• 한국방사선학회논문지
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• 제14권4호
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• pp.455-465
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• 2020

본 연구 목적은 선형가속기 표적(Target) 및 조준기(Collimator)에서 발생한 광중성자 특성을 평가하는 것이다. 전산모사 설계는 첫째, 표적은 단일물질 표적과 복합물질 표적으로 구성하였다. 둘째, 조준기 종류에 따라 원뿔형(Cone beam) 조준기와 부채꼴(Fan beam) 조준기로 구성하였다. 셋째, 부채꼴 조준기의 물질을 납(Pb)으로만 구성된 단일물질과 텅스텐(W)과 납으로 구성된 복합물질 조준기로 구성하였다. 연구 방법은 표적으로부터 100 cm 거리에서 가상의 구(Sphere) 표면에서 F2 Tally를 이용하여 광중성자 발생률과 에너지 스펙트럼을 계산하였다. 그 결과 광중성자 발생률은 첫째, 표적에 따라서는 20% 차이가 발생하였다. 둘째, 조준기의 종류에 따라서는 10% 차이가 발생하였다. 셋째, 조준기 물질에 따라서는 40% 차이가 발생하였다. 광중성자 스펙트럼에서도 평균 광중성자 플럭스(Flux)가 광중성자 발생량과 유사한 경향으로 나타났다. 이러한 결과로 9 MeV 선형가속기 광중성자 발생은 표적보다는 조준기에 의해 광중성자 발생이 증가하며, 조준기의 종류보다는 물질에 영향을 더 크게 받는 것을 확인할 수 있었다. 광중성자 발생이 적은 표적 및 조준기를 선택하여 운영하는 것이 가장 적극적인 방사선 방호가 될 것이다. 따라서, 본 연구는 컨테이너 보안 검색용 선형가속기 도입 및 운영 그리고 방사선 방호에 유용한 자료가 될 수 있을 것으로 생각된다.

• Radiation Oncology Journal
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• 제21권2호
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• pp.158-166
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• 2003

목적: 본 연구는 새로운 근접치료선원의 이론적 설계를 통해 출력변조를 이용한 혈관내 방사선치료를 제안한다. 대상 및 방법: 제시된 이론적 선원은 기존의 선원과는 달리 선원물질과 차폐물질(스테인리스 스틸, 또는 텅스텐) 둘 다로 구성되며 이는 방위방향으로 비대칭적 방사선량을 제공할 수 있게 한다. 따라서, 방위방향으로 선원의 방향과 체류시간을 조절함으로써 출력변조를 통한 근접치료가 가능해진다. Novoste Beta-Cath system에서 사용하는 Sr-90/Y 전자방출 선원과 유사한 모양의 두 가지 단순화한 선원을 연구의 대상으로 고려하였다. 첫 번째 선원은 선원물질과 차폐물질이 각각 반씩 차지하며, 두 번째 선원은 1/4은 선원물질로, 나머지 3/4은 차폐물질로 구성된다. 두 선원에 대해 방위 및 방사방향으로의 선량분포를 MCNP 몬테 카를로 코드를 이용하여 계산하였다. 결과: 선원이 혈관내의 중심에 위치하지 않게 되는 가상조건에서의 선량 최적화 계산을 시도한 결과, 혈관내벽에 미치는 선량의 최고치와 최저치의 차이가 87$\%$에서 7$\%$까지 줄어들 수 있음을 보였다. 결론: 본 연구에서 제시된 이론적 선원은 선량적 관점에서의 적합성 여부에 관해 매우 고무적인 결과를 보여 줌으로써 출력변조를 통한 혈관내 근접방사선치료의 가능성을 나타내었다. 본 과제의 다음 단계는 굵기가 가는 맥관 내에서 선원의 위치를 파악하여 그를 방위방향으로 정확하게 회전시킬 수 있는 방사선 전달 체계의 개발이라 할 수 있다.

• 대한방사선기술학회지:방사선기술과학
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• 제43권5호
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

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

• Radiation Oncology Journal
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• 제19권1호
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• pp.66-73
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• 2001

