Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations
TL;DR: In this article, a nuclear coupled thermal-hydraulic model was developed to simulate core-wide and regional stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients.
Abstract: The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.
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TL;DR: A systematic overview of all key two-phase instabilities focusing on the fundamental mechanisms leading to their occurrence is provided, with emphasis on how these mechanisms may change depending on whether flow may be classified as macro- or micro-channel.
Abstract: Study of two-phase flow instabilities began in the late 1920′s, and in the nearly 100 years since, significant progress has been made in both experimental and theoretical understanding of them. Despite these advances, many key deficiencies remain, solution of which will provide appreciable value for system designers looking to leverage phase change heat transfer technologies in a safe and repeatable manner. The present review provides a systematic overview of all key two-phase instabilities focusing on the fundamental mechanisms leading to their occurrence. Emphasis is placed on how these mechanisms may change depending on whether flow may be classified as macro- or micro-channel, particularly relevant due to the modern proliferation of parallel micro-channel heat sinks. Extensive literature surveys are performed for each instability type, and strengths and weaknesses of existing literature assessed. Focus is placed on providing recommendations for future work based on the status of current literature. Important takeaways include the significant mechanistic differences for Density Wave Oscillations and Parallel Channel Instability between macro- and micro-channels, the need for better understanding of the role of parallel micro-channels on external pressure curves (impacting Ledinegg instability and Pressure Drop Oscillations), and the influence of size and position of compressible volume on Pressure Drop Oscillations.
74 citations
TL;DR: In this article, the authors investigated the parallel channel density wave instability in the CANDU supercritical water reactor (SCWR) using a 1-D thermal-hydraulic model in the time domain.
Abstract: The present work investigates the parallel channel density wave instability in the CANDU supercritical water reactor (SCWR) using a 1-D thermal-hydraulic model in the time domain. The model, which was used to analyze a single channel in-phase mode of instability in the SCWR, is extended for parallel channels in the reactor core to determine both in-phase and out-of-phase modes of oscillations considering the origination of the instabilities purely due to the thermal-hydraulic feedbacks without taking into account the influence of neutronic reactivity feedback effects. The capability of the proposed model to analyze the parallel channel instability is validated against the available experimental data. Extensive numerical investigations are carried out to determine the marginal stability boundary for each of the modes of oscillations in the working regime of the CANDU SCWR. The effect of the asymmetric heating power on the instability thresholds is also studied. Finally, the relative dominance between the in-phase and out-of-phase modes of oscillations at the operating regime of the CANDU SCWR is determined.
26 citations
TL;DR: In this paper, a simple one-dimensional numerical thermal-hydraulic model based on a finite-difference scheme has been developed to predict steady and unsteady thermal hydraulic behavior of supercritical water flow at various operating conditions.
Abstract: The present paper is aimed at the development of numerical models to predict steady and unsteady thermal-hydraulic behaviour of supercritical water flow at various operating conditions. A simple one-dimensional numerical thermal-hydraulic model based on a finite-difference scheme has been developed. A detailed CFD analysis based on two turbulence models, Reynolds Stress Model and k–ω SST model, has also been presented in this paper. Seven experimental cases of steady state and vertically up flowing supercritical water in circular tubes operated at various working regimes, such as normal and deteriorated heat transfer regions, are used to validate the numerical models. Comparisons for steady state flow show good agreement between the numerical and experimental results for all normal heat transfer cases and most of the deteriorated heat transfer cases. Next, the numerical models are used for transient simulations. Three case studies are undertaken with a purpose to quantify the time dependent responses from both the 1-D model and CFD model. The comparisons carried out for both the normal and deteriorated heat transfer conditions show a good agreement between the two numerical models.
26 citations
TL;DR: In this paper, a 1-D thermal-hydraulic model, THRUST, is developed to simulate and analyze the CANDU supercritical water reactor (SCWR) from the thermodynamic point of view without considering the effect of neutronic coupling.
Abstract: In this paper, a 1-D thermal-hydraulic model, THRUST, is developed to simulate and analyze the CANDU supercritical water reactor (SCWR) from the thermodynamic point of view without considering the effect of neutronic coupling. THRUST, where a characteristic-based finite difference scheme is used, is validated against the available numerical results. The model is, then, used for the analysis of the CANDU SCWR with a primary focus to determine the conditions for potential density wave oscillations. Extensive numerical studies are performed to obtain the marginal stability boundary in the operating regime of the reactor. The effect of various parameters, such as mass flow rate, operating pressure, axial heat flux profile, local pressure drop coefficient, and friction factor, on the stability thresholds of the reactor have been investigated.
