Paul Coddington

A study of the performance of void fraction correlations used in the context of drift-flux two-phase flow models

Drift-flux models have traditionally been and are currently used in thermal-hydraulic analysis codes in the nuclear and other industries to analyze the behavior of systems during a wide variety of transient conditions. Their simplicity and closeness to experimental data, compared to two-fluid models, and their robustness, make them a cost-effective and efficient choice, although these models are generally limited to co-current flow. The drift-flux models are based on correlations to compute the void fraction distribution and slip in two-phase flow needed to obtain the relative velocity between the phases. Thus, the accuracy of the correlations has a decisive role in determining the correct transport of vapor along the system and, subsequently, in the prediction of the correct response of nuclear or industrial systems. This paper presents the results of an evaluation of the accuracy of a range of widely used void fraction correlations based on the Findlay–Zuber drift-flux model. The 13 correlations presented in this paper, a sub-set of all considered, can loosely be termed as ‘wide range void correlations’, since, as shown in this paper, they are those able to perform reasonably well for the wide range of experimental conditions used in the assessment. The size of the experimental database allowed a detailed statistically based comparison of the performance of all the correlations assessed. The void fraction data is taken from rod bundle, level swell and boil-off experiments performed within the last 10–15 years at 9 experimental facilities in France, Japan, Switzerland, the UK and the USA. The pressure and mass fluxes of the analyzed experiments range from 0.1 to 15 MPa and from 1 to 2000 kg m−2 s−1, respectively. Finally, the assessment of a widely used correlation against experimental transient void fraction data has been performed. The selected correlation is that of Chexal–Lellouche, currently used in the system codes RETRAN-3D and RELAP-5. The results show the performance of the correlation when used in the context of a system code and two different drift-flux model approaches, namely, an algebraic slip calculation and the calculation of the slip velocity based on the solution of a differential slip equation. The accuracy of the predictions shows that it is possible to use a drift-flux approach even for the analysis of rapid transients.

Trends and needs in experimentation and numerical simulation for LWR safety

Abstract The very complex phenomena that need to be considered in safety analyses require use of sophisticated analytical tools. Basically, one-dimensional (1D) system codes have been used for a long time and have reached a degree of maturity. There are, however, limits to their capabilities and further developments are underway; these are outlined. The development of new generations of tools and methods can profit from the availability of increasingly powerful computers and advances in multiphase flow, information technology and numerical techniques. Three-dimensional (3D) situations need also to be addressed more frequently now. Certain developments in these directions that are already taking place in various EURATOM research programs and elsewhere are briefly reviewed; case studies of applications are discussed and lessons drawn. Future safety analyses for nuclear power plants may include use of Computational Fluid Dynamics (CFD) for parts of the primary system and the containment. First applications in this direction have already been made. Although 3D, single-phase CFD computations are commonplace, the size of the systems considered make these quite challenging. The real challenges lie, however, in two-phase flow CFD applications that are still at their very infancy. Coupling of neutronic and thermal-hydraulic codes is also necessary for certain problems.

Evaluation of the Slip Options in RETRAN-3D

The RETRAN-3D code provides the user with a range of options for calculating the liquid-vapor slip. In the three-equation model, for example, two drift flux correlations are available, while the four-equation model includes an additional momentum equation with three interphase friction options. To assess the adequacy of these options, RETRAN-3D has been evaluated against a wide range of rod bundle void fraction data. The data used for this analysis includes information on 83 experiments from nine facilities performed in four different countries. The results of the assessment show that all options provide an excellent prediction of the data for conditions typical of boiling water reactor normal operation. However, there is a progressive worsening of the predictive quality of all options, except that of the Chexal-Lellouche correlation, as first the flow rate and second the system pressure is reduced. At low mass fluxes and pressur there is some overprediction by the Chexal-Lellouche correlation, while at very low pressures the code fails to reach a converged solution. An assessment of the five-equation subcooled boiling model shows an overprediction of the void fraction for negative values of the equilibrium quality.

Development of a drift-flux model for heavy liquid metal/gas flow

The gas lift pump concept based on the bubbling of an inert gas into the primary reactor coolant to enhance natural circulation is currently considered in a number of PbBi-cooled reactor concepts. Thus, the analysis of available void fraction data and the development of two-phase heavy liquid metal/gas flow calculational models have become an important issue in the study of advanced nuclear reactor systems. In the absence of the detailed two-phase flow information needed to develop a flow regime map and the associated interfacial relations, drift-flux models have often been used in the thermal-hydraulic analysis of nuclear and other systems. Accordingly, we consider, in the current paper, the analysis of five sets of experimental data with different geometries, working fluids, flow rates and void fraction ranges, with a view to obtaining a best fit to the data in the form of a drift-flux model. The results of the analysis show that, for systems with flowing fluid, it is possible to represent the heavy liquid metal void fraction data in the form of a drift-flux correlation with a residual error of as low as 0.016, thus offering an improvement over existing void correlations.

Analysis and sensitivity studies of postulated SB-LOCAs in the Mühleberg (KKM) BWR/4 by TRAC-BF1

Abstract In this work, we shall report on the analysis of a number of postulated small breaks ranging from 1 to 10% of the recirculation line flow area at the Muhleberg (KKM) BWR/4 in Switzerland by using the TRAC-BF1 code. The analysis was performed by assuming both limited as well as full (nominal) emergency core cooling systems (ECCS) availability, and also limited availability of some of the safety relief valves used in the automatic depressurization system. Through these assumptions, we consider the response of the plant to multiple failures and therefore we extend our analysis beyond the normal “design basis”. Furthermore, in order to provide a measure of the “uncertainty” of the results, a sensitivity study was performed by varying some plant parameters as well as physical models in the code. To the best of our knowledge, this is the first time that such a systematic analysis of a wide spectrum of postulated small breaks in a commercial BWR has been pursued and the results of this work show that even under the most pessimistic assumptions on the availability of plant safety systems as well as of plant parameters and physical models in the code, the peak clad temperatures never exceed 1000 K.

