• Journal of Radiation Protection and Research
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• v.40 no.1
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• pp.1-9
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• 2015

Compton cameras overcome several limitations of conventional mechanical collimation based gamma imaging devices, such as pin-hole imaging devices, due to its electronic collimation based on coincidence logic. Especially large-scale Compton camera has wide field of view and high imaging sensitivity. Those merits suggest that a large-scale Compton camera might be applicable to monitoring nuclear materials in large facilities without necessity of portability. To that end, our research group have made an effort to design a large-scale Compton camera for safeguard application. Energy resolution or position resolution of large-area detectors vary with configuration style of the detectors. Those performances directly affect the image quality of the large-scale Compton camera. In the present study, a series of Geant4 Monte Carlo simulations were performed in order to examine the effect of those detector parameters. Performance of the designed large-scale Compton camera was also estimated for various monitoring condition with realistic modeling. The conclusion of the present study indicates that the energy resolution of the component detector is the limiting factor of imaging resolution rather than the position resolution. Also, the designed large-scale Compton camera provides the 16.3 cm image resolution in full width at half maximum (angular resolution: $9.26^{\circ}$) for the depleted uranium source considered in this study located at the 1 m from the system when the component detectors have 10% energy resolution and 7 mm position resolution.

• Progress in Medical Physics
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• v.19 no.4
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• pp.291-297
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• 2008

The purpose of this study is to optimize a parallel-hole collimator for small gamma camera having the pixellated crystal array and evaluate the effect of crystal-collimator misalignment on the image quality using a simulation tool GATE (Geant4 Application for Tomographic Emission). The spatial resolution and sensitivity were measured for the various size of hexagonal-hole and matched square-hole collimators with a Tc-99m point source and the uniformity of flood image was estimated as a function of the angle between crystal array and collimator by misalignment. The results showed that the spatial resolution and sensitivity were greatly improved by using the matched collimator and the uniformity was reduced by crystal-collimator misalignment.

• Progress in Medical Physics
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• v.20 no.1
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• pp.7-13
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• 2009

This work is for the preliminary study for the calibration of an $^{192}Ir$ brachytherapy source based on an absorbed dose to water standards. In order to calibrate brachytherapy sources based on absorbed dose to water standards using a clyndirical ionization chamber, the beam quality correction factor $k_{Q,Q_0}$ is needed. In this study $k_{Q,Q_0}s$ were determined by both Monte carlo simulation and semiexperimental methods because of the realistic difficulties to use primary standards to measure an absolute dose at a specified distance. The 5 different serial numbers of the PTW30013 chamber type were selected for this study. While chamber to chamber variations ran up to maximum 4.0% with the generic $k^{gen}_{Q,Q_0}$, the chamber to chamber variations were within a maximum deviation of 0.5% with the individual $k^{ind}_{Q,Q_0}$. The results show why and how important ionization chambers must be calibrated individually for the calibration of $^{192}Ir$ brachytherapy sources based on absorbed dose to water standards. We hope that in the near future users will be able to calibrate the brachytherapy sources in terms of an absorbed dose to water, the quantity of interest in the treatment, instead of an air kerma strength just as the calibration in the high energy photon and electron beam.

• Journal of Radiation Protection and Research
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• v.24 no.4
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• pp.205-213
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• 1999

Effective dose conversion coefficients from unit activity radionuclides contaminated on the ground surface were calculated by using MCNP4A rode and male/female anthropomorphic phantoms. The simulation calculations were made for 19 energy points in the range of 40 keV to 10 MeV. The effective doses E resulting from unit source intensity for different energy were compared to the effective dose equivalent $H_E$ of previous studies. Our E values are lower by 30% at low energy than the $H_E$ values given in the Federal Guidance Report of USEPA. The effective dose response functions derived by polynomial fitting of the energy-effective dose relationship are as follows: $f({\varepsilon})[fSv\;m^2]=\;0.0634\;+\;0.727{\varepsilon}-0.0520{\varepsilon}^2+0.00247{\varepsilon}^3,\;where\;{\varepsilon}$ is the gamma energy in MeV. Using the response function and the radionuclide decay data given in ICRP 38, the effective dose conversion coefficients for unit activity contamination on the ground surface were calculated with addition of the skin dose contribution of beta particles determined by use of the DOSEFACTOR code. The conversion coefficients for 90 important radionuclides were evaluated and tabulated. Comparison with the existing data showed that a significant underestimates could be resulted when the old conversion coefficients were used, especially for the nuclides emitting low energy photons or high energy beta particles.

