Simulated salt waste

Leaching of salts Simulated salt waste streams (both supernatant and sludge)... 

Water used in the experiments was doubly distilled and passed through an ion exchange unit. The conductivity was approximately 1 x 10"6 S/m. Simulated HLLW consisted of 21 metal nitrates in an aqueous 1.6 M nitric acid solution as shown in Table 1 and was supplied by EBARA Co. (Tokyo, Japan). Concentrations were verified by AA for Na and Cs with 1000 1 dilution and by ICP for the other elements with 100 1 dilution. Total metal ion concentration was 98,393 ppm. The experimental apparatus consisted of nominal 9.2 cm3 batch reactors (O.D. 12.7 mm, I.D. 8.5 mm) constructed of 316 stainless steel with an internal K-type thermocouple for temperature measurement. Heating of each reactor was accomplished with a 50%NaNO2 + 50% KNO 2 salt bath that was stirred to insure uniform temperature. Temperature in the bath did not vary more than 1 K. The reactors were loaded with the simulated HLLW waste at atmospheric conditions according to an approximate calculated pressure. Each reactor was then immersed in the salt bath for 2 min -24 hours. After a predetermined time, the reactor was removed from the bath and quenched in a 293 K water bath. The reactor was opened and the contents were passed through a 0.1 pm nitro-ceflulose filter while diluting with water. Analysis of the liquid was performed with methods in Table 1. Analysis of filtered solids were carried out with X-ray diffraction with a CuK a beam and Ni filter. Reaction time was defined as the time that the sample spent at the desired temperature. Typical cumulative heat-up and cool-down time was on the order of one minute. Results of this work are reported in terms of recoveries as defined by ... 

Because stripping back into a salt solution would not be acceptable in this case, ABEC resins were tested for their ability to remove TcO.r from the simulated Hanford wastes. The results based on batch contacts... 

Simulated solutions of Russian waste and INL (Idaho National Laboratory, USA) waste were used as feed solutions. Extraction of cesium was 98.4%, and of strontium, 98.1%. A problem with low solubility of the crown ethers (<180 mg/L for all type of solutions) was shown in the tests. The authors pointed out the main positive features of the proposed flowsheet a salt-free strip product with very low nitric acid concentration (<0.1 M), possibility to extract both cesium and strontium, and low losses of extractants. A last step in the modification of this solvent was the addition of polyalkylphosphonitrilic acid to the mixture of crown ethers.65 Positive results were obtained for extraction of not only Cs and Sr, but also MAs from simulated HLW. 

This assessment included results of leach tests with glass contain ing either actual(2,3) or simulated(4) radioactive waste in leachants that simulated groundwaters of possible geologic repositories salt, basalt, tuff, and granite. In support of the environmental... 

Leaching of contaminants Eission products Tc, partitioned from high-level waste tanks, debris from contaminated pipes from K-25 plant at Oak Ridge Cs, from salt supernatant and sludge, silico-titanates, and wastewater Radioactive components Ra, Fernald silo waste, transuranics, simulated and actual Rocky Flats ash waste, wastewater... 

Synthesis efforts in this project to date focused on producing a sorbent with increased 90Sr and actinide removal performance. Specific types of sorbents produced and evaluated for removal performance include sodium nonatitanate, metal-substituted sodium nonatitanates, crystalline silicotitanates, titanosilicates having a pharmacosiderite structure and heteropolyniobates. Table 2 provides a list of sorbent materials tested. Performance testing featured a simulated waste solution comprised of the major anionic components of SRS waste solutions as the respective sodium salts and specific amounts of strontium and actinide elements. 

Table 3 provides a summary of the simulated waste solution used to evaluate new sorbent materials. Testing also featured actual tank waste material diluted to the same sodium concentration (5.6 M) as that provided by the simulated waste solution in Table 3. The actual waste solution, however, contains a different salt composition than that of the simulated waste.

Service testing to simulate ash/salt deposit corrosion is of importance to a number of industries. The fossil-fired power generation industry must deal with what is called "fuel ash corrosion fixrm sulfur- and vanadium-containing fuels and alkali, chlorine, and sulfur in coal. The gas turbine industry must deal with "hot corrosion" problems arising fixjm sulfur in fuel and sodium salts from ingested air. Waste incineration environments can become even more complex with refuse containing sulfur, chlorine, phosphorus, and numerous metallic elements. 

Figure 6 Influence of the Presence of Nitrate Salts on the Permeation of Cs from Simulated Medium-active Liquid Wastes to Deionized Water by 1,3-Calix[4]-A/5 -crown-6 2,2l,3-Calix[4]-W -o-benzo-crown-6 5 and l,3-Calix[4]- /y-naphthyl-crown-6 6 (10 M) in 2-NPOE.

Demixing of a ternary solution water / chloroform / TBP. Assisted extraction of the U02(N03)2 salt The demixing of more concentrated solutions has also been simulated, to investigate what happens beyond the first layer formed at the interface, A first "computer demixing experiment" of I ns has been performed with a chloroform / water binary mixnire containing 30 TBP molecules as solute. In relation with the PUREX treatment of nuclear wastes 03,ii4 second simulation of 5.1 ns has been repeated with [U02(N03)2]5 as solute. Indeed, in this process, a nitric acid solution of the cations is contacted with a solution of TBP in kerosene into which U02(N03)2(TBP)2 is extracted... 

The properties of the fuel salt used in these simulations are summarized in Table 7.2. The fuel salt considered in the simulations is a molten binary fluoride salt with 77.5 mol% of lithium fluoride the other 22.5 mol% is a mixture of heavy nuclei fluorides. This proportion, maintained throughout the reactor evolution, leads to a fast neutron spectrum in the core as shown in Fig. 7.2. Thus this MSFR system combines the generic assets of fast neutron reactors (extended resource utilization, waste minimization) and those associated with a liquid-fueled reactor. 

The knowledge of real-phase behaviour of electrolyte solutions provides a basis for the design and the simulation of many processes in biological and chemical engineering. Otherwise, salts are systematically used for the recovery of biomolecules and as auxiliary material in separation units. Thus, technical applications of systems containing electrolytes can be found in waste-water and drinking-water treatment, fertilizer production,... 

Other leach methods, used in the European Community, call for chloride salts and nitrate salts to be added to the deionized-water leaching fluid. These leachates are believed to give a more accurate simulation of actual leaching processes in waste disposal areas in these geographic areas. Indeed, acid rain consisting of primarily nitrate anions, and soils with high chloride contents are better represented by these fluids than by the USEPA or other leachate compositions. 

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