Lance J. Agee

A Drift-Flux Model of Two-Phase Flow for RETRAN

A drift-flux model of two-phase flow has been incorporated into the RETRAN computer program. The drift-flux equations allow more accurate prediction of the void fraction than the homogeneous mixture model equations that form the basic RETRAN two-phase flow model. In addition, the drift-flux model is more physically based than empirical correlations of void fraction. Comparisons of model predictions with experimental data show good agreement for both subcooled and saturated boiling of water. The drift-flux model as developed and incorporated into RETRAN is given in this paper. 27 refs.

Comparisons of the RETRAN two-phase flow model with experimental data☆

Abstract The RETRAN computer program was designed for thermal-hydraulic analysis of nuclear steam supply systems at steady-state and off-normal conditions. A homogeneous equilibrium mixture model (HEMM) of two-phase flow is used in the program. The constitutive models for wall friction, two-phase friction multiplier, heat transfer coefficient, and critical heat flux are compared with steady-state experimental data in this report. The void fraction predicted by the HEMM is also compared with data. The results of the data comparisons indicate that the constitutive models are accurate representations and that the RETRAN numerical scheme provides an excellent framework for data comparisons.

Overview of 1988-1989 RETRAN activity

This paper reviews the major RETRAN-related activities during 1988-1989. These activities fall into two broad areas: the use of RETRAN-02 by the nuclear industry and the development, verification, and validation of RETRAN-03.

An analytical method of integrating the thermal-hydraulic conservation equations

Abstract This report presents an analytical method of integrating the conservation equations solved by current thermal-hydraulic programs. The main advantages of this method are: (a) its numerical stability, (b) it has a rational automatic time step, (c) it results in a reduction in computer running time, (d) it allows for a major reduction in core storage.

Main-Steam-Line-Break Analysis of TMI-1 Using RETRAN-3D

A main-steam-line-break accident analysis for Three Mile Island Unit I is performed with point kinetics and three-dimensional kinetics with RETRAN-3D MOD002. These analyses were performed to demonstrate differences in results that can be expected due to different reactor kinetics models. To illustrate the difference in kinetics models, the RETRAN-3D models used for both analyses were the same with the exception of the reactor core modeling. The key assumptions and methods used to model loop mixing are described. The point-kinetics analysis demonstrates a significant return-to-power following the reactor trip while the three-dimensional kinetics case does not. This study shows that three-dimensional core transient modeling provides margin to recriticality over a point-kinetics approach. Such margin is desirable to allow for power uprate and extended refueling cycles.

Realistic Scoping Study of Reactivity Insertion Accidents for Typical PWR and BWR Cores

The rod ejection accident in a pressurized water reactor and the control rod drop accident in a boiling water reactor are analyzed in this paper, both in a best-estimate (realistic) and a conservative manner. CORETRAN, a modem three-dimensional time-dependent nodal code, is used for all simulations. In all considered cases, the resulting peak fuel enthalpy is far less than the current licensing limit of 180 cal/g. The advantage of using a three-dimensional code over the classical point-kinetics approach can be summarized: The power peak is nominally a factor of 10 times lower, and the pulse is 10 times wider. Therefore, a three-dimensional approach predicts a much milder event. Sensitivity studies were performed to identify the influence of several parameters on the reactivity insertion simulations.

The RETRAN-3D Code; Pressurized Water Reactor Multidimensional Neutron Kinetics Applications

A RETRAN-03 computer code version has been developed to analyze reactor transients requiring three-dimensional reactor core neutronics models. The new code will enable the user to couple a complex RETRAN nuclear steam supply system model to a detailed multidimensional neutronics core model. The neutronics model is based on a three-dimensional nodal model using the analytic nodal method that allows a detailed three-dimensional representation of the core but requires less computational effort than conventional fine-mesh finite difference methods. The model uses a full two-group diffusion equation implementation coupled to six delayed neutron groups. Two representative analyses were used as evaluation cases. The work involved the first use of the RETRAN-03 advanced system analysis code using three-dimensional neutronics methods. The purpose of these studies was to gain experience in RETRAN-3D modeling methods and to compare the results with previous calculations as part of a code verification effort. The work has led to a new capability for the RETRAN-03 code, enabling the user to examine the core behavior in more detail than in previous versions and to study transients that involve nonsymmetric core behavior.

Evaluation of the RETRAN-3D Wall Friction Models and Heat Transfer Coefficient Correlations

Abstract This paper presents an evaluation of many of the RETRAN-3D two-phase pressure drop and heat transfer models by comparing model prediction to a large body of experimental data. RETRAN-3D has been used to evaluate multiple two-phase pressure drop models utilizing an extensive experimental two-phase pressure drop database. The experimental pressure drop data cover both heated and adiabatic tests in upflow and horizontal configurations for a wide range of key parameters such as pressure, mass flux, quality, and pipe diameters. Two RETRAN-3D two-phase friction options and the Friedel two-phase friction model are tested and compared to the data. For the two-phase friction models compared herein, the modified Baroczy model available in RETRAN-3D is the best choice for all adiabatic and diabatic situations. The RETRAN-3D code has also been used to simulate a wide variety of heat transfer experiments. These heat transfer data cover single-phase and two-phase conditions over a large range of pressure, heat flux, and mass flux values. The performance of the RETRAN-3D default forced convection heat transfer coefficient correlations is evaluated. The Petukhov correlations provide comparable results for single-phase liquid, but the Dittus-Boelter model provides markedly better statistics for single-phase vapor. The RETRAN five-equation model that combines the Dittus-Boelter and Thom correlations provides the best overall subcooled and saturated boiling statistics and scatter chart behavior.

A probabilistic approach for the evaluation of reactivity insertion accident effects

Reported fuel failures at low peak enthalpies for highly exposed fuel during fast reactivity transients promoted the evaluation of reactivity insertion accidents (RIAs) in light water reactors with an approach different from the ultraconservative point-kinetics licensing evaluations performed in the past. On the basis of realistic estimates for the rod worth and plant conditions, an evaluation of the consequences of the RIAs has been performed. For the pressurized water reactor and boiling water reactor (BWR) cases, rod worth became the most important parameter affecting the severity of the accident. In BWRs high subcooling can adversely affect the consequences of the event. The RIA analyses have been performed using an estimation of the distribution for these parameters in an actual plant during startup. The results show that when assumptions consistent with operating procedures are used, the probability of a significant enthalpy increase due to an RIA is greatly reduced.

Overview of electric power research institute nuclear safety analysis activities

An overview is presented of the important RETRAN-related events that have occurred since the previous International RETRAN Conference, which was held in 1991. Summarized are the following: (a) an overview of the Electric Power Research Institutes (EPRIs) nuclear safety analysis activities, including the evolving integration of EPRI codes; (b) utilities using RETRAN-02 for licensing analyses and the current status of the RETRAN-3D design review and anticipated submittal to the U.S. Nuclear Regulatory Commission; and (c) key features and capabilities of RETRAN-3D and new applications of either RETRAN-02 and/or RETRAN-3D.