Integral loop test and assessment of modified RELAP5/MOD3.3 for RCS coolant inventory during a LBLOCA

This study investigates experimentally and analytically the thermal hydraulic phenomena during the transition from design basis accident (DBA) to beyond-DBA, particularly, the depletion of core coolant inventory. To investigate the overall thermal hydraulic behavior after safety injection (SI) failure during a large-break loss-of-coolant accident (LBLOCA) in a cold leg, an integral loop test was performed at the Seoul National University Integral Test Facility (SNUF), which was scaled down to 1/6.4 in length and 1/178 in area from the advanced power reactor 1400 MWe (APR1400) according to the three-level scaling method. The plant condition at 200.0 s as the base case and those at 625.0 and 1950.0 s as test cases after the initiation of LBLOCA were applied as initial conditions in the experiments, respectively. The experimental results showed that the sweepout increased the coolant flow discharged to the break depending on the steam flow rate in intact cold legs and the initial downcomer coolant level and expedited the depletion of the core coolant inventory.In the meantime, since RELAP5/MOD3.3 uses the average properties of donor volume as those of its connected junction, this scheme causes the mass and the energy flux in a junction to be calculated incorrectly if significant phase separation occurs in the donor volume such as in the downcomer during the LBLOCA. The sweepout model was developed and implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory during the LBLOCA. To assess the applicability of the modified RELAP5/MOD3.3 to the actual system, the experiments in SNUF were simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the discharge flow rate at the break larger than that of the experiment. On the other hand, the modified one calculated the discharge flow rate more similar to that of the experiment than the original one did. As a result, the modified RELAP5/MOD3.3 reduced the coolant flow discharged to the break to delay the initiation time of heater heat-up after SI failure during LBLOCA in a cold leg. This improved RELAP5/MOD3.3 will support a more realistic thermal hydraulic analysis in an integrated analysis system.     

Assessment of the GOTHIC code for prediction of hydrogen flame propagation in small scale experiments

With the rising concerns regarding the time and space dependent hydrogen behavior in severe accidents, the calculation for local hydrogen combustion in compartment has been attempted using CFD codes like GOTHIC. In particular, the space resolved hydrogen combustion analysis is essential to address certain safety issues such as the safety components survivability, and to determine proper positions for hydrogen control devices as e.q. recombiners or igniters. In the GOTHIC 6.1b code, there are many advanced features associated with the hydrogen burn models to enhance its calculation capability.In this study, we performed premixed hydrogen/air combustion experiments with an upright, rectangular shaped, combustion chamber of dimensions 1 m × 0.024 m × 1 m. The GOTHIC 6.1b code was used to simulate the hydrogen/air combustion experiments, and its prediction capability was assessed by comparing the experimental with multidimensional calculational results. Especially, the prediction capability of the GOTHIC 6.1b code for local hydrogen flame propagation phenomena was examined. For some cases, comparisons are also presented for lumped modeling of hydrogen combustion. By evaluating the effect of parametric simulations, we present some instructions for local hydrogen combustion analysis using the GOTHIC 6.1b code. From the analyses results, it is concluded that the modeling parameter of GOTHIC 6.1b code should be modified when applying the mechanistic burn model for hydrogen propagation analysis in small geometry.     

CFD Analysis and Assessment for Cross-Flow Phenomena in VHTR Prismatic Core

The very high temperature gas-cooled reactor (VHTR) is a uranium-fueled, graphite-moderated, and helium-cooled reactor envisioned as one of the promising future nuclear reactor concepts due to its high efficiency, safety, and variety of applications. In this reactor concept, core bypass flow and cross flow are currently considered as key issues because of their inherent uncertainties and complication of prediction. Recently, the computational fluid dynamics (CFD) method has received a great deal of attention as a method for understanding the flow behavior in the VHTR core, including the bypass and the cross flow. However, validation of this method has not been sufficiently conducted yet based on real experimental data. For this reason, prediction capability of the CFD method was validated in this study by comparing the predictions with the existing multi-hole experimental data obtained by Groehn as a part of the core thermal-hydraulics design study for the NHDD PMR-200. A two-stacked fuel block with wedge-shaped cross gap was simulated for computational domain as the experimental setup. The flow loss coefficients and the velocity distributions of the cross flow from the experiment were compared to the CFD predictions extensively. As a result, good agreements between the CFD predictions and the experimental results were observed, confirming prediction capability of the CFD method for the complicated cross flow in the VHTR core. Furthermore, the velocity distributions and pressure distribution in the cross gap were investigated to identify the characteristics of the cross flow.     

Simulation of wall film condensation with non-condensable gases using wall function approach in component thermal hydraulic analysis code CUPID

In the containment of nuclear power plants, wall film condensation occurs with non-condensable gases under accident conditions. With non-condensable gases, condensation heat transfer on the containment wall can be degraded significantly because of the accumulation of non-condensable gases near the condenser wall; therefore, an investigation into the wall condensation heat transfer is of great importance to nuclear reactor safety. In this study, wall film condensation with non-condensable gases was simulated using the CUPID code. To evaluate the heat and mass transfer coefficients, a wall function approach was adopted to save the computational cost. To validate the model, a COPAIN condensation experiment was simulated using CUPID. The calculation results were compared with the COPAIN experiment data and results from the commercial CFD code (STAR-CCM+) results, which used the resolved boundary layer approach. From the comparison, good agreements were obtained between the CUPID code and the other results.     

