Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

March-Leuba, J.; Rey, JoséM.

doi:10.1016/0029-5493(93)90061-d

Cited by 131 publications

(27 citation statements)

References 5 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The system will move from one solution to the other, driven by feedback mechanisms. March-Leuba and Rey [23] presented a detailed explanation of the DWO and the feedback mechanisms, which is driven by the interaction of inertia and friction for the thermo-hydraulic modes. In a nuclear reactor another feedback mechanism is present: the neutronic feedback which couples the instant fluid density to the power production through the moderation.…”

Section: Introductionmentioning

confidence: 99%

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

T’Joen

Rohde

2012

Nuclear Engineering and Design

View full text Add to dashboard Cite

The HPLWR (High Performance Light Water Reactor) is the European concept design for a SCWR (SuperCritical Water Reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction -mass flow rate or gravity -mass flow rate feedback of the system), or they can be coupled thermo-hydraulic -neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic -neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 seconds. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models.

show abstract

Section: Introductionmentioning

confidence: 99%

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

T’Joen

Rohde

2012

Nuclear Engineering and Design

View full text Add to dashboard Cite

show abstract

“…In reactor physics, criticality problems typically require the calculation of the fundamental eigenvalue and associated eigenmode to determine cycle lengths, reactivity margins and power distributions. For analyzing the unstable patterns encountered in Boiling Water Reactors (BWR) [84,85], often modal analysis is performed that requires solutions to large eigenvalue problems. Reactor instabilities are local power variations occurring due to periodic flow oscillations and can sometimes grow undamped.…”

Section: Criticality Eigenproblem and Modal Analysismentioning

confidence: 99%

Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

Ragusa¹,

Siegel²,

Ruggieri³

2010

View full text Add to dashboard Cite

Executive SummaryThe objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.For decades, the modeling of nuclear cores has been divided into several distinct domains of physics: neutronics, hydraulics, heat transfer, … Yet, these physical models describe physical processes that are tightly intertwined and rely heavily on the solution field of one another. In the last decade or so, various existing mono-disciplinary codes have been coupled together in a naive "black-box" fashion, where the output of one code serves as the input of another code, producing nonlinearly inconsistent multiphysics solutions. Such schemes, which are still the main coupling paradigm today for solving nonlinear nuclear reactor physics equations, are based on a linearization of the coupling terms that is never resolved and that can lead to a loss of accuracy and stability in the solution procedure. In order to address the inconsistencies of traditional coupling strategies, a 3-D test-bed code based on reduced physical models has been developed. It includes several physic components, namely, multigroup neutron diffusion, monophasic fluid conservation laws (mass, momentum, energy), nonlinear heat conduction. This test-bed code was used to demonstrate the loss of accuracy order due to traditional operator-split techniques used in conventional coupling schemes; to implement Jacobian-free full resolution coupling schemes; to develop adaptive preconditioning techniques; and to investigate high-order time discretizations and adaptive time stepping strategies. The test-bed code employs state-of-the-art algorithms and libraries, namely the Jacobian0free Newton-Krylov technique to solve consistently fully implicit tightly coupled simulations and the ANL's PETSc library of linear and nonlinear solvers for scalability. We have demonstrated that the multiphysics solution procedure yields highly accurate solutions in space and time and is amenable for space/time adaptivity. This portion of the work is discussed in Part A of this report.It is widely agreed that high fidelity simulations is an utmost important tool for the conception, design and validation of complex physical systems such as nuclear reactors. Examples of such thrust can be seen in the ASCI program and the SciDAC program. GNEP strongly advocates the development of the next-generation simulation software with a far wider range of applicability than conven...

show abstract

“…Two decades later, March-Leuba et al 1993 published a new review on the instabilities associated to the coupling of thermo-hydraulics to neutronics in Boiling Water Reactors (BWRs), therefore compound instabilities. Authors were motivated by a series of instability events in operating BWRs, like the case in Caorso, Italy, 1984, andin LaSalle, US, 1988.…”

Section: Introductionmentioning

confidence: 99%

Experimental Analysis of a Single-Phase Passive Cooling System

Bastos

Couto²,

Gonçalves³

et al. 2018

R. bras. Ener. Renov.

View full text Add to dashboard Cite

The Fukushima Daiichi Accident highlighted the need for new sustainable technologies with high reliability for removing thermal load, in the thermal power area, with focus on nuclear power plants. This technology is designed for heat transfer from a hot source to a cold source by natural convection, without the need of active components, such as pumps or ventilators, reducing costs and improving reliability. In order to analyze the system parameters of such passive systems, with focus on its thermo-hydraulic stability, an experimental campaign was performed using a reduced model built at State University of Rio de Janeiro – UERJ – with a Single-phase Passive Cooling System. Thus, the objective of this work is the experimental characterization of such systems for the analysis of the physical phenomena that drives the flow to unstable regimes and also to validate a 1D numerical model developed within this research project to simulate this kind of systems. Keywords: Natural Convection, Thermal Energy, Nuclear systems.

show abstract

Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

Cited by 131 publications

References 5 publications

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

Experimental Analysis of a Single-Phase Passive Cooling System

Contact Info

Product

Resources

About