Effects of condensation modeling on transient behavior of pressurized water reactors

In simulating pressurized water reactor (PWR) transients with large-scale systems codes such as TRAC and RELAP, the effect of condensation has been recognized as a controlling mechanism in the prediction of plant response. For transients involving contraction of or loss of primary coolant, the rate of condensation (primarily in the pressurizer) controls the system refill characteristics. Several separate but interacting phenomena occur during the process of pressurizer refill: steam compression, system heat losses, thermal stratification or mixing of liquid, and condensation. The relative importance of each of these processes and the degree of interaction between them during different transients is very complex. The existing condensation models do not adequately describe the interplay between these effects and this leads to uncertainties in the predicted system response. Further experimental data and code assessment are required to provide data necessary for improving condensation models. Three examples of transients involving uncertainties introduced by condensation modeling are (1) pressurized thermal shock (PTS) transients, (2) small break loss-of-coolant accidents (SBLOCA), and (3) steam generator tube ruptures (SGTR)

Computer simulation of SiO_x structure based on thin film Si 2p peaks of X-ray photoelectron spectroscopy

SIGLEAvailable from British Library Document Supply Centre- DSC:DX172495 / BLDSC - British Library Document Supply CentreGBUnited Kingdo

Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

Comparisons of TRAC-PF1 calculations with semiscale Mod-3 small-break tests S-07-10D, S-SB-P1, and S-SB-P7. [PWR]

Semiscale Tests S-07-10D, S-SB-P1, and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Test S-07-10D simulated an equivalent pressurized-water-reactor (PWR) 10% communicative cold-leg break for an early pump trip with an emergency core coolant (ECC) injected only into the intact-loop cold leg. Tests S-SB-P1 and S-SB-P7 simulated 2.5% communicative cold-leg breaks for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7. [PWR]

Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

TRAC-PF1 choked-flow model

The two-phase, two-component choked-flow model implemented in the latest version of the Transient Reactor analysis Code (TRAC-PF1) was developed from first principles using the characteristic analysis approach. The subcooled choked-flow model in TRAC-PF1 is a modified form of the Burnell model. This paper discusses these choked-flow models and their implementation in TRAC-PF1. comparisons using the TRAC-PF1 choked-flow models are made with the Burnell model for subcooled flow and with the homogeneous-equilibrium model (HEM) for two-phae flow. These comparisons agree well under homogeneous conditions. Generally good agreements have been obtained between the TRAC-PF1 results from models using the choking criteria and those using a fine mesh (natural choking). Code-data comparisons between the separate-effects tests of the Marviken facility and the Edwards' blowdown experiment also are favorable. 10 figures

A three-dimensional free-Lagrange code for multimaterial flow simulations

A time-dependent, three-dimensional, compressible, multicomponent, free-Lagrange code is currently under development at the Los Alamos National Laboratory. The code uses fixed-mass particles (called mass points) surrounded by median Lagrangian cells. These mass points are free to change their nearest-neighbor connections as they follow the fluid motion, which ensures accuracy in the differencing of equations and allows us to simulate flows with extreme distortions. All variables, including velocity, are mass-point centered and time-advancement is performed using the finite-volume technique. The code conserves mass, momentum, and energy exactly, except in some pathological situations. We utilize the Voronoi connections algorithm for Delaunay tetrahedralization of the median mesh during mesh generations and mesh reconnections. The code is highly vectorized and utilizes all eight processors on a Cray YMP. Also, we have recently mapped the code to the massively parallel Connection Machine. Some of the applications for the free-Lagrange method include atomspheric and ocean-circulation models, oil-reservoir and high-velocity impact simulations. These applications are in addition to our standard model problems of high-explosive driven shock-wave problems that involve high degree of deformation, shear flow, and turbulent mixing. 9 refs., 12 figs

A Parallel Unstructured-Mesh Methodology for Device-Scale Combustion Calculations

At Los Alamos we are developing a parallel, unstructured-mesh, finite-volume CFD methodology for the simulation of chemically reactive flows in complex geometries. The methodology is embodied in the CHAD (Computational Hydrodynamics for Advanced Design) code. In this report we give an overview of the CHAD numerical methodology and present parallel scaling results for calculations of flows in a four-valve diesel engine