• 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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• 제55권6호
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• pp.2288-2297
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• 2023

Current designs of both large reactor units and small modular reactors utilize a nuclear fuel with increasing enrichment. This increasing demand for better nuclear fuel utilization is a challenge for nuclear fuel handling facilities. The operation with higher enriched fuels leads to reduced reserves to legislative and safety criticality limits of spent fuel transport, storage and final disposal facilities. Design changes in these facilities are restricted due to a boron content in steel and aluminum alloys that are limited by rolling, extrusion, welding and other manufacturing processes. One possible solution for spent fuel pools and casks is the burnup credit method that allows decreasing very high safety margins associated with the fresh fuel assumption in spent fuel facilities. This solution can be supplemented or replaced by an alternative solution based on placing the neutron absorber material directly into the fuel assembly, where its efficiency is higher than between fuel assemblies. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the fuel assemblies. The paper summarizes possibilities of fixed neutron absorbers for various nuclear fuel and fuel handling facilities. Moreover, an absorber material was optimized to propose alternative options to boron. Multiple effective absorbers that do not require steel or aluminum alloy compatibility are discussed because fixed absorbers are placed inside zirconium or steel cladding.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1895-1901
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• 2024

The fission model, G4ParaFissionModel, was enhanced in this study, mainly focusing on refining the energy dependence of the peak-to-valley ratio in the mass distribution and the energy dependence of the average total kinetic energy (TKE). The enhanced model was employed to investigate the characteristics of fission products from ²³⁵U(n, f) reaction. The calculated results, including fission yield, TKE distribution, prompt fission neutron and gamma spectra, were compared with both evaluated and experimental data. The comparison shows that these physical observables related nuclear data, which are of importance for developments of the nuclear power and physics, can be reasonably well reproduced.

• Journal of Radiation Protection and Research
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• 제32권1호
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• pp.15-20
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• 2007

An empirical formula fur the neutron dose equivalent at the maze entrance of medical accelerator treatment rooms was derived on the basis of a Monte Carlo simulation. The simulated neutron dose equivalents around the Varian medical accelerator by the MCNPX code were employed. Two cases of target rotational planes were considered: parallel and perpendicular to maze walls. Most of the maximum neutron dose equivalents at the doorway were found when the target rotational planes were parallel to maze walls and the beams were directed to the inner maze entrances. The neutron dose equivalents at the outer maze entrances were calculated for about 698 medical accelerator facilities which were generated from the geometry configurations of running treatment rooms, based on such gantry rotation that produces the maximum neutron dose at the doorway. The results calculated with the empirical formula in this study were compared with those calculated by the Kersey method for 7 operating facilities. It was found that the maximum disagreement between the calculation of this study and that of the Kersey method was a factor of 8.54 with the value calculated by the Kersey method exceeding that of this study. It was concluded that the kersey method estimated the neutron dose equivalent at the doorway computed by MCNPX more conservatively than this study technique.

• 한국전기전자재료학회논문지
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• 제20권5호
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• pp.467-470
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• 2007

Application field of neutron beam is very broad including industry, medicine and science. But the research and development and use of neutron beam is restricted within in narrow limits in this country, because neutron beam facility is insufficient - a big research facility of nuclear reactor(HANARO) and some small industrial facilities which use radioisotope neutron source are available. This paper compare and investigate the results of experiment and numerical analysis of the discharge in the spherically convergent beam fusion device which were expected as a portable neutron source. The spherically convergent beam fusion device will offer stability in neutron production, possibility of movement for convenience, low construction cost and higher neutron flux than radioisotope neutron source. The star mode discharge which efficiently generate neutron, were observed at both results.

• Journal of Radiation Protection and Research
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• 제14권2호
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• pp.51-62
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• 1989

The high efficiency moderated-type neutron spectrometer and doseequivalent counter were developed for the measurement of low level environmental neutrons. By using these detectors, the neutron energy spectra and dose equivalent rates due to skyshine effect were measured in the environment surrounding the accelerator facilities and also the altitude variation of cosmic ray neutrons in the aircraft flying over Japan.

