Numerical analysis of RBHT reflood experiments using MARS 1D and 3D modules

The Rod Bundle Heat Transfer (RBHT) program was performed experimentally to analyze the reflood heat transfer phenomena under the conditions of reflood phase following a hypothesized loss of coolant accident (LOCA) by the team of Penn State University. In order to verify the experimental data using a numerical analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) assessment of the RBHT experimental data was carried out for the flooding rates of 0.0254 and 0.1524 m/sec with the upper plenum pressure of 276 kPa. The RBHT experimental data of Tests 1285 and 1383 were compared with the calculation results of the MARS 1D and 3D modules. The MARS code shows a good agreement in the general trend of the peak cladding temperatures although there are limitations in predicting accurate quenching time for both modules. However, in comparison to the MARS 1D module simulation, the MARS 3D module shows the improved calculation capability in that the code can capture local enhanced heat transfer with implication of spacer grids. Moreover, the temperature profiles simulated by the 3D module show the accurate prediction at which the local peak temperatures occur. For more enhanced simulations, local flow parameters such as cross flow and vortex flow need to be analyzed for a more accurate prediction of quenching behavior.     

Phase change heat transfer device for process heat applications

The next generation nuclear plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼1300 K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+ m) dictated by safety and licensing mandates.The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a thermal device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via ‘pumping a fluid’, a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization/condensing process. The condensate is further returned to the hot source by gravity, i.e., without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.     

Two phase natural circulation and the heat transfer in the passive residual heat removal system of an integral type reactor

An investigation of the thermal hydraulic characteristics and the natural circulation performance in the passive residual heat removal system (PRHRS) for an integral type reactor have been carried out using the VISTA facility and the calculated results using the MARS code, which is a best estimate system analysis code have been compared with the experimental results. The VISTA facility consists of the primary, secondary, and the PRHRS circuits, to simulate the SMART design verification program. The experimental results show that the fluid is well stabilized in the PRHRS loop and the PRHRS heat exchanger accomplishes well its functions in removing the transferred heat from the primary side in the steam generator as long as the heat exchanger is submerged in the water in the emergency cooldown tank (ECT). The decay heat and the sensible heat can be sufficiently removed from the primary loop with the operation of the PRHRS. The MARS code predicts reasonably well the characteristics of the natural circulation in the PRHRS. From the calculation results, most of the heat transferred from the primary system is removed at the PRHRS heat exchanger by a condensation heat transfer.     

Investigation of mixed convection in a large rectangular enclosure

This experimental research investigates mixed convection and heat transfer augmentation by gaseous forced jets in a large enclosure, at conditions simulating those of passive containment cooling systems for Gen III+ passively safe reactors. The experiment is designed to measure the key parameters governing heat transfer augmentation by forced jets, and to investigate the effects of geometric factors, including the jet diameter, jet injection orientation, interior structures, and enclosure aspect ratio. The tests cover a variety of injection modes leading to flow configurations of interest for mixing and stratification phenomena in containments under accident conditions. Correlations for heat transfer augmentation by forced jets are developed and compared with experimental data. The characteristic recirculation speed inside the enclosure is introduced and analyzed. Steady stratified temperature distributions are compared with model simulations of the BMIX++ code.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I)

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include:

1. Froth level (or dry-out level) calculation

2. Transition and mixing between convection flow regimes in convective heat transfer

3. Radiant heat transfer between solid walls and flowing gas

4. Heat generation by zirconium-water reaction

5. Crucibilization effect of zirconium-oxide layer

6. Steam starvation effect on zirconium-water reaction.

This code does not calculate motion of fuel rod material but predicts the beginning of relocation. The major affecting models, froth level calculation model, heat transfer model and crucibilization model, are verified through analyses of experiments. This code can be used for thermal hydraulic analysis of a severe accident and fuel damage experiment until significant material relocation occurs.     

Heat transfer coefficients under LOCA conditions in containment buildings

A model to calculate local heat transfer coefficients between the containment atmosphere and the walls of a pressurized water reactor containment building after a loss-of-coolant accident has been developed. The new calculation is based on the containment wall and atmosphere bulk temperatures, mass ratio of steam to air, and condensing or convective heat transfer conditions. Comparison with the Carolinas Virginia Tube Reactor Containment Tests shows good agreement. The model has been implemented into the containment code TECAR.     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

Modeling of turbulent condensation heat transfer in the boiling water reactor primary containment

A condensation heat transfer model is developed for the purpose of predicting the atmosphere temperature response within the primary containment of a boiling water reactor during the initial forced convection heat transfer period following a postulated loss-of-coolant accident. The model utilizes simultaneous heat and mass transfer for the process of condensation in the presence of a non-condensible gas. The gas-vapor diffusion layer formed is in the mode of turbulent, forced convection. The predicted heat transfer is determined to be diffusion controlled with negligible resistance being contributed by the condensate film. The model is qualified through the analysis of the response of a containment test facility; the results compare favorably with experimental observations made by the General Electric Co. Predicted temperature responses for a typical containment are also shown and compared with those obtained through use of the Uchida heat transfer correlations.     

