• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.178-183
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• 2007

Neutrons are high LET (linear energy transfer) radiation and cause more damage to the target cells than x-rays or gamma rays. The damage from neutrons is generally considered fatal to a cell and neutrons have a greater tendency to cause cell death through direct interaction on DNA. We performed experiments to measure growth delay ratio and oxygen enhancement ratio (OER) in mouse model system. We inoculated EMT-6 cells to the right hind leg of BALB-c mouse and X-rays and neutron beams were given when the average volume of tumors reached $200-300mm^3$. We irradiated 0, 11, 15.4 Gy of X-ray and 0, 5, 7 Gy of fast neutron beam at normoxic and hypoxic condition. The volume of tumors was measured 3 times per week. In x-ray experiment, growth delay ratio was 1.34 with 11 Gy and 1.33 with 15.4 Gy in normoxic condition compared to in hypoxic condition, respectively. In neutron experiment, growth delay ratio was 0.94 with 5 Gy and 0.98 with 7 Gy, respectively. The OER of neutron beam was 0.97. The neutron beam was more effective than X-ray in the control of hypoxic tumors.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.328-332
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• 2016

Background: Boron Neutron Capture Therapy (BNCT) is a new radiation therapy. In BNCT, there exists some very critical problems that should be solved. One of the severest problems is that the treatment effect cannot be known during BNCT in real time. We are now developing a SPECT (single photon emission computed tomography) system (BNCT-SPECT), with a cadmium telluride (CdTe) semiconductor detector. BNCT-SPECT can obtain the BNCT treatment effect by measuring 478 keV gamma-rays emitted from the excited state of $^7Li$ nucleus created by $^{10}B(n,{\alpha})$ $^7Li$ reaction. In the previous studies, we investigated the feasibility of the BNCT-SPECT system. As a result, the S/N ratio did not meet the criterion of S/N > 1 because deterioration of the S/N ratio occurred caused by the influence of Compton scattering especially due to capture gamma-rays of hydrogen. Materials and Methods: We thus produced an arrayed detector with two CdTe crystals to test cross talk phenomenon and to examine an anti-coincidence detection possibility. For more precise analysis for the anti-coincidence detection, we designed and made a collimator having a similar performance to the real BNCT-SPECT. Results and Discussion: We carried out experiments with the collimator to examine the effect of cross talk of scattering gamma-rays between CdTe elements more practically. As a result of measurement the coincidence events were successfully extracted. Conclusion: We are now planning to carry out evaluation of coincidence rate from the measurement and comparison of it with the numerical calculations.

• 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.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.161-163
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• 2002

The BNCT(Boron Neutron Capture Therapy) facility has been developed in Hanaro(High-flux Advanced Neutron Application Reactor), a research reactor of Korea Atomic Energy Research Institute. A typical tangenial beam port is utilized with this BNCT facility. Thermal neutrons can be penetrated within the limits of the possible maximum instead of being filtered fast neutrons and gamma rays as much as possible using the silicon and bismuth single crystals. In addition to, the liquid nitrogen (LN$_2$) is used to cool down the silicon and bismuth single crystals for the increase of the penetrated thermal neutron flux. Neutron beams for BNCT are shielded using the water shutter. The water shutter was designed and manufactured not to interfere with any other subsystem of Hanaro when the BNCT facility is operated. Also, it is replaced with conventional beam port plug in order to cut off helium gas leakage in the beam port. A circular collimator, composed of $\^$6/Li$_2$CO$_3$ and polyethylene compounds, is installed at the irradiation position. The measured neutron flux with 24 MW reactor power using the Au-198 activation analysis method is 8.3${\times}$10$\^$8/ n/cm$^2$ s at the collimator, exit point of neutron beams. Flatness of neutron beams is proven to ${\pm}$ 6.8% at 97 mm collimator. According to the result of acceptance tests of the water shutter, the filling time of water is about 190 seconds and drainage time of it is about 270 seconds. The radiation leakages in the irradiation room are analyzed to near the background level for neutron and 12 mSv/hr in the maximum for gamma by using BF$_3$ proportional counter and GM counter respectively. Therefore, it is verified that the neutron beams from BNCT facility in Hanaro will be enough to utilize for the purpose of clinical and pre-clinical experiment.

• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.219-226
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• 1997

Ultralow background gamma-ray measurement system was installed to measure and analyze gamma-rays emitted from environmental and swipe samples. The background reduction techniques applied on this system are the passive shielding to surround the HPGe detector, an active external anticosmic shield to shield cosmic-rays and the nitrogen gas supply to minimize the introduction of ubiquitous radon decay nuclei. The performance test result showed that the system background at energies between 50 keV and 2 MeV is reduced about $10^{-2}$ order and the MDA is so low as to be suitable for the environmental sample analysis. But it is appeared that the neutron produced by cosmic-ray increases the background at low energy region.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.470-477
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• 2018

With a highly functional methyl vinyl silicone rubber (VMQ) matrix and filler materials of $B_4C$, PbO, and benzophenone (BP) and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were $323.6^{\circ}C$ and $335.3^{\circ}C$, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am-Be neutron source. The transmission of ${\gamma}$-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.178-184
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• 2020

The traditional methods for radiation shield design always only focus on either the structure or the components of the shields rather than both of them at the same time, which largely affects the shielding performance of the facilities, so in this paper, a novel method for designing the structure and components of shields simultaneously is put forward to enhance the shielding ability. The method is developed by using the genetic algorithm (GA) and the MCNP software. In the research, six types of shielding materials with different combinations of elements such as polyethylene (PE), lead (Pb) and Boron compounds are applied to the radiation shield design, and the performance of each material is analyzed and compared. Then two typical materials are selected based on the experiment result of the six samples, which are later verified by the Compact Accelerator Neutron Source (CANS) facility. By using this method, the optimal result can be reached rapidly, and since the design progress is semi-automatic for most procedures are completed by computer, the method saves time and improves accuracy.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.23-33
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• 1982

The characteristic thermoluminescence responses of Teflon thermoluminescent dosimeters to radiations have been studied by the variation of radiation qualities as well as the high dose radiations. The change in the sensitivity of TLDs for different radiation qualities were studied through not only the photon energy dependence but also the change of supralinearity on the photon energy dependence, by exposing $^{60}Co$ gamma rays, the effective X-rays of 44keV, 69keV, 108keV, and thermal neutron of 0.04 eV. The results were as the following: The TL response of $T-CaSO_4$: Dy as a function of absorbed dose was linear up to about 5 Gy, and the response beyond 5Gy was supralinear for $^{60}Co$ gamma rays. The supralinearity of T-LiF-7 became noticeably apparent more than that of $T-CaSO_4$:Dy and also the lower the LET of radiation became the higher the supralinear effects were. No supralinearity appeared for the thermal neutron irradiations equivalent to 10Gy of $^{60}Co$ gamma rays. The relative sensitivities (Rs), which depended on the doses of $^{60}Co$ gamma rays to the TLDs of T-LiF-7 and T-$CaSO_4$:Dy could be, respectively, approximated to the following empirical formula fitted by the least square method: $$R_{LiF}=1.021-0.04581\;logD+0.402(logD)^2-0.405(logD)^3,\;\;5{\times}10^3{\geq}D{\geq}1(Gy)$$ $$R_{CaSO_4}=0.976-0.3241\;logD+0.262(logD)^2-0.298(logD)^3,\;5{\times}10^3{\geq}D{\geq}1(Gy)$$.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2613-2620
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• 2023

Excellent thermal neutron absorption performance of boron expands the potential use of boron rich slag to prepare epoxy resin matrix nuclear shielding composites. However, shielding attenuation behaviors and mechanism of the composites against gamma rays are unclear. Based on the radiation protection theory, Phy-X/PSD, XCOM, and ⁶⁰Co gamma ray source were integrated to obtain the shielding parameters of boron rich slag/epoxy resin composites at 0.015-15 MeV, which include mass attenuation coefficient (µ_t), linear attenuation coefficient (µ), half value thickness layer (HVL), electron density (N_eff), effective atomic number (Z_eff), exposure buildup factor (EBF) and exposure absorption buildup factor (EABF).µ_t, µ, HVL, N_eff, Z_eff, EBF and EABF are 0.02-7 cm²/g, 0.04-17 cm^-1, 0.045-20 cm, 5-14, 3 × 10²³-8 × 10²³ electron/g, 0-2000, and 0-3500. Shielding performance is BS4, BS3, BS3, BS1 in descending order, but worse than ordinary concrete. µ and HVL of BS1-BS4 for ⁶⁰Co gamma ray is 0.095-0.110 cm^-1 and 6.3-7.2 cm. Shielding mechanism is main interactions for attenuation gamma ray by BS1-BS4 are elements with higher content or higher atomic number via Photoelectric Absorption at low energy range, and elements with higher content via Compton Scattering and Pair Production in Nuclear Field at middle and higher energy range.