LMFBR fuel process

R C LLOYD and E. b, CLAYTON, 6itb Safety Data Applicable to LMFBR Fuel Processing, Nuct ScL Eng., 59,21 (1976).. ... 

Plants in the UK, USSR and France are now reprocessing irradiated UO2/PUO2 fuels and LMFBR fuel reprocessing has been the subject of international conferences. " The plant at Cap la Hague, France, employs a 30% TBP solution and no U/Pu separation is undertaken, so that a mixed U/Pu product is obtained. Fluoride is added to the process feed to complex zirconium and suppress its extraction." The Dounreay plant in the UK employs a 20% TBP/OK solution and uses sulfuric acid to effect the U/Pu separation. TBP poorly extracts Pu or U from sulfuric acid solutions, but in mixed HNO3/H2SO4 the equilibria shown in equations (206) and (207) must be considered. The equilibrium constants for these reactions, Kj, and K i, are given... 

The present work at Rocky Flats is an extension of the Argonne work and is directed to development of a proliferation resistant pyrochemical process for LMFBR fuels. This article describes a conceptual pyrochemical process and preliminary engineering concepts for coprocessing uranium and plutonium in spent LMFBR core-axial blanket and radial blanket fuels using the Salt Transport Process. 

Criteria used to develop the conceptual pyrochemical process for coprocessing uranium and plutonium from spent LMFBR fuel include the following ... 

Changes in fuel assembly size will increase or decrease plant throughput. Processing more than one fuel assembly per batch will require an increase in vessel size. Additional capacity also can be achieved by establishing parallel operations. The single fuel assembly per batch operation can process up to 200 tons per year of mixed oxide LMFBR fuel or up to eight fuel assemblies per day. 

Aqueous Processing of LMFBR Fuels, Technical Assessment and Experimental Program" ORNL-4436, June 1970. 

Baumgartner, F., Ochsenfeld, W., and Schmieder, H., "Development Work on Reprocessing of Oxidic LMFBR Fuel by the Purex Process," Paper presented at 82nd AIChE Mtg., Atlantic City, N.J., 1976. 

Richardson, G. L. The Effect of High Solvent Irradiation Exposure on TBP Processing of Spent LMFBR Fuels , HEDL-TME 73-51, 1973 Appendix A. 

Extraction Process Having Application to the Reprocessing of LMFBR Fuels, Report ORNL 746, 1972. 

The following discussion of reprocessing LMFBR fuels outlines the principal process steps, lists the main problem areas, and discusses possible solutions. Since 1973, international dissemination of reprocessing information has been restricted. This discussion of reprocessing LMFBR fuel is thus less complete and less up to date than would be desired. 

Figure 10.29 shows the principal steps in applying the Purex process to irradiated LMFBR fuel, step 7 of Fig. 10.28. The flow scheme and the compositions and locations of solvent, scrubbing, and stripping streams have been taken from the process flow sheet of a 1978 Oak Ridge report [Oil] describing a planned experimental reprocessing facility designed for 0.5 MT of uranium-plutonium fuel or 0.2 MT of uranium-plutonium-thoiium fuel per day. As that report gave process flow rates only for the uranium-plutonium-thorium fuel. Fig. 10.29 does not give flow rates for the uranium-plutonium fuel of present interest. This flow sheet shows the codecontamination step, in which flssion products are separated from uranium and plutonium the partitioning step, which produces an aqueous stream of partially decontaminated...

Figure 10.29 Principal steps in Purex process for LMFBR fuel. F.P. = fission products HAN = hydroxylamine nitrate ...

G9. Goode, J. H., and S. D. Clinton Aqueous Processing of LMFBR Fuels—Technical Assessment and Experimental Program Definition, Report ORNL-4436, secs. 4.4 and... 

G16. Gronier, W. S. Calculation of the Transient Behavior of a Dilute-Purex Solvent Extraction Process Having Application to the Reprocessing of LMFBR Fuels, Report ORNL 746, Apr. 1972. 

Experiments were recently completed to provide data on the effectiveness of soluble poisons on the criticality of plutonium-uranium solution mixtures. Although some data have been reported on Pu solutions, none have been available for U + Pu solutions. The current experiments provide data on systems typical of LMFBR fuel compositions. Gadolinium was chosen for this study because of its high neutron cross section, high solubility, and compatibility in the separation process i.e., gadolinium is more easily separable from Pu than cadmium or boron. The data can have applications in criticality control and prevention and serve also as experimental benchmarks for validating calculational techniques and cross-section sets. 

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