Reactor cavity cooling system (Rccs) experimental characterization

The development of a new concept for a high temperature gas cooled reactor is strictly correlated to the Reactor Cavity Cooling System (RCCS) design. This is the latest heat sink designed to ensure the cooling down of the vessel and other structural materials during an accident scenario. An experimental facility was built at Texas A&M to quantitatively analyze heat transfer phenomena and the air flow regime inside the reactor cavity. The thermal measurements were performed using 18 thermocouples mounted on the vessel surface, 8 on the external surface of one of the cooling pipes and 24 on a movable rack that captures the axial temperature profile inside the cavity. Flow regime measurements were performed with Particle Tracking Velocimetry techniques (PTV), using a high speed camera and spraying a special dust in the cavity for the tracking. The results demonstrate that the main heat transfer mode inside the cavity is the radiation (about 80%); also, they show the complexity of the flow regime inside the cavity due to natural circulation. The experimental conditions were a vessel surface temperature of about 300 °C heated at fixed power and different flow rates for the cooling pipes.     

Modeling condensation heat transfer on a horizontal finned tube in the presence of noncondensable gases

European designs for the next generation of nuclear reactors incorporate innovative passive systems in their containments to enhance heat removal by condensation under postulated accident conditions. These systems consist of several units of cross-flow finned tube bundles internally cooled with water. So far most of the studies that have been addressed to the issue of heat transfer onto finned surfaces under condensing conditions have involved refrigerants and pure vapor conditions. This study presents a model (HTCFIN) capable of predicting condensation of a cross-flow air–steam mixture onto a single horizontal finned tube. The comparison of HTCFIN predictions to the available databases shows its acceptable accuracy in a wide range of conditions and allows an interpretation of the influence of major variables acting on the scenario. As a consequence, HTCFIN model represents a step forward in the present theoretical capability to estimate heat transfer within containments of next generation of European reactors in the case of a hypothetical accident.     

Investigations on post-dryout heat transfer in bilaterally heated annular channels

Post-dryout heat transfer in bilaterally heated vertical narrow annular channels with 1.0, 1.5 and 2.0 mm gap size has been experimentally investigated with deionized water under the condition of pressure ranging from 1.38 to 5.9 MPa and low mass flow rate from 42.9 to 150.2 kg/m²s. The experimental data was compared with well known empirical correlations including Groeneveld, Mattson, etc., and none of them gave an ideal prediction. Theoretical investigations were also carried out on post-dryout heat transfer in annular channels. Based on analysis of heat exchange processes arising among the droplets, the vapor and two tube walls of annular channel, a non-equilibrium mechanistic heat transfer model was developed. Comparison indicated that the present model prediction showed a good agreement with our experimental data. Theoretical calculation result showed that the forced convective heat transfer between the heated wall and vapor dominate the overall heat transfer. The heat transfer caused by the droplets direct contact to the wall and the interfacial convection/evaporation of droplets in superheated vapors also had an indispensable contribution. The radiation heat transfer would be neglected because of its small contribution (less than 0.11%) to the total heat transfer.     

Entropy generation minimization (EGM) to optimize mass flow rate in dual channel cable-in-conduit conductors (CICCs) used for fusion grade magnets

Dual channel cable-in-conduit conductors (CICCs) used in tokamaks such as International Thermonuclear Experimental Reactor (ITER) consist of annular channel packed with superconducting strands and a clear central channel separated by a spiral from the annular channel. Supercritical helium (SHe) operating at 4.5 K and 0.5 MPa is used for forced convective cooling of CICC. Pressure drop is inevitable in the process of forced convective cooling, leading to the development of velocity gradients and temperature gradients. These velocity gradients and thermal gradients result in entropy generation in CICCs.The present work aims at estimating volumetric rate of entropy generation (EG) in dual channel CICC. Subsequently, entropy generation minimization (EGM) technique is used to find optimum mass flow rate at which volumetric rate of EG is minimum. Pumping power and heat transfer corresponding to minimum rate of EG are also calculated. Computational fluid dynamics (CFD) is used as a tool to estimate EG as the analytical solution for turbulent forced convective flows requires inaccurate simplifications. A three dimensional model of dual channel CICC is developed in GAMBIT-2.1 and solved using a compatible solver FLUENT-6.3.26. The annular region of CICC is assumed to be porous and the central channel is assumed as clear region for EG analysis using CFD. The pressure gradients and heat transfer coefficient estimated from the simulations are validated against relevant experimental results available in the literature. The effect of mass flow rate on volumetric rate of EG in turbulent forced convective flow is studied using CFD.     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

Sensitivity study on depressurized LOFC accidents with failure of RCCS in a modular gas-cooled reactor

A modular gas-cooled reactor design with a thermal output of 600 MWt and a core exit temperature of 950 °C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multicomponent problems was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600 °C at about 50 h at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.     