Interfacial heat transfer of condensing bubble in subcooled boiling flow at low pressure

The interfacial heat transfer coefficient is an important parameter for the analysis of multi-phase flow. In subcooled boiling flow, bubbles condense through the interface of phases and the interfacial heat transfer determines the condensation rate which affects the two-phase parameters such as void fraction and local liquid temperature. Thus, the present experiments are conducted to correlate the interfacial heat transfer coefficient at low pressure in the subcooled boiling flow. The local liquid temperature is measured by microthermocouple and the bubble condensation rate is estimated by orthogonal, two-image processing. The condensate Nusselt number, which is a function of bubble Reynolds number, local liquid Prandtl number, and local Jacob number, is obtained from the experimental results. The bubble history is derived from the newly proposed correlation and the condensate Nusselt number is compared with the previous models.     

Experimental analysis on mixed convection in reactor cavity cooling system of HTGR for hydrogen production

To achieve the massive production of hydrogen, a thermo-chemical cycle coupled to a High Temperature Gas-cooled nuclear Reactor (HTGR) is considered as one of the systems exhibiting high application potential. The Reactor Cavity Cooling System is a key safety system that ensures the integrity of the HTGR during accident conditions; therefore, it is necessary to verify its performance for the safety of the HTGR as well as for the reliability of the coupled hydrogen production system. The RCCS performance depends on the heat transfer rate of riser ducts; however, the mixed convection that is likely to occur in the riser ducts can complicate the verification process. In this study, an experimental facility was constructed to investigate heat transfer experiment inside a riser duct and experiments with various heat flux and flow rate conditions were carried out. The experimental results demonstrated that the mixed convection occurred under certain experimental conditions in the riser duct, which resulted in heat transfer deterioration. The evaluated heat transfer coefficients from the experimental results were not consistent with those predicted by extant mixed convective heat transfer correlations which were derived from data obtained for different test section configurations. Therefore, a modified correlation was proposed to fit the experimental data for the RCCS riser duct with an average error of 6.06%. The correlation will contribute to the verification of RCCS performance and the credibility of the HTGR-coupled hydrogen production system.     

Assessment of the MARS code for design of water pool-type reactor cavity cooling system in HTGR

This paper reports an experimental and numerical study on the assessment of the MARS code as a tool for analyzing the water pool-type reactor cavity cooling system (RCCS), which was developed by Seoul National University (SNU). A series of experiments were performed to determine the heat removal capability of the proposed RCCS and assess the capability of MARS code to predict the forced convective, natural convective and radiative heat transfer under normal operation conditions and boiling heat transfer during accident conditions in the RCCS. In the loss of forced convection (LOFC) accident experiment performed at the integral effect test facility called RCCS-SNU, the MARS code underestimated the vapor generation rate at the inner wall of the water pool. Therefore, the newly developed models of the bubble departure and lift-off diameters were implemented into the MARS code to make a better prediction of the vapor generation rate. The improved MARS code was assessed again using the experimental data of the LOFC accident conditions in the RCCS-SNU facility.     

Development of bubble departure and lift-off diameter models in low heat flux and low flow velocity conditions

The forces acting on a bubble were analyzed to determine the criteria for bubble departure and lift-off from a heated wall. Mechanistic models for the bubble departure and lift-off diameters were developed from the analysis results to predict the evaporation heat flux under low heat flux and low flow velocity conditions. The developed model for the departure diameter depends strongly on the contact angle, but the model for the lift-off diameter was found to be a function of the lift-off number, which indicates the ratio between the shear lift force and the growth force. Wall boiling experiments were performed to measure bubble departure and lift-off diameters using digital image processing. From the experimental data, the developed models were validated, and the results showed good agreement between the experimental data and the predicted diameters.     

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor,REX-10

The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional four-equation drift-flux model for two-phase flows. It also consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.     

Development of linked analysis system for PWR severe accidents utilizing RELAP5 and MAAP4

The purpose of this study is to develop a severe accident (SA) analysis method that is more reliable thorough transferring the physical status of the plant predicted by RELAP5 computer code to MAAP4 computer code. The methodology of the linkage analysis is developed and the criterion of linkage time is suggested to utilize the RELAP5 thermal–hydraulic calculation to the maximum degree possible and thereby guarantee the continuity of calculation for hydrogen generation. The MAAP4 calculations after data transfer show the physically proper results based on RELAP5 data. Comparison with other code results for TMI-2 accident reveals that the result from the RELAP5–MAAP4 linked analysis lay in the span given by a number of results of TMI calculation from other SA code systems. The results of this study are expected to improve the SA analysis methodology by analyzing an SA scenario with more reliable thermal–hydraulic initial conditions.