• Nuclear Engineering and Technology
• /
• v.47 no.7
• /
• pp.939-944
• /
• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.715-727
• /
• 2024

During fast neutron imaging, besides the dark current noise and readout noise of the CCD camera, the main noise in fast neutron imaging comes from high-energy gamma rays generated by neutron nuclear reactions in and around the experimental setup. These high-energy gamma rays result in the presence of high-density gamma white spots (GWS) in the fast neutron image. Due to the microscopic quantum characteristics of the neutron beam itself and environmental scattering effects, fast neutron images typically exhibit a mixture of Gaussian noise. Existing denoising methods in neutron images are difficult to handle when dealing with a mixture of GWS and Gaussian noise. Herein we put forward a deep learning approach based on the Swin Transformer UNet (SUNet) model to remove high-density GWS-Gaussian mixture noise from fast neutron images. The improved denoising model utilizes a customized loss function for training, which combines perceptual loss and mean squared error loss to avoid grid-like artifacts caused by using a single perceptual loss. To address the high cost of acquiring real fast neutron images, this study introduces Monte Carlo method to simulate noise data with GWS characteristics by computing the interaction between gamma rays and sensors based on the principle of GWS generation. Ultimately, the experimental scenarios involving simulated neutron noise images and real fast neutron images demonstrate that the proposed method not only improves the quality and signal-to-noise ratio of fast neutron images but also preserves the details of the original images during denoising.

• 한국초전도학회:학술대회논문집
• /
• 2008.08a
• /
• pp.20-20
• /
• 2008

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.1069-1070
• /
• 2005

• The Transactions of the Korean Institute of Electrical Engineers C
• /
• v.52 no.12
• /
• pp.545-552
• /
• 2003

When a Si PIN diode is exposed to fast neutrons, it results in displacement damage to the Si lattice structure of the diode. Defects induced from structural dislocation become effective recombination centers for carriers which pass through the base of a PIN diode. Hence, increasing the resistivity of the diode decreases the current for the applied forward voltage. This paper involves the development of a neutron sensor based on the phenomena of the displacement effect damaged by neutron exposure. The neutron effect on the semiconductor was analyzed. Several PIN diode arrays with various thickness and cross-section area of the intrinsic layer(I layer) were fabricated. Under irradiation tests with a neutron beam, the manufactured diodes have a good linearity to neutron dose and show that the increase of thickness of I layer and the decrease of cross-section of PIN diodes improve the sensitivity. Newly developed PIN diodes with thicker I layer and various cross section, were retested and then showed the best neutron sensitivity at the condition that the I layer thickness was similar to a side length. On the basis of two test results, final discrete PIN diodes with a rectangular shape were manufactured and the characteristics as neutron detectors were analyzed through the neutron beam test using on-line electronic dosimetry system. Developed PIN diode shows a good linearity as dosimetry in the range of 0 to 1,000cGy(Tissue) and its neutron sensitivity is 13mV/cGy at constant current of 5mA, that is three times higher than that of commercially available neutron detectors. And the device shows little dependency on the orientation of the neutron beam and a considerable stability in annealing test for a long period.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2364-2370
• /
• 2021

The ⁿMCP (Neutron sensitive microchannel plate) combined with advanced readout electronics is widely used in energy selective neutron imaging because of its good spatial and timing resolution. Neutron detection efficiency is a crucial parameter for the ⁿMCP. In this paper, a mathematical model based on the oblique cylindrical channel and elliptical pore was established to calculate the neutron absorption probability, the escape probability of charged particles and overall detection efficiency of ⁿMCP and analyze the effects of neutron incident position, pore diameter, wall thickness and bias angle. It was shown that when the doping concentration of the ⁿMCP was 10 mol%, the thickness of ⁿMCP was 0.6 mm, the detection efficiency could reach maximum value, about 24% for thermal neutrons if the pore diameter was 6 ㎛, the wall thickness was 2 ㎛ and the bias angle was 3 or 6°. The calculated results are of great significance for evaluating the detection efficiency of the ⁿMCP. In a subsequent companion paper, the mathematical model would be extended to the case of the spatial resolution and detection efficiency optimization of the coating ⁿMCP.

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

• Journal of Sensor Science and Technology
• /
• v.22 no.3
• /
• pp.223-226
• /
• 2013

We designed a single polyethylene bonner sphere with several thermo-luminescence dosimeters (TLD), for measurement of neutron energy spectrum. For the separation of the neutron dosage in the neutron-gamma mixed field, we used 21 ea TLD-600s and TLD-700s, respectively. Because, TLD-600 is sensitive to neutron and gamma rays, and, TLD-700 is sensitive only to gamma-rays, we could determine the each dose by neutron and gamma rays. The neutron response function of the bonner sphere with TLDs was calculated by MCNPX (ver. 2.5.0) Monte Carlo simulation in the energy range from $10^{-1}$ to 20 MeV. For the Cf-252 standard neutron source in KRISS, we could estimate the neutron energy spectrum by unfolding method using the response function.

• The Journal of Korean Society for Radiation Therapy
• /
• v.9 no.1
• /
• pp.71-81
• /
• 1997

I. Objective and Importance of the Project We have been using MC-50 cyclotron and NT-50 neutron therapy machine for treating cancer patients since 1986 at Korea Cancer Center Hospital. It is mandatory to measure accurately the dose distribution and the total absorbed dose of fast neutron for putting it to the clinical use. At present the methods of measurement of fast neutron are proposed largely by American Associations of Physicists in Medicine (Task Group 18), European Clinical Neutron Dosimetry Group, and International Commission on Radiation Units and Measurements. The complexity of measurement, however, induce the methodological differences between them. In our study, therefore, we tried to establish a unique technique of measurement by means of measuring the emitted doses and the dose distribution of fast neutron beam from neutron therapy machine, and to invent a standard method of measurement adequate to our situation. II. Scope and Contents of the Project For establishing a unique technique of measurement and inventing a standard method of measurement of fast neutron beam, 1. to grasp the physical characteristics of neutron therapy machine 2. to study the principles for measrement of fast neutron beam 3. to get the dose distribution (dose rate, percent-depth dose, flatness etc) throught the actual measurement 4. to compare our data with those being cited world-widely.

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