• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.203-209
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• 2004

In-phantom neutron flux distribution is measured at the HANARO BNCT irradiation facility. The measurements are performed with Au foil and wires. The thermal neutron flux and Cd ratio obtained at the HANARO BNCT facility are $1.19{\times}10^9\;n/cm^{2}s$ and 152, respectively, at 24 MW reactor power. The measured in-phantom neutron flux has a maximum value at a depth of 3 mm in the phantom and then decreases rapidly. The maximum flux is about $25\%$ larger than that of the phantom surface, and the measured value at a depth of 22 mm in the phantom is about a half of the maximum value. In addition, the neutron beam is limited well within the aperture of the neutron collimator. The two-dimensional in-phantom neutron flux distribution is determined. Significant neutron irradiation is observed within 20 mm from the phantom surface. The measured neutron flux distribution can be utilized in irradiation planning for a patient.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1075-1080
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• 2019

This paper presents a study on thermal neutron reflection properties of neutron guide for cylinder, spindle, elliptic and parabolic geometries using $^{241}Am-Be$ neutron source (5.2 Ci) and $BF_3$ detector, whereas neutron guide is important instrument for transportation of neutrons. To this goal, the required inner and outer radii of neutron guide have been calculated to achieve the highest guided thermal neutron flux based on MCNPX Monte Carlo code. The maximum flux of cylinder geometry with a length 50 cm has been obtained at an inner radius 9 cm and an outer radius 21 cm. Also, the maximum value of thermal neutron albedo is $0.46{\pm}0.001$ at 12 cm thickness of parabolic guide.

• Bulletin of the Korean Chemical Society
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• v.27 no.12
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• pp.2061-2063
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• 2006

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.937-942
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• 1995

Great interest has prompted Boron Neutron Capture Therapy (BNCT) as a new treatment for brain tumors. The use of $^{252}$Cf as a neutron source for BNn makes the in-hospital treatments of tumors to be possible. Newly proposed subcritical multiplying assemblies (SMA) are explored to improve relatively tow neutron fluxes of the source and construct the feasibilities of $^{252}$Cf as a neutron source. The MCNP code has been used to evaluate the effective multiplication factor of the entire system and the intensities and percentages of epithermal neutron flux at the patient-end surface of the system. The neutron beam using SMA shows the epithermal neutron flux enhancement of about 13 times as large as the beam without using SMA. It is expected that the neutron beam proposed in this research will be more effective for treatment of tumors due to the increased therapeutic neutron fluxes.

• Journal of Radiation Protection and Research
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• v.36 no.2
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• pp.59-63
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• 2011

It is important to reduce indirect scattered neutron beside direct neutron of chosen energy for designing a neutron-reference-field laboratory with neutron produced from a nuclear reaction by a accelerator. Therefore MCNPX simulation was performed with various conditions for obtaining such condition. At first in the original laboratory condition we calculated the direct neutron flux which was inserted in chamber (virtual chamber composed of air) of 0 degree (proton moving direction) for neutron flux measurement and the scattered neutron flux which is inserted in the chamber after scattering wall or bottom. In the result, the scattered neutron which was inserted after scattering bottom is more than that which was inserted after scattering the others. Therefore MCNPX simulation was again performed with removing the concrete bottom and with removing the concrete bottom and digging 1 m in the ground. In the result of removing concrete bottom and digging 1 m in the ground, scattered neutron which was inserted after scattering bottom became less than that which was inserted after scattering the others.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1871-1877
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• 2023

The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized ¹⁸⁶W and ⁸⁹Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers-lead moderator, graphite reflector, and polyethylene absorbent- is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for ¹⁸⁸W and ⁹⁰Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.203-210
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• 2018

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

• Progress in Medical Physics
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• v.29 no.4
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• pp.101-105
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• 2018

In this paper, a neutron moderation system for boron neutron capture therapy (BNCT) based on a $^{252}Cf$ neutron source is proposed. Different materials have been studied in order to produce a high percentage of epithermal neutrons. A moderator with a construction mixture of $AlF_3$ and Al, three reflectors of $Al_2O_3$, BeO, graphite, and seven filters (Bi, Cu, Fe, Pb, Ti, a two-layer filter of Ti+Bi, and a two-layer filter of Ti+Pb) is considered. The MCNPX simulation code has been used to calculate the neutron and gamma flux at the output window of the neutronic system. The results show that the epithermal neutron flux is relatively high for four filters: Ti+Pb, Ti+Bi, Bi, and Ti. However, a layer of Ti cannot reduce the contribution of ${\gamma}$-rays at the output window. Although the neutron spectra filtered by the Ti+Bi and Ti+Pb overlap, a large fraction of neutrons (74.95%) has epithermal energy when the Ti+Pb is used as a filter. However, the percentages of the fast and thermal neutrons are 25% and 0.5%, respectively. The Bi layer provides a relatively low epithermal neutron flux. Moreover, an assembly configuration of 30% $AlF_3+70%$ Al moderator/$Al_2O_3$ reflector/a two-layer filter of Ti+Pb reduces the fast neutron flux at the output port much more than other assembly combinations. In comparison with a recent model suggested by Ghassoun et al., the proposed neutron moderation system provides a higher epithermal flux with a relatively low contamination of gamma rays.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.787-791
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• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.575-600
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• 2024

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.