• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.167-174
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• 2023

Background: A nondestructive test is commonly used to inspect the surface defects and internal structure of an object without any physical damage. X-rays generated from an electron accelerator or a tube are one of the methods used for nondestructive testing. The high penetration of X-rays through materials with low atomic numbers makes it difficult to discriminate between these materials using X-ray imaging. The interaction characteristics of neutrons with materials can supplement the limitations of X-ray imaging in material discrimination. Materials and Methods: The radiation image acquisition process for air-cargo security inspection equipment using X-rays and neutrons was simulated using a GEometry ANd Tracking (Geant4) simulation toolkit. Radiation images of phantoms composed of 13 materials were obtained, and the R-value, representing the attenuation ratio of neutrons and gamma rays in a material, was calculated from these images. Results and Discussion: The R-values were calculated from the simulated X-ray and neutron images for each phantom and compared with those obtained in the experiments. The R-values obtained from the experiments were higher than those obtained from the simulations. The difference can be due to the following two causes. The first reason is that there are various facilities or equipment in the experimental environment that scatter neutrons, unlike the simulation. The other is the difference in the neutron signal processing. In the simulation, the neutron signal is the sum of the number of neutrons entering the detector. However, in the experiment, the neutron signal was obtained by superimposing the intensities of the neutron signals. Neutron detectors also detect gamma rays, and the neutron signal cannot be clearly distinguished in the process of separating the two types of radiation. Despite these differences, the two results showed similar trends and the viability of using simulation-based radiation images, particularly in the field of security screening. With further research, the simulation-based radiation images can replace ones from experiments and be used in the related fields. Conclusion: The Korea Atomic Energy Research Institute has developed air-cargo security inspection equipment using neutrons and X-rays. Using this equipment, radiation images and R-values for various materials were obtained. The equipment was reconstructed, and the R-values were obtained for 13 materials using the Geant4 simulation toolkit. The R-values calculated by experiment and simulation show similar trends. Therefore, we confirmed the feasibility of using the simulation-based radiation image.

• 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.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.253-264
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• 1988

Our experience of neutron noise analysis in French-type 900 MWe pressurized water reactor (PWR) is presented. Neutron noise analysis is based on the technique of interpreting the signal fluctuations of ex-core detectors caused by core reactivity changes and neutron attenuation due to lateral core motion. It also provides advantages over deterministic dynamic-testing techniques because existing plant instrumentation can be utilized and normal operation of the plant is not disturbed. The data of this paper were obtained in the ULJIN unit 1 reactor during the start-up test period and the statistical descriptors, useful for our purpose, are power spectral density (PSD), coherence function (CF), and phase difference between detectors. It is found that core support barrel (CSB) motions induced by coolant flow forces and pressure pulsations in a reactor vessel were indentified around 8 Hz of frequency.

• Composites Research
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• v.36 no.6
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• pp.435-440
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• 2023

This study aimed to enhance the thermal conductivity of B₄C/Al composite materials, commonly used in transport/storage containers for spent nuclear fuel, by incorporating both boron carbide (B₄C) and cubic boron nitride(cBN) as reinforcing agents in an aluminum (Al) matrix. The composite materials were successfully manufactured through a stir casting process and practical neutron-absorbing materials were obtained by rolling the fabricated composite ingot. The evaluation of the thermal conductivity of the fabricated composites was carried out because thermal conductivity is critical for neutron absorbing materials. The thermal conductivity measurement results indicated an approximately 3% increase in thermal conductivity under the same volume fraction when compared to composite materials using only B₄C particles. Through neutron absorption cross-sectional area calculations, it was confirmed that the neutron absorption capability decreased to a negligible level. Based on the findings of this study, new design approaches for neutron absorption materials are proposed, contributing to the development of high-performance transport/storage containers.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.442-449
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• 2010

