Radioactive Waste Samples Research Articles

Incorporation of trace metals in Hanford waste boehmite mineral phases and dissolution rate impacts

AbstractData from radioactive Hanford tank waste samples was analyzed to assess whether trace metals present were incorporated into the bulk boehmite matrix or were simply intermingled with the bulk material and subsequent impacts to the boehmite dissolution rate. Results suggest that chromium is primarily blended into the bulk boehmite, with a small fraction present on the surface of the solids. However, increasing the level of chromium incorporation 30‐fold decreased the dissolution rate of boehmite by only 8%, suggesting incorporation of chromium into these samples had a minimal impact on the dissolution rate. Iron was also found to be incorporated in the boehmite solids. Silicon appears to be simply intermingled and not blended into the boehmite crystals.

Removal of ionic and colloidal 110 mAg from radioactive wastewater using radiografted chelating adsorbents

This research investigates the performance of new radiografted materials as adsorbents for ionic and colloidal 110 mAg, a prominent pollutant in pressurized water reactor (PWR) wastewater. A comparative analysis was conducted to evaluate their effectiveness against commercially available resins. Due to the increasing demand for new nuclear power plants and the growing global dependence on nuclear energy, it is crucial to minimize radioactive waste production. The study was conducted on simulated inactive and radioactive waste samples with trace amounts of 110 mAg that are hard or impossible to detect by standard analytic techniques. The synthesized materials were labeled PGE (amine functional groups) and PGM (sulfhydryl and amine functional groups). Functionalization with amine and sulfhydryl groups was chosen since it is well known that these groups strongly interact with Ag in both ionic and colloidal form. The tested materials demonstrate exceptional performance as adsorbents by completely removing (100 %) both ionic (PGM) and colloidal (PGE) forms of 110 mAg. Additionally, depending on the adsorbed species excellent elution of 110 mAg was achieved (>94 % for ions and >50 % for colloids) implying possible material reusability. Manufacturing cost of such adsorbents is orders of magnitude lower per kg (10 – 50 $/kg), compared to prices of commercial resins (100 – 10000 $/kg). These results indicate that such inexpensive materials could potentially be used in an efficient and selective way to treat PWR wastewater for 110 mAg contamination while reducing the operating cost and minimizing resulting solid waste.

Determination of 93Zr, 237Np and Th radionuclides in radioactive waste and mineral samples: extension of the method for determination of actinides

A novel method has been developed for the simultaneous determination of difficult-to-measure (DTM) nuclides 93Zr, 237Np, 232Th, 230Th and 228Th using a DGA resin® (containing N,N,N′,N′-tetra-n-octyl-diglycolamide) column for simultaneous separation of Zr and actinides followed by purification of Zr, Np and Th on a TEVA resin® (containing Aliquat 336) column. Samples are destroyed by fusion with sodium hydroxide or by acid destruction using HNO3, HCl and HF. Pre-concentration procedure tailored to the extraction chromatography is based on co-precipitation of Zr and actinides and removal of alumina, silica, iron and calcium. The concentrations of 93Zr, 237Np and Th nuclides are determined by ICP-MS and α spectrometry, respectively. High chemical recoveries (≥ 73%) and separation factors were obtained for Zr, Np and Th both in evaporation concentrates of an NPP and various samples of mineral origin such as soil and concrete. The procedure can be easily combined with the simultaneous separation of other actinides (Pu, Am, U) using the same DGA column that has been described in our previous papers (Cassette et al. in Appl Radiat Isot 68:122–130, 2010 and Groska et al. in J Radioanal Nucl Chem 309(3):1145–1158, 2016).

Unusually Kinetically Inert Monocationic Neptunyl Complex with a Fluorescein-Modified 1,10-Phenanthroline-2,9-dicarboxylate Ligand: Specific Separation and Detection in Gel Electrophoresis

We found a singly charged Np(V)O2+ complex with unprecedented kinetic inertness in aqueous solution, one million times slower than the widely accepted fast kinetics of neptunyl complexes. An inert NpO2+ complex with a fluorescent 1,10-phenanthroline-2,9-dicarboxylate derivative was found by kinetic selection using polyacrylamide gel electrophoresis (PAGE) from a small chemical library. Autoreduction from Np(VI)O22+ to Np(V)O2+ via complexation was observed. A remarkably small spontaneous dissociation rate constant of 8 × 10-6 s-1 (half-life of 23 h) was determined using PAGE. Selective detection of Np(V)O2+ was achieved in PAGE with a detection limit of 68 pmol dm-3 (17 fg). This system was successfully applied to simulated radioactive waste samples. Our finding that electron-rich NpO2+ forms a uniquely inert complex with no strong electrostatic interaction reveals a new aspect of actinide chemistry for developing a novel separation system of real radioactive material samples.

