To confirm the effectiveness of the UNEX-extractant under operating con­ditions, special tests of a prolonged continuous process duration (66 hours) were performed at RI and the Idaho National Laboratory. Over this period of time, the extractant made 89 recycles. The tests demonstrated that:

• extractable components are not accumulated in the recycled extractant;

• concentration of the most water-soluble component (PEG-400) can be maintained by its addition to the fed strip agent;

• increase in the degree of radionuclide concentration in the strip product is attained due to a temperature rise to 60°C at the stripping operation stage;

• rates of recovery of radionuclides (Cs — 99.78%, Sr — 99.984%, Eu — 99.997%) allow the transfer of the raffinate into the LLW category;

• elimination of the extractant regeneration operation from the improved flowsheet did not result in accumulation of extractable components in the recyclable organic phase;

• as far as metal macroimpurities were concerned, Ba and Pb were almost completely extracted during the tests, whilst Ca, Fe, K, Mo were partially extracted.

The results presented have already been partly published [26-29]. All the UNEX tests up to 2001 year were reviewed in [30] (see Table 9.17). The UNEX process was patented in Russia and in the US. [31]

In order to verify the UNEX process within the framework of DOE Project EM-50/JCCEM under contract with Sandia National Laboratories’ “Testing and demonstration of UNEX-process technology”, an extraction setup was created in hot cells at the Central Plant Laboratory of the Radiochemical Plant (RChP) at Mining Chemical Combine (MCC), Zheleznogorsk, Russia. The EZ-33 centrifugal contactors developed and manufactured by NIKIMT were the main extraction process equipment used. The UNEX-process was tested in two stages — on simulated and actual waste feeds [32]. The experimental objective was to: determine the techno­logical parameters of the UNEX process when treating simulated and actual liquid HLW; to investigate the distribution of stable and radioactive

Flowsheet test Test 1 Test 2 Test 3 Test 4 Test 5 Test 6 Solvent wash 3 M HN03 5 M HN03 2 M HN03 None None None Radionuclide Cs 99.4% 137Cs 99.95% 137Cs 99.4% Cs > 97.5% Cs 99.95% 137Cs 99.99% removal Sr 99.97% soSr 99.985% soSr 99.995% Sr > 99.993% Sr > 99.999% soSr 99.73% efficiencies Eu 4.3% to >99.92% Alpha 95.2% Alpha 99.96% Eu 17.2% to 34.1% Nd > 98.3% Ce > 99.6% Alpha > 99.9% Matrix metal Zr 52% Zr >97.7% Zr 87% Zr < 6.4% Zr 3% Zr 0.7% Removal Mo <3.1% Mo 19% Mo 32% Mo < 19.2% Mo > 2% Mo 12% efficiencies Fe 10% Fe 6.9% Fe 8% Fe < 13.2% Fe 9% Fe 2% Ba 99.5% Ba >87% Ba > 99% Ba > 99.6% Ba 99.7% Ba, Pb 100% Pb 99.8% Pb >98.5% Pb > 98.8% Pb > 99.94% К 50% Mn 23% К 20% К 17% Ca 10% К 28% Notes Two tests with 4 hour test with 3 hour test with 66 hr run time w/ 4 hour test with 3 hour test light phase solvent recycle. solvent recycle. solvent recycle. solvent recycle. with solvent solvent; flooding Flooding was No No precipitation No precipitation recycle. No and precipitation observed in the precipitation or or flooding was or flooding was precipitation observed in strip actinide strip flooding was observed. observed. or flooding section of initial section. observed. was test. observed.

aGN = guanidine nitrate, bGC = guanidine carbonate, CAHA = acetohydroxamic acid, dDTPA = diethylenetriamine pentaacetic acid, eANN = aluminium nitrate.

elements over the various stages of the extraction units; to check the opera­tional reliability of the EZ-33 centrifugal contactors with high radiation loadings under prolonged operational conditions.

Initially a simulated solution, corresponding to real process waste but composed of stable macrocomponents, was treated. The duration of the experiments was 48 hours and the volume of treated solution was 29 l. The process flowsheet of the setup involved the following operations: extraction of target elements (18 stages); extract scrubbing (2 stages); combined strip­ping of target elements (16 stages).

In the second part of the experiment, actual high-level waste from a radiochemical plant (Pr. 501) was used, with the following composition: Pu

— 152 mg/l; U — 102 mg/l; Np — 0.38 mg/l; Am — 68 mg/l; Th — < 0.5 mg/l; Fe

— 645 mg/l; Cr — 340 mg/l; Mn — 1,2 g/l; Ni — 510 mg/l; Pb — 6 mg/l; Al — 220 mg/l; Ba — 12 mg/l; Ca — 230 mg/l; F-ion — 2,7 g/l; Si — 100 mg/l; HNO3

— 69 g/l; NaNO3 — 47 g/l; 90Sr — 0.3 Ci/l; gross a-activity — 1,4 Ci/l.

The specific volumetric activities of the radionuclides in Bq/l were:-125Sb

— 4.89*107; 134Cs — 1.02*107; 137Cs — 5.3*108; 155Eu — 5.82*108; 154Eu — 3.37*107; 144Ce — 9.31*109; 60Co — 2.67*107; 106Ru — 2.07*109. In the second test 77 l of actual waste were treated. The time operating on the actual solution was 128 hours, and the solvent made 51 recycles. Results of checks on the extrac­tion properties of the recycled solvent and its chemical analysis during the test did not reveal any considerable deviations from initial characteristics. Some decrease of Sr and Cs distribution coefficients were noted — by 10 and 18.5% of their initial values, respectively; however, this fact did not affect the process parameters as a whole.

On treatment of actual waste, the recovery rates of target radionuclides, as calculated from their content in the raffinate, was: Pu — 99.76%; Am — >99.02%; Sr-90 — 99.99%; U — >99.0%; Eu-155 — >99.91%; Eu-154 — >98.75%; Cs-137 — 99.95%; Ce-144 — >99.98%. Al, Cr, Ru and Sb are not recovered totally (their content in the strip product is below detection levels), while Pb and Ba are extracted and are removed with the strip product. Up to 20% Ca, 25% Fe, 23% Mn, 5% Co and 3% Ni (measured as a percentage of their content in the feed solution) are also withdrawn with the strip product.