• Journal of the Korean Society of Radiology
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• v.17 no.3
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• pp.473-479
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• 2023

In this study, the dose of photoneutrons generated during radiotherapy of prostate cancer using high energy was measured using a photo-stimulated luminescence dosimeter. In addition, this study was intended to study the probability of side effects occurring in the abdomen. A medical linear accelerator capable of generating 15 MV energy, True Beam STx (Varian Medical Systems, USA) and a radiation treatment planning system (Eclipse, Varian Medical Systems, USA) were used. A human body phantom was installed on the couch of the linear accelerator, and an Albedo Neutron Optical Stimulation Luminescence Neutron Detector (Landauer Inc., IL, USA) was used to measure the photoneutron dose. The photoneutron dose value in the abdomen of VMAT and 3C-CRT was 52.8 mSv, more than twice as high as VMAT compared to 3D-CRT. During radiotherapy of prostate cancer, the probability of causing side effects in the abdomen due to light neutron dose was calculated to be 3.2 per 1,000 for VMAT and 1.4 for 3D-CRT. By studying the abdomen, which has a major side effect that can occur during radiotherapy of prostate cancer, it is expected that it will be used as a meaningful study to study the quality of life and stochastic effect of prostate cancer patients

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.9-14
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• 2009

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

• Nuclear Engineering and Technology
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• v.7 no.4
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• pp.277-284
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• 1975

A result of measurement for the relative dose distribution of neutron gamma mixed radiation field in the biological irradiation facility installed at TRIGA Mark-III reactor is described. The relative dose distributions of neutron-gamma mixed radiation field in the biological exposure room have been experimentally determined using a thermoluminescent dosimeter. Presented herein in graphical forms are the experimental results obtained. It as observed that the region commonly having the characteristics of rather homogeneous horizontal and lateral dose distributions is confined to the area bounded by the two planes horizontally parallel to the beam direction with heights of about 40 cm and 130 cm, respectively, at distances beyond 100 cm from the segmentary surface of the aluminum pool liner projected into the the exposure room, while other areas show a steeper gradient in dosage, especially the places adjacent to the segment of the aluminum pool liner and near the inner po${\gamma}$lion of the concrete walls of the exposure room.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.77-83
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• 1997

Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be $300{\mu}Sv/h$ and the maximum calculated dose rate at 100 cm from the cask is $97{\mu}Sv/h$. These dose rates remain within allowable values specified in related regulations.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.71-79
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• 1993

Form the pure Maxwellian distribution(kT= 1.42MeV), the effects upon calibration factors of encapsulating a $^{252}Cf$ spontaneous fission neutron source were investigated to establish a standard neutron field in the Secondary Standard Dosimetry Laboratory at Korea Atomic Energy Research Institute(KAERI). A Monte Carlo code MCNP was used in simulating the encapsulation SR-Cf-100 and SR-Cf-1273 to be real conditions. The anisotropy(FI) and fluence-to-dose equivalents conversion factors$(H/{\Phi})$ were evaluated and compared with other results. As the results, the FI was determined to be 1.061 at ${\theta}=90^{\circ}$ with ${\pm}0.2%$ statistical error and the $(H/{\Phi})$ was evaluated to be $333.9 [pSv\;cm^2]\;with\;{\pm}0.5%$ statistical error, which is lower by 1.8% than that recommended by the ISO 8529. This means physically that the neutron spectrum of the unmoderated $^{252}Cf$ source in KAERI is a little more softened than that by the ISO.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1998-2004
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• 2019

The Multi-Physics (MP) instrument is one of 20 neutron spectrometers planned in the China Spallation Neutron Source (CSNS). This paper presents a shielding calculation for the MP instrument using Monte Carlo codes MCNPX and FLUKA. First, the neutrons that escape from the CSNS decoupled water moderator and are delivered to the beam line of the MP instrument are calculated to use as the source term of the shielding calculation. Then, to validate the calculation method based on multiple variance reduction techniques, a cross check between MCNPX and FLUKA codes is performed by comparing the calculation results of the dose rate distribution on a simplified beam line model. Finally, a complete geometry model of the MP instrument is set up, and the primary parameters for the shielding design are obtained according to the calculated dose rate map considering different worst-case scenarios.

