Lower Plenum Research Articles

305 並列細管流路内の気液二相流に関する研究(OS9 オーガナイズドセッション《熱・流体・エネルギー変換工学の進展》)

The aim of this study is to investigate that the diameter and the number of the flow channel, the position is what effect the pressure in lower plenum and quantity of liquid falling by using multiple flow channels, the smaller tube diameter. As a experiment condition, equipment consists of ten 1mm diameter narrow channels. We investigated pressure in lower plenum and quantity of liquid falling through an experiment and calculation. As a result, the following facts became clear. Pressure in lower plenum and quantity of liquid falling were decided by the number of glass tubes.

20813 BWR複雑流路内における溶融燃料落下挙動(086-4【熱流体工学のフロンティア(4)】)

It is important to estimate the behavior of molten core jet in RPV (Reactor Pressure Vessel) of BWR when severe accident occurred as can be seen from Fukushima Danchi nuclear power plant accident. Thus we are developing a simulation code to estimate the behavior of molten core falling in a lower plenum of BWR. Since there are many complicated structures in the plenum, it is expected that molten core would be injected in the complicated flow channel. The objective of the present study is to investigate the influence of complicated structures m a lower plenum on jet by experiment. To investigate it, we observe jet injection which simulated the molten core falling accident experimentally. We constructed a model of a lower plenum and a steady jet injection equipment. Results of preliminary experiment showed that jet behavior is affected by complicated flow structure

The Effective Convectivity Model for Simulation of Molten Metal Layer Heat Transfer in a Boiling Water Reactor Lower Head

This paper is concerned with the development of approaches for assessment of core debris heat transfer and Control Rod Guide Tube (CRGT) cooling effectiveness in case of a Boiling Water Reactor (BWR) severe accident. We consider a hypothetical scenario with stratified (metal layer atop) melt pool in the lower plenum. Effective Convectivity Model (ECM) and Phase-Change ECM (PECM) are developed for the modeling of molten metal layer heat transfer. The PECM model takes into account reduced convection heat transfer in mushy zone and compositional convection that enables simulations of noneutectic binary mixture solidification and melting. The ECM and PECM are (i) validated against relevant experiments for both eutectic and noneutectic mixtures and (ii) benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is then applied to the analysis of heat transfer in a stratified heterogeneous debris pool taking into account CRGT cooling. The PECM simulation results show apparent efficacy of the CRGT cooling which can be utilized as Severe Accident Management (SAM) measure to protect the vessel wall from focusing effect caused by metallic layer.

Isothermal Air-Ingress Validation Experiments

Idaho National Laboratory has conducted air-ingress experiments as part of a campaign to validate computational fluid dynamics (CFD) calculations for very high-temperature gas-cooled reactor (VHTR) analysis. An isothermal test loop was designed to recreate exchange or stratified flow that occurs in the lower plenum of VHTR after a break in the primary loop allows helium to leak out and reactor building air to enter the reactor core. The experiment was designed to measure stratified flow in the inlet pipe connecting to the lower plenum of the General Atomics gas turbine-modular helium reactor (GT-MHR). Instead of helium and air, brine and sucrose were used as heavy fluids, and water was used as the lighter fluid to create, using scaling laws, the appropriate flow characteristics of the lower plenum immediately after depressurization. These results clearly indicate that stratified flow is established even for very small density differences.Corresponding CFD results were validated with the experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations. The calculated current speed showed very good agreement with the experimental data, indicating that current CFD methods are suitable for simulating density gradient stratified flow phenomena in an air-ingress accident.

LES of the Flow Inside the Lower Plenum of an Advanced Gas-Cooled Reactor with Conjugate Heat Transfer

The details of flow and conjugate heat transfer in Advanced Gas-cooled Reactors are currently not completely understood because of the reactor complexity and the difficulty to examine the flow while the reactors are in operation. This paper proposes to simulate the flow using Large Eddy Simulation for the turbulence and to compute the conjugate heat transfer by coupling a multi-purpose CFD software with a transient thermal software dedicated to complex solid geometries. Two configurations are used, 1/8th of the full reactor and the full reactor itself. It is shown that simulating 1/8th of the domain only does not allow to reproduce the agitated patterns observed near the centre of the under-dome, where the temperature is the highest.

Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC

Abstract The research field focussing on the investigations and the analyses of postulated severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises of the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. This paper is focused on the possibilities for accident management measures in case of severe accidents. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accidents, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed.

