• Nuclear Engineering and Technology
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• 제46권2호
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• pp.125-146
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• 2014

This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3650-3659
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• 2022

We proposed a modification of a double cascade scheme to enrich reprocessed uranium. Such a cascade scheme represents a combination of one cascade with "broadening" of the flow and an ordinary three-flow cascade. A calculation and optimization method has been developed for the proposed scheme according to various efficiency criteria. It is shown that the proposed scheme makes it possible to obtain low-enriched uranium of commercial quality using reprocessed uranium of different initial compositions. For example, the enrichment of reprocessed uranium, which had gone through five consequent recycles, was considered. The proposed scheme allowed to enrich it with simultaneous fulfillment of restrictions on isotopes ²³²U, ²³⁴U, and ²³⁶U. Such results indicate the scheme's applicability under conditions of multiple recycling of uranium in reactor fuel. Computational experiments have shown that in the proposed modification, a noticeable saving of natural uranium in the cycle (~18%) can be achieved, provided that the additional consumption of separative work does not exceed 10%, compared with the case of enrichment of natural uranium to obtain LEU of equivalent quality.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1044-1062
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• 2017

The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U-Mo fuel. Plate-type U-7 wt.% Mo/Al-5 wt.% Si, referred to as U-7Mo/Ale5Si, dispersion fuel with a uranium loading of $8.0gU/cm^3$, was selected to achieve higher fuel efficiency and performance than are possible when using $U_3Si_2/Al$ dispersion fuel. To qualify the U-Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U-7Mo/Al-5Si dispersion fuel ($8gU/cm^3$), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U-7Mo/Al-5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U-Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U-Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.