25 citations
TL;DR: An ocean-based thermal-hydraulic system analysis code was developed based on RELAP5/MOD3 code by adding additional force model of ocean condition and control volume coordinate solver model as mentioned in this paper.
Abstract: An ocean-based thermal-hydraulic system analysis code was developed based on RELAP5/MOD3 code by adding additional force model of ocean condition and control volume coordinate solver model. The natural circulation operation characteristics of integrated pressurized water reactor (IPWR) under ocean conditions were studied and the effects of inclination and rolling motions were analyzed. The results conclude that, the inclination condition can reduce the mass flow rate of reactor core and lead to inconsistent coolant flow rates of the left and right loops, furthermore, it affects the heat transfer of once-through steam generators (OTSGs). In the case of rolling motion, the additional pressure drop of the loop is dominated by tangential force, and flow oscillation of different loops cancel each other due to the symmetrical arrangement of the reactor. The off-center roll axis location, the combination of the inclination and rolling motion, both can break the thermal-hydraulic symmetry among different loops and enlarge fluctuation amplitude of the core flow rate.
20 citations
References
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TL;DR: In this paper, the authors identify the causes and mechanisms of thermal-hydrodynamic instabilities in boiling flow in a water-cooled reactor, an evaporator, or an electronic cooling system.
Abstract: Boiling flow in a water-cooled reactor, an evaporator, or an electronic cooling system is susceptible to thermal-hydrodynamic instabilities, which may cause flow oscillations of constant amplitude or diverging amplitude. These oscillations could induce boiling crisis, disturb control systems, or cause mechanical damage. This paper identifies the causes and mechanisms of these instabilities. Based on their mechanisms, various types of instabilities are classified and tabulated. The parametric effects on flow instability, observed experimentally, are systematically presented. Various analytical techniques for predicting the instability threshold are reviewed in terms of their applicability and accuracy.
708 citations
Abstract: A new technique is developed for solving the equations of two-phase fluid dynamics. This technique involves a semi-implicit differencing of the field equations and a variation of the Newton Gauss Seidel iterative method for solving at each time level the resulting system of algebraic equations. Although the technique can be applied to any of several sets of equations representing two-phase flow, including the two-fluid equations, numerical results are presented here for the drift-flux approximation in one dimension. Significant advantages of the method are its stability, ease of programming for complicated flow networks, and ease of extension to problems in two or three dimensions.
150 citations
TL;DR: A review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs) can be found in this article.
Abstract: This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation in an extended operation domain with increased void and thereby increased void reactivity feedback and which often have thinner fuel rods and thereby decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of “unexpected” instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, the subject of BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a “new and improved” state of the art has emerged recently.
142 citations
TL;DR: A novel method of solution of the finite difference equations was deviced and incorporated, and many of the approximations that are common in other stability codes are avoided.
Abstract: A simple code, called SPORTS ∗ , has been developed for two-phase stability studies. A novel method of solution of the finite difference equations was deviced and incorporated, and many of the approximations that are common in other stability codes are avoided. SPORTS is believed to be accurate and efficient, as small and large time-steps are permitted, and hence suitable for micro-computers.
66 citations
TL;DR: In this paper, an eigenvalue problem governing BWR core nuclear thermal-hydraulic modes which result in out-of-phase power oscillations is formulated based on the linearization approximation to nonlinear feedback terms and the very simple models for neutronics and thermalhydraulics.
Abstract: An eigenvalue problem governing BWR core nuclear thermal-hydraulic modes which result in out-of-phase power oscillations is formulated. This formulation is based on the linearization approximation to nonlinear feedback terms and the very simple models for neutronics and thermal-hydraulics. The eigenvalue problem in 5 × 5 matrix formulation can be easily solved without using a computer. A series of the calculations are carried out, at a high-power and low-core-flow condition, to investigate the dependence of the eigenvalues and eigenfunctions on the void reactivity coefficient and the subcriticality of spatial neutronic modes, where the latter parameter is identical to the eigenvalue separation of the higher-harmonic neutronic mode. These results show that the threshold value of the void coefficient for initiating the unstable out-of-phase oscillation strongly depends on the subcriticality. The oscillation mode becomes more unstable with an increase in the absolute value of the negative void coefficient, whereas the mode becomes more stable, almost linearly, with increasing subcriticality. The resonant frequency of the oscillation and the phase shifts between the nuclear thermal-hydraulic variables are consistent with previous measured or calculated values.
55 citations