B. R. Sehgal

Numerical investigation of bubble growth and detachment by the lattice-Boltzmann method

A numerical study has been performed to investigate the characteristics of bubble growth on, and detachment from, an orifice. The FlowLab code, which is based on a lattice-Boltzmann model of two-ph ...

Experimental and analytical studies of melt jet-coolant interactions : a synthesis

Abstract Instability and fragmentation of a core melt jet in water have been actively studied during the past 10 years. Several models, and a few computer codes, have been developed. However, there are, still, large uncertainties, both, in interpreting experimental results and in predicting reactor-scale processes. Steam explosion and debris coolability, as reactor safety issues, are related to the jet fragmentation process. A better understanding of the physics of jet instability and fragmentation is crucial for assessments of fuel-coolant interactions (FCIs). This paper presents research, conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning molten jet-coolant interactions, as a precursor for premixing. First, observations were obtained from scoping experiments with simulant fluids. Second, the linear perturbation method was extended and applied to analyze the interfacial-instability characteristics. Third, two innovative approaches to computational fluid dynamics (CFD) modeling of jet fragmentation were developed and employed for analysis. The focus of the studies was placed on (a) identifying potential factors, which may affect the jet instability, (b) determining the scaling laws, and (c) predicting the jet behavior for severe accident conditions. In particular, the effects of melt physical properties, and the thermal hydraulics of the mixing zone, on jet fragmentation were investigated. Finally, with the insights gained from a synthesis of the experimental results and analysis results, a new phenomenological concept, named ‘macrointeractions concept of jet fragmentation’ is proposed.

Effect of fluid Prandtl number on heat transfer characteristics in internally heated liquid pools with Rayleigh numbers up to 1012

This paper presents an analysis of effects of the fluid Prandtl number (Pr) on natural convection heat transfer in volumetrically heated liquid pools. Experimental and computational studies performed in the past are reviewed, with particular emphasis on the analysis of Pr number effects. As a practical exercise, numerical analysis is performed for two-dimensional square, semicircular and elliptical enclosures, and for three-dimensional semicircular and hemispherical cavities, to investigate the physics of the effect of the Pr number on heat transfer in internally heated liquid pools with Rayleigh numbers up to 1012. It was found that the fluid Prandtl number has a small effect on heat transfer in the convection-dominated regions (near the top surface and side walls) of the enclosures. The decrease of the Pr number leads to the decrease of the top and side wall Nusselt (Nu) numbers. The effects of the Pr number on the Nu number at the bottom surface of the enclosures are found to be significant and they become larger with increasing Rayleigh numbers. Two physical mechanisms, i.e. thermal diffusivity and kinematic viscosity phenomena, have been proposed to explain the fluid Prandtl number effects. Calculational results have been used to quantify the significance and the area of influence for each mechanism. Also, strong dependence on the geometry (curvilinearity) of the downward cooled pool surface has been found.

An integrated structure and scaling methodology for severe accident technical issue resolution : Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.

Challenges left in the area of in-vessel melt retention

The in-vessel melt retention becomes an important safety objective for the present or future middle power nuclear plants, so care has to be taken in the evaluation of the various phenomena related to ensuring the feasibility of this objective. Since the prediction of the relevant phenomena has to be performed for the prototypical accident conditions, the applicability of the measured data or of the correlations derived from these measurements have to be established and the uncertainties determined. In this context, most uncertainties are introduced by the non-prototypicalities in the experiments. The paper describes the major findings from the OECD RASPLAV project and discusses the remaining challenges left in the area of in-vessel molten corium coolability.

Accomplishments and challenges of the severe accident research

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research.

Core melt spreading on a reactor containment floor

The ex-vessel core melt spreading, cooling and stabilization is proposed for a nuclear power plant containment design. Clearly, the retention and coolability of the decay-heated core debris is very much the focal point in the proposed new and advanced designs so that, in the postulated event of a severe accident, the containment integrity is maintained and the risk of radioactivity releases is eliminated. The work reported here includes three tasks (i) to review related methodology and data base, (ii) to develop the scaling methodology and (iii) to validate the assessment methodology developed by the authors. The study is based on state-of-the-art knowledge of the melt spreading phenomenology, in particular, and, of severe accident phenomenology in general.

On heat transfer characteristics of real and simulant melt pool experiments

Abstract This paper presents results of analytical studies on natural convection heat transfer in scaled and/or simulant melt pool experiments related to the pressurized water reactor in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing the relevant physical processes in prototypical situations is examined. Experiments and experimental approaches are analyzed by focusing on their ability to represent prototypical situations. Calculations are performed to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. In the present analysis, Rayleigh numbers are limited to 10 12 , where uncertainties in turbulence modelling do not override other uncertainties. Calculations are performed to explore limitations of using side wall heating and direct electrical heating. The need for further experimental and analytical efforts is also discussed.

Coupled thermal structural analysis of LWR vessel creep failure experiments

Considering the hypothetical core melt down scenario for a light water reactor (LWR) the failure mode of the reactor pressure vessel (RPV) has to be investigated to determine the loadings on the containment. The failure of reactor vessel retention (FOREVER)-experiments, currently underway, are simulating the thermal and pressure loadings on the lower head for a melt pool with internal heat sources. Due to the multi-axial creep deformation of the vessel with a non-uniform temperature field these experiments are an excellent source of data for validation of numerical creep models. Therefore, a finite element (FE) model has been developed based on a commercial multi-purpose code. Using the computational fluid dynamics (CFD) module the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are performed using a new numerical approach, which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a three-dimensional array is developed where the creep strain rate is evaluated according to the values of the actual total strain, temperature and equivalent stress. Care has to be exercised performing post-test calculations particularly in the comparisons of the measured data and the numerical results. Considering the experiment FOREVER-C2, for example, the recorded creep process appears to be tertiary, if a constant temperature field is assumed. But, small temperature increase during the creep deformation stage could also explain the observed creep behavior. Such considerations provide insight and better predictive capability for the vessel creep behavior during prototypic severe accident scenarios.

On lattice Boltzmann modeling of phase transition in an isothermal non-ideal fluid

A new lattice Bolztmann BGK model for isothermal non-ideal fluid is introduced and formulated for an arbitrary lattice, composed of several D-dimensional sublattices. The model is a generalization ...