No abstract available

[en] A program was initiated to develop wrought alloys with a better combination of corrosion resistance and high-temperature strength under advanced-reactor conditions than commercial alloys currently available. This program culminated in 1980 with the design, melting, and fabrication of 10 experimental Ni-Cr-Mo-W-Al-Ti-Zr-C alloys and the initiation of efforts to evaluate their corrosion and mechanical behavior. A previous report presented data on room- and elevated-temperature tensile strength, creep-rupture behavior, impact strength, and compatibility in air and simulated reactor helium for times up to 4000 h. This report presents the results of 10,000-h corrosion tests in a simulated advanced-HTGR helium environment. In these tests, all of the experimental alloys exhibited excellent resistance to carburization in advanced-reactor helium at temperatures up to 900⁰C (1652⁰F)

No abstract available

[en] Fuel rod thermal conductivity is an important parameter for the design and thermal analysis of large HTGR (LHTGR) cores. An investigation has been conducted to measure the thermal conductivity of fuel rods fabricated to be representative of fuel rods for a LHTGR core. The experimental data obtained on unirradiated prototype LHTGR fuel rods and the theoretical basis for predicting the effect of neutron irradiation are presented. Thermal conductivity curves are given for LHTGR fuel rods operating under HTGR conditions

[en] HTGR fuel element block candidate graphites H-327, H-451, and TS-1240 and two experimental laboratory-prepared graphites, H451L and ''X,'' were irradiated isothermally at about 1226 K in the High Flux Isotope Reactor (HFIR) to HTGR end-of-life fluences of approximately 8 x 10²⁵ n/m² (E >29fJ)/sub HTGR/ to compare dimensional changes. Graphite ''X'' was prepared from the calcined petroleum coke used in candidate graphite SO818, which was not available in time for the experiment. Dimensional charge data are presented and discussed

[en] The thermal conductivities of six types of unirradiated prototype large HTGR fuel rods, representing the maximum and minimum fissile and fertile fuel loadings for a large HTGR core, have been measured in vacuum and in helium by a radial heat flow direct heating method at 950⁰ to 1250⁰C. Results are presented and discussed

No abstract available

No abstract available

[en] Nuclear graphites H-451, lot 440 [Great Lakes Carbon Corporation (GLCC)], and SO818 [Airco Speer Division, Air Reduction Corporation (AS)] are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table

[en] Some versions of the HTGR generate high primary coolant gas temperatures (850⁰ to 950⁰C) and exchange this heat, through intermediate heat exchangers (IHX's), to a secondary loop for higher temperature process heat applications. Although IHX's for these systems are typically pressure-balanced (low-stress) units, their design involves several challenges, including the potential interactions between structural materials and impurities present in the HTGR primary coolant. Considerable work is required to qualify materials for IHX applications, including detailed mechanical property characterization, determination of environmental influences on performance, provision of welding materials and procedures for producing joints of adequate strength and integrity, and provisions for wear protection. Some of the work currently under way addressing these issues is described