• Journal of Sensor Science and Technology
• /
• v.18 no.5
• /
• pp.402-407
• /
• 2009

In this study, plastic optical fiber sensors which can measure thermal neutron dose in a mixed neutron-gamma field are developed and characterized. Using $^{252}Cf$ and $^{60}Co$ sources, the scintillators suitable for thermal neutron detection, are tested and the scintillating lights generated from a plastic optical fiber sensor in the Kyoto University Critical Assembly (kuca) core are measured. Also, the distributions of thermal neutron and gamma-ray are measured in a mixed field as a function of the distance from the center of the reactor core at KUCA and the distribution of thermal neutron is obtained using a subtraction method. Sensitivity of the fiber-optic radiation sensor system is about 0.49 V/mW according to power of the KUCA core and its relative error is about 1.2 %.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1807-1816
• /
• 2020

In this work, a Gaussian Process (Kriging) approach is proposed to provide efficient dose mapping for complex radiation fields using limited number of responses. Given a few response measurements (or simulation data points), the proposed approach can help the analyst in completing a map of the radiation dose field with a 95% confidence interval, efficiently. Two case studies are used to validate the proposed approach. The First case study is based on experimental dose measurements to build the dose map in a radiation field induced by a D-D neutron generator. The second, is a simulation case study where the proposed approach is used to mimic Monte Carlo dose predictions in the radiation field using a limited number of MCNP simulations. Given the low computational cost of constructing Gaussian Process (GP) models, results indicate that the GP model can reasonably map the dose in the radiation field given a limited number of data measurements. Both case studies are performed on the nuclear engineering radiation laboratories at the University of Sharjah.

• Korean Journal of Materials Research
• /
• v.27 no.6
• /
• pp.345-349
• /
• 2017

The effects of electron beam(EB) irradiation on the electrical and optical properties of InGaZnO(IGZO) thin films fabricated using a sol-gel process were investigated. As the EB dose increased, the electrical characteristic of the IGZO TFTs changed from semiconductor to conductor, and the threshold voltage values shifted to the negative direction. X-ray photoelectron spectroscopy analysis of the O 1s core level showed that the relative area of oxygen vacancies increased from 14.68 to 19.08 % as the EB dose increased from 0 to $1.5{\times}10^{16}electrons/cm^2$. In addition, spectroscopic ellipsometer analysis showed that the optical band gap varied from 3.39 to 3.46 eV with increasing EB dose. From the result of band alignment, it was confirmed that the Fermi level($E_F$) of the sample irradiated with $1.5{\times}10^{16}electrons/cm^2$ was located at the closest position to the conduction band minimum(CBM) due to the increase of electron carrier concentration.

• Radiation Oncology Journal
• /
• v.6 no.1
• /
• pp.85-91
• /
• 1988

For the physical characterization of neutron beam, dosimetric measurements had been performed to obtain physical data of KCCH cyclotron-produced neutrons for clinical use. The results are presented and compared with the data of other institutions from the literatures. The central axis percent depth dose, build-up curves and open and wedge isodose curve values are intermediate between that of a 4 and 6 MV X-rays. The build-up level of maximum dose was at 1.35cm and entrance dose was approximately $40\%$. Flatness of the beam was $9\%$ at Dmax and less $than{\pm}3\%$ at the depth of $80\%$ isodose line. Penumbra begond the $20\%$ line is wider than corresponding photon beam. The output factors ranged 0.894 for $6\times6cm$ field to 1.187 for $30\times30cm$ field. Gamma contamination of neutron beam was $4.9\%$ at 2 cm depth in $10\times10cm$ field.

• The Journal of Korean Society for Radiation Therapy
• /
• v.17 no.2
• /
• pp.141-145
• /
• 2005

Purpose : It is known that the neutron is generally generated from the photon, its energy is larger than 10 MV. The neutron is leaked in the container inspection system installed at the customs though its energy is below 9 MV. It is needed that the spacial effect of the neutrons released from radiation treatment machine, linac, installed in the medical canter. Materials and Methods : The medical linear accelerator(Clinac 1800, varian, USA) was used in the experiment. Measuring neutron was used bubble detector(Bubble detector, BDPND type, BTI, Canada) which was created bubble by neutron. The bubble detector is located on the medical linear accelerator outskirt in three different distance, 30, 50, 120 cm and upper, lower four point from the iso-center. In addition, for effect on protect material we have measured eight points which are 50 cm distance from iso-center. The SAD(source-axis-distance), distance from photon source to iso-center, is adjusted to 100 cm and the field size is adjusted to $15{\times}15cm^2$. Irradiate 20 MU and calculate the dose rate in mrem/MU by measuring the number of bubble. Results : The neutron is more detected at 5 position in 30, 50 cm, 7 position in 120 cm and with wedge, and 2 position without mount. Conclusion : Though detection position is laid in the same distance in neutron measurement, the different value is shown in measuring results. Also, neutron dose is affected by the additional structure, the different value is obtained in each measurement positions. So, it is needed to measure and evaluate the neutron dose in the whole space considering the effect of the distance, angular distribution and additional structure.

