• Progress in Medical Physics
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• v.27 no.4
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• pp.232-235
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• 2016

The purpose of this study is to develop the treatment planning system (TPS) based on Monte-Carlo simulation for BNCT. In this paper, we will propose a method for dose estimation by Monte-Carlo simulation using the CT image, and will evaluate the accuracy of dose estimation of this TPS. The complicated geometry like a human body allows defining using the lattice function in MCNPX. The results of simulation such as flux or energy deposition averaged over a cell, can be obtained using the features of the tally provided by MCNPX. To assess the dose distribution and therapeutic effect, dose distribution was displayed on the CT image, and dose volume histogram (DVH) was employed in our developed system. The therapeutic effect can be efficiently evaluated by these evaluation tool. Our developed TPS could be effectively performed creating the voxel model from CT image, the estimation of each dose component, and evaluation of the BNCT plan.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1406-1416
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• 2019

In this research, a multi-slit prompt-gamma camera was developed to locate the distal dose falloff of the proton beam spots in spot scanning proton therapy. To see the performance of the developed camera, therapeutic proton beams were delivered to a solid plate phantom and then the prompt gammas from the phantom were measured using the camera. Our results show that the camera locates the 90% distal dose falloff (= d90%), within about 2-3 mm of error for the spots which are composed $3.8{\times}10^8$ protons or more. The measured location of d90% is not very sensitive to the irradiation depth of the proton beam (i.e., the depth of proton beam from the phantom surface toward which the camera is located). Considering the number of protons per spot for the most distal spots in typical treatment cases (i.e., 2 Gy dose divided in 2 fields), the camera can locate d90% only for a fraction of the spots depending on the treatment cases. However, the information of those spots is still valuable in that, in the multi-slit prompt-gamma camera, the distal dose falloff of the spots is located solely based on prompt gamma measurement, i.e., not referring to Monte Carlo simulation.

• Journal of Radiation Protection and Research
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• v.12 no.2
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• pp.1-8
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• 1987

The relative biological effectiveness has been measured for the biological characterization of $p^+(50.5\;MeV)$ Be neutron of KCCH-Cyclotron prior to clinical application. Measured RBE of mouse jejunal crypt cell in single whole body irradiation was 2.8. This RBE value is changed differently in different biologic systems such as mouse jejunal crypt cells, intestine and bone marrow in different irradiation method, so that in fractionated irradiation RBE is variable to the different fraction size and total dose, and also variable to the number of fractions.

• Journal of radiological science and technology
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• v.43 no.5
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.275-283
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• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.114-122
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• 2024

Irradiation-induced defects and the precipitates in the wrapper material of the Indian Fast Breeder Test Reactor (FBTR), SS 316 are analyzed using the synchrotron source-based Angle Dispersive X-Ray Diffraction (ADXRD) technique with X-rays of energy 17.185 keV (wavelength ~0.72146 Å). The differences and similarities in the high displacement damage samples as a function of dpa (displacement per atom) and dpa rate in the range of 2.9 × 10^-7- 9 × 10^-7 dpa/s are studied. Ferrite and M₂₃C₆ are commonly observed in the present set of high displacement damage 40-74 dpa SS 316 samples irradiated at temperatures in the range of 400-483 ℃. Also, the dislocation density has increased as a function of the irradiation dose. The X-ray diffraction peak profile parameters quantified such as peak shift and asymmetry show that the irradiation-induced defects are sensitive to the dpa rate-irradiation temperature combinations. The increase in yield strength as a function of displacement damage is also correlated to the dislocation density.

• Korean Journal of Veterinary Research
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• v.41 no.4
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• pp.571-578
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• 2001

To evaluate if the apoptotic fragment assay could be used to estimate the dose prediction after radiation exposure, we examined apoptotic mouse crypt cells per 1,000 cells after whole body $^{60}Co$ $\gamma$-rays and 50MeV ($p{\rightarrow}Be^+$) cyclotron fast neutron irradiation in the range of 0.25 to 1 Gy, respectively. The incidence of apoptotic cell death rose steeply at very low doses up to 1 Gy, and radiation at all doses tigger rapid changes in crypt cells in stem cell region. These data suggest that apoptosis may play an important role in homeostasis of damaged radiosensitive target organ by removing damaged cells. The curve of dose-effect relationship for the data of apoptotic fragments was obtained by the linear-quadratic model $y=0.18+(9.728{\pm}0.887)D+(-4.727{\pm}1.033)D^2$ ($r^2=0.984$) after $\gamma$-rays irradiation, while $y=0.18+(5.125{\pm}0.601)D+(-2.652{\pm}0.7000)D^2$ ($r^2=0.970$) after neutrons in mice. The dose-response curves were linear-quadratic, and a significant dose-response relationship was found between the frequency of apoptotic cell and dose. These data show a trend towards increase of the numbers of apoptotic crypt cells with increasing dose. Both the time course and the radiation dose-response curve for high and low linear energy transfer (LET) radiation modalities were similar. The relative biological effectiveness (RBE) value for crypt cells was 2.072. In addition, there were significant peaks on apoptosis induction at 4 and 6h after irradiation, and the morpholoigcal findings of the irradiated groups were typical apoptotic fragments in crypt cells that were hardly observed in the control group. Thus, apoptosis in crypt cells could be a useful in vivo model for studying radio-protective drug sensitivity or screening test, microdosimetric indicator and radiation-induced target organ injury. Since the apoptotic fragment assay is simple, rapid and reproducible in the range of 0.25 to 1 Gy, it will also be a good tool for evaluating the dose response of radiation-induced organ damage in vivo and provide a potentially valuable biodosimetry for the early dose prediction after accidental exposure.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.584-589
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• 2003

The hot cell facility which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations performed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}$, $2.97{\times}10^{-2}$ and $1.01{\times}10^{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}$, $2.99{\times}10^{-3}$ and $7.88{\times}10^{-2}$ mSv/h, respectively The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources, and penetration and toboggan will be partly reinforced by lead shield.