HTGR risk assessment study

Conference · Fri Jul 01 00:00:00 EDT 1977

OSTI ID:5450842

Joksimovic, V.; Houghton, W. J.; Emon, D. E.

The Accident Initiation and Progression Analysis (AIPA) Study being performed for the high-temperature gas-cooled reactor (HTGR) utilizes methods of probabilistic risk assessment similar to those used in the Reactor Safety Study (WASH-1400). Analytical methods are employed for the treatment of common cause failures involving human errors and other dependencies in redundant systems. The original objective of the AIPA study was to provide guidance for safety research and development programs for HTGRs. Core heatup events appear to be relatively important though the risk from such accidents in the HTGR is found to be low. Plateout and fission product transport within the prestressed concrete reactor vessel (PCRV) during core heatup rank high as important subjects for safety research and development to confirm the degree of retention of fission products. Conclusions about the relative and, in particular, absolute safety of HTGRs are being reached with care. Results of the study to date provide a reasonable and strong indication that the HTGR has excellent relative and promising absolute safety characteristics because of its inherent safety features that reduce the potential for fission product release.

• Research Organization: General Atomics, San Diego, CA (United States)
• DOE Contract Number: EY-76-C-03-0167-051
• OSTI ID: 5450842
• Report Number(s): GA-A-14336; CONF-770625-4; TRN: 78-001539
• Resource Relation: Conference: International conference on nuclear systems reliability engineering and risk assessment, Gatlinburg, TN, USA, 20 Jun 1977
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
HTGR TYPE REACTORS
REACTOR ACCIDENTS
RISK ASSESSMENT
FISSION PRODUCT RELEASE
PROBABILITY
REACTOR SAFETY
SYSTEM FAILURE ANALYSIS
ACCIDENTS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
REACTORS
SAFETY
SYSTEMS ANALYSIS
220900* - Nuclear Reactor Technology- Reactor Safety
210300 - Power Reactors
Nonbreeding
Graphite Moderated