Investigation of a new approach for 36Cl determination in solid samples using plastic scintillators

This work reports a new approach for the determination of 36Cl in radioactive waste samples from nuclear decommissioning, wherein novel plastic scintillator (PS) materials were used for the concentration of 36Cl prior to the detection with scintillation counting. Different plastic scintillator (PS) materials were tested for their selective absorption and detection of 36Cl activity in solid samples. PS microspheres (PSm), cross-linked PSm (CPSm) and PS resin have been investigated. PS resin was identified as the most suitable material for 36Cl analysis. Pyrolysis and subsequent trapping of the volatile elements in a bubbler was used. The trapping solution was finally loaded onto a cartridge of the PS resin. Scintillation counting and ion chromatography were used to determine the activity concentration and the chemical recovery, respectively.

Hybrid analytical method for calibrating a standard NaI(Tl) gamma-ray scintillation detector using a lateral hexagonal radioactive source

In various fields of radiation and nuclear physics, research and its applications help in the management and control of human activities, especially in the analysis of nuclear waste and measurement of low activity environmental samples. The current study uses a hybrid analytical technique and a 152Eu lateral hexagonal radioactive source to calibrate a standard cylindrical NaI(Tl) γ-ray scintillation detector. The technique is a combination of experimental measurements and theoretical calculations because for each environmental and/or radioactive waste sample, a separate reference calibration source equivalent to samples of different compositions is required, which needs additional funding. This method is suitable for any source–detector geometry. It can be used to calculate the geometric, total, and full-energy peak efficiency and the P/T ratio using the model probability of detecting γ-radiation photons, in particular from a lateral hexagonal radioactive source. The detection probability can be mainly influenced by the direction of the γ-quantum motion. This can help build more realistic and complex models than the typical point source case, where simple integration is used to numerically calculate the efficiency of the detection system. Using this method, the self-attenuation and coincidence-summing effect can be calculated for any gamma source matrices. The results obtained can show how useful this technique can be for calibrating complex gamma-ray detectors with any type of gamma-ray source.

Sorbent extraction behavior of cesium and strontium from nitric acid solutions using a new high thermal stability material

In this study a new low-cost carbonaceous material was prepared from husks of opuntia-ficus-indica as a starting material (precursor) which was accomplished by chemical activation route using H3PO4 impregnation. The material has been identified by different analytical tools. The sorption performance of Cs(I) and Sr(II) from HNO3 solutions was examined through batch system. Variations of the distribution coefficients (Kd) as a function of HNO3 concentration in the range 0.001–5.0 M were presented. Some of separation probabilities were suggested. The results attained signals that the Sr(II) selectivity is higher than that of Cs(I) at high molarities. The retention capacity (qe) of Cs(I) and Sr(II) ions increased with growing temperature. The capacities at 0.001 M HNO3 are 34 and 108 mg/g for Cs(I) and Sr(II), respectively. Whereas, at 2.0 M HNO3 capacities were about 4 and 37 mg/g for each of Cs(I) and Sr (II), respectively. This studies demonstrates that the prepared carbonaceous sorbent is an economically effective sorbent for retention of Cs(I) and Sr(II) species from HNO3 solutions. Cs(I) and Sr(II) removal potential was tested from simulated low- and intermediate-level radioactive waste samples.

Active single-shot interrogation of natural uranium in industrial radioactive waste using plasma focus device

An Active Interrogation by Delayed Neutron Counting (AIDNEC) system for the assay of natural uranium in industrial waste in form of civil debris has been developed using the plasma focus (PF) device as a single-shot neutron source. The methodology of the technique involves irradiation of fissile mass by neutrons from a PF device and counting the delayed neutron emitted as a result of induced fission. In-house developed medium energy (17 kJ) PF device and 3He detector banks have been used for neutron irradiation of waste samples containing fissile masses and counting of the delayed neutrons respectively. The detector geometry and the moderators for the source and delayed neutrons have been optimized by Monte Carlo simulations. In this work, the results of these simulations and experimental observations are presented. The industrial radioactive waste samples used in this work were in the form of sand, bricks, chunks of concrete, etc. having a sample mass of few tens of kg. With D2 gas operated PF device (average yield (2.1 ± 0.8) ×108 neutrons/shot) the detection limit is determined to be 170 g of natural uranium, i.e., 1.2 g of 235U. The feasibility of the technique using the same PF device operated with D–T gas (14.06 MeV neutrons) has been assessed by Monte Carlo simulations. As D–T fusion provides two orders more neutron yield as compared to D2 fusion, the detection limit of the present system with D–T gas is estimated to be 3.3 g of natural uranium in ∼40 kg waste. Considering this low detection limit, this technique can be used for the assay of nuclear material in all types of industrial radioactive waste produced in the fuel fabrication/handling or similar facilities.

