Exploring the Use of Alumina Nanofluid as Emergency Coolant for Nuclear Fuel Bundle

Abstract Nanofluids are suspensions of nanosized particles in any base fluid that show significant enhancement of their heat transfer properties at modest nanoparticle concentrations. Due to enhanced thermal properties at low nanoparticle concentration, it is a potential candidate for utilization in nuclear heat transfer applications. In the last decade, there have been few studies which indicate possible advantages of using nanofluids as a coolant in nuclear reactors during normal as well as accidental conditions. In continuation with these studies, the utilization of nanofluids as a viable candidate for emergency core cooling in nuclear reactors is explored in this paper by carrying out experiments in a scaled facility. The experiments carried out mainly focus on quenching behavior of a simulated nuclear fuel rod bundle by using 1% Alumina nanofluid as a coolant in emergency core cooling system (ECCS). In addition, its performance is compared with water. In the experiments, nuclear decay heat (from 1.5% to 2.6% reactor full power) is simulated through electrical heating. The present experiments show that, from heat transfer point of view, alumina nanofluids have a definite advantage over water as coolant for ECCS. Additionally, to assess the suitability of using nanofluids in reactors, their stability was investigated in radiation field. Our tests showed good stability even after very high dose of radiation, indicating the feasibility of their possible use in nuclear reactor heat transfer systems.

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Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

10.38032/jea.2020.02.004 ◽

2020 ◽

Vol 01 (02) ◽

pp. 53-60

Author(s):

Pronob Deb Nath ◽

Kazi Mostafijur Rahman ◽

Md. Abdullah Al Bari

Keyword(s):

Nuclear Reactor ◽

Cooling System ◽

Computational Study ◽

Reactor Core ◽

Primary Circuit ◽

Feed Water ◽

Pressurized Water ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

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Numerical Analysis of Natural Circulation Phenomena of Supercritical Fluids

Volume 6: Beyond Design Basis Events; Student Paper Competition ◽

10.1115/icone21-16452 ◽

2013 ◽

Cited By ~ 1

Author(s):

Jeffrey Samuel ◽

Glenn Harvel ◽

Igor Pioro

Keyword(s):

Scaling Laws ◽

Nuclear Reactor ◽

Nuclear Reactors ◽

Natural Circulation ◽

Operating Conditions ◽

Passive Safety System ◽

Test Loop ◽

R 134A ◽

Carbon Dioxide Co2 ◽

Core Cooling

The feasibility of operating with natural circulation as the normal mode of core cooling has been successfully demonstrated for a few small sized nuclear reactors. Natural circulation is being considered for cooling the core of a nuclear reactor under normal operating conditions in several advanced reactor concepts being developed today. Although studies have been conducted in natural circulation for many decades, using natural circulation as the primary cooling mechanism for nuclear reactors or as a passive safety system requires a comprehensive understanding of local and integral system phenomena, validated benchmark data, accurate predictive tools, and reliability analysis methods. As full-scale experiments of supercritical water are expensive, scaling laws can be applied to develop test matrices using modelling fluids to reproduce similar conditions in a scaled-down experimental thermalhydraulic loop. The main aim of this work is to understand the natural circulation phenomena by analyzing water and modelling fluids such as Carbon dioxide (CO2) and Freon 134a (R-134a). The use of the modelling fluids at subcritical, pseudocritical and supercritical pressures is discussed along with fluid-to-fluid scaling techniques. The results from a one-dimensional numerical model developed using MATLAB to calculate the steady-state mass flow rate and heat transport characteristics of an experimental natural circulation test loop are presented and analyzed.

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Basic analysis on heat transfer phenomena in natural circulation for liquid sodium

Journal of Physics Conference Series ◽

10.1088/1742-6596/2072/1/012012 ◽

2021 ◽

Vol 2072 (1) ◽

pp. 012012

Author(s):

R Wulandari ◽

S Permana ◽

Suprijadi

Keyword(s):

Heat Transfer ◽

Nuclear Reactor ◽

Cooling System ◽

Natural Circulation ◽

Liquid Sodium ◽

Navier Stokes ◽

Circulation System ◽

Fluid Temperature ◽

Navier Stokes Equation ◽

Basic Equations

Abstract Natural convention, the heat transfer on fluid due to density differences that can be caused by differences in fluid temperature. One example application of natural convection is cooling system, such as nuclear reactor cooling system. The purpose of this study is to analysis the basic characteristic heat transfer of sodium liquid in the natural circulation system for steady state analysis and transient characteristic with Finite Element Method. The selected module is the Non-Isothermal FLow (NITF) module. This module is a combination of three basic equations, namely the continuity equation, the Navier-Stokes equation, and the dynamic equation of heat transfer in fluid. The simulation model measures 1.5 x 2 (m) with sodium liquid (Na) as a fluid.

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POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Acta Polytechnica CTU Proceedings ◽

10.14311/ap.2016.4.0008 ◽

2016 ◽

Vol 4 ◽

pp. 8 ◽

Cited By ~ 1

Author(s):

Vojtěch Caha ◽

Jakub Krejčí

Keyword(s):

Heat Transfer ◽

Heat Flux ◽

Oxide Layer ◽

Critical Heat Flux ◽

Nuclear Reactor ◽

Nuclear Reactors ◽

Convection Flow ◽

Fuel Cladding ◽

Loss Of Coolant Accident ◽

Nuclear Reactor Safety

The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

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On a Long Term Strategy for the Success of Nuclear Power

10.20944/preprints201704.0050.v1 ◽

2017 ◽

Author(s):

Bruno Merk ◽

Dzianis Litskevich ◽

Karl R. Whittle ◽

Mark Bankhead ◽

Richard Taylor ◽

...

