Research Challenges of Heat Transfer to Supercritical Fluids

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
X. Cheng ◽  
X. J. Liu
Keyword(s):  
Heat Transfer ◽  
Power Systems ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Superconducting Magnets ◽  
Accurate Determination ◽  
Model Development ◽  
Experimental Studies ◽  
Reactor Core ◽  
Prediction Method

Supercritical fluids (SCFs) become more and more important in various engineering applications. In nuclear power systems, SCFs are considered as coolant of the reactor core such as the supercritical water-cooled reactor (SCWR), superconducting magnets and blankets in the fusion reactors, or as fluid in the energy conversion systems of the next generation nuclear reactors. Accurate determination of heat transfer and the temperature of the structural material (e.g., fuel rod cladding) is of crucial importance for the system design. Thus, extensive studies on heat transfer to SCFs have been carried out in the past five decades and are still ongoing worldwide. However, no breakthrough is recognized or expected in the near future. In this paper, the status, main challenges, and future R&D needs are briefly reviewed. Three aspects are taken into consideration, i.e., experimental studies, numerical analysis, and model development for the prediction of heat transfer coefficient (HTC). Several key challenges and also the important subjects of the future R&D needs are identified. They are (a) data base for turbulence quantities, (b) multisolution of wall temperature, (c) extensive Reynolds-averaged Navier–Stokes (ERANS) method, and (d) new prediction method for HTC.

Application of Supercritical Fluids in Thermal- and Nuclear-Power Engineering

2021 ◽  
pp. 601-658
Author(s):  
Igor L. Pioro
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Thermal Power Plants ◽  
Power Cycles

Supercritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide, and helium are considered and discussed. Also, particularities of heat transfer at Supercritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. Supercritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs, and SCW in SCWRs) and/or working fluids in power cycles (helium, mixture of nitrogen (80%) and helium (20%), nitrogen and carbon dioxide in Brayton gas-turbine cycles, and SCW/“steam” in Rankine cycle).

scholarly journals Small break loss of coolant accident of 200 cm² in cold leg of primary loop of Angra2 nuclear power reactor evaluation

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
EDUARDO MADEIRA BORGES ◽  
GAIANÊ SABUNDJIAN
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Power Reactor ◽  
Reactor Core ◽  
Flow Regimes ◽  
Nuclear Power Reactor ◽  
Primary Loop ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor

2017 ◽  
Author(s):  
Yang Lyu ◽  
Xiao Liang
Keyword(s):  
Liquid Metal ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Fourth Generation ◽  
Reactivity Effect ◽  
The Core ◽  
Reactivity Coefficient

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.

Uses of Supercritical Fluids and Their Characteristics Within Deteriorated Heat Transfer Region

2014 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Data Sets ◽  
Operational Parameters ◽  
Transfer Coefficients ◽  
Environmentally Sustainable ◽  
Energy Demands ◽  
Transfer Region ◽  
Nuclear Systems

Taking into account the expected increase in global energy demands and increasing climate change issues there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a big part in being part of the energy mix since it offers a relatively clean, safe and reliable source of energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as flexibility in design to adapt to different countries and situational needs. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF, 2002). The main design goals for the Generation IV nuclear system concept were to improve economic gains, enhance safety, extend sustainability, and strengthen proliferation resistance. To achieve high thermal efficiencies of up to 45–50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed. An important step towards development of SCF applications in novel Generation IV Nuclear Power Plant (NPP) designs and in other industries; is to understand the thermal hydraulic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. Heat transfer under SC turbulent conditions is generally very complex and is extremely sensitive to the test geometry and operational flow parameters. Detailed sets of experiments have been conducted around the world in tubes and other geometries with SCFs to study the basics of heat transfer. By variation of operational parameters and test geometries, fundamental heat transfer data sets are collected that will help in our understanding of SC heat transfer phenomena. Applications for SCFs are not limited to power industry; recent advancements have indicated the use of SCFs in a much wider range of applications due to its unique and attractive heat transfer characteristics. In this paper, proposed uses of SCFs are presented and basic concept of DHT within SCFs is analyzed.

scholarly journals On the possibility of participation of npps with vverin emergency frequency regulation in power system

2019 ◽  
Vol 21 (3) ◽  
pp. 99-108
Author(s):  
V. A. Khrustalev ◽  
M. V. Garievskii ◽  
I. A. Rostuntsova ◽  
A. V. Portyankin
Keyword(s):  
Power Systems ◽  
Power System ◽  
Nuclear Power ◽  
Power Plants ◽  
Frequency Control ◽  
Reactor Core ◽  
Operating Conditions ◽  
Frequency Regulation ◽  
Primary Control ◽  
Main Circulation

