• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Particle and aerosol research
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• v.13 no.1
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• pp.1-10
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• 2017

Accurate prediction of size, shape and velocity of a bubble rising through a liquid pool is very important for predicting the particulate removal efficiency in pool scrubbing, for designing engineering safety features to prepare for severe accidents in nuclear power plants, and for predicting the emission of fission products from MCCI (molten core-concrete interaction) process during severe accidents. In this review article, previous studies on the determination of the size, shape and rising velocity of a bubble in liquid are reviewed. Various theoretical and parameterization formulas calculating the bubble size, shape and velocity from physical properties of liquid and gas flowrate are compared. Recent studies tend to suggest simple parameterizations that can easily determine the bubble shape and rising velocity without iteration, whereas iteration has to be performed to determine the bubble shape and velocity in old theories. The recent parameterizations show good agreement with measured data obtained from experiments conducted using different liquid materials with very diverse physical properties, proving themselves to be very useful tools for researchers in related fields.

• Journal of Powder Materials
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• v.27 no.2
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• pp.119-125
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• 2020

One of the promising candidates for accident-tolerant fuel (ATF), a ceramic microcell fuel, which can be distinguished by an unusual cell-like microstructure (UO₂ grain cell surrounded by a doped oxide cell wall), is being developed. This study deals with the microstructural observation of the constituent phases and the wetting behaviors of the cell wall materials in three kinds of ceramic microcell UO₂ pellets: Si-Ti-O (STO), Si-Cr-O (SCO), and Al-Si-Ti-O (ASTO). The chemical and physical states of the cell wall materials are estimated by HSC Chemistry and confirmed by experiment to be mixtures of Si-O and Ti-O for the STO; Si-O and Cr-O for SCO; and Si-O, Ti-O, and Al-Si-O for the ASTO. From their morphology at triple junctions, UO₂ grains appear to be wet by the Si-O or Al-Si-O rather than other oxides, providing a benefit on the capture-ability of the ceramic microcell cell wall. The wetting behavior can be explained by the relationships between the interface energy and the contact angle.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.223-236
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• 1977

An integral computer code has been developed for a mechanical and thermal design and performance analysis of an oxide fuel rod in a pressurized water reactor. The code designated as FROD 1.0 takes into account the phenomena of radial power depression within the pellet, cracking, densification and swelling of the pellet, fission gas release, clad creep, pellet-clad contact, heat transfer to coolant and buildup of corrosion layers on the clad surface. The FROD 1.0 code yields two-dimensional temperature distributions, dimensional changes, stresses, and internal pressure of a fuel rod as a function of irradiation time within a reasonable computation time. The code may also be used for the analyses of oxide fuel rods in other thermal reactors. As an application of FROD 1.0 the behavior of fuel rod loaded in the first core of Go-ri Nuclear Power Plant Unit 1 is predicted for the two power histories corresponding to steady state operation and Codition II of the ANS Classification. The results are compared with the design criteria described in the Final Safety Analysis Report and a discrepancy between these two values is discussed herein.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2628-2641
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• 2023

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.69-76
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• 2006

TRISO(tri-isotropic)-coated fuel particle technology is utilized owing to its higher stability at a high temperature and Its efficient retention capability for fission products In the HTGR(high temperature gas-reeled reactor). The typical spherical TRISO fuel panicle with a diameter of about 1mm is composed of a nuclear fuel kernel and outer coating layers. The outer coating layers consist of a buffer PyC(pyrolytic carbon) layer, Inner PyC(1-PyC) layer, SiC layer, and outer PyC(O-PyC) layer Most of the Inspection Items for the TRTSO-coated fuel particle depend on destructive methods. The coating thickness of the TRISO fuel particle can be nondestructively measured by the X-ray radiography without generating radioactive wastel. In this study, the coaling thickness for the simulated TRISO-coated fuel particle with $ZrO_2$ kernel Instead of $%UO_2$ kernel was measured by using micro-focus X-ray radiography with micro-focus X-ray generator and flat panel detector The radiographic image was also enhanced by image processing technique to acquire clear boundary lines between coating layers. The coaling thickness wat effectively measured by applying the micro-focus X-ray radiography The inspection process for the TRISO-coated fuel particles will be improved by the developed micro-focus X-ray radiography and digital image processing technology.

