Review of concrete properties for prestressed concrete pressure vesssels

Technical Report · Fri Oct 01 04:00:00 EDT 1976

DOI:https://doi.org/10.2172/7343708· OSTI ID:7343708

Nanstad, R. K.

The desire for increasing power output along with safety requirements has resulted in consideration of the prestressed concrete pressure vessel (PCPV) for most current nuclear reactor systems, as well as for the very-high-temperature reactor for process heat and as primary pressure vessels for coal conversion systems. Results are presented of a literature review to ascertain current knowledge regarding plain concrete properties under conditions imposed by a mass concrete structure such as PCRV. The effects of high temperature on such properties as strength, elasticity, and creep are discussed, as well as changes in thermal properties, multiaxial behavior, and the mechanisms thought to be responsible for the observed behavior. In addition, the effects of radiation and moisture migration are discussed. It is concluded that testing results found in the technical literature show much disagreement as to the effects of temperature on concrete properties. The variations in concrete mixtures, curing and testing procedures, age at loading, and moisture conditions during exposure and testing are some of the reasons for such disagreement. Test results must be limited, in most cases, to the materials and conditions of a given test rather than applied to such a general class of materials such as concrete. It is also concluded that sustained exposure of normal concretes to current PCRV operating conditions will not result in any significant loss of properties. However, lack of knowledge regarding effects of temperatures exceeding 100/sup 0/C (212/sup 0/F), moisture migration, and multiaxial behavior precludes a statement advocating operation beyond current design limits. The report includes recommendations for future research on concrete for PCPVs.

• Research Organization: Oak Ridge National Lab., Tenn. (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 7343708
• Report Number(s): ORNL/TM-5497
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200* -- Nuclear Reactor Technology-- Components & Accessories
36 MATERIALS SCIENCE
360603 -- Materials-- Properties
BUILDING MATERIALS
CONCRETES
CONTAINERS
DESIGN
DOCUMENT TYPES
MATERIALS
MATERIALS TESTING
MECHANICAL PROPERTIES
MECHANICAL TESTS
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
PRESTRESSED CONCRETE
RADIATION EFFECTS
REVIEWS
TEMPERATURE DEPENDENCE
TESTING