• Nuclear Engineering and Technology
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• v.52 no.3
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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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• v.55 no.6
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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.

• Journal of Radiation Protection and Research
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• v.32 no.1
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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.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.20 no.5
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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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• v.14 no.2
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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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• v.38 no.3
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• pp.143-148
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• 2013

Operational nuclear facilities such as nuclear power plants and particle accelerators show various neutron spectra according to the type of facilities and specific position. Necessities of neutron dose management and neutron monitoring for radiation protection of radiation workers in such a kind of facilities have continuously increased in recent years. Bonner sphere spectrometers are widely used for measurement of neutron spectra. Data on response function of neutron detector, default neutron spectra and count rates of Bonner sphere spectrometer are required to obtain unfolded neutron spectra in specific workplaces. In this study, we carried out measurement of neutron spectra produced in MC50 cyclotron using Bonner sphere spectrometer with LiI scintillation detector. Additionally, we estimated quantitative data on neutron flux, mean neutron energy and ambient dose equivalent rate according to the incident proton energies and positions in MC50 cyclotron.

• Nuclear Engineering and Technology
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• v.46 no.4
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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.

• Journal of the Korea Concrete Institute
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• v.28 no.2
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• pp.197-204
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• 2016

Concrete walls of neutron generating facilities such as fusion reactors and fission reactors become radioactive by neutron irradiation. Both low-activation and neutron shielding are a critical concern at the dismantling stage after the shutdown of facilities with a requirement of radioactive waste management. To tackle this, two types of additives were investigated in fabricating mortar specimens: synthetic high polymers and boron carbide. It is well known that a hydrogen atom is effective in neutron shielding by an elastic scattering because its mass is almost the same as that of the neutron. And boron is an effective neutron absorber with a big neutron absorption cross section. In this study, the effect of the type, shape, and size of polymers were investigated as well as that of boron carbide. Total 16 mix designs were prepared to reveal the effect of polymers on mechanical properties and neutron shielding performance. The neutron does equivalent of polymers-based mortar for fast neutrons decreased by 36 %, and the count rate of boron carbide-based mortar with regard to thermal neutrons decreased by 90 % compared to conventional mortar. These results showed that a combination of polymers and boron carbide compounds has potential to reduce the thickness of neutron shields as well as radioactive waste from reactors.

• Nuclear Engineering and Technology
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• v.55 no.8
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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.

• Journal of Radiation Protection and Research
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• v.42 no.3
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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.