• 한국의학물리학회지:의학물리
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• 제18권2호
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• pp.87-92
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• 2007

최대출력 30 MW, 하나로(HANARO) 다목적 연구용 원자로의 접선 중성자공에 붕소중성자포획치료(Boron Neutron Capture Therapy, BNCT)를 위한 열중성자 조사장치가 개발되었다. BNCT 조사장치에서는 서로 다른 물리적 특성과 생물학적 효과비를 가진 여러 성분의 방사선이 방출되기 때문에 정확한 투여선량을 결정하기 위해서는 각 성분의 정량적 분석이 필수적이다. 따라서 본 연구에서는 방사화 분석, 열형광선량계 및 이온전리함 등 여러 유형의 검출기를 사용하여 BNCT 조사장치에서 방출되는 열중성자 및 감마선 혼합장의 선량 성분을 분리, 측정하였다. 선량측정은 물 속에 함유된 불순물과 중성자의 이차반응을 최소화하기 위해 증류수를 채운 물팬텀을 이용하였다. 그리고 측정 결과는 MCNP4B 전산계산의 결과와 상호 비교하였다. 측정 결과 열중성자속은 물팬텀 10 mm와 20 mm 깊이에서 각각 $1.02E9n/cm^2{\cdot}s$과 $6.07E8n/cm^2{\cdot}s$이었고, 고속중성자선량율은 10 mm 깊이에서 0.11 Gy/hr로 미세하였다. 감마선량률은 물팬텀 20 mm 깊이에서 5.10 Gy/hr로 나타났다. 측정된 중성자와 감마선량값은 MCNP의 결과와 5% 이내로 잘 일치하였고, 열중성자속은 14%의 비교오차를 나타내었다. 이러한 결과들은 중성자 검출의 난이도를 고려할 때 충분히 신뢰할 수 있는 수준이라 판단되며, BNCT 임상 연구를 위한 선량평가 자료로 활용할 수 있을 것으로 사료된다.

• 한국방사선학회논문지
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• 제13권2호
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• pp.253-260
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• 2019

방사선 선량 증강은 물질과의 상호작용 단면적을 높여 국소 부위에 대한 선량을 증가시키는 방법으로 선에너지부여 및 상대적 생물학적 효과비 증가로 치료가능비 향상에 기여할 수 있다. 선량 증강에 대한 선행 연구는 X, ${\gamma}$선에 대한 보고가 주를 이루고 있으나, 본 연구에서는 MCNP6를 이용한 몬테칼로 시뮬레이션을 바탕으로 양성자 선원에 대해 선량 증강 현상을 분석하였다. 수학적 모델 방법에 따라 확산된 피크의 양성자 선원에 대한 에너지 분포와 상대적 강도를 산출하였으며, 금, 아이오딘, 가돌리늄의 선량 증강 물질에 대한 선량증강비와 깊이 변화에 따른 에너지 분포를 평가하였다. 금을 이용한 증강 현상에서 1.085-1.120배, 가돌리늄에서는 1.047-1.091배의 선량증강비를 나타내었다. 또한 깊이에 따른 흡수에너지 변화로 인해 실질 비정과 95% 선량 구간의 감소를 나타내었으며, 이는 선량 증강 현상과 더불어 종양조직에 불확실한 선량 전달로 이어질 수 있으므로 증강 물질의 질량 저지능으로부터 확산된 피크 구간의 적절한 보정이 필요할 것으로 사료된다. 본 연구에서 모의모사를 통한 선량 증강 현상의 분석은 실질적인 증강 효과 확인을 위한 체내 외 실험의 기초자료로써 활용될 것으로 기대한다.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.197-202
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• 1997

ORIGEN2코드의 검증계산을 통해 PWR 사용 후 핵연료 조성핵종의 핵종량에 대한 핵임계측면에서 보수성을 가지는 안전인자를 산출하였고, MCNP코드의 검증계산으로 95/95 신뢰구간에서의 계산오차를 구하였다. 이를 바탕으로 직경이 1.2567cm이고 길이가 380.5cm인 196 개 금속봉을 장전한 캐니스터 ( 금속저장체 )가 x-y 방향으로 무한히 배열된 경우에 대해 캐니스터의 두께, 간격 및 외부의 공기중 수분농도에 따른 핵임계 안전해석을 수행하였다. 그 결과, 캐니스터의 두께가 7mm일 때 공기중 수분농도가 0.30 g/㎤이고 캐니스터간의 간격이 6.0cm인 경우의 최종핵 임계도값은 0.94130로서 최대허용핵임계값 (0.942)보다 적은 값을 보였다.

