• 대한방사선기술학회지:방사선기술과학
• /
• 제45권5호
• /
• pp.433-438
• /
• 2022

Targeted radionuclides treatment (TRT) requires the establishment of treatment plans that consider various factors, such as the type of radionuclides, target organs, and administration methods. For this reason, in this study, the absorption dose of a single cell was analyzed according to the type of radioisotope used to treat target radionuclides. In this study, a simulation was performed on beta rays used in the treatment of target radionuclides at the cell level using MCNPX (ver. 2.5.0). First, according to the calculation formula, the beam path according to the type of radioisotope for treatment was calculated. Second, the amount of self-radiation by beta rays emitted from cell diameters of 5 ㎛ and 10 ㎛ cell nuclei was evaluated. As a result, it showed a high range proportional to the maximum energy of the beta-ray, and the highest self-dose distribution from 177 Lu radiation sources among therapeutic radioisotopes. This was analyzed as a result that is inversely proportional to the maximum energy of the beta-ray, and it suggests that the selection of a nuclide considering the range of the beta-ray is necessary in the treatment of target radionuclides in the future.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3888-3896
• /
• 2022

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm k_eff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of ⁵²Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.

• Nuclear Engineering and Technology
• /
• 제54권12호
• /
• pp.4722-4730
• /
• 2022

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation - Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the depletion matrix are excluded in the corrector, and hence the corrector calculation involves only the reaction components, which can be efficiently solved with the Talyor Expansion Method (TEM). The numerical test shows that the new scheme substantially reduces computing time without loss of accuracy in comparison with the conventional scheme using CRAM (Chebyshev Rational Approximation Method), especially when the substep calculations are applied. The depletion calculation and source term estimation capability of BESNA are verified and validated through several problems, where results from BESNA are compared with those calculated by other codes as well as measured data. The analysis results show the computational efficiency of the new modified scheme and the reliability of BESNA in both isotopic predictions and source term estimations.

• Nuclear Engineering and Technology
• /
• 제55권10호
• /
• pp.3525-3534
• /
• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• 대한방사선기술학회지:방사선기술과학
• /
• 제44권5호
• /
• pp.489-494
• /
• 2021

In this study, radioactivity quantitative analysis was performed on radon contained in cigarette, and the effective dose was calculated using the result value to determine the amount of exposure caused by smoking. A total of 5 types of cigarettes were sampled. Cigarette smoke was collected by using activated carbon, and tobacco were measured by homogenizing for quantitative analysis. For each sample, Bi-214 and Pb-214 were subjected to gamma nuclide analysis to observe the uranium-based radioactive material contained in cigarette, and a measurement time of 30,000 seconds was set for the sample based on the results of previous studies. As a result of measuring the radioactivity of tobacco, a maximum of 0.715 Bq/kg was derived, and in the case of cigarette smoke measured using activated carbon, a maximum of 3.652 Bq/kg was derived. Using this measurement, the average effective dose to the lungs is 0.938 mSv/y, and it was found that there is a possibility of receiving exposure up to 1.099 mSv/y depending on the type of tobacco. It was found that the exposure dose due to cigarette occupies a large proportion of the annual effective dose limit for the general public. Therefore, more diverse studies on radioactive substances in cigarette are needed, and measures to monitor and reduce the incidental exposure to radon should be established.

• 한국환경농학회지
• /
• 제26권2호
• /
• pp.186-191
• /
• 2007

1998년부터 2006년까지 국내 방사능 교차분석에 참여하여 제주지역의 환경방사능 감시를 위한 방사능 분석 기술의 능력 검증과 신뢰도를 확보하기 위하여 수행되었다. 전베타 방사능 분석 시료는 공기부유진 필터와 물이었고, 감마 분석은 토양과 물 시료 중 자연 및 인공 방사성 핵종들이었다. 전베타 방사능 분석 값은 1998년과 1999년 물 시료를 제외하고는 모두 신뢰도 범위내의 값을 가졌고, 감마 핵종은 토양 시료 중의 $^{40}K$와 $137^{CS}$ 그리고 물 시료 중 몇 개의 핵종을 제외하고는 대부분 매우 우수한 평가를 받았다. 따라서 방사선 이상 사고시 원자력 안전을 위한 제주지역의 환경방사능 감시를 위한 신뢰도를 확보하여 자체적으로 환경방사능을 분석할 수 있는 능력을 함양하였다.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.507-513
• /
• 2022

