An overview on rod-bundle thermal-hydraulic analysis

Sha, W.T.

doi:10.1016/0029-5493(80)90018-7

Cited by 41 publications

(11 citation statements)

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“…From thermal hydraulic considerations, the safety requirements of the nuclear reactor systems to be ensured must be able to perform under all different conditions during normal operation, operational transients, anticipated operational occurrences, design basis accidents and under extreme emergency situations by incorporating the engineered safety systems by passive means. This can usually be achieved via thermal hydraulic analysis (Sha, 1980) where such analysis is performed using analysis system, subchannel or computational fluid dynamics (CFD) codes to estimate the various thermal hydraulic safety parameters like critical heat flux (CHF), critical power, fuel centre line temperature, fuel surface temperature, subchannel maximum temperature and bulk coolant outlet temperature (Chelemer et al, 1977). In this paper, the past, present and future challenges of thermal hydraulic analysis for nuclear reactor systems are reviewed based on the design and operation of existing Generation II, III and III+ and advanced Generation IV reactor technologies.…”

Section: Introductionmentioning

confidence: 99%

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

Yeoh

2019

Exp. Comput. Multiph. Flow

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Thermal hydraulic analysis of nuclear reactor core and its associated systems can be performed using analysis system, subchannel or computational fluid dynamics (CFD) codes to estimate the different thermal hydraulic safety margins. The safety margins and operating power limits under different conditions of the primary as well as secondary cooling system such as the system pressure, coolant inlet temperature, coolant flow rate, and thermal power and its distributions are considered as key parameters for thermal hydraulic analysis. Considering the complexity of rod bundle geometry, boiling heat transfer and different turbulent scales bring about the many challenges in performing the thermal hydraulic analysis to ensure the safe design and operation of nuclear reactor systems under normal and abnormal conditions. A comprehensive review is presented of past, present and future challenges in state-of-the-art thermal hydraulic analysis covering various aspects of experimental, analytical and computational approaches.

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Section: Introductionmentioning

confidence: 99%

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

Yeoh

2019

Exp. Comput. Multiph. Flow

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“…The safety assessment of a nuclear reactor core and its associated components strongly relies on thermal hydraulic analysis of the coolant by means of experimental investigation or numerical simulations (Sha, 1980;Yadigaroglu et al, 2003). Restricted by the computational power in the 1960s to 1980s, the thermal hydraulic calculations were mainly performed using the best-estimate system codes such as RELAP5 (RELAP5 Development Team, 1995), ATHLET (Lerchl et al, 2012), CATHARE (Bestion, 1990) and TRAC (Liles and Mahaffy, 1986), and sub-channel codes such as COBRA (Rowe, 1967), VIPRE (Stewart et al, 1983) and MATRA (Hwang et al, 2008).The former are usually used to analyse the overall behaviour of the whole system under different operating conditions, whereas the latter provide a relatively detailed thermal hydraulic analysis at the fuel channel level by solving 1-D transport equations based on individual flow passages formed between fuel rods or fuel rods and walls, i.e.…”

Section: Introductionmentioning

confidence: 99%

Sub-channel CFD for nuclear fuel bundles

Liu

Moulinec

et al. 2019

Nuclear Engineering and Design

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This paper presents a novel Computational Fluid Dynamics (CFD)-based sub-channel framework for nuclear power plants, which combines the advantageous features of modern CFD and traditional 1-D subchannel codes. The new method is capable of producing CFD-level 3-D results with locally desirable refinement when coupled with embedded resolved models, but because a very coarse mesh is used over most part of the domain, the computing cost is typically significantly smaller than that of a conventional CFD simulation. In this new Sub-Channel CFD (SubChCFD), a dual-mesh system is used, comprising, (i) a filtering mesh which aligns with the mesh used in typical sub-channel codes, enabling the use of existing engineering correlations to account for integral wall friction and heat transfer effects, and (ii) a computing mesh, which provides a platform for the solution of the governing equations with turbulence modelled using a mixinglength-type model. The method has been implemented in an open-source finite volume CFD code Code_Saturne and validated initially using a 5×5 bare bundle case based on the OECD/NEA MATiS-H benchmarking experiment. It has been found that SubChCFD is able of satisfactorily predicting both the velocity and temperature fields. To further investigate the performance for complex flow conditions, SubChCFD was applied to two full 3-D cases. The first is a 5×5 rod bundle case with local blockage in one of the sub-channels, creating significant localised cross flows. The second is a two-parallel-assembly channel with different input mass flow rates at the inlet of each assembly, allowing strong inter-assembly mixing. For both cases, SubChCFD has produced results which agree well with experiments and simulations using resolved CFD. It has also been demonstrated that SubChCFD exhibits excellent flexibility in comparison with traditional sub-channel codes and that it has the potential to serve as a substitution to sub-channel codes.

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“…In 1978, Butterworth [20] developed a three-dimensional model for heat transfer in tube bundles. From 1980 to 1982, porous media model and distribution resistance concept was improved by Sha [21] and Sha et al [22], and the concept of surface permeabilities was introduced to account for the anisotropy of tube bundles porosities. Prithiviraj and Andrews [23][24][25] developed a three-dimensional CFD method (named as HEATX) based on the distributed resistance concept along with volumetric porosities and surface permeabilities to simulate flow and heat transfer in STHXsSB.…”

Section: Introductionmentioning

confidence: 99%