• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.71-79
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• 2023

Air ingress incidents are major safety accidents in very-high-temperature reactors (VHTRs). Air containing a high volume fraction of oxygen may cause severe oxidation of core components at the VHTR, especially for the significantly thin alloy tube wall in the intermediate heat exchanger (IHE). The research objects of this study are Inconel 617 and Incoloy 800H, two candidate alloys for IHE in VHTR. The air ingress accident scenario is simulated with high-temperature air flow at 950 ℃. A continuous oxide scale was formed on the surfaces of both the alloys after the experiment. Because the oxide scale of Inconel 617 has a loose structure, whereas that of Incoloy 800H is denser, Inconel 617 exhibited significantly more severe internal oxidation than Incoloy 800H. Further, Inconel 617 showed a significant decrease in ultimate tensile strength and plasticity after aging for 200 h, whereas Incoloy 800H maintained its tensile properties satisfactorily. Through control experiment under vacuum, we preliminarily concluded that serious internal oxidation is the primary reason for the decline in the tensile properties of Inconel 617.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.292-295
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• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3928-3942
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• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.271-278
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• 2020

The investigation into a full-scale 27 m high, by 6 m wide, thermosyphon loop. The simulation model is based on a one-dimensional axially-symmetrical control volume approach, where the loop is divided into a series of discreet control volumes. The three conservation equations, namely, mass, momentum and energy, were applied to these control volumes and solved with an explicit numerical method. The flow is assumed to be quasi-static, implying that the mass-flow rate changes over time. However, at any instant in time the mass-flow rate is constant around the loop. The boussinesq approximation was invoked, and a reasonable correlation between the experimental and theoretical results was obtained. Experimental results are presented and the flow regimes of the working fluid inside the loop identified. The results indicate that a series of such thermosyphon loops can be used as a cavity cooling system and that the one-dimensional theoretical model can predict the internal temperature and mass-flow rate of the thermosyphon loop.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.248.1-248
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• 1999

Preliminary analysis of in-reactor thermal performance of three MOX fuel rods, which are going to be irradiated in the Halden reactor beginning in the first Quarter of the year 2000 under the framework of the OECD Halden Reactor Programme, have been conducted by using the computer code COSMOS to ensure their safe operation. Parametric studies have been carried out to investigate the effect of uncertainties on in-reactor behavior by considering the four kinds of uncertainties; thermal conductivity, linear power, manufacturing parameters, and model constants. The analysis shows that, in the case of annular MOX -1 fuel, calculation results for thermal performance vary widely depending on the selection of model constants for fission gas release (FGR). On the contrary, the thermal performance of solid MOX - 3 fuel does not depend on the choice of FGR constants to a large extent as MOX-I, because the fuel temperature is very high in the MOX-3 irrespective of the choice of FGR constants and hence the capacity of grain boundaries to retain gas atoms is not large enough to accommodate the number of gas atoms reaching the grain boundaries. It is planned that when the data on microstructure and thermal conductivity for each type of MOX fuel are available, new analysis will be made using these information. In addition, FGR model constants will be derived from the measured fuel centerline temperature, rod internal pressure and other related data.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2859-2874
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• 2024

This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as thermal boundary conditions were used, and ABAQUS thermal and structural analyses were performed. After full ablation, the temperature of the inner surface in the thinnest section remained high (920 ℃), but the stress remained relatively low (less than 6 MPa). At the outer surface, the stress was as high as 250 MPa; however, the resulting plastic strain was small owing to the low temperature of 200 ℃. Variations in stress, inelastic strain, and temperature with time in the thinnest section suggest that the plastic and creep strains are saturated owing to stress relaxation, resulting in low cumulative damage. Thus, the lower head of the vessel can maintain its structural integrity under LBLOCA with IVR-ERVC conditions. The sensitivity analysis of internal pressure indicates the occurrence of failure in the thinnest section at an internal pressure >9.6 MPa via local necking followed by failure due to high stresses.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2001.05a
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• pp.332-337
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• 2001

The objective of research is to clarify the interaction between dislocations and precipitates during high temperature creep deformation behaviors of high Mn austenitic steels. After measuring the internal stress in minimum creep rate at 873K, a transmission electron microscope (TEM) observation was performed to investigate the interaction between dislocations and precipitates during high temperature creep deformation. The band width of effective stress and internal stress increased when the nitride precipitates distribute more densely. Fine nitrides disturbed the dislocation movement with pinning the dislocations and perfect dislocations were separated into Shockley partial dislocations by fine nitrides. Coarse nitrides disturbed the dislocation movement with climb mechanism.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.11
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• pp.2866-2874
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• 2000

This paper is concerned with the optimum design for external reinforcement of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.