High-Temperature Gas-Cooled Reactor Safety studies for the Division of Accident Evaluation. Quarterly progress report, April 1-June 30, 1982
Continuing work on High-Temperature Gas-Cooled Reactor (HTGR) severe accident analyses included a study of a hypothetical large-scale release following a permanent loss-of-coolant accident at the Fort St. Vrain reactor and further development of the ORECA code for siting studies of the 2240-MW(t) cogeneration plant HTGR. Work on fission-product release and transport included investigations of alternative iodine chemistry scenarios and an analysis of the major areas of uncertainties in release predictions during severe accidents. Code development work showed further progress in steam generator modeling, development of a multiloop HTGR simulation, and testing of an alternative simplified core model. 5 figures.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 6468970
- Report Number(s):
- NUREG/CR-2874-Vol.2; ORNL/TM-8443-Vol.2; ON: DE83004612
- Country of Publication:
- United States
- Language:
- English
Similar Records
High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, April 1-June 30, 1981
High-temperature gas-cooled-reactor safety studies for the Division of Accident Evaluation. Quarterly progress report, July 1-September 30, 1982. Volume 3
Related Subjects
210300 -- Power Reactors
Nonbreeding
Graphite Moderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ATWS
COGENERATION
DEUS
ENERGY SYSTEMS
FISSION PRODUCT RELEASE
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HTGR TYPE REACTORS
POWER GENERATION
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
RESEARCH PROGRAMS
SAFETY
SIMULATION
STEAM GENERATION