• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.942-948
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• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Progress in Medical Physics
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• v.18 no.2
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• pp.87-92
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• 2007

A thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the HANARO, 30 MW multi-purpose research reactor. Mixed beams with different physical characteristics and relative biological effectiveness would be emitted from the BNCT irradiation facility, so a quantitative analysis of each component of the mixed beams should be performed to determine the accurate delivered dose. Thus, various techniques were applied including the use of activation foils, TLDs and ionization chambers. All the dose measurements were perform ed with the water phantom filled with distilled water. The results of the measurement were compared with MCNP4B calculation. The thermal neutron fluxes were $1.02E9n/cm^2{\cdot}s\;and\;6.07E8n/cm^2{\cdot}s$ at 10 and 20 mm depth respectively, and the fast neutron dose rate was insignificant as 0.11 Gy/hr at 10 mm depth in water The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water Good agreement within 5%, has been obtained between the measured dose and the calculated dose using MCNP for neutron and gamma component and discrepancy with 14% for fast neutron flux Considering the difficulty of neutron detection, the current study support the reliability of these results and confirmed the suitability of the thermal neutron beam as a dosimetric data for BNCT clinical trials.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.113-122
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• 1990

After four years of planning, equipment acquisition, facility construction and beam testing, the KCCH cyclotron facility was put into operation in November1986. Now the KCCH cyclotron(MC-50) has been used for four years in neutron therapy and radioisotope production. Up to December 1989, 179(1852 sessions) patient have undergone neutron therapy. Radioisotope production for nuclear medicine use was started from March 1989 after extensive work to overcome target transport, target melting, beam diagnostic and chemical processing problems. This status report introduces the cyclotron facility, and the experiences of neutron therapy and isotope production with the MC-50 cyclotron. Besides, the operation results and the general troubles of the MC-50 during 1989 are summarized. Total operation time was 1252.5 hours. Four hundred hours were used for neutron therapy of 599 treatment sessions and 832.5 hours for radioisotope production. Total amount of produced raioisotope was 1695 mCi(Ga-67 : 1478mCi, Tl-201 : 107 mCi, I-123 : 25mCi, In-111 : 85mCi). Twenty hours were used for scheduled beam testing. In 1989, 882% of the planned operation were performed on schedule and this rats is improved remarkably compared to 71.0% in 1988.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2334-2337
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• 2022

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

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

• Journal of radiological science and technology
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• v.38 no.1
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• pp.39-49
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• 2015

One of properties which X-ray and Neutron can be applied nondestructive test is penetration into the object with interaction leads to decrease in intensity. X-ray interaction with the matter caused by electrons, Neutron caused by atoms. They share applications in nondestructive test area because of their similarities of interaction mechanism. Grating interferometer is the one of applications produces phase contrast image and dark field image. It is defined by Talbot interferometer and Talbot-Lau interferometer according to Talbot effect and Talbot-Lau effect respectively. Talbot interferometer works with coherence beam like X-ray, and Talbot-Lau has an effect with incoherence beam like Neutron. It is important to expect the interference in grating interferometer compared normal nondestructive system. In this paper, simulation works are conducted according to Talbot and Talbot-Lau interferometer in case of X-ray and Neutron. Variation of interference intensity with X-ray and Neutron based on wave theory is constructed and calculate elements consist the system. Additionally, Talbot and Talbot-Lau interferometer is simulated in different kinds of conditions.

• Radiation Oncology Journal
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• v.11 no.1
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• pp.127-132
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• 1993

Total of 53 patients of unresectable and recurrent rectal cancer treated with neutron beam during the period from Oct.1987 to Apr.1992 were analyzed. Dose fractionation for the neutron only group was 1.5 Gy per fraction,3 fraction per week,21 Gy/41/2 wks for 42 patients out of 53 ($76{\%}$). Neutron only but modified fractionation schedule ($10{\%}$ more or less of total dose) was applied for 9 patients, and mixed beam (neutron boost) was for 4 patients, Complete tumor response was obtained in 40 patients ($76{\%}$, response rate). Local control rate was 28 out of 53 ($53{\%}$). Statistically significant better prognostic factors for local control were age below 49 years old (15/22, $68{\%}$) than above 50 years old (13/31, $42{\%}$), male (20/32, $63{\%}$) than female (8/21, $38{\%}$), tumor size less than 5 cm and non-metastatic (16/24, $67{\%}$) than size more than 5 cm or metastatic (12/29, $41{\%}$). Major complication had developed in 7 patients ($13{\%}$). Two year overall survival rate by Kaplan-Meier method was $30{\%}$, but it was rised to, $47{\%}$ when the turner was less than 5 cm non-metastatic.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3413-3420
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• 2021

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D₁) and output (D₂) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D₁ = 1.8 cm and D₁ = 9.4 cm), previous design of the TRR with D₁ = 3 cm and D₁ = 9.4 cm, and another design with D₁ = 5 cm and D₁ = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2683-2689
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• 2024

This work investigated the most suitable type of degrader (Cu, Al or Nb) and its thickness, taking into consideration the salient aspects of concrete activation for high-purity ⁸⁹Zr production by ⁸⁹Y(p,n)⁸⁹Zr nuclear reaction. The MCNP and FISPACT codes were used to determine the optimal degrader thickness and the radioactivity of shielding concrete by neutron activation, respectively. The results showed that the optimal thickness of the beam degraders was 1.16, 3.19, and 1.33 mm for Cu, Al, and Nb, respectively. The neutron production rate per proton and the energy and angular distributions of neutrons varied depending on the type of degrader. Considering the radioactivity of the target-room concrete and the amount of radioactive waste expected to be generated, the use of a 1.33-mm-thick Nb degrader for ⁸⁹Zr production was determined to be the best choice.

• 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.