Prestressed-concrete reactor vessel (PCRV) design-concept study for small HTGR steam-cycle plant
A study was performed to identify a prestressed-concrete reactor vessel (PCRV) concept for a high-temperature gas-cooled reactor (HTGR) steam-cycle plant rated on the order of 100 MW(e). The selected configuration, considered optimum for a small HTGR, is based on a single-cavity approach and uses technology and experience gained from the design, construction, and operation of the PCRV for the Fort St. Vrain nuclear generating station in the USA. A feature inherent in the HTGR nuclear-heat-source design, and in concert with enhanced safety goals, is the ability to remove decay heat by passive means (i.e., natural circulation). The use of a PCRV, in conjunction with a reactor core of small thermal rating, and passive cooling capability, even under the most extreme conditions, results in a plant with benign characteristics that can be classified as super-safe.
- Research Organization:
- GA Technologies, Inc., San Diego, CA (USA)
- DOE Contract Number:
- AT03-76SF70046
- OSTI ID:
- 6119094
- Report Number(s):
- GA-A-16986; CONF-830805-8; ON: DE83009101
- Resource Relation:
- Conference: 7. international conference on structural mechanics in reactor technology, Chicago, IL, USA, 22 Aug 1983
- Country of Publication:
- United States
- Language:
- English
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CONTAINMENT SYSTEMS
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PROCESS HEAT REACTORS
AFTER-HEAT REMOVAL
NATURAL CONVECTION
CONTAINERS
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CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
REACTOR COMPONENTS
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210900* - Nuclear Power Plants- Process Heat Reactors- (-1987)
210300 - Power Reactors
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