• Journal of the Korean Society of Radiology
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• v.13 no.1
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• pp.73-80
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• 2019

Neutrons are generated by the nuclear reaction, which is absorbed into the concrete wall and causes the activation during cyclotron operation. The purpose of this study is to investigate the effect of neutron activation and radiative concrete on concrete type. This experiment used Monte Carlo simulation and RESRAD model. The results of the experiment showed that the higher the content of Fe in concrete, the greater the shielding rate. The effect of $^{56}Fe(n,\;2np)^{54}Mn$ reaction on workers is also increased. However, radioactive nuclides have low activity and have very low impact on workers. Radioactive concrete should be treated as general wastes with less than its self-disposal tolerance level, and it should be recycled to the surface such as road repair rather than landfill to minimize the effect of $^{14}C$.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2000-2010
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• 2021

Mechanical moduli, such as Young's modulus (E), Bulks modulus (B), Shear modulus (S), longitudinal modulus (L), Poisson's ratio (σ) and micro Hardness (H) were theoretically calculated for (100-x)TeO₂+x MgO glasses, where x = 10, 20, 30, 40 and 45 mol%, based on the Makishima-Mackenzie model. The estimated results showed that the mechanical moduli and the microhardness of the glasses were improved with the increase of the MgO contents in the TM glasses, while Poisson's ratio decreased with the increase in MgO content. Moreover, the radiation shielding capacity was evaluated for the studied TM glasses. Thus, the linear attenuation coefficient (LAC), mass attenuation coefficient (MAC), transmission factor (TF) and half-value thickness (𝚫_0.5) were simulated for gamma photon energies between 0.344 and 1.406 MeV. The simulated results showed that glass TM10 with 10 mol % MgO possess the highest LAC and varied in the range between 0.259 and 0.711 cm^-1, while TM45 glass with 45 mol % MgO possess the lowest LAC and vary in the range between 0.223 and 0.587 cm^-1 at gamma photon energies between 0.344 and 1.406 MeV. Furthermore, the BXCOM program was applied to calculate the effective atomic number (Z_eff), equivalent atomic number (Z_eq) and buildup factors (EBF and EABF) of the glasses. The effective removal cross-section for the fast neutrons (ERCSFN, ∑_R) was also calculated theoretically. The received data depicts that the lowest ∑_R was achieved for TM10 glasses, where ∑_R = 0.0193 cm² g^-1, while TM45 possesses the highest ERCSFN where ∑_R = 0.0215 cm² g^-1.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3080-3087
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• 2023

Thermoelectric (TE) materials working in radioisotope thermoelectric generators are irradiated by neutrons throughout its service; thus, investigating the neutron irradiation stability of TE devices is necessary. Herein, the influence of neutron irradiation with fluences of 4.56 × 10¹⁰ and 1 × 10¹³ n/cm² by pulsed neutron reactor on the electrical and thermal transport properties of n-type Bi₂Te_2.7Se_0.3 and p-type Bi_0.5Sb_1.5Te₃ thermoelectric alloys prepared by cold-pressing and molding is investigated. After neutron irradiation, the properties of thermoelectric materials fluctuate, which is related to the material type and irradiation fluence. Different from p-type thermoelectric materials, neutron irradiation has a positive effect on n-type Bi₂Te_2.7Se_0.3 materials. This result might be due to the increase of carrier mobility and the optimization of electrical conductivity. Afterward, the effects of p-type and n-type TE devices with different treatments on the output performance of TE devices are further discussed. The positive and negative effects caused by irradiation can cancel each other to a certain extent. For TE devices paired with p-type Bi_0.5Sb_1.5Te₃ and n-type Bi₂Te_2.7Se_0.3 thermoelectric legs, the generated power and conversion efficiency are stable after neutron irradiation.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1587-1592
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• 2023

Time-encoded imagers (TEI) are important class of instruments to search for potential radioactive sources to prevent illicit transportation and trafficking of nuclear materials and other radioactive sources. The energy of the radiation cannot be known in advance due to the type and shielding of source is unknown in practice. However, the response function of the time-encoded imagers is related to the energy of neutrons or gamma-rays. An improved image reconstruction method based on MLEM was proposed to correct for the energy induced response difference. In this method, the count vector versus time was first smoothed. Then, the preset response function was adaptively corrected according to the measured counts. Finally, the smoothed count vector and corrected response were used in MLEM to reconstruct the source distribution. A one-dimensional dual-particle time-encode imager was developed and used to verify the improved method through imaging an Am-Be neutron source. The improvement of this method was demonstrated by the image reconstruction results. For gamma-ray and neutron images, the angular resolution improved by 17.2% and 7.0%; the contrast-to-noise ratio improved by 58.7% and 14.9%; the signal-to-noise ratio improved by 36.3% and 11.7%, respectively.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3648-3658
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• 2023

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly containing a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm² of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.92-96
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• 1986

The $^{93}Nb(n,n\alpha)^{89m}Y$, $^{93}Nb(n,{\alpha})^{90m}Y$ and $^{93}Nb(n,2n)^{92m}Nb$ cross sections at a neutron energy of 14.6 MeV have been measured relative to the $^{27}Al(n,p)^{27}Mg$ and $^{27}Al(n,{\alpha})^{24}Na$ cross sections. A small accelerator utilizing $T(D,n)^4He$ reaction was used as a neutron source and the neutron energy spread is about 0.4MeV at the sample. All induced activities were measured with a 70cc HPGe detector in the same geometry.

• Membrane Journal
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• v.34 no.2
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• pp.114-123
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• 2024

Hydrogen isotopes can be categorized into light hydrogen, heavy hydrogen, and tritium based on the number of neutrons, each of which is used in specific fields. Specifically, deuterium is of interest in the electronics industry, nuclear energy industry, analytical technology industry, pharmaceutical industry, and telecommunications industry. Conventional methods such as cold distillation, thermal cycling absorption processes, Girdler sulfide processes, and water electrolysis have their own advantages and disadvantages, leading to the need for alternative technologies with high separation and energy efficiency. In this context, membrane-based hydrogen isotope separation is one of the promising solutions to reduce energy consumption. In this review, we will present the state-of-the-art in hydrogen isotope separation using membranes and their operating principles. The technology for separating hydrogen isotopes using membranes is just beginning to be conceptualized, and many challenges remain to be overcome. However, if achieved, the economic benefits are expected to be significant. We will discuss future research directions for this purpose.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.