Conceptual study on air ingress mitigation for VHTRs

An air ingress accident following a postulated pipe break is considered a critical event for a very high temperature gas-cooled reactor (VHTR) safety. Following helium depressurization, it is anticipated that air will enter the core through the break leading to oxidation of the in-core graphite structures. Under extreme circumstances and without mitigation features this accident may lead to exothermic chemical reactions between graphite and oxygen depending on the accident scenario and the design. Under extreme circumstances (beyond design basis), a loss of structural integrity may occur in some core structures and lead to elevated release of radiological inventory for the fuel matrix.This paper discusses various air ingress mitigation concepts applicable for the VHTRs that would prevent core damage even in the most extreme scenarios. The study begins with identifying important factors (or phenomena) associated with the air ingress accident using root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air ingress mitigation ideas were conceived and developed.Among them, two concepts were finally evaluated as effective candidates. One concept is to inject helium directly into the lower plenum (direct in-vessel injection); the other concept is to enclose the reactor with a non-pressure boundary with an opening at the bottom (ex-vessel enclosure). Computational fluid dynamics (CFD) methods were used to evaluate these concepts for proof of these principles. Results indicate that both concepts can effectively suppress air ingress. In the first concept, the buoyancy of the injected helium replaces the air in the core and the upper part of lower plenum. This prevents air from moving into the reactor core and shows the most potential for mitigating graphite oxidation within the vessel. In the second concept, air ingress is limited by molecular diffusion through the opening at the enclosure bottom, a very slow process that allows sufficient time for the core to cool. Existing VHTR concepts can be modified to accommodate either of these mitigation concepts.

CORIUM BEHAVIOR IN THE LOWER PLENUM OF THE REACTOR VESSEL UNDER IVR-ERVC CONDITION: TECHNICAL ISSUES

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

Cooling capacity of plate type research reactors during the natural convective cooling mode

Numerical investigations on the cooling capacity of plate type research reactors during the natural convective cooling mode are fulfilled in the present study. The cooling capacity, as a thermal core design criterion, is determined by the onset of nucleate boiling (ONB) temperature margin. A simplified model, consisting of the lower plenum, the core, the chimney, the flap valve, and the pool, is simulated by RELAP5/MOD3 and NATCON. The axial and radial peaking factors for the power distribution are applied to investigate the cooling capacity on the hot spot in a fuel assembly. Several convective heat transfer correlations developed by Dittus and Boelter (1930), Petukhov (1970), and Sudo et al. (1985) are implemented into the simulations; then the coolant and surface temperatures and ONB temperature margin, as a function of core power, are obtained from the simulations. The comparisons between RELAP5/MOD3 and NATCON simulations are in good agreement, and the lowest cooling capacity, which limits the permissible core power during the natural convection cooling, is found with Sudo heat transfer correlation.

Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

1905 ABWR下部プレナム内流路における複雑流動構造(OS7-1 流れの計測と制御I,OS7 流れの計測と制御)

In order to achieve increase of power outputs of an ABWR, it is extremely important to evaluate coolant flow in a lower plenum. Numerical simulation is helpful to predict the coolant flow. However the simulation results considerably depends on models, thus verification experiment is needed. Hence, the objective of the present study is the construction of the benchmark of CFD code for clarification the flow structure of a lower plenum. We constructed a 1/10 model of a lower plenum to conduct flow experiment for validation of CFD analysis. In the experiment, it turned out that coolant flow distribution becomes to be uniform at core support beam, and there are complicated flows around side entry orifices. These flow structures are also described by CFD analysis. As good performance of analysis, we investigated flow distribution to the core particularly by using CFD results, and mechanisms of these flows were discussed.

Experimental Investigation of Rising Gas Bubble Characteristics from a Vertical Tube under CCFL Condition

This paper describes an experimental study of gas/liquid countercurrent flow in a vertical circular tube. CCFL experiments were carried out with three different water levels in the upper plenum, two different tube diameters. Measurements were made for liquid and gas flow rates, time variations of pressure at locations of the upper entry of the tube and lower plenum. Visual observations were also conducted to investigate the relationship between rising gas bubble characteristics and time variation of gas pressure at the upper entry of the tube. The results indicate that one bubble formation cycle (e.g., bubble growth, expansion, and detachment into the water pool) corresponds to one pressure fluctuation cycle. For the 20 mm diameter tube, it was confirmed that there was a characteristic waiting time between bubble cycles in which no bubble was formed at the upper entry of the tube. The waiting time is a favorable time for a liquid introduction into the tube from the upper plenum. The bubble volumes are compared with existing bubble formation correlations.

