• 대한화학회지
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• 제7권1호
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• pp.38-41
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• 1963

Nb is one of the trouble-some fission products in the reprocessing of nuclear fuels. In this paper, the extraction of Nb from 1, 2, 3, 4, 6 and 9N $HNO_3$ solution by mixtures of TBP and DBP in dodecane are reported. Sums of the concentration of TBP and DBP are kept to 20%. When the concentrations of DBP are lower the $2{\times}10^{-2}%$, distribution ratios are almost same, and ratios increase abruptly and the slope is about 2.5 at between $2{\times}10^{-2}$ to $4{\times}10^{-1}%$, then slope falls down to about 0.5. There is aging effect on mixture of TBP and DBP.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.470-475
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• 1998

The marginal nuclear criticality analysis for the high density spent fuel storage at a PWR plant was carried out by using the HELIOS and CASMO-3 codes. More than 20 % of the calculated reactivity saving effect is observed in this analysis. This mainly comes from the adoption of some important fission products and B-10 in the criticality analysis. By taking burnup and boron credits, the high capacity of the spent fuel storage rack can be more fully utilized, reducing the space of storage. Larger storage for a given inventory of spent fuel should result in remarkable cost savings and mort importantly reduce the risks to the public and occupational workers.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.63-70
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• 2009

A steam generator tube rupture (SGTR) accident, which is a partial reactor building bypass scenario, has a low probability and high consequences. SAMG has been used to manage the progression of severe accidents and the release of fission products induced by an SGTR at the Wolsong plants. Four of the six SAGs in the SAMG are used to manage the progression of a severe accident induced by an SGTR at the Wolsong plants. The results of the ISAAC code calculation have shown that the proper use the SAMG can stop a severe accident from progressing and keep the reactor building intact during a severe accident. These results confirm that the SAMG is an effective means of managing the progression of severe accidents initiated by an SGTR at the Wolsong plants.

• Corrosion Science and Technology
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• 제5권6호
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• pp.201-205
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• 2006

HYPER(Hybrid Power Extraction Reactor) is the accelerator driven subcritical transmutation system developed by KAERI(Korea Atomic Research Institute). HYPER is designed to transmute long-lived transuranic actinides and fission products such as Tc-99 and I-129. Liquid lead-bismuth eutectic (LBE). Has been a primary candidate for coolant and spallation neutron target due to its appropriate thermal-physical and chemical properties, However, it is very corrosive to the common steels used in nuclear installations at high temperature. This corrosion problem is one of the main factors considered to set the upper limits of temperature and velocity of HYPER system. In this study, a parametric study for a corrosion model was performed. And a preliminary corrosion model was also developed to predict the corrosion rate in isothermal Pb and LBE flow loops.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.647-652
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• 1998

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identified the possible even scenarios following the loss of shutdown cooling. The Thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior, From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determined the containment closure time to prevent the uncontrolled released of fission products to atmosphere, These data provide useful information to the abnormal procedure to cope with event.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2008년도 춘계학술대회 논문집
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• pp.473-476
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• 2008

In this paper, we designed the main mechanism of high throughput device for rod-cuts of spent fuel. For this, we analyzed the mechanical methods(slitting, ball mill, roller straightening) and chemical methods(muffle furnace, rotary kiln). As the results, methods of ball drop and rotary drum for concepts design were selected in the analysis step. For enhancement of oxidation rate, we devised the blades on the reactor with mesh type. Also, for enhancement of decladding rate, we designed ball size and rotation reactor with mesh type and devised the vacuum system for fission products. Mechanisms of oxidation and recovery can simultaneously handle the rod-cuts of spent fuel and independently recover. The results of mechanism design can be used for scale-up of high throughput device.

• Bulletin of the Korean Chemical Society
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• 제19권10호
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• pp.1090-1094
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• 1998

Acid-catalyzed benzidine rearrangements of new unsymmetrical diazanes 1-3, prepared from the reduction of corresponding diazenes 4-6, were carried out in ethanolic solutions. The results are as follows; rearrangement of (3-carbomethoxyphenyl)(3-methoxyphenyl)diazane 1 gave 4,4'-diamino-2-carbomethoxy-2'-methoxybiphenyl 12 (p-benzidine type) in 71% and 10-amino-3-methoxyphenanthridin-6(5H)-one 13, 8-amino-3-methoxyphenanthridin-6(5H)-one 14 in 7.1% and 3.4%, respectively. Product 13 and 14 were formed by the condensation reaction of primarily formed o-benzidine and diphenyline type product, respectively. (5-Carbomethoxy-2-chlorophenyl)(4-methoxyphenyl)diazane 2 and (5-carbomethoxy-2-methylphenyl)(4-methoxyphenyl)diazane 3 underwent mainly disproportionations to give fission amines and corresponding diazenes in about 53% and 40% yields, respectively. The results obtained from the rearrangements of diazanes 1-3 indirectly indicated the importance of disproportionations to understand the benzidine rearrangements. The structures of benzidine rearrangement products were determined by usual NMR techniques such as DEPT, 2D H-H COSY, H-C COSY, 2D NOESY, and Gaussian function multiplication.

• Nuclear Engineering and Technology
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• 제37권1호
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• pp.101-108
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• 2005

The neutron induced nuclear cross-section data for Pd-105, Ag-109, Xe-131, and Cs-133 were calculated and evaluated from an unresolved energy to 20 MeV. The energy dependent optical model potential parameters were extracted based on recent experimental data and applied up to 20 MeV. A spherical optical model and a statistical model for the equilibrium energy, and a multistep direct and a multistep compound model for the pre-equilibrium energy were used in the calculation. The direct capture model was recently introduced for fast neutron capture. The theoretically calculated cross-sections were compared with the experimental data and the evaluated files. The total and capture cross-sections calculated using the model were in good agreement with the reference experimental data. The evaluated cross-section results were compiled in ENDF-6 format and merged with the resonance component, already adopted in the ENDF/B-VI release 8. New data library files covering from thermal to 20 MeV were created. They are at the preliminary stage of an ENDF/B- VII release.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.370-381
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• 2002

The neutron induced nuclear data for Pr-141, Nd-143, Nd-145, Sm-147 and Sm-149 were calculated and evaluated from 10 keV to 20 MeV. The energy dependent optical model potential parameters were extracted based on the recent experimental data and applied up to 20 MeV. The s-wave strength function was calculated. Spherical optical model , statistical model in equilibrium energy, multistep direct and multistep compound model in pre-equilibrium energy and direct capture model were introduced in Empire calculation. The theoretically calculated cross sections were compared with the experimental data and the evaluated files. The model calculated total and capture cross sections were in good agreement with the reference experimental data. The capture cross sections in pre-equilibrium were enhanced in recent released Empire version. The evaluated cross section results were compiled to ENDF-6 format and will improve the ENDF/B-Vl.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.565-572
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• 2013

The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade $SiC_f/SiC$ composites are briefly reviewed. A CVI-processed $SiC_f/SiC$ composite with a PyC or $(PyC-SiC)_n$ interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.