Experimental and simulated assay of actinides in a real waste package

The non-destructive control of nuclear wastes is important for their management and the non-proliferation. Among the methods using the nuclear radiation as an investigation, the Instrumental Photon Activation Analysis (IPAA) seems to be a promising way to quantify the masses of actinides present in bulky packages of nuclear waste. The IPAA method consists in irradiating actinides with photons of high energy to produce photofission reactions. The counting of the delayed neutrons, produced by these photofission reactions, allows to locate and to quantify the mass of actinides by tomography. At this end, we use a simulation tool named OPERA to obtain the information necessary for the tomographic restitution and an experimental installation based on a LINear ACcelerator (LINAC) to perform the measurements. The high-energy photons (11 MeV) are produced by Bremsstrahlung, thanks to a tungsten target placed in front of the LINAC. In this paper, we present the first experimental results obtained on a real package of nuclear waste. We establish that, for this waste package, the limit of detection, in terms of mass of actinides, is about 1 g. Furthermore, these results show the good agreement between the experiment and the simulation that provides a localization of actinides by tomography.     

Kinetic and thermodynamic studies for cesium removal from low-level liquid radioactive waste using impregnated polymeric material

A novel polymeric impregnated material was prepared by loading 4,4′(5′)di-t-butylbenzo-18-crown-6 (DtBB18C6) onto poly(acrylamide-acrylic acid-acrylonitrile)-N,N′-methylenediacrylamide [P(AM-AA-AN)-DAM]. The sorption of ¹³⁷Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was studied using batch equilibrium technique with respect to the pH, contact time, and temperature. The applicability to treatment of low-level liquid radioactive waste was examined. The free energy (ΔG ⁰), enthalpy (ΔH ⁰), and entropy (ΔS ⁰) of the sorption were calculated. These parameters showed that the sorption of ¹³⁷Cs onto P(AM-AA-AN)-DAM/DtBB18C6 was spontaneous and endothermic in nature. The Lagergren first-order, pseudo-second-order, and homogeneous particle diffusion models were tested kinetically to describe the reaction mechanism. The experimental data were fitted well by the pseudo-second-order kinetic model.     

Construction of complementary cumulative distribution functions for comparison with the EPA release limits for radioactive waste disposal

The US Environmental Protection Agency (EPA) has promulgated a standard for the geologic disposal of radioactive waste (40 CFR 191) that requires the construction of a complementary cumulative distribution function (CCDF) that defines the probability of exceeding radionuclide releases of various sizes. Construction of this CCDF involves three interrelated activities: development of scenarios that describe the various disruptions that could occur at a waste disposal site, determination of probabilities for scenarios, and calculation of radionuclide releases for scenarios. The computational procedures being used in the performance assesement for the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico to bring these three activities together to produce the CCDF in 40 CFR 191 are described and illustrated with results obtained in a recent analysis.     

Effect of thermal annealing on the properties of borobasalt systems containing simulated radioactive waste from pyroelectrochemical technology

The effect of thermal annealing performed in the temperature range 400–800°С on the hydrolytic durability of borobasalt systems containing simulated radioactive waste from pyroelectrochemical technology was studied. Three types of simulated waste were examined: type 1, waste from reprocessing of reactor-grade plutonium; type 2, waste from reprocessing of weapons-grade plutonium; type 3, waste from gallium recovery from type 2 waste (virtually pure americium phosphate). Depending on the waste type and temperature, annealing can both increase and decrease the host material leach rate.     

Cs sorption to potential host rock of low-level radioactive waste repository in Taiwan: Experiments and numerical fitting study

A reliable performance assessment of radioactive waste repository depends on better knowledge of interactions between nuclides and geological substances. Numerical fitting of acquired experimental results by the surface complexation model enables us to interpret sorption behavior at molecular scale and thus to build a solid basis for simulation study. A lack of consensus on a standard set of assessment criteria (such as determination of sorption site concentration, reaction formula) during numerical fitting, on the other hand, makes lower case comparison between various studies difficult. In this study we explored the sorption of cesium to argillite by conducting experiments under different pH and solid/liquid ratio (s/l) with two specific initial Cs concentrations (100 mg/L, 7.5 × 10⁻⁴ mol/L and 0.01 mg/L, 7.5 × 10⁻⁸ mol/L). After this, numerical fitting was performed, focusing on assessment criteria and their consequences. It was found that both ion exchange and electrostatic interactions governed Cs sorption on argillite. At higher initial Cs concentration the Cs sorption showed an increasing dependence on pH as the solid/liquid ratio was lowered. In contrast at trace Cs levels, the Cs sorption was neither s/l dependent nor pH sensitive. It is therefore proposed that ion exchange mechanism dominates Cs sorption when the concentration of surface sorption site exceeds that of Cs, whereas surface complexation is attributed to Cs uptake under alkaline environments. Numerical fitting was conducted using two different strategies to determine concentration of surface sorption sites: the clay model (based on the cation exchange capacity plus surface titration results) and the iron oxide model (where the concentration of sorption sites is proportional to the surface area of argillite). It was found that the clay model led to better fitting than the iron oxide model, which is attributed to more amenable sorption sites (two specific sorption sites along with larger site density) when using clay model. Moreover, increasing s/l ratio would produce more sorption sites, which helps to suppress the impact of heterogeneous surface on Cs sorption behavior under high pH environments.     

