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ORNL-4191 - the Molten Salt Energy Technologies Web Site

ORNL-4191 - the Molten Salt Energy Technologies Web Site

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uranium has been recovered by fluorination. EX-<br />

perirnents have been performed using lithium dis-<br />

solved in molteri bismuth as a reductant. Although<br />

<strong>the</strong> results are complicated by an unexplained loss<br />

of metallic lithium to <strong>the</strong> salt phase, <strong>the</strong> distribu-<br />

tion of rare earths between <strong>the</strong> salt and metal<br />

phases can he correlated with <strong>the</strong> lithiuin metal<br />

concentration in <strong>the</strong> metal phase.<br />

25. Modifications to MSWE Fuel Processing<br />

facility far o Short Deerry Cycle<br />

Provisions are being made for processing <strong>the</strong><br />

MSRE fuel salt for uranium recovery on <strong>the</strong> shortest<br />

possible cycle afler shutdown of <strong>the</strong> MSRE in early<br />

1968. The flush salt Twill be processed first, and<br />

<strong>the</strong>n <strong>the</strong> fuel salt will be treated with H,-HF to<br />

establish <strong>the</strong> oxygen concentration. Allowing time<br />

for <strong>the</strong>se operations, <strong>the</strong> fuel salt may be fluorinat-<br />

ed after 35 days (initial plans called for a 90-day<br />

coaling time). This shorter cooling time requires<br />

some modification of <strong>the</strong> processing facility at <strong>the</strong><br />

MSRE. The higher concentration of iodine requires<br />

improvement of <strong>the</strong> off-gas system, and <strong>the</strong> pres-<br />

12<br />

ence of molybdenum requires increased shielding<br />

around <strong>the</strong> UF, product absorbers,<br />

26. Preparation of 233~~,-7~i~<br />

~uei<br />

Concentrate fur <strong>the</strong> MSWK<br />

Refueling and operating <strong>the</strong> MSRE with 23 3U<br />

fuel early in 1968 is planned; this will require<br />

approximately 40 kg of 233U as 233UF -7LiF (27<br />

and 73 mole X) eutectic salt. This fuel concen-<br />

trate will he prepared in a cell in <strong>the</strong> 'TURF build-<br />

ing because of <strong>the</strong> radiation from <strong>the</strong> 232U daugii-<br />

ters in <strong>the</strong> 233U The uranium will arrive as an<br />

oxide ia cans, which will be opened and dumped<br />

into a reaction vessel. Lithium fluoride will be<br />

added, and <strong>the</strong> mixture will bp treated with hydra-<br />

gen and finally HF to produce <strong>the</strong> eutectic melt.<br />

Three 12-kg 233U batches will be prepared for <strong>the</strong><br />

major additions to <strong>the</strong> barren MSRE salt and one<br />

7-kg 233U batch will be loaded into 60 enriching<br />

capsules. The engineering design is almost com-<br />

plete, and most of <strong>the</strong> equipment has been fabri-<br />

cated.

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