• Korean Journal of Materials Research
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• v.3 no.3
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• pp.276-281
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• 1993

In order to study the sintering kinetics of the ternery system, U-Ce-O, its shrinkage rate with time and temperature at the Ar atmoshpere were measured by using the dilatometer. At the sintering kinetics of U-Ce-O, sintering rate curve revealed bimodal and the first extreme point at bimodal curve was affected by the $UO_2$ the second one was due to the $CeO_2$. The sintering of $(U, Ce)O_2$ was delayed as increasing the $CeO_2$. At the same lOwt. % content, the highest sintering rate was observed at the $(U, Ce)O_2$ sample ball-milled for 4 hours.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.214-220
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• 2019

Thermochemical cycles have been predominantly used for energy transformation from heat to stored chemical free energy in the form of hydrogen. The thermochemical cycle based on uranium (UTC), proposed by Oak Ridge National Laboratory, has been considered as a better alternative compared to other thermochemical cycles mainly due to its safety and high efficiency. UTC process includes three steps, in which only the first step is unique. Hydrogen production apparatus with hectogram reactants was designed in this study. The results showed that high yield hydrogen was obtained, which was determined by drainage method. The results also indicated that the chemical conversion rate of hydrogen production was in direct proportion to the mass of $Na_2CO_3$, while the solid product was $Na_2UO_4$, instead of $Na_2U_2O_7$. Nevertheless the thermochemical cycle used for hydrogen generation can be closed, and chemical compounds used in these processes can also be recycled. So the cycle with $Na_2UO_4$ as its first reaction product has an advantage over the proposed UTC process, attributed to the fast reaction rate and high hydrogen yield in the first reaction step.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.820-828
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• 2018

The growing nuclear threat has amplified the need for developing diverse and accurate nuclear forensics analysis techniques to strengthen nuclear security measures. The work presented here is part of a research effort focused on developing a methodology for reactor-type discrimination of weapons-grade plutonium. To verify the developed methodology, natural $UO_2$ fuel samples were irradiated in a thermal neutron spectrum at the University of Missouri Research Reactor (MURR) and produced approximately $20{\mu}g$ of weapons-grade plutonium test material. Radiation transport simulations of common thermal reactor types that can produce weapons-grade plutonium were performed, and the results are presented here. These simulations were needed to verify whether the plutonium produced in the natural $UO_2$ fuel samples during the experimental irradiation at MURR was a suitable representative to plutonium produced in common thermal reactor types. Also presented are comparisons of fission product and plutonium concentrations obtained from computational simulations of the experimental irradiation at MURR to the nondestructive and destructive measurements of the irradiated natural $UO_2$ fuel samples. Gamma spectroscopy measurements of radioactive fission products were mostly within 10%, mass spectroscopy measurements of the total plutonium mass were within 4%, and mass spectroscopy measurements of stable fission products were mostly within 5%.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1236-1249
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• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• Journal of the Korean Chemical Society
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• v.56 no.1
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• pp.62-67
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• 2012

A series of homo binuclear complexs of the type $[M_2(L/L^')(NO_3)n].mH_2O$, [where $M=U{O_2}^{2+},\;Th^{4+},\;ZrO^{2+}$] and $[(VO)_2(L/L^')(SO_4)_2]{\cdot}2H_2O$, L=1,5,6,9,12,15,16,20 octaaza-7,813,14-tetraphenyl-2,4,17,19-tetramethyl-1,4,6,8,12,14,16,19-docosaoctene (OTTDO) or L'=10:11;21:22-dibenzo-1,5,6,9,12,15,16,20-octaaza-7,813,14-tetraphenyl-2,4,17,19-tetramethyl-1,4,6,8,12,14,16,19-docosaoctene (DOTTOT), n=4 for $U{O_2}^{2+}$, $ZrO^{2+}$ n=8 for $Th^{4+}$ m=1,2,3 respectively, have been synthesized in template method from ethylenediamine/orthophenylene diamine, benzil monohydrazone and acetyl acetone and characterized on the basis of elemental analysis, thermal analysis, molar conductivity, magnetic moment, electronic, infrared, $^1H$-NMR studies. The results indicate that the VO(IV) ion is penta co-ordinated yielding paramagnetic complexes; $UO_2(VI)$, ZrO(IV) ions are hexa co-ordinated where as Th(IV) ion is octa co-ordinated yielding diamagnetic complexes of above composition. The fungi toxicity of the ZrO(IV) and VO(IV) complexes against some fungal pathogen has been studied.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.273-282
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• 2012