Keyword(s):

Nuclear Fuel ◽

Nuclear Power ◽

Nuclear Reactor ◽

Spent Nuclear Fuel ◽

Nuclear Reactors ◽

Theoretical Perspective ◽

Strategic Development ◽

Safety And Reliability ◽

Water Reactors

The current generation of nuclear reactors are evolutionary in design, mostly based on the technology originally designed to power submarines, and dominated by Light Water Reactors. The aims of the GenIV consortium are driven by sustainability, safety and reliability, economics, and proliferation resistance. The aims are extended here to encompass the ultimate and universal vision for strategic development of energy production, the ‘perpetuum mobile’ – at least as close as possible. We propose to rethink nuclear reactor design with the mission to develop a system which uses no fresh resources and produces no fresh waste during operation as well as generates power safe and reliably in economic way. The results of the innovative simulations presented here demonstrate that, from a theoretical perspective, it is feasible to fulfil the mission through the reuse of spent nuclear fuel from currently operating reactors as the fuel for a new reactor. The produced waste is less burdensome than current spent nuclear fuel which is used as feed to the system. However, safety, reliability and operational economics will need to be demonstrated to create the basis for the long term success of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source.

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Thermophysical Studies for Safety Assessment of Promising Nuclear Power Plants

Nuclear and Radiation Safety ◽

10.32918/nrs.2017.2(74).03 ◽

2017 ◽

pp. 14-19

Author(s):

A. Avramenko ◽

M. Kovetska ◽

A. Kravchuk ◽

Yu. Kovetska

Keyword(s):

Heat Transfer ◽

Nuclear Power ◽

Safety Assessment ◽

Power Plants ◽

Nuclear Reactor ◽

Geometrical Parameters ◽

Assembly Design ◽

Core Cooling ◽

Transfer Problems

The paper describes the role of thermophysical studies for safety assessment and improvement of nuclear reactor technologies. The research considers the issues of using nanofluids for core cooling purposes and analyzes heat transfer problems in promising technologies of generation IV reactor designs with helium coolant and supercritical water. Besides, original results of calculations related to degraded heat transfer in seven-rod fuel assembly design with VVER-SKD geometrical parameters were presented.

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assurance of emergency core cooling system in nuclear reactors

Radiation Protection and Environment ◽

10.4103/0972-0464.146466 ◽

2014 ◽

Vol 37 (1) ◽

pp. 52

Author(s):

AnilKumar Saxena

Keyword(s):

Cooling System ◽

Nuclear Reactors ◽

Core Cooling

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Heat-Transfer at Supercritical Pressures

2010 14th International Heat Transfer Conference, Volume 7 ◽

10.1115/ihtc14-23403 ◽

2010 ◽

Cited By ~ 2

Author(s):

Igor Pioro

Keyword(s):

Heat Transfer ◽

Supercritical Water ◽

Power Plants ◽

Cooling System ◽

Nuclear Reactors ◽

Turbine Blades ◽

Convection Heat Transfer ◽

Steam Generators ◽

Capital Costs ◽

The 1950S

The first works devoted to the problem of heat transfer at supercritical pressures started as early as the 1930s. E. Schmidt and his associates investigated free-convection heat transfer to fluids at the near-critical point with the objective of developing a new effective cooling system for turbine blades in jet engines. In the 1950s, the idea of using supercritical “steam”-water appeared to be rather attractive for steam generators / turbines to increase thermal efficiency of fossil-fired power plants. Intensive work on this subject was mainly performed in the former USSR and in the USA in the 1950s–1980s. Therefore, the most investigated flow geometry at supercritical pressures is circular tubes with water as the coolant. Currently, using supercritical “steam” in fossil-fired power plants is the largest industrial application of fluids at supercritical pressures. At the end of the 1950s and the beginning of the 1960s, some studies were conducted to investigate the possibility of using supercritical water as a coolant in nuclear reactors. Several concepts of nuclear reactors were developed. However, this idea was abandoned for almost 30 years, and then regained momentum in the 1990s as a means to improve the performance of water-cooled nuclear reactors. Main objectives of using supercritical water in nuclear reactors are increasing the efficiency of modern nuclear power plants, which is currently 30–35%, to circa 43–50%, and decreasing operational and capital costs by eliminating steam generators, steam separators, steam dryers, etc. Therefore, objectives of the current paper are to assess the work that was performed and to understand specifics of heat transfer at supercritical pressures.

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Calculations of Combined Radiation and Convection Heat Transfer in Rod Bundles Under Emergency Cooling Conditions

Journal of Heat Transfer ◽

10.1115/1.3450569 ◽

1976 ◽

Vol 98 (3) ◽

pp. 414-420 ◽

Cited By ~ 34

Author(s):

K. H. Sun ◽

J. M. Gonzalez-Santalo ◽

C. L. Tien

Keyword(s):

Heat Transfer ◽

Heat Transfer Coefficient ◽

Transfer Coefficient ◽

Cooling System ◽

Radiation Heat Transfer ◽

Convection Heat Transfer ◽

Transfer Coefficients ◽

Operation Conditions ◽

Water Reactors ◽

Core Cooling

A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle.

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