A characteristics-based implicit finite-difference scheme for the analysis of instability in water cooled reactors
TL;DR: In this article, a transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain.
Abstract: The objective of the paper is to analyze the thermally induced density wave oscillations in water cooled boiling water reactors A transient thermal hydraulic model is developed with a characteristics-based implicit finite-difference scheme to solve the nonlinear mass, momentum and energy conservation equations in a time-domain A two-phase flow was simulated with a one-dimensional homogeneous equilibrium model The model treats the boundary conditions naturally and takes into account the compressibility effect of the two-phase flow The axial variation of the heat flux profile can also be handled with the model Unlike the method of characteristics analysis, the present numerical model is computationally inexpensive in terms of time and works in a Eulerian coordinate system without the loss of accuracy The model was validated against available benchmarks The model was extended for the purpose of studying the flow-induced density wave oscillations in forced circulation and natural circulation boiling water reactors Various parametric studies were undertaken to evaluate the model's performance under different operating conditions Marginal stability boundaries were drawn for type-I and type-II instabilities in a dimensionless parameter space The significance of adiabatic riser sections in different boiling reactors was analyzed in detail The effect of the axial heat flux profile was also investigated for different boiling reactors
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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: In this paper, the system control equations of parallel channels were established based on the homogeneous flow model in two-phase region Semi-implicit finite-difference scheme and staggered mesh method were used to discretize the equations, and the difference equations were solved by chasing method Cosine, bottom-peaked and toppeaked heat fluxes are used to study the influence of non-uniform heating on 2-phase flow instability of the parallel channels system.
Abstract: Two-phase flow instability in parallel channels heated by axial non-uniform heat flux has been theoretically studied in this paper The system control equations of parallel channels were established based on the homogeneous flow model in two-phase region Semi-implicit finite-difference scheme and staggered mesh method were used to discretize the equations, and the difference equations were solved by chasing method Cosine, bottom-peaked and top-peaked heat fluxes were used to study the influence of non-uniform heating on two-phase flow instability of the parallel channels system The marginal stability boundaries (MSB) of parallel channels and three-dimensional instability spaces (or instability reefs) under different heat flux conditions have been obtained Compared with axial uniform heating, axial non-uniform heating will affect the system stability Cosine and bottom-peaked heat fluxes can destabilize the system stability in high inlet subcooling region, while the opposite effect can be found in low inlet subcooling region However, top-peaked heat flux can enhance the system stability in the whole region In addition, for cosine heat flux, increasing the system pressure or inlet resistance coefficient can strengthen the system stability, and increasing the heating power will destabilize the system stability The influence of inlet subcooling number on the system stability is multi-valued under cosine heat flux
19 citations
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.
19 citations
TL;DR: In this article, the existence of Ledinegg and dynamic instability phenomena in a rectangular shape of a natural circulation loop at supercritical condition using supercritical water as a working fluid.
Abstract: In this present work mathematical and numerical analysis are carryout to determine the existence of Ledinegg and dynamic instability phenomena in a rectangular shape of natural circulation loop at supercritical condition using supercritical water as a working fluid. A mathematical model has been developed based on the thermal hydraulic (TH) conversion equations of mass, energy and momentum with and without considering the heating structure. Then this dimensional mathematical model has been validated with the SPORT and SUCLIN benchmark model. A rectangular shape with constant diameter of supercritical water natural circulation (RSCWNCL) is investigated and various simulations are performed to find a threshold stability boundary (TSB) at supercritical conditions. Numerical simulations have been carried out to determine the effect of inlet pressure, hydraulic loop diameter and loop riser height on the Ledinegg instability.
11 citations
TL;DR: In this paper, the typical heat flux profiles in SHHX were separated and its effects on PCDWO were studied, and it was found that the heat flux profile will have a distinct effect on PC DWO, and the marginal stability boundaries (MSB) will move clearly as heat flow profile changes.
Abstract: Because of the heat transfer coefficient varying much from single-phase to two-phase and changes of operation conditions, heat flux profiles in sodium heated heat exchanger(SHHX) should be very different from the usually studied uniform profile for analyzing the parallel channel density wave oscillation (PCDWO). In this paper, the typical heat flux profiles in SHHX were separated and its effects on PCDWO were studied. The results showed that the heat flux profile will have a distinct effect on PCDWO, and the marginal stability boundaries (MSB) will move clearly as heat flux profile changes. However, when the operation condition points of the SHHX were taken into account, in most cases the stability boundaries of different heat flux profiles would follow the movement of operation points, making the heat flux profile effect be a factor enhancing the SHHX stability. Because of the heat flux profile effects, impacts of many parameters on stability like water inlet temperature would be suppressed, and only the water mass flow, sodium mass flow and water pressure could have distinct effects. Further discussion was performed and the two phase region was found to be dominate in determining the heat flux profile effect, and the normalized boiling onset location was believed to be a better parameter than the subcooling number for analyzing the heat flux profile effect. It was found that the normalized boiling onset location at the intersection point of the MSB curves of two different heat flux profiles may not change when inlet resistance coefficient varied and might decrease with increase of outlet resistance coefficient, and that the stability could be strongly enhanced by large two phase normalized heat flux. On the basis, a criterion that can be used to estimate the effect of heat flux profile at different boiling onset locations for a certain heat flux profile, to compare roughly the relative stability of different heat flux profiles under different conditions or to analyze qualitatively the slopes of the MSB curves was proposed. This study will provide a better understanding of the heat flux profile effects on PCDWO.
8 citations
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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: In this paper, a parametric study of coupled neutronic-thermohydraulic stability of natural circulation boiling water reactors (BWRs) is performed as an example of the stability characteristics of the Dutch Dodewa
Abstract: A parametric study of coupled neutronic-thermohydraulic stability of natural circulation boiling water reactors (BWRs) is performed As an example, the stability characteristics of the Dutch Dodewa
67 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 article, a theoretical model describing coupled neutronic-thermohydraulic power oscillations in natural circulation boiling water reactors (BWRs) was developed and the governing equations for the thermohydrauli...
Abstract: A theoretical model describing coupled neutronic-thermohydraulic power oscillations in natural circulation boiling water reactors (BWRs) is developed. The governing equations for the thermohydrauli...
66 citations
TL;DR: In this paper, thermal and stability analyses for a two-phase natural circulation mechanism were performed for both existing and next-generation light water reactors, and the results showed that natural circulation is an important passive heat-removal mechanism.
Abstract: Natural circulation is an important passive heat-removal mechanism in both existing and next-generation light water reactors. Thermal and stability analyses are performed for a two-phase natural ci...
38 citations