Numerical Simulation of Full Scale Upper Plenum Test Facility Experiments

Höhne, Thomas; Deendarlianto,

doi:10.11159/jffhmt.2017.003

Cited by 2 publications

(1 citation statement)

References 8 publications

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“…Countercurrent flows of water and steam are of great importance in the field of safety analysis of nuclear reactors. CCFL can occur in the hot-leg of a pressurized water reactor PWR during a loss of coolant accident, LOCA or SBLOCA and in the event of loss of residual removal system (loss of RHR) [4], [5].…”

Section: Methodsmentioning

confidence: 99%

Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors

Morghi¹,

Mesquita

2021

Braz. J. Rad. Sci.

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After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermofluidodynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it are characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences.

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Section: Methodsmentioning

confidence: 99%

Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors

Morghi¹,

Mesquita

2021

Braz. J. Rad. Sci.

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CFD Applications to Pressurized Thermal Shock‐Related Phenomena

Okagaki,

Hibiki,

Sibamoto

2024

International Journal of Energy Research

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In pressurized water reactor accident scenarios, the injection of water from the emergency core cooling system (ECCS) (ECC injection) might induce a pressurized thermal shock (PTS), affecting the reactor pressure vessel (RPV) integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal–hydraulic and structural analyses. The thermal–hydraulic approach is particularly crucial, and reliable computational fluid dynamic (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer (DC). Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD simulation can benefit from complement evaluations with thermal–hydraulic system analysis codes. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the Rossendorf coolant mixing model (ROCOM), transient two-phase flow (TOPFLOW), upper plenum test facility (UPTF), and large-scale test facility (LSTF). The results revealed that in single-phase flow CFD simulation, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using large eddy simulation (LES) have been reported. For two-phase flow analysis of air–water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k–ε and shear stress transport (SST) k–ω models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam–water conditions, many studies have used two-fluid models and Reynolds-averaged Navier-Stoke (RANS), and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation.

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Numerical Simulation of Full Scale Upper Plenum Test Facility Experiments

Abstract: -

Cited by 2 publications

References 8 publications

Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors

Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors

CFD Applications to Pressurized Thermal Shock‐Related Phenomena

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