic parameters of a nuclear reactor are essential in reactor dynamic and safety related characteristics. Kinetic parameters of the Dalat nuclear research reactor were evaluated using the MCNP6.3 code and a new nuclear data library (JENDL-5). Numerical calculations were performed for the core configuration consisting of 92 low-enriched uranium fuel bundles for obtaining the effective delayed neutron fraction beff, the neutron generation time L and the prompt neutron lifetime lp. Two methods were used to calculate the beff: the adjoint weighted method based on perturbation theory and adjoint

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33

Stoll, W. "Material Matters." MRS Bulletin 23, no. 3 (1998): 6–16. http://dx.doi.org/10.1557/s0883769400029894.

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The following article is based on a talk for Symposium X presented by Wolfgang Stoll, Chief Scientific Advisor and Consultant in Siemens, Germany, at the 1996 MRS Fall Meeting.Since 1941 when Glenn Seaborg first isolated plutonium in milligram quantities, the total amount converted through neutron capture in U-238 has increased worldwide to about 1,200 tons and continues to grow about 70 tons/year. What was fissioned in situ in operating nuclear power stations is roughly equivalent to 5 billion tons of black coal, while the fission energy contained in those 1,200 tons unloaded in spent fuel is

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Widdicombe, Teyen, J. Stephen Herring, and Brad Kirkwood. "The Bimodal Epithermal Astronuclear Reactor (BEAR): A Long-Lived, Universal Space Nuclear Powerplant." Nuclear Science and Technology Open Research 2 (October 1, 2024): 4. http://dx.doi.org/10.12688/nuclscitechnolopenres.17481.2.

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A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecra

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Widdicombe, Teyen, J. Stephen Herring, and Brad Kirkwood. "The Bimodal Epithermal Astronuclear Reactor (BEAR): A LONG-LIVED, Universal Space Nuclear Powerplant." Nuclear Science and Technology Open Research 2 (February 1, 2024): 4. http://dx.doi.org/10.12688/nuclscitechnolopenres.17481.1.

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A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecra

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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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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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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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van der Marck, Steven C., and Nicola L. Asquith. "THE TCA BENCHMARK FOR VALIDATION OF TEMPERATURE FEEDBACK CALCULATIONS." EPJ Web of Conferences 247 (2021): 10010. http://dx.doi.org/10.1051/epjconf/202124710010.

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The TCA benchmark was investigated as a possible candidate for validation of temperature feedback calculations. This benchmark has low-enriched uranium fuel, light water moderation and reflection, and a temperature range of 20–60 °C. The use of three different nuclear data libraries was considered, viz. ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0. Since the results were not as good as hoped for, additional studies were performed to identify the cause(s) of discrepancies. The benchmark values depend on a choice of delayed neutron data, so it was investigated whether this could be the cause of discre

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Merk, Bruno, Anna Detkina, Seddon Atkinson, Dzianis Litskevich, and Gregory Cartland-Glover. "On the Dimensions Required for a Molten Salt Zero Power Reactor Operating on Chloride Salts." Applied Sciences 11, no. 15 (2021): 6673. http://dx.doi.org/10.3390/app11156673.

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Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for

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dos Santos Araújo, Matheus Henrique, Graiciany De Paula Barros, Geovana Loren da Cruz, Keferson Almeida Carvalho, Vitor Silva, and Andre Augusto Campagnole dos Santos. "Initial Safety Parameter Evaluation of a PWR Loaded with Thorium and Reprocessed Fuel<i></i>." Brazilian Journal of Radiation Sciences 12, no. 4B (Suppl.) (2025): 2688. https://doi.org/10.15392/2319-0612.2024.2688.

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The once-through cycle (OTC) of nuclear fuel results in storing large quantities of high-radioactive isotopes. Alternatively, the closed cycle (CC), which involves reprocessing and reusing spent nuclear fuel, improves fuel utilization and reduces high-level radioactive waste. This study evaluates the feasibility of incorporating reprocessed fuel into a Pressurized Water Reactor (PWR) core. The PWR core was simulated based on the component dimensions, material definitions, and fuel compositions described in the available Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS). T

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Gates, J. T., A. Denig, R. Ahmed, V. K. Mehta, and D. Kotlyar. "Low-enriched cermet-based fuel options for a nuclear thermal propulsion engine." Nuclear Engineering and Design 331 (May 2018): 313–30. http://dx.doi.org/10.1016/j.nucengdes.2018.02.036.

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Kazansky, Yury A., Nikita O. Kushnir, and Ekaterina S. Khnykina. "Multiple usage of thorium-based fuel in a VVER-1000 reactor." Nuclear Energy and Technology 9, no. 2 (2023): 93–98. http://dx.doi.org/10.3897/nucet.9.101762.

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This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are 232Th and fissile isotopes

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Kazansky, Yury A., Nikita O. Kushnir, and Ekaterina S. Khnykina. "Multiple usage of thorium-based fuel in a VVER-1000 reactor." Nuclear Energy and Technology 9, no. (2) (2023): 93–98. https://doi.org/10.3897/nucet.9.101762.

