Relevant bibliographies by topics / Low-enriched cermet-based uranium fuel

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Academic literature on the topic 'Low-enriched cermet-based uranium fuel'

Author: Grafiati
Published: 5 June 2025
Last updated: 1 August 2025

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

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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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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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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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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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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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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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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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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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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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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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Books on the topic "Low-enriched cermet-based uranium fuel"

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Skupio, Rafał. Zastosowanie nieinwazyjnych pomiarów rdzeni wiertniczych do zwiększenia informacji na temat parametrów skał zbiornikowych. Instytut Nafty i Gazu - Państwowy Instytut Badawczy, 2022. http://dx.doi.org/10.18668/pn2022.237.

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The research carried out in the monograph aimed to create a measurement and interpretation system which is to obtain reliable results of well logging with the accuracy of laboratory measurements. Continuous core measurements allow for the generation of logging results without the impact of the borehole and facilitate the depth matching of the core to well log data. Four main chapters can be distinguished in this work: research methodology with a description of the devices used; partial results of core measurements made on various types of rocks; a proposal for a research system, and comprehens
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Book chapters on the topic "Low-enriched cermet-based uranium fuel"

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Ohara, Masato, Akifumi Ogawa, Takanori Kitada, and Satoshi Takeda. "Heterogeneous Reactivity Effect Analysis of Pu Spots Considering Grain Size Distribution Based on MOC." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_61.

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AbstractIn the process of Mixed Oxide (MOX) fuel fabrication, plutonium grains, called plutonium spots (Pu spots), occur in MOX fuel because uranium and plutonium cannot be mixed completely. The previous study showed that the prediction accuracy of criticality is improved by considering the heterogeneity of Pu spots in MOX fuel in the analysis of some critical experiments. The analysis of the heterogeneity has been performed by Monte Carlo calculations, to ensure that core geometry and self-shielding effects are accurately considered in nuclear calculations. However, the results of Monte Carlo
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Heimberger, Lisa A., and Tunc Aldemir. "The Sensitivity of Research Reactor Based Bnct Irradiation Facility Performance to Core Conversion from High to Low Enriched Uranium Fuels." In Advances in Neutron Capture Therapy. Springer US, 1993. http://dx.doi.org/10.1007/978-1-4615-2978-1_8.

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Pyeon, Cheol Ho, Go Chiba, Tomohiro Endo, and Kenichi Watanabe. "Approach-to-Criticality Experiment." In Reactor Laboratory Experiments at Kyoto University Critical Assembly. Springer Nature Singapore, 2024. http://dx.doi.org/10.1007/978-981-97-8070-9_3.

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AbstractIn nuclear reactors, sustainable fission chain reactions by neutrons are realized without any neutron supply from external sources. In other words, nuclear reactors are systems to generate energy and to utilize radiation rays such as neutrons through properly controlling sustainable fission chain reactions. Based on this consideration, the first step to study nuclear reactors would be to know or understand which conditions in geometry, composition, size, etc. are required to make reactors in a critical state, how reactors can reach a critical state, and how reactors behave during appro
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Dalnoki-Veress, Ferenc, and Miles A. Pomper. "International Efforts to Minimize the Use of HEU." In The Oxford Handbook of Nuclear Security. Oxford University Press, 2023. http://dx.doi.org/10.1093/oxfordhb/9780192847935.013.38.

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Abstract The chapter highlights global efforts to reduce highly enriched uranium (HEU) use in the civilian sector, focusing on the success achieved and the challenges that remain. In recent decades, the United States and Russia have successfully shut down, or converted to low enriched uranium (LEU) fuel, nearly 100 research reactors that formerly used HEU. The chapter explores the progress in converting high-performance reactors in Europe and the US, as well as other HEU-fuelled reactors in Russia, South Africa, Kazakhstan, and Japan. The chapter also examines the global success of LEU-based m
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Mangaiyarkkarasi, J., and J. Shanthalakshmi Revathy. "Innovations in Nuclear Fuel Technology." In Pathways to a Carbon-Free Future Through Advanced Nuclear Systems. IGI Global, 2025. https://doi.org/10.4018/979-8-3373-0685-8.ch005.