• Nuclear Engineering and Technology
• /
• v.49 no.1
• /
• pp.209-215
• /
• 2017

Fast neutron irradiation was used to improve the switching speed of a 600-V nonpunch-through insulated gate bipolar transistor. Fast neutron irradiation was carried out at 30-MeV energy in doses of $1{\times}10^8n/cm^2$, $1{\times}10^9n/cm^2$, $1{\times}10^{10}n/cm^2$, and $1{\times}10^{11}n/cm^2$. Electrical characteristics such as current-voltage, forward on-state voltage drop, and switching speed of the device were analyzed and compared with those prior to irradiation. The on-state voltage drop of the initial devices prior to irradiation was 2.08 V, which increased to 2.10 V, 2.20 V, 2.3 V, and 2.4 V, respectively, depending on the irradiation dose. This effect arises because of the lattice defects generated by the fast neutrons. In particular, the turnoff delay time was reduced to 92 nanoseconds, 45% of that prior to irradiation, which means there is a substantial improvement in the switching speed of the device.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.293-302
• /
• 2019

Many researches have been done to develop and improve the performance of personal (individual) dosimeter response to cover a wide of neutron energy range (from thermal to fast). Depending on the individual category of the dosimeter, the semiconductor sensor has been used to simplify and lightweight. In this plan, it's very important to have a fairly accurate counting of doses rate in different energies. With a general design and single-sensor simulations, all optimal thicknesses have been extracted. The performance of the simulation scheme has been compared with the commercial and laboratory samples in the world. Due to the deviation of all dosimeters with a flat energy response, in this paper, has been used an idea of one semi-conductor sensor to have the flat energy-response in the entire neutron energy range. Finally, by analyzing of the sensors data as arrays for the first time, we have reached a nearly flat and acceptable energy-response. Also a comparison has been made between Lucite-PMMA ($H_5C_5O_2$) and polyethylene-PE ($CH_2$) as a radiator and $B_4C$ has been studied as absorbent. Moreover, in this paper, the effect of gamma dose in the dosimeter has been investigated and shown around the standard has not been exceeded.

• Nuclear Engineering and Technology
• /
• v.51 no.8
• /
• pp.2018-2025
• /
• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2767-2773
• /
• 2021

We used the GEANT4 Monte Carlo MC Toolkit to simulate carbon ion beams incident on water, tissue, and bone, taking into account nuclear fragmentation reactions. Upon increasing the energy of the primary beam, the position of the Bragg-Peak transfers to a location deeper inside the phantom. For different materials, the peak is located at a shallower depth along the beam direction and becomes sharper with increasing electron density NZ. Subsequently, the generated depth dose of the Bragg curve is then benchmarked with experimental data from GSI in Germany. The results exhibit a reasonable correlation with GSI experimental data with an accuracy of between 0.02 and 0.08 cm, thus establishing the basis to adopt MC in heavy-ion treatment planning. The Kolmogorov-Smirnov K-S test further ascertained from a statistical point of view that the simulation data matched the experimentally measured data very well. The two-dimensional isodose contours at the entrance were compared to those around the peak position and in the tail region beyond the peak, showing that bone produces more dose, in comparison to both water and tissue, due to secondary doses. In the water, the results show that the maximum energy deposited per fragment is mainly attributed to secondary carbon ions, followed by secondary boron and beryllium. Furthermore, the number of protons produced is the highest, thus making the maximum contribution to the total dose deposition in the tail region. Finally, the associated spectra of neutrons and photons were analyzed. The mean neutron energy value was found to be 16.29 MeV, and 1.03 MeV for the secondary gamma. However, the neutron dose was found to be negligible as compared to the total dose due to their longer range.

• Progress in Medical Physics
• /
• v.25 no.3
• /
• pp.151-156
• /
• 2014

The purpose of this study was to confirm the feasibility of imaging of therapy region from the boron neutron capture therapy (BNCT) using the measurement of the prompt gamma ray depending on the neutron flux. Through the Monte Carlo simulation, we performed the verification of physical phenomena from the BNCT; (1) the effects of neutron according to the existence of boron uptake region (BUR), (2) the internal and external measurement of prompt gamma ray dose, (3) the energy spectrum by the prompt gamma ray. All simulation results were deducted using the Monte Carlo n-particle extended (MCNPX, Ver.2.6.0, Los Alamos National Laboratory, Los Alamos, NM, USA) simulation tool. The virtual water phantom, thermal neutron source, and BURs were simulated using the MCNPX. The energy of the thermal neutron source was defined as below 1 eV with 2,000,000 n/sec flux. The prompt gamma ray was measured with the direction of beam path in the water phantom. The detector material was defined as the lutetium-yttrium oxyorthosilicate (Lu0,6Y1,4Si0,5:Ce; LYSO) scintillator with lead shielding for the collimation. The BUR's height was 5 cm with the 28 frames (bin: 0.18 cm) for the dose calculation. The neutron flux was decreased dramatically at the shallow region of BUR. In addition, the dose of prompt gamma ray was confirmed at the 9 cm depth from water surface, which is the start point of the BUR. In the energy spectrum, the prompt gamma ray peak of the 478 keV was appeared clearly with full width at half maximum (FWHM) of the 41 keV (energy resolution: 8.5%). In conclusion, the therapy region can be monitored by the gamma camera and single photon emission computed tomography (SPECT) using the measurement of the prompt gamma ray during the BNCT.