Fuel behavior analysis code FEMAXI-FBR development and validation for core disruptive accident

Japan Nuclear Energy Safety Organization (JNES) is a Technical Safety Organization (TSO) for Japanese Nuclear Regulation Authority to provide technical information on safety reviews including incident and accident analyses of nuclear power plants including Fast Breeder Reactor (FBR). JNES has develo...

Full description

Saved in:
Bibliographic Details
Published in:Progress in nuclear energy (New series) 2015-07, Vol.82, p.80-85
Main Authors: Okawa, T., Tatewaki, I., Ishizu, T., Endo, H., Tsuboi, Y., Saitou, H.
Format: Article
Language:English
Subjects:
Accident analysis
Code development
Core disruptive accident
Fast reactor fuel
Fuel behavior
Fuels
Mathematical models
Modules
Nuclear engineering
Nuclear power generation
Nuclear reactor components
Nuclear reactors
Citations: Items that this one cites
Items that cite this one
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary:Japan Nuclear Energy Safety Organization (JNES) is a Technical Safety Organization (TSO) for Japanese Nuclear Regulation Authority to provide technical information on safety reviews including incident and accident analyses of nuclear power plants including Fast Breeder Reactor (FBR). JNES has developed an ASTERIA-FBR code system as an integrated Core Disruptive Accident (CDA) analysis code of the regulatory side for the purpose of evaluating the Unprotected Loss-of-Flow (ULOF) accident of FBR cores. The ASTERIA-FBR consists of three main modules, a thermal-hydraulics calculation module; CONCORD, a fuel behavior calculation module; FEMAXI-FBR, and a neutronic calculation module; dynamic GMVP. The FEMAXI-FBR is based on a LWR fuel behavior simulation code FEMAXI-6 and modified the material properties, the calculation models for steady state and transient operational conditions. The FEMAXI-FBR has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient test data. Furthermore, the fuel temperature calculation model was validated by B5D Power-To-Melt (PTM) test at the experimental fast reactor JOYO in Japan. Through these validation, the code is found to be capable of evaluating FBR fuel pin behavior from normal to accidental conditions precisely.
ISSN:0149-1970
DOI:10.1016/j.pnucene.2014.11.002