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Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 34

Book Description
The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (id est, contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.

Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 34

Book Description
The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (id est, contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.

Bulletin of the Atomic Scientists

Bulletin of the Atomic Scientists PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 56

Book Description
The Bulletin of the Atomic Scientists is the premier public resource on scientific and technological developments that impact global security. Founded by Manhattan Project Scientists, the Bulletin's iconic "Doomsday Clock" stimulates solutions for a safer world.

Post-Dryout Heat Transfer

Post-Dryout Heat Transfer PDF Author: G.F. Hewitt
Publisher: Routledge
ISBN: 1351423320
Category : Science
Languages : en
Pages : 448

Book Description
The study of post-dryout heat transfer has generated great interest because of its importance in determining maximum clad temperature in nuclear reactor loss-of-coolant accidents (LOCAs). An associated phenomenon, the deterioration of heat transfer in boiling, is significant to other industrial sectors. This book provides comprehensive coverage of post-dryout heat transfer, discussing such essential topics as post-dryout heat transfer in dispersed flow, interpretation and use of transient data in surface rewetting by reinstatement of flow or by reducing heat flux, rod bundles, two-phase flow occurrences in the post-dryout region, various methods for predicting ""inverted annular flow,"" and new experiments for measuring thermodynamic nonequilibrium with probes in the channel. The book also presents a basis for independent safety assessment of nuclear reactors and chemical plant systems where post-dryout heat transfer may occur. Post-Dryout Heat Transfer will be a useful reference for researchers and professionals in the nuclear and chemical production industries.

Pellet-clad Interaction in Water Reactor Fuels

Pellet-clad Interaction in Water Reactor Fuels PDF Author:
Publisher: OECD Publishing
ISBN:
Category : Business & Economics
Languages : en
Pages : 562

Book Description
This publication sets out the findings of an international seminar, held in Aix-en-Provence, France in March 2004, which considered recent progress in the field of pellet-clad interaction in light water reactor fuels. It also reviews current understanding of relevant phenomena and their impact on the nuclear fuel rod under the widest possible conditions, and about both uranium-oxide and mixed-oxide fuels.

Modelling of Fuel Behaviour in Design Basis Accidents and Design Extension Conditions

Modelling of Fuel Behaviour in Design Basis Accidents and Design Extension Conditions PDF Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201080202
Category :
Languages : en
Pages : 218

Book Description
This publication is the result of an IAEA technical meeting and reports on Member States' capabilities in modelling, predicting and improving their understanding of the behaviour of nuclear fuel under accident conditions. The main results and outcomes of a coordinated research project (CRP) on this topic are also presented.

Natural Circulation in Water Cooled Nuclear Power Plants

Natural Circulation in Water Cooled Nuclear Power Plants PDF Author: International Atomic Energy Agency
Publisher: IAEA
ISBN:
Category : Business & Economics
Languages : en
Pages : 656

Book Description
Describes the state of knowledge of natural circulation in water cooled nuclear power plants and passive system reliability. The publication presents information on phenomena, models, predictive tools and experiments that currently support design and analysis of natural circulation systems, and highlights areas where additional research is needed.

Structural Materials for Generation IV Nuclear Reactors

Structural Materials for Generation IV Nuclear Reactors PDF Author: Pascal Yvon
Publisher: Woodhead Publishing
ISBN: 0081009127
Category : Technology & Engineering
Languages : en
Pages : 686

Book Description
Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials. Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors. Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates Written by an expert in that particular area

Safety Margins of Operating Rectors

Safety Margins of Operating Rectors PDF Author: M. Antila
Publisher: International Atomic Energy Agency
ISBN: 9789201181022
Category : Business & Economics
Languages : en
Pages : 145

Book Description
This TECDOC deals with a basic concept of safety margins and their role in assuring safety of nuclear Installations. The document describes capabilities of thermal hydraulic computer codes used to determine safety margins, evaluation of uncertainties, methods for safety margin evaluation and utilization of safety margins in operation and modifications of nuclear power plants.

Failure Analysis of Heat Treated Steel Components

Failure Analysis of Heat Treated Steel Components PDF Author: Lauralice de Campos Franceschini Canale
Publisher: ASM International
ISBN: 1615030980
Category : Technology & Engineering
Languages : en
Pages : 651

Book Description


Status of Fast Reactor Research and Technology Development

Status of Fast Reactor Research and Technology Development PDF Author: International Atomic Energy Agency
Publisher:
ISBN: 9781523130191
Category : Fast reactors
Languages : en
Pages : 0

Book Description
"Based on a recommendation from the Technical Working Group on Fast Reactors, this publication is a regular update of previous publications on fast reactor technology. The publication provides comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors. The main issues of discussion are experience in design, construction, operation and decommissioning, various areas of research and development, engineering, safety and national strategies, and public acceptance of fast reactors. In the summary the reader will find national strategies, international initiatives on innovative (i.e. Generation IV) systems and an assessment of public acceptance as related to fast reactors."--Résumé de l'éditeur.