• Progress in Medical Physics
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• v.21 no.4
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• pp.311-322
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• 2010

In proton therapy, the analysis of secondary particles is important due to delivered dose outside the target volume and thus increased potential risk for the development of secondary cancer. The purpose of this study is to analyze the influence of secondary particles from proton beams on fluence and energy deposition in the presence of inhomogeneous material by using Geant4 simulation toolkit. The inhomogeneity was modeled with the condition that the adipose tissue, bone and lung equivalent slab with thickness of 2 cm were inserted at 30% (Plateau region) and 80% (Bragg peak region) dose points of maximum dose in Bragg curve. The energy of proton was varied with 100, 130, 160 and 190 MeV for energy dependency. The results for secondary particles were presented for the fluence and deposited energy of secondary particles at inhomogeneous condition. Our study demonstrates that the fluence of secondary particles is neither influenced insertion of inhomogeneties nor the energy of initial proton, while there is a little effect by material density. The deposited energy of secondary particles has a difference in the position placed inhomogeneous materials. In the Plateau region, deposited energy of secondary particles mostly depends on the density of inserted materials. Deposited energy in the Bragg region, in otherwise, is influenced by both density of inserted material and initial energy of proton beams. Our results suggest a possibility of prediction about the distribution of secondary particles within complex heterogeneity.

• Progress in Medical Physics
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• v.15 no.4
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• pp.215-219
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• 2004

Due to their excellence for the high-energy therapy range of photon beams, researchers show increasing interest in applying MOSFET dosimeters to low- and medium-energy applications. In this energy range, however, MOSFET dosimeter is complicated by the fact that the interaction probability of photons shows significant dependence on the atomic number, Z, due to photoelectric effect. The objective of this study is to develop a very detailed 3-dimensional Monte Carlo simulation model of a MOSFET dosimeter for radiological characterizations and calibrations. The sensitive volume of the High-Sensitivity MOSFET dosimeter is very thin (1 ${\mu}{\textrm}{m}$) and the standard MCNP tallies do not accurately determine absorbed dose to the sensitive volume. Therefore, we need to score the energy deposition directly from electrons. The developed model was then used to study various radiological characteristics of the MOSFET dosimeter. the energy dependence was quantified for the energy range 15 keV to 6 MeV; finding maximum dependence of 6.6 at about 40 keV. A commercial computer code, Sabrina, was used to read the particle track information from an MCNP simulation and count the tracks of simulated electrons. The MOSFET dosimeter estimated the calibration factor by 1.16 when the dosimeter was at 15 cm depth in tissue phantom for 662 keV incident photons. Our results showed that the MOSFET dosimeter estimated by 1.11 for 1.25 MeV photons for the same condition.

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

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.199-205
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• 2014

In the field of neutron detections, $^3He$ gas, the so-called "the gold standard," is the most widely used material for neutron detections because of its high efficiency in neutron capturing. However, from variable causes since early 2009, $^3He$ is being depleted, which has maintained an upward pressure on its cost. For this reason, the demands for $^3He$ replacements are rising sharply. Research into neutron converting materials, which has not been used well due to a neutron detection efficiency lower than the efficiency of $^3He$, although it can be chosen for use in a neutron detector, has been highlighted again. $^{10}B$, which is one of the $^3He$ replacements, such as $BF_3$, $^6Li$, $^{10}B$, $Gd_2O_2S$, is being researched by various detector development groups owing to a number of advantages such as easy gamma-ray discrimination, non-toxicity, low cost, etc. One of the possible techniques for the detection is an indirect neutron detection method measuring secondary radiation generated by interactions between neutrons and $^{10}B$. Because of the mean free path of alpha particle from interactions that are very short in a solid material, the thickness of $^{10}B$ should be thin. Therefore, to increase the neutron detection efficiency, it is important to make a $^{10}B$ thin film. In this study, we fabricated a $^{10}B$ thin film that is about 60 um in thickness for neutron detection using well-known technology for the manufacturing of a thin electrode for use in lithium ion batteries. In addition, by performing simple physical tests on the conductivity, dispersion, adhesion, and flexibility, we confirmed that the physical characteristics of the fabricated $^{10}B$ thin film are good. Using the fabricated $^{10}B$ thin film, we made a proportional counter for neutron monitoring and measured the neutron pulse height spectrum at a neutron facility at KAERI. Furthermore, we calculated using the Monte Carlo simulation the change of neutron detection efficiency according to the number of thin film layers. In conclusion, we suggest a fabrication method of a $^{10}B$ thin film using the technology used in making a thin electrode of lithium ion batteries and made the $^{10}B$ thin film for neutron detection using suggested method.