The purpose of the article is to study the possibility and feasibility of participation of nuclear power plants (NPPs) with VVER in emergency frequency control in power systems with a high proportion of nuclear power units and, at the same time, of reducing the power consumption for the own needs of the main circulation pumps during modes with power below nominal. To solve these problems, it was proposed to increase the achievable speeds of power gain (load increase) due to the installation of frequency controlled drives of the MCP. Large system frequency variations (caused by large imbalances between generation and demand) may jeopardize electrical equipment, in terms of maintaining stable and reliable operating conditions. For NPPs, the task of preventing or localizing accidents is even more important than for TPPs, since in case of major system accidents, it is possible to completely stop external power supply of the NPPs own needs. Thus, besides the requirements for the primary control of the frequency of NPPs with VVER, today we need more stringent requirements for their emergency acceleration and mobility. The operation of NPPs with long-term non-recoverable active power shortage causes a decrease in the speed of the main circulation pumps of NPPs with VVER and a decrease in the coolant flow rate. It is shown that the installation of variable frequency drives of the MCPs at NPP with VVER is appropriate not only to save energy consumption for their drive in partial modes, but also to increase the power of NPP above the nominal (without reducing the reserve before the heat exchange crisis in the reactor core) for the elimination of system accidents, and thus to improve the safety of the NPPs included in the power system.

scholarly journals Nanofluid Heat Transfer Enhancement for Nuclear Reactor Applications

2009 ◽  
Author(s):  
Jacopo Buongiorno ◽  
Lin-wen Hu
Keyword(s):  
Heat Transfer ◽  
Power Systems ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Colloidal Dispersions ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Primary Coolant ◽  
The Past ◽  
Flux Limit

Colloidal dispersions of nanoparticles are known as ‘nanofluids’. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks (i.e., erosion, settling, clogging) that hindered the use of particle-laden fluids in the past. At MIT we have been studying the heat transfer characteristics of nanofluids for the past five years, with the goal of evaluating their benefits for and applicability to nuclear power systems (i.e., primary coolant, safety systems, severe accident mitigation strategies). This paper will summarize the MIT research in this area with particular emphasis to boiling behavior, including, prominently, the Critical Heat Flux limit and quenching phenomena.

scholarly journals Evaporation and Condensation Devices for Passive Heat Removal Systems in Nuclear Power Engineering

2017 ◽  
Author(s):  
A. Gershuni ◽  
E. Pismennyi ◽  
A. Nishchik
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Experimental Studies ◽  
Heat Removal ◽  
Nuclear Power Engineering ◽  
Power Engineering ◽  
Engineering Properties ◽  
Two Phase ◽  
Evaporation And Condensation ◽  
Heat Transfer Capacity

The paper justifies advantages of evaporation and condensation heat transfer devices as perspective means of passive heat removal and thermal shielding in nuclear power engineering. The main thermophysical factors that limit heat transfer capacity of evaporation and condensation systems have been examined in the research. The results of experimental studies of heat engineering properties of elongated (8-m) vertically oriented evaporation and condensation devices (two-phase thermosyphons), which showed a high enough heat transfer capacity, as well as stability and reliability both in steady state and in start-up modes, are provided. The paper presents the examples of schematic designs of evaporation and condensation systems for passive heat removal and thermal shielding in application to nuclear power equipment.

scholarly journals Direct Numerical Simulation of the Flow Through a Randomly Packed Pebble Bed

2020 ◽  
Vol 142 (4) ◽  
Author(s):  
Mustafa Alper Yildiz ◽  
Gerrit Botha ◽  
Haomin Yuan ◽  
Elia Merzari ◽  
Richard C. Kurwitz ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Direct Numerical Simulation ◽  
Power Systems ◽  
Nuclear Power ◽  
Flow Behavior ◽  
Reactor Core ◽  
Reynolds Numbers ◽  
Fourth Generation ◽  
Flow Through ◽  
Flow Domain

Abstract The proposition for molten salt and high-temperature gas-cooled reactors has increased the focus on the dynamics and physics in randomly packed pebble beds. Research is being conducted on the validity of these designs as a possible contestant for the fourth-generation nuclear power systems. A detailed understanding of the coolant flow behavior is required in order to ensure proper cooling of the reactor core during normal and accident conditions. In order to increase the understanding of the flow through these complex geometries and enhance the accuracy of lower-fidelity modeling, high-fidelity approaches such as direct numerical simulation (DNS) can be utilized. Nek5000, a spectral-element computational fluid dynamics (CFD) code, was used to develop DNS fluid flow data. The flow domain consisted of 147 pebbles enclosed by a bounding wall. In the work presented, the Reynolds numbers ranged from 430 to 1050 based on the pebble diameter and inlet velocity. Characteristics of the flow domain such as volume averaged porosity, axial porosity, and radial porosity were studied and compared with correlations available in the literature. Friction factors from the DNS results for all Reynolds numbers were compared with correlations in the literature. The first- and second-order statistics show good agreement with the available experimental data. Turbulence length scales were analyzed in the flow. Reynolds stress anisotropy was characterized by utilizing invariant analysis. Overall, the results of the analysis in this study provide deeper understanding of the flow behavior and the effect of the wall in packed beds.

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