• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.895-898
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• 1995

베타선 흡수선량의 정밀측정에 사용되는 평판형전리함의 일종인 외삽형전리함(Extrapolation Chamber)으로 베타선 이온화전류의 정확한 측정을 위하여는 이온수집 단면적의 정화한 평가가 필요하다. 이를 위하여 표준 폴리축전기와 표준전압원을 이용한 충전용량 측정법으로 실제 유효 단면적을 구하였으며 측정결과는 기계적 측정치와 약 10.5%의 차이가 있음이 밝혀졌다.

• Analytical Science and Technology
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• v.21 no.2
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• pp.117-122
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• 2008

Dual scintillator for simultaneous alpha- and beta-ray counting used by detection materials of a surface contamination monitor was developed. In this study, preparation method was not a heat melting method but a solvent method, by which the counting material was manufactured by dissolving the polymer materials with solvent. It was simplified the preparation process. Plastic scintillator for beta-ray counting was prepared by solidifying the casting solution mixed with organic scintillator, polymer, and solvent. ZnS(Ag) scintillator layer was prepared by screen printing the paste solution mixed with ZnS(Ag), paste, and solvent onto the plastic layer. The good counting ability for alpha- and beta-ray using the ZnS(Ag)/plastic dual scintillator prepared and possibility for the counting material of surface contamination monitor was confirmed.

• Journal of Radiation Protection and Research
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• v.26 no.4
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• pp.375-384
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• 2001

It is possible to count and perform quench correction on two ${\beta}$ -label samples so long as the maximum ${\beta}$-energies are sufficiently different. However, when 4he conventional technique is applied to the radioassay of a mixture of more than three nuclides, the reliability of the activities determined is considerably reduced, resulting from the large overlapping of liquid scintillation pulse height distributions of each nuclide. A technique that allows the activities of multiple ${\beta}$-labeled samples to be radioassayed was proposed by using the least square method. The technique was applied to mixture samples of $^3H,\;^{14}C,\;^{36}Cl$, and $^{90}Sr$. The analytical values were in good agreement with the reference values within 7% relative error.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.159-165
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• 2005

The characterization of radiological contamination inside pipes generated during the decommission of a nuclear facility is necessary before pipes can be recycled or disposed. But, existing direct measurements of radioactive contamination level using the survey-meter can not estimate the characteristic of contamination on a local area such as the pipe inside. Moreover, the measurement of surface contamination level using the indirect methods has many problems of an application because of the difficulty of collecting sample and contamination possibility of a worker when collecting sample. In this work, plastic scintillator was simulated by using Monte Carlo simulation method for detection of beta radiation emitted from internal surfaces of small diameter pipe. Simulation results predicted the optimum thickness and geometry of plastic scintillator at which energy absorption for beta radiation was maximized. In addition, the problem of scintillator processing and transferring the detector into the pipe inside was considered when fabricating the plastic detector on the basis of simulation results. The characteristic of detector fabricated was also estimated. As a result, it was confirmed that detector capability was suitable for the measurement of contamination level. Also, the development of a detector for estimating the radiological characteristic of contamination on a local area such as the pipe inside was proven to be feasible.

• Analytical Science and Technology
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• v.20 no.6
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• pp.460-467
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• 2007

Dose rate conversion factor was calculated to estimate the absorbed effective annual doses for soils for the beta-rays and gamma-rays, which were emitted from $^{238,235}U$, $^{232}Th$, and $^{40}K$ isotopes. The most recent data of the emitted energies per decay, half-lifes, and branching ratios, which were obtained from National Nuclear Data Center, were used. When this factor and the effective annual doses for the beta-rays and the gamma-rays of natural radioisotopes were compared with those of Aitken, these of $^{238}U$, $^{232}Th$ and $^{40}K$ are estimated to have good agreements but a large difference is shown in this for $^{235}U$. Through the calculations of effective annual doses by using these factor and the measurements of gamma-ray spectra for soils, which were extracted from prehistoric remains (Mansuri) on Osong, Chungchengbuk-do, The annual effective doses were obtained to be 3.8~5.9 mGy/yr. Also, when these doses including decay elements upper Rn were compared with those on all isotopes, the differences within 9~30 % were obtained. The analysis method of the annual effective doses for the beta-rays and the gamma-rays of the natural isotopes of soils was established by this dose rate conversion factor.

