Published 1977 | Version v1
Report Restricted

Effect of steam oxidation on the tensile strength of HTGR structural graphites

Creators

Description

The core support system of the General Atomic Company design High Temperature Gas-Cooled Reactor (HTGR) contains type PGX graphite as core support blocks. The change in ultimate tensile strength of PGX graphite specimens with oxidation (burnoff) has been determined in a safety-related experimental program at Brookhaven National Laboratory(BNL). It is shown that Fe, an impurity in PGX graphite, plays a key role in the rate of oxidation. The subsequent failure of the graphite specimens is dependent upon the total weight loss due to oxidation. The results indicate that the loss in tensile strength is exponentially related to the percent burnoff (weight loss), and that grain orientation of the specimen has a significant effect

Availability note (English)

MF available from INIS under the Report Number; Available from NTIS., PC A02/MF A01.

Files

Restricted

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Additional details

Publishing Information

Imprint Pagination
16 p.
Report number
BNL-NUREG--23000

Conference

Title
Thermal reactor safety meeting.
Dates
31 Jul - 5 Aug 1977.
Place
Sun Valley, Idaho, United States of America (USA).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
9352700
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S37: INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL CHEMISTRY;
Resource subtype / Literary indicator
Conference
Descriptors DEI
GRAPHITE; HTGR TYPE REACTORS; OXIDATION; PERFORMANCE TESTING; REACTOR CORES; SUPPORTS; TENSILE PROPERTIES
Descriptors DEC
CARBON; CHEMICAL REACTIONS; ELEMENTS; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; MECHANICAL PROPERTIES; MECHANICAL STRUCTURES; NONMETALS; REACTOR COMPONENTS; REACTORS; TESTING

Optional Information

Secondary number(s)
CONF-770708--3.