• 방사선산업학회지
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• 제6권1호
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• pp.95-100
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• 2012

In this study, there has been investigated the simulation of irradiation dose using Monte Carlo methodology for the biological control of wooden cultural property. In the evaluation of fungal contamination on wooden cultural properties, Cladosporium tenuissimum, Aspergillus versicolor, Penicillium sp. were mainly identified from the Gimjeotgae. But these microorganisms were completely inactivated by 20 kGy gamma-rays. For dosimetry simulation of wooden cultural properties, Monte Carlo methodology with MCNP was used. The radiation absorbed dose distribution was predicted at 8.2~18.9 kGy. These results show that irradiation is effective for biologic control of wooden cultural properties and Monte Carlo methodology is useful for non-destructive conservation and preservation of wooden cultural properties.

• 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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• 제48권5호
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.93-98
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• 2006

The radial uniformity of neutron irradiation in silicon ingots for neutron transmutation doping (NTD) at HANARO is examined by both calculations and measurements. HANARO has two NTD holes named NTD1 and NTD2. We have been using the NTD2 hole for 5 in. NTD commercial service, and we intend to use two holes for 6 in. NTD. The objective of this study is to predict the radial uniformity of 6 in. NTD at the two holes. The radial neutron flux distributions inside single crystal and noncrystal silicon loaded at the NTD2 hole are calculated by the VENTURE code. For NTD1, the radial distributions of the reaction rate for a 6 in. NTD with a neutron screen are calculated by MCNP, and measured by gold wire activation. The results of the measurements are compared with those of the calculations. From the VENTURE calculation, it is confirmed that the neutron flux distribution in the single crystal silicon is much flatter than that in the non-crystal silicon. The non-uniformities of the measurements for radial neutron irradiation are slightly larger than those of the calculations. However, excluding local dips in the measurements, the overall trends of the distributions are similar. The radial resistivity gradient (RRG) for a 5 in. silicon ingot is estimated to be about $1.5\%$. For a 6 in. ingot, the RRG of a silicon ingot irradiated at HANARO is predicted to be about $2.1\%$. Also, from the experimental results, we expect that the RRG would not be larger than $4.4\%$.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1425-1430
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• 2024

This study develops superheated emulsion detectors that are both sensitive to fast neutrons, and thermal neutrons owing to the exergonic ⁶₃Li(n, α)³₁H capture reaction caused by the ⁶Li-containing compound dispersed throughout the gel-like medium. The experimental research was conducted on two SEDs. One detector was an ordinary Freon-12 detector and the other was a Freon-12 detector containing 3.4 % (by weight) LiCl. In order to investigate the sensitivity of lithium-containing SEDs to thermal neutrons, two types of SEDs were simultaneously exposed to various flux levels of thermal neutrons from ²⁴¹Am-Be neutron source inside a cylindrical tank filled with water. A Boron-lined proportional counter was used to estimate the thermal neutron flux and the relevant MCNP code was developed for flux and dose calculations in the prepared set-up around ²⁴¹Am-Be source. The results demonstrate that there is a proportional relationship between the variations of SED response and the change in thermal neutron flux and dose. Also, the sensitivity of SED was estimated.

• 방사성폐기물학회지
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• 제6권1호
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• pp.11-15
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• 2008

고밀도폴리 에틸렌 고건전성용기에 담은 폐수지의 운반을 위해 원전 폐수지의 방사능 분석결과를 사용하여 운반물 등급 분류방법을 도출하였다. 원전 폐수지의 방사능 분석결과로부터 폐수지 내 핵종 존재비를 구하였고, 폐수지의 표면선량률로 핵종재고량을 평가하기 위해 MCNP 코드로 방사능대선량 환산인자를 모사하였다. 이로부터 고밀도폴리에틸렌 고건전성용기에 담은 폐수지에 대한 A형 운반물과 B형 운반물의 경계값은 1.19 TBq 이고 이를 표면선량으로 환산한 결과는 124.2 mSv/h임을 알 수 있었다.