G.H. Su

Analysis of KROTOS Steam Explosion Experiments Using the Improved Fuel-Coolant-Interaction Code TEXAS-VI

Abstract In the present study, we propose a new fragmentation criterion for the explosion phase to take account of the effect of partial fuel melt solidification on the rapid fragmentation process. This new criterion judges whether or not the explosive fragmentation can occur by comparing the impact stress induced by vapor film collapse and water jet impingement with the fracture toughness of the corium crust layer. The fragmentation criterion was incorporated into the revised Thermal EXplosion Analysis Simulation (TEXAS) fuel-coolant-interaction (FCI) model TEXAS-VI and combined with the previously proposed fuel particle solidification model and the fragmentation criterion for the mixing phase. TEXAS-VI was compared to KROTOS alumina test K-44 and corium tests K-52 and K-53, and good agreement was obtained. The simulation results indicate that TEXAS-VI has the capability to consider the effect of partial solidification for both the mixing and the explosion phases of the FCI process and can capture the effect of fuel solidification, which reduces corium-water explosion energetics. Experiments K-52 and K-53 also demonstrate the ability of TEXAS-VI to model the effects of ambient pressure on energetics.

Thermal-Hydraulic Analyses of Transportable Fluoride-Salt-Cooled High-Temperature Reactor with CFD Modeling

Abstract The technology for the 20-MW(thermal) Transportable Fluoride Salt–Cooled High-Temperature Reactor (TFHR) is proposed by Massachusetts Institute of Technology for off-grid applications such as Antarctic bases and remote mining sites. The preliminary thermal-hydraulic analyses and improvements based on a 1/12th full-core model were performed using three-dimensional computational fluid dynamics (CFD). A benchmark study was conducted by comparing the CFD results against empirical correlations and experimental data obtained by Cooke, Silverman, and Grele. In the 1/12th full-core analysis, three practical considerations that may challenge the TFHR temperature limits are evaluated as bounding analysis. These include (1) helium gap between fuel compact and graphite block, (2) thermal conductivity degradations of graphite matrix due to neutron irradiation, and (3) full-core scale power distribution obtained from neutronic calculations. These design considerations lead to insufficient margin between the normal operating condition and the predefined thermal limits. In this context, additional design features are implemented to improve the thermal-hydraulic safety of the TFHR. First, bypass flow in the interstitial gaps between the active core and the reflector is found capable of reducing the temperature peaks at the core periphery. Second, improvements of the flow distribution from the central downcomer to individual coolant channels enable a higher mass flow rate to the regions with compromised cooling access. Overall, thermal-hydraulic performance was significantly improved with a fuel temperature margin from 10 to 150 K and a coolant temperature margin from 16 to 160 K, as well as the more uniform temperature distribution across the reactor core. Furthermore, thermal-hydraulic safety can be maintained at a 20% overpower operating condition [i.e., 24 MW(thermal)]. Overall, this study provides an engineering basis for the TFHR thermal-hydraulic design to improve its safety margin.

Experimental Investigation of Air-Water CCFL in the Pressurizer Surge Line of AP1000

