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

• Resources Recycling
• /
• v.19 no.2
• /
• pp.45-53
• /
• 2010

During the biomass gasification, tar generation is typically accompanied, which causes many problems, such as pipe plugging and equipment fouling. In the experiments, activated carbon was applied to the upper reactor of the two-stage gasifier in order to remove the tar generated during gasification. In addition, the effects of the upper-reactor temperature and equivalence ratio on the producer gas characteristics (composition, tar content and lower heating value) were investigated. To investigate the effect of the upper reactor-temperature, experiments were performed at 743, 793, $838^{\circ}C$, respectively. To examine the influence of the equivalence ratio, a comparison experiment was carried out at a equivalence ratio of 0.17. In all experiments, tar contents in the producer gases were below $2mg/Nm^3$. The maximum LHV of the producer gas was above $10MJ/Nm^3$, which is much higher than the typical LHV($3\sim6MJ/Nm^3$) in the air gasification of biomass.

• Journal of the Semiconductor & Display Technology
• /
• v.15 no.1
• /
• pp.41-46
• /
• 2016

In this study, gas flow and temperature distribution in the multi-wafer planetary CVD reactor for the Si epitaxial growth were analyzed. Although the structure of the reactor was simplified as the first step of the study, the three-dimensional analysis was performed taking all these considerations of the revolution of the susceptor and the rotation of satellites into account. From the analyses, a reasonable velocity field and temperature field were obtained. However, it was found that analyses including the upper structure of the reactor were required in order to obtain more realistic temperature results. DCS mole fraction above the satellite surface and the susceptor surface without satellite was compared in order to check the gas species mixing. We found that satellite rotation helped gases to mix in the reactor.

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

• Journal of environmental and Sanitary engineering
• /
• v.17 no.3
• /
• pp.46-51
• /
• 2002

This study was performed to treat a sewage at the upper stream of dam using modified sequencing batch reactor, During the operating period, average $COD_{cr}$, removal efficiency was about 85% but average T-N and ${PO_4}^{3-}-P$ removal efficiencies were 43% and 30% respectively. Because the organic matter was very low compared with nitrogen and phosphorous in influent($BOD_{5}/{NH_4}^{+}-N{\;}={\;}2,{\;}BOD_{5}/{PO_4}^{3-}-P{\;}={\;}15.6$), nitrogen and phosphorus removal efficiency was relatively low. Average nitrogen removal efficiency was 50 % at $10^{\circ}C$ or above and it was 36 % at $10^{\circ}C$ or below. As reactor was located in outdoor without any thermostat, temperature decreased at least $2.4^{\circ}C$ in the winter season. Therefore, if we would apply this modified sequencing batch reactor to sewage which concentration of organic matter was very low compared with nitrogen and phosphorous, we have to addition of external carbon and installation of any thermostat.

• Applied Chemistry for Engineering
• /
• v.10 no.8
• /
• pp.1200-1209
• /
• 1999

Prediction model was composed by optimal parameter estimation from best fitting on reactant temperature profile, inlet and outlet temperature of coolant and yield of dual fixed-bed catalytic reactor(FBCR) which was measured in the industrial field. In order to design the FBCR which could obtain maximum conversion and yield, we investigated the effect of catalyst bed length and reactor radius changes. An uniform activity FBCR showed the best performance at z = 2.8 m of total catalysst bed length in case of reactor radius r = 0.01241 m and z =2.80 m(upper layer: 1.88 m, lower layer: 0.92 m) under reactor radius r = 0.01254 m for a dual activities FCBR. In case of reactor radius changes, the axial temperature profile and maximum radial temperature was rapidly risen for radius increase. The reactor radius decrease showed the opposite result.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.357-373
• /
• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
• /
• v.44 no.5
• /
• pp.543-550
• /
• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• Nuclear Engineering and Technology
• /
• v.33 no.4
• /
• pp.397-408
• /
• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1821-1829
• /
• 2023

The temperature distribution of the reactor pool under natural circulation induced by the RVACS operation was experimentally studied. According to the Bo' based similarity law, which could reproduce the temperature distribution of the working fluid under natural circulation, SINCRO-2D facility was designed based on the PGSFR. It was reduced to 1 : 25 in length scale, having water as a simulant of the sodium, which is the original working fluid. In general, temperature was stratified, however, effect of the natural circulation flow could be observed by the entrainment of the stratified temperature. Relative cooling contribution of the upper plenum (narrow gap) and lower plenum was approximately 0.2 and 0.8, respectively. In the range of decay heat from 0.2% to 1.0%, only the magnitude of the temperature was changed, while the normalized temperature maintained. Boundary temperature distribution change made a global temperature offset of the pool, without a significant local change. Therefore, the decay heat and cooling boundary condition had no significant effect on temperature distribution characteristics of the pool within the given range of the decay heat and boundary temperature distribution.