Desorption of Technetium from Active Carbon with Alkaline Thiocyanate Solution

The desorption behavior of Tc from active carbon was studied as a fundamental study on separating Tc from high-level radioactive liquid waste. Technetium adsorbed from 0.5 M HNO, solution could be desorbed more effectively with alkaline thiocyanate solution than with HN0₃. The adsorption behavior of Tc on the active carbon previously treated with alkaline thiocyanate solution was also studied for the purpose of a recycle use of the active carbon.

The distribution coefficient of Tc, Kd, depended sensitively on pH in the presence of KSCN. The value of Kd exceeded 10⁵ ml/g below pH 2 and became less than 10 ml/g above pH 4. It was found by spectrophotometric analysis that this large difference in Kd was due to the change of the chemical form of Tc. Below pH 2, Tc was adsorbed in the form of thiocyanate complexes, [TC(NCS)₆]^2- and [Tc(NCS)₆]^3-. Above pH 4, these complexes were hardly formed and TcO₄ ^? was desorbed by anion exchange reaction with thiocyanate ion.     

Adsorption Behavior of Americium on Zeolites

Zeolites are expected to be used in buffer materials as an admixture to bentonite to increase their ion exchange ability. The behavior of ion exchange adsorption of Am³⁺ on various zeolites was examined by batch and column methods. The distribution coefficient of Am³⁺, K_d, increased with equilibrium pH from 1.5 to 4.0. L zeolite and mordenite having large pore size yielded relatively large values of K _d over 10³ at pH 3.0. The K _d for L zeolite was unaffected by the concentration of coexisting cations, i.e. Na⁺ up to 0.1 M, and K⁺, Ca²⁺ and Mg²⁺ less than 0.01 M. The Am³⁺ ion was not detected in the effluent passed through the L zeolite column up to 90 bed volumes, even when the feed solution contained a relatively large amount of Na⁺-0.5 M. Americium ion adsorbed on zeolites could be quantitatively eluted with 0.1 M nitric acid.     

Characteristics of Cast Stone cementitious waste form for immobilization of secondary wastes from vitrification process

The high-temperature in vitrification process of radioactive wastes could cause radioactive technetium (⁹⁹Tc) in secondary liquid wastes to become volatile. Solidified cementitious waste forms at low temperature were developed to immobilize radioactive secondary waste. This research focuses on the characterization of a cementitious waste form called Cast Stone. Properties including compressive strength, surface area, phase composition, and technetium leaching were measured. The results indicate that technetium diffusivity is affected by simulant type. Additionally, ettringite and AFm (Al₂O₃–Fe₂O₃–mono) main crystalline phases were formed during hydration. The Cast Stone waste form passed the qualification requirements for a secondary waste form, which are compressive strength of 3.45 MPa and technetium diffusivity of 10^?9 cm²/s. Cast Stone was found to be a good candidate for immobilizing secondary waste streams.     

Application of “Hydration Model” to Evaluate Gas Phase Transfer of Ruthenium and Technetium from Reprocessing Solutions

In order to evaluate the amounts of gas phase transferred ruthenium (Ru), and technetium (Tc), simulations were made for the continuous evaporator used in a reprocessing plant to concentrate high level liquid waste. The concentrations and activities of nitric acid and water, which controlled the reaction rate and gas-liquid equilibrium in the evaporator solution, were evaluated using the previously developed “Hydration Model”. When the feed solution contained 2.7 M (=mol/dm³) of nitric acid, the nitric acid concentration in the evaporator solution reached its maximum at the concentration factor (CF) of 6 (CF: concentration ratio of FPs in evaporator and feed solutions). The activities of nitric acid and water were saturated at values of 0.01 and 0.43, respectively, after the CF reached 6. The simulation predicted decontamination factors DFs of 2×10⁵ and 8×10³ for Ru and Tc, respectively, for a typical evaporation conditions with an operational pressure of 6,700 Pa, and FPs of 0.02 to 1.4 M. The simulation results agreed with the verification experiment, which will be described in next paper, within a factor of 2 for the amount of gas-phase transferred Ru during evaporation. The factor for the amount of gas-phase transferred Tc was estimated as 5 from the measurement error in the gas-liquid equilibrium constant.     