목적 : 붕소-중성자 포획치료법(Boron Neutron Capture Therapy, BNCT)을 위해 원자력병원 싸이클로트론에서 발생되는 최대에너지 34.4 MeV의 속중성자(Fast neutron)를 70 cm 파라핀으로 감속시킨 후 선량 특성을 조사하였다. 그 결과를 토대로 열외중성자(Epithermal neutron) 선량 측정법에 대한 프로토콜을 확립하여 원자로에서 방출되는 열외 중성자 선량 특성 평가의 기초를 삼고, 가속기를 이용한 BNCT 연구에 대한 타당성 여부를 조사하고자 한다. 대상 및 방법 : 공기 중 선량 및 물질 내 선량 분포 측정을 위해 Unidos 10005 (PTW, Germany) 전기계와 조직 등가 물질인 A-150 플라스틱으로 제작된 IC-17 (Far West, USA) 및 IC-18, ElC-1 이온함을 사용하였고, 감마선의 측정을 위해서는 마그네슘으로 제작된 IC-l7M 이온함을 이용하였으며 조직등가 기체와 아르곤 기체를 분당 5cc 씩 주입하며 측정하였다. 중성자, 광자, 전자가 혼합된 장의 모의 수송 해석을 위해 이용되는 Monte Carlo N-Particle (MCNP) transport code를 사용하여 2차원적 선량 분포 및 에너지 분포를 계산하였으며 이 결과를 측정값과 비교하였다. 결과 : BNCT에서의 유효 치료 깊이인 물 팬텀 4 cm에서의 선량은 치료기 1 MU 당 $6.47\times10^{-3}\;cGy$로 미세하였으며, 이때 감마 오염도(contamination)는 $65.2{\pm}0.9\%$로 중성자보다는 감마선에 의한 선량 기여분이 우세하였다. 깊이에 따른 선량 분포 특성에서는 중성자 선량은 선형적으로 감쇠 되었고, 감마선량은 지수적으로 보다 급격히 감쇠되는 경향을 보였으며 전체 선량의 $D_{20}/D_{10}$은 0.718 이었다. MCNP에 의한 에너지 분포 전산 계산의 결과 2.87 MeV 이하에서 중성자 피크가 나타났으며, 저에너지 영역에서는 감마선이 연속적으로 분포되는 양상을 보였다. 결론 : 벽 물질이 서로 다른 두 개의 이온함을 사용한 직접 선량 측정과 MCNP 전산 시뮬레이션을 이용한 공간 선량분포 계산으로 미세 속중성자 빔에 대한 선량 특성을 파악할 수 있었으며, 원자로 열외중성자 주(Epithermal neutron column)에 대한 선량 평가 자료로 확보하였다. 아울러 가속기에 대한 연구가 진행되어 고전압, 고전류를 발생시키는 전원 공급장치와 표적핵(Target) 물질이 개발되고 비스무스나 납 등에 의해 감마 오염도를 줄일 경우, 싸이크로트론에 의한 보론-중성자 포획치료도 가능해질 것으로 판단된다.

• Journal of Biosystems Engineering
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• 제39권3호
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• pp.205-214
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• 2014

Purpose: Phytosanitary irradiation treatment can effectively control regulated pests while maintaining produce quality. The objective of this study was to establish the best irradiation treatment for mangosteen, a popular tropical fruit, using a Monte Carlo simulation. Methods: Magnetic resonance image (MRI) data were used to generate a 3-D geometry to simulate dose distributions in a mangosteen using a radiation transport code (MCNP5). Microsoft Excel with visual basic application (VBA) was used to divide the image data into seed, flesh, and rind. Radiation energies used for the simulation were 10 MeV (high-energy) and 1.35 MeV (low-energy) for the electron beam, 5 MeV for X-rays, and 1.25 MeV for gamma rays from Co-60. Results: At 5 MeV X-rays and 1.25 MeV gamma rays, all areas (seeds, flesh, and rind) were irradiated ranging from 0.3 ~ 0.7 kGy. The average doses decreased as the number of fruit increased. For a 10 MeV electron beam, the dose distribution was biased: the dose for the rind where the electrons entered was $0.45{\pm}0.03$ kGy and the other side was $0.24 {\pm}0.10$ kGy. Use of an electron kinetic energy absorber improved the dose distribution in mangosteens. For the 1.35 MeV electron beam, the dose was shown only in the rind on the irradiated side; no significant dose was found in the flesh or seeds. One rotation of the fruit while in front of the beam improved the dose distribution around the entire rind. Conclusion: These results are invaluable for determining the ideal irradiation conditions for phytosanitary irradiation treatment of tropical fruit.

• 한국의학물리학회지:의학물리
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• 제33권4호
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• pp.129-135
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• 2022

Purpose: A full-energy-peak (FEP) efficiency correction is required through a Monte Carlo simulation for accurate radioactivity measurement, considering the geometrical characteristics of the detector and the sample. However, a relative deviation (RD) occurs between the measurement and calculation efficiencies when modeling using the data provided by the manufacturers due to the randomly generated dead layer. This study aims to optimize the structure of the detector by determining the dead layer thickness based on Monte Carlo simulation. Methods: The high-purity germanium (HPGe) detector used in this study was a coaxial p-type GC2518 model, and a certified reference material (CRM) was used to measure the FEP efficiency. Using the MC N-Particle Transport Code (MCNP) code, the FEP efficiency was calculated by increasing the thickness of the outer and inner dead layer in proportion to the thickness of the electrode. Results: As the thickness of the outer and inner dead layer increased by 0.1 mm and 0.1 ㎛, the efficiency difference decreased by 2.43% on average up to 1.0 mm and 1.0 ㎛ and increased by 1.86% thereafter. Therefore, the structure of the detector was optimized by determining 1.0 mm and 1.0 ㎛ as thickness of the dead layer. Conclusions: The effect of the dead layer on the FEP efficiency was evaluated, and an excellent agreement between the measured and calculated efficiencies was confirmed with RDs of less than 4%. It suggests that the optimized HPGe detector can be used to measure the accurate radioactivity using in dismantling and disposing medical linear accelerators.