Effects of Fuel and Coolant Temperatures and Neutron Fluence on CANDLE Burnup Calculation
スポンサーリンク
概要
- 論文の詳細を見る
A new method is developed to analyze CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup, where the microscopic group cross-sections are evaluated at every space mesh by TLLI (Table Look-up and Linear Interpolation) method, and used to analyze a fast reactor with natural uranium as a fresh fuel. The results are compared with the conventional method, where only one set of the microscopic group cross-sections is employed, to investigate the effects of space-dependency of the microscopic group cross-sections and feasibility of the old method.The differences of the effective neutron multiplication factor, burning region moving speed, spent fuel burnup and spatial distributions of nuclide densities, neutron fluence and power density may be considerable from the reactor designer point. However, they are small enough when we study about the characteristics of CANDLE burnup for different designs.
- 一般社団法人 日本原子力学会の論文
- 2006-02-25
著者
-
UDAGAWA Yutaka
Japan Atomic Energy Agency
-
Sekimoto Hiroshi
Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology
-
Udagawa Yutaka
Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology
関連論文
- Stress Intensity Factor at the Tip of Cladding Incipient Crack in RIA-Simulating Experiments for High-Burnup PWR Fuels
- Evaluation of Initial Temperature Effect on Transient Fuel Behavior under Simulated Reactivity-Initiated Accident Conditions
- Breeding Capability and Void Reactivity Analysis of Heavy-Water-Cooled Thorium Reactor
- ICONE11-36126 Examination of applicability of IK method in the negative reactivity measurements
- ICONE11-36061 APPLICATION OF CANDLE BURNUP TO BLOCK-TYPE HIGH TEMPERATURE GAS COOLED REACTOR
- A New Estimation Method for Nuclide Number Densities in Equilibrium Cycle
- Radioactive Waste Transmutation and Safety Potentials of the Lead Cooled Fast Reactor in the Equilibrium State
- Comparison of the Burnup Characteristics and Radiotoxicity Hazards of Rock-like Oxide Fuel with Different Types of Additives
- Application of Monte Carlo Method to Solve the Neutron Kinetics Equation for a Subcritical Assembly
- Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code
- Effect of Cooling History on Cladding Ductility under LOCA Conditions
- Measurement of keV-Neutron Capture Cross Sections and Capture Gamma-Ray Spectra of ^Bi
- Feasible Region of Design Parameters for Water Cooled Thorium Breeder Reactor
- Effects of Fuel and Coolant Temperatures and Neutron Fluence on CANDLE Burnup Calculation
- Sensitivities of Some Characteristics of Nuclear Equilibrium State to One-Group Constants
- Transport Equivalent Diffusion Constants for Reflector Region in PWRs
- ICONE11-36045 Removal of Polonium Contamination by Lead-Bismuth Eutectic in Nuclear Systems
- Calculational Method of One-Group Nuclear Constants in Nuclear Equilibrium State
- Design Concept of Fast Spectrum Pulse Reactor with Packed Core of Coated Dilute Fuel Particles
- Exact Error Estimation for Solutions of Nuclide Chain Equations
- Characteristics of Several Equilibrium Fuel Cycles of PWR
- Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
- Fast neutron spectrum in lithium fluoride pile with D-T neutron source.
- New method to analyze equilibrium cycle of pebble-bed reactors.
- Scalar fast neutron spectra in graphite-reflected lithium fluoride pile with D-T neutron source.
- Burn-off and Production of CO and CO2 in the Oxidation of Nuclear Reactor-Grade Graphites in a Flow System.