• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.91-102
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• 2015

A thermal-hydraulic design and performance analysis computer code for a once-through steam generator using straight tubes is developed. To benchmark the developed physical models and computer code, an once-through steam generator developed by other designer is simulated and the calculated results are compared with the design data. Also, the same steam generator is analyzed with the best-estimate thermal-hydraulic system code, MARS, for the code-to-code validation. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by the developed code show general agreements with the published design data as well as the analysis results of MARS. It is demonstrated that the developed code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

• Computers and Concrete
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• v.15 no.1
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• pp.103-126
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• 2015

Thermal effects are significant loads for assessing concrete dam behaviour during operation. A new methodology to estimate thermal loads on concrete dams taking into account processes which were previously unconsidered, such as: the evaporative cooling, the night radiating cooling or the shades, has been recently reported. The application of this novel approach in combination with a three-dimensional finite element method to solve the heat diffusion equation led to a precise characterization of the thermal field inside the dam. However, that approach may be computationally expensive. This paper proposes the use of a new one-dimensional model based on an explicit finite difference scheme which is improved by means of the reported methodology for computing the heat fluxes through the dam faces. The improved model has been applied to a case study where observations from 21 concrete thermometers and data of climatic variables were available. The results are compared with those from: (a) the original one-dimensional finite difference model, (b) the Stucky-Derron classical one-dimensional analytical solution, and (c) a three-dimensional finite element method. The results of the improved model match well with the observed temperatures, in addition they are similar to those obtained with (c) except in the vicinity of the abutments, although this later is a considerably more complex methodology. The improved model have a better performance than the models (a) and (b), whose results present larger error and bias when compared with the recorded data.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.17-28
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• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• 한국신재생에너지학회:학술대회논문집
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• 2005.06a
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• pp.697-700
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• 2005

Aquifer Thermal Energy Storage (ATES) system can be very cost-effective and renewable energy sources, depending on site-specific parameters and load characteristics. In order to develop an ATES system which has certain hydrogeological characteristics, understanding of the thermo hydraulic processes of an aquifer is necessary for a proper design of an aquifer heat storage system under given conditions. The thermo hydraulic transfer for heat storage is simulated using FEFLOW according to two sets of pumping and waste water reinjection scenarios of heat pump operation in a two layered confined aquifer. In the first set of model, the movement of the thermal front and groundwater level are simulated by changing the locations of injection and pumping well in seasonal cycle. However, in the second set of model the simulation is performed in the state of fixing the locations of pumping and injection well. After 365 days simulation period, the temperature distribution is dominated by injected water temperature and the distance from injection well. The small temperature change is appears on the surface compared to other slices of depth because the first layer has very low porosity and the transfer of thermal energy are sensitive at the porosity of each layer. The groundwater levels and temperature changes in injection and pumping wells are monitored to validate the effectiveness of the used heat pump operation method and the thermal interference between wells is analyzed.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.28 no.11
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• pp.1322-1331
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• 2004

This paper presents studies on the design of water hydraulic system and components to replace oil with tap-water as the pressure transmission medium in hydraulic systems. In order to improve the performance of water hydraulic system, the thermal and hydraulic properties of tap-water are first investigated. Based on these characteristics, the design parameters, such as the clearances of the moving parts, the cross-sectional area of pipes and relative roughness, are proposed so that the performance of water hydraulic system is the same as that of oil. In addition, the operating ranges, which show the possibility of using water hydraulic system, are examined.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.977-986
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• 2019

In the thermal-hydraulic system codes, such as MARS and RELAP5/MOD3, the Savannah River Laboratory (SRL) model has been adopted as a subcooled boiling model. It, however, has been shown that the SRL model cannot take into account appropriately the effects of inlet liquid velocity and hydraulic diameter on axial void fraction development. To overcome the problems, Ha et al. (2018) proposed a modified SRL model, which is applicable to low-pressure and low-Pe conditions (P < 9.83 bar and $Pe{\leq}70,000$) only. In this work, the authors extended the modified SRL model by proposing a new net vapor generation (NVG) model and a wall evaporation model so that the new subcooled boiling model can cover a wide range of thermal-hydraulic conditions with pressures ranging from 1.1 to 69 bar, heat fluxes of $97-1186kW/m^2$, Pe of 3600 to 329,000, and hydraulic diameters of 5-25.5 mm. The new model was implemented in the MARS code and has been assessed using various subcooled boiling experimental data. The results of the new model showed better agreements with measured void fraction data, especially at low-pressure conditions.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.