• Korean Journal of Animal Reproduction
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• v.17 no.4
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• pp.357-363
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• 1994

To investigate the effect of extracellular matrix proteins on the in vitro development of blastomeres isolated from 2, 4, and 8-cell embryos(termed 1/2, 1/4 and 1/8 blastomeres, respectively) of ICR strain mice, those were cultured in fibronectin, gelatin, or collagen precoated culture dishes containing 1.5ml of NaHCO3-BMOC-3 medium at 37$^{\circ}C$ for 72 hrs, under the atmosphere at 5% CO2. and 95% air. Fibronectin, gelatin, or collagen significantly(P<0.01) increased and blastocyst formation rate compared with controls in 1/2(65.3, 59.2, 60.7% vs. 21.6%), 1/4(63.7, 53.4, 57.1% vs. 26.3%), and 1/8 blastomeres(61.1, 52.3, 53.7% vs. 19.1%). Both the nuclear number(P<0.05) and diameter of blastocysts(P<0.01) developed from balstomeres were significantly affected by the origin of blastomeres. The nuclear number of blastocysts developed from 1/2, 1/4, and 1/8 blastomeres ranged 29.3$\pm$1.6, 24.5$\pm$1.3, and 20..$\pm$1.2, respectively. And the diameter of those blastocysts was 88.3$\pm$2.4, 57.6$\pm$2.1, 39.8$\pm$1.9${\mu}{\textrm}{m}$, respectively.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.452-461
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• 1998

The pin power distribution is determined by analyzing the rotational gamma scanning data for 36 element fuel bundle of HANARO. A fission monitor of Nb$^{95}$ is chosen by considering the criteria of the half-life, fission yield, emitting ${\gamma}$-ray energy and probability. The ${\gamma}$-ray spectra were measured in Korea Atomic Energy Research Institute(KAERI) by using a HPGe detector and by rotating the fuel bundle at steps of 10$^{\circ}$. The counting rates of Nb$^{95}$ 766 keV ${\gamma}$-rays are determined by analyzing the full absorption peak in the spectra. A 36$\times$36 response matrix is obtained from calculating the contribution of each rod at every scanning angle by assuming 2-dimensional and parallel beam approximations for the measuring geometry. In terms of the measured counting rates and the calculated response matrix, an inverse problem is set up for the unknown distribution of activity concentrations of pins. To select a suitable solving method, the performances of three direct methods and the iterative least-square method are tested by solving simulation examples. The final solution is obtained by using the iterative least-square method that shows a good stability. The influences of detection error, step size of rotation and the collimator width are discussed on the accuracy of the numerical solution. Hence an improvement in the accuracy of the solution is proposed by reducing the collimator width of the scanning arrangement.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.521-528
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• 2014

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and $Y_2O_3$ particles of $10{\mu}m$ were selected for ODS treatment using LBS. Through the LBS method, the $Y_2O_3$ particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at $500^{\circ}C$ was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive $Y_2O_3$ particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.277-284
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• 1994

Much attention has been given to predict the near-field mass transfer of a single radioactive species from a waste solid into surrounding porous medium. But only limited considerations have been given to predict the coupled mass transfer of species with a radioactive decay chain. In this study we present an analysis assuming that the members of a decay chain dissolve congruently with a solubility-limited matrix. We give general, non-recursive analytic solutions for the transport of a radioactive decay chain in a finite porous medium when nuclides are released congruently with the matrix. As an illustration we consider the decay chain $^{234}$ Ulongrightarrow$^{230}$ Thlongrightarrow$^{226}$ Ra from spent fuel. These solutions may be useful and potentially important in performance assessment of radioactive waste repositories.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.229-246
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• 1996

A mechanistic model has been developed to predict the release behavior of fission gas during steady-state irradiation of LWR UO$_2$ fuel. Under the assumption that UO$_2$ grain surface is composed of fourteen identical circular faces and grain edge bubble can be represented by a triangulated tube around the circumference of three circular grain faces, it introduces the concept of continuous formation of open grain edges tunnels that is proportional to grain edge swelling. In addition, it takes into account the interaction between the gas release from matrix to grain boundary and the reintroduction of gas atoms into the matrix by the irradiation-induced re-solution of grain face bubbles. It also treats analytically the behavior of intragranular, intergranular, and grain edge bubbles under the assumption that both intragranular and intergranular bubbles are uniform in both radius and number density. Comparison of the present model with experimental data shows that the model's prediction produces reasonable agreement for fuel with centerline temperatures of 1000 to 140$0^{\circ}C$, wide scatter band for fuel with centerline temperatures lower than 100$0^{\circ}C$, and underprediction for fuel with centerline temperatures higher than 140$0^{\circ}C$.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.247-247
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• 1999

