Computer analysis of HTGR core reactions for steam or air inleakage accidents

Conference · Sun Dec 31 23:00:00 EST 1972

OSTI ID:4393657

Barsell, A W; Chmielewski, R D; Moffette, T R

From national topical meeting on mathematical models and computational techniques for analysis of nuclear systems; Ann Arbor, Michigan, USA (8 Apr 1973). In mathematical models and computational techniques for analysis of nuclear systems. The OXIDE computer code developed for analyzing transient effects of accidental inleakage of steam and/or air to the primary coolant of a high- temperature gas-cooled reactor (HTGR) is described. A core spatial analysis is made of gas diffusion; steamgraphite, oxygen-graphite, and steam-carbide chemical reactions; and heat transfer within the bulk moderator and fuel rods. Application to HTGR accident sequences for safety evaluations is described. 18 references. (auth)

• Research Organization: Gulf General Atomic Co., San Diego, CA; American Nuclear Society. Michigan Section
• NSA Number: NSA-29-007060
• OSTI ID: 4393657
• Report Number(s): CONF-730414--P1
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

*HTGR TYPE REACTORS-- REACTOR SAFETY
*PRIMARY COOLANT CIRCUITS-- LEAKS
MATHEMATICAL MODELS
N77300 --Reactors--Power Reactors
Non-breeding
Graphite Moderated
N77900* --Reactors--Reactor Safety & Environmental Aspects
O CODES
REACTOR ACCIDENTS
TRANSIENTS