• Progress in Medical Physics
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• v.32 no.4
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• pp.145-152
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• 2021

Purpose: During the treatments of cancer patients with a linear accelerator (LINAC) using photon beams with energies ≥8 MV, the components inside the LINAC head get activated through the interaction of photonuclear reaction (γ, n) and neutron capture (n, γ). We used spectroscopy and measured the dose rate for the LINAC in operation after the treatment ended. Methods: We performed spectroscopy and dose rate measurements for three units of LINACs with a portable high-purity Germanium (HPGe) detector and a survey meter. The spectra were obtained after the beams were turned off. Spectroscopy was conducted for 3,600 seconds, and the dose rate was measured three times. We identified the radionuclides for each LINAC. Results: According to gamma spectroscopy results, most of the nuclides were short-lived radionuclides with half-lives of 100 days, except for ⁶⁰Co, ⁶⁵Zn, and ¹⁸¹W nuclides. The dose rate for three LINACs obtained immediately in front of the crosshair was in the range of 0.113 to 0.129 µSv/h. The maximum and minimum dose rates measured on weekends were 0.097 µSv/h and 0.092 µSv/h, respectively. Compared with the differences in weekday data, there was no significant difference between the data measured on Saturday and Sunday. Conclusions: Most of the detected radionuclides had half-lives <100 days, and the dose rate decreased rapidly. For equipment that primarily used energies ≤10 MV, when the equipment was transferred after at least 10 minutes after shutting it down, it is expected that there will be little effect on the workers' exposure.

• Applied Chemistry for Engineering
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• v.8 no.3
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• pp.524-532
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• 1997

ffects of radiation on the thermal and mechanical properties of modified epoxy resin and hydrogenated bisphenol-A type epoxy resin based neutron shielding materials to be used for radioactive material shipping and storage casks have been investigated. The onset temperatures of the shielding materials of KNS(Kaeri Neutron Shield)-201 and KNS-302 increased with the radiation dose, but those of KNS-202 and KNS-301 decreased at radiation dose above 0.5 MGy. In addition, the radiation dose rarely affected the change of weight of shielding materials with the variation in temperature. At radiation dose up to 0.1 MGy, thermal conductivities of shielding materials were not affected. The thermal expansion coefficients of the shielding materials of KNS-301 and 302 were affected to a less extent than those of KNS-201 and 202 by radiation. At radiation dose up to 0.1 MGy, the tensile strength, compressive strength and flexural strength of the shielding materials of KNS-202 and KNS-301 and 302 increased with the radiation dose. In contrast, those of KNS-201 decreased with an increase in the radiation dose. In addition, the amount of radiation dose on the shielding materials did not result in a measurable loss of specific gravity, weight and hydrogen content.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.34-48
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• 1982

The prime purpose of this study is to realize an index quantity, absorbed dose index, defined by the ICRU for the characterization of ambient radiation level at any location for the purpose of radiation protection. The experiment has been designed to be carried out in two phases, namely, preliminary and main experiment. In the primary study a 30cm diameter sphere of polyethylene was used, while in the main experiment that of tissue equivalent material was fabricated and used. Both experiments were performed in the gamma-ray fields of $^{137}Cs\;and\;^{60}Co$, and in a neutron beam of thermal column of the TRIGA MARK-II research reactor. In the measurement of gamma-ray absorbed dose TLD-700 $(^{7}LiF)$ chips were used, and for the neutron dose both Au activation foils and TLD chips (TLD-600 $(^{6}LiF)$ and TLD-700 for the discrimination of gamma-ray contribution) were used. Theoretical assessment of the absorbed dose in the sphere phantom has been carried out in accordance with the Ehrlich's idea that deduced on the basis of Burlin's cavity theory in the case of gamma-ray irradiation. For the analysis of neutron dose fluence-KERMA rate conversion method was used. The explanation on the dose assessment is given in detail. Results obtained were numerically and statistically analyzed and the depth dose distributions are presented in the graphic forms with normalized values. In the concluding remarks, the possibility and difficulty of realizing the index quantity, including questions and problems to be solved are mentioned.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.