El Pueblo : diario republicano de Valencia: El Pueblo : diario republicano de Valencia - Año XXIX Número 10939 - 1922 diciembre 4 (04/12/1922)

Studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two /sup 238/PuO/sub 2/ pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported

General-purpose heat source project and space nuclear safety and fuels program. Progress reportt, January 1980

This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are the general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work

General-purpose heat source project and space nuclear safety and fuels program. Progress report

This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed hear are of a continuing nature. Results and conclusions described may change as the work continues

Milliwatt generator project. Progress report, April-September 1981

This formal biannual report covers the effort related to the Milliwatt Generator Project (MWG) carried out for the Department of Energy, Office of Military Applications, by the Los Alamos National Laboratory. Most of the studies discussed here are of a continuing nature. Results and conclusions may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work

REACTOR IRRADIATION OF URANIUM-IMPREGNATED GRAPHITE AT 1500 C TO 10% BURNUP

Two type-AUC graphite fuel elements loaded by solution impregnation to an average concentration of 0.115 g/cc of 93.13% enriched U converted to UC and UC/sub 2/ were irradiated at temperatures of about 1500 deg C to a 10.2% maximum burnup, corresponding to an irradiation level of 219 kwh/cc or 2.45 x 10/sup 19/ fissions/cc of fuel element. Post-irradiation measurements of the elements showed dimensional changes of -4.3 and -4.8% with the grain, and --0.8 to -2.5% across the grain. Weight losses were 3.2 and 5.1% for the individual elements with approximately 11% of the total U being lost from the elements. With-the- grain thermal conductivity at nominal room temperature was reduced by a factor of approximates 7 and electrical conductivities by factors of 3.4 to 8.3, also at room temperature. Impact strength appeared to be somewhat improved by irradiation. Migration of U within the element was detected by radiographic density observations but not evaluated quantitatively. As anticipated, fission product release was high. (auth

PURIFICATION AND CONCENTRATION OF SOLVENT EXTRACTED PLUTONIUM BY OXALATE PRECIPITATION

A production-scale procedure for purifying and concentrating Pu contained in the stripping solution from a TBP extraction process is described. Conclusions from exploratory tests on variables affecting the efficiency are included. (auth

Pyrometallurgical Purification of Plutonium Reactor Fuels

Pyrometallurgical methods studied are liquation, self-drossing and filtration, slagging by the addition of oxide, carbide, or halide, liquid metal extraction, complete conversion to halide, followed by filtration and selective reduction of the plutonium, and electrorefininng. Experimental techniques and results are presented for each method. (auth