Adsorption of Fission Products on Alumina in Organic Solutions

Metal nitrates are readily adsorbed on alumina in organic solutions such as acetone and n-butylacetate. Adsorption isotherms for such compounds as uranyl nitrate, thorium nitrate and cerium nitrate (III) were examined, and the adsorption was found to have occured monomolecularlly on the surface of the alumina.

The weakest adsorption strength on alumina was found to be that of uranyl nitrate, and the adsorbed uranyl nitrate could be substituted very easily by other nitrates such as of fission products.

The logarithmic decrease of the Kd values of fission products on alumina with increase of uranyl nitrate concentration in n-butylacetate was found to correspond to the increase of the solubility of nitrates of fission products (Ce, Sr and Zr) with increase of uranyl nitrate concentration.     

Sorption Behavior of Carrier-Free Technetium-95m on Minerals,Rocks and Backfill Materials under both Oxidizing and Reducing Conditions

The sorption behavior of ^95mTc on various sorbents, including minerals, rocks and backfill materials was studied in both oxidizing (0.16 M-mol·dm^?3-NO^? ₃), reducing (0.1 M and NaBH₄) solutions. Under oxidizing condition, distribution coefficient K _d ranged from 0 to 4.5cm³·g^?1 for minerals and rocks. Under reducing condition, the values exceeded 140 cm³·g^?1 for active carbon, and ranged from 13 to 70 cm³·g^?1 for minerals and rocks, and in both cases showing a tendency to decrease with increase of coexisting salt concentration and of sorbent particle size. Similarly under reducing condition, the Freundlich sorption isotherms of Tc on active carbon and on bentonite presented a break in slope when the concentration of unsorbed Tc remaining in solution lowered to a certain level. The slopes before and after the break were more moderate on active carbon and on bentonite than on other minerals and rocks, in which case no break in slope took place on the isotherms within the range covered in the present experiment.     

Feasibility studies on the selective separation of fission palladium(II) by isoHex-BTP/SiO2-P adsorbent from HLLW

Aiming at the selective recovery of fission palladium(II) from high-level liquid waste (HLLW), the silica/polymer (SiO₂-P)-based isoHex-BTP adsorbent (isoHex-BTP/SiO₂-P) was synthesized by impregnating complexing agent isoHex-BTP into the multiporous SiO₂-P inert support. The feasibility of separation of Pd(II) from HLLW by isoHex-BTP/SiO₂-P was evaluated by batch experiment method. The results showed that isoHex-BTP/SiO₂-P exhibited much higher adsorption selectivity for Pd(II) than the other fission products, even Am(III) and Pu(IV) presented in HLLW. The ideal nitric acid concentration for the adsorption of Pd(II) by the adsorbent was shown to be ≥2 mol dm^–3. The adsorption of Pd(II) fits well to the pseudo-second-order kinetic model and Langmuir isotherm model. Quantitative Pd(II) desorption was achieved by using 0.5 mol dm^?3 SC(NH₂)₂ - 0.1 mol dm^?3 HNO₃ solution.     

Separation of Heat-Generating Nuclides from High-Level Liquid Wastes through Zeolite Columns

Heat-generating nuclides Cs and Sr were separated from simulated high-level liquid wastes (HLLW) by successive adsorption on columns of ferrierites (F) and zeolite A. Adsorbed Cs and Sr were efficiently eluted with NH₄NO₃ and EDTA solutions, respectively, yielding the recovery over 96%. A simulated waste solution containing 29 components was denitrated with a formic acid up to pH 7.92. The amounts of nuclides adsorbed from this denitrated solution were experimentally estimated to be 0.33 mmol Cs/g·F and 0.19 mmol Sr/g·A, respectively.     

Preliminary Observations of Single Crystals of UO2 by Transmission Electron Microscope

Adsorption behavior of inorganic ions were systematically surveyed through radiochemical methods. Adsorbents used were zirconium phosphate, silica gel and charcoal, and the aqueous phases were 0.01N-10N nitric acid and 0.01M-10M ammonium nitrate solutions. The distribution ratios of about 50 chemical elements from Na to Cm were determined by batch method using radioactive tracers.