Influence of sorption parameters on cesium-137 removal using modified activated carbon obtained from corchorus olitorius stalks

Abstract Removal of 137Cs radionuclides from the environment has engrossed the concern of researchers after Fukushima accident. The leakage of radioactive cesium ions can lead up to surface and groundwater contamination, and this leads to pollution of drinking water sources. In this work, corchorus olitorius stalks has been used as a novel precursor for production of low-cost meso porous a ctivated c arbon ( Meso-AC ) and HNO3/H2O2- m odified Meso-AC ( m-Meso-AC ). The physicochemical properties of all adsorbents were evaluated. The influences of sorption parameters and presence of some ligands (humic acid, fulvic acid, and EDTA) on the sorption of 137Cs were studied. The maximum 137Cs capacity of m-Meso-AC was found to be 58.74 mg/g. Efficiency of the new adsorbent to remove 137Cs radionuclides from natural waters (tap, river, and groundwater) was investigated. The studies showed that new adsorbent could be used as promising material for the retention of 137Cs from real radioactive waste and natural water samples.

Environmental impact assessment of waste-water: radionuclides use in hospitals (Abruzzo, Italy, 2000-2015).

Radionuclides are increasingly used in hospitals for diagnostic and therapeutic purposes, such as functional research, diagnostic imaging, and in the performance of radioiodine therapy. Their use produces radioactive waste, and risks environmental contamination. The present study involves 486 samples of radioactive waste produced in hospitals in Abruzzo, Italy, during 2000 - 2015. Measurements were carried out with the gamma spectrometry technique: Germanium detector (PTG) with 8000 acquisition channels, power range 59 - 1836 keV, resolution 1 keV, overall efficiency 30%, measurement time 60000 Seconds. The radionuclides involved were as follows: 131I, 99mTc, 67Ga, 201Tl, 123I, 111In, with substantial activity in 44 samples. Checks allowed us to certify the levels of radioactive concentration in waste-water, plan for their suppression, and optimise the management procedures.

Determination of 94Nb in radioactive wastes using extraction chromatography and gamma spectrometry

A radiochemical methodology for the determination of 94Nb in low-level radioactive wastes from nuclear power plant was proposed. Although 94Nb is a strong gamma emitter, its concentration in radioactive waste samples is usually several orders of magnitude lower than that of other β–γ emitters, whose emissions interferes in the detection of the emission lines of 94Nb. The procedure involves acid digestion, precipitation, cation exchange chromatography by using Amberlite IRA120 resin, extraction chromatography by using TEVA resin to isolate the Nb and the gamma spectrometry to its measurement. The chemical yield was 70% in average. Samples of evaporator concentrate and spent resin were analyzed. For the samples of the evaporator concentrate, the results obtained were below LD = 9.59 × 10−4 Bq g−1. For the spent resin an average result of 1.54 × 102 Bq g−1 was obtained.

Determination of 129I using volatilization method and liquid scintillation spectrometry

A simple and rapid separation method for 129I determination in radioactive waste samples was developed. Suitable conditions for iodine volatilization were tested. Iodine was trapped in 1.5 mol L−1 NaOH and precipitated as PdI2·H2O by addition of PdCl2 with recoveries higher than 80%. The method was applied for analysis of contaminated soil, radioactive sludge, evaporator concentrate and heterogeneous waste samples from nuclear power plants in Slovak Republic. 129I was measured on liquid scintillation counter TRI CARB 2900 TR using Ultima Gold AB scintillation cocktail.