• Nuclear Engineering and Technology
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• v.20 no.2
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• pp.88-104
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• 1988

The FUel Rod Analysis(FURA) code is developed using two-dimensional finite element methods for axisymmetric and plane stress analysis of fuel rod. It predicts the thermal and mechanical behavior of fuel rod during normal and load follow operations. To evaluate the exact temperature distribution and the inner gas pressure, the radial deformation of pellet and clad, the fission gas release are considered over the full-length of fuel rod. The thermal element equation is derived using Galerkin's techniques. The displacement element equation is derived using the principle of virtual works. The mechanical analysis can accommodate various components of strain: elastic, plastic, creep and thermal strain as well as strain due to swelling, relocation and densification. The 4-node quadratic isoparametric elements are adopted, and the geometric model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The pellet cracking and crack healing, pellet-cladding interaction are modelled. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behavior accurately and stably. The pellet and cladding model has been compared with both analytical solutions and experimental results. The observed and predicted results are in good agreement. The general behavior of fuel rod is calculated by axisymmetric system and the cladding behavior against radial crack is used by plane stress system. The sensitivity of strain aging of PWR fuel cladding tube due to load following is evaluated in terms of linear power, load cycle frequency and amplitude.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.79-84
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• 2007

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.301-308
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• 2018

Anthropogenic radioactive noble gases formed by nuclear fission are significant indicators used to monitor the nuclear activity of neighboring countries. In particular, radioactive xenon, owing to its abundant generation and short half-life, can be used to detect nuclear testing, and radioactive krypton has been used as a tracer to monitor the reprocessing of nuclear fuels. Released radioactive noble gases are in the atmosphere at infinitesimal amounts due to their dilution in the air and their short half-life decay. Therefore, to obtain reliable and significant data when performing measurement of noble gases in the atmosphere, the minimum detectable activity (MDA) for noble gases should be defined as low as possible. In this study, the MDA values for radioactive xenon and krypton were theoretically obtained based on the BfS-IAR system by collecting both noble gases simultaneously. In addition, various MDA methods, confidence level and analysis conditions were suggested to reduce and optimize MDA with an assessment of the factors affecting MDA. The current investigation indicated that maximizing the pretreatment efficiency and performance maintenance of the counter were the most important aspects for Xe. In the case of Kr, since sample activities are much higher than those of Xe, it is possible to change the target MDA or to simplification of the analysis system.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.444-455
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• 1991

An accurate and fast running NEDAR model for calculating radial power profile throughout fuel life in both solid and annular pellets for existing and advanced CANDU-PHWR-fuel was developed in this work. This model contains resultant flux depression equations and neutron depression data tables which have been developed for CANDU-PHWR fuel of pellet with the diameter 8.0 to 19.5 mm and enrichment 0.71-6.0 wt % U-235, over a bumup range of 0 to 840 MWh /kgU (35000 MWD/T). In order to obtain the neutron flux distribution in the fuel pellet, the CE-HAMMER physics code was run for a neutron flux spectrum appropriate to a CANDU-PHWR to give predictions of radial power profile for several ranges of fuel design parameters. The results, which were calculated by the CE-HAMMER physics code, were fitted to an equation which is solved in terms of Bessel and exponential functions in order to obtain the parameters, $textsc{k}$, $\beta$ and λ in the resultant equation. The present NEDAR model produce a radial profile which, when normalized to unity at the pellet surface, is slightly higher than the profile of the original ELESIM data table. The predictions of the fission gas release by KAFEPA-NEDAR are in slightly better agreement with the experiments than those of ELESIM. The NEDAR model described in this study has been shown to provide an effective, reliable, and accurate method for determining radial power profiles in CANDU-PHWR fuel rods without incurring a significant increase in computing time.