• Radiation Oncology Journal
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• v.11 no.1
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• pp.35-42
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• 1993

The dose response of the number of micronuclei in cytokinesis-blocked (CB) lymphocytes after in vitro irradiation with $\gamma$-rays and neutrons in the 5 dose ranges was studied for a heterogeneous population of 4 donors. One thousand binucleated cells were systematically scored for micronuclei. Measurements performed after irradiation showed a dose-dependent increase in micronuclei (MN) frequency in each of the donors studied. The dose-response curves were analyzed by a linear-quadratic model, frequencies per 1000 CB cells were ($0.31{\pm}0.049$)D+($0.0022{\pm}0.0002)D^2+(13.19{\pm}1.854) (r^2=1.000,\;X^2=0.7074,\;p=0.95$) following $\gamma$ irradiation, and ($0.99{\pm}0.528$)\;D+(0.0093{\pm}0.0047)\;D^2+(13.31{\pm}7.309)\;(r^2=0.996,\;X^2=7.6834,\;p=0.11) following neutrons irradiation (D is irradiation dose in cGy). The relative biological effectiveness (RBE) of neutrons compared with $\gamma$-rays was estimated by best fitting linear-quadratic model. In the micronuclei frequency between 0.05 and 0.8 per cell, the RBE of neutrons was $2.37{\pm}0.17$. Since the MN assay is simple and rapid, it may be a good tool for evaluating the $\gamma$-ray and neutron response.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2078-2084
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• 2020

Magnesium oxide (MgO) and Zinc oxide (ZnO) nanoparticles (NPs) have been successfully synthesized by solid-solid reaction method. The structural properties of ZnO and MgO NPs were studied using X-ray diffraction (XRD) and scanning electron microscopy (SEM). XRD results indicated a formation of pure MgO and ZnO NPs. The mean diameter values of the agglomerated particles were around to be 70 and 50 nm for MgO and ZnO NPs, respectively using SEM analysis. Further, a wide-range of nuclear radiation shielding investigation for gamma-ray and fast neutrons have been studied for Magnesium oxide (MgO) and Zinc oxide (ZnO) samples. FLUKA and Microshield codes have been employed for the determination of mass attenuation coefficients (μ_m) and transmission factors (TF) of Magnesium oxide (MgO) and Zinc oxide (ZnO) samples. The calculated values for mass attenuation coefficients (μ_m) were utilized to determine other vital shielding properties against gamma-ray radiation. Moreover, the results showed that Zinc oxide (ZnO) nanoparticles with the lowest diameter value as 50 nm had a satisfactory capacity in nuclear radiation shielding.

• Nuclear Engineering and Technology
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• v.3 no.4
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• pp.185-197
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• 1971

Korea's TRIGA Mark-Ⅱ reactor was primarily designed in 1950's and was constructed in 1962 for 100 kw thermal output, but it was upgraded to 250 kw in July 1969. Nevertheless, the shield remains unchanged, although the radiation level has increased. The result of computation On this paper shows that, with the existing shield, it is safe for the fast neutrons even after the power upgrading by 2.5 times. It is, however, somewhat dangerous for the gamma rays which are comprised of primary and secondary. For the analysis of the reactor shielding design, an attempt is made for the computation toward the horizontal direction. From theoretical point of view, it can be concluded that some layer of additional shield must be reinforced to the existing concrete in order to be radiologically safe in the reactor hall.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.323-331
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• 2015

Zirconium alloys are widely used in nuclear reactors as structural materials. During the operation, they are exposed to fast neutrons. Ion irradiation is used to simulate the damage introduced by neutron irradiation. In this article, we briefly review the neutron irradiation damage of zirconium alloys, then summarize the effect of ion irradiation on microstructural evolution, mechanical and corrosion properties, and their relationships. The microstructure components consist of dislocation loops, second phase precipitates, and gas bubbles. The microstructure parameters are also included such as domain size and microstrain determined by X-ray diffraction and the S-parameter determined by positron annihilation. Understanding the relationships of microstructure and properties is necessary for developing new advanced materials with higher irradiation tolerance.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2607-2612
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• 2020

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu-Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section Σ_R (cm^-1) of fast neutrons and total attenuation coefficient μ (cm^-1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section Σ_R (cm^-1) to 0.141 compared to 0.094 of ordinary concrete.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.189-199
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• 2016

Fast neutrons with a broad energy spectrum, with which it is possible to evaluate nuclear data for various research fields such as medical applications and the development of fusion reactors, can be generated by irradiating proton beams on target materials such as beryllium. To generate short-pulse proton beam, we adopted a deflector and slit system. In a simple deflector with slit system, most of the proton beam is blocked by the slit, especially when the beam pulse width is short. Therefore, the available beam current is very low, which results in low neutron flux. In this study, we proposed beam modulation using a buncher cavity to increase the available beam current. The ideal field pattern for the buncher cavity is sawtooth. To make the field pattern similar to a sawtooth waveform, a multiharmonic buncher was adopted. The design process for the multiharmonic buncher includes a beam dynamics calculation and three-dimensional electromagnetic simulation. In addition to the system design for pulsed proton generation, a test bench with a microwave ion source is under preparation to test the performance of the system. The design study results concerning the pulsed proton beam generation and the test bench preparation with some preliminary test results are presented in this paper.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.818-824
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• 2019

In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81-1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the $Z_{eff}$ of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE ($C_2F_4$, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons ($\sum_{R}$) were computed to investigate the neutron attenuation characteristics. It is found that the $\sum_{R}$ values of the eight investigated polymers are close and ranged from $0.07058cm^{-1}$ for PVDC to $0.11510cm^{-1}$ for PA-6.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.942-948
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• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Nuclear Engineering and Technology
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• v.2 no.2
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• pp.85-95
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• 1970

An attempt has been made to do interpretation of the fast neutron dose with two threshold detectors incorporated with the Harwell criticality locket. This method is based on the assumption that the spectral distribution of fission neutrons in criticality accidents may be governed by one spectral parameter. The surface-absorbed dose for a unit fission neutron fluence seems to be insensitive to spectral shifts of the fission neutron spectrum. The average cross-sections for the activation detectors, however, are considerably changed with the neutron spectral shape, which may lead to a large error in calculating the dose from the reaction rate if one uses a fixed value for the average cross sections regardless of the neutron spectral distribution. Besides, the doses calculated from three representative formulae for fission neutron spectra have been compared : these formulae are Watt, Cranberg at al. and Maxwellian forms. The results obtained front the Maxwellian formula show a departure from the Watt and Cranberg's, both being similarly close.