• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2215-2221
• /
• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1901-1910
• /
• 2023

The safety-class (1E) digital control system (DCS) of nuclear power plant characterized structural multiple redundancies, therefore, it is important to quantitatively evaluate the reliability of DCS in different degree of backup loss. In this paper, a reliability evaluation model based on T-S fuzzy fault tree (FT) is proposed for 1E DCS of nuclear power plant, in which the connection relationship between components is described by T-S fuzzy gates. Specifically, an output rejection control system is chosen as an example, based on the T-S fuzzy FT model, the key indicators such as probabilistic importance are calculated, and for a further discussion, the T-S fuzzy FT model is transformed into Bayesian Network(BN) equivalently, and the fault diagnosis based on probabilistic analysis is accomplished. Combined with the analysis of actual objects, the effectiveness of proposed method is proved.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.410-418
• /
• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1791-1801
• /
• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Korean Journal of Digital Imaging in Medicine
• /
• v.15 no.2
• /
• pp.25-30
• /
• 2013

The purpose of this study was to provide information about exhibits for nuclear energy in science museums. This analysed form and content of exhibits of science museums. The subjects were exhibits of 3 science museums; Seoul science park, Seoul national science museum, Gwachen national science museum. The research results were as follows: First, 3 science museums had similar methods of exhibits and types of explanations because of speciality of theme. 3 science museums had mostly fixed exhibit. Panel was the most types of explanations in 3 science museums. Second, 3 science museums had similar contents of exhibits. They dealt with nuclear power generation and radiation. However, Some parts such as radioactive waste, nuclear fusion generation had different. This study suggests that exhibits for nuclear energy of science museums use a variety of methods and types of explanation. Also, science museums need to increase exhibits for nuclear energy.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.942-948
• /
• 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
• /
• v.56 no.6
• /
• pp.2332-2342
• /
• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.121-122
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.125-130
• /
• 2010

Sipping measurements were implemented at the Open Pool Australian Light water reactor (OPAL) to test for failure in reactor fuel elements. Fission product released by the fuel element into the pool water was measured using both High Purity Germanium (HPGe) detection via samples and a NaI(Tl) detection in-situ with the sipping device. Results from two fuel elements are presented.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.05a
• /
• pp.204-205
• /
• 2018