• 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.

• 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.

• 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.

• Journal of Radiopharmaceuticals and Molecular Probes
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• v.8 no.2
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• pp.113-118
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• 2022

Gadolinium neutron capture therapy (Gd-NCT) is a precision radiation therapy that kills cancer cells using the neutron capture reaction that occurs when ¹⁵⁷Gd hits thermal neutrons. ¹⁵⁷Gd has the highest thermal neutron capture cross section of 254,000 barns among stable isotopes in the periodic table. Another stable isotope, ¹⁵⁵Gd, also has a high thermal neutron trapping area (~ 60,700 barns), so gadolinium that exists in nature can be used as a Gd-NCT drug. Gd-NCT is a mixed kinetic energy of low-energy and high-energy ionizing particles, which can be uniformly distributed throughout the tumor tissue, thereby solving the disadvantage of heterogeneous dose distribution in tumor tissue. The Gd complexes of small-sized molecule are widely used as contrast agents for magnetic resonance imaging (MRI) in clinical practice. Therefore, these compounds can be used not only for diagnosis but also therapy when considering the concept of Gd-NCT. This multifunctional trial can look forward to new medical advance into NCT clinical practices. In this review, we introduce gadolinium compounds suitable for Gd-NCT and describe the necessity of image guided Gd-NCT.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3591-3598
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• 2023

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF₃ gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the ²⁴¹Am-Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 mSv/h. The results measured by the above methods could provide guidance for linac-related research.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.185-190
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• 2016

Background: In order to measure neutron energy spectra, the conventional Bonner Sphere Spectrometers (BSS) are widely used. In this spectrometer, several measurements with different size Bonner spheres are required. Operators should, therefore, place these spheres in several times to a measurement point where radiation dose might be relatively high. In order to reduce this effort, novel neutron energy spectrometer using an onion-like single Bonner sphere was proposed in our group. This Bonner sphere has multiple sensitive spherical shell layers in the single sphere. In this spectrometer, a band-shaped thermal neutron detection medium, which consists of a LiF-ZnS mixed powder scintillator sheet and a wavelength-shifting (WLS) fiber readout, was looped to each sphere at equal angular intervals. Amount of LiF neutron converter is reduced near polar region, where the band-shaped detectors are concentrated, in order to uniform the directional sensitivity. The LiF-ZnS mixed powder has an advantage of extremely high light yield. However, since it is opaque, scintillation photons cannot be collect uniformly. This type of detector shows no characteristic shape in the pulse height spectrum. Subsequently, it is difficult to set the pulse height discrimination level. This issue causes sensitivity fluctuation due to gain instability of photodetectors and/or electric modules. Materials and Methods: In order to solve this problem, we propose to replace the LiF-ZnS mixed powder into a flexible and Transparent RUbber SheeT type $LiCaAlF_6$ (TRUST LiCAF) scintillator. TRUST LiCAF scintillator can show a peak shape corresponding to neutron absorption events in the pulse height spectrum. Results and Discussion: We fabricated the prototype detector with five sensitive layers using TRUST LiCAF scintillator and conducted basic experiments to evaluate the directional uniformity of the sensitivity. Conclusion: The fabricated detector shows excellent directional uniformity of the neutron sensitivity.