Development and Preliminary Assessment of a Three-Dimensional Thermal Hydraulics Code, Cupid

Jeong, Jae Jun; Yoon, Han Young; Park, Ik Kyu; Cho, Hyoung Kyu; Lee, Hee-Dong

doi:10.5516/net.2010.42.3.279

Cited by 33 publications

(11 citation statements)

References 20 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Xn should be decreased along the flow path due to the evaporation of the water vapor and its theoretical minimum is determined by the local saturation pressure of the water. The results reveal that the CUPID code works well with noncondensable gas [11]. In summary, the results of the verification show that the CUPID governing equations for two-phase flows were solved correctly on both structured and unstructured meshes.…”

Section: Verificationmentioning

confidence: 48%

“…The validation process is a work in progress for a limited range of two-phase flow conditions [11]. Basically, the CUPID code was designed for the transient analysis of a two-phase flow at component scale and has both features of porous and open media approach.…”

Section: Validation Status and Strategymentioning

confidence: 99%

“…The Korea Atomic Energy Research Institute has been developing a three-dimensional thermal-hydraulic code, called CUPID, which was motivated from practical needs for the realistic simulation of two-phase flows in nuclear reactor components [10,11]. The need for a multi-dimensional analysis is further increasing with advanced design features, such as a direct vessel injection (DVI) system, a gravity-driven safety injection system, and a passive secondary cooling system.…”

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

The Cupid Code Development and Assessment Strategy

Jeong¹,

Yoon²,

Park³

et al. 2010

Nuclear Engineering and Technology

View full text Add to dashboard Cite

Section: Verificationmentioning

confidence: 48%

Section: Validation Status and Strategymentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

The Cupid Code Development and Assessment Strategy

Jeong¹,

Yoon²,

Park³

et al. 2010

Nuclear Engineering and Technology

View full text Add to dashboard Cite

“…They have been verified against standard conceptual problems of single-and two-phase flows and validated for thermal-hydraulic experiments in our previous studies [3][4][5]. The assessment strategy for the future verification and validation was outlined in Jeong et al [2].…”

Section: Introductionmentioning

confidence: 98%

Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Cho

Yun

Yoon

et al. 2012

Journal of Nuclear Science and Technology

View full text Add to dashboard Cite

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the downcomer boiling experiment (DOBO) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change from a bubbly flow to churn and mist flows and a circulation of liquid accelerated by bubbles. The twophase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling.

show abstract

“…Inside the water pool, the hydraulic driving force is very small compared with conventional active systems, and the fluid flow is multi-dimensional with subcooled boiling and flashing. Thus, a sophisticated two-phase thermal-hydraulic analysis is required to assess the test data and to evaluate the performance of the passive safety system [2]. This technical need for the multi-dimensional analysis motivated the development of a three-dimensional component analysis code, named CUPID [3], which is in progress at KAERI.…”

Section: Introductionmentioning

confidence: 99%

Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

Cho

Lee

Yoon

et al. 2013

Journal of Nuclear Science and Technology

View full text Add to dashboard Cite

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermalhydraulic code, named CUPID, has been developed. In the present study, the CUPID code was applied for the simulation of the PASCAL test facility constructed with an aim of validating the cooling and operational performance of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor + (APR + ), which is intended to completely replace the conventional active auxiliary feedwater system. This paper introduces the simulation results for the passive condensate cooling tank (PCCT) of the PASCAL facility performed with the CUPID code in order to investigate the thermal-hydraulic phenomena in the PCCT. The simulation showed that the important thermal-hydraulic characteristics in the PCCT, such as two-phase natural circulation and boil-off phenomena, can be successfully reproduced by CUPID. Two important validation parameters, collapsed water level and local liquid temperature, were quantitatively well captured in the simulation. This paper presents the description of the PASCAL test facility, the physical models of the CUPID code, and its simulation result for the PCCT.

show abstract

Development and Preliminary Assessment of a Three-Dimensional Thermal Hydraulics Code, Cupid

Cited by 33 publications

References 20 publications

The Cupid Code Development and Assessment Strategy

The Cupid Code Development and Assessment Strategy

Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

Contact Info

Product

Resources

About