Plutonium experimental design

Sorption Prediction Equations. Equations predicting radioelement distribution coefficients, K s, as arithmetic functions of component concentrations were obtained for sorption of strontium, neptunium, plutonium, and americium on two Hanford sediments. These equations, presented in Table VH and derived from statistical fits of Box-Behnken experimental designs, were generated for strontium in terms of sodium ion, HEDTA, and EDTA concentrations. Prediction equations for neptunium and plutonium sorption were derived from NaOH, NaA102, HEDTA, and EDTA concentrations. Americium sorption prediction equations were based on NaOH, HEDTA, and EDTA concentrations. 

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...

Switzerland Within the framework of the CAPRA project, the fuel option for amplified plutonium consumption is being studied. In the area of materials for actinide transmutation, the following tasks has been completed in 1994 (1) preparatory experiments and solubility tests for (Ui, PUJ O2 (0.25 < x < 0.65 and for (Uj., PU,J N (0.25 < x< 0.75), as possible materials for the efficient fission of plutonium in a fast neutron flux (2) fabrication of pure PuN-microspheres for ceramic-metal fuel (3) design calculations for sphere-pac segments, based on the idea of a ceramic-metal fiiel (4) material preparation of (U, Zr) N and pelletization tests of TiN and (U, Zr)N for the irradiation experiment in the reactor PHENIX (5) experimental preparations of (Ce,U) O2, (Ce,U,Pu)02 and (Ce, PU)02 for the CAPRA core with lower Pu content, and (6) cleaning of americium from waste streams of the plutonium separation equipment (extraction chromatography). 

Two plutonium compositions were used, as listed in Fig. 1, vdiich also shows the designs of the coextruded aluminum fuel assemblies. Hie experimental lattices consisted of 19 of these fuel assemblies in hexagonal arrays at 12.12- or 14.00-in. pitches. To provide an easily calculated radial boundary, the fuel lattices were extended for two additional rings with nonfissioning iron-lithium assemblies having very nearly the same absorption cross section as the fuel. 

For plutonium re cle desi calculations, the same precisions are wanted as those required and usually achieved for uranium lattices 0.5 to 1% on k, 4 to S% on power distributions and 10% on control rod worth. To check if they can be acMeyed by the design codes, experimental techniques should allow the determination of the involved characteristics within error margins substantially smaller than the wanted theoretical precisions. 

Criticality safety data have been accumulated from around the world, with the help of such unclassified reports as the three-volume publication by Koponen et al., which contains indexes, concordances, and abstracts of reports published up to 1978. We have analyzed the publication and found, as shown in Fig. 1, that the amount of experimental data for criticality safety is less than 50% of the whole criticality data base and that 13, 16, and 28% of the data are on mixtures of plutonium and uranium, solutions of plutoiuum, and solutions of uranium, respectively. In those data for mixtmes and solutions, we could not find data suffldent-for reak>nable and effective design and operation of out-of-reactor nuclear fadlities for light water reactor (LWR) and fast brewer reactor (FBR) fuel cydes. ... 

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