• 한국가시화정보학회지
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• 제18권1호
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• pp.44-49
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• 2020

Neutron radiography has been widely used in many research areas due to its different characteristics from X-rays. Neutron tomography is a powerful tool because it can clearly show the inside of an object that the eye cannot see. However, generally, commercial software is used for the reconstruction of neutron tomography. It means that maintenance costs are incurred and analysis is inefficient in some cases. In this respect, our own image processing code is required to reconstruct neutron images efficiently. In this study, an image processing code is developed for reconstruction of cross-sectional images from neutron radiography taken from the side of the object. Using the developed code, cross-sectional images of the sample are successfully reconstructed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.515-519
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• 1997

The purpose of this study is to reduce the fast neutron fluence at the reactor pressure vessel(RPV) and to provide a basis for plant-life extension. In this study, different neutron absorbers were employed in the core outer assemblies of Kori Unit 1 Cycle 14. The modified assemblies were used to calculate fast neutron fluence at the RPV and to evaluate reduction of outer assembly power and total power in core. By comparison with the case of no suppression fixture, the fast neutron fluence of a case with two rows stainless steel around the assembly with natural uranium pins is decreased by 85.8%. It is noted that the modification of outer assembly is more efficient than the previous low leakage loading pattern (LLLP) applied to Kori Unit 1. Also, compared fast neutron fluence in Cycle 1 with Cycle 14, fast neutron fluence at the RPV between Cycle 1 and Cycle 14 is not significantly different. It is found that LLLP applied to the Kori Unit 1 has not contributed to fast neutron fluence reduction at the RPV.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.1002-1007
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• 2020

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 10⁹ (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.170-175
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• 2019

Displacement cross-sections for an advanced assessment of radiation damage rates were obtained for a number of materials using the arc-dpa model at neutron incident energies from $10^{-5}eV$ to 10 GeV. Evaluated data files, CEM03 and ECIS codes, and an approximate approach were applied for the calculation of recoil energy distributions in neutron induced reactions. Three sets of displacement cross-sections based on the use of low-energy data from JEFF-3.3, ENDF/B-VIII.0, and JENDL-4.0u were prepared. Files contain also cross-sections calculated using the standard NRT model. Special efforts were made to estimate the uncertainty of obtained displacement cross-sections.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1426-1432
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• 2018

This paper presents a neutron detector configuration using EJ-301 scintillation liquid, a R9420 photo-multiplier and a homemade preamplifier. The detector qualities which include the energy linearity, efficiency response and neutron/gamma discrimination are guaranteed for neutron detection in the energy range from 0 to 3000 keVee. Regarding the neutron/gamma discrimination capability, four pulse shape discrimination (PSD) methods which are the threshold crossing time (TCT), pulse gradient analysis (PGA), charge comparison (CC) and correlation pattern recognition (CPR), were evaluated and discussed; among of these, the CPR method provides the best neutron/gamma discrimination.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.811-816
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• 2022

This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction β_eff and neutron generation time L. The modified code, OpenMC(Time-Dependent) or OpenMC(TD), was then used to calculate the effective delayed neutron fraction by using the prompt method, while the neutron generation time was estimated using the pulsed method, fitting Λ to the decay of the neutron population. OpenMC(TD) is intended to serve as an alternative for the estimation of kinetic parameters when licensed codes are not available. The results obtained are compared to experimental data and MCNP calculated values for 18 benchmark configurations.

• 방사선산업학회지
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• 제17권1호
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• pp.93-100
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• 2023

This study analyzes the neutron shielding performance of Soft Magnetic Material and proposes a military application. In general, the military protection facility has been constructed with thick concrete, so Soft Magnetic Material, consisting of boron, was used with concrete in this study. To do so, Monte-Carlo N-Particle (MCNP) was applied to simulate the Watt-fission neutron spectrum of ²³⁵U and ²³⁹Pu. As a result, a configuration of polyethylene and Soft Magnetic Material is evaluated about four times better than borated polyethylene concerning the atomic weight of boron inside each shielding material. Also, when a nuclear weapon explosion is simulated in MCNP, 1 mm of Soft Magnetic Material with 20 cm of concrete shows about 55% more additional neutron shielding performance compared to when Soft Magnetic Material is not used. In this work, the neutron shielding performance of Soft Magnetic Material could be identified and Soft Magnetic Material would be useful for neutron shielding if applicable to concrete structure.

• Journal of Radiation Protection and Research
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• 제40권2호
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• pp.79-86
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• 2015

헬륨-3는 높은 반응효율, 장시간 사용가능성, 감마선에 대한 낮은 반응확률 등과 같은 장점들을 가지고 있기 때문에 대부분의 중성자 검출기의 반응물질로 사용되어 왔다. 그러나 지난 몇 년 사이 전세계적인 헬륨-3의 부족으로 인해 기체의 수급이 어려워지고 있고 이에 따라 가격이 급격히 증가하게 되었다. 이러한 이유로 헬륨-3 대체 물질들을 이용한 고효율의 중성자 검출기의 개발에 대한 연구가 많은 연구그룹에 의해 활발히 진행되기 시작하였다. 이러한 연구에서는 다양한 물질들을 이용하고 있으며, 이 중에서 붕소-10은 다른 대체물질과 비교할 때 상대적으로 높은 중성자 반응확률, 낮은 감마반응효율, 물질의 안정성, 가격적 이점 그리고 기존 헬륨-3를 이용한 검출기의 계측회로의 재활용 가능성 등과 같은 장점들 때문에 많은 연구그룹에서 붕소-10을 이용한 중성자 검출기 개발을 진행하고 있다. 본 논문에서는 중성자 검출기에 사용될 수 있는 붕소-10 박막을 개발하고 이에 대한 성능평가를 수행하였다. 중성자 검출기의 반응물질로 붕소-10을 사용하기 위해서는 중성자와 붕소-10이 반응하여 생성되는 이차방사선을 측정할 수 있어야 한다. 본 연구에서 활용한 기체충진형 중성자 검출기의 경우 붕소-10을 얇은 박막 형태로 제작하여 중성자와 반응하여 생성된 이차방사선이 기체를 이온화 시켜서 생성되는 이온쌍을 측정하는 방법을 이용한다. 그러므로 중성자 반응효율과 이차방사선의 재흡수율을 고려한 붕소-10(탄화붕소)의 적절한 두께를 선정할 필요가 있다. 이를 위해서 본 논문에서는 몬테칼로 기법을 이용하는 MCNP6를 이용하여 다양한 두께에 따른 중성자신호수집효율의 변화를 계산하였다. 또한, 스퍼터링 기법을 이용하여 다양한 두께의 박막을 제작하고 이를 이용하여 중성자 반응신호를 측정하였다. 그리고 제작된 박막의 2차원 모니터링을 위한 다중선 비례계수기의 적용가능성을 타진하기 위해 제작된 붕소박막이 설치된 2차원 다중선 비례계수기를 제작하고 중성자 응답 특성을 평가하였다.

• 한국초전도학회:학술대회논문집
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• 한국초전도학회 2009년도 Korea Superconductivity Society Meeting 2009
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• pp.105-105
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• 2009

• 한국초전도학회:학술대회논문집
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• 한국초전도학회 2009년도 Korea Superconductivity Society Meeting 2009
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• pp.52-52
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• 2009