• The Journal of Korean Society for Radiation Therapy
• /
• v.17 no.2
• /
• pp.141-145
• /
• 2005

Purpose : It is known that the neutron is generally generated from the photon, its energy is larger than 10 MV. The neutron is leaked in the container inspection system installed at the customs though its energy is below 9 MV. It is needed that the spacial effect of the neutrons released from radiation treatment machine, linac, installed in the medical canter. Materials and Methods : The medical linear accelerator(Clinac 1800, varian, USA) was used in the experiment. Measuring neutron was used bubble detector(Bubble detector, BDPND type, BTI, Canada) which was created bubble by neutron. The bubble detector is located on the medical linear accelerator outskirt in three different distance, 30, 50, 120 cm and upper, lower four point from the iso-center. In addition, for effect on protect material we have measured eight points which are 50 cm distance from iso-center. The SAD(source-axis-distance), distance from photon source to iso-center, is adjusted to 100 cm and the field size is adjusted to $15{\times}15cm^2$. Irradiate 20 MU and calculate the dose rate in mrem/MU by measuring the number of bubble. Results : The neutron is more detected at 5 position in 30, 50 cm, 7 position in 120 cm and with wedge, and 2 position without mount. Conclusion : Though detection position is laid in the same distance in neutron measurement, the different value is shown in measuring results. Also, neutron dose is affected by the additional structure, the different value is obtained in each measurement positions. So, it is needed to measure and evaluate the neutron dose in the whole space considering the effect of the distance, angular distribution and additional structure.

• Journal of Radiation Protection and Research
• /
• v.26 no.2
• /
• pp.93-99
• /
• 2001

Voxel head phantom for overcoming the limitation of mathematical phantom in depleting anatomical details was constructed and example dose calculation for BNCT was performed. The repeated structure algorithm of the general purpose Monte Carlo code, MCNP4B was applied for yokel Monte Carlo calculation. Simple binary yokel phantom and combinatorial geometry phantom composed of two materials were constructed for validating the voxel Monte Carlo calculation system. The tomographic images of VHP man provided by NLM(National Library of Medicine) were segmented and indexed to construct yokel head phantom. Comparison of doses for broad parallel gamma and neutron beams in AP and PA directions showed decrease of brain dose due to the attenuation of neutron in eye balls in case of yokel head phantom. The spherical tumor volume with diameter, 5cm was defined in the center of brain for BNCT dose calculation in which accurate 3 dimensional dose calculation is essential. As a result of BNCT dose calculation for downward neutron beam of 10keV and 40keV, the tumor dose is about doubled when boron concentration ratio between the tumor to the normal tissue is $30{\mu}g/g$ to $3{\mu}g/g$. This study established the voxel Monte Carlo calculation system and suggested the feasibility of precise dose calculation in therapeutic radiology.

• Nuclear Engineering and Technology
• /
• v.13 no.4
• /
• pp.264-276
• /
• 1981

A simulation procedure which can represent time-dependent nuclear characteristics of TRIGA Mark-III reactor is developed. CITATION, a multi-group diffusion-depletion program, has been utilized as calculational tool. The group structure employed in this study consists of 7 groups: -3-fast and 4-thermal-which is conventionally utilized in TRIGA type reactor analysis. Three-dimensional nuclear characteristics are synthesized by combining results from two-dimensional plane calculation and two-dimensional cylinder calculation, since direct three-dimensional approach is not yet possible. An effort ia made to develope a method which can extract effective zone and group dependent bucklings by neutron diffusion theory rather than conventional zone and/or group independent Ducklings by neutron transport theory, since neutron leakage is quite high for small core such as research reactors. It is turned out that the method developed in this study gives satisfactory results. The calculation is performed under assumptions that all control rods are fully withdrawn, that no samples are inserted in the irradiation holes and that the core is located in the center of the reactor pool. Burnup-dependent variation of core excess reactivity, time dependent change of Xe-135 poisoning and reactivity worth of rotary specimen rack are calculated and compared with operation records. Neutron flux and power distribution as well as neutron spectrum in each irradiation .facility are presented.

• Korean Journal of Optics and Photonics
• /
• v.19 no.6
• /
• pp.435-438
• /
• 2008

D2O, which is used in nuclear power generation, is slightly different from $H_2O$. $D_2O$ consists of deuterium (D), which is an isotope of hydrogen (H) and has one more neutron than H. $D_2O$ is heavier by about 11% than $H_2O$, and $D_2O$ is present in water in natureat about 0.002%. Its melting point and boiling point are $3.81^{\circ}C$ and $101.42^{\circ}C$, respectively. $D_2O$ is harmful to the human body if it replaces water in the human body by more than $25%{\sim}50%$. We have measured the index of refractive and power absorption of 0%, 25%, 50%, 75%, and 100% of $D_2O$ in $H_2O$ using terahertz time-domain spectroscopy, and we have found that the refractive index decreases and power absorption also decreases as the concentration of $D_2O$ increases.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2008.11a
• /
• pp.346-347
• /
• 2008

• Journal of Korea Water Resources Association
• /
• v.49 no.2
• /
• pp.133-144
• /
• 2016