Sorption of radionuclides by microorganisms from a deep repository of liquid low-level waste

Sorption of radionuclides on a biomass of aerobic microorganisms isolated from deep repositories of liquid low-level waste was examined. Each strain exhibited substantially different sorption of different radionuclides. In neutral medium, the maximal proportion of the recovered radionuclides was 77, 92, 76, 72, and 33% for Pu, Np, U, Am, and Sr, respectively. None of the isolated strains sorbed Cs. The highest sorption ability was exhibited by Pseudomonas genus bacteria. The results obtained are indicative of different mechanisms of interactions of microorganisms with radionuclides.     

纳滤法去除模拟放射性废水中Sr~(2+)的研究

采用XN45聚酰胺复合纳滤膜处理含锶(Sr2+)的模拟放射性废水.考察了初始离子浓度、溶液pH、离子强度和投加络合剂等情况下对Sr2+截留率的影响.结果表明,Sr2+截留率随着初始离子浓度和离子强度的增加而有所降低;pH升高时,截留率在一定pH范围内先降低后升高,在等电点(pH 5)附近达到最小值;在分别以聚丙烯酸和乙二胺四乙酸二钠(EDTA-2Na)作为络合剂的情况下,Sr2+的截留率均得到了提高,其中在[PAA]/[Sr](质量浓度比,下同)=5,pH7或[EDTA-2Na]/[Sr]=4,pH=9的情况下,Sr2+截留率均达到98%以上.     

Retrospective dosimetry: dose evaluation using unheated and heated quartz from a radioactive waste storage building

In the assessment of dose received from a nuclear accident, considerable attention has been paid to retrospective dosimetry using heated materials such as household ceramics and bricks. However, unheated materials such as mortar and concrete are more commonly found in industrial sites and particularly in nuclear installations. These materials contain natural dosemeters such as quartz, which usually is less sensitive than its heated counterpart. The potential of quartz extracted from mortar in a wall of a low-level radioactive-waste storage facility containing distributed sources of 60Co and 137Cs has been investigated. Dose-depth proliles based on small aliquots and single grains from the quartz extracted from the mortar samples are reported here. These are compared with results from heated quartz and polymineral fine grains extracted from an adjacent brick, and the integrated dose recorded by environmental TLDs.     

Durability of class C fly ash belite cement in simulated sodium chloride radioactive liquid waste: influence of temperature

This work is a continuation of a previous durability study of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) that is very rich in sulphate salts. The same experimental methodology was applied in the present case, but with a SRLW rich in sodium chloride. The study was carried out by testing the flexural strength of mortars immersed in simulated radioactive liquid waste that was rich in chloride (0.5M), and demineralised water as a reference, at 20 and 40 degrees C over a period of 180 days. The reaction mechanism of chloride ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated chloride radioactive liquid waste (SCRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive Friedel's salt inside the pores; accordingly, the microstructure was refined.     

Separation of critical radioactive and non-radioactive species from aqueous waste streams