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This chapter highlights the development of nuclear fuel and the current status with special reference to accident tolerable fuels and efficiency and sustainability. Low-gamma materials like chromium-coated zirconium and FeCrAl alloys enhance the safety by operation at harsh conditions and long-lived harmless materials; sophisticated fuels like HALEU and thorium-based fuels increase energy extraction efficiency, decrease radioactive waste, and help in long reactor cycles. Technologies in recycling also advance sustainability since areas of uranium usage are reduced and unsafe burial also elimin
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McElroy, Michael B. "Power from the Sun Abundant But Expensive." In Energy and Climate. Oxford University Press, 2016. http://dx.doi.org/10.1093/oso/9780190490331.003.0015.

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As discussed in the preceding chapter, wind resources available from nonforested, nonurban, land-based environments in the United States are more than sufficient to meet present and projected future US demand for electricity. Wind resources are comparably abundant elsewhere. As indicated in Table 10.2, a combination of onshore and offshore wind could accommodate prospective demand for electricity for all of the countries classified as top- 10 emitters of CO2. Solar energy reaching the Earth’s surface averages about 200 W m– 2 (Fig. 4.1). If this power source could be converted to electricity w
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Conference papers on the topic "Low-enriched cermet-based uranium fuel"

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Vinson, D. W., J. I. Mickalonis, and R. L. Sindelar. "Initial Evaluation of the Effect of Neutron Absorbers on the Corrosion of Uranium-Aluminum Alloys." In CORROSION 2001. NACE International, 2001. https://doi.org/10.5006/c2001-01146.

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Abstract Aluminum-based spent nuclear fuel from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Monitored Geologic Repository. The melt-dilute treatment technology has been developed to consolidate fuel assemblies by a melting/casting process in which depleted uranium is added to reduce enrichment below 20% 235U. The melt-dilute product is essentially a binary uranium-aluminum alloy to which neutron absorber materials may be readily added. Demonstration of the compatibility and effectiveness of neutron-absorbing additio
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Downar, Thomas J., Sean M. McDeavitt, S. T. Revankar, A. A. Solomon, and T. K. Kim. "Thoria-Based Cermet Nuclear Fuel: Neutronics Fuel Design and Fuel Cycle Analysis." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22305.

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Cermet nuclear fuels have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle provide intriguing options for using thoria based cermet nuclear fuel in advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or ad
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Turner, Kyler K., Gary L. Solbrekken, and Charlie W. Allen. "Thermal-Mechanical Response of Non-Uniformly Heated Nominally Flat and Curved LEU Foil Based Target." In ASME 2012 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/imece2012-88939.

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Technetium-99m is a radiopharmaceutical currently used in 85% of all diagnostic imaging procedures. The relative long lived parent isotope of technetium-99m is molybdenum-99, which is commonly produced by irradiating highly enriched uranium. In accordance with the Department of Energy: National Nuclear Security Administration’s Global Threat Reduction Initiative an effort is underway to develop low enriched uranium based molybdenum-99 production concepts. Achieving comparable molybdenum-99 yields in a low enriched uranium target effectively mandates the use of a high density metal low enriched
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Roh, Hee Seok, Walid Mohamed, and Hakan Ozaltun. "Effects of Fission Profiles on the Performance of a U-10Mo Fuel Plate." In ASME 2019 Power Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/power2019-1895.

