• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.295-300
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• 2015

The precise determination of the activity for each radionuclide in environmental samples requires the self-absorption correction factor. In this research, we derived the self-absorption correction factor for three p-type high purity germanium detectors using the Monte Carlo code MCNPX. These detectors have different characteristics such as crystal diameter, height and size of the core. We compared the calculated full-energy peak efficiency with the experimental value using a standard sample with $1g/m^3$ density and verified the modeling. We simulated the dependency of the full-energy peak efficiency on the 0.3, 0.6, 0.9, 1.0, 1.2 and $1.5g/m^3$ samples and obtained the corresponding self-absorption correction factor. The self-absorption correction factors calculated for the three detectors differ by less than 1% over most of the energy range and sample densities considered. This indicates that the self-absorption correction factors are independent of the crystal characteristics of HPGe detector.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Journal of radiological science and technology
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• v.40 no.4
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• pp.605-611
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• 2017

The structure of the actual detector was computed using the code of the PENELOPE. Using the standard mixed sources (450, 1,000 ml), compare the effectiveness of each energy according to various densities and height of the PENELOPE computer simulation, and calculate the effectiveness of the various environmental specimens and apply them to various environmental specimens to determine the lower limit. The values obtained by the obtained value were obtained by applying the obtained efficiency to the actual environmental specimens and obtaining the lower limit values. The density correction factor is 1.155 g of the density correction factor of $0.4g/cm^3$ (59.54keV), 1.153 (661 keV), $1.06g/cm^3$ 1.064 (1,836.04keV), 1.03, and 1.033. It was confirmed that the radioactivity concentration of environmental samples decreased as the amount of specimen was measured increases, and the MDA value decreased as time measured increases.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.12-17
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• 1989

The correction factor of neutron fraction absorbed by $^{55}$ Mn in the MnSO$_4$ bath was determined for the absolute measurement of neutron emission rate by using the solution circulation-type manganese sulfate bath system. For the determination of this correction factor, I/f, the atomic number desnsity and the effective neutron capture cross section data of Mn, S and impurity elements in the MnSO$_4$ solution were determined. For the atomic number density determination, the MnSO$_4$ solution concentration was determined by using the volumetric EDTA titration and gravimetric method. The impurity contents were analyzed by using the ICP method. For the calculation of effective neutron capture cross sections, a FORTRAN computer program EASCAL was developed in this study. in which Westcott's parameters and Axton's empirical relations are used.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.309-317
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• 1997

The measured radioactivity of gamma-emitting radionuclides in each radioactive waste drum using the non-destructive waste assay method is underestimated than real radioactivity in radioactive waste drum because the gamma-rays are attenuated within the medium. Therefore, the measured radioactivity should be corrected for the attenuation of gamma-rays. For the correction of the attenuation of gamma-rays, the attenuation correction method should be applied differently by considering the distribution and density of medium in radioactive wastes drum generated from nuclear power plants. In this study, the model drums were fabricated for simulating five types of radioactive waste drums generated from nuclear power plant and the optimum methods of the attenuation correction were experimentally determined to analyze the activity of radionuclides in the waste drum accurately using the segmented gamma scanning system. With the determination of the attenuation correction methods from the experimental results the transmission method and the average density method for the miscellaneous waste drum, the transmission method and the differential peak absorption method for the shielded miscellaneous waste drum were used to measure the density of medium in waste drums. Also, the average density method and the differential peak absorption method for the spent resin drum, the paraffin solidified drum, and the spent filter drum were used.

• Proceedings of the Korea Water Resources Association Conference
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• 2017.05a
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• pp.282-282
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• 2017

모래, 실트 및 자갈과 같은 비점착성 유사는 하천에서의 이동 형태에 따라 소류사와 부유사로 구분된다. 부유사는 난류로 인해 흐름 내에서 부유 상태로 이동하는 유사로, 대부분의 자연 하천에서 유사는 부유사 형태로 이송된다. 유수동역학적 조건 하에서 이동하는 부유사의 입도 분포는 유사 입자의 부유와 퇴적에 따라 불규칙적으로 변화하기 때문에 여러 연구에서 주요한 문제로 다뤄지고 있다. 부유사의 입도 분포는 흐름 유속, 부유사의 부유 높이, 하상 재료의 특성 등에 따라 변화하며, 로그 정규분포를 따르는 것으로 알려져 있다. 이에 본 연구에서는 여러 다양한 하천 흐름 조건에서 부유사의 입도 분포를 모의할 수 있는 입도 분포 모형에 관한 개념적 틀(Framework)을 제안한다. 유사 입자의 입도 분포 모의는 추계학적 방법의 적용을 통해 얻어진다. 본래 점착성 유사의 입도 분포를 모의하기 위한 추계학적 입도 분포 모형으로부터 제안된 개념적 틀로, 다양한 흐름 조건 하에서 특정 확률 분포형을 띠는 입도 분포를 모의할 수 있다. 점착성 유사의 이동 모형에서는 점착성을 띠는 유사 입자들의 응집 현상에 따른 크기 변화를 모의하기 위한 응집 모형이 필수적이다. 시간에 따른 크기 변화를 모의하는 응집 모형에서, 흐름 내 여러 특성들에 의해 결정되는 응집 인자와 달리 파괴 인자의 경우 불규칙적 난류 운동으로 인해 무작위한 특성을 띤다. 모형에서 요구되는 파괴 인자를 특정 확률 분포형을 띠는 난수로 고려함으로써 점착성 유사의 입도 분포 모형이 개발되었다. 이 때, 점착성 유사는 프랙탈 구조를 가지는 것으로 가정하기 때문에 크기에 따라 밀도와 침강 속도가 변화한다. 반면 비점착성 유사는 크기에 따른 밀도 변화가 일어나지 않으므로, 고정된 밀도와 프랙탈 차원을 적용하여 점착성 유사의 입도 분포모형으로부터 비점착성 유사의 입도 분포 모의가 가능할 것으로 판단된다. 이러한 추계학적 방법의 적용을 통해, 하나의 경계 조건으로 대변되는 하상 특성에 따른 단점 또한 보완될 것으로 예측된다. 예를 들어 로그 정규 분포를 띤다고 가정할 때 보정을 통해 결정해야하는 변수는 평균과 분산으로 두 개가 요구된다. 유사의 평균 크기로부터 확률분포형의 평균값이 결정되면, 하상에 존재하는 유사의 특성에 따른 입도 분포의 분산은 난수의 분산을 결정함으로써 모의할 수 있다.

