NDLI: Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

Please wait, while we are loading the content...

ANES/ASME Joint National Solar Energy Week

ASME Citrus Engineering Symposium

ASME India Oil and Gas Pipeline Conference

ASME International Mechanical Engineering Congress and Exposition

ASME International Pipeline Geotechnical Conference

ASME Power Conference

ASME Turbine Blade Tip Symposium

ASME/IEEE Joint Rail Conference

ASME/JSME Thermal Engineering Joint Conference

ASME/NRC Pump and Valve Symposium

ASME/USCG Workshop on Marine Technology and Standards

CANEUS: MNT for Aerospace Applications

Dynamic Systems and Control Conference

Energy Nanotechnology International Conference

Energy Sustainability

Engineering Systems Design and Analysis

Engineering Technology Conference on Energy

Fluid Power Systems Technology

Fluids Engineering Division Summer Meeting

Frontiers in Biomedical Devices

Gas Turbine India Conference

Global Congress on NanoEngineering for Medicine and Biology

Heat Transfer Summer Conference

High Temperature Reactor Technology

Information Storage and Processing Systems

Integrated Nanosystems: Design, Synthesis, and Applications

Internal Combustion Engine Division Fall Technical Conference

Internal Combustion Engine Division Spring Technical Conference

International Conference on Fluidized Bed Combustion

International Conference on Fuel Cell Science, Engineering and Technology

International Conference on Integration and Commercialization of Micro and Nanosystems

International Conference on Micro/Nanoscale Heat Transfer

International Conference on Nanochannels, Microchannels, and Minichannels

International Conference on Nuclear Engineering

16th International Conference on Nuclear Engineering(ICONE16)

10th International Conference on Nuclear Engineering(ICONE10)

17th International Conference on Nuclear Engineering(ICONE17)

12th International Conference on Nuclear Engineering(ICONE12)

14th International Conference on Nuclear Engineering(ICONE14)

18th International Conference on Nuclear Engineering(ICONE18)

2013 21st International Conference on Nuclear Engineering(ICONE21)

2014 22nd International Conference on Nuclear Engineering(ICONE22)

ASME 2015 Nuclear Forum collocated with the ASME 2015 Power Conference, the ASME 2015 9th International Conference on Energy Sustainability, and the ASME 2015 13th International Conference on Fuel Cell Science, Engineering and Technology(NUCLRF2015)

2016 24th International Conference on Nuclear Engineering(ICONE24)

ASME 2017 Nuclear Forum collocated with the ASME 2017 Power Conference Joint With ICOPE-17, the ASME 2017 11th International Conference on Energy Sustainability, and the ASME 2017 15th International Conference on Fuel Cell Science, Engineering and Technology(NUCLRF2017)

2017 25th International Conference on Nuclear Engineering(ICONE25)

2018 26th International Conference on Nuclear Engineering(ICONE26)

Volume 6A: Thermal-Hydraulics and Safety Analyses

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management

Volume 6B: Thermal-Hydraulics and Safety Analyses

Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory

Uncertainty Evaluation and Sensitivity Analysis Under Accident Scenarios

Multi-Dimensional Heterogeneous Resonance Integral Tables Generated for Embedded Self-Shielding Method Towards Irregular Lattices

Application of the Dynamic Rod Worth Measurement Method on WWER

Modelling and Simulation Activities in Support of the UK Nuclear R&D Programme on Digital Reactor Design

Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC

Advances in FEMN: The Code for Nuclear Noise Analysis Based on ICEM-CFD

The Design of Nuclear Noise Frequency Spectrum Analysis System and Helium Bubble Movement Behavior Computation for Narrow Channel Core

Effect of Al2O3 and SiC Content in Pack Cementation Powders on the Microstructure of SiC Coatings on HTR Graphite Spheres

Modular Boiling Water Reactor Concept

Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction

A Space-Time Parallel Method to Solve Space-Dependent Neutron Kinetics Equations in Hexagonal-Z Geometry

Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor

Thermal and Mechanics Analysis Code of the PWR Nuclear Fuel Performance Based on FEM

Isogeometric Multi-Level Iterative Solution Algorithms With Applications in Nuclear Reactor Physics

Polygonal Virtual Element Spatial Discretisation Methods for the Neutron Diffusion Equation With Applications in Nuclear Reactor Physics

Interior Penalty Schemes for Discontinuous Isogeometric Methods With an Application to Nuclear Reactor Physics

Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry

Fuel Cycle Economy of Accident Tolerant Fuel Assemblies

Physics Design of Special Epithermal Neutron Beam Based on Multi D-D Reaction Neutron Tubes

Strength Limit of Thimble Tube With Material Plasticity Under Bending Moment and Axial Compression Force

Preliminary Neutron Simulation of Ceramic Fast Reactor

Development and Application of a 2D/1D Fusion Code With Leakage Reconstruction Method

Research on the Effect of Grid Clamping Failure on Flow Elastic Stability and Vortex Shedding of Fuel Rod

Development and Verification of a New Nuclear Data Processing System NECP-Atlas

Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys

Investigation on Breakaway Oxidation Behavior of CZ Alloys

Creation of an OpenFOAM Fuel Performance Class Based on FRED and Integration Into the GeN-Foam Multi-Physics Code

A Monte Carlo Method for Simulating Stochastic Neutron Fields in Criticality Transients

Analysis on Pellet-Cladding Interaction of Fuel Rod During Power Ramp of NHR200-II

Finite Element Analysis for Fuel Assembly Structural Behavior

A Theoretical Model of the Stress Intensity Factor Threshold of DHC for Fuel Cladding Tube

A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification

Propagation of Nuclear Data Uncertainties in PWR Pin-Cell Burnup Calculations via Stochastic Sampling

Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

Enhancement of Stochastic Sampling Capability in RMC Code

Fast Sub-Grid Scale Finite Element Method for the First Order Neutron Transport Equation

Experiences on Radioactive Materials Safe Transport in CIRP

Study on Variation of HPGE Detector Dead Layer Thickness and its Effect on the Measurements of the Detector Response and Samples Characterization Using Monte Carlo Simulation

Uranium Enrichment Measurement Using Enrichment-Meter Approach

Conceptual Core Design of HAPPY200 Reactor

Analyzing the Impact of Solutes on PKA Spectrum for Simulation of Neutron Induced-Radiation Damage in Zr-Based Metals

Development of a Spatial Domain Decomposition Scheme for Monte Carlo Neutron Transport

One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation

Preliminary Study on Thermal Performance of Inert Matrix Disperse Pellet Using FEA Method

Analysis and Improvement of Global-Local Self-Shielding Calculation Scheme for AIC Control Rods

Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT

Criticality Safety Issues in Nuclear Design of HP-SMTCs Space Reactor

Analysis on the Dynamic Buckling Behavior of the Spacer Grid Structure

Foreseen Capabilities, Bottlenecks Identification and Potential Limitations of Serpent MC Transport Code in Large-Scale Full 3-D Burnup Calculations

Influence of Xe-135 Dynamic Behavior on Core Operation Safety for a Molten Salt Reactor

First Steps Towards the Development of a 3D Nuclear Fuel Behavior Solver With OpenFOAM

Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation

Experimental Research and Development of Safety Analysis Systems for Advanced Types of Fuel Rods

Internal Coupling Between Neutronics and Thermal-Hydraulics With RMC/CTF and Validation Using VERA Benchmarks

Assessment of Failure Modes of Monolithic Fuel Plates

Experimental and Modeling Research on Leading Betavoltaic Technology

Greatly Enhanced Thermal Conductivity of Fully Inert Matrix Dispersion Pellet (IMDP) Produced by Spark Plasma Sintering (SPS) Technique

Volume 7: Decontamination and Decommissioning, Radiation Protection, and Waste Management; Mitigation Strategies for Beyond Design Basis Events

Volume 8: Computational Fluid Dynamics (CFD); Nuclear Education and Public Acceptance

