• Nuclear Engineering and Technology
• /
• 제40권1호
• /
• pp.93-98
• /
• 2008

Several adhesive tapes have been studied in terms of their suitability for securing gold wires into positions for neutron flux measurements in the reactor core and irradiation facilities surrounding the core of the Open Pool Australian Light water (OPAL) reactor. Gamma ray spectrometry has been performed on each irradiated tape in order to identify and quantify activated components. Numerous metallic impurities have been identified in all tapes. Calculations relating to both the effective neutron shielding properties of the tapes and the error in measurement of the $^{198}Au$ activity caused by superfluous activity due to residual tape have been made. The most important identified effects were the prolonged cooling times required before safe enough levels of radioactivity to allow handling were reached, and extra activity caused by residual tape when measured with an ionisation chamber. Knowledge of the most suitable tape can allow a minimal contribution due to these effects, and the use of gamma spectrometry in preference to ionisation chamber measurements of the flux wires is shown to make all systematic errors due to the tape completely negligible.

• Nuclear Engineering and Technology
• /
• 제55권11호
• /
• pp.4246-4251
• /
• 2023

The optimized design of a Neutron Activation Analysis (NAA) system, including Delayed Gamma NAA (DGNAA) and Prompt Gamma NAA (PGNAA), has been proposed in this research based on Mevex Linac with 5 MeV electron energy and 50 kW power as a neutron source. Based on the MCNPX 2.6 simulation, the optimized configuration contains; tungsten as an electron-photon converter, BeO as a photoneutron target, BeD₂ and plexiglass as moderators, and graphite as a reflector and collimator, as well as lead as a gamma shield. The obtained thermal neutron flux at the beam port is equal to 2.06 × 10⁹ (# /cm².s). In addition, using the optimized neutron beam, the detection limit has been calculated for some elements such as H-1, B-10, Na-23, Al-27, and Ti-48. The HPGe Coaxial detector has been used to measure gamma rays emitted by nuclides in the sample. By the results, the proposed system can be an appropriate solution to measure the concentration and toxicity of elements in different samples such as food, soil, and plant samples.

• 대한약학회:학술대회논문집
• /
• 대한약학회 2003년도 Proceedings of the Convention of the Pharmaceutical Society of Korea Vol.2-2
• /
• pp.109.1-109.1
• /
• 2003

The aim of this study was to investigate the biological effects of blood and testis by neutron or gamma-ray irradiation in black mouse. Six-week-old C57BL male mice were irradiated with neutron (flux: 1.036739E+09) or Co60 gamma rays(dose rate: lGy/min.) The irradiation method of animal was abdominal irradiation and dose of irradiation was 10 and 20 Gy added with 5 and 15Gy in neutron irradiation.. After that, the mice were sacrificed 3 days later. Blood and testis were taken and then composition of blood in blood cell were investigated. (omitted)

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3824-3832
• /
• 2022

Prompt capture gammas are an important part of the fission reactor gamma field. Because some of the structural materials after neutron capture can emit photons with high energies forming the dominant component of the gamma spectrum in the high energy region, the following study of the high energy capture gamma was carried out. High energy gamma radiation may play a major role in areas of the radiation sciences as reactor dosimetry. The HPGe measurements and calculations of the high-energy aluminum capture gamma were performed at two moderator levels in the VR-1 pool-type reactor. The result comparison for nominal levels was within two sigma uncertainties for the major 7.724 MeV peak. A larger discrepancy of 60% was found for the 7.693 MeV peak. The spectra were also measured using a stilbene detector, and a good agreement between HPGe and stilbene was observed. This confirms the validity of stilbene measurements of gamma flux. Additionally, agreement of the wide peak measurement in 7-9.2 MeV by stilbene detector shows the possibility of using the organic scintillators as an independent power monitor. This fact is valid in these reactor types because power is proportional to the thermal neutron flux, which is also proportional to the production of capture gammas forming the wide peak.

• 한국의학물리학회지:의학물리
• /
• 제25권3호
• /
• pp.151-156
• /
• 2014

본 연구의 목적은 붕소 중성자 포획 치료 시 집적된 붕소 영역에서 중성자 선속의 변화와 그에 따른 방출된 즉발 감마선의 검출 시뮬레이션을 통하여 치료 영역에 대한 영상화의 가능성을 확인하고자 함이다. 전산 모사를 통하여 (1) 붕소 유무에 따른 중성자의 영향, (2) 내부와 외부에서의 즉발 감마선량 검출, (3) 즉발 감마선에 대한 에너지 스펙트럼 검출을 수행하였다. 모든 전산 모사는 Monte Carlo n-particle extended (MCNPX, Ver.2.6.0, Los Alamos National Laboratory, Los Alamos, NM, USA)를 이용하여 가상의 물 팬텀과 열중성자(thermal neutron) 소스, 붕소 영역을 지정하였다. 열중성자의 에너지는 1 eV 이하의 에너지였으며 선속은 2,000,000 n/sec.로 설정하였다. 이 때, 발생된 즉발 감마선의 검출은 물 팬텀과 수직 방향으로 위치시키고 납으로 둘러싸인 lutetium-yttrium oxyorthosilicate (Lu0,6Y1,4Si0,5:Ce; LYSO) 섬광체 검출기를 이용하였다. 붕소가 존재하는 영역인 5 cm 깊이에서의 28 분할로서 대략 0.18 cm의 bin을 도출하여 붕소 영역의 얕은 깊이에서부터 급격하게 저하되는 것을 확인하였다. 또한 붕소 영역이 시작되는 지점인 9 cm 깊이에서 감마선의 피크 레벨을 확인하였다. 그리고 478 keV 지점에서 정확한 즉발 감마선 피크가 관찰되는 것을 확인하였다. 478 keV의 즉발 감마선 피크는 41 keV의 반치폭으로 에너지 분해능 값은 8.5%로 측정되었다. 결론적으로 붕소 중성자 포획 치료 시 발생되는 즉발 감마선의 계측으로 치료가 행해지는 부위를 감마 카메라 또는 단일 광자 방출 단층 촬영 기기에서 영상화할 수 있는 가능성을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제50권1호
• /
• pp.182-189
• /
• 2018