The separation of radioactive and non-radioactive species from the simulated DOE neutralized current acid waste (NCAW) stream was studied. Cation and anion species were referred to their possible basic compounds, and divided into seven groups (nitrate, phosphate, sulfate, fluoride, nitrite, carbonate, and hydroxide). The nitrate group (the major anion in the DOE waste streams) contains several cations species, while the rest of the groups are only in the form of sodium. The precipitation measurements were conducted in three experimental stages. In the first stage, the precipitation of sodium sulfate, sodium phosphate, sodium-sulfate-phosphate, and aluminum nitrate systems were studied using isopropylamine (IPA) as a precipitation solvent. The objectives of this stage were to evaluate the precipitation ability of IPA in precipitating these compounds individually, and to validate the consistency of the analytical instruments and the employed experimental procedure. Tests performed on the acquired data indicated a high level of experimental consistency. The removal of phosphate, sulfate and aluminum were very high. In the second stage, the precipitation studies were conducted on the: (1) nitrate group alone; (2) binary groups containing the groups of nitrate-phosphate, nitrate-sulfate, nitrate-fluoride, nitrate-nitrite, and nitrate-carbonate; (3) combined nitrate, phosphate, sulfate, and fluoride groups and (4) combined nitrate, phosphate, sulfate, fluoride, nitrite, and carbonate groups. IPA was used as a precipitation solvent. The objectives of this stage were to evaluate the interactions of these groups in the absence of the hydroxide group (e.g. DOE acid-dissolved sludge and acidified supernate streams), and the influence of such interactions on the individual removal of the targeted species. The removal of the aluminum, phosphate, fluoride, and alkaline cations were significantly high (reached 99.9%). The removal of sulfate were moderately high (reached 87%), and the removal of nitrate and alkali cations including cesium were to some extent low (reached about 50%).In the third stage, the precipitation of inorganic species from the simulated NCAW stream was studied using IPA and ethylamine (EA). The precipitation process is very feasible for reducing the radioactivity contents of alkaline cations. However, the process is less effective in separating alkali cations including cesium. The removal of polyvalent transition metals such as aluminum ion is negatively influenced by the significant presence of hydroxide. While the process is effectively capable of separating phosphate, fluoride, and sulfate, it is significantly less effective in separating nitrate and nitrite. A previously derived thermodynamics framework was used to model the precipitation measurements. The framework provided two predictive equations (the 2-Suffix and 3-Suffix equations). Both equations were reasonably adequate for predicting the solubility phase behavior of tested inorganic species in a mixed-solvents mixture as well as for estimating optimum interaction parameters. However, the 3-Suffix equation was better than the 2-Suffix equation. The parameters were useful for estimating the: (1) precipitation fractions (%P) of the studied species, for instance, at different concentration levels of similar targeted species, or in different waste streams with similar or approximate abundance of species, or at different solvents volume ratio (V(r)) where no experimental data are available.     

Resistance of class C fly ash belite cement to simulated sodium sulphate radioactive liquid waste attack

The resistance of class C fly ash belite cement (FABC-2-W) to concentrated sodium sulphate salts associated with low level wastes (LLW) and medium level wastes (MLW) is discussed. This study was carried out according to the Koch and Steinegger methodology by testing the flexural strength of mortars immersed in simulated radioactive liquid waste rich in sulphate (48,000 ppm) and demineralised water (used as a reference), at 20 degrees C and 40 degrees C over a period of 180 days. The reaction mechanisms of sulphate ion with the mortar was carried out through a microstructure study, which included the use of Scanning electron microscopy (SEM), porosity and pore-size distribution and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated sulphate radioactive liquid waste (SSRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive ettringite inside the pores and an alkaline activation of the hydraulic activity of cement promoted by the ingress of sulphate. Accordingly, the microstructure was strongly refined.     

Irradiated and metamict materials: relevance to radioactive waste science

The relation of the structures of metamict mineral phases and non-crystalline solids is discussed; the question of radiation effects in solidified nuclear waste forms is considered in this regard. In ceramics, if the actinides can be concentrated in radiation-resistant phases such as uraninite or monazite, the overall radiation effects on the ceramics should be quite small.     

Decontamination of liquid radioactive waste concentrates from 60Co using diethyldithiocarbamates

Oxidized bottom residue (BR) from installations for reprocessing trap waters of the Leningrad NPP were treated to remove ⁶⁰Co by coprecipitation with transition metal diethyldithiocarbamates (DDTC). Coprecipitation of ⁶⁰Co with manganese and cobalt diethyldithiocarbamates was examined in relation to pH of solution, temperature, amount of diethyldithiocarbamates, reagent ratio, and number of coprecipitation steps. The effect of diethyldithiocarbamates on the efficiency and conditions of the solution decontamination from ^134,137Cs on nickel potassium ferrocyanides was studied. Oxidative treatment of liquid radioactive waste concentrates to break down organic complexes is a necessary condition for ⁶⁰Co removal by coprecipitation with transition metal diethyldithiocarbamates. From a solution initially containing 56 kBq dm^{−3 60}Co, ⁶⁰Co can be removed to a residual content of 0.12–0.16 kBq dm⁻³ by two-step coprecipitation with Co(DDTC)₂. The coprecipitation with Co(DDTC)₂ does not affect the efficiency of the subsequent removal of Cs radionuclides on nickel potassium ferrocyanides.     

