• 한국방사선학회논문지
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• 제1권1호
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• pp.45-49
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• 2007

동경공업대학교의 3MV 펠레트론가속기를 사용하여 10에서 90keV 영역에 대하여 $^{197}Au$의 중성자포획 스펙트럼을 측정하였다. 중성자 펄스빔은 $^7Li(p,n)^7Be$반응을 통하여 발생되었다. 사용되어진 양성자 빔의 폭은 1.5-ns였다. 금 시료에 입사된 중성자의 에너지 스펙트럼은 $^6Li$-glass 섬광검출기의 중성자 비행시간법을 사용하여 측정하였다. 금 시료의 중성자포획에 의해서 발생된 감마선은 anti-Compton NaI(TI) 검출장비를 사용하여 측정되었다. 본 연구에서는 5개의 중성자 에너지 역영을 선택했고, 각각의 에너지영역에서 얻어진 감마선파고스펙트럼을 표시하였다. 본 연구에서 얻어진 스펙트럼은 처음으로 얻어진 결과이며, 중성자 결합에너지부근에 몇 개의 천이 피크가 보인다.

• 한국방사선학회논문지
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• 제10권7호
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• pp.489-494
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• 2016

사이클로트론에서 발생되는 핵반응은 불필요한 중성자를 발생시키며, 이로 인해 주변 물질들이 방사화되게 된다. 방사화된 물질은 방사선피폭의 원인으로, 공기가 방사화 되었을 경우 인체에 흡입되어 내부피폭을 발생 시킨다. 이에 본 연구에서는 16.5 MeV의 초소형 사이클로트론의 운영에 따른 내부 공기의 방사화를 분석하고자 하였다. 실험결과 초소형 사이클로트론의 핵반응으로 발생되는 방사화는 종사자에게 매우 낮은 내부피폭을 발생시키는 것을 확인할 수 있었으며, 방사화로 인하여 발생된 방사능을 법적 기준치와 비교하여 보았을 때 기준치 이하로 법적 관리의 대상에서 제외 될 수 있음을 알 수 있었다. 하지만, 사이클로트론의 에너지가 높아짐에 따라 내부피폭의 위험성은 더욱 높아질 우려가 있으며, 이에 따라 국내에 정립 되어 있지 않는 방사선 관련 시설의 환기설비에 대한 기준이 필요할 것으로 사료되었다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.584-589
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• 2003

차세대관리 종합공정 실증시설 핫셀은 최대 1,385 TBq의 선원의 취급시에도 방사선 선량율을 법규에서 규제하는 허용치 이하로 차폐능을 가질 수 있도록 설계되고 있다. 선량 제한 설계치를 만족시키기 위하여 각 구역에 대한 차폐보강 방안이 수립되었으며, 이의 검증을 위하여 QAD-CGGP 및 MCNP-4C 코드를 이용하여 차폐 계산을 수행하여, 핫셀의 차폐 설계에 대한 안전성을 평가하였다. 핫셀 외벽에 대한 차폐 평가를 수행한 결과, QAD-CGGP 코드에 의한 작업구역에 대한 감마선 평가 결과는 $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ mSv/h, MCNP-4C 코드는 $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ mSv/h 이었으며, 서비스 구역은 $1.01{\times}10^{-2}$, $7.88{\times}10^{-2}$ mSv/h로 평가되었다 중성자에 의한 선량률은 감마선에 의한 선량률의 약 20% 이하치를 나타내는 것을 알 수 있었으며, 차폐벽의 각종 Penetration 및 Toboggan 경우 부분적인 납 차폐보강이 필요하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.614-620
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• 2003

원자력연구소에서는 국내 원전에서 배출된 사용후핵연료를 IMEF M6 핫셀에서 건식 재가공하여 건식공정 산화물핵연료를 개발하였다. 개발된 핵연료의 성능을 검증하기 위해서는 실제 상용로와 동일한 고온고압 조건하에서 조사시험이 필요하나 국내에는 이러한 조사시설을 갖추지 못하고 있으므로 핵연료 성능의 검증이 어렵던 차에 한$\cdot$카$\cdot$미 IAEA간의 국제공동연구 과제진도회의에서 AECL측은 중성자비를 받지 않고 캐나다 NRU에서 건식공정 산화물핵연료를 조사시험을 할 수 있다고 제안하였다. NRU 조사시험을 하고자 하는 핵연료는 건식공정 산화물핵연료봉 10개(약 6kgU)이며 운반물 분류등급에 따라 제7종 위험물로 핵분열성물질에 해당한다. 일반적으로 소량의 방사성물질을 운반할 경우에는 비용뿐 아니라 수송기간 측면에서 항공수송이 선박수송에 비해 유리한 것으로 알려져 있어 항공기를 이용한 건식공정 산화물핵연료의 해외 수송방안을 검토하였다. 검토결과, 현재 건식공정 산화물핵연료봉 10개를 운반할 수 있는 적절한 항공수송용 수송용기가 없어 항공수송이 불가능한 것으로 조사되었다. 선박을 이용한 해외 수송방안은 가능하나 이 경우에는 전용선박을 사용해야 함으로 비용이 많이 수요되는 것으로 분석되었다.

