Dissolution behavior of FFTF fuel

These tests, using FFTF fuel, show that fuel fragmentation and dislodgement from the cladding occurs rather early in the dissolution. The large surface areas of the fuel fragments will lead to rapid dissolution, certainly more rapid than would be expected if the fuel remained within the cladding and dissolved from the open ends, as is sometimes assumed

Secondary cleanup of Idaho Chemical Processing Plant solvent

Solvent from the Idaho Chemical Processing Plant (ICPP) (operated by Westinghouse Idaho Nuclear Company, Inc.) has been tested to determine the ability of activated alumina to remove secondary degradation products - those degradation products which are not removed by scrubbing with sodium carbonate

Solubility of krypton in hydrofracture grout at elevated pressures

The solubilities of krypton in water, simulated waste solution, and simulated grout at about 25/sup 0/C and to pressures of 150 atm have been determined. The results of these studies show that preliminary calculations of krypton solubility based on the aqueous component of the hydrofracture grout were overly pessimistic. The volume of noble gas generated annually by the reference reprocessing plant would be soluble in the annual hydrofracture grout injection at ORNL at about 10 atm. The amount of krypton in the gas phase would depend on the amount of air in the hydrofracture grout mixture. At 34 atm, and with a small air volume relative to the injected krypton, the krypton would constitute about 30% of the gas bubbles. The disposal of krypton via injection with hydrofracture grout seems to be a viable process. The next logical steps would be to determine the krypton diffusion rate at injection conditions, and possibly to perform a test injection. At present, the schedule for future work is uncertain since funds for this project have been reduced significantly

Calculated and experimental studies of nonequilibrium solvent extraction of uranium-thorium and uranium-zirconium

The nonequilibrium simultaneous transfer of ions in solvent extraction has been examined experimentally and these results have been compared with the calculated behavior determined using transfer rate constants. In the Th-U system when transferring from 2 M HNO/sub 3/ into 30% tributyl phosphate (TBP) in normal hydrocarbon diluent (NPH), the thorium approaches equilibrium faster than uranium over much of the transfer region. Thus, nonequilibrium operation will not increase the separation factor between uranium and thorium over that attained at equilibrium. However, in the Zr-U system when transferring from 3.4 M HNO/sub 3/ into 30% TBP-NPH, the separation factor between uranium and zirconium is increased over that attained at equilibrium over much of the transfer region due to the relatively slow transfer of ZrOH/sup 3 +/, one of the two extractable forms of zirconium. Thus, the uranium-zirconium separation factor can potentially be increased by the use of short, nonequilibrium residence times

Improved Purex solvent scrubbing methods

Studies of hydrazine and hydroxylamine salts as solvent scrubbing agents that can be decomposed into gases are summarized. Results from testing of countercurrent scrubbers and solid sorber columns that produce lesser amounts of permanent salts are reported. The status of studies of the acid-degradation of paraffin diluent and the options for removal of long-chain organic acids is given

Solvent degradation and cleanup: a survey and recent ORNL studies

This paper surveys the mechanisms for degradation of the tributyl phosphate and diluent components of Purex solvent by acid and radiation, reviews the problems encountered in plant operations resulting from the presence of these degradation products, and discusses methods for minimizing the formation of degradation products and accomplishing their removal. Scrubbing solutions containing sodium carbonate or hydroxylamine salts and secondary cleanup of solvents using solid sorbents are evaluated. Finally, recommendations for improved solvent cleanup are presented. 50 references, 4 figures, 3 tables

Recovery of uranium from 30 vol % tributyl phosphate solvents containing dibutyl phosphate

A number of solid sorbents were tested for the removal of uranium and dibutyl phosphate (DBP) from 30% tributyl phosphate (TBP) solvent. The desired clean uranium product can be obtained either by removing the DBP, leaving the uranium in the solvent for subsequent stripping, or by removing the uranium, leaving the DBP in the solvent for subsequent treatment. The tests performed show that it is relatively easy to preferentially remove uranium from solvents containing uranium and DBP, but quite difficult to remove DBP preferentially. The current methods could be used by removing the uranium (as by a cation exchange resin) and then using either an anion exchange resin in the hydroxyl form or a conventional treatment with a basic solution to remove the DBP. Treatment of a solvent with a cation exchange resin could be useful for recovery of valuable metals from solvents containing DBP and might be used to remove cations before scrubbing a solvent with a basic solution to minimize emulsion formation. 6 refs., 9 figs

Dissolution of low burnup Fast Flux Test reactor fuel

The first Fast-Flux Test Facility reactor fuel (mixed (U,Pu)O/sub 2/ composition) has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 1997/sup 0/C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 95/sup 0/C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 29/sup 0/C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 95/sup 0/C dissolution contained the equivalent of 198 mg of /sup 239/Pu per 100 g of hulls, while the cladding from the 29/sup 0/c experiments contained only 0.21 mg of /sup 239/Pu per 100 g of hulls. 9 references, 5 figures