MARITIME GAS-COOLED REACTOR PROGRAM. STRENGTH AND STRUCTURE OF SINTERED BEO MODERATOR CERAMICS
The development of beryllium oxide moderator ceramics for a gas-cooled reactor requires a knowledge of their strength and structure. Reactor service conditions of simultaneous irradiation, thermal stress, and exposure to hot gas have an important effect on moderator life and the cost of core maintenance. Purposes of this study are: 1) to select strength-test methods for evaluating small ceramic specimens used in capsule irradiation tests; 2) to assess the effects of grain size on the strength of beryllia ceramics; 3) to provide strength data needed for the design of beryllia moderator shapes. Tests were made on specimens to determine strength and to determine the modulus of elasticity in tension. The bend test gave more significant results and less instrumental scatter than the other methods. The ratio of internal tensile strength as determined by diametral loading, to bend strength was about 0.6 for the berryllium- oxide specimens.
- This document is available for review at the Department of Commerce Library, Main Commerce Building, Washington, D.C., under reference number GA-2727. Presented at the Pacific Coast Regional Meeting of the American Ceramic Society, San Francisco, California, October 25-28, 1961.
General DynamicsQuincy, MA USA 02169
- Quirk, J
- Lofftus, F H
- Publication Date: 1961-10-24
- Features: Figures; Photos; References; Tables;
- Pagination: 17 p.
- TRT Terms: Elasticity (Mechanics); Nuclear reactors; Radioactive materials; Thermal stresses
- Old TRIS Terms: Nuclear reactor materials
- Subject Areas: Marine Transportation; Materials; Vehicles and Equipment;
- Accession Number: 00026429
- Record Type: Publication
- Source Agency: Maritime Administration
- Report/Paper Numbers: GA-2727
- Contract Numbers: AT(04-3)-187
- Files: TRIS, USDOT
- Created Date: May 11 1973 12:00AM