Counter Current Flow Limitation of Gas-Liquid Two-Phase Flow in Nearly Horizontal Pipe

Experimental work about counter current two-phase flow of air and gas in nearly horizontal pipe has been performed. The work was performed in a 1.1 m long circular transparent acrylic pipe with 50 mm inner diameter, in two inclination angle settings (20° and 10° from horizontal). The smooth liquid and air inlet was used. Porous liquid inlet and a nozzle connected with calm section were used as liquid and gas inlet. The effect of liquid properties is examined by using five different working fluids (Water, two different concentration of butanol and glycerin aqua solutions). As for results. (1) CCFL causes a drastic change in the delivered liquid to the lower plenum. (2) The effect of inclination angle is significantly observed. The flooding gas superficial velocity decreases with inclination angle. (3) The liquid viscosity affects the flooding phenomena.

Examples of unsteady CFD validation system response quantities in a cylinder array

A validation study for several CFD models of the time-varying flow through a confined bank of cylinders is presented. The geometry is cylinders arranged on equilateral triangles with pitch to diameter ratio of 1.7 to represent a scaled subsection of the lower plenum of a high temperature reactor. Time-resolved Particle Image Velocimetry (PIV) measurements, coupled with time-varying pressure measurements along the facility walls, are compared to the Unsteady Reynolds-Averaged Navier–Stokes (URANS) k − ω model and two variations of a Detached Eddy Simulation (DES) model. Spatial and temporal validation system response quantities (SRQs) on both the local and global scales were used for validation. The DES model accurately predicted frequencies present in the pressure along the walls next to the cylinders in the first and the last cylinder, yet predicted other dominant frequencies in the remaining cylinders that were not found in the experiment. As expected, the temporal behavior of the DES was generally far superior to that of the URANS model. A grid convergence study shows typical global quantities (such as pressure losses) converge well while temporal quantities converge poorly for the same grids.

Large eddy simulations of jet flow interaction within staggered rod bundles

This work is a numerical investigation of experimental work on dual isothermal jet injection into a staggered rod bundle, mimicking phenomena present in the lower plenum of the very high temperature gas cooled reactor. Large eddy simulations were used to model the unsteady flow of the experimental conditions. As a basis, steady state simulations as well as time averaged comparisons were made with the standard and realizable k–ɛ formulations. Grid sensitivity studies were performed over six grids with the most refined grid comprised of roughly 26.5 million semi-structured polyhedral cells. The continuous wavelet transform was used for instantaneous signal decomposition of results taken within the domain in an effort to devise a better approach to validating CFD predictions of the complex flows in the lower plenum of the VHTR.

CFD simulations of separate effects in an HTGR lower plenum under air ingress condition

The local heat transfer in an HTGR lower plenum is investigated at the Czech Technical University (CTU) in Prague with collaboration with the U.S. Nuclear Regulatory Commission (NRC). NRC has initiated efforts to build a technical infrastructure necessary to support licensing activities for HTGR's. These efforts include the development and evaluation of computational fluid dynamics (CFD) tools to analyze the system during various conditions. In this case air ingress condition is investigated. The local heat transfer under air ingress condition is strongly affected by the physical phenomena such as molecular diffusion and natural circulation. These separate effects are not yet well understood. Therefore, in the first phase of research it is necessary to perform separate-effects simulations. The CFD solver is exercised and the results are compared with experimental data. The separate-effect benchmarking of the CFD code includes simulations of two separate phases of the JAERI experiments: isothermal molecular diffusion of a binary gas mixture and non-isothermal diffusion and natural circulation of a binary gas mixture. Computational model of experimental apparatus consists of a vertical inverted U-tube connected at the bottom to a cylindrical tank. Commercial CFD solver Fluent 6.3 was used for the simulations. The influence of the size of computational mesh and length of time step of unsteady solver was studied. For the simulation of isothermal molecular diffusion the coarser mesh with about 9000 hexahedral cells was suitable. The length of time step about 0.05s seems to be optimal. The non-isothermal diffusion and natural circulation simulations were made and compared with experimental data. The onset time of the natural circulation was simulated in good agreement with experiment with small deviation of 2.3%.

