• Nuclear Engineering and Technology
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• 제51권1호
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1736-1746
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• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.753-761
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• 2024

Introduction: In this study, TLD 600 and TLD 700 pairs were used to measure the neutron dose of Elekta Precise medical linac. To this end, the optimum moderate thickness for the conversion of fast to thermal neutrons were evaluated. Materials and methods: ²⁴¹Am-Be and ²⁵²Cf sources were simulated to calculate the optimum thicknesses of the moderator for the conversion of maximum fast neutrons (FN) into thermal neutrons (TN). Pair TLDs were used to measure F&TN doses for three different field sizes at four depths of the medical linac. Results: The maximum thickness of the moderator was optimized at 6 cm. The measurement results demonstrated that the TN dose increased with the expansion of field size and depth. The FN dose, which was converted TN, exhibits behaviors comparable to the TN due to its nature. Conclusion: This study presents the optimum thickness for the moderator to convert FN into TN and measure F&TN using TLDs.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.402-409
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• 2024

High energy neutron irradiations impact on structural and electrical properties of alumina are studied with particular emphasis on real time in-situ radiation induced conductivity measurement in low flux region. Polycrystalline Al₂O₃ samples are subjected to high energy neutrons produced from D-T neutron generator and Am-Be neutron source. 14 MeV neutrons from D-T generator are chosen to study the role of fast neutron irradiation in the structural modification of samples. Real time in-situ electrical measurement is performed to investigate the change in insulation resistance of Al₂O₃ due to radiation induced conductivity at low flux regime. During neutron irradiation, a significant transient decrease in insulation resistance is observed which recovers relative higher value just after neutron exposure is switched off. XRD results of 14 MeV neutron irradiated samples suggest annealing effect. Impact of relatively low energy neutrons on the structural properties is also studied using Am-Be neutrons. In this case, clustering is observed on the sample surface after prolonged neutron exposure. The structural characterizations of pristine and irradiated Al₂O₃ samples are performed using XRD, SEM, and EDX. The results from these characterizations are analysed and interpreted in the manuscript.

• 대한방사선기술학회지:방사선기술과학
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• 제37권3호
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• pp.223-231
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• 2014

의료용의 선형가속장치는 방사선치료에 있어서 그 활용도가 가장 높으며, 최근에는 10MV 이상의 고에너지 광자선을 이용한 치료가 보편화되고 있다. 그러나 광핵반응에 의한 광중성자가 생성됨으로써 방사선 방호측면에서 많은 문제를 야기 시키고 있다. 이에 본 연구는 MCNPX 프로그램을 이용하여 치료실 내의 위치별 중성자의 특성을 분석하였다. 그 결과, 광중성자의 생성 개수도 선원 중심점을 기준으로 거리가 멀어질수록 감소됨을 알 수 있었다. 그리고 10MV에서 20MV로 에너지가 높아짐에 따른 열중성자와 속중성자의 비율은 큰 차이가 없었지만, 선원 중심점으로부터 거리가 이격될수록 열중성자의 비율이 높아지는 현상을 확인할 수 있었다.

• 한국의학물리학회:학술대회논문집
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• 한국의학물리학회 2002년도 Proceedings
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• pp.157-160
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• 2002

The dosage of intra-operative BNCT using near-threshold $^{7}$ Li(p,n)$^{7}$ Be direct neutrons was evaluated with the calculation method validated with the phantom experiment. The production of both neutrons by near-threshold $^{7}$ Li(p,n)$^{7}$ Be and gamma rays by $^{7}$ Li(p,p'gamma)$^{7}$ Li in a Li target was calculated using Lee's method and their transport in the phantom was calculated with MCNP-4B. As a result, the region satisfying the requirements of the protocol in intra-operative BNCT for brain tumors in Japan was acknowledged to be comparable to present BNCT, for the proton energy of 1.900 MeV for example. A boron-dose enhancer (BDE) introduced in this study to increase $^{10}$ (n,$\alpha$)$^{7}$ Li dose in a living body was effective. The void used to increase doses in deep regions was also valid with the BDE. It was found that intra-operative BNCT using near-threshold $^{7}$ Li(p,n)$^{7}$ Be direct neutrons is feasible.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.

• Journal of Radiation Protection and Research
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• 제23권1호
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• pp.17-23
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• 1998

자주달개비 4430 클론에 있어서 $^{252}Cf$ 중성자에 의해 유발되는 유전자 돌연변이의 선량반응 관계를 확립함과 동시에 TSH 분홍돌연변이를 유발함에 있어서 X-선에 대비한 중성자의 생물학적 효과비를 산출하였다. 돌연변이의 분석은 공시재료의 방사선조사 후 4주 이상 이뤄졌으며 최대변이율을 보이는 조사후 6일부터 13일까지의 분석결과로부터 돌연변이빈도를 산출하였다 가용한 실험범위 내에서 중성자 선량의 증가에 따라 돌연변이빈도의 선형적 증가 양상을 나타내었다. X-선에 비교하여 산출한 중성자의 RBE값은 3.1에서 6.8의 범위를 보였으며 이러한 값은 일반적으로 인지되고 있는 중성자의 RBE값에 비해 상당히 낮은 값이다.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.133-137
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• 2001

Spatial and time distributions of x-rays and neutrons from Hokkaido University 45 MeV electron linac facility were measured and compared with the calculation. In the calculation, x-rays in a Pb-target were evaluated using the EGS-code. The x-rays and the neutrons from the target to the facility building boundary and skyshine process outside the facility building were simulated with the EGS and the MCNP respectively.