Simulation of containment jet flows including condensation

The validation of a CFD code for light-water reactor containment applications requires among others the presence of steam in the different flow types like jets or buoyant plumes and leads to the need to simulate condensation phenomena.In this context the paper addresses the simulation of two “HYJET” experiments from the former Battelle Model Containment by the CFD code CFX. These experiments involve jet releases into the multi-compartment geometry of the test facility accompanied by condensation of steam at walls and in the bulk gas. In both experiments mixtures of helium and steam are injected. Helium is used to simulate hydrogen. One experiment represents a fast jet whereas in the second test a slow release of helium and steam is investigated. CFX was earlier extended by bulk and wall condensation models and is able to model all relevant phenomena observed during the experiments. The paper focuses on the simulation of the two experiments employing an identical model set-up. This provides together with other validation exercises the information on how well a wider range of flowing conditions in a full containment simulation can be covered with a single set of models (e.g. turbulence and condensation model). Some aspects related to numerical and modelling uncertainties of CFD calculations are included in the paper by investigating different turbulence models together with the modelling errors of the differencing schemes applied.     

Three-gas mixture plume inducing mixing and stratification in a multi-compartment containment

A project named SETH (SESAR Thermal Hydraulics (SESAR: Senior Group of Experts on Nuclear Safety Research)) has been performed under the auspices of 15 OECD countries, with the aim of creating an experimental database suitable to assess the 3D code capabilities in analyzing physical key phenomena relevant for LWR safety analysis. This paper presents the results of OECD/SETH Project Test 25 performed in the PANDA facility (located at PSI in Switzerland). Test 25 consists of two phases for duration of approximately 2 h for each phase. During Phase 1 a steam-helium mixture (helium is used to simulate hydrogen) is released in a containment compartment initially filled with only air. This phase simulates the hydrogen released in a postulated accident due to the fuel cladding metal-water (MW) reaction. During Phase 2 only steam is released in the containment and the scenario that occurs after the MW reaction is exhausted is addressed. Temperature and gas concentration measurements obtained at several locations in the containment compartments allow the recognition of the complex stratification pattern evolution during the test period. The analysis of Test 25 carried out with the GOTHIC code is also presented in this paper. The analysis shows that the prediction of three-gas (hydrogen, steam, and air) stratification pattern in a multi-compartment geometry for scenarios characterized by the evolution of the density difference between and inside compartments and in particular of the hydrogen accumulation in a dead-end volume is a challenging task also for a code having 3D capabilities.     

Simulation of containment cooling with outside spray after a core meltdown

In the most severe hypothetical loss of coolant accident, the reactor core melts and falls into the containment sump, there evaporating much of the sump water and raising the pressure within the containment building. One possible method to remove the decay heat is to cool the steel containment shell with an outside spray system. To perform the structural analysis needed to confirm the integrity of the containment, the thermal profile in the containment wall must first be found. The purpose of this work is to develop a computer code to calculate this transient profile. Other aspects such as hydrogen build-up are not considered in this code.The method uses relationships for the natural convection-partial condensation phenomena occurring at the containment internal surfaces, iteratively coupled to a one-dimensional heat balance at the wall to solve for the wall temperature as a function of angular position. A differential calculation as a function of time treats the thermodynamic changes within the containment as quasi-steady state. The result is a quick-running code capable of analyzing the temperature transient for several hours following the LOCA with a few minutes of computing time.     

Validation of the CFX4 CFD code for containment thermal-hydraulics

In order to determine the risk associated with the presence of hydrogen in a nuclear power plant containment during a hypothetical severe accident, thermal hydraulic codes are used. Amongst other codes, NRG uses the commercial computational fluid dynamics code CFX4 for this purpose. Models to describe condensation have been implemented by user coding. This paper describes these models. In addition, an overview is given of validation activities with the CFX4 model. Experimental results from the following sources have been used: the Kuhn condensation model; the PHEBUS test facility; the PANDA test facility; and the TOSQAN, MISTRA, and THAI test facilities within the OECD International Standard Problem 47. The CFX4 model predictions are fairly good. Deviations originate primarily from the applied wall treatment. Several recommendations for further development are therefore proposed.     

