Technique for Determining Local Nucleate and Film Boiling Correlations for Large Diameter Tubes

2020 ◽  
Author(s):  
Chukwudi Azih ◽  
Reilly MacCoy ◽  
Hazem Mazhar ◽  
Chris Fraser
Keyword(s):  
Local Heating ◽  
Large Diameter ◽  
Pool Boiling ◽  
Heat Removal ◽  
Nuclear Industry ◽  
Film Boiling ◽  
Safety Margins ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Water Moderator

Abstract The boiling behaviour on the surface of large diameter tubes is known to be strongly dependent on the local orientation of the surface around the circumference. Understanding such local variations in boiling behaviour is of particular interest in the CANDU® nuclear industry, where part of the heat removal path for cooling the fuel under postulated accident conditions is via pool-boiling on the surface of the 132 mm diameter calandria tube that is immersed in a pool of the heavy water moderator. While the average pool boiling behaviour of the calandria tubes has been well studied with integrated experiments, local boiling correlations have not been developed for prototypical diameter calandria tubes. Local boiling correlations will allow for more detailed modelling of postulated accident scenarios and better quantification of safety margins. In this study, the nucleate and film pool-boiling characteristics of a large diameter Zircaloy-2 CANDU® calandria tube were tested in a pool of subcooled water. A novel technique is developed to derive the local boiling curve along the circumference of the tube, involving the local heating, measurement of the wall temperature, and heat flux. The adequacy of the technique is determined by comparing the local boiling curves with previous experiments in which averaged pool boiling characteristics were obtained.

scholarly journals Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Coatings ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 557
Author(s):  
Egor Kashkarov ◽  
Bright Afornu ◽  
Dmitrii Sidelev ◽  
Maksim Krinitcyn ◽  
Veronica Gouws ◽  
...  
Keyword(s):  
Protective Coatings ◽  
Zirconium Alloys ◽  
Nuclear Industry ◽  
Normal Operation ◽  
Severe Accidents ◽  
Recent Developments ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Directed Research

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
A. Khanna ◽  
C. Allison
Keyword(s):  
Heat Flux ◽  
Cooling System ◽  
Heat Removal ◽  
Hydraulic Modeling ◽  
Carbon Fiber Composites ◽  
High Heat Flux ◽  
Vacuum Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Microstructured Surfaces for Enhanced Pool Boiling Heat Transfer

2011 ◽  
Author(s):  
Kuang-Han Chu ◽  
Ryan Enright ◽  
Evelyn N. Wang
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Boiling Heat Transfer ◽  
Pool Boiling ◽  
Heat Removal ◽  
Design Guidelines ◽  
Complete Wetting ◽  
Structured Surfaces ◽  
Microstructured Surfaces ◽  
Wide Range

We experimentally investigated pool boiling on microstructured surfaces which demonstrate high critical heat flux (CHF) by enhancing wettability. The microstructures were designed to provide a wide range of well-defined surface roughness to study roughness-augmented wettability on CHF. A maximum CHF of 196 W/cm2 and heat transfer coefficient (h) greater than 80 kW/m2K were achieved. To explain the experimental results, a model extended from a correlation developed by Kandlikar was developed, which well predicts CHF in the complete wetting regime where the apparent liquid contact angle is zero. The model offers a first step towards understanding complex pool boiling processes and developing models to accurately predict CHF on structured surfaces. The insights gained from this work provide design guidelines for new surface technologies with higher heat removal capability that can be effectively used by industry.

Numerical Simulation of Pool Boiling: A Review

2013 ◽  
Vol 135 (6) ◽  
Author(s):  
Vijay K. Dhir ◽  
Gopinath R. Warrier ◽  
Eduardo Aktinol
Keyword(s):  
Numerical Simulation ◽  
Conjugate Heat Transfer ◽  
Numerical Models ◽  
Pool Boiling ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Gravity Level ◽  
Designed Experiments ◽  
Wall Superheat ◽  
Multiple Bubbles

A review of numerical simulation of pool boiling is presented. Details of the numerical models and results obtained for single bubble, multiple bubbles, nucleate boiling, and film boiling are provided. The effect of such parameters such as wall superheat, liquid subcooling, contact angle, gravity level, noncondensables, and conjugate heat transfer are also included. The numerical simulation results have been validated with data from well designed experiments.