Effect of thermal annealing on the properties of borophosphate systems containing simulated Cs + Sr radioactive waste fractions

The effect of thermal annealing on the resistance of borophosphate systems to crystallization and the effect of crystallization on the hydrolytic stability of the materials were examined. Systems containing simulated Cs + Sr fraction were studied. Annealing at 400–500°C leads to a noticeable decrease in the leach rate. Annealing at a higher temperature, 600°C, leads to crystallization of the glass and to a considerable increase in the leach rate.     

Zero-order release of theophylline from a core-in-cup tablet in sequenced simulated gastric and intestinal fluid

Core-in-cup tablets containing theophylline were evaluated for their dissolution characteristics in sequenced simulated gastric fluid (SGF) followed by simulated intestinalfluid (SIF). Core-in-cup tablets containing 10% w/w, 20% w/w, and 30% w/w acacia as binder were evaluated for their effects on the time course of release of theophylline. This was done to optimize a formula that could release theophylline at a zero-order rate of release for 8-16 hr in simulated gastrointestinal fluids. Theophylline was released and dissolved from the core-in-cup tablets at a rate that is more consistent with a zero-order dissolution rate than a first-order dissolution rate in both SIG and SIF. The dissolution rates of theophylline from the 10%, 20%, and 30% acacia core-in-cup tablets were 0.87 mg/min, 0.53 mg/min, and 0.27 mg/min, respectively in SGF, and 0.61 mg/min, 0.30 mg/min, and 0.20 mg/min, respectively in SIF. The results indicate that a concentration of 32% w/w acacia in the core tablet will release theophylline at a rate of 0.14 mg/min in SGF for 2 hr followed by SIF for 10 hr.     

Low-temperature immobilization of liquid radioactive waste using silica from concentrated hydrothermal solutions

Experiments on vitrification of simulated liquid radioactive waste (LRW) were performed. To obtain solid glass-like phases, LRW components were treated with aqueous-alcoholic hydrochloric acid solutions containing hydrolyzates of alkyl silicates and aqueous silica sols obtained by membrane concentration of a natural hydrothermal solution. The pH of the mixture was 1.5–4. The characteristics of the solid samples obtained were studied by X-ray phase analysis, thermogravimetry, and scanning electron microscopy. A procedure was developed for low-temperature (5–60°C) immobilizaiton of low-level LRW using aqueous silica sols.     

Modeling of paraffin deposition in flow of a gas-oil mixture in tubes

The paper presents a scheme of the formation and growth of paraffin deposits on internal walls of the hoisting column of a well with allowance for diffusive transport of the particles of a solid phase to the wall of the hoisting column, for temperature and concentration conditions for these particles (that ensure the formation of a solid phase), and for thermal effects of the paraffin formation.     

Advances in cement solidification technology for waste radioactive ion exchange resins: a review

Treatment and disposal of waste radioactive ion exchange resins is one of the most urgent problems for nuclear industries in China. Cement solidification technology has many advantages, such as requiring simple equipment, easy scaling-up, low working temperature, no trouble of gas cleaning and low cost. It is a suitable technology for treatment of waste radioactive resins, and has been widely used. The new developments and theoretical basis of cement solidification of radioactive resins were introduced in this paper. The cement solidification technology suitable for China and the questions needed to solve were also discussed.     

Concept of autonomous processing of liquid radioactive waste from ship nuclear power installations

The concept of autonomous processing of liquid radioactive waste from ship nuclear power installations is substantiated, and its implementation in the form of a module membrane sorption installation consisting of micro- and ultrafiltration, reverse-osmosis, and ion-exchange modules is suggested. Data on treatment of liquid radioactive waste of complex physicochemical composition using this installation are reported.     

Chromatographic separation of 65Zn from radioactive tellurium waste solution using alumina column

High-purity ⁶⁵Zn was separated from a mixture containing ^{121,121m,123m,127}Te, ⁶⁵Zn, ⁵⁴Mn, ⁶⁰Co, ^110m Ag, ¹²⁵Sb, and ¹³⁴Cs using a small chromatographic alumina column. Samples of aged radioactive tellurium waste were dissolved in alkali solutions (1 and 5 M NaOH) and fed into preconditioned 1.0 g alumina columns at ～50°C. The columns were washed with 1 M NaOH or successively with 5 and 1 M NaOH. ⁶⁵Zn was quantitatively retained in the alumina column in the course of feeding and washing the column with 1 M NaOH. Solutions of NH₄Cl-NH₃, NH₄Cl-HCl, and HCl were studied as eluents for ⁶⁵Zn from the alumina column. 2 M HCl ensured the highest elution yield (88.7 ± 1.7%) with the ⁶⁵Zn radionuclidic purity of 99.4 ± 0.02%.