LBE-water interaction in LIFUS 5 facility under different operating conditions

In the frame of the IP-EUROTRANS Project, an experimental program, focused on studying the LBE/water interaction has been performed using the LIFUS 5 facility available at ENEA-Brasimone. The physical effects and the possible consequences of this interaction have been evaluated over a wide range of different conditions. Besides the experimental activities, a numerical simulation activity has been performed with SIMMER code in order to better investigate the thermo-hydraulic phenomena involved in the interaction and to confirm the capabilities of the code to simulate this kind of phenomena. The experimental and the calculated results in terms of pressure and temperature evolutions in the system show a good agreement.     

应用GOTHIC8.0程序模拟非能动安全壳冷却系统冷凝和蒸发现象的适用性研究

在非能动安全壳冷却系统（PCS）设计基准事故的排热过程中,安全壳内壁面蒸汽冷凝现象和安全壳外壁面水膜蒸发现象是两种非常关键的排热途径。本文应用GOTHIC8.0程序模拟了安全壳内壁面蒸汽冷凝和安全壳外壁面水膜蒸发传热过程,并通过蒸汽冷凝试验和水膜蒸发试验数据,对GOTHIC程序的模拟结果进行了分析和评价。研究结果表明：GOTHIC程序的蒸汽冷凝模型可较好地模拟蒸汽冷凝传热现象;水膜蒸发模型明显低估了水膜蒸发换热量,这对设计基准事故安全壳完整性分析是非常保守的,建议对GOTHIC程序进行适当开发,更好地模拟水膜蒸发换热过程。     

A model for tritium transport in fusion reactor components: The FUS-TPC code

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead–lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO.     

Experimental Investigation of the Steam Condensation Phenomena due to a Multi-hole Sparger

KAERI has performed a series of steam condensation tests to assess the performance of a unit cell sparger that will be used in the APR1400 reactor. A simplified I-sparger was used for the steady state steam condensation tests to study the characteristics of the condensation phenomena due to a multi-hole sparger and to provide test data for a code development and verification. A range of steam mass fluxes for the steady state condensation tests were selected to define the transition region from the condensation oscillation regime to the stable condensation regime. Condensation loads and a variation of the frequencies of the pressure waves due to a steam condensation are analyzed. In addition, the local temperature distribution near the sparger discharge holes is discussed and a condensation regime map for a multi-hole sparger has been suggested.     

Two new passive safety systems for LWR applications

In a research activity that SIET has been conducting for years about safety systems for light water reactors (LWRs), attention has been paid to developing two passive injection systems representing an innovative solution in mitigating the consequences of loss of coolant accidents. Both systems allow the completely passive injection of cold water into a pressurised vessel. They are triggered by a low-level signal and work on the base of phenomena like natural circulation and condensation. The simplest system, Sistema Iniezione Passiva 1 (SIP-1), injects water contained in a tank into a circuit at the same pressure as the circuit. The most complex system, injection cyclic system (ICS), injects cold water, by filling cyclically a proper tank with the water stored in an atmospheric pressure pool. Thanks to the ENEA sponsorship, this activity has been conducted in three steps: the definition of the conceptual design of the systems; the application of the Relap5 code to simulate their behaviour; and the proposal of their specific applications to pressurised and boiling LWR. In this paper, both systems are presented in their structural and operating characteristics together with the main results of the code application for their simulation. Some proposals of application of SIP-1 and ICS to pressurised water reactors and boiling water reactors are also shown. The developments and reached goals of the prosecution of the research are also summarised here, together with future needs.     

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.     