Protective coatings on zirconium-based alloys as accident-tolerant fuel (ATF) claddings

2017 ◽  
Vol 35 (3) ◽  
pp. 141-165 ◽  
Author(s):  
Chongchong Tang ◽  
Michael Stueber ◽  
Hans Juergen Seifert ◽  
Martin Steinbrueck
Keyword(s):  
Protective Coatings ◽  
Future Research ◽  
Coating Materials ◽  
Research Activities ◽  
High Temperature Steam ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Surface Modified ◽  
Deposition Technology ◽  
Zr Alloy

AbstractSurface-modified zirconium (Zr)-based alloys, mainly by fabricating protective coatings, are being developed and evaluated as accident-tolerant fuel (ATF) claddings, aiming to improve fuel reliability and safety during normal operations, anticipated operational occurrences, and accident scenarios in water-cooled reactors. In this overview, the performance of Zr alloy claddings under normal and accident conditions is first briefly summarized. In evaluating previous studies, various coating concepts are highlighted based on coating materials, focusing on their performance in autoclave hydrothermal corrosion tests and high-temperature steam oxidation tests. The challenges for the utilization of coatings, including materials selection, deposition technology, and stability under various situations, are discussed to provide some valuable guidance to future research activities.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Tube Heat Exchanger ◽  
Removal Capacity ◽  
Safety Margins ◽  
Final Design ◽  
Safety Features ◽  
Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

Subcooled Pool Boiling Experiments on Horizontal Heaters Coated With Carbon Nanotubes

2009 ◽  
Vol 131 (7) ◽  
Author(s):  
V. Sathyamurthi ◽  
H-S. Ahn ◽  
D. Banerjee ◽  
S. C. Lau
Keyword(s):  
Heat Transfer ◽  
Carbon Nanotubes ◽  
Heat Flux ◽  
Pool Boiling ◽  
Type A ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Test Fluid ◽  
Type B ◽  
Coated Surfaces

Pool boiling experiments were conducted with three horizontal, flat, silicon surfaces, two of which were coated with vertically aligned multiwalled carbon nanotubes (MWCNTs). The two wafers were coated with MWCNT of two different thicknesses: 9 μm (Type-A) and 25 μm (Type-B). Experiments were conducted for the nucleate boiling and film boiling regimes for saturated and subcooled conditions with liquid subcooling of 0–30°C using a dielectric fluorocarbon liquid (PF-5060) as test fluid. The pool boiling heat flux data obtained from the bare silicon test surface were used as a base line for all heat transfer comparisons. Type-B MWCNT coatings enhanced the critical heat flux (CHF) in saturated nucleate boiling by 58%. The heat flux at the Leidenfrost point was enhanced by a maximum of ∼150% (i.e., 2.5 times) at 10°C subcooling. Type-A MWCNT enhanced the CHF in nucleate boiling by as much as 62%. Both Type-A MWCNT and bare silicon test surfaces showed similar heat transfer rates (within the bounds of experimental uncertainty) in film boiling. The Leidenfrost points on the boiling curve for Type-A MWCNT occurred at higher wall superheats. The percentage enhancements in the value of heat flux at the CHF condition decreased with an increase in liquid subcooling. However the enhancement in heat flux at the Leidenfrost points for the nanotube coated surfaces increased with liquid subcooling. Significantly higher bubble nucleation rates were observed for both nanotube coated surfaces.

Tackling Coolant Freezing in Generation-IV Molten Salt Reactors

2017 ◽  
Author(s):  
N. Le Brun ◽  
A. Charogiannis ◽  
G. F. Hewitt ◽  
C. N. Markides
Keyword(s):  
Molten Salt ◽  
Removal Rate ◽  
Heat Removal ◽  
Internal Flow ◽  
Experimental System ◽  
High Heat ◽  
Diagnostic Techniques ◽  
Piping System ◽  
Cold Surface ◽  
Accident Conditions

In this study we describe an experimental system designed to simulate the conditions of transient freezing which can occur in abnormal behaviour of molten salt reactors (MSRs). Freezing of coolant is indeed one of the main technical challenges preventing the deployment of MSR. First a novel experimental technique is presented by which it is possible to accurately track the growth of the solidified layer of fluid near a cold surface in an internal flow of liquid. This scenario simulates the possible solidification of a molten salt coolant over a cold wall inside the piping system of the MSR. Specifically, we conducted measurements using water as a simulant for the molten salt, and liquid nitrogen to achieve high heat removal rate at the wall. Particle image velocimetry and planar induced fluorescence were used as diagnostic techniques to track the growth of the solid layer. In addition this study describes a thermo-hydraulic model which has been used to characterise transient freezing in internal flow and compares the said model with the experiments. The numerical simulations were shown to be able to capture qualitatively and quantitatively all the essential processes involved in internal flow transient freezing. Accurate numerical predictive tools such the one presented in this work are essential in simulating the behaviour of MSR under accident conditions.

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