• Fire Science and Engineering
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• v.16 no.2
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• pp.51-58
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• 2002

For Halon-1301 regulated by Montreal Protocol and $CO_2$as its alternatives, the empirical equations of density, viscosity, and enthalpy were correlated in terms of temperature. They were obtained by regression analysis from the experimental data in the literature. The empirical equation of density was expressed as compressibility factor by the second- order function of temperature. The empirical equation of viscosity was formulated as a power function, and a correction factor was considered to cover the wider range of temperature. Finally, heat capacity as well as enthalpy were well fitted by empirical form of the second-order temperature. The correlation coefficients of the empirical equations in this work were more than 0.99.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.228-233
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• 2013

It has been continuously generated the requirement for the replacement of the main components such as a steam generator due to the deterioration of the nuclear power plant all around the world. Also, a large amount of radioactive metal was generated during the decommissioning in a short period. It is required to make an accurate measurement of the residual radioactivity for recycling the metal waste for releasing from regulatory control. In planning the measurement procedures, the influence of geometry, self-absorption, density and other relevant factors on the representativeness of the measurements should be considered for the decommissioning metal waste. In this study, the method for measurement procedures, the source term evaluation, the ways to secure representative samples, the measurement device for wide area and the self-absorption correction factors for different density were evaluated. The metal samples for measurement were prepared for securing the simple geometry and representative by melting process. The developed correction method for measuring the radioactivity a variety density of metal waste could improve the reliability of the evaluation results for clearance.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2011.04a
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• pp.362-365
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• 2011

본 논문에서는 통계적인 방법을 이용하여 점탄성 제진재인 합성고무의 물성에 대한 변동성을 평가하는 방법을 제안하고 측정데이터를 이용하여 합성고무에 대한 평가를 수행하였다. 고무 물성의 불확실성 인자로는 외기 온도의 변화와 실험 데이터의 오차 및 점탄성 제진모델의 오차를 고려하였다. 고무는 분수차 미분 모델로 표현되었고 온도의 영향은 비선형 이동계수모델을 도입하여 복소계수로 나타내어 동강성과 감쇠를 표현하였다. 이러한 물성모델을 바탕으로 고무에 대한 물성 실험데이터와 물성계수의 확률밀도함수 사이에 정의된 우도함수를 최대화하는 통계적 보정방법을 이용하여 물성모델의 물질계수들에 대한 변동성을 추정하였다.

• Journal of Radiation Protection and Research
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• v.23 no.3
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• pp.139-147
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• 1998

Purpose: Tissue inhomogeneity such as lung affects tumor dose as well as transmission dose in new concept of on-line dosimetry which estimates tumor dose from transmission dose using the new algorithm. This study was carried out to confirm accuracy of correction by tissue density in tumor dose estimation utilizing transmission dose. Methods: Cork phantom (CP, density $0.202\;gm/cm^3$) having similar density with lung parenchyme and polystyrene phantom (PP, density $1.040\;gm/cm^3$) having similar density with soft tissue were used. Dose measurement was carried out under condition simulating human chest. On simulating AP-PA irradiation, PPs with 3 cm thickness were placed above and below CP, which had thickness of 5, 10, and 20 cm. On simulating lateral irradiation, 6 cm thickness of PP was placed between two 10 cm thickness CPs additional 3 cm thick PP was placed to both lateral sides. 4, 6, and 10 MV x-ray were used. Field size was in the range of $3{\times}3$ cm through $20{\times}20$ cm, and phantom-chamber distance (PCD) was 10 to 50 cm. Above result was compared with another sets of data with equivalent thickness of PP which was corrected by density. Result: When transmission dose of PP was compared with equivalent thickness of CP which was corrected with density, the average error was 0.18 (${\pm}0.27$) % for 4 MV, 0.10 (${\pm}0.43$) % for 6 MV, and 0.33 (${\pm}0.30$) % for 10 MV with CP having thickness of 5 cm. When CP was 10 cm thick, the error was 0.23 (${\pm}0.73$) %, 0.05 (${\pm}0.57$) %, and 0.04 (${\pm}0.40$) %, while for 20 cm, error was 0.55 (${\pm}0.36$) %, 0.34 (${\pm}0.27$) %, and 0.34 (${\pm}0.18$) % for corresponding energy. With lateral irradiation model, difference was 1.15 (${\pm}1.86$) %, 0.90 (${\pm}1.43$) %, and 0.86 (${\pm}1.01$) % for corresponding energy. Relatively large difference was found in case of PCD having value of 10 cm. Omitting PCD with 10 cm, the difference was reduced to 0.47 (${\pm}$1.17) %, 0.42 (${\pm}$0.96) %, and 0.55 (${\pm}$0.77) % for corresponding energy. Conclusion When tissue inhomogeneity such as lung is in tract of x-ray beam, tumor dose could be calculated from transmission dose after correction utilizing tissue density.