Experimental verification of the fast reactor safety analysis code SIMMER-III for transient bubble behavior with condensation

Experimental verification of a reactor safety analysis code, SIMMER-III, was undertaken for transient behaviors of large-scale bubbles with condensation. The present study aimed to verify the code for numerical simulations of relatively short-time-scale multi-phase, multi-component hydraulic problems. Among these, vaporization and condensation, or simultaneous heat and mass transfer, play important roles. In this study, a series of transient bubble behavior experiments dedicated to condensation phenomena with noncondensable gases was carried out. In the experiments, a pressurized mixture of noncondensable gas and steam was discharged as a large-scale single bubble into a cylindrical pool filled with stagnant subcooled water. The concentration of noncondensable gas was taken as an experimental parameter as was the species of noncondensable gas. The characteristics of transient behavior of large-scale bubbles with condensation observed in the experiments were estimated through experimental analyses using SIMMER-III. In the experiments with steam condensation, dispersion of the gas mixture discharged into the liquid pool was accompanied by vapor condensation at the bubble surface. SIMMER-III simulations suggested that the noncondensable gas had a less inhibiting effect on the condensation of large-scale bubbles. This is a different characteristic to that of the quasi-steady condensation of small-scale bubbles observed in our previous experiments.     

Numerical Analysis of the PHEBUS Containment Thermal Hydraulics

This paper describes numerical analysis of the PHEBUS FP containment thermal-hydraulics. PHEBUS FP is an international project undertaken with the aim of evaluating the behavior of radioactive fission products released from a LWR pressure vessel into the containment vessel during a hypothetical severe accident. Six integral in-pile tests have been planned and are being carried out at Cadarache, France. The European Union, the United States, Canada, Korea and Japan are participating in this project. From Japan, the Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute are collaborating the other parties involved in the project.

Since the behavior of fission products is strongly dependent on the surrounding environmental conditions, accurate prediction of the thermal-hydraulics in the containment vessel is essential to accurately evaluate the behavior. Characteristics of condensation heat transfer in the presence of noncondensable gases play a key role in the PHEBUS thermal-hydraulics, especially under the condition of high noncondensable gas mass fraction. Many models for condensation heat transfer in the presence of noncondensable gases have been proposed. However, these models were not found suitable for PHEBUS analysis, because they were focused on the low noncondensable gas mass fraction condition.

In this study, a single-phase multi-component code, TFLOW-FP has been newly developed to predict thermal-hydraulics in the PHEBUS FP containment. Moreover, a new degradation factor correlation for the condensation heat transfer coefficient due to the presence of noncondensable gases has also been developed and incorporated into the code. This code was applied for analysis of the thermal-hydraulic benchmark tests and the first in-pile test, FPTO. The results show that the code can predict the total pressure, gas temperature distributions, the relative humidity in the containment vessel and steam condensation rate on the surface of condenser rods very well.     

Improvement of a physical model for blockage formation of solid–liquid mixture flow with freezing for core safety evaluation of SFRs

The SIMMER code has been developed to analyze event progression during core disruptive accidents (CDAs) in sodium-cooled fast reactors. One of the key phenomena during CDAs is the discharge of molten fuel from the core region which reduces the reactivity effectively. The discharge flow is inhibited by blockage formation due to freezing of the molten fuel. Then, the blockage formation is enhanced by unmolten fuel which forms solid–liquid mixture flow with the molten fuel. A physical model for blockage formation of solid–liquid mixture flow with freezing in the SIMMER code is improved in this study to dissolve some inconsistencies between the modeling and the physical phenomena involved in the solid–liquid mixture flow with freezing for more precise evaluation of CDA. The improved model is validated with a systematical procedure through a benchmark analysis of an experiment. Consequently, experimental penetration behaviors are simulated reasonably by the SIMMER code analysis with the improved model while excessive blockage formation occurred in the analysis with the original model.     

Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.     

Analytical thermal hydraulic model for oscillatory condensation of steam in presence of air

An analytical thermal hydraulic model has been developed from fundamental conservation laws, for the process of oscillatory condensation of steam in a pool of water, in presence of non-condensable gas (air). The oscillatory condensation phenomena addressed here is steam chugging, with an emphasis laid on studying the effect of small amount of air present in steam, on the phenomena. The objective of developing the model is to present an approximation of the real phenomena and to obtain an analytical solution. At the outset, a parametric study was conducted by using the developed model to capture and identify the salient features of steam chugging and compare the wave shapes obtained with those available in open literature. Subsequently, the effect of presence of air in steam was studied in detail using the non-condensable gas model. An attempt has been made to show numerically that the presence of a small amount of air in steam would effectively stabilize condensation and prevent inception of chugging. Typical results are presented in this paper to bring out the difference in oscillatory behavior due to presence and absence of non-condensable gas.     