Predictive study of the incipient point of net vapor generation in low-flow subcooled boiling

A predictive model of the initial point of net vapor generation, incipient point of net vapor generation (IPNVG) for low-flow subcooled boiling is developed in this paper. The IPNVG established in this model meets both the thermodynamic and hydrodynamic conditions. The thermodynamic condition is described by the heat balance at IPNVG. The amount of heat for steam generation is equal to that for bubble condensation at IPNVG. The force balance of the detached bubbles at IPNVG or the hydrodynamic condition is established to provide the diameter of the detached bubbles and the interfacial heat transfer coefficient for the calculation of heat transfer at IPNVG. This mechanism of the present model makes it applicable to low-flow subcooled boiling. Several coefficients involved in the proposed model are identified by Freon-12 experimental data. This model is compared with the experimental data obtained from different works. The data cover two working media of steam-water and Freon-12, two flow conditions of natural and forced circulations and relatively wide ranges of pressure, mass flux and heat flux. The predictions of this model agree with the data quite well.     

An experimental study on evaporative heat transfer coefficient and applications for passive cooling of AP600 steel containment

After TMI and Chernobyl accidents, many efforts have been made to enhance the nuclear safety with passive features. Among such passive features, the passive containment cooling system (PCCS) has been suggested by Westinghouse in the AP600 plant. The containment with PCCS is a dual containment, and consists of a stainless steel vessel and a concrete wall. In the gap between these structures, air and water can counter-currently pass and cool the steel surface. This paper experimentally investigates evaporative heat and mass transfer at the surface of a falling water film with counter-current air flow in a vertical duct with one-side heated plate. Experiments included various conditions of mass flow rate of film and air. Experimental results show the strong effects of water temperature and air mass flow rate, but little effect of the water flow rate. Also, simple analyses based on heat and mass transfer analogy were performed to evaluate the experimental results. With experimental data, a new correlation on evaporative mass transfer coefficient was developed, and with the correlation, the containment pressure and temperature was calculated for the design basis accident of AP600 by the use of CONTEMPT4/MOD5 code implementation.     

一维开式自然循环系统分析程序的实验验证

基于华龙一号非能动安全壳热量导出系统（PCS）综合性能实验装置实验结果,对采用基于漂移流模型开发的华龙一号PCS程序（PCS?NCCP）进行验证,对比分析了设计工况及非设计工况下PCS?NCCP程序计算值与实验值之间的误差。结果显示,所开发的PCS?NCCP程序能模拟PCS的排热能力、稳态运行特性和动态响应特性,程序计算值能很好地跟踪实验的趋势和幅值变化,绝大部分计算误差落在±20%范围内,验证了PCS?NCCP程序的准确性。     

Modifications to COBRA-TF to model dispersed flow film boiling with two flow,four field Eulerian–Eulerian approach – Part 1

In the case of a postulated loss of coolant accident (LOCA) in a nuclear reactor, an accurate prediction of clad temperature is needed to determine the safety margins. During the reflood phase of the LOCA, when the local void fraction is greater than 80% with the wall temperature above minimum film boiling temperature (T_min), the heat transfer process is dispersed flow film boiling (DFFB). This study has been performed to model DFFB in the reflood phase of a LOCA in a pressurized water reactor (PWR) rod bundle. The COBRA-TF computer code is utilized, since it has a detailed reflood package which takes into account the effect of spacer grids on the local heat transfer. The COBRA-TF code has also been improved to include a four field Eulerian–Eulerian modeling for the two-phase dispersed flow film boiling heat transfer regime. The modifications include adding a small droplet field to COBRA-TF as the fourth field. In addition, the spacer grid models of COBRA-TF have been revised and modified. In the first part of the paper, the results of the code predictions are presented by comparing the experimental data from rod bundle heat transfer (RBHT) experiments with the results of code simulations performed with original and modified code. Measurements and calculations for the heater rod, vapor temperatures and quench front progression have been compared and the results are described in detail. The results of the analysis performed with the modified code indicate the improvement in code predictions for the rod surface temperature, vapor temperature and quench front behavior. The results also indicate the need for improvement in the entrainment and interfacial drag models for the drop fields. The effects of spacer grids on the heat transfer, the models improved and developed for spacer grids and the results of the code calculations with these models are described in the part 2 of the paper.     

An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.     

实验模拟非能动安全壳冷却系统中因破口喷射诱发的传热增长(英文)

介绍在自然对流中由强迫喷射所造成的传热增长。实验研究了在具有垂直冷却面的大型矩形腔体中因强迫射流造成的混合对流传热。在模拟实际非能动安全壳冷却系统及接近实际安全壳分隔区域尺寸的条件下,测量了控制强迫射流传热的关键参数,研究了包括喷射直径、喷射方向、内部构件和腔体比例在内的几何因子的影响。本实验包括了多种射流模式,有助于揭示新一代固有安全型反应堆在事故条件下内部的混合与分层现象。通过控制方程的相似律分析,可预言混合对流传热由阿基米德数和几何因子控制。利用混合对流传热的组合律及飘浮型和碰撞型射流的数学模型,推导出了传热增长关系式,并经过了实验数据测试。通过对实验结果的分析阐明了喷射直径、喷射方向、内部构件和腔体比例对传热增长的影响。