When X-ray energy above 8 MV is used, photoneutrons are generated by the photonuclear reaction, which activates the components of linear accelerator (linac). Safely managing the radioactive material, when disposing linac or replacing components, is difficult, as the standards for the radioactive material management are not clear in Korea. We surveyed the management status of radioactive components occurred from medical linacs in Korea. And we also measured the activation of each part of the discarded Elekta linac using a survey meter and portable High Purity Germanium (HPGe) detector. We found that most medical institutions did not perform radiation measurements when disposing of radioactive components. The radioactive material was either stored within the institution or collected by the manufacturer. The surface dose rate measurements showed that the parts with high surface dose rates were target, primary collimator, and multileaf collimator (MLC). ⁶⁰Co nuclide was detected in most parts, whereas for the target, ⁶⁰Co and ¹⁸⁴Re nuclides were detected. Results suggest that most institutions in Korea did not have the regulations for disposing radioactive waste from linac or the management procedures and standards were unclear. Further studies are underway to evaluate short-lived radionuclides and to lay the foundation for radioactive waste management from medical linacs.

• Nuclear Engineering and Technology
• /
• 제16권2호
• /
• pp.64-69
• /
• 1984

방사성폐기물 처분과 관련하여 산청, 온양 그리고 무안에서 채취한 국산점토에 대한 Cs-137 및 Sr-90의 흡착특성과 이들 핵종의 점토층이동에 대하여 고찰하였다. 흡착분배계수(Ksorp)를 회분식 흡착실험으로 결정한 결과 Cs-137의 경우 8,000-17,000ml/g 그리고 Sr-90의 경우 10,000-15,000m1/gr 범위의 값이었다. 이때 액상의 초기농도는 0.l$\mu$Ci/ml이었다. 산청과 온양의 점토는 흡착성능이 우수하였으나 무안의 점토는 현저하게 낮았다. 이것은 무안점토에 다량 존재하는 석영성분때문인 것으로 생각되었다. 이상의 흡착특성을 Freundlich형의 형태로 다음과 같이 표시할 수 있었다. $C_{R}$=18.0 $C_{A}$$^{0.74}$ : Cs-137, $C_{R}$=0.84 $C_{A}$$^{0.45}$ : Sr-90. 이 관계식을 BOX모델에 적용하여 점토층내에서의 핵종이동을 모사한 결과 국산점토가 처분장의 충진제로서 효과적임을 확인하였다.하였다.하였다.

• 한국산학기술학회논문지
• /
• 제4권3호
• /
• pp.223-229
• /
• 2003

ICRP-60 신권고에 맞게 수정된 방사선피폭선량 계산프로그램인 K-DOSE 60을 4개 발전소 (고리, 월성, 울진, 영광)에 적용하여 주변주민 최대개인피폭선량을 평가하였다 핵종, 장기, 경로별로 결정변수값들을 도출한 결과, 경로는 성인의 경우 농작물, 유아의 경우 우유가, 핵종은 ³H, /sup 133/Xe, /sup 60/Co (고리 1,2발전소), /sup 14/C, /sup 41/Ar(월성 1,2발전소)로 나타났으며, 모든 장기에 대한 피폭선량차이가 매우 작았다. "출력 대 입력" 변동에 근거한 민감도 분석결과, 핵종의 화학적 형태가 타 입력변수들보다 10² factor만큼 높은 민감도를 보였으며, 전이/농축계수에 의한 민감도는 섭취량이나 방출량의 그것에 비하여 상대적으로 매우 낮았다. PCC를 이용한 상관성 민감도의 경우는 4가지 입력변수 모두 0.97 이상의 높은 상관성을 보였다.

• 한국의료질향상학회지
• /
• 제20권1호
• /
• pp.60-73
• /
• 2014

Objectives: This study aims at decreasing spatial dose rate through work improvement whilst spatial dose rate is the cause of increasing personal exposure dose which occurs in the process of handling radioisotope. Methods: From February 2013 until July 2013, divided into "before" and "after" the improvement, spatial dose rate in laboratory of nuclear medicine was measured in gamma image room, PET/CT-1 image room, and PET/CT-2 image room as its locations. The measurement time was 08:00, 12:00 and 17:00, and SPSS 21.0 USA was opted for its statistical analysis. Result: The spatial dose rate at distribution worktable, injection table, the entrance to the distribution room, and radioisotope storage box, which had showed high spatial dose rate, decreased by more than 43.7% a monthly average. The distribution worktable, that had showed the highest spatial dose rate in PET/CT-1 image room, dropped the rate to 42.3% as of July. The injection table and distribution worktable in the PET/CT-2 image room also showed the decline of spatial dose rate to 89% and 64.4%, respectively. Conclusion: By improving distribution process and introducing proper radiation shielding material, we were able to drop the spatial dose rate substantially at distribution worktable, injection table, and nuclide storage box. However, taking into account of steadily increasing amount of radioisotope used, strengthening radiation related regulations, and safe utilization of radioisotope, the process of system improvement needs to be maintained through continuous monitoring.