In the zirconium phosphate system, Cs shows the highest Kd value and most bivalent species are not adsorbed.

A few elements are adsorbed on silica gel from acid solution. High Kd values are shown by Zr, Nb and Pa.

In the charcoal system, most chemical elements are adsorbed to some extent except for alkali and alkaline earth ions.     

Adsorption mechanism of silica/polymer-based 2,6-bis(5,6-diisohexyl-1,2,4-triazin-3-yl)pyridine adsorbent towards Ln(III) from nitric acid solution

2,6-Bis(5,6-diisohexyl-1,2,4-triazin-3-yl)pyridine (isoHexyl-BTP) is a nitrogen-donor chelating ligand which shows high extraction selectivity for minor actinides (MA) over lanthanides (Ln). We synthesized a macroporous sillica/polymer-based isoHexyl-BTP adsorbent (isoHexyl-BTP/SiO₂-P) for the separation of MA(III) from Ln(III) in high-level liquid waste. The work focused on the isoHexyl-BTP/SiO₂-P adsorption mechanism towards Ln(III) in nitric acid solution and the coordination chemistry between Ln(III) and isoHexyl-BTP/SiO₂-P through batch adsorption experiments, extended X-ray absorption fine spectroscopy, acid-base titration and ion chromatography. It was found that both H⁺ and NO₃^? directly participated in the adsorption reaction. For the middle and heavy Ln(III), Ln(isoHexyl-BTP/SiO₂-P)₃(NO₃)₃·3HNO₃ was supposed as the main product of the adsorption. The Ln(III)-N (Ln(isoHexyl-BTP/SiO₂-P)₃³⁺) bond length of the first coordination layer decreased as the atomic number of Ln(III) increased, which explains the increased adsorption affinity of isoHexyl-BTP/SiO₂-P towards middle and heavy Ln(III) in relatively high concentration nitric acid as the Ln(III) atomic number increased.     

Free Energy of Formation for Solid Lithium Nitride

Breakthrough properties of Cs were examined at 298 K using columns packed with K⁺ forms of ferrierites. Breakthrough curves showed symmetrical S-shaped profiles at lower space velocity (SV) below 50/h. The adsorption zone of Cs increased linearly with flow rate and exceeded the length of packed column over 50/h. The S-shaped curves were obtained at lower HN0₃ concentrations below 1 mol/dm³, whereas Cs readily flowed out through the column in the presence of 5 mol/dm³ HNO₃. The break point was shifted to lower bed volumes with increasing concentration of coexisting cations; the amounts of Cs adsorbed were lowered in the presence of competing cations in the order, Na⁺>K⁺>Rb⁺. Cesium adsorbed on ferrierites could be quantitatively eluted with 1 mol/dm³ KNO₃ as an eluent. The selective removal of Cs was achieved by passing the simulated radioactive liquid wastes through the ferrierite columns.     

Ion Exchange Separation for Decontamination of Centrifuge Enrichment Plant

Ion exchange separation of uranyl ion (UO²⁺ ₂) from metal cations has been carried out by the columnar operation using ion exchange resins in 0.1 mol/dm³ sulfuric acid medium. Uranyl ion was adsorbed in an anion exchange resin and the rest metal cations, Fe(III), Al(III), Cr(III), Ni(II) and Cu(II) ions, were adsorbed in a cation exchange resin in this system. Desorption of uranyl ion and metal cations adsorbed in the resins were tested by 2 mol/dm³ sulfuric acid solution. Desorbed elements were confirmed to be precipitated by appropriate alkaline solutions. On the basis of the results obtained, a concept was made on a decontamination system for uranium-contaminated waste solution from centrifuge enrichment plant.     

Reduction of Cesium Leaching Ratio from Cementitious Resin Forms Using Natural Active Silica

Cesium adsorption behavior of active silica, which is a natural acid clay composed of cristobalite and quartz, was evaluated for its applicability as Cs adsorbent to be added to cementitious waste forms containing spent ion exchange resin. Since active silica carried the Cs exchangeable silanol group (—SiOH) originally, the Cs distribution coefficient was remarkably high (<10⁴). It increased in saturated Ca(OH)₂ solution, simulating the cement paste, due to formation of new silanol groups. With its addition to the cementitious forms with ¹³⁴Cs adsorbed ion exchange resin solidified by slag cement, the Cs leaching ratio was reduced to below 1/10 that without active silica.     

