• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.293-302
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• 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
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• v.51 no.8
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• pp.2018-2025
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• 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
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• v.53 no.8
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• pp.2767-2773
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• 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
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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.1
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• pp.21-29
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• 2014

A new cold neutron triple-axis spectrometer (Cold-TAS) was recently constructed at the 30 MWth research reactor, HANARO. The spectrometer, which is composed of neutron optical components and radiation shield, required a redesign of the segmented monochromator shield due to the lack of adequate support of its weight. To shed some weight, lowering the height of the segmented shield was suggested while adding more radiation shield to the top cover of the monochromator chamber. To investigate the radiological effect of such change, we performed MCNPX simulations of a few different configurations of the Cold-TAS monochromator shield and obtained neutron and photon intensities at 5 reference points just outside the shield. Reducing the 35% of the height of the segmented shield and locating lead 10 cm from the bottom of the top cover made of polyethylene was shown to perform just as well as the original configuration as radiation shield excepting gamma flux at two points. Using gamma map by MCNPX, it was checked that is distribution of gamma. Increased flux had direction to the top and it had longer distance from top of segmented shield. However, because of reducing the 35% of the height, height of dissipated gamma was lower than original geometry. Reducing the 35% of the height of the segmented shield and locating lead 10cm from the bottom of the top cover was selected. After changing geometry, radiation dose was measured by TLD for confirming tester's safety at any condition. Neutron(0.21 ${\mu}Svhr^{-1}$) and gamma(3.69 ${\mu}Svhr^{-1}$) radiation dose were satisfied standard(6.25 ${\mu}Svhr^{-1}$).

• Journal of radiological science and technology
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• v.36 no.2
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• pp.165-173
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• 2013

This study was the estimation of the dose distribution for proton, prompt gamma rays and proton induced neutron particles, in case of exposing the proton beam to polymer gel dosimeter and water phantom. The polymer gel dosimeter was compositeness material of Gelatin, Methacrylic acid, Hydroquinone, Tetrakis and Distilled water. The density of gel dosimeter was $1.04g/cm^3$ which was similar to water. The 72, 116 and 140 MeV proton beams were used in the simulation. Proton beam interacted with the nuclei of the phantom and the nuclei in excited states emitted prompt gamma rays and proton induced neutron particles during the process of de-excitation. The proton particles, prompt gamma rays, proton induced neutron particles were detected by polymer gel dosimeter and water phantom, respectively. The gap of the axis for gel was 2 mm. The Bragg-peak for proton particles in gel dosimeter was similar to water phantom. The dose distribution for proton and prompt gamma rays in gel dosimeter and water phantom was approximately identical in case of 72, 116 and 140 MeV for proton beam. However, in case of proton induced neutron particles for 72, 116 and 140 MeV proton beam, particles were not detected in gel dosimeter, while the Water phantom absorbed neutron particles. Considering the resulting data, gel dosimeter which was developed in the normoxic state attentively detected the dose distribution for proton beam exposure except proton induced neutron particles.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.324-324
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• 1998

• Journal of Radiation Protection and Research
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• v.23 no.1
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• pp.17-23
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• 1998

Dose response relationships for one of biological end-points (gene mutation) in somatic cells of Tradescantia 4430 clones were studied using neutrons coming out of a californium-252 isotopic source. And the relative biological effectiveness (RBE) of neutrons in relation to X-rays in the induction of TSH pink mutations was assessed. Inflorescences were irradiated with X-ray from X-ray generator and neutrons from $^{252}Cf$ source. Irradiated cuttings were incubated with aeration in neutrient solution under the controlled condition. For more than 4 weeks after irradiation cell mutations were scored. Pink mutation frequencies were calculated from the pooled data for the peak interval (days 6 to 13 post-irradiation). Somatic cell mutations in TSH showed linear dose response relationships in the range of neutron doses available for the experiment. The RBE values estimated for neutrons in relation to X-rays were in the range 3.1 to 6.8, which were much lower than normally recognized value.

• Nuclear Engineering and Technology
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• v.4 no.1
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• pp.39-45
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• 1972

In neutron shielding, the scattering effect is equally important as the attenuations in shielding materials. In the present study, the scattered dose equivalent was measured using a Rem counter for water, paraffin, borated paraffin, ordinary and heavy concrete, lead, iron, and tissue equivalent material in three different angles: 45$^{\circ}$, 90$^{\circ}$, and 135$^{\circ}$, respectively. The measurements were performed for the neutron, having the energies of 0.5, 1, 2, 5, and 18 MeV, which are produced from the Van do Graaff accelerator. The scattered dose equivalent ratios were increased with increasing the thickness of scattering materials and saturated at a certain thickness although they were different from one to other materials under study. The ratios were large for lead and iron while they were small for the hydrogen containing materials such as water and paraffin etc.

• The Journal of the Korea Contents Association
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• v.15 no.1
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• pp.331-337
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• 2015

By comparing and evaluating the neutron dose when running cyclotron in fixed type measuring machine and portable type measuring machine, we performed this study to see if there is any difference between the doses in both measuring devices. We performed correction 3 times for the portable type during the experiment. We measured the values 12 times for each correction, so total 36 times by using the fixed type and portable type once every 2 weeks from June 2012 to February 2014. The statistics for the portable type and fixed type for each experiment showed 0.186, 0.511, and 0.057, which are not statistically significant. Therefore, we have come to the conclusion that unless there is deviation based on the correction period of the portable type, correction of fixed type wouldn't be necessary. Also, reduction of social costs may be achieved if we install an alarm meter, which is relatively low priced compared to the high cost neutron measuring machine, when detecting higher level than the interested standard level.