Blanket salts

A fwo fluid design uses a fuel salt that carries the fissile UF4, but no fertile material and a separate blanket salt for fhe fertile ThF4. As is produced in the blanket, it is transferred to the fuel salf by a relatively simple fluorination process. Fluorination is accomplished by bubbling fluorine gas through the salt, thereby converting UF4 to volatile UFg. Gaseous UFg is collected and converted back to UF4 by a well-established process before being added to fuel salf. 

Temperature coefficient of reactivity -17.5 pcm/K —14.6 pcm/K -4.3 pcm/K blanket salt positive void coefficient -2.4 pcm/K recent indications maybe positive -6.7 pcm/K... 

Mid-1960s efforts at ORNL sought to bypass this issue by using two separate liquids, a fuel salt containing the produced and a blanket salt containing the thorium to produce... 

The blanket salt would require periodic fluorination to remove the produced (although often rapid Ta removal was also suggested). The fuel salt, in a continuous side stream, would be first fluorinated to remove and then salt processed by vacuum distillation where at high temperature (1(XX)°C) the carrier salt would evaporate off and leave behind most fission products in the still bottoms. 

Blanket salt containing fertile material, typically Thp4, can be employed in another fashion. In this more commonly proposed area, designs call for two separate fluids, one blanket and one fuel salt both flowing in and out of the core region. 

Another MSR variation was conceived wherein a fertile blanket salt is only used to partially or fully surround the core region which contains the fuel salt—a mix of fertile and fissile. This was termed a one-and-a-half fluid design by ORNL. 

The decontaminated diluent salt would then be used to produce make-up fuel for the FUJI-233Um. Specifically, the fertile salt would be charged to the storage tank of the AMSB to keep the U content in the target/blanket salt constant, such as -0.5 mol. %. The salt taken from the storage tank will be enriched by adding the removed U and sent as the fuel-salt to the FUJI power station sites, see Fig. XXX-4. 

D-plan Dry reprocessing of spent fuel and target/blanket salts including not only molten salt but solid fuel of ordinary reactors such as the LWR, fast breeder reactors (FBR), heavy water reactors (HWR), etc. for producing molten fluoride fuel salt of the FUJI or target/blanket salt of the AMSB ... 

Fertile blanket salt, initial composition (mol%) LiF-ThF4 (77.5-22.5%)... 

Other systems that also contain radioactive materials have to be studied, in particular the fertile blanket salt system, including the storage and processing of the associated gases, and all of the related intersystem transfers. 

Fig. 14-4. Initial fuel regeneration in two-region, homogeneous, molten fluoride-salt reactors fueled with U . Total power, 600 Mw (heat) external fuel volume, 339 ft core and blanket salts No. 1.

The effects of changes in the compositions of the fuel and blanket salts are indicated in the following description of the results of a series of calculations for which salts with more favorable melting points and viscosities were assumed. The BeFa content was raised to 37 mole % in the fuel salt... 

Blanket salt No. 2 13 mole % ThF4 + 16 mole % BeF2 + 71 mole % LiF. 

Fuel salt 53 mole % NaF 4- 46 mole % BeF2 - - 1 mole % (TI1F4 4- UF4). Blanket salt 58 mole % NaF 4- 35 mole % BeF2 4- 7 mole % TI1F4. 

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