Thermal-hydraulic noise analysis of a VVER-1000 reactor with nanofluid as coolant

Nourollahi, Reza; Esteki, Mohammad Hossein; Jahanfarnia, Gholamreza

doi:10.1016/j.pnucene.2018.06.010

Cited by 8 publications

(2 citation statements)

References 13 publications

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“…As previously anticipated, two of the key choices to be made when nanofluids considered in In some notably studies ( [8,10,12,[30][31][32][33][34] among others), due to lack of data, the values of specific heat capacity of nanoparticles at 25 o C were incorrectly adopted and used for calculations in NPPs under operational temperatures (around 300 o C). This incorrect input parameter along with applied BCs (inlet mass flux instead of inlet velocity that is a design parameter) have affected the results achieved in those articles, damaged their conclusions about using nanofluids and wrongly proposed some incorrect modifications.…”

Section: On the Use Of Boundary Conditions And Specific Heat Of Nanoparticlesmentioning

confidence: 99%

On the use of boundary conditions and thermophysical properties of nanoparticles for application of nanofluids as coolant in nuclear power plants; a numerical study

Bafrani

Noori-kalkhoran

Gei

et al. 2020

Progress in Nuclear Energy

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In the first part of the present study, a thermal-hydraulic subchannel code hereafter called 'SUBTHAC' is developed to evaluate the enhancement effects of nanoparticles in core heat transfer. The first version of SUBTHAC (V1.0) can analyze the steady state flow of coolant with Al2O3, TiO2 or CuO as nanoparticles (other types of nanoparticles can be added by the user). Different output profiles can be selected such as fluid temperature, pressure and velocity for each subchannel, clad outside temperature for each fuel rod, axial and lateral mass flow, etc. SUBTHAC uses a dedicated algorithm to solve the subchannel equations and, unlike many other codes, allows for thermophysical parameters of nanoparticles to be a function of the temperature, leading to improvement the accuracy of results. Results computed by SUBTHAC for base fluid (pure water) are validated against those obtained by COBRA-EN code. In the next step, with the aim of validating the capability of nanofluid analysis of SUBTHAC code, its nanofluids results have been validated against reference CFD simulations. After the validation, comprehensive numerical comparisons are conducted to assess the enhancement of thermal-hydraulic parameters by using nanofluids. It is shown that, among Al2O3, TiO2 and CuO nanofluids with volumetric concentration in the range of 1-5%, TiO2-3% and CuO-3% are the best choices to increase fluid outlet temperature and decrease clad temperature, respectively. Using nanofluids with a concentration higher than 3% volumetric is not justifiable as the core pressure drop increases up to more than 20%.In the second part of the manuscript, some relevant remarks are put forward on the assignment of boundary conditions (BC, i.e. inlet velocity/inlet mass flux/inlet Reynolds number) and the adoption of reliable values for specific heat capacity of nanoparticles in operational temperature of NPPs. The effects of using the above boundary conditions and incorrect values of the specific heat (as adopted in the literature so far) are depicted by presenting some profiles of coolant and clad temperature. Selecting different BCs and incorrect values of specific heat for nanoparticles can jeopardize the results of calculations.

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Section: On the Use Of Boundary Conditions And Specific Heat Of Nanoparticlesmentioning

confidence: 99%

On the use of boundary conditions and thermophysical properties of nanoparticles for application of nanofluids as coolant in nuclear power plants; a numerical study

Bafrani

Noori-kalkhoran

Gei

et al. 2020

Progress in Nuclear Energy

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“…The past research also provides perspectives for future directions. In the early phases of sub-channel thermal hydraulic analysis, the focus of research articles was mostly on analytical and experimental work done for the creation of sub-channel analysis codes, [23], [24], [25], [26], [27], [28], [29], [30], [31], [32], [33], [34], [35], [36], [37], [38], [39], [40], [41], [42]. These attempts were made to increase the accuracy of CFD predictions and to derive models from CFD for novel fuel geometries, decreasing the time needed for fuel assembly design and optimization, improving safety, and running the fewest number of tests possible to conserve resources and time.…”

Section: Introductionmentioning

confidence: 99%

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Mollah

2023

1790-5044

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A nuclear power reactor's primary use is to generate thermal energy, which in turn produces electricity. The primary heat source is a nuclear fission event occurring inside the fuel rod. The convection heat is transmitted through the coolant by the heat energy generated at the fuel rod wall boundary. Better heat transfer is produced in the flow area by turbulence and irregularity. As a result, turbulent flow heat transfer may present a significant challenge when predicting and assessing the thermal performance of nuclear power reactors. Computational techniques in convective heat transfer have become indispensable for solving challenging issues in the fields of science and engineering thanks to the development of current sophisticated numerical methods and high-performance computer hardware. The development of novel computational techniques and models for complicated transport and multi-physical phenomena is constantly in demand throughout applicable disciplines. This chapter's objective is to provide some recent developments in computational techniques for convective heat transfer, taking into account research interests in the community of mass and heat transfer, and to showcase relevant applications in nuclear power plant engineering domains including future directions. This study describes the most recent advancements in nuclear reactor convective heat transfer research utilizing the computational fluid dynamics (CFD) method, particularly at Ansys Fluent. This work examines the convective heat transfer and fluid dynamics fluid dynamics for turbulent flows across three rod bundle sub-channels that are typical of those employed in the PWR-based VVER type reactor. In this paper, CFD analysis is carried out using the software tool Ansys Fluent. Temperature distribution profile, velocity profile, pressure drop, and turbulence properties were investigated in this study. Boundary conditions i.e. temperature, velocity, pressure, heat flux, and heat generation rate were applied in the sub-channel domain. The main obstacles and bright spots for the CFD methods in nuclear reactor engineering are discussed, which helps to further its further uses. We intend to research a full-length fuel bundle model for VVER-1200 in the future to gather specific fluid characteristic data and use the findings to analyze safety and operate nuclear power facilities in Bangladesh. This paper presents a thorough analysis of the sub-channel thermal hydraulic codes used in nuclear reactor core analysis. This review discusses several facets of previous experimental, analytical, and computational work on rod bundles and identifies potential future directions based on those earlier studies.

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Applications of hybrid nanofluids in different fields

Jamil

Ali

2020

Hybrid Nanofluids for Convection Heat Transfer

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Thermal-hydraulic noise analysis of a VVER-1000 reactor with nanofluid as coolant

Cited by 8 publications

References 13 publications

On the use of boundary conditions and thermophysical properties of nanoparticles for application of nanofluids as coolant in nuclear power plants; a numerical study

On the use of boundary conditions and thermophysical properties of nanoparticles for application of nanofluids as coolant in nuclear power plants; a numerical study

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Applications of hybrid nanofluids in different fields

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