Prestressed-concrete reactor vessel (PCRV) design-concept study for small HTGR steam-cycle plant

Conference · Tue Feb 01 04:00:00 EST 1983

OSTI ID:6119094

Venkatesh, M C; McDonald, C F

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
• Country of Publication: United States
• Language: English

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