Solvent Extraction Behavior of Macroamounts of Elements in 100 % TBP-Nitric or Hydrochloric Acid Systems

The acid dependence curves of 15 inorganic ions in macroamounts were surveyed for the system of 100% TBP-HC1 and HNO₃, and compared with results obtained in tracer work previously reported. In most cases, theKd values measured in a system containing 0.1m/l salt are little different from those determined by radioactive tracers. The acid dependence behavior of ferric and zinc chloride were especially studied with varying metal ion concentrations. Additional experiments were carried out for ferric and zinc chloride in the system of 1 % (w/v) TBPO and 5 % (w/v) TOPO toluene-HCl. In these cases theKd values lowered with increasing chloride concentration.     

Effect of Temperature on the Entrainment of Ruthenium,Technetium and Selenium in Continuous Calcination of Simulated High-Level Liquid Waste

The effect of ternperature on the entrainment of Ru, Tc and Se was studied in calcinations experiments performed with continuous feeding of simulated high-level liquid waste containing radioactive tracers. The calcining temperature ranged from 300 to 800°C. The entrainment ratios showed a tendency to peak around 300–400°C followed by a dip and a second rise beyond 500°C. The volatilized fraction of Tc and Se increased with rising calcining temperature, whereas Ru volatilization decreased with rising temperature in the range covered by the experiment, i.e. above 300°C.     

Recovery of palladium by silica/polymer-based 2,6-bis(5,6,7,8-tetrahydro-5,8,9,9-tetramethyl-5,8-methano-1,2,4-benzotriazin-3-yl)pyridine adsorbents from high level liquid waste

The adsorption properties of a porous silica-based adsorbent 2,6-bis(5,6,7,8-tetrahydro-5,8,9,9-tetramethyl-5,8-methano-1,2,4-benzotriazin-3-yl)pyridine (Me₂-CA-BTP/SiO₂-P) towards Pd(II) were investigated. The adsorption ability of Me₂-CA-BTP/SiO₂-P towards Pd(II) increased dramatically with the increase of HNO₃ concentration and showed a saturated uptake of Pd(II) after the HNO₃ concentration reached 3 mol/dm³. It was found that nitrate ion would participate in the adsorption reaction. Palladium adsorption kinetic and isotherm were in accordance with the pseudo-second-order rate law and Langmuir isotherm adsorption model, respectively. Thiourea could effectively elute the adsorbed palladium from Me₂-CA-BTP/SiO₂-P. Adsorption studies with high level liquid waste elements showed a high selectivity towards Pd(II) over other fission product elements in ≥3 mol/dm³ HNO₃ solution. The stability of Me₂-CA-BTP/SiO₂-P against HNO₃ was satisfied for the reason of the adsorption performance keeping for Pd(II) after a long-time contact with HNO₃ solution. The adsorbent performed good γ-irradiation stability in dry state and low concentrations of HNO₃. The presence of HNO₃ enhanced the radiolysis and the adsorbent still reserved a presentable adsorption amount after γ-irradiation in 3 mol/dm³ HNO₃.     

Magnesium potassium phosphate ceramic for Tc immobilization

Technetium-99, present in the US Department of Energy’s (DOE) high-level waste (HLW) as a by-product of fission reactions, poses a serious environmental threat because it has a long half-life, is highly mobile in its soluble Tc⁷⁺ oxidation state and is volatile at high temperatures. Magnesium potassium phosphate (MKP) ceramics have been developed to treat ⁹⁹Tc that has been partitioned and eluted from simulated high-level tank wastes by means of sorption processes. Waste forms were fabricated by adding MKP binder and a reducing agent (SnCl₂) to the ⁹⁹Tc-containing aqueous waste. In addition, waste forms were fabricated by first precipitating ⁹⁹Tc from the waste and subsequently solidifying it in MKP. ⁹⁹Tc loadings in the waste forms were as high as 900 ppm by weight. Waste form performance was established through various strength, leaching, and durability tests. Long-term leaching studies, as per the ANS 16.1 procedure, showed leachability indices between 11 and 14 for ⁹⁹Tc under ambient conditions. The normalized leach rate for ⁹⁹Tc, according to the product consistency test, was as low as 1.1 × 10⁻³ g/m² d. The waste forms exhibited a compressive strength of ≈30 MPa and were durable in an aqueous environment. Containment of ⁹⁹Tc in MKP ceramics is believed to be due to a combination of appropriate reducing environment (determined from Eh-pH measurements) and microencapsulation in a dense matrix.     

