• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.230-237
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• 2020

In this part, an implicit time dependent solution is presented for the Boltzmann transport equation discretized by the analytic coarse mesh finite difference method (ACMFD) over the spatial domain as well as the simplified P₃ (SP₃) for the angular variable. In the first part of this work we proposed a SP₃-ACMFD approach to solve the static eigenvalue equations which provide the initial conditions for temp dependent equations. Having solved the 3D multi-group SP₃-ACMFD static equations, an implicit approach is resorted to ensure stability of time steps. An exponential behavior is assumed in transverse integrated equations to establish a relationship between flux moments and currents. Also, analytic integration is benefited for the time-dependent solution of precursor concentration equations. Finally, a multi-channel one-phase thermal hydraulic model is coupled to the proposed methodology. Transient equations are then solved at each step using the GMRES technique. To show the sufficiency of proposed transient SP₃-ACMFD approximation for a full core analysis, a comparison is made using transport peers as the reference. To further demonstrate superiority, results are compared with a 3D multi-group transient diffusion solver developed as a byproduct of this work. Outcomes confirm that the idea can be considered as an economic interim approach which is superior to the diffusion approximation, and comparable with transport in results.

• Progress in Medical Physics
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• v.27 no.4
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• pp.232-235
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• 2016

The purpose of this study is to develop the treatment planning system (TPS) based on Monte-Carlo simulation for BNCT. In this paper, we will propose a method for dose estimation by Monte-Carlo simulation using the CT image, and will evaluate the accuracy of dose estimation of this TPS. The complicated geometry like a human body allows defining using the lattice function in MCNPX. The results of simulation such as flux or energy deposition averaged over a cell, can be obtained using the features of the tally provided by MCNPX. To assess the dose distribution and therapeutic effect, dose distribution was displayed on the CT image, and dose volume histogram (DVH) was employed in our developed system. The therapeutic effect can be efficiently evaluated by these evaluation tool. Our developed TPS could be effectively performed creating the voxel model from CT image, the estimation of each dose component, and evaluation of the BNCT plan.

• The Korean Journal of Nuclear Medicine
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• v.28 no.2
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• pp.227-233
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• 1994

Irradiation of 20mg of natural $Dy(NO_3)_3$ in a neutron flux of $2{\times}10^{13}n/cm^2$ sec for 4 hours gave 5.76 Ci of $^{165}Dy$(specific activity, 610mCi/mg Dy) with high radionuclidic purity (>99.9 %). $^{165}Dy-MA$ was prepared in a quantitative yield by reacting the aqueous solution of $^{165}Dy(NO_3)_3$ with sodium borohydride solution in 0.2N NaOH. Coulter particle analyzer exhibited mean particle size of $2.6{\mu}m$ (range $1{\sim}6{\mu}m$), Even though the $^{165}Dy-MA$ suspension in saline was stored at $37^{\circ}C$ for 24 hours or autoclaved at $121^{\circ}C$ for 30minutes, there was no significant change in particle size and leakage problem indicating the prepared $^{165}Dy-MA$ is sufficiently stable. In-vivo retention studies were carried out by administering $^{165}Dy-MA$ into the knee joint space of normal rabbits. Gamma camera analysis showed high retention in joint space of normal rabbits. Gamma camera analysis showed high retention in joint space even at 24 hours after administration (> 99.9%). The ease with which the $^{165}Dy-MA$ can be made in the narrow size range and their high invitro and vivo stability make them attractive agents for radiation synovectomy.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• Proceedings of the SAREK Conference
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• 2007.11a
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• pp.305-309
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• 2007

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. The Commissioning of the FTL is to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

• Proceedings of the SAREK Conference
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• 2007.11a
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• pp.310-315
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• 2007

