Assessment of <Superscript>137</Superscript>Cs decontamination of concrete by the reagent method

The particulars of radioactive contamination of concrete wastes by ¹³⁷Cs were studied. x-Ray phase analysis and chemical analysis show that clayey materials, including Al₂O₃, Fe₂O₃, K₂O, and MgO, on which ¹³⁷Cs sorption is possible, were present in the concrete wastes. The content and form in which ¹³⁷Cs was found in radioactive concrete wastes from nuclear power facilities as well as in model samples were determined. When the wastes were treated with nitric acid the binder dissolved and a polydisperse suspension formed. The suspension consisted of three phases: solution, fine suspension, and rapidly settling precipitate of heavy filler particles. x-Ray phase analysis was performed and the ¹³⁷Cs mass ratio and distribution in the phases were determined. The possibility of decontaminating the concrete by a reagent method was evaluated.     

Removal of Liquid and Solid Impurities from Uranyl Nitrate Hexahydrate Crystalline Particles in Crystal Purification Process

The purification behavior of uranyl nitrate hexahydrate (UNH) was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide (MOX) fuel in batch experiments. The UNH crystal recovered from the MOX fuel dissolver solution containing simulated fission products (FPs) was purified by a sweating and melt filtration process. Although the decontamination factors (DFs) of Pu, Cs, and Ba did not change in the sweating process, that of Eu increased with increases in temperature and time. These results indicate that liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs, and Ba were minimally affected in the batch experiments. On the other hand, the DF of Ba increased with 0.45 and 5.0 μ filters in the melt filtration process. Since Pu and Cs formed as Cs₂Pu(NO₃)₆ in the course of U crystallization and was accompanied with the UNH crystal, these behaviors were similar to each other. Although the DFs of Pu and Cs did not change with the 5.0 μ filter, it increased approximately twofold with the 0.45 μ filter. The particle size of Cs₂Pu(NO₃)₆ is relatively small and might pass through the 5.0 μ filter in the melt filtration process. The liquid impurities as Eu remained in the molten UNH crystal with some filters.     

Use of γ Radiation from Short-Lived Fission Products for Detecting <Superscript>235</Superscript>U and <Superscript>239</Superscript>Pu

It is suggested that γ radiation with E _γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect ²³⁵U and ²³⁹Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 365–370, May 2005.     

Adsorption behavior of radioactive cesium by non-mica minerals

We studied the adsorption behavior of radioactive cesium (Cs) by the non-mica minerals kaolinite, halloysite, chlorite, montmorillonite, mordenite, MnO₂, TiO₂, Al₂O₃, and FeOOH to elucidate the environmental behavior of radioactive Cs fallout from the Fukushima Daiichi Nuclear Power Plant in the Tohoku region of Japan. The adsorption and desorption experiments of Cs on the minerals were carried out at the Cs concentrations 1 × 10^?4, 1 × 10^?5, and 2 × 10^?9 mole L^?1 at pH 5.5. The desorption of Cs from the minerals was examined using 0.1 mole L^?1 LiCl, NaCl, KCl, RbCl, and CsCl solutions. The sequential desorption was examined using a 0.1 mole L^?1 LiCl solution, a 1 mole L^?1 KCl solution, and a 1 mole L^?1 HCl solution. The distribution coefficient (K _d) for the minerals at the Cs concentration 10^?9 mole L^?1 was in the order of mordenite > illite > montmorillonite, sericite, MnO₂, kaolinite, and halloysite > chlorite, TiO₂, Al₂O₃, and FeOOH, differing from the order observed at higher Cs concentrations. After the sequential desorption by the three reagent solutions, the residual fraction of Cs was higher at the Cs concentration 10^?9 mole L^?1 than at higher concentrations. Approximately 40%, 40%, 50%, and 25% of the adsorbed Cs were residual in montmorillonite, mordenite, MnO₂, and kaolinite, respectively, after the sequential desorption. These results strongly suggest that (1) radioactive Cs at 10^?9 mole L^?1 is more strongly associated with the non-mica minerals than at higher concentrations of 1 × 10^?4 and 1 × 10^?5 mole L^?1, and (2) the non-mica minerals montmorillonite, mordenite, kaolinite, and MnO₂ contributed to the fixation of the radioactive Cs fallout on Fukushima soil.     

