• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.423-429
• /
• 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
• /
• 제54권5호
• /
• pp.1674-1685
• /
• 2022

The ball valve is an important device in the pipeline transportation system of nuclear power plants. Its operational stability and safety directly affect the normal working of nuclear power plants. In this study, the transient numerical simulation of the opening and closing process of a ball valve was conducted on the basis of the flow interruption capability experiment of the ball valve by using the moving mesh method and inlet and outlet variable boundary conditions. The flow rate and pressure difference with time of the opening and closing process of the ball valve were studied. The internal flow characteristics of the ball valve under different relative openings were analyzed in conjunction with the typical back-step flow structure. Results show that the transient numerical results agree well with the experimental results. The internal flow characteristics of the ball valve are similar at the same opening during opening and closing process. At small opening, the spool and outlet channels easily form a back-step flow structure. The disappearance and generation of backflow vortices during opening and closing occur at 85% opening and 75% opening, respectively. With the decrease in opening degree, the difference in vortex core area in the flow channel of the ball valve spool in the opening and closing process gradually appears. The research results provide some reference value for the design and optimization of ball valves.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.84-96
• /
• 2022

In the worst case, a temporary ignition source (also known as transient combustibles) between two electrical panels can damage both panels. Mitigation strategies for electrical panel fires were previously developed using fire modeling and risk analysis. However, since they do not comply with deterministic fire protection requirements, it is necessary to analyze the boundary values at which combustibles may damage targets depending on various factors. In the present study, a sensitivity analysis of input variables related to the damage threshold of two electrical panels was performed for dimensionless geometry using a Fire Dynamics Simulator (FDS). A new methodology using a damage evaluation map was developed to assess the damage of the electrical panel. The input variables were the distance between the electrical panels, the vertical height of the fuel, the size of the fire, the wind speed and the wind direction. The heat flux was determined to increase as the vertical distance between the fuel and the panel decreased, and the largest heat flux was predicted when the vertical separation distance divided by one half flame length was 0.3-0.5. As the distance between the panels increases, the heat flux decreases according to the power law, and damage can be avoided when the distance between the fuel and the panel is twice the length of the panel. When the wind direction is east and south, to avoid damage to the electrical panel the distance must be increased by 1.5 times compared to no wind. The present scale model can be applied to any configuration where combustibles are located between two electrical panels, and can provide useful guidance for the design of redundant electrical panels.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.49-60
• /
• 2022

Motivated by the high-resolution properties of high-order Weighted Essentially Non-Oscillatory (WENO) and flux limiter (FL) for steep-gradient problems and the robust convergence of Jacobian-free Newton-Krylov (JFNK) methods for nonlinear systems, the preconditioned JFNK fully implicit high-order WENO and FL schemes are proposed to solve the transient two-phase two-fluid models. Specially, the second-order fully-implicit BDF2 is used for the temporal operator and then the third-order WENO schemes and various flux limiters can be adopted to discrete the spatial operator. For the sake of the generalization of the finite-difference-based preconditioning acceleration methods and the excellent convergence to solve the complicated and various operational conditions, the random vector instead of the initial condition is skillfully chosen as the solving variables to obtain better sparsity pattern or more positions of non-zero elements in this paper. Finally, the WENO_JFNK and FL_JFNK codes are developed and then the two-phase steep-gradient problem, phase appearance/disappearance problem, U-tube problem and linear advection problem are tested to analyze the convergence, computational cost and efficiency in detailed. Numerical results show that WENO_JFNK and FL_JFNK can significantly reduce numerical diffusion and obtain better solutions than traditional methods. WENO_JFNK gives more stable and accurate solutions than FL_JFNK for the test problems and the proposed finite-difference-based preconditioning acceleration methods based on the random vector can significantly improve the convergence speed and efficiency.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.1-10
• /
• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.374-385
• /
• 2022

