• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4052-4059
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• 2021

The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO₂)-molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 ℃ was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO₂-Mo is presented. UO₂-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 ℃. The experimental investigations suggest that UO₂-Mo composite pellets have favorable microstructural features compared to the UO₂ pellet.

• 한국분말재료학회지
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• 제16권4호
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• pp.254-261
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• 2009

In this study, we investigated the unit process parameters in spherical $UO_2$ kernel preparation. Nearly perfect spherical $UO_3$ microspheres were obtained from the 0.6M of U-concentration in the broth solution, and the microstructure of the $UO_2$ kernel appeared the good results in the calcining, reducing, and sintering processes. For good sphericity, high density, suitable microstructure, and no-crack final $UO_2$ microspheres, the temperature control range in calcination process was $300{\sim}450^{\circ}C$, and the microstructure, the pore structure, and the density of $UO_2$ kernel could be controlled in this temperature range. Also, the concentration changes of the ageing solution in aging step were not effective factor in the gelation of the liquid droplets, but the temperature change of the ageing solution was very sensitive for the final ADU gel particles.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.315-322
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• 2015

The research carried out on thorium-based fuels indicates that these fuels can be considered as economic alternatives with improved physical properties and proliferation resistance issues. In the current study, neutronic assessment of $UO_2$ in comparison with two $(Th-^{233}U)O_2$, and $(Th-^{235}U)O_2$ thorium-based fuel loads in a heavy water research reactor has been proposed. The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with $UO_2$ fuel loading. By contrast, $^{235}U$-containing thorium-based fuel and $^{235}U$-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to $UO_2$ fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.90-99
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• 1994

손상 핵연료에서 발생하는 주요한 현상중의 하나는 수중기의 분해로 갭에 존재하는 산소에 의해 $UO_2$ 가 $UO_{2+}$x/로 산화되고, 이로 인해 결정립내에서의 핵분열기체 확산계수가 증가하여 결과적으로 핵분열 기체의 방출이 증대하는 현상이다. 본 논문은 일반적인 원자로 운전 조건하에서 원자로 및 손상 핵연료의 운전조건을 고려하여 소결체의 산화거동을 모사하고 이를 바탕으로 소결체 산화가 핵분열기체의 방출 중대에 미치는 영향을 분석하였다. 소결체 산화거동은 갭에는 150기압의 포화된 순수한 수증기만이 존재한다는 가정하에 분석하였고, 산화에 의한 핵분열기체의 방출 증대 효과를 정량적으로 분석하기 위해 방출중대 인자를 도입하였다. 실험 치와 비교한 결과 방출증대 인자는 소결체 산화에 의한 핵분열기체의 방출증대 효과를 잘 예측하였다.

• Nuclear Engineering and Technology
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• 제36권4호
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• pp.338-345
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• 2004

This study examines the sensitivity of several fabrication parameters for dry process fuel, using a random sampling technique. The in-pile performance of dry process fuel with irradiation was calculated by a modified ELESTRES code, which is the CANDU fuel performance code system. The performance of the fuel rod was then analyzed using a Monte Carlo simulation to obtain the uncertainty of the major outputs, such as the fuel centerline temperature, the fission gas pressure, and the plastic strain. It was proved by statistical analysis that for both the dry process fuel and the $UO_2$ fuel, pellet density is one of the most sensitive parameters, but as for the fission gas pressure, the density of the $UO_2$ fuel exhibits insensitive behavior compared to that of the dry process fuel. The grain size of the dry process fuel is insensitive to the fission gas pressure, while the grain size of the $UO_2$ fuel is correlative to the fission gas pressure. From the calculation with a typical CANDU reactor power envelop, the centerline temperature, fission gas pressure, and plastic strain of the dry process fuel are higher than those of the $UO_2$ fuel.

• 한국세라믹학회지
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• 제47권6호
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• pp.590-597
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• 2010

HTGR using a TRISO coated particles as nuclear raw fuel material can be used to produce clean hydrogen gas and process heat for a next-generation energy source. For these purposes, a TRISO coated particle was prepared with 3 pyro-carbon (buffer, IPyC, and OPyC) layers and 1 silicone carbide (SiC) layer using a CVD technique on a spherical $UO_2$ kernel surface as a fissile material. In this study, a spherical $UO_2$ particle was prepared using a modified sol-gel method with a vibrating nozzle system, and TRISO coating fabrication was carried out using a fluidized bed reactor with coating gases, such as acetylene, propylene, and methyltrichlorosilane (MTS). As the results of this study, a spherical $UO_2$ kernel with a sphericity of 1+0.06 was obtained, and the main process parameters in the $UO_2$ kernel preparation were the well-formed nature of the spherical ADU liquid droplets and the suitable temperature control in the thermal treatment of intermediate compounds in the ADU, $UO_3$, and $UO_2$ conversions. Also, the important parameters for the TRISO coating procedure were the coating temperature and feed rate of the feeding gas in the PyC layer coating, the coating temperature, and the volume fraction of the reactant and inert gases in the SiC deposition.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1002-1008
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• 2016

An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated $UO_2$ fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated $UO_2$ fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of $UO_2$ fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.309-316
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• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• 한국세라믹학회지
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• 제46권6호
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• pp.574-582
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• 2009

$UO_2$ microsphere particles, core material of HTGR(High Temperature Gas Reactor) nuclear fuel, were prepared using by the GSP(Gel Supported Precipitation) method which is a modified-method of the wet sol-gel process. The spherical shape of initial liquid ADU droplets from the vibration nozzle system was continuously kept till the conversion to the final $UO_2$ microsphere. But the size of a final $UO_2$ microsphere was shrunken to about 25% of an initial ADU droplet size. Also, we found that the composition of dried-ADU gel particles was constituted of the very complicated phases, coexisted the U=O, C-H, N-H, N-O, and O-H functional groups by FT-IR. The important factors for obtain the no-crack $UO_2$ microsphere during the thermal treatment processes must perfectly wash out the remained-$NH_4NO_3$ within the ADU gel particle in washing process and the selections of an appropriate heating rate at a suitable gas atmosphere, during the calcining of ADU gel particles, the reducing of $UO_3$ particles, and the sintering of $UO_2$ particles, respectively.