• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.441-456
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• 2001

ABB-CE＇s regionalization methodology for the core inlet flow distribution map is reviewed. This methodology merges the test data of fuel assembly locations which are either in symmetry or strongly correlated with others. It increases the number of available test data for each regional flow factor It makes up effectively for the deficiency due to limited number of test data. It also contributes to making the core inlet flow distribution smoother not only locally but also over the entire core, and to relieving the impacts of test errors that may happen due to some do- calibrated local pressure measurement taps. As a result, the core inlet How distribution data becomes more statistically useful and thus the conservatism involved in handling the core inlet flow factors for the thermal margin analysis is expected to be reduced. Meanwhile, the regionalized map may lose the unique local characteristics in core inlet flow distribution too much. By an alternative approach introduced in the present work, it is shown that such a disadvantage can be mitigated somewhat if the engineering judgement is made more

• Nuclear Engineering and Technology
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• v.53 no.1
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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.

• Journal of Advanced Marine Engineering and Technology
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• v.39 no.6
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• pp.604-608
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• 2015

In this paper, we study the characteristics of heat transfer for an insulated multi-core tube using glass wool as an insulator for the multi-core tube. By performing experiments and modeling, we examine the variations in the temperature characteristics of hydraulic oil inside the multi-core tube with atmosphere temperature, inlet temperature, and the flow rate of hydraulic oil for the insulated multi-core tube that we developed. When the minimum inlet flow rate of hydraulic oil employed within the scope of the research is 0.29 l/min, the temperature difference obtained in the experiments and numerical analysis was a maximum of $3^{\circ}C$. For a constant atmospheric temperature, as the inlet temperature of the hydraulic oil increases, the outlet temperature of the hydraulic oil will also increase, regardless of its inlet flow rate. Further, when the inlet flow rate of the hydraulic oil is more than 1.01 l/min, the effect of the atmospheric temperature on the temperature drop of the hydraulic oil is low.

• Journal of Korea Water Resources Association
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• v.52 no.4
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• pp.235-243
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• 2019

Recently, urban inundation was frequently occurred due to the intensive rainfall exceeding marginal capacity of the flood control facility. Furthermore, needs for the underground storage facilities to mitigate urban flood are increasing according to rapidly accelerating urbanization. Thus, in this study, drainage efficiency in drain tunnel connecting to underground storage was investigated from the air-core measurements in the drop shaft against two types of inlet structure. In case of the spiral inlet, the multi-stage structure is introduced at the bottom of the inlet to improve the vortex induction effect at low inflow discharge (multi-stage spiral inlet). The average cross-sectional area of the air-core in the multi-stage spiral inlet is 10% larger than the tangential inlet, and show the highly air emission effect and the highly inflow efficiency at the high inflow discharge. In case of the tangential inlets, the air emission effect decreased after exceeding the maximum inflow discharge, which is required to maintain the inherent performance of the tangential inlet. From the measurements, the empirical formula for the cross-sectional area of the air-core according to locations inside the drop shaft was proposed in order to provide the experimental data available for the inlet model used in experiments.

• Journal of computational fluids engineering
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• v.20 no.1
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.