Implementation of an Improved Interfacial Mass and Energy Transfer Model in RETRAN-3D

RETRAN-3D-MOD002.0 includes a five-equation flow field model to extend the codes analytical capabilities to situations in which thermodynamic nonequilibrium phenomena are important. Evaluation of this models performance against several depressurization and repressurization transients has shown severe numerical and convergence problems related to the calculation of the interfacial energy and mass transfer. To remove these code limitations, a new interfacial mass and energy transfer model has been developed and implemented in RETRAN-3D. This model calculates the phase change based on the net heat transfer to the liquid-vapor interface at saturation. The heat transfer for each phase is equal to the product of the interfacial area density, a heat transfer coefficient, and the difference between the interface and the bulk temperature of the respective phase. A flow regime map based on the work of Taitel and Dukler is used to identify the flow regime in a control volume and to select the appropriate correlations for these quantitie Assessment of the new models performance includes the simulation of an experimental depressurization transient, OMEGA test 9; a turbine trip transient in a BWR/4; and a very fast depressurization transient, the Edwards pipe problem. The results are free from the previous numerical problems and show a good agreement with experimental values.

The analysis of pressurized water reactor and boiling water reactor reactivity transients with CORETRAN and RETRAN-3D

Abstract A code environment based on the CORETRAN and RETRAN-3D codes for the three-dimensional (3-D) kinetic analysis of transients in Swiss light water reactors is currently being developed and implemented within the STARS project at the Paul Scherrer Institute (PSI). As a first step in the application of these codes, an assessment of both codes for the analysis of reactivity-initiated transients in pressurized water reactors (PWRs) was performed. For that purpose, the Nuclear Energy Agency benchmark exercises, consisting of rod ejection and uncontrolled rod bank withdrawal transients, were selected. These analyses showed that very satisfactory results could be obtained with both CORETRAN and RETRAN-3D. In this paper, a summary of the PWR results, along with an emphasis of important modeling options that were identified during that work, is presented. As a second step, it was considered important to assess both codes for boiling water reactor (BWR) reactivity transients. Therefore, in addition, the analysis of a hypothetical beyond-design-basis rod drop accident for a Swiss BWR core at end of cycle is presented in this paper. This transient, which was previously analyzed with another 3-D code at PSI, shows that also for BWRs, both codes give satisfactory results.

Simulation and Analysis of Experimental Blowdown and Small-Break Loss-of-Coolant Scenarios with RETRAN-3D

Abstract RETRAN-3D, a system analysis code currently employed by the nuclear industry in studies covering a wide variety of operational and accident scenarios, has not been extensively validated for application to loss-of-coolant accident (LOCA) scenarios. The results of the in-depth analysis of two experimental loss-of-coolant transients, namely, Test No. 9 in the French OMEGA facility, and the International Standard Problem 26 (ISP-26) in the Japanese ROSA-IV Facility are discussed. The OMEGA test simulated the blowdown phase of a double-ended cold-leg break, whereas the ISP-26 test simulated a small break (5%) in a full height, volume (1/48), and power (~1/342) scaled facility representing a typical two (or four)-loop pressurized water reactor (PWR) system. The RETRAN-3D results for the OMEGA test show good estimates of the important system parameters, with the best agreement corresponding to the use of the dynamic-slip flow model. A sensitivity analysis on the break flow showed that the Henry/Fauske-Isoenthalpic Expansion critical flow model yields the best results, which are significantly improved with a refined nodalization upstream of the break. The ISP-26 was also simulated using the dynamic-slip flow model. The results indicate that the code is able to calculate a small-break LOCA with a model including the main PWR system components and to reproduce the principal physical processes in a reasonable manner. In summary, this assessment shows the ability of RETRAN-3D to model LOCA scenarios in a reasonable way and also points to areas where further model improvement could result in more accurate simulations.

VIPRE-02 Subchannel Validation Against NUPEC BWR Void Fraction Data

A void fraction validation study is undertaken for the VIPRE-02 code due to the importance of void feedback when the code is coupled to the neutronic code ARROTTA in the new code package CORETRAN. This first stage is based on the steady-state boiling water reactor void,fraction data obtained from the Nuclear Power Energy Corporation (Japan) experimental program. The code bundle-averaged void prediction is found to be adequate for the different assembly types studied, but at the subassembly level, the code underpredicts substantially the void for regions with low power-to-flow ratios. This is believed to be due to the lack of a lateral void drift model. The importance of these regional void underpredictions is negligible for the standard 8 × 8 assembly, but their impact on neutronics still needs assessing for the newer assembly types.

Comparative transient analysis of critical and subcritical 80-MW Pb-Bi eutectic-cooled reactor systems

A consistent analytical comparison has been made of the transient behavior of critical and subcritical fast-spectrum reactor systems, the basic core design assumed in each case being that of the 80-MW(thermal) mixed-oxide-fueled, Pb-Bi-cooled, Experimental Accelerator Driven System (XADS). The transient calculations were performed using the FAST code system developed at the Paul Scherrer Institute. The present study demonstrates a high level of self-protection of both the critical and subcritical systems over a wide range of postulated events, including transient overpower due to reactivity insertion, loss of flow, station blackout, loss of coolant, and core overcooling accidents. The relative advantages and shortcomings of the two system types, from the viewpoint of transient behavior, are discussed on the basis of the corresponding simulation results obtained.