• Proceedings of the Korean Vacuum Society Conference
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• 2013.02a
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• pp.549-549
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• 2013

고전압 펄스 플라즈마를 액체 속에서 발생시켜 수소 스펙트럼의 광학적 특성을 연구하였다. 고전압 펄스 발생 장치인 막스 제네레이터는 용량이 $0.5{\mu}F$인 축전기 5개로 이루어져 있다. 각각의 축전기는 전원 장치를 이용하여 저항을 통해 병렬로 충전되며, 방전 시에는 불꽃 방전 스위치에 의해 동시에 직렬로 연결되어 고전압을 발생시킨다. 따라서, 출력 전압과 전류는 40kV, 3 kA이며 총 에너지는 약 125 J이다. 직육면체 모양의 폴리카보네이트 용기 내부의 양쪽면에는 탐침 모양의 전극이 구성되어 있으며 전극 사이에서 고전압을 가진 플라즈마가 형성된다. 실험에서 액체로는 증류수를 사용하였다. 액체 방전 시 발생하는 수소 스펙트럼을 관측하기 위해 초점거리 30 cm의 monochromator를 이용하였고, 수소 알파선의 656.3 nm와 수소 베타선의 434.1 nm를 관측하였다. 전자 밀도의 측정법으로는 Stark broadening법을 이용하여 측정하였으며, 전자 온도는 Stark profile의 상대적인 전자 밀도의 비를 이용하여 계산하였다. 전자밀도는 실험조건에서 약 $3{\times}10^{15}cm^{-3}$, 전자온도는 약 2.5 eV가 측정되었다.

• Analytical Science and Technology
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• v.5 no.4
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• pp.381-387
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• 1992

Optimum counting condition of a low level liquid scintillation counter for radiocarbon dating has been determined to minimize the contribution of tritium activity and get the highest FOM by the variation of scintillation solutions, scintillation solution to benzene ratio, and ${\beta}$-ray spectrum counting ranges. Under the condition, background count rate has been reduced to $2.26{\pm}0.03cpm$, and the maximum measurable age reached to about 40,000 years. And the practicality of counting system has been verified by the analysis of shell and peat samples.

• Journal of Energy Engineering
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• v.13 no.2
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• pp.93-100
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• 2004

The smear media possible to sampling and radiation detection was prepared and evaluated for the surface contamination using indirect method. The films were made by impregnating Cerium Activated Yttrium Silicate (CAYS) in a polysulfone membrane. The membranes used solution as a dimethylformamide (DMF) and methylene chloride (MC), polysulfone as a polymer matrix and CAYS as a inorganic scintillator. The proximity membranes were prepared with single- and double-layered structure. The solidified methods were immersion to the nonsolvent bath such at water and ethanol and solvent evaporation. The measurement of the photon produced by interaction with radiation and inorganic scintillator used a photomultiflier tube (PMT), amplifier, and counter. In the comparison with the low background alpha/beta counter, the counter rate using inorganic scintillator proximity membrane for the $\^$14/C surface contamination was about 50%. Also. the $^3$H counting results revealed that the prepared membranes were efficient to monitor the surface contaminated with the low energy be-ray emitter nuclides.

• Analytical Science and Technology
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• v.18 no.6
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• pp.495-499
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• 2005

Notwithstanding antipathies against radiation, radiations are extensively used in various fields with development of the radiation technology. In order to safely manage such radiation it is necessary for development of the radiation measurement materials. In this paper a thin film type of plastic scintillator using in the contamination measurement devices was prepared. The plastic organic scintillator was prepared by simple solvent methods. It was possible to prepare in easy way and in various forms. The polysulfone which has high transparency and solubility was chosen. As the radiation measurement scintillator, the optical properties and radiation detection abilities were estimated.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.248-254
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• 2005

This paper presents a quantitative method of sequential separation of $^{90}Sr,\;^{241}Am$ and Pu nuclides with an anion exchange resin and a Sr-Spec resin. The Pu isotopes were purified with an anion exchange resin. The americium and strontium fractions were separated from the matrix elements with an oxalate co-precipitation method. Americium fraction was separated from the strontium fraction with iron co-precipitation method and purified from lanthanides with anion exchange resin. Strontium-90 was purified from other hindrance elements with the Sr-Spec resin after oxalate co-precipitation. The measurement of Pu and Am isotopes was carried out by an ${\alpha}$-spectrometer. Strontium-90 was measured by a liquid scintillation counter. The radiochemical procedure of $^{90}Sr,\;^{241}Am$ and Pu nuclides investigated in this study has been validated by application to IAEA-Reference soils.