This paper describes a new mechanistic thermal conductivity model considering the heterogeneous microstructure of MOX fuel. Even though the thermal conductivities of MOX have been investigated numerously by experimental measurements and theoretical analyses, they show the large scattering making the performance analysis of MOX fuel difficult. Therefore, a thermal conductivity model that depends on the heterogeneous microstructure of MOX fuel has been developed by using a general two-phase thermal conductivity model. In order to apply this model for developing the thermal conductivity for heterogeneous MOX fuel, the fuel is assumed to consist of Purich particles and U02 matrix including Pu02 in solid solution. Since little relevant data on Purich particles is available, FIGARO and SiemensKWU results are only used to characterize the microstructure of unirradiated and irradiated fuel. Philliponneaus and HALDEN models are selected for the local thermal conductivities for Purich particles and matrix, respectively. Then by combining the two models, overall thermal conductivity of MOX fuel is obtained. The new proposed model estimates the MOX thermal conductivity about 10% less than the value of U02 fuel, which is in the range of MOX thermal conductivity from HALDEN. The developed thermal conductivity model has been incorporated into KAERIs fuel performance code, COSMOS, and then verified using the measured data in the FIGARO program. Comparison of predicted and measured temperatures shows the reasonable agreement within acceptable error bounds together with satisfactory results for the fission gas release and gap pressure.essure.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2204-2212
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• 2022

Recently, the demand for alpha imaging detectors for quantifying the distributions of alpha particles has increased in various fields. This study aims to reconstruct a high-resolution image from an alpha imaging detector by applying a super-spatial resolution method combined with the maximum-likelihood expectation maximization (MLEM) algorithm. To perform the super-spatial resolution method, several images are acquired while slightly moving the detector to predefined positions. Then, a forward model for imaging is established by the system matrix containing the mechanical shifts, subsampling, and measured point-spread function of the imaging system. Using the measured images and system matrix, the MLEM algorithm is implemented, which converges towards a high-resolution image. We evaluated the performance of the proposed method through the Monte Carlo simulations and phantom experiments. The results showed that the super-spatial resolution method was successfully applied to the alpha imaging detector. The spatial resolution of the resultant image was improved by approximately 12% using four images. Overall, the study's outcomes demonstrate the feasibility of the super-spatial resolution method for the alpha imaging detector. Possible applications of the proposed method include high-resolution imaging for alpha particles of in vitro sliced tissue and pre-clinical biologic assessments for targeted alpha therapy.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10a
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• pp.133-136
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• 1996

The nozzle dam task is essentially needed to maintain and repair nuclear power plants. For this task, an 8-dof redundant robot is studied with a local path-planning method[l] which is effective to find the optimal joint path in the constrained environment. In this paper, the method[l] is improved practically with the weight matrix and efficient algorithm to find working set. The effectiveness of the proposed method is demonstrated by simulation and animation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.467-472
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• 1996

This study discusses special aspects of severe accident related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor.$^{1, 2)}$ The analytical comparison of three elements core to former two elements case is conducted including evaluation of suitable nuclear cross-section sets to account for the effects of system configulation, fuel and moderator mixture temperature, material dispersion and the other thermal-hydraulics. Three elements core ANS reactor is the alternative core design which was proposed as a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium (former uranium fuel is the baseline design value of 93%) A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies still on geometry, material constituents, and thermal-hydraulic conditions are verified. Therefore, the concepts of mitigative design features are qualified.d.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.689-694
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• 2013

A duplex stainless steel sheet with 1 wt.% gadolinium was fabricated for a neutron absorbing material with high strength, excellent corrosion resistance, and low cost as well as high neutron absorption capability. The microstructure of the as-cast specimen has typical duplex phases including 31% ferrite and 69% austenite. Main alloy elements like chromium (Cr), nickel (Ni), and gadolinium (Gd) are relatively uniformly distributed in the matrix. Gadolinium rich precipitates were present in the grains and at the grain boundaries. The solution treatment at $1070^{\circ}$ for 50 minutes followed by the hot-rolling above $950^{\circ}$ after keeping the sheet at $1200^{\circ}$ for 1.5 hours are important points of the optimum condition to produce a 6 mm-thick plate without cracking.