Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

Conference · Tue Jan 01 00:00:00 EST 1991

OSTI ID:5177041

Haggag, F M; Nanstad, R K ^[1]; Byrne, S T ^[2]

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1. Oak Ridge National Lab., TN (United States)
2. ABB/Combustion Engineering, Inc., Windsor, CT (United States)

The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low (121 to 210{degrees}C (250--410{degrees}F)) compared to those for commercial light-water reactors (LWRs) ({approximately}288{degrees}C (550{degrees}F)). The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 {times}10{sup 18} neutron/cm{sup 2} (0.68 {times} 10{sup 18} neutron/cm{sup 2} (>1 MeV)) at temperatures of 288, 204, 163, and 121{degrees}C (550, 400, 325, and 250{degrees}F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: USDOE; USDOE, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 5177041
• Report Number(s): CONF-910808-5; ON: DE91018849
• Resource Relation: Conference: 5. international symposium on environmental degradation on materials in nuclear power systems - water reactors, Monterey, CA (United States), 25-29 Aug 1991
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
HTGR TYPE REACTORS
REACTOR VESSELS
RADIATION EFFECTS
STEEL-ASTM-A508
STEEL-ASTM-A533-B
WELDED JOINTS
CHARPY TEST
NEUTRON FLUENCE
REACTOR MATERIALS
TENSILE PROPERTIES
ALLOYS
CARBON STEELS
CONTAINERS
DESTRUCTIVE TESTING
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
IMPACT TESTS
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
LOW ALLOY STEELS
MATERIALS
MATERIALS TESTING
MECHANICAL PROPERTIES
MECHANICAL TESTS
REACTORS
STEELS
TESTING
210300* - Power Reactors
Nonbreeding
Graphite Moderated
360103 - Metals & Alloys- Mechanical Properties
360106 - Metals & Alloys- Radiation Effects