• Journal of Radiation Protection and Research
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• 제38권3호
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• pp.143-148
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• 2013

원자력발전소 및 입자 가속기 등 상용화된 원자력관계시설에서는 해당 시설 내부의 구조 및 위치 등에 따라 다양한 에너지를 가진 중성자스펙트럼을 나타내며 이에 대한 중성자선량 평가 및 작업자의 방사선방호 조치에 대한 수요는 점차 증가하고 있다. 중성자스펙트럼을 측정하기 위해서 일반적으로 보너구 스펙트로메터가 많이 이용된다. 보너구 스펙트로메터를 이용한 중성자 스펙트럼 측정의 경우 중성자검출기의 반응함수, 초기추정 스펙트럼, 계수율 측정값을 사용하여 중성자스펙트럼 언폴딩을 수행하게 된다. 본 연구에서는 MC50 사이클로트론에서 발생하는 중성자스펙트럼을 LiI 섬광검출기 기반의 보너구 스펙트로메터를 이용하여 측정하였으며 표적에 입사되는 양성자 에너지와 위치에 따른 중성자선속, 중성자 평균에너지 및 선량률 자료를 정량적으로 평가하였다.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.501-512
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• 2014

HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.

• 콘크리트학회논문집
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• 제28권2호
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• pp.197-204
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• 2016

본 논문에서는 타 재료에 비해 수소 원소 함유량이 높아 고속 중성자 차폐에 유리한 합성 고분자 화합물과 중성자 포획단면적이 큰 붕소화합물을 각각 혼입한 모르타르를 대상으로 중성자 차폐성능을 분석하였다. 합성 고분자 화합물의 종류, 형상, 크기, 함량 및 붕소화합물의 함량에 따라 총 16개 모르타르 배합을 설계하였으며, 각 배합의 슬럼프 플로우, 28일 압축 및 인장강도를 측정하고, 고속 중성자 및 열중성자에 대한 차폐실험을 수행하였다. 합성 고분자 화합물의 선량 투과율은 모르타르 대비 최대 38.5%까지 감소하였으며, 공기 중 선량의 26.3%까지 차폐하였다. 붕소화합물을 혼입한 모르타르의 열중성자 차폐율은 최대 90.3%로 대부분의 열중성자를 차폐하였다. 비록 화합물 혼입에 의해 모르타르의 기본 특성은 저하되었으나, 표면 개질, 특수 혼화제 첨가 등 지속적인 연구를 통하여 성능 저하를 최소화할 수 있을 것으로 판단되며, 중성자 투과성능 역시 다양한 타입의 시험체 조합을 통한 레이어 시스템 도입 등으로 다양한 투과성능에 따른 맞춤형 설계를 제공할 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2984-2996
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• 2023

A 100 kW thermal power pool-type light water reactor and Pu(Be) as a fast neutron source were used to determine the appropriate carrier for irradiating boron-containing samples with neutron beams. The tested materials (carriers) were subjected to neutron beams in the reactor's tangential channel. The geometrical arrangement of experimental facilities relative to the neutron beam trajectory, as well as the effect of sample thickness on the count rate, were investigated. The majority of the detectable charged particles emitted by the neutron beam's interaction with tested materials and the detector's detecting layer are protons (recoiled hydrogen) and particles generated in nuclear reactions (protons and alpha particles), respectively. Stopping and Range of Ions in Matter (SRIM) software was used to do theoretical calculations for the range of expected released particles in various materials, including human tissue. The results of measurement and calculation are in good agreement. According to experiments and theoretical calculations, the number of protons emitted by tissue-like materials may commit a dose comparable to that of boron capture reactions. Furthermore, the range of protons is significantly larger than that of alpha particles, which most probably changes dose distribution in healthy cells surrounding the tumor, which is undesirable in the BNCT approach.