• Nuclear Engineering and Technology
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• 제53권1호
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2625-2634
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• 2022

Discretization errors are extremely challenging conundrums of discrete ordinates calculations for radiation transport problems with void regions. In previous work, we have presented a multi-collision source method (MCS) to overcome discretization errors, but the efficiency needs to be improved. This paper proposes a goal-oriented algorithm for the MCS method to adaptively determine the partitioning of the geometry and dynamically change the angular quadrature in remaining iterations. The importance factor based on the adjoint transport calculation obtains the response function to get a problem-dependent, goal-oriented spatial decomposition. The difference in the scalar fluxes from one high-order quadrature set to a lower one provides the error estimation as a driving force behind the dynamic quadrature. The goal-oriented algorithm allows optimizing by using ray-tracing technology or high-order quadrature sets in the first few iterations and arranging the integration order of the remaining iterations from high to low. The algorithm has been implemented in the 3D transport code ARES and was tested on the Kobayashi benchmarks. The numerical results show a reduction in computation time on these problems for the same desired level of accuracy as compared to the standard ARES code, and it has clear advantages over the traditional MCS method in solving radiation transport problems with reflective boundary conditions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• 한국태양에너지학회 논문집
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• 제40권4호
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• pp.23-33
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• 2020

The loss in photovoltaic power due to hailstorms has been highlighted as a major issue in the sustained growth of the PV power plant industry. This study investigates the safety of a solar module by conducting a numerical analysis of a hail test according to the IEC 61215 standard. Our study aims to elucidate the detailed behavior between the ice and solar modules and the micro-cracks forming on solar modules during hailstorms. To analyze the impact of hail, we used the ANSYS AUTODYN software to evaluate the impact characteristics on a solar module with different front glass thicknesses. The simulations show that a solar module with a glass thickness of 4.0 mm results in excellent durability against hail. The results indicate the feasibility of using simulations to analyze and predict micro-cracks on solar modules tailored to various conditions, which can be used to develop new solar modules.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.484-488
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• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.

• 한국안전학회지
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• 제16권4호
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• pp.168-174
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• 2001

For human error analysis, the structure and situation of a task should be analyzed in advance. The paper introduces Structured Information Analysis (SIA) as a task analysis method for error analysis, and delineates the result of application on the emergency procedure of Korean Standard Nuclear Plants (KSNPs). From the task analysis about emergency procedure of KSNP, total 72 specific task goals were identified in the level of system function, and 86 generic tasks were classified from the viewpoint of physical sameness of the task description. Human errors are dependent on task types so that the result of task analysis would be used as a basis for the error analysis on the emergency tasks in nuclear power plants.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.244-250
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• 1996

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1192-1205
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• 2016

Field programmable gate array (FPGA)-based systems are thought to be a practical option to replace certain obsolete instrumentation and control systems in nuclear power plants. An FPGA is a type of integrated circuit, which is programmed after being manufactured. FPGAs have some advantages over other electronic technologies, such as analog circuits, microprocessors, and Programmable Logic Controllers (PLCs), for nuclear instrumentation and control, and safety system applications. However, safety-related issues for FPGA-based systems remain to be verified. Owing to this, modeling FPGA-based systems for safety assessment has now become an important point of research. One potential methodology is the dynamic flowgraph methodology (DFM). It has been used for modeling software/hardware interactions in modern control systems. In this paper, FPGA logic was analyzed using DFM. Four aspects of FPGAs are investigated: the "IEEE 1164 standard," registers (D flip-flops), configurable logic blocks, and an FPGA-based signal compensator. The ModelSim simulations confirmed that DFM was able to accurately model those four FPGA properties, proving that DFM has the potential to be used in the modeling of FPGA-based systems. Furthermore, advantages of DFM over traditional reliability analysis methods and FPGA simulators are presented, along with a discussion of potential issues with using DFM for FPGA-based system modeling.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2905-2918
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• 2023

The valve assembly used in nuclear power plants is important safety-related equipment. In the new standard, the physical attributes are measured using a valve diagnosis test, which is used in the expansion to other non-tested valves using a quantitative test-basis methodology. With a motor-operated actuator, the state of stem's lubrication is related to physical attributes such as the stem factor and the friction coefficient. This study analyzed the numerical transient of fluid and solid lubrication with a squeeze film effect due to the loading rate on the stem and the stem nut using the experimental data. The differential equation that governs the motion mechanism of the stem and stem nut is established and analyzed. The flow rate, the fluid and the solid contact forces are calculated with the friction coefficient. Finally, we found that a change in the friction coefficient results from a change of the shear force in the solid contact mode during the interchange process between the solid contact mode and the fluid contact mode. The qualitative understanding of the squeeze film effect is expanded quantitatively for forces, thread surface distance, velocity, and acceleration, with consideration of the metal solid contact and fluid contact.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.459-465
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• 2017

The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.