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Abstract In order to convert the high-performance research reactors from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, U-Mo alloy-based fuels in monolithic form have been proposed. These plate-type fuels consist of a high density and low enriched uranium (LEU) foil coated with a diffusion barrier and encapsulated with the aluminum cladding. The performance of the fuel plate has been evaluated by many studies through experimental tests and numerical analyses. When evaluating the performance of a fuel, it is expensive and time-consuming to consider a variation of several parame
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Boubcher, Mustapha, Zhiliang Meng, Catherine M. Cottrell, and Sermet Kuran. "Advanced Fuel CANDU Reactor (AFCR) Core Physics With High-Burnup Recycled Uranium and Uranium-Thorium Fuel." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16662.

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The CANDU reactor has flexibility for using alternative fuels, such as Recycled Uranium (RU), low-enriched uranium and thorium (LEU/Th) and plutonium-thorium (Pu/Th) based fuels. This capability results from a versatile pressure tube design, simple fuel bundle, on-power refuelling, and high neutron economy. The Advanced Fuel CANDU Reactor (AFCR), which retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design, is a design jointly undertaken by Candu Energy and its Chinese partners led by CNNC. The objective is to use RU that is planned to be
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Cui, Deyang, Xiangzhou Cai, Jingen Chen, and Chenggang Yu. "Analysis of Sustainable Thorium Fuel Utilization in Molten Salt Reactors Starting From Enriched Uranium." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67177.

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Molten salt reactor (MSR), as one of the six systems selected by the Generation IV International Forum (GIF) for future advantaged reactors research and development (R&D), has excellent performances such as high inherent safety, desirable breeding capacity, low radioactive waste production, flexible fuel cycle and non-proliferation. Meanwhile, thorium, as an appealing alternative nuclear fuel to uranium, is more abundant than uranium in the earth’s crust. Realization of thorium fuel cycle in MSRs will greatly contribute to sustainable energy supply for global development. The objective of
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Turner, Kyler K., Gary L. Solbrekken, and Charlie W. Allen. "Thermal-Mechanical Analysis of Varying Boundary Conditions on a LEU Foil Based Molybdenum-99 Plate Processing Target." In ASME 2010 International Mechanical Engineering Congress and Exposition. ASMEDC, 2010. http://dx.doi.org/10.1115/imece2010-39393.

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Technetium-99m is a diagnostic radio-pharmaceutical that is currently used in 85% of the United States diagnostic imaging procedures [1]. All supplies of technetium-99m’s parent isotope molybdenum-99 currently originate from the irradiation of high enriched uranium (HEU) in nuclear reactor facilities located outside the United States. In accordance with the Global Threat Reduction Initiative all uranium used in future molybdenum-99 production will use low enriched uranium (LEU). Conversion to LEU material effectively mandates using LEU in the form of a metal foil as opposed to current powder b
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Britton, Kyle Anthony, and Zeyun Wu. "A Neutronics Feasibility Study of the TRIGA LEU Fuel in the 20MWt NIST Research Reactor." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82433.

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The National Bureau of Standards reactor (NBSR) at the National Institute of Standards and Technology (NIST) is under conversion from high enriched uranium (HEU) to the low enriched uranium (LEU) schema under the Reduced Enrichment for Research and Test Reactors program (RERTR) as a part of the Global Threat Reduction Initiative (GTRI). The conversion of the high performance research reactors (HPRR) such as NBSR is a challenging task due to the high flux need (2.5 × 1014 n/cm2-s for the NBSR), as well as other neutronics performance characteristics requirements without significant changes to t
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Ozaltun, Hakan, and Pavel Medvedev. "Effects of the Foil Flatness on the Stress-Strain Characteristics of U10Mo Alloy Based Monolithic Mini-Plates." In ASME 2014 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/imece2014-36605.

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The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabri
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Ozaltun, Hakan, Herman Shen, and Pavel Medvedev. "Numerical Simulation for Mechanical Behavior of U10Mo Monolithic Miniplates for Research and Test Reactors." In ASME 2011 International Mechanical Engineering Congress and Exposition. ASMEDC, 2011. http://dx.doi.org/10.1115/imece2011-62114.

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This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stress
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