Recently, the neutron irradiation for large diameter silicon (Si)-ingots of more than 8" diameter is requested to satisfy the demand for the neutron transmutation doping silicon (NTD-Si). By increasing the Si-ingot diameter, the radial non-uniformity becomes larger due to the neutron attenuation effect, which results in a limit of the feasible diameter of the Si-ingot. The current evaluation method has a certain limit to precisely evaluate the radial uniformity of Si-ingot because the current evaluation method does not consider the effect of the Si-ingot diameter on the radial uniformity. The objective of this study is to propose a new evaluation method of radial uniformity by improving the conventional evaluation approach. To precisely predict the radial uniformity of a Si-ingot with large diameter, numerical verification is conducted through comparison with the measured data and introducing the new evaluation method. A new concept of a gradient is introduced as an alternative approach of radial uniformity evaluation instead of the radial resistivity gradient (RRG) interpretation. Using the new concept of gradient, the normalized reaction rate gradient (NRG) and the surface normalized reaction rate gradient (SNRG) are described. By introducing NRG, the radial uniformity can be evaluated with one certain standard regardless of the ingot diameter and irradiation condition. Furthermore, by introducing SNRG, the uniformity on the Si-ingot surface, which is ignored by RRG and NRG, can be evaluated successfully. Finally, the radial uniformity flattening methods are installed by the stainless steel thermal neutron filter and additional Si-pipe to reduce SNRG.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1323-1330
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• 2021

This work reported the radiation shielding characteristic of the bismuth titanium vanadium sodium tellurite glass system. The density of the specially-developed glass samples was increased from 2.21 to 4.01 g/cm³ with the addition of Bi₂O₃, despite the fact the molar volume is decease within 85.43-54.79 cm³/mol. The WinXcom program was used to approximate the effect of Bi₂O₃ on the gamma radiation shielding parameters of bismuth titanium vanadium sodium tellurite glasses. The ㎛ values decrease with the increase of Bi₂O₃ concentration. The computed data shows that the glass sample with 20 mol.% of Bi₂O₃ content has the greatest radiation attenuation performance in comparison to other selected glasses. The Bi₂O₃-TiO₂-V₂O₅-Na₂O-TeO₂ glass system shows excellent neutron shielding material with high long-term light transmittance and discharge resistance and could be potentially used as transparent radiation-resistant shielding glass applications.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.949-959
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• 2021

The current study aims to explore the shielding properties of multi-component borate-based glass series. Seven glass-samples with composition of (80-y)H₃BO₃-10ZnO-10Na₂O-yBaO where (y = 0, 5, 10, 15, 20, 25 and 30 mol.%) were synthesized by melt-quench method. Various shielding features for photons, neutrons, and protons were determined for all prepared samples. XCOM, Phy-X program, and SRIM code were performed to determine and explain several shielding properties such as equivalent atomic number, exposure build-up factor, specific gamma-ray constants, effective removal cross-section (Σ_R), neutron scattering and absorption, Mass Stopping Power (MSP) and projected range. The energy ranges for photons and protons were 0.015-15 MeV and 0.01-10 MeV, respectively. The mass attenuation coefficient (μ/ρ) was also determined experimentally by utilizing two radioactive sources (¹⁶⁶Ho and ¹³⁷Cs). Consistent results were obtained between experimental and XCOM values in determining μ/ρ of the new glasses. The addition of BaO to the glass matrix led to enhance the μ/ρ and specific gamma-ray constants of glasses. Whereas the remarkable reductions in Σ_R, MSP, and projected range values were reported with increasing BaO concentrations. The acquired results nominate the use of these glasses in different radiation shielding purposes.

• Journal of radiological science and technology
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• v.26 no.3
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• pp.51-57
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• 2003

Mitigation of fast neutron irradiation damage on reactor vessel and improvement of mechanical integrity are desired for the successful plant life-time extension. In this study, the performance of metallic hydride for this application is reviewed and compared. First, selected prospective metallic hydrides are evaluated by MCNP code and put into the attenuation test using Cf-252 neutron source. Since for the reactor application high moderation and reflection with no absorption are favored, Z factor is introduced for the evaluation. According to the Z value estimation $ZrD_2$ and $TiD_2$ are turned out to be the most favorable fast neutron shielding materials. More thorough evaluation by computer simulation and experimentally, will be followed.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.119-125
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• 2023

Burnup of spent nuclear fuel should be determined accurately for the safety storage of spent nuclear fuel. In this study, a gamma detection system was developed as a part of basic research to measure the burnup of spent nuclear fuel, and its performance was evaluated using a calibration source. The prototype of the gamma detection system was based on a semiconductor sensor using a CZT (Cadmium Zinc Telluride). For quantitative evaluation, tests were conducted using ¹³⁷Cs, ¹³⁴Cs and ²⁵²Cf calibration source. In the performance evaluation, Its field applicability was verified by assessing the energy resolution, the detection linearity and the shielding attenuation according to the nuclide.