Radioactive waste management of Y-90 microspheres used for radioembolization technique

Introduction The radioembolization technique is used for the treatment of liver tumors, with the implantation of glass microspheres doped with yttrium-90 (Y-90). Specific radiation protection procedures, particularly those regarding radioactive waste managements are required. Several publications indicate the presence of radionuclidic impurities in those microspheres, resulting in additional radiation safety concerns, with possible impact on the minimum radioactive waste storage time. Purpose Determination of the effective half-life time of Y-90 radioactive waste samples, in order to establish a minimum storage time before discharge, and the identification of potential radionuclide impurities of Y-90. Materials and methods Exposure rate and surface contamination measurements were performed on plastic containers containing residues from Y-90 microsphere delivery procedures. The gamma energy spectrum analysis of the decayed Y-90 samples was also obtained using a NaI(Tl) detector. Results Waste emission values higher than those expected for a pure Y-90 decay were measured, even after 30 days after the procedure, suggesting the presence of impurities. An larger effective half-life time days was determined when compared with pure Y-90. Spectrum peaks were observed suggesting the presence of Y-88. Conclusion For samples decayed up to 3 months, these findings impose an increase of the waste decay time for Y-90 waste containers when compared with pure Y-90. Spectrum analysis suggests the presence of Y-88. Further analysis are being made in samples decayed greater than 6 months, that include the spectrum acquisition in order to assess the presence of other potential impurities. Disclosure The authors have noting to disclose.

Preparation of activated carbon from doum stone and its application on adsorption of 60Co and 152+154Eu: Equilibrium, kinetic and thermodynamic studies

Removal of radionuclides from wastewater before discharging to environment is necessary for the safety of living beings. Activated carbon prepared from doum stone (DS), an agricultural waste by-product, has been used for the sorption of 60Co and 152+154Eu radionuclides from aqueous solutions. DS has been characterized by different analytical tools. The efficiency of the adsorbent was investigated using batch sorption technique under different experimental conditions. The equilibrium sorption data were analyzed using Freundlich and Langmuir isotherm models. The kinetic experimental data were analyzed using four kinetic models including pseudo-first order, pseudo-second order, Elovich and intraparticle diffusion model to examine the mechanism of sorption and potential rate-controlling step. The maximum capacity of DS was found to be 121 mg/g and 166 mg/g for cobalt and europium, respectively. Sorption efficiency of DS to remove 60Co, 152+154Eu and 134Cs from real radioactive wastewater and environmental (river water and sea water) samples was also investigated. The results revealed that the prepared DS as low cost could be used as a promising material for the simultaneous removal of different radionuclides such as 60Co, 152+154Eu and 134Cs, or trivalent actinides such as 241, 242m,243Am from real radioactive waste and environmental water samples.

Determination of actinides in radioactive waste after separation on a single DGA resin column

A new procedure was developed for the separation of uranium, thorium, plutonium, neptunium and americium based on extraction chromatography using the N,N,N′,N′-tetraoctyldiglycolamide sorbed onto Amberchrom CG-71 (available as DGA resin® from Triskem International). Radioactive waste samples were wet ashed, actinides were preconcentrated by coprecipitation with ferrous hydroxide. The dissolved precipitate was loaded from a reductive media on the column and the actinides were sequentially eluted changing the nature and the acidity of the eluents, the oxidation states of the given actinides, the concentration of the complexing agent, and the temperature. Alpha sources were prepared by micro-coprecipitation from the strip solutions without further processing. High recoveries and decontamination factors were obtained in each alpha source.

Erratum to: Rapid quantification of alpha emitters in low- and intermediate-level dry radioactive waste

In the original publication, the units of the activities in Tables 2 and 3 are incorrect and they need to be corrected to Bq/kg. In addition, ‘Activity of Am’ in the right column in Table 3 needs to be corrected to ‘Activity’. Therefore, corrected Tables 2 and 3 are given below: Table 2 Activity measurement of Am and Cm isotopes in the radioactive dry waste sample using alpha spectrometry Sample # Activity (Bq/kg) Am recovery (%) Am Cm Cm

Rapid separation of 99Tc, 90Sr, 55Fe, 94Nb, and 59,63Ni in radioactive waste samples

A procedure for the rapid separation of 99Tc, 90Sr, 55Fe, 94Nb, and 59,63Ni in radioactive waste samples was developed for measuring their activity. After separating 99Tc on an anion-exchange resin column, 90Sr was separated as 90Sr(CaC2O4) co-precipitates, and 59,63Ni, 55Fe, and 94Nb were sequentially separated on an anion-exchange resin column in the presence of oxalate. The separated 90Sr and 59,63Ni were further purified on extraction chromatographic resin columns, respectively. The chemical yields of the separation procedure for Re, Sr, Fe, Nb, and Ni, which were added as chemical carriers, are 99.3 ± 0.3 %, 94.8 ± 0.6 %, 