A new effective methodology for optimizing the enrichment of low-enriched zones as well as gadolinia fuel ($UO_2/Gd_2O_3$) rod designs in PLUS7 fuel assemblies was developed to minimize the maximum peak power in the core and to maximize the cycle lifetime. An automated link code was developed to integrate the genetic algorithm (GA) and the core design code package of ALPHA/PHOENIX-P/ANC and to generate and evaluate the candidates to be optimized efficiently through the integrated code package. This study introduces an optimization technique for the optimization of gadolinia fuel rod designs in order to effectively reduce the peak powers for a few hot assemblies simultaneously during the cycle. Coupled with the gadolinia optimization, the optimum enrichments were determined using the same automated code package. Applying this technique to the reference core of Ulchin Unit 4 Cycle 11, the gadolinia fuel rods in each hot assembly were optimized to different numbers and positions from their original designs, and the maximum peak power was decreased by 2.5%, while the independent optimization technique showed a decrease of 1.6% for the same fuel assembly. The lower enrichments at the fuel rods adjacent to the corner gap (CG), guide tube (GT), and instrumentation tube (IT) were optimized from the current 4.1, 4.1, 4.1 w/o to 4.65, 4.2, 4.2 w/o. The increase in the cycle lifetime achieved through this methodology was 5 effective full-power days (EFPD) on an ideal equilibrium cycle basis while keeping the peak power as low as 2.3% compared with the original design.

• Analytical Science and Technology
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• v.14 no.1
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• pp.28-33
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• 2001

The ion chromatography for the separation of plutonium species which are suggested to be $Pu^{3+}$, $Pu^{4+}$, $PuO_2{^+}$ and $PuO_2{^{2+}}$ in natural water was studied. Two separation methods were performed; 1) two-column method containing each of $SiO^-$ and SiO-$SO_3{^-}$ cation exchanger, 2) IC with AG11 column and the eluent of oxalate/nitric acid. Separation conditions for $Eu^{3+}$, $Th^{4+}$, $NpO_2{^+}$, $UO_2{^{2+}}$ in place of plutonium species were acquired from preliminary tests. When these conditions were applied to separate the plutonium species, two-column method was separated them successfully. However, the IC method with oxalate eluent was difficult in the separation of plutonium species due to the change of $Pu^{3+}$ and $PuO_2{^{2+}}$ to $Pu^{4+}$ and $PuO_2{^+}$ respectively.

• Journal of the Korean Chemical Society
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• v.54 no.4
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• pp.402-410
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• 2010

New complexes cis-[$Mo_2O_5(Hapdhba)_2$], trans-[$UO_2(Hapdhba)_2$], [Pd(Hapdhba)Cl($H_2O$)], [Pd(bpy)(Hapdhba)]Cl, [Ag(bpy)(Hapdhba)], [$Ru(Hapdhba)_2(H_2O)_2$], [$Rh(Hapdhba)_2Cl(H_2O)$] and [Au(Hapdhba)$Cl_2$] are reported, where $H_2$apdhba is the Schiff-base derived from 2-aminopyrimidine and 2,4-dihydroxy benzaldehyde. The complexes were characterized by IR, electronic, $^1H$ NMR and mass spectra, conductivity, magnetic and thermal measurements. The inhibitive effect of $H_2$apdhba for the corrosion of copper in 0.5 M HCl was also determined by potentiodynamic polarization measurements.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.