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation

Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues

Volume 9: Student Paper Competition

Volume 1: Operations and Maintenance, Engineering, Modifications, Life Extension, Life Cycle, and Balance of Plant; Instrumentation and Control (I&C) and Influence of Human Factors; Innovative Nuclear Power Plant Design and SMRs

International Conference on Offshore Mechanics and Arctic Engineering

International Conference on Radioactive Waste Management and Environmental Remediation

International Design Engineering Technical Conferences and Computers and Information in Engineering Conference

International Electronic Packaging Technical Conference and Exhibition

International Heat Transfer Conference

International Joint Power Generation Conference

International Joint Tribology Conference

International Manufacturing Science and Engineering Conference

International Pipeline Conference

International Solar Energy Conference

International Symposium on Flexible Automation

Multifunctional Nanocomposites and Nanomaterials International Conference

Noise Control and Acoustics Division Conference

North American Waste-to-Energy Conference

Pressure Technology

Pressure Vessels and Piping Conference

Rail Transportation Division Conference

Small Modular Reactors Symposium

Smart Materials, Adaptive Structures and Intelligent Systems

Summer Bioengineering Conference

Turbo Expo: Power for Land, Sea, and Air

Verification and Validation

Water Quality, Drought, Human Health and Engineering Conference

Wind Energy Symposium

World Conference on Innovative Virtual Reality

World Tribology Congress

Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Horie, Hideki Takeuchi, Yutaka Takiwaki, Kenya Sebe, Fumie Kakiuchi, Kazuo Sato, Hisaki
Copyright Year	2018
Abstract	Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.
Sponsorship	Nuclear Engineering Division
File Format	PDF
ISBN	9780791851456
DOI	10.1115/ICONE26-81923
Volume Number	Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory
Conference Proceedings	2018 26th International Conference on Nuclear Engineering
Language	English
Publisher Date	2018-07-22
Publisher Place	London, England
Access Restriction	Subscribed
Subject Keyword	Water Temperature Hydrogen Japan Light water reactors Chemical reactions 2011 Fukushima nuclear disaster High temperature Fuels Steel Scram Steam Probabilistic risk assessment Oxidation Silicon Thermal hydraulics Nuclear power stations Damage Cladding systems (building) Accidents
Content Type	Text
Resource Type	Article

Central Library (ISO-9001:2015 Certified)
Indian Institute of Technology Kharagpur
Kharagpur, West Bengal, India | PIN - 721302

See location in the Map
03222 282435
Mail: support@ndl.gov.in

Sl.	Authority	Responsibilities	Communication Details
1	Ministry of Education (GoI), Department of Higher Education	Sanctioning Authority	https://www.education.gov.in/ict-initiatives
2	Indian Institute of Technology Kharagpur	Host Institute of the Project: The host institute of the project is responsible for providing infrastructure support and hosting the project	https://www.iitkgp.ac.in
3	National Digital Library of India Office, Indian Institute of Technology Kharagpur	The administrative and infrastructural headquarters of the project	Dr. B. Sutradhar bsutra@ndl.gov.in
4	Project PI / Joint PI	Principal Investigator and Joint Principal Investigators of the project	Dr. B. Sutradhar bsutra@ndl.gov.in Prof. Saswat Chakrabarti will be added soon
5	Website/Portal (Helpdesk)	Queries regarding NDLI and its services	support@ndl.gov.in
6	Contents and Copyright Issues	Queries related to content curation and copyright issues	content@ndl.gov.in
7	National Digital Libarray of India Club (NDLI Club)	Queries related to NDLI Club formation, support, user awareness program, seminar/symposium, collaboration, social media, promotion, and outreach	clubsupport@ndl.gov.in
8	Digital Preservation Centre (DPC)	Assistance with digitizing and archiving copyright-free printed books	dpc@ndl.gov.in
9	IDR Setup or Support	Queries related to establishment and support of Institutional Digital Repository (IDR) and IDR workshops	idr@ndl.gov.in