Abstract Countercurrent flow limitation (CCFL) may occur under certain flow conditions in the surge line, restricting the draining of water from the pressurizer and thus affecting the coolant inventory and water level in the reactor pressure vessel (RPV). The complexity of the AP1000 pressurizer surge line structure makes predicting CCFL fairly difficult, and there are still not enough CCFL studies on this complex structure. Based on an extensive literature survey, the authors of this paper are particularly aware of the need for improved CCFL models for the pressurizer surge line of AP1000. To investigate the CCFL phenomenon in the surge line assembly fixture of AP1000, a whole-visual test model of the surge line is designed with a scaling ratio of 1:4, and a test loop is established to carry out visualization experiments with air-water countercurrent flow (CCF). The whole-visual test section made of acrylic material is composed of a pressurizer simulator, a surge line tube, a hot leg T-type tube, and an RPV simulator. The air-water CCF experiments are conducted at atmospheric pressure and room temperature with the pressurizer simulator water level varying from 150 to 900 mm. The visual CCF experimental processes and CCFL phenomena are filmed by a high-speed camera and analyzed in detail. The pressure drops at different CCFL locations are measured and evaluated to explore the relationships between the CCFL characteristics and flow patterns in the surge line. The development process of the CCFL is defined as the CCFL region, which can be divided into different regions according of the changes in water mass flow and CCF flow behavior. The CCFL data are analyzed and compared using the air and water superficial velocities to study the effects of hysteresis and water level. Small discrepancies are found between the data of different water levels, reflecting the small but not-negligible influence of the upper tank water level. Empirical models for the CCFL in the surge line assembly fixture are explored preliminarily using Kutateladze-type correlation and Froude-Ohnesorge correlation. Deficiencies still exist in the present semiempirical models, inspiring a more in-depth study on the empirical models for CCFL in the surge line assembly fixture that considers the complex two-phase flow behaviors in the upper tank and near the joint between the upper tank and surge line tube. The present CCFL data are compared broadly and in detail with groups of CCFL data of similar former experiments to demonstrate the applicability of the present air-water CCFL data to the development of a CCFL prediction model for the prototype large-diameter surge line assembly fixture of the AP600/AP1000. We will perform much more experimental and theoretical work to study the detailed mechanism of these special phenomena and to develop a more applicable CCFL model for the geometry and conditions of the prototype large-diameter surge line assembly fixture.

Transient Safety Analysis of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor Using RELAP5-3D

Abstract A transportable fluoride-salt-cooled high-temperature reactor (TFHR) design with 20-MW(thermal) rated power and 18-month fuel cycle is proposed for off-grid applications. One of the design goals of the compact reactor core is potential transport by truck, rail, or air. Full-core thermal-hydraulic analyses and improvements using three-dimensional computational fluid dynamics (CFD) were performed previously to demonstrate the feasibility of a TFHR design at a nominal power of 20 MW(thermal). In this paper, the best-estimate system code Reactor Excursion Leak Analysis Program (RELAP5-3D) is adopted to study the transient behavior of this TFHR design and the safety characteristics of the primary loop system during accident conditions. The modeling results of the steady state were verified using CFD results with consideration of radial heat conduction between heat transfer unit cells. Four most challenging accidents of anticipated transient without scram were analyzed, as well as parametric studies of some key factors. These accidents include unprotected reactivity insertion accident (URIA), unprotected loss of heat sink (ULOHS), unprotected loss of flow (ULOF), and a combination accident of ULOF and ULOHS. The results indicate that transient temperature limits are not exceeded during the most severe accidents. They indicate satisfactory transient performance of the TFHR design. The transient temperature limit of structure material Hastelloy N, based on embrittlement phenomena, poses the most limiting constraint due to the small temperature margin of about 20 K in the accident combination of ULOF and ULOHS. Overall, TFHR is a sound reactor design from a thermal-hydraulic viewpoint.

Analysis of Flow Instabilities in Forced-Convection Steam Generator

Because of the practical importance of two-phase instabilities, substantial efforts have been made to date to understand the physical phenomena governing such instabilities and to develop computational tools to model the dynamics. The purpose of this paper is to present a numerical model for the analysis of flow-induced instabilities in forced-convection steam generator. The model is based on the assumption of homogenous two-phase flow and thermodynamic equilibrium of the phases. The thermal capacity of the heater wall has been included in the analysis. The model is used to analysis of flow instabilities in steam generator and to study the effects of system pressure, mass flux, inlet temperature and inlet/outlet restriction, gap size, the ratio of do/di and the ratio of qi/qo on system behavior.Copyright

Experimental Study on Dryout Point in Vertical Narrow Annulus Under Low Flow Conditions