Implementation of a generalized diffusion layer model for condensation into MELCOR

Condensation of co-current steam-noncondensable gas mixtures in vertical tubes is an important, yet difficult to model, component of many passive nuclear reactor cooling systems. The stagnant film model, which is used by the severe accident code MELCOR, gains its name by assuming that the gas-vapor film formed along the condensation surface is stagnant. Liao developed a generalized diffusion layer model that removes limitations of the stagnant film model and considers additional phenomena to improve predictive capabilities for condensation heat transfer with noncondensable gases. Similarities between the formulations of the stagnant film model and generalized diffusion layer model allow for the generalized diffusion layer model to be implemented into MELCOR. Input decks representing experimental facilities that produced co-current condensation data have been created to analyze and validate the generalized diffusion layer model implemented in MELCOR. The experimental data span a wide range of noncondensable gas mass fractions and include condensation mass transfer both on a vertical flat plate and in vertical tubes. MELCOR predictions of the condensation mass flux are seen to improve when using the generalized diffusion layer model instead of the stagnant film model.     

Overview of SAFER03 Assessment Studies

The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.     

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.     

CFD simulation of steam jet-induced thermal mixing in subcooled water pool

This study examined the IRWST thermal mixing phenomena induced by a steam jet in a subcooled water pool. Due to the limitation of the current CFD code to simulate condensation, the steam condensation region model was developed to evaluate the thermal mixing phenomena. Within this region, all the steam was condensed into water, and the steam mass and energy inputs were treated as the source. This calculation was treated using single-phase CFD methods. The benchmark calculation for a thermal mixing experiment in the water tank was performed to develop an optimized 3D evaluation methodology of the thermal-hydraulic behavior in APR1400 IRWST. Steam discharge through the sparger and condensation phenomenon was modeled with the choking flow and thermal mixing model in the quenching tank using CFX11.Three types of thermal mixing experiments, local phenomena test, thermal mixing tests in cylindrical water pool and annulus water pool, were designed to provide data representative of the behavior of the prototype for CFD simulations of the thermal-hydraulic behavior in IRWST. A comparison of the calculated and experimentally measured temperature profiles showed some disagreement particularly around the sparger. The main reason for this disagreement was caused by the difference in the test and simulating conditions at the tank wall. However, moving away from the sparger, the trends of the temperature rise became similar to that in the experiment. Despite these problems, this model is the best way of evaluating the thermal mixing phenomena caused by a steam jet in a subcooled water pool.     

Analysis of hydrogen risk mitigation with passive autocatalytic recombiner system in CPR1000 NPP during a hypothetical station blackout

Hydrogen safety has attracted extensive concern in severe accident analysis especially after the Fukushima accident. In this study, a similar station blackout as happened in Fukushima accident is simulated for CPR1000 nuclear power plant (NPP) model, with the computational fluid dynamic code GASFLOW. The hydrogen risk is analyzed with the assessment of efficiency of passive autocatalytic recombiner (PAR) system. The numerical results show that the CPR1000 containment may be damaged by global flame acceleration (FA) and local detonation caused by hydrogen combustion if no hydrogen mitigation system (HMS) is applied. A new condensation model is developed and validated in this study for the consideration of natural circulation flow pattern and presence of non-condensable gases. The new condensation model is more conservative in hydrogen risk evaluation than the current model in some compartments, giving earlier starting time of deflagration to detonation transition (DDT). The results also indicate that the PAR system installed in CPR1000 could prevent the occurrence of the FA and DDT. Therefore, HMS such as PAR system is suggested to be applied in NPPs to avoid the radioactive leak caused by containment failure.