Stepped decontamination of <Superscript>137</Superscript>Cs-contaminated sandy soil

The possibilities of reagent-based processing of sandy soil contaminated with ¹³⁷Cs are examined. To attain a high decontamination factor, the process is conducted in several steps. First, water-gravity separation is used to separate finely dispersed matter (first step). The main goal of the next (second) reagent-based processing step is to destroy the strongly bound fragments and hydroxide films with which a part of the finely dispersed fraction is bound. At the next step, part of the ¹³⁷Cs is extracted into the reagent solution from the purely sandy fraction; repeated water-gravity separation (third step) makes it possible to completely separate the finely dispersed substance. The subsequent steps of reagent-based processing give ¹³⁷Cs concentration in the sandy fraction of the initial soils that meets the sanitary norms.     

Separation of Fission Products from Irradiated Uranium Dioxide in Fused Nitrate Systems

The separation of fission products from irradiated UO₂ in fused nitrate systems was studied by the following procedure:

Dissolution of UO₂ in fused NH₄NO₃; adsorption of fission products on glass powder; fused salt chromatography with γ-Al₂O₃; evaporation from fused nitrates; and precipitation of U-compound (probably alkali uranate, M₂U₂O₇) in a fused LiNO₃-KNO₃ mixture. Radiochemically pure Zr-Nb was selectively separated from the fused NH₄NO₃ melt by adsorption. Ru and I were completely evaporated from fused LiNO₃-KNO₃ mixture at above 280°C and from fused NH₄NO₃at above 240°C, respectively. By means of chromatography with a γ-Al₂O₃ column, Zr-Nb and rare earths were completely separated from U in fused NH₄NO₃ or fused NH₄NO₃-LiNO₃ medium at 150° to 180°C, and further, Cs, Ba and Ru were also expected to be separable from U under suitable condition. When the U-compound was precipitated in fused LiNO₃-KNO₃ at 350°C, Cs, Sr, Ba and Ru were fairly well separated from U.     

<Superscript>137</Superscript>Cs content and vertical migration in littoral soil of two lakes in Bryansk Oblast

The dynamics and present state of the radioactive contamination with ¹³⁷Cs of littoral soil of Lake Kozhanovskoe and Lake Svyatoe on the Besed’ River are presented. The parameters of the vertical migration of ¹³⁷Cs in soil-the rate of directional transfer with soil moisture, the diffusion coefficient, and the average velocity of vertical migration-are estimated by comparing the experimental and model distributions of ¹³⁷Cs content over soil depth. It is shown that at the present time the two 5-cm layer of soil can contain 20–90% ¹³⁷Cs depending on the type of soil and landscape. The average values of the diffusion coefficient, the rate of directional transfer, and the vertical migration velocity for 12-, 13-, and 20-year periods after the Chernobyl accident are 0.1–2.8 cm²/yr, 0.1–0.3 cm/yr, and 0.1–0.8 cm/yr, respectively. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 306–311, May, 2007.     

Immobilization of Cs and Sr to HZr2(PO4)3 using an autoclave

The proton-type crystalline zirconium phosphate, HZr₂(PO₄)₃, was prepared by a thermal decomposition of NH₄Zr₂(PO₄)₃ at about 450 °C, where NH₄Zr₂(PO₄)₃ was obtained in advance by a hydrothermal synthesis using a mixed solution of ZrOCl₂, H₃PO₄ and H₂C₂O₄. Cs or Sr ion was immobilized to HZr₂(PO₄)₃ by mixing HZr₂(PO₄)₃ with an aqueous solution of CsNO₃ or Sr(NO₃)₂ under the molar ratio CsNO₃/HZr₂(PO₄)₃ = 1.0 or Sr(NO₃)₂/HZr₂(PO₄)₃ = 0.5. The mixtures were treated thermally in an autoclave at different temperatures from 200 to 275 °C and Arrhenius equation was applied to the Cs and Sr immobilization process to HZr₂(PO₄)₃. The activation energy for the immobilization process of Cs or Sr was estimated as 179 kJ mol^?1 and 186 kJ mol^?1, respectively.     