The FTL has been developed to be able to irradiate test fuels at the irradiation hole(IR1 hole) by considering its utility and user's irradiation requirements. FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. For safety assurance IPS is designed to have dual stainless steel pressure vessel and OPS is composed of main cooling water system, emergency cooling water system, LMP(letdown, make-up, purification) system, etc. FTL Conceptual design was set up in 2001, basic design had completed including a design requirement, basic piping & instrument diagram (P&ID), and the detail design in 2004. In 2005, the development team carried out purchase and manufacture hardware and make a contract for construction work. FTL construction work began on August, 2006 and ended on March, 2007. After FTL development which is expected to be finished by 2008, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

• Journal of Magnetics
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• v.21 no.2
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• pp.183-186
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• 2016

In this study, we report on an investigation of proton radiation resistance of 410 martensitic stainless steels under 3 MeV proton with the doses ranging from $1.0{\times}10^{15}$ to $1.0{\times}10^{17}p/cm^2$ at the temperature 623 K. Vibrating sample magnetometer (VSM) and X-ray diffractometer (XRD) were used to study the variation of magnetic properties and structural damages by virtue of proton irradiation, respectively. VSM and XRD analysis revealed that the 410 martensitic stainless steels showed proton radiation resistance up to $10^{17}p/cm^2$. Proton energy degradation and flux attenuations in 410 stainless steels as a function of penetration depth were calculated by using Stopping and Range of Ions in Matter (SRIM) code. It suggested that the 410 stainless steels have the radiation resistance up to $5.2{\times}10^{-3}$ dpa which corresponds to neutron irradiation of $3.5{\times}10^{18}n/cm^2$. These results could be used to predict the maintenance period of SUS410 stainless steels in fission power plants.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.99-110
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• 1989

The optimal control of xenon concentration in a nuclear reactor is posed as a linear quadratic regulator problem with state feedback control. Since it is not possible to measure the state variables such as xenon and iodine concentrations directly, implementation of the optimal state feedback control law requires estimation of the unmeasurable state variables. The estimation method used is based on the Luenberger observer. The set of the reactor kinetics equations is a stiff system. This singularly perturbed system arises from the interaction of slow dynamic modes (iodine and xenon concentrations) and fast dynamic modes (neutron flux, fuel and coolant temperatures). The singular perturbation technique is used to overcome this stiffness problem. The observer-based controller of the original system is effected by separate design of the observer and controller of the reduced subsystem and the fast subsystem. In particular, since in the reactor kinetics control problem analyzed in the study the fast mode dies out quickly, we need only design the observer for the reduced slow subsystem. The results of the test problems demonstrated that the state feedback control of the xenon oscillation can be accomplished efficiently and without sacrificing accuracy by using the observer combined with the singular perturbation method.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.15 no.1
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• pp.67-74
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• 2002

A phosphor for Plasma Display Panel, BaMgAl$_{10}$ O$_{17}$ :Eu$^{2+}$, showing a blue emission band at about 450nm was prepared by a solid-state reaction using BaCO$_3$, $Al_2$O$_3$, MgO, Eu$_2$O$_3$ as starting materials wish flux AlF$_3$. The study of the behaviour of Eu in BAM phosphor was carried out by the photoluminescence spectra and the Rietveld method with X-ray and neutron powder diffraction data to refine the structural parameters such as lattice constants, the valence state of Eu, the preferential site of Mg atom and the site fraction of each atom. The phenomenon of the concentration quenching was abound 2.25~2.3wt% of Eu due to a decrease in the critical distance for energy transfer of inter-atomic Eu. Through the combined Rietveld refinement, R-factor, R$_{wp}$, was 8.11%, and the occupancy of Eu and Mg was 0.0882 and 0.526 at critical concentration. The critical distance of Eu$^{2+}$ in BAM was 18.8$\AA$ at 2.25% Eu of the concentration quenching. Furthermore, c/a ratio was decreased to 3.0wt% and no more change was observed over that concentration. The maximum entropy electron density was found that the modeling of $\beta$-alumina structure in BaMgAl$_{10}$ O$_{17}$ :Eu$^{2+}$correct coincided showing Ba, Eu, O atoms of z= 1/4 mirror plane.e.ane.e.