Natural circulation patterns in the VHTR air-ingress accident and related issues

Abstract Natural circulation patterns in the VHTR during a hypothetical air-ingress accident have been investigated using computational fluid dynamic (CFD) methods in order to compare results from the previous 1-D model which was developed using GAMMA code for the air-ingress analyses. The GT-MHR 600 MWt reactor was selected to be the reference design and modeled by a half symmetric 3-D geometry using FLUENT 6.3, a commercial CFD code. CFD simulations were carried out as the steady-state calculation, and the boundary conditions were either assumed or provided from the 1-D GAMMA code results. Totally, 12 different cases have been reviewed, and many notable results have been obtained through this work. According to the simulations, natural circulation patterns in the reactor were quite different from the previous 1-D assumptions. A large re-circulation flow with thermal stratification phenomena was clearly observed in the hot-leg and the lower plenum in the 3-D model. This re-circulation flow provided about an order faster air-ingress speed (0.46 m/s in superficial velocity) than previously predicted by 1-D modeling (0.02–0.03 m/s). It indicates that the 1-D air-ingress modeling may significantly distort the air-ingress scenario and consequences. In addition, complicated natural circulation patterns are eventually expected to result in very complex graphite oxidations and corrosion behaviors.

Experimental results on the coolability of a debris bed with down comer configurations

Abstract In case of a severe accident, a debris bed is formed from a mixture of molten core and the residual water in the lower plenum of the reactor pressure vessel. The presence of decay heat in a debris bed poses a critical threat to the reactor pressure vessel. To avoid any damage to the reactor pressure vessel, the removal of decay heat from the debris bed is of great importance. In order to investigate experimentally the long-term coolability of debris beds, the non-nuclear test facility “DEBRIS” has been established at IKE. Experimental investigations of coolability limits for such a debris bed, based on two different bed configurations and at various thermo-hydraulic conditions, are carried out at IKE. This paper presents the experimental results for multidimensional cooling effects on boiling and dryout tests with different bed configurations and different system pressures. Two different down comer configurations (cylindrical and perforated cylindrical tubes with inner diameters of 10 mm each) are used to investigate the multidimensional cooling effects. The down comer is concentrically installed inside the debris bed which is contained in a cylindrical crucible with an inner diameter of 125 mm. Different bed configurations, e.g. polydispersed particle bed with spherical particles 2, 3 and 6 mm in diameters and irregular particles of equivalent diameters 2–10 mm, have been used with a bed height of 640 mm. A layer of ceramic balls of diameter 4 mm with 40% porosity is used to make up a water pool at the bottom of the bed. The bottom inflow via the down comer tube as well as the lateral inflow of water through the perforated down comer tube into the bed improves the coolability of the debris bed, and therefore an increase of the dryout heat flux can be observed. Experimental results also show that the system pressure has no significant effect on the characteristics of pressure gradients inside the bed, whereas with increasing system pressure the coolability limits are increased.

Dynamic characteristics of a perforated cylindrical shell for flow distribution in SMART

The System-integrated Modular Advanced ReacTor (SMART) is a small nuclear reactor under development in Korea. It is equipped with a perforated cylindrical shell, which is called a flow skirt, in the lower plenum of the reactor for uniform flow distribution and to prevent inflow of debris into the core. This perforated cylindrical shell can be excited by external forces such as seismic or pump pulsation loads. The dynamic characteristics of the perforated cylindrical shell must be identified for further dynamic analysis. This research explores the modal analysis of the scaled-down flow skirt model submerged in coolant water. For the numerical simulation, finite element analysis is carried out to extract modal characteristics of the structure considering the fluid–structure interaction and we introduce the NAVMI factor for similarity analysis. In the finite element model, the whole shape of the perforated cylindrical shell is simulated instead of using the effective material properties. In addition, a 1/12 scaled-down flow skirt is manufactured, and an experiment is designed using an exciter and waterproof accelerometers for the modal test. Due to excellent agreement between the modal test results and the finite element analysis results such as natural frequencies and mode shapes, the finite element model is validated and can be used to predict the dynamic characteristics of the real flow skirt. Moreover, the natural frequency of the real flow skirt can be calculated from the NAVMI factor and is in good agreement with the FEM result.

Buoyancy‐driven Mixing Studies of Natural Circulation Flows using Rossendorf Coolant Mixing Model Experiments and CFD

AbstractCoolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the deborated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5‐scaled Rossendorf Coolant Mixing Model (ROCOM) mixing test facility. Thermalhydraulics analyses showed that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re‐establish simultaneously and the deborated slugs are shifted towards the reactor pressure vessel. In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 and 2 % density differences between the deborated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.