Syntheses and crystal structures of Eu(III) and Sm(III) perrhenate complexes with 2,2′-(imino)bis(N,N′-diethylacetamide)

Europium(III) and Sm(III) perrhenate complexes with 2,2′-(imino)bis(N,N′-diethyl-acetamide) (IDEA) were synthesized. From the single crystal X-ray analyses, the synthesized complexes were found to be [Eu(IDEA)₃][Eu(NO₃)₄(ReO₄)₂]_0.38[Eu(NO₃)₅(ReO₄)]_0.62 (I) and [Sm(IDEA)₃][Sm(NO₃)₄ (ReO₄)₂]_0.23[Sm(NO₃)₅(ReO₄)]_0.77 (II). In the cationic parts [M(IDEA)₃]³⁺ (M = Eu and Sm), M(III) is nine-coordinated with three tridentate IDEA ligands. The anionic parts are located crystallographically at the same position in the unit cell. The occupancies of [M(NO₃)₄(ReO₄)₂]^3? and [M(NO₃)₅(ReO₄)]^3? are 38% and 62% for Eu(III), 23% and 77% for Sm(III), respectively. The main crystallographic parameters are as follows: space group P2₁/c, unit cell parameters a = 12.3811 (4) Å, b = 32.0748 (8) Å, c = 16.7287 (4) Å, β = 112.358 (1)°, Z = 4, V = 6143.9 (3) ų for (I), and space group P2₁/c, unit cell parameters a = 12.3597 (2) Å, b = 32.1599 (6) Å, c = 16.6762 (4) Å, β = 112.431 (1)°, Z = 4, V = 6127.0 (2) ų for (II). Infrared (IR) absorption spectra of the complexes I and II support the coordination of IDEA and ReO₄^? to M(III) through each carbonyl oxygen and one oxygen.     

<Superscript>59</Superscript>Ni Accumulation in Corrosion-Resistant Steel under the Action of Neutron Flux

A γ-ray line with energy E_γ = 11.3 MeV was detected during an experiment, performed on a nuclear reactor, investigating the characteristics of the energy spectrum of γ-rays. The most likely source of this line is radiative capture of thermal neutrons by ⁵⁹Ni nuclei, which accumulated in the corrosion-resistance steel as a result of the more than 20 years of irradiation in the reactor, via the reaction ⁵⁸Ni(n, γ)⁵⁹Ni. It was found that for thermal-neutron fluence 10²¹ cm⁻² the ⁵⁹Ni concentration is 0.47% of the ⁵⁸Ni concentration. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 268–272, October, 2005.     

Direct measurements of the<Superscript>90</Superscript>Sr activity in water using Cherenkov radiation

A computational-experimental investigation of Cherenkov radiation due to⁹⁰Sr−⁹⁰Y in water samples was performed. The Monte Carlo method was used to simulate the generation of photons from the decay of⁹⁰Sr−⁹⁰Y and other isotopes in water in the range 220–600 nm. The Cherenkov radiation was measured using a low-noise photomultiplier with a tellurium-rubidium photocathode on a MgF₂ entrance window. Experiments on measuring the amplitude distributions and counting rates due to Cherenkov radiation from the radioactive solutions of⁹⁰Sr−⁹⁰Y,¹³⁷Cs−¹³⁷Ba, and⁴⁰K were performed. The sensitivity and lowest measurable activity for water samples of⁹⁰Sr−⁹⁰Y was estimated on the basis of the results obtained, 4 figures, 3 tables, 11 references. Russian Science Center “Kurchatov Institute.” Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 282–286, April, 2000.     

Stability constants of nitrate complexes of trivalent plutonium in aqueous and tributyl phosphate solutions

The extractability of nitrate complexes of trivalent plutonium from an aqueous phase at various concentrations was investigated. It was shown that tributyl phosphate (TBP) will extract trivalent plutonium from nitric acid solutions in the form of Pu(NO₃)₃. TBP at nitrate ion concentrations up to 1.2 M. Under the conditions studied, the distribution coefficient of trivalent plutonium depended very little on the concentration of hydrogen ions. The stability constants of the complexes Pu(NO₃)₃ · 3TBP, Pu(NO₃)₃, Pu(NO₃) ₂ ⁺ and Pu(NO₃)²⁺ were determined and equaled 0.75 ± 0.1; 14.4 ± 0.8; 14.3 ± 0.8 and 5.9 ± 0.5.     