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This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are <sup>232</sup>Th and fissi

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Courtin, Fanny, Camille Laguerre, Philippe Miranda, Christine Chabert, and Guillaume Martin. "Pu multi-recycling scenarios towards a PWR fleet for a stabilization of spent fuel inventories in France." EPJ Nuclear Sciences & Technologies 7 (2021): 23. http://dx.doi.org/10.1051/epjn/2021022.

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Nuclear scenario studies are performed to explore the impact of possible evolutions of nuclear fleets. The nuclear fuel cycle simulation tool COSI, developed by CEA, is used to model these dynamic scenarios and to evaluate them with respect to uranium and plutonium management, fuel reprocessing and waste production. In recent years, scenarios have focused on transitions from the current nuclear French fleet to a deployment of SFR. However, the French Multi-annual Energy Planning has recently postponed the deployment of this technology to the second half of the 21st century. Alternative solutio

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Gusev, V. E. "On the problems of reusing reprocessed uranium by enrichment in schemes based on ordinary cascades." Journal of Physics: Conference Series 2147, no. 1 (2022): 012004. http://dx.doi.org/10.1088/1742-6596/2147/1/012004.

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Abstract The problem of spent nuclear fuel attracts considerable attention while its quantity is accumulating worldwide. The problem of long-term supply of the fresh fuel also remains important. One of the strategies to solve both problems is reusing the spent nuclear material. The uranium, in this way, could be recycled multiple times in light-water reactors. In order to recycle the uranium, it is extracted from the irradiated fuel during the reprocessing and then enriched in 235U, taking into account the limitations on reactor-born isotopes 232,236U in the final product. The only way to do t

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Reda, Sonia M., Ibrahim M. Gomaa, Ibrahim I. Bashter, and Esmat A. Amin. "Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library." Science and Technology of Nuclear Installations 2021 (April 14, 2021): 1–9. http://dx.doi.org/10.1155/2021/8838097.

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In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed using MCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. The effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. The reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. The calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff an

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Kazazyan, V. T., A. P. Malykhin, E. F. Vaitsetskaya, N. M. Dneprovskaya, I. E. Rubin, and N. A. Tetereva. "Preliminary analysis of the possibility of using REMIX fuel in VVER-1200 reactors of the Belarusian NPP." Proceedings of the National Academy of Sciences of Belarus, Physical-Technical Series 67, no. 1 (2022): 57–64. http://dx.doi.org/10.29235/1561-8358-2022-67-1-57-64.

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The transition from conventional uranium to regenerated fuel, which uses reprocessed spent fuel and enriched natural uranium, improves fuel efficiency and reduces the amount of spent nuclear fuel (SNF). Based on the analysis of published materials concerning mainly the fuel cycles of the VVER-1000 reactor, it was concluded that the most suitable in the conditions of the Republic of Belarus is the use of REMIX fuel. To confirm this conclusion in relation to the VVER-1200 reactors of the Belarusian NPP, computational studies were carried out within the framework of the State program “Scienceinte

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Jaroszewicz, Janusz, Zuzanna Marcinkowska, and Krzysztof Pytel. "Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis." Nukleonika 59, no. 2 (2014): 43–52. http://dx.doi.org/10.2478/nuka-2014-0009.

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Abstract The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed

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Dekusar, V., and O. Gurskaya. "ON THE ISSUE OF PLUTONIUM COST IN A TWO-COMPONENT NUCLEAR POWER SYSTEM." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (2021): 25–33. http://dx.doi.org/10.55176/2414-1038-2021-2-25-33.

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A possible approach to accounting for the specific present value of plutonium produced in fast reactors of a two-component nuclear power system (NPS) with thermal and fast reactors is described. The approach is based on taking into account the additional income that can be obtained by selling at the market price the natural uranium released when thermal reactors are replaced with fast reactors with MOX-fuel based on plutonium produced in NPS. At the same time, along with the sale of natural uranium, the sale at market value of other products made on its basis, for example, enriched uranium or

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Pomysukhina, Alina Ye, Yury P. Sukharev, and German N. Vlasichev. "Evaluation of the neutronic performance of a fast traveling wave reactor in the Th-U fuel cycle." Nuclear Energy and Technology 6, no. 2 (2020): 77–82. http://dx.doi.org/10.3897/nucet.6.54629.

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The possibility for all of the uranium or thorium fuel to be used nearly in full is expected in traveling wave reactors. A traveling wave reactor core with a fast neutron spectrum in a thorium-uranium cycle has been numerically simulated. The reactor core is shaped as a rectangular prism with a seed region arranged at one of its ends for the neutron fission wave formation. High-enriched uranium metal is used as the seed region fuel. Calculated power density dependences and concentrations of the nuclides involved with the transformation chain along the core at a number of time points have been

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