• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.11a
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• pp.420-423
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• 2002

In recent years, there has been keen interest in developing f1at panel detectors for all modalities of radiology, including gerneral radiology, fluoroscopy(angiography and cardiology), electronic portal imaging, and mammography. In this paper, we report the new hybrid x-ray detector consisted of CsI(Tl) photoemission layer and a-Se photoconductor layer to resolve conventional x-ray detector such as the direct detector using a-Se and the indirect detector using CsI(Tl)/a-Si. To design the structure of CsI(Tl)/a-Se detector, the penetrated energy spectrum and absorption fraction was estimated using MCNP 4C code. Experimental results showed that the absorption fraction of $500{\mu}m-Se$ film and $150{\mu}m-CsI\left(Tl \right)/a-Se\left( 30{\mu}m \right)$ film is 70% at 70 kVp. The absorption energy is 90% at $350{\mu}m-CsI(Tl)$.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.37 no.5C
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• pp.444-452
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• 2012

The detection performances, in this paper, are derived according to combination of the multiband GNSS signals in broadband jamming environments. The detection probabilities depending on the false alarm probabilities are derived and presented via Monte-Carlo simulation under the assumption as follows: the GNSS signals are perfectly orthogonal and simultaneously received by the receiver using non-coherent correlation.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.330-338
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• 2016

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

• Journal of Korean Society for Quality Management
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• v.48 no.4
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• pp.597-608
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• 2020

Purpose: The purpose of this paper is to present a probability distribution of the burst pressure of API 5L X65 pipes for the reliability assessment of corroded gas pipelines. Methods: Corrosion is a major cause of weakening the residual strength of the pipe. The mean residual strength on the corrosion defect can be obtained using the burst pressure code. However, in order to obtain the pipe reliability, a probability distribution of the burst pressure should be provided. This study is concerned with estimating the burst pressure distribution using Monte Carlo simulation. A response surface method is employed to represent the distribution parameter as a model of the corrosion defect size. Results: The experimental results suggest that the normal or Weibull distribution should be suitable as the probability distribution of the burst pressure. In particular, it was shown that the probability distribution parameters can be well predicted by using the depth and length of the corrosion defect. Conclusion: Given a corrosion defect on the pipe, its corresponding burst pressure distribution can be provided at instant. Subsequently, a reliability assessment of the pipe is conducted as well.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.30-35
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• 2022

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Journal of Radiation Protection and Research
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• v.27 no.3
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• pp.155-164
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• 2002

Methodology to evaluate the effective doses to adults undergoing various diagnostic x-ray examinations were established by Monte Carlo simulation of the x-ray examinations. Anthropomorphic mathematical phantoms, the MIRD5 male phantom and the ORNL female phantom, were used as the target body and x-ray spectra were produced by the x-ray spectrum generation code SPEC78. The computational procedure was validated by comparing the resulting doses to the results of NRPB studies for the same diagnostic procedures. The effective doses as well as the organ doses due to chest, abdomen, head and spine examinations were calculated for x-rays incident from AP, PA, LLAT and RLAT directions. For instance, the effective doses from the most common procedures, chest PA and abdomen AP, were 0.029 mSv and 0.44 mSv, respectively. The fact that the effective dose from PA chest x-ray is far lower than the traditional value of 0.3 mSv(or 30 mrem), which results partly from the advances of technology in diagnostic radiology and partly from the differences in the dose concept employed, emphasizes necessities of intensive assessment of the patient doses in wide ranges of medical exposures. The methodology and tools established in this study can easily be applied to dose assessments for other radiology procedures; dose from CT examinations, dose to the fetus due to examinations of pregnant women, dose from pediatric radiology.

• Journal of Korean Society of Steel Construction
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• v.20 no.1
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• pp.21-31
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• 2008

response surface method (RSM) is widely used to evaluate th e extremely smal probability of ocurence or toanalyze the reliability of very complicated structures. Althoug h Monte-Carlo Simulation (MCS) technique can evaluate any system, the procesing time of MCS dependson the reciprocal num ber of the probability of failure. The stochastic finite element method could solve thislimitation. However, it is limit ed to the specific program, in which the mean and coeficient o f random variables are programed by a perturbation or by a weigh ted integral method. Therefore, it is not aplicable when erequisite programing. In a few number of stage analyses, RSM can construct a regresion model from the response of the c omplicated structural system, thus, saving time and efort significantly. However, the acuracy of RSM depends on the dist ance of the axial points and on the linearity of the limit stat e functions. To improve the convergence in exact solution regardl es of the linearity limit of state functions, an improved adaptive response surface method is developed. The analyzed res ults have ben verified using linear and quadratic forms of response surface functions in two examples. As a result, the be st combination of the improved RSM techniques is determined and programed in a numerical code. The developed linear adapti ve weighted response surface method (LAW-RSM) shows the closest converged reliability indices, compared with quadratic form or non-adaptive or non-weighted RSMs.

• Progress in Medical Physics
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• v.28 no.3
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• pp.77-82
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• 2017

We proposed the objective ROC analysis method based on the setting of threshold value for evaluation of single photon emission computed tomography (SPECT) image. This proposed ROC analysis method uses the quantification computational threshold value to each signal on the SPECT image. The SPECT images for this study were acquired by using Monte Carlo n-particle extended simulation code (MCNPX, Ver. 2.6.0, Los Alamos National Laboratory, USA). The basic SPECT detectors and specific water phantom were realized in the simulation, and we could get the simulation results by the simulation operation. We tried to analyze the reconstructed images using threshold value application based objective ROC method. We can get the accuracy information of reconstructed region in the image. This proposed ROC technique can be helpful when we have to evaluate the weak signal for the NM image. In this study, the proposed threshold value based computational ROC analysis method can provide better objectivity than the conventional ROC analysis method.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.127-133
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• 1997

In this study an EGS4 simulation code was used to calculate real energy spectrum from measured ${\gamma}$-ray energy spectrum obtained using 2-dimensional radiation monitoring system. As a result, the $39{\times}39$ response matrix was calculated the energy range of 0.1 to 2 MeV which energy interval of 50 keV The real energy spectrum for Co-60 radioisotope was calculated using inverse of response matrix. It was confirmed that the calculated response matrix was useful to the analysis of the measured energy spectrum for the radiation monitoring system.