Applicability of a catalytic reduction method using a palladium–copper catalyst and hydrazine for the denitration of a highly concentrated nitrate salt solution

Denitration of a highly concentrated sodium nitrate (NaNO₃) aqueous solution via a catalytic reduction method using a palladium–copper catalyst supported on carbon powder (Pd–Cu/C) and hydrazine (N₂H₄) was investigated. It was demonstrated that nitrate ion (NO₃ ^?) in a 5 mol L^?1 NaNO₃ solution was completely reduced through an intermediate nitrite ion (NO₂ ^?) to nitrogen compounds such as nitrogen, nitrous oxide, and ammonia. By comparing the reaction rates of NO₃ ^? and NO₂ ^? obtained using catalysts with various Pd–Cu compositions and different reductants (hydrogen (H₂) or N₂H₄), it was determined that the catalyst with a molar ratio of Pd:Cu = 1:0.66 provides the maximum reaction rates for NO₃ ^? and NO₂ ^? using N₂H₄, and that not only the reactions of NO₃ ^? and NO₂ ^? but also that of N₂H₄ were affected by the Pd–Cu composition.     

Analysis of the characteristics of <Superscript>238</Superscript>U, <Superscript>232</Superscript>Th, <Superscript>237</Superscript>Np,and <Superscript>231</Superscript>Pa radiators for fission ionization chambers for measuring the neutron flux density in the ITER diverter

Computational results, obtained by analyzing possible schemes of nuclear transformations of each of four threshold fission radiators ²³⁸U, ²³²Th, ²³⁷Np, and ²³¹Pa, for fission ionization chambers are presented. The influence of the nuclear reactions (n, ƒ), (n, γ), and (n, 2n) on the characteristics of fission ionization chambers is taken into account in the nuclear transformation schemes for all four radiators. The results are presented in the form of a dependence of the sensitivity of the fission ionization chambers on the neutron fluence in the range 10²¹–10²⁴ cm⁻². The effect of 0.2 and 1 g/cm² thick boron screens is examined. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 42–47, January, 2009.     

Study of the vaporization of Ba from UO2 nuclear fuel particles by high-temperature mass spectrometry

In connection with improving the retention of solid fission products in gas-cooled high-temperature reactor fuels, the vaporization of Ba from UO₂ model nuclear fuel particles with and without a pyrocarbon coating was studied by high-temperature mass spectrometry using a Knudsen cell. The UO₂ kernels of the particles were doped with BaO. In addition, some of them contained Al₂O₃. Whereas BaO mainly evaporated from the surface of the kernels as BaO, only Ba could be observed over the coated particles. Moreover, the BaO vapor pressure over kernels with and without the addition of Al₂O₃ was determined. From this it was determined that the BaO vapor pressure could be diminished by approximately two orders of magnitude by the admixture of Al₂O₃. Finally it was proved that the diminution of the BaO vapor pressure was caused by the formation of the compound BaAl₂O₄.     

Phase study on solid fission products,Ba, Sr and Zr in oxide fuel

Behavior of pseudo-fission products (Ba, Sr, and Zr) as oxides in UO₂ has been investigated by means of high-temperature X-ray diffraction and microprobe analysis. Two identifiable compounds were formed as reaction products in the mixed oxides which initially consisted of UO₂, BaO, SrO, and ZrO₂ powder. These compounds were present and were identified crystallographically as (Ba, Sr) ZrO₃ and (Ba, Sr)UO₃ after heat treatment of the powders at 1500°C for 30 min. Both compounds are isostructural with perovskite, CaTiO₃, and the lattice parameters of both (Ba, Sr)ZrO₃ and (Ba, Sr)UO₃ decrease with increasing content of Sr. (Ba, Sr)UO₃ is decomposed almost completely at 1800°C while (Ba, Sr) ZrO₃ is stable up to 2000°C. The behavior of Ba, Sr, and Zr in fuel under irradiation is discussed.     

