Jae Jun Jeong

DEVELOPMENT AND PRELIMINARY ASSESSMENT OF A THREE-DIMENSIONAL THERMAL HYDRAULICS CODE, CUPID

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.

MULTI-SCALE THERMAL-HYDRAULIC ANALYSIS OF PWRS USING THE CUPID CODE

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a systemscale code, MARS.

RECENT IMPROVEMENTS IN THE CUPID CODE FOR A MULTI-DIMENSIONAL TWO-PHASE FLOW ANALYSIS OF NUCLEAR REACTOR COMPONENTS

The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of a two-phase flow in light water nuclear reactor components. It can provide both a component-scale and a CFD-scale simulation by using a porous media or an open media model for a two-phase flow. In this paper, recent advances in the CUPID code are presented in three sections. First, the domain decomposition parallel method implemented in the CUPID code is described with the parallel efficiency test for multiple processors. Then, the coupling of CUPID-MARS via heat structure is introduced, where CUPID has been coupled with a system-scale thermal-hydraulics code, MARS, through the heat structure. The coupled code has been applied to a multi-scale thermal-hydraulic analysis of a pool mixing test. Finally, CUPID-SG is developed for analyzing two-phase flows in PWR steam generators. Physical models and validation results of CUPID-SG are discussed.

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

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal-hydraulic 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.

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

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 two-phase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling.

IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operators action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

DEVELOPMENT OF A COMPUTER PROGRAM TO SUPPORT AN EFFICIENT NON-REGRESSION TEST OF A THERMAL-HYDRAULIC SYSTEM CODE

During the development process of a thermal-hydraulic system code, a non-regression test (NRT) must be performed repeatedly in order to prevent software regression. The NRT process, however, is time-consuming and labor-intensive. Thus, automation of this process is an ideal solution. In this study, we have developed a program to support an efficient NRT for the SPACE code and demonstrated its usability. This results in a high degree of efficiency for code development. The program was developed using the Visual Basic for Applications and designed so that it can be easily customized for the NRT of other computer codes.

Preliminary Analysis of the CANDU Moderator Thermal-Hydraulics using the CUPID Code

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code, CUPID

For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID . It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model.

The MARS Simulation of the ATLAS Main Steam Line Break Experiment

A main steam line break (MSLB) test at the ATLAS facility was simulated using the best-estimate thermal-hydraulic system code, MARS-KS. This has been performed as an activity at the third domestic standard problem for code benchmark (DSP-03) that has been organized by Korea Atomic Energy Research Institute (KAERI). The results of the MSLB experiment and the MARS input data prepared for the previous DSP-02 using the ATLAS facility were provided to participants. The preliminary MSLB simulation using the base input data, however, showed unphysical results in the primary-to-secondary heat transfer. To resolve the problems, some improvements were implemented in the MARS input modelling. These include the use of fine meshes for the bottom region of the steam generator secondary side and proper thermal-hydraulics calculation options. Other input model improvements in the heat loss and the flow restrictor models were also made and the results were investigated in detail. From the results of simulations, the limitations and further improvement areas of the MARS code were identified.