• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.756-768
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• 2023

Neutron transport calculations are extremely challenging due to the high computational cost of large and complex problems. A multilevel octree grid algorithm (MLTG) of discrete ordinates method was developed to improve the modeling accuracy and simulation efficiency on 3-D Cartesian grids. The Balakovo-3 VVER-1000 neutron dosimetry benchmark is calculated to verify and validate this numerical technique. A simplified S₂ synthetic acceleration is used in the MLTG calculation method to improve the convergence of the source iterations. For the triangularly arranged fuel pins, we adopt a source projection algorithm to generate pin-by-pin source distributions of hexagonal assemblies. MLTG provides accurate geometric modeling and flexible fixed source description at a lower cost than traditional Cartesian grids. The total number of meshes is reduced to 1.9 million from the initial 9.5 million for the Balakovo-3 model. The numerical comparisons show that the MLTG results are in satisfactory agreement with the conventional S_N method and experimental data, within the root-mean-square errors of about 4% and 10%, respectively. Compared to uniform fine meshing, approximately 70% of the computational cost can be saved using the MLTG algorithm for the Balakovo-3 computational model.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• Journal of IKEEE
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• v.26 no.1
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• pp.129-132
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• 2022

In this paper, a technique for optimizing the sub-array system structure that can minimize the side lobe level of the phased-array antenna is proposed. Optimization of the proposed array antenna structure is to adjust the spacing between sub-arrays and sub-arrays by using a hexagonal array structure of one sub-array and a hexagonal sub-array for six hexagonal arrays, and thus the entire phased array antenna system of the radiation pattern was optimized. Compared to the 2-dimensional planar antenna system, the proposed technique maintains a gain of 24.3 dBi and a half-power beam-width of 8.46 degrees without change, and only reduces -3.4 dB and -6.5 dB in the x-axis and y-axis directions, respectively.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2147-2157
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• 2023

The aggregate-forming minerals in concrete undergo volume swelling and microstructure change under neutron irradiation, leading to degradation of physical and mechanical properties of the aggregates and concrete. A comprehensive investigation of volume change and elastic property variation of major aggregate-forming minerals is still lacking, so molecular dynamics simulations have been employed in this paper to improve the understanding of the degradation mechanisms. The results demonstrated that the densities of the selected aggregate-forming minerals of similar atomic structure and chemical composition vary in a similar trend with deposited energy due to the similar amorphization mechanism. The elastic tensors of all silicate minerals are almost isotropic after saturated irradiation, while those of irradiated carbonate minerals remain anisotropic. Moreover, the elastic modulus ratio versus density ratio of irradiated minerals is roughly following the density-modulus scaling relationship. These findings could further provide basis for predicting the volume and elastic properties of irradiated concrete aggregates in nuclear facilities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.303-313
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• 2023

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

• Journal of IKEEE
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• v.28 no.1
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• pp.20-25
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• 2024

This paper presents a new dualband shared-aperture microstrip antenna to operate in the S-Band of 2 GHz and X-Band of 8 GHz, for a Low Earth Orbit satellite antenna system. The proposed antenna incorporates two types of patches those are a rectangular loop-shaped for the S-Band and a square patch for the X-Band. Each patch are optimized for its respective operating band with minimal interference. The proposed antenna achieves a bandwidth of 16 MHz in the S-Band and 572 MHz in the X-Band. The highest gain is measured 7.14 dBi at 1.99 GHz and 7.95 dBi at 7.88 GHz. The proposed antenna exhibits half power beamwidths of 85 degree and 80 degree at 1.99 GHz and 7.88 GHz, respectively. The proposed dualband shared-aperture microstrip antenna may be a good candidate for as a feeding system of a dualband reflectarray antenna With its unidirectional radiation pattern from excellent agreement between simulation and measurement results.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3097-3111
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• 2024

Since the level 2 PSA of OPR-1000 was the requirement for regulatory purposes, Cs-137 release estimation was contained as the Nuclear Safety Act of ROK in which the Cs-137 release frequency exceeding 100 TBq was determined to happen less than 1.0E-6 per year after the Fukushima Daiichi Accident. However, Cs-137 release estimation from the conventional level 2 PSA of OPR-1000 provided uncertainty due to dominant accident sequence consideration. Thus, this study aimed to develop systematic methods through the overall framework to quantify realistic uncertainty concerns of radioactive material release using sensitivity and uncertainty analysis methods and apply them to OPR-1000. This framework helped to quantify confidential value for the Cs-137 release under the BEPU approach using both parametric and non-parametric methods to cover both realistic and conservative points. Uncertainty propagation analysis showed the unexpected uncertainty increase of Cs-137 release exceeding 100 TBq. The non-parametric uncertainty analysis provided higher conservative concerns for safety than the realistic concerns in terms of economics when compared with the parametric uncertainty analysis. Wilks' uncertainty analysis showed the importance to consider conservative Cs-137 release in order to reach the higher safety need. Sensitivity analysis showed reasonable relationships between engineering safety parameters with the Cs-137 release.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3236-3241
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• 2024

Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.