Wide Range Wave-Number Spectra of Place-to-Place Variations in Terrestrial Gamma-Ray Exposure Rate

The precipitation behavior of Pu. Np and Am during the denitration of high-level radioactive liquid waste (HLW) by formic acid was studied using a simulated HLW. The dissolution of the precipitate formed in denitrated HLW was also studied using oxalic acid to recover transuranium (TRU) elements from the precipitate. In the denitration, the precipitated fractions of TRU elements increased with decreasing acidity of the denitrated HLW. In the denitration at [HCOOH]/[HNO₃]=1.5, which was adopted in the partitioning process developed in JAERI, the precipitated fractions of Np and Am were only 0.6% and 0.06%, respectively, whereas that of Pu was 90 %. The precipitation fractions of Pu and Np did not depend on their concentrations in the range of 6x10^?5–6x10^?4 M for Pu and 10^?5–10^?3 M for Np. Plutonium was not precipitated itself by polymerization or hydrolysis but coprecipitated with other elements such as Mo and Zr. It was found that the precipitate formed during the denitration of 1 l of HLW could be dissolved in a 800 ml of 0.5 M oxalic acid solution.     

Removal of Radioactive Ions from Nuclear Waste Solutions by Electrodialysis

Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained ¹³⁷Cs⁺, ⁹⁰Sr²⁺, ¹⁰⁶Ru³⁺ or fission products (F.P.) as active ions and NaCl, Na₂SO₄ or Ca(NO₃)₂ as inactive coexisting salts. The decontamination factor of the active ions was in the order: ¹³⁷Cs⁺ (greater than 99%) ⁹⁰Sr²⁺F.P. ¹⁰⁶Ru³⁺. The dialysis time required to attain the saturation was the shortest for monovalent cations K⁺, Cs⁺ and Na⁺, intermediate for divalent cation Sr²⁺, and the longest for trivalent cation Ru³⁺. The ratio of the decontamination factor of an active ion ηA to the desalination factor of an inactive ion A₊ was nearly equal to unity for ²⁴Na, ⁴²K, ¹³⁷Cs and ^S),Sr. On the other hand, the apparent selective permeability of an active ion (A⁺) against Na⁺ ion, T_A+ was higher than unity for all the active ions tested, and was in the order of ¹³⁷>Cs⁹⁰Sr>⁴²>K²⁴Na, where T^A _Na+ is defined by the ratio of γA to γ_Na+ with γA being the ratio of dilution of A in the diluate and γ_Na+ being that of Na⁺ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentrations of the coexistent salts or on the concentration of the active ions.     

Influence of Monoisodecyl Phosphoric Acid on Extraction of Neptunium with Diisodecyl Phosphoric Acid

The influence of monoisodecy1 phosphoric acid (MIDPA) on the separation of Np with diisodecy1 phosphoric acid (DIDPA) was studied from the aspects of the extraction rate and the back-extraction with H₂C₂O₄ solution. This study is important for the development of the partitioning method for high-level waste because MIDPA is one of the radiolysis products of DIDPA The increase in the MIDPA concentration was found to accelerate the extraction of Np initially in the pentavalent oxidation state. When the MIDPA concentration was 0.1 M, the extraction rate was ten times faster in the absence of H₂O₂ and twice faster in the presence of 0.5 M H₂O₂. It was also found that the extraction mechanism in the aspect of Np redox reactions did not change even in the presence of MIDPA

On the other hand, Np became less back-extracted with increasing MIDPA concentration. However, the tolerable concentration of MIDPA, determined by the required distribution ratio for the Np back-extraction, is higher than the estimated concentration of MIDPA produced radioiytically in the partitioning process.