• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• Journal of computational fluids engineering
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• v.12 no.1
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• pp.22-27
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• 2007

Automobile underhood thermal and airflow simulation h α s been performed by using a commercial CFD program, FLUENT. To implement the radiation heat transfer effect to the underhood thermal and flow field, Discrete Ordinates Method(DOM) was used. The cooling fan was modeled by using the Multiple Reference Frame(MRF) technique. For the implementation of the heat exchangers, such as radiator and condenser, which are located in the front side of vehicle, the effectiveness-NTU model was used. The pressure drop throughout the heat exchangers was modeled as Porous media. For the validation of the current computational method, the coolant temperature at the inlet port of the radiator was compared with experimental data, and less than 3% error was observed. Finally, the composed model was used for the cooling fan spec determination process in the development of a new vehicle, and the results showed that the current CFD method could be successfully applied to the vehicle development process.

• Journal of Institute of Convergence Technology
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• v.4 no.2
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• pp.61-65
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• 2014

Pressure distribution around rotating impeller blades in centrifugal pump has been main issue for design of efficient and high performance automotive water pump. In addition, pressure losses of inlet water pipes should be considered to reduce additional pressure drop and design high performance engine cooling system. In this paper, pressure distribution inside water pump and pressure drop between inlet and outlet of water pump are investigated numerically to design plastic water pump for clean diesel engine application. And the inlet geometry of water pump was considered to analysis the effect of inlet water pipe geometry on pressure distribution around impeller blades and outlet pressure. The prediction results are compared with experimental data to validate and determine optimal operation condition without water pump cavitation. Major design parameters such as blade angle, volute geometry, system pressure, and coolant flow rate are considered to confirm applying possibility of plastic blades to the clean diesel engine.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.35-46
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• 1988

The prediction of clad surface temperatures is important to the design and the safety anlaysis of nuclear reactor. The accurate prediction requires the detailed knowledge of the flow structure and heat transfer, which is complicate due to anisotropic turbulent phenomena. A two-equation model including anisotropic eddy viscosity model is applied to forecast the velocity distribution. And the temperature field is calculated with uniform wall heat flux. The Galerkin's weighted residual finite element method has been used to calculate the turbulent quantities right up to the wall. The numerical results show good agreement with available data and that turbulence anisotropy strongly affects on the mean flow and thus the temperature field. And Nu-P/D correlation is established for sodium coolant in close-packed equilateral triangular bundle in the P/D range of 1.05 to 1.30.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.

• Tribology and Lubricants
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• v.39 no.4
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• pp.123-132
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• 2023

In this study, the expansion, stress, and operating clearance of bearing elements during operation are observed using the inner/outer ring temperature test data of a 3.0×10⁶ DN-class roller bearing. The operating clearance characteristics of inner-/outer-ring cooling (IORC) bearings are compared to those of inner-ring cooling (IRC) bearings. For IRC bearings, the thermal expansion of the outer ring is the most important factor in clearance variation. As a result, the operating clearance is less than the initial clearance of 61 ㎛, and the operating clearance decreases to 0.5 ㎛ at 25,500 rpm. Conversely, the temperature of the outer ring of IORC bearings is lower than that of IRC bearings, so the operating clearance is kept smaller. When the coolant flow rate to the outer ring is approximately 1.5 to 2.0 L/min, the temperature difference between the inner and outer rings is minimized and the operating clearance is maintained at a significantly lower level than IRC bearings. Small operating clearances are expected to be effective in reducing cage slip and skid damage in roller bearings. The results and analysis procedures of this study can be utilized to design of bearing clearance, lubricant flow rate, and assembled interference in the early design stage of aerospace roller bearings.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.126-139
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• 1994

Best-estimate thermal-hydraulic codes, CATHARE2 V1.2 and RELAP5/MOD3, hate been assessed against the BETHSY 6.2 tc six-inch cold leg break loss-of-coolant accident (LOCA) test. Main objective is to analyze the overall capabilities of the two codes on physical phenomena of concern during the small break LOCA i.e. two-phase critical flow, depressurization, core water level de-pression, loop seal clearing, liquid holdup, etc. The calculation results show that the too codes predict well both in the occurrences and trends of major two-phase flow phenomena observed. Especially, the CATHARE2 calculations show better agreements with the experimental data. However, the two codes, in common, show some deviations in the predictions of loop seal clearing, collapsed core water level after the loop seal clearing, and accumulator injection behaviors. The discrepancies found from the comprision with the experimental data are larger in the RELAP5 results than in the CATHARE2. To analyze the deviations of the two code predictions in detail, several sensitivity calculations have been performed. In addition to the change of two-phase discharge coefficients for the break junction, fine nodalization and some corrections of the interphase drag term are made. For CATHARE2, the change of interphase drag force improves the mass distribution in the primary side. And the prediction of SG pressure is improved by the modification of boundary conditions. For RELAP5, any single input change doesn't improve the whole result and it is found that the interphase drag model has still large uncertainties.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.97-106
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• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.