Thermal annealing of an embrittled reactor vessel: Feasibility and methodology: Final report

Technical Report · Sun Jan 01 04:00:00 EST 1989

OSTI ID:6371821

Mager, T R; Lott, R G

The annealing and reirradiation response of irradiated pressure vessel weldments has been studied using microhardness techniques. Eight different weldments from nuclear pressure vessel surveillance programs were selected for study. Three weldments that were originally irradiated in a test reactor program and an A302B plate were also studied. The microhardness specimens were taken from Charpy specimens originally tested in the reactor pressure vessel surveillance programs. The Charpy specimens had experienced fast neutron fluences ranging from 2.5 x 10/sup 18/ n/cm/sup 2/ to 8.84 x 10/sup 19/ n/cm/sup 2/ (E>1 MeV) prior to the preparation of the microhardness specimens. The original fluences were determined by the availability of surveillance capsule material. Specimens were also prepared from unirradiated archive material. These specimens were then annealed for 168 hours (1 week) at temperatures ranging from 350/degree/C to 450/degree/C and then reirradiated in a test reactor. The test reactor reirradiations were performed in two capsules. The high fluence capsule was irradiated to a neutron fluence of 5.4 x 10/sup 19/ n/cm/sup 2/, while the low fluence capsule was irradiated to 2.1 x 10/sup 19/ n/cm/sup 2/. Microhardness measurements were performed at each stage of the process to monitor the annealing effectiveness and the benefits of annealing after long term reirradiation. The results of this study are provided in this report. A procedure for monitoring the embrittlement of an annealed pressure vessel is also outlined. This procedure is based on the benefit factor methodology developed for use in the model of embrittlement. Examples of how this procedure might be applied to a reactor vessel with a limited number of remaining surveillance specimens are provided. 3 refs., 6 figs., 2 tabs.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (USA); Westinghouse Electric Corp., Pittsburgh, PA (USA). Power Systems Div.
• OSTI ID: 6371821
• Report Number(s): EPRI-NP-6113-M; ON: TI89006853
• Country of Publication: United States
• Language: English

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36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106 -- Metals & Alloys-- Radiation Effects
ALLOYS
ANNEALING
BWR TYPE REACTORS
CONTAINERS
DOCUMENT TYPES
EMBRITTLEMENT
FERRITIC STEELS
HARDNESS
HEAT TREATMENTS
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
MATERIALS TESTING
MECHANICAL PROPERTIES
MICROHARDNESS
PRESSURE VESSELS
PROGRESS REPORT
PWR TYPE REACTORS
RADIATION EFFECTS
REACTOR MONITORING SYSTEMS
REACTORS
STEELS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
WELDED JOINTS