Selective separation and recovery of cesium by ammonium tungstophosphate-alginate microcapsules

A fine crystalline ammonium tungstophosphate (AWP) exchanger with high selectivity towards Cs⁺ was encapsulated in biopolymer matrices (calcium alginate, CaALG). The characterization of the AWP-CaALG microcapsule was examined using SEM/WDS, IR and DTA/TG analyses, and the selective separation and recovery of ¹³⁷Cs were examined by the batch and column methods using simulated (SHLLW) and real high-level liquid waste (HLLW). The free energy (ΔG°) of the ion exchange (NH₄⁺ ↔ Cs⁺) for fine AWP crystals was determined at −13.2 kJ/mol, indicating the high selectivity of AWP towards Cs⁺. Spherical and elastic AWP-CaALG microcapsules (∼700 μm in diameter) were obtained and fine AWP crystals were uniformly immobilized in alginate matrices. Relatively large K_d values of Cs⁺ above 10⁵ cm³/g were obtained in the presence of 10⁻³–1 M Ca(NO₃)₂, resulting in a separation factor of Cs/Rb exceeding 10². The irradiated samples (⁶⁰Co, 17.6 kGy) also exhibited large K_d values exceeding 10⁵ cm³/g in the presence of 2.5 M HNO₃. The K_d values in the presence of 0.1–9 M HNO₃ for 67 elements were determined and the order of K_d value was Cs⁺ ? Rb⁺ > Ag⁺. The breakthrough curve of Cs⁺ had an S-shaped profile, and the breakpoint increased with decreasing flow rate; the breakpoint and breakthrough capacity at a flow rate of 0.35 cm³/min for the column (0.7 g AWP-CaALG) were estimated at 25.2 cm³ and 0.068 mmol/g, respectively. Good breakthrough and elution properties were retained even after thrice-repeated runs. The uptake (%) of Cs⁺ in SHLLW (28 metal components-1.92 M HNO₃, SW-11, JAEA) was estimated at 97%, and the distribution of Cs⁺ and Zr/Ru into the AWP and alginate phases, respectively, were observed by WDS analysis. Further, the selective uptake of ¹³⁷Cs exceeding 99% was confirmed by using real HLLW (FBR “JOYO”, JAEA). AWP-CaALG microcapsules are thus effective for the selective separation and recovery of Cs⁺ from HLLWs.     

Behavior of Iodine-131 in Dissolution of Irradiated Uranium Dioxide

A hundred milligram of UO₂ was irradiated In JRR-4 (burn up 0.017%) and dissolved In boiling 3 M-HNO₃ (20 ml) to study the behavior of fission-product iodine (¹³¹I, ? 1 ng) During the dissolution, 80–90% of the iodine volatilized together with ¹³³Xe. Zeolite 13X (–4g) trapped completely the iodine in the dissolver off-gas Blowing a NO₂ flow through the solution was effective for expulsion of the remaining iodine and finally 90–95% of the initial iodine was removed from the solution with the aid of addition of KIO₃. The Iodine species in the solution includes I₂, I^?, IO₃ ^? (IO₄ ^?), organic iodides and other unknown species The expulsion operations caused the unknown species to increase In relative abundance. This tendency was also noticed for iodine of higher concentration in 3 M-HNO₃.     

Correcting the isotopic composition of regenerated uranium with respect to <Superscript>232</Superscript>U by a centrifuge method with introduction of a carrier gas

Theoretical and experimental studies dealing with correcting the isotopic composition of regenerated uranium with respect to ²³²U by a centrifuge method with introduction of a carrier gas are reported. In order to increase the efficiency of separating ²³²U from the spent uranium and reduce the loss of ²³⁵U, the use of a carrier gas is proposed – the gaseous compound ¹²C₈H₃F₁₃, which is inert to uranium hexafluoride, and whose molecular weight, M_c = 346 amu, matches that of ²³²UF₆. Freon, C₈H₃F₁₃, is shown not to decompose during operation in the rotor of a centrifuge or to interact with the centrifuge material. The measured absorption parameters of freon on sodium fluoride NaF confirm the feasibility of efficient separation of a mixture of uranium hexafluoride and freon with return of the freon to the separation process. It is shown that introducing a carrier gas into the centrifuge technology can yield some new results: lowering the radioactivity of the commercial product, normalizing the overall radiation situation during production, increasing the recovered ²³⁵U in the commercial product, and reducing the volume of radioactive waste. The recovery of ²³⁵U in the commercial product can be increased to 99% or more. Then the ²³²U content in the commercial product is ∼2·10⁻⁸% or a factor of 10 less than the maximum allowable content of 2·10⁻⁷%. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 150–156, September, 2008.