In this study, Frequency Domain Reflectometry (FDR) and COSMIC-ray soil moisture (SM) stations were installed at Sungkyunkwan University in Suwon, South Korea. To provide reliable information about SM, soil property test, time series analysis of measured soil moisture, and comparison of measured SM with satellite-based SM product are conducted. In 2014, six FDR stations were set up for obtaining SM. Each of the stations had four FDR sensors with soil depth from 5 cm to 40 cm at 5~10 cm different intervals. The result showed that study region had heterogeneous soil layer properties such as sand and loamy sand. The measured SM data showed strong coupling with precipitation. Furthermore, they had a high correlation coefficient and a low root mean square deviation (RMSD) as compared to the satellite-based SM products. After verifying the accuracy of the data in 2014, four FDR stations and one COSMIC-ray station were additionally installed to establish the Soil Moisture site with FDR and COSMIC-ray, called SM-FC. COSMIC-ray-based SM had a high correlation coefficient of 0.95 compared with mean SM of FDR stations. From these results, the SM-FC will give a valuable insight for researchers into investigate satellite- and model-based SM validation study in South Korea.

• Computational Structural Engineering
• /
• v.3 no.3
• /
• pp.141-146
• /
• 1990

This study is on a seismic analysis of absorber rod in KMRR Reactivity Control Mechanism. The model being studied is two coaxial tubes(control absorber rod and flow tube) immersed in the water and partially coupled(overlap) by water gap. The hydrodynamic mass effects by the water in each surrounding conditions are considered in the model. The natural frequencies, stresses and displacements of the system due to Safe Shutdown Earthquake are computed in the cases of in-phase modes and out-of-phase modes of two coaxial tubes. The results show that maximum stresses are well below the allowable limit but the maximum displacements at the ends of both tubes are so much that the absorber rod contacts with the flow tube(or surrounding wall).

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.432-432
• /
• 2005

사용후핵연료시료 중에 함유된 요오드(I-127 및 129)를 정량하기 위하여 화학적 방법으로 분리 회수하고 중성자방사화분석법을 이용하였다. 사전실험으로 모의사용후핵연료를 이용하여 회수율을 측정하였다. 모의 및 실제사용후핵연료시료를 $90^{\circ}C$에서 8 M $HNO_3$ 용액으로 용해하고 용해 후 용해용액 중의 잔류 요오드, 응축 및 휘발된 요오드 각각을 정량하였다. 응축 요오드는 핵연료 용해 후 재증류하여 회수하였다. 잔류 및 응축 요오드는 시료의 산화상태를 조절한 후 용매추출로 요오드를 회수한 다음 이온교환 또는 침전법으로 방사화학적으로 분리한 후 중성자방사화분석(RNAA)으로 정량하였다. 제작한 이온교환분리관 및 여과키트에 요오드를 흡착 또는 침전시켜 분리한 다음 중성자조사를 위한 삽입체(Insert)로 이용하였다. 휘발 요오드는 제조한 흡착체(Ag-silica gel)를 담은 흡착관에 포집하고 홉착체를 구간별 균질시료로 만든 다음 비파괴중성자 방사화분석(INAA)으로 정량하였다. 침전 및 흡착 요오드의 화학적 특성을 EPMA(electron probe microanalysis) 분석으로 조사하였다. 요오드 정량결과를 다른 방법으로 비교분석하기 위하여 음이온교환수지상에서 요오드를 정제 및 회수하기 위한 용리거동을 조사하였다.

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.252-262
• /
• 1992

The intent of this study is to develop an efficient calculation method which can be used to analyze the heterogeneous time-dependent reactor problems. By using the nodal theory one can not only reduce the calculational efforts, but accurately determine the group dependent flux densities averaged over the entire homogeneous nodes. This method uses correction factors(called“discontinuity factors”) in a rigorous manner to obtain the relationship between the node-averaged flux and the surface-averaged fluxes and currents. The discontinuity factors are calculated from the node-averaged fluxes, diffusion coefficients, and the discontinuity factors of the previous time step. The test results for two benchmark problems demonstrate the accuracy and efficiency of the method developed for the transient application in which assembly-size nodes can be used.

• Journal of Radiation Protection and Research
• /
• v.21 no.1
• /
• pp.27-39
• /
• 1996

The change in spent fuel characteristics by DUPIC fuel cycle(burnup of spent PWR fuel again in CANDU) is examined with time elapse since discharge. Major characteristics examined include isotopic concentration, radioactivity, decay heat radiotoxicity and radiation source-term of spent fuel material, which is existing in a type of spent PWR and DUPIC fuel. Behaviors of major nuclides contributing to such changes are also analyzed in terms of radionuclide concentration. From the analysis, the change in radionuclide concentration by DUPIC shows approximately 2% decrease in actinides concentration and 20% increase in fission products concentration. Radioactivity and decay heat of spent DUPIC fuel does not depend upon radionuclides concentrations, which is a unique in sence of general characteristics of spent fuel. In terms of gamma spectrum, spent DUPIC fuel shows lower values than that of spent PWR fuel by 40 to 50% in the range of $0.01{\sim}0.575$ MeV but much higher over 3.5MeV. Neutron Intensities of both spent fuels are mainly determined by $({\alpha},\;n)$ reaction and spontaneous fission reaction of actinides. Of them, especially, the spontaneous fission reaction Is a major neutron source-term, which causes that neutron intensities of spent DUPIC fuel $having{\sim}3.3$ times higher Cm-244 concentration are ${\sim}4$ times higher than that of spent PWR fuel.