In order to investigate the application of 3D base-seismic isolation system in nuclear power plants (NPPs), comprehensive analysis of constitution and design theory for 3-dimensional combined isolation bearing (3D-CIB) was presented and derived. Four different vertical stiffness of 3D-CIB was designed to isolate the nuclear island (NI) building. This paper aimed at investigating the isolation effectiveness of 3D-CIB through modal analysis and dynamic time-history analysis. Numerical results in terms of dynamic response of 3D-CIB, relative displacement response, acceleration and floor response spectra (FRS) of the superstructure were compared to validate the reliability of 3D-CIB in mitigating seismic response. The results showed that 3D-CIB can significantly attenuate the horizontal acceleration response, and a fair amount of the vertical acceleration response reduction of the upper structure was still observed. 3D-CIB plays a significant role in reducing the horizontal and vertical FRS, the vertical FRS basically do not vary with the floor height. The smaller the vertical stiffness of 3D-CIB is, the better the vertical isolation effectiveness is, whereas, it will increase the displacement and the rocking effect of superstructure. Although the advantage of 3D-CIB is that the vertical stiffness can be flexibly adjusted, it should be designed by properly accounting for the balance between the isolation effectiveness and displacement control including rocking effect. The results of this study can provide the technical basis and guidance for the application of 3D-CIB to engineering structure.

• Nuclear Engineering and Technology
• /
• 제54권8호
• /
• pp.2792-2800
• /
• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Nuclear Engineering and Technology
• /
• 제54권6호
• /
• pp.2230-2243
• /
• 2022

The tin tailing processing industry in Malaysia has operated with minimal regard and awareness for material management and working environment safety, impacting the environment and workers in aspects of radiation and heavy metal exposure. RIA was conducted where environmental samples were analyzed, revealing concentrations of ²²⁶Ra, ²³²Th and ⁴⁰K between the range of 0.1-10.0, 0.0-25.7, and 0.1-5.8 Bq/g respectively, resulting in the AED exceeding UNCEAR recommended value and regulation limit enforced by AELB (1 mSv/y). Ra_eq calculated indicates that samples collected pose a significant threat to human health from gamma-ray exposure. Assessment of heavy metal content via pollution indices of soil and sediment showed significant contamination and enrichment from processing activities conducted. As and Fe were two of the highest metals exposed both via soil ingestion with an average of 4.6 × 10^-3 mg/kg-day and 1.4 × 10^-4 mg/kg-day, and dermal contact with an average of 5.6 × 10^-4 mg/kg-day and 6.0 × 10^-4. mg/kg-day respectively. Exposure via accidental ingestion of soil and sediment could potentially cause adverse non-carcinogenic and carcinogenic health effect towards workers in the industry. Correlation analysis indicates the presence of a relationship between the concentration of NORM and trace elements.

• 웰빙융합연구
• /
• 제5권2호
• /
• pp.43-50
• /
• 2022

Purpose: The small-scale environmental impact assessment conducted during the development project stage has focused on the preservation of the natural environment centered on non-urban areas, due to the nature of urbanization, health problems for citizens of high-density urban areas have a limitation in that they are relatively neglected. In the case of strategic environmental impact assessment and environmental impact assessment in urban areas, there is no basis for evaluation in urban areas because there are exceptions to be excluded from the target projects or there are no target project regulations for buildings. Therefore, in this research, we examined the problems with the target project such as the current environmental impact assessment, and tried to establish a system improvement plan that can solve them. Research design, data and methodology: After reviewing the current environmental impact assessment-related laws (including enforcement ordinances) and national land planning laws (including enforcement ordinances), exceptions such as environmental impact assessment in urban areas were identified and problems were identified. Based on this, an amendment to the Enforcement Decree was proposed to provide institutional support for the expansion of target projects such as environmental impact assessment in urban areas. Results and Conclusions: Through this research, it is expected that the projects subject to environmental impact assessment on development projects in urban areas directly related to the health of the people will be expanded, and the net function of the environmental impact assessment system will be maximized.

• Nuclear Engineering and Technology
• /
• 제54권10호
• /
• pp.3909-3918
• /
• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.