A neutron generation system was developed to induce fissile fission in a lead slowing down spectrometer (LSDS) system. The source neutron is one of the key factors for LSDS system work. The LSDS was developed to quantify the isotopic contents of fissile materials in spent nuclear fuel and recycled fuel. The source neutron is produced at a multilayered target by the (e,${\gamma}$)(${\gamma}$,n) reaction and slowed down at the lead medium. Activation analysis of the target materials is necessary to estimate the lifetime, durability, and safety of the target system. The CINDER90 code was used for the activation analysis, and it can involve three-dimensional geometry, position dependent neutron flux, and multigroup cross-section libraries. Several sensitivity calculations for a metal target with different geometries, materials, and coolants were done to achieve a high neutron generation rate and a low activation characteristic. Based on the results of the activation analysis, tantalum was chosen as a target material due to its better activation characteristics, and helium gas was suggested as a coolant. In addition, activation in a lead medium was performed. After a distance of 55 cm from the lead surface to the neutron incidence, the neutron intensity dramatically decreased; this result indicates very low activation.

• 지구물리와물리탐사
• /
• 제18권3호
• /
• pp.154-161
• /
• 2015

본 연구에서는 중성자-감마 스펙트럼검층 존데 설계를 목적으로 Monte Carlo 시물레이션을 이용하여 열중성자 반응의 우세한 영역 파악 및 포획감마 스펙트럼의 에너지피크 값에 기초한 지층 구성 원소 구분을 수행하였다. 14 MeV 에너지준위의 중성자를 방출하는 중성자발생장치를 선원으로 이용하여 선원으로부터 10 cm 간격으로 12개의 중성자 검출기들을 배열함으로써 거리에 따른 열중성자 양을 측정하였다. 시추공 영향 저감을 위해 존데모형에 차폐재를 적용하여 보다 정확한 열중성자 측정을 수행하여 열중성자 반응이 우세한 위치를 분석한 뒤, 이 위치에서 검출된 포획감마 에너지 스펙트럼을 분석하여 지층을 구성하는 주요 원소 및 그 양을 확인하였다. 본 연구 결과는 중성자-감마 스펙트럼검층 존데의 신호대잡음 비 향상과 포획감마 검출기 최적 위치 선정에 도움이 될 것으로 판단된다.

• 분석과학
• /
• 제5권4호
• /
• pp.367-372
• /
• 1992

방사선 중성자 선원의 에너지 스펙트럼을 측정하기 위하여 액체 섬광 검출기(NE213)와 펄스모형 분리장치를 감마선 선원과 중성자 선원을 이용하여 그 감응 특성을 분석하였다. Am-Be 선원을 이용하여 이 장치에 대한 "Figure of Merit"을 측정한 결과 1.13 이었다. 이 값은 단색 에너지 중성자 선원인 $^{12}C(d,\;n)^{13}N$에서의 1.3 과 상당히 유사한 값을 보여 준다. NE213-PSD 장치의 성능 시험을 위한 이 실험결과는 중성자-감마 혼합 방사선장에서 스펙트럼의 측정과 중성자 에너지 스펙트럼과 속밀도 측정표준을 확립하는 데 기술적으로 유용하게 쓰일 것이다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1999년도 춘계학술발표회요약집
• /
• pp.29-29
• /
• 1999

• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1769-1780
• /
• 2022

A new method of dose calculation algorithm, called GPU-accelerated Monte Carlo and collapsed cone Convolution (GMCC) was developed to improve the calculation speed of BNCT treatment planning system. The GPU-accelerated Monte Carlo routine in GMCC is used to simulate the neutron transport over whole energy range and the Collapsed Cone Convolution method is to calculate the gamma dose. Other dose components due to alpha particles and protons, are calculated using the calculated neutron flux and reaction data. The mathematical principle and the algorithm architecture are introduced. The accuracy and performance of the GMCC were verified by comparing with the FLUKA results. A water phantom and a head CT voxel model were simulated. The neutron flux and the absorbed dose obtained by the GMCC were consistent well with the FLUKA results. In the case of head CT voxel model, the mean absolute percentage error for the neutron flux and the absorbed dose were 3.98% and 3.91%, respectively. The calculation speed of the absorbed dose by the GMCC was 56 times faster than the FLUKA code. It was verified that the GMCC could be a good candidate tool instead of the Monte Carlo method in the BNCT dose calculations.