Prediction of dryout point is experimentally investigated with deionized water upflowing through narrow annular channel with 1.0 mm and 1.5 mm gap respectively. The annular with narrow gap is bilaterally heated by AC current power supply. The experimental conditions covered a range of pressure from 0.8 to 3.5 MPa, mass flux from 26.6 to 68.8 kg/m2 s and wall heat flux from 5.0 to 50.0 kW/m2 . The location of dryout is obtained by observing a sudden rise in surface temperature, Kutateladze correlations is cited, however proved to be not a proper one for annuli gap and modified to predict the location of dryout. Considering in detail the effects of the geometry of annuli, pressure, mass flux and heat flux on dryout, an empirical correction is finally developed to predict dryout point in narrow annular gap under low flow condition which has a good agreement with experimental data.Copyright

Effect of Rolling Motion on Flow Instability of Parallel Rectangular Channels of Natural Circulation

In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.Copyright

Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method

Abstract In detailed previous work by the authors, an innovative decay heat removal (DHR) system has been proposed and designed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The current paper studies the inadvertent actuation of one DHR system train during I2S-LWR normal operation due to a false signal or operator action. The RELAP5 code is used to perform a one-dimensional study, and important thermal-hydraulic characteristics, including primary loop coolant flow rate, pressure, temperature, DHR primary-side flow rate, and coolant temperature, are achieved during this transient. Then, a detailed computational fluid dynamics simulation utilizing STARCCM+ is carried out to investigate the coolant mixing characteristics in the downcomer and lower plenum and obtain the local thermal-hydraulic conditions at the reactor core inlet. It is found that as a consequence of inadvertent DHR actuation, the maximum overcooling at the reactor core inlet is about 3 K, which would not result in significant reactivity insertion. Furthermore, a more severe transient of inadvertent DHR operation with intermediate loop break is studied, and the results show that this would not lead to more significant overcooling to the I2S-LWR core compared with inadvertent DHR operation without intermediate loop break. This work is an indispensable supplement for DHR system comprehensive assessment in the I2S-LWR project.

Numerical Simulation on Collapse of Vapor Bubble Using Particle Method

In this paper, the collapse of a void bubble filled with vapor content is numerically investigated using a novel moving particle semi-implicit with meshless advection by flow-directional local grid (MPS-MAFL) method. The interfacial velocity, collapse time, bubble shape variation, peak pressure, rebound bubble radius, and other interesting parameters were obtained and are discussed profoundly. The vapor bubble undergoes several cycles of oscillation with reduced amplitude during the whole collapse process, which is similar to cavitation bubble collapse. The computational results show that the bubble collapse time is linearly proportional to the initial bubble size, which agrees with the Rayleigh equation. The minimum rebound bubble radius ratio is less affected by initial bubble size for a large bubble. Comparison work was also conducted against experimental data by Board and Kimpton. The comparison revealed that the MPS method supplied with an adiabatic compression assumption for vapor content is more suitable to evaluate the collapse behaviors of a low-pressure vapor bubble. This work is helpful for further application of the moving particle semi-implicit with meshless advection using flow-directional local grid (MPS-MAFL) method to solving complicated bubble dynamics.

Numerical Research on Oscillation of Two-Phase Flow Under Rolling Motion

Two-phase flow instability and dynamics of a parallel multichannel system has been theoretically studied under periodic excitation induced by rolling motion in the present research. Based on the homogeneous flow model considering the rolling motion, the parallel multichannel model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The influences of the inlet, upward sections, and heating power on the flow instability under rolling motion have been analyzed. The marginal stability boundary (MSB) under rolling motion condition is obtained. The unstable regions occur in both low and high equilibrium quality and inlet subcooling regions. The multiplied period phenomenon occurs in the high equilibrium quality region and the chaos phenomenon appears on the right of MSB. The concept of stability space is presented.© 2010 ASME