• Title/Summary/Keyword: Neutron Flux

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Measurements of In-phantom Neutron Flux Distribution at the HANARO BNCT Facility

  • Kim Myong Seop;Park Sang Jun;Jun Byung Jin
    • 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.

Thermal neutron albedo and flux for different geometries neutron guide

  • Azimkhani, S.;Rezaei Ochbelagh, D.;Zolfagharpour, F.
    • 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.

Application of data-driven model reduction techniques in reactor neutron field calculations

  • Zhaocai Xiang;Qiafeng Chen;Pengcheng Zhao
    • Nuclear Engineering and Technology
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    • v.56 no.8
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    • pp.2948-2957
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    • 2024
  • High-order harmonic techniques can be used to recreate neutron flux distributions in reactor cores using the neutron diffusion equation. However, traditional source iteration and source correction iteration techniques have sluggish convergence rates and protracted calculation periods. The correctness of the implicitly restarted Arnoldi method (IRAM) in resolving the eigenvalue problems of the one-dimensional and two-dimensional neutron diffusion equations was confirmed by computing the benchmark problems SLAB_1D_1G and two-dimensional steady-state TWIGL using IRAM. By integrating Galerkin projection with Proper Orthogonal Decomposition (POD) techniques, a POD-Galerkin reduced-order model was developed and the IRAM model was used as the full-order model. For 14 macroscopic cross-section values, the TWIGL benchmark problem was perturbed within a 20% range. We extracted 100 sample points using the Latin hypercube sampling method, and 70% of the samples were used as the testing set to assess the performance of the reduced-order model The remaining 30% were utilized as the training set to develop the reduced-order model, which was employed to rebuild the TWIGL benchmark problem. The reduced-order model demonstrates good flexibility and can efficiently and accurately forecast the effective multiplication factor and neutron flux distribution in the core. The reduced-order model predicts keff and neutron flux distribution with a high degree of agreement compared to the full-order model. Additionally, the reduced-order model's computation time is only 10.18% of that required by the full-order model.The neutron flux distribution of the steady-state TWIGL benchmark was recreated using the reduced-order model. The obtained results indicate that the reduced-order model can accurately predict the keff and neutron flux distribution of the steady-state TWIGL benchmark.Overall, the proposed technique not only has the potential to accurately project neutron flux distributions in transient settings, but is also relevant for reconstructing neutron flux distributions in steady-state conditions; thus, its applicability is bound to increase in the future.

Epithermal Neutron Flux Enhancement Using SMA in Designing a Cf-Based Neutron Beam for BNCT

  • Kim, Do-Heon;Kim, Jong-Kyung
    • 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.

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MCNPX Simulation of Scattered Neutron Distribution in Experimental Room for the Neutron Reference Field of Monoenergetic Neutron below 2.5 MeV (2.5 MeV 이하 단색 중성자 표준장에 대한 중성자 실험실내의 산란 중성자 분포 전산모사)

  • Park, Jung-Hun;Kim, Gi-Dong
    • 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.

Nano Yttrium-90 and Rhenium-188 production through medium medical cyclotron and research reactor for therapeutic usages: A Simulation study

  • Abdollah Khorshidi
    • 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.

Calibration of digital wide-range neutron power measurement channel for open-pool type research reactor

  • Joo, Sungmoon;Lee, Jong Bok;Seo, Sang Mun
    • 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.

Assembly Neutron Moderation System for BNCT Based on a 252Cf Neutron Source

  • Gheisari, Rouhollah;Mohammadi, Habib
    • 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.

Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

  • Didi, Abdessamad;Dadouch, Ahmed;Jai, Otman;Tajmouati, Jaouad;Bekkouri, Hassane El
    • 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.