• Nuclear Engineering and Technology
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• 제2권3호
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• pp.149-161
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• 1970

국내에서 산출되는 광물골재를 사용하여 방사선 차폐용 중차폐 콘크리트를 제조하고 감마선에 대한 차폐효능을 규명하는 동시에 방사선 차폐체로서의 활용가능성을 검토하였다. 10여종의 각기 다른 광물골재를 수집하여 방사선 차폐용 골재로서의 사용타당성을 검토하기 위한 물리시험과 화학분석이 실시되었고 이 결과를 토대로 최적한 골재를 선택하여 중차폐 콘크리트가 제조되었다. 차폐용 콘크리트를 제조하는데 골재의 배합비, 물-세멘트 비율, 세골재, 조골재 비율 등을 달리해주므로써 방사선 차폐효과가 달라지는 현상을 실험적으로 구해 보았고 그 결과 차폐체의 비중이 높고 균질성이 좋은 중차폐체의 설계 조건을 유도해 낼수 있었다. 각기 다른 중차폐체에 대한 차폐효능 실험은 60Co 감마선원을 사용한 방사선 투과시험법으로 구했다. 실험을 통하여 중차폐체에 대한 방사선 차폐능과 차폐콘크리트의 비중, 차폐가격등을 분석하므로써 차폐설계상 최적의 공간배치로서 가장 경제적으로 차폐치를 설계할 수 있는 최적의 조건을 얻을 수 있었다.

• Advances in materials Research
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• 제3권2호
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• pp.61-75
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• 2014

This study aimed to investigate the suitability of some concrete components for producing "high-performance heavy density concrete" using different types of aggregates that could enhances the shielding efficiency against ${\gamma}$-rays. 15 mixes were prepared using barite, magnetite, goethite and serpentine aggregates along with 10% silica fume, 20% fly ash and 30% blast furnace slag to total OPC content for each mix. The mixes were subjected to compressive strength at 7, 28 and 90 days. In some mixes, compressive strengths were also tested up to 90 days upon replacing sand with the fine portions of magnetite, barite and goethite. The mixes containing magnetite along with 10% SF reaches the highest compressive strength exceeding over M60 requirement by 14% after 28 days. Whereas, the compressive strength of concrete containing barite was very close to M60 and exceeds upon continuing for 90 days. Also, the compressive strength of high-performance concrete incorporating magnetite fine aggregate was significantly higher than that containing sand by 23%. On the other hand, concrete made with magnetite fine aggregate had higher physico-mechanical properties than that containing barite and goethite. High-performance concrete incorporating magnetite fine aggregate enhances the shielding efficiency against ${\gamma}$-rays.

• Journal of Radiation Protection and Research
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• 제3권1호
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• pp.29-32
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• 1978

감마선을 방출(放出)하는 방사성발기물(放射性發棄物)드럼 5292개($42{\times}42{\times}3$) 저장시설(貯藏施設)의 적정(適正) 콘크리트 차폐체(遮蔽體) 두께를 산출(算出)하였다. 발기물(發棄物)이 여러가지 종류(種類)의 방사성원소(放射性元素)로 구성(構成)되어 있다고 할때 평균(平均)한 감마선 에너지와 개개(個個) 감마선 에너지에 대하여 계산(計算)한 결과(結果)를 서로 비교(比較)하였다. 그 결과(結果) 적정차폐체(適正遮蔽體)의 두께는 50cm 정도(程度)로 판명(判明)되었다. 그런데 평균(平均) 감마선 에너지에 근거(根據)하여 계산(計算)한 선량치(線量値)는 개개(個個) 감마선 에너지에 대한 값보다 동일(同一)두께의 차폐체(遮蔽體)에 대해서 훨씬 적었다.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2437-2443
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• 2024