• 한국결정학회지
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• 제11권3호
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• pp.162-166
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• 2000

KD₂PO₄ single crystals were grown from D₂O with reagent KH₂PO₄ and the crystal structure was determined by X-ray and neutron diffraction methods. The crystals are tetragonal at room temperature, I42d, with lattice parameters of a=7.4633(7), c=6.9785(5) Å and Z=4. Intensity data were collected on an Enraf-nonius CAD4 diffractometer with a graphite monochromated MoK/sub α/ radiation (λ=0.7107Å) and on the neutron four circle single crystal diffractometer with Ge(331) monochromated neutron beam (λ=0.997Å). The structure was refined by full-matrix least-square to final R and wR values of 0.030 and 0.072, respectively, for 204 observed reflections with I>2σ(I) by X-ray diffraction and to final R=0.041 and wR=0.096 for 144 observed relfecdtions by neutron diffraction. The O…O distance of 2.516(4)Å obtained by X-ray diffraction is the same as that of 2.515(4)Å by neutron diffraction. On the other hand, the O-D/H distance of 0.84(4)Å by X-ray diffraction is considerably shorter than 1.029(7) Åby neutron diffraction. Hydrogen and deuterium can be readily distinguished by neutrons. In this crystal 66% of H-positions were substituted by D and the rest 34% occupied by H. The phase transition temperature of DKDP obtained with deuteration levels is f193K. This value agrees fairly well with the result of DSC measurement. The nuclear density distribution by neutron diffraction provides an observation of the disordered state of D/H in KD₂PO₄ at room temperature.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.531-544
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• 2009

This paper is intended to provide key issues and current research outcomes on accelerator-based Boron Neutron Capture Therapy (BNCT). Accelerator-based neutron sources are efficient to provide epithermal neutron beams for BNCT; hence, much research, worldwide, has focused on the development of components crucial for its realization: neutron-producing targets and cooling equipment, beam-shaping assemblies, and treatment planning systems. Proton beams of 2.5 MeV incident on lithium target results in high yield of neutrons at relatively low energies. Cooling equipment based on submerged jet impingement and micro-channels provide for viable heat removal options. Insofar as beam-shaping assemblies are concerned, moderators containing fluorine or magnesium have the best performance in terms of neutron accumulation in the epithermal energy range during the slowing-down from the high energies. NCT_Plan and SERA systems, which are popular dose distribution analysis tools for BNCT, contain all the required features (i.e., image reconstruction, dose calculations, etc.). However, detailed studies of these systems remain to be done for accurate dose evaluation. Advanced research centered on accelerator-based BNCT is active in Korea as evidenced by the latest research at Hanyang University. There, a new target system and a beam-shaping assembly have been constructed. The performance of these components has been evaluated through comparisons of experimental measurements with simulations. In addition, a new patient-specific treatment planning system, BTPS, has been developed to calculate the deposited dose and radiation flux in human tissue. It is based on MCNPX, and it facilitates BNCT efficient planning based via a user-friendly Graphical User Interface (GUI).

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

• Nuclear Engineering and Technology
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• 제27권1호
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• pp.74-83
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• 1995

원자로 압력용기에서의 정화한 속중성자 조사량의 계산은 발전소 압력용기 surveillance program의 핵심적인 부문이다. 최근 기존의 ENDF /B-III~V에 있는 철의 핵단면적 자료가 압력용기와 같은 철이 포함된 구조물에서 속중성자속을 낮게 평가하는 것으로 알려지고 있다. 본 논문에서는 ENDF /B-IV와 VI의 철(Fe) 자료의 비교를 위해 영광3/4호기 모델과 2개의 ENDF/B 파일에 있는 각각의 철자료를 이용하여 47-에너지그룹 핵단면적집 (CXFe-IV와 CXFe-VI )을 만들었다. CXFe-IV와 CXFe-VI를 사용하여 수행한 DOT4.3 계산결과에 의하면 압력용기 취화해석에 중요한 속중성자속(E 〉 1.0 MeV) 계산에서 ENDF /B-VI의 철자료를 사용한 경우가 ENDF /B-IV의 철자료를 사용한 경우보다 압력용기 내부표면에서 7.6%, 외부표면에서 20% 높게 나타났다.

• Journal of Radiation Protection and Research
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• 제42권3호
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• pp.141-145
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• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.344-349
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• 2016

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.