• Nuclear Engineering and Technology
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• 제41권4호
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• pp.521-530
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• 2009

Neutron science and technology, the utilization of neutron beams for a wide variety of scientific and engineering research ranging from materials and life science to industrial applications, has been one of the key elements of modem science and technology. Currently, the neutron science and technology in Korea is in rapid growth with the operation of the 30 MW High-flux Advanced Neutron Application Reactor (HANARO) at the Korea Atomic Energy Research Institute, which is one of the most powerful nuclear research reactors in the world. Furthermore, a state of the art HANARO cold neutron research facility, which will open a new era for the neutron science and technology in Korea, is expected to become available in 2010. In this paper, the progress of neutron science and technology in Korea is reviewed and its unprecedented new opportunities and challenges in coming years are presented.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1277-1281
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• 2020

A novel ceramic Gas Electron Multiplier (GEM) has been developed to meet the demand of high counting rate for the neutron detection which is an alternative to ³He-based detector at China Spallation Neutron Source (CSNS). An experiment was performed to measure the neutron transmittance of ceramic-GEM and FR4-GEM at the small angle neutron scattering (SANS) instrument. The result showed the ceramic-GEM has higher transmittance and less self-scattering especially for cold neutrons. One single ceramic GEM could give a gain of 10²-10⁴ in the mixture gas of Ar and CO₂ (90%:10%) and its energy resolution was about 27.7% by using ⁵⁵Fe X ray of 5.9 keV. A prototype has been developed in order to investigate the performances of the ceramic GEM-based neutron detector. Several neutron beam tests, including detection efficiency, spatial resolution, two-dimensional imaging, and wavelength spectrum, were carried out at CSNS and China Mianyang Research Reactor (CMRR). The results show that the ceramic GEM-based neutron detector is a good candidate to measure the high intensity neutrons.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.151-158
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• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1280-1286
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• 2023

MgF₂ as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF₂ were calculated, based on the phonon expansion model. The structural properties of MgF₂ were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF₂. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF₂ were reduced by 27.7% and 8.2%, respectively.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.863-868
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• 2020

Co Self-Powered Neutron Detector (SPND) is confronted with the problem of material consumption, which causes the response current can neither reflect the change of neutron flux in time nor be proportional to the neutron flux. In this paper, a deconvolution-based method is established to solve this problem. First of all, a step signal of neutron flux is taken as an example to analyze its performance. When the material consumption of Co SPND is 10%, after compensation, the response current can be in correspondence of neutron flux. Finally, the effects of this model in different Signal-to-Noise Ratio are analyzed, which fully confirms the truth of its excellent performance for compensating Co SPND's signal.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.787-791
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• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.277-284
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• 2017

This paper presents the establishment and characterization of a neutron calibration field using a bare $^{252}Cf$ source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.633-646
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• 2020

Imaging and scattering techniques using thermal neutrons allow to analyze complex specimens in scientific and industrial researches. Owing to this advantage, there have been a considerable demand for neutron facilities in the industrial sector. Among neutron sources, an accelerator driven compact neutron source is the only one that can satisfy the various requirements-construction budget, facility size, and required neutron flux-of industrial applications. In this paper, a target, moderator, and reflector (TMR) system for low-energy proton-accelerator driven compact thermal neutron source was designed via Monte Carlo simulations. For 10-30 MeV proton beams, the optimal conditions of the beryllium target were determined by considering the neutron yield and the blistering of the target. For a non-borated polyethylene moderator, the neutronic properties were verified based on its thickness. For a reflector, three candidates-light water, beryllium, and graphite-were considered as reflector materials, and the optimal conditions were identified. The results verified that the neutronic intensity varied in the order beryllium > light water > graphite, the compacter size in the order light water < beryllium < graphite and the shorter emission time in the order graphite < light water < beryllium. The performance of the designed TMR system was compared with that of existing facilities and were laid between performance of existing facilities.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2019-2028
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• 2024

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.179-180
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• 2018

Al - MMC, used as a neutron absorber, shows pitting corrosion and/or galvanic corrosion in 3.5wt.% NaCl solution. If pitting corrosion penetrates the core, neutron absorption performance could be affected. Galvanic corrosion was observed between Al 5052 and Al 1070, and pitting corrosion was observed around $B_4C$.