This work synthesized four glass samples with a fixed ratio of PbO to Na₂O and a variable ratio of BaO to B₂O₃. The linear attenuation coefficient (LAC) (μ, cm^-1) and additional attenuator parameters were determined experimentally using a semiconductor detector and different gamma sources. The comparison was carried out between the experimental and the XCOM calculated results, with good agreement emerging between the two results. The impacts of the BaO substituting for the B₂O₃ on fabricated PNBB glasses' radiation shielding properties were discussed. By increasing the BaO substitution concentration between 10 and 25 mol.%, the LAC μ values (cm^-1) increased by 76.60 %, 13.81 %, 12.56 %, and 12.52 % for the respective γ-ray energies of 0.059, 0.662, 1.173, and 1.332 MeV. The μ value reduction with raised gamma energy values increased the values of the calculated half-value thickness (Δ_0.5) as a result of the μ and Δ_0.5 values' reverse proportionality. Other shielding parameters such as the lead equivalent thickness (Δ_eq) and radiation protection efficiency were also determined for the present PNBB glass samples.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.1049-1061
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• 2022

Neutron and gamma-ray shielding properties of Inconel 718 reinforced B₄C (0-25 wt%) were investigated using PSD software. Mean free path (MFP), linear and mass attenuation coefficients (LAC,MAC), tenth-value and half-value layers (TVL,HVL), effective atomic number (Z_eff), exposure buildup factors (EBF), and fast neutron removal cross-sections (FNRC) values were calculated for 0.015-15 MeV. It was found that MAC and LAC increased with the decrease in the content of B₄C compound by weight in Inconel 718. The EBFs were computed using G-P fitting method for 0.015-15 MeV up to the penetration depth of 40 mfp. HVL, TVL, and FNRC values were found to range between 0.018 cm and 3.6 cm, between 2.46 cm and 12.087 cm, and between 0.159 cm^-1 and 0.194 cm^-1, respectively. While Inconel 718 provides the maximum photon shielding property since it offered the highest values of MAC and Z_eff and the lowest value of HVL, Inconel 718 with B₄C(25 wt%) was observed to provide the best shielding material for neutron since it offered the highest FNRC value. The study is original in terms of several aspects; moreover, the results of the study may be used in nuclear technology, as well as other technologies including nano and space technologies.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3317-3323
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• 2022

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

• Journal of Radiation Protection and Research
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• 제43권2호
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• pp.75-84
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• 2018

Background: Gamma-ray spectrometry helps in radiation shielding problems and different applications of radioisotopes. Experimental arrangements including broad beam geometries are widely used. The aim is to investigate and evaluate the ${\gamma}-ray$ spectra via attenuation by environmental materials. Materials and Methods: The photo peak to nominated parts in the ${\gamma}-ray$ spectra and the attenuation coefficients ${\mu}_b/{\rho}$ from broad beam geometries are measured for the materials water, soil, sand and cement at the energies 0.662, 1.25, and 1.332 MeV with a $3{^{\prime}^{\prime}}{\times}3{^{\prime}^{\prime}}$ NaI(Tl) detector. Results and Discussion: The ${\gamma}-ray$ spectra vary according to changes in the effective atomic number $Z_{eff}$ of the attenuator, the photon energy and the solid angle. The peak to total ratios are the most sensitive parts to variations in the experimental conditions and overturn in the region 0.663 MeV to 1.332 MeV. This is indicated as inversion trend. The results are discussed in view of $Z_{eff}$ and the experimental conditions. The intensity build-up is larger at the lower energy and larger scattering angles in agreement with Klein-Nishina formula and other results. The build-up factor B is$${\sim_=}$$1 at high ${\gamma}-energies$ and small scattering angles. Conclusion: The sensitivity to material characteristics decrease gradually from peak: to total, to Compton valley, to Compton plateau ratios. Rigorous collimation is necessary at small energies. Cement, of the largest $Z_{eff}$, is characterized by the maximum broad beam mass attenuation coefficients ${\mu}_b/{\rho}$. The obtained results provide information to decide for the suitable experimental set-up based on aim of the work.

• 대한방사선기술학회지:방사선기술과학
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• 제43권5호
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• pp.375-382
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• 2020

PET/CT is a medical equipment that detects 0.511 MeV of gamma rays. The radiation workers are inevitably exposed to ionizing radiation in the process of handling the isotope. Accordingly, PET/CT workers use syringe shields made of lead and tungsten to protect their hands. However, lead and tungsten are known to generate very high scattering particles by interacting with gamma rays. Therefore, in this study, we tried to find out the effect on the scattering particles emitted from the syringe shield. In the experiment, first, the exposure dose to the hand (Rod phantom) was evaluated according to the metal material (lead, tungsten, iron, stainless steel) using Monte Carlo simulation. The exposure dose was compared according to whether or not plastic is attached. Second, the exposure dose of scattering particles was measured using a dosimeter and lead. As a result of the experiment, the shielding rate of plastics using the Monte Carlo simulation showed the largest difference in dose of about 40 % in lead, and the lowest in iron, about 15 %. As a result of the dosimeter test, when the plastic tape was wound on lead, it was found that the reduction rate was about 15 %, 28 %, and 39 % depending on the thickness. Based on the above results, it was found that 0.511 MeV of gamma ray interacts with the shielding tool to emit scattered rays and has a very large effect on radiation exposure. However, it was considered that the scattering particles could be sufficiently removed with plastics with a low atomic number. From now on, when using high-energy radiation, the shielding tool and the skin should not be in direct contact, and should be covered with a material with a low atomic number.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.269-274
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• 2022

An artificial neural network (ANN) that identifies radionuclides from low-count gamma spectra of a NaI scintillator is proposed. The ANN was trained and tested using simulated spectra. 14 target nuclides were considered corresponding to the requisite radionuclide library of a radionuclide identification device mentioned in IEC 62327-2017. The network shows an average identification accuracy of 98.63% on the validation dataset, with the gross counts in each spectrum N_c = 100~10000 and the signal to noise ratio SNR = 0.05-1. Most of the false predictions come from nuclides with low branching ratio and/or similar decay energies. If the N_c>1000 and SNR>0.3, which is defined as the minimum identifiable condition, the averaged identification accuracy is 99.87%. Even when the source and the detector are covered with lead bricks and the response function of the detector thus varies, the ANN which was trained using non-shielding spectra still shows high accuracy as long as the minimum identifiable condition is satisfied. Among all the considered nuclides, only the identification accuracy of ²³⁵U is seriously affected by the shielding. Identification of other nuclides shows high accuracy even the shielding condition is changed, which indicates that the ANN has good generalization performance.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2842-2849
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• 2024

The amorphous glasses containing PbO, ZnO, MgO, and B₂O₃ have been fabricated using the melt quenching technique. The structural properties have been analysed using the Fourier-transform infrared (FTIR) and Raman spectroscopy. Derivative of Absorption Spectra Fitting (DASF) method have been used to estimate the band gap energy from the UV-Vis absorption data which decreases from 3.02 eV to 2.66 eV with increasing the concentration of the PbO.The four glass samples 0.284 and 0.826 MeV showed unique variations in terms of gamma attenuation ability. LZMB4 glass sample proved to be the mist effective in terms of shielding of gamma radiation as it requires little distance compared to LZMB3, LZMB2 and LZMB1 to attenuate. RPE revealed a raise with increase in the thickness of the material and reduces as the energy raises. TF is superior in LZMB1 compared to LZMB2, LZMB3 and LZMB4, confirming that, LZMB4 will attenuate better. The Z_Eff of the materials was seen falling as the energy increases, confirming that the linear attenuation coefficient of the glass materials decreases when the energy is increased. The results confirmed that, glass material LZMB4 is the best option especially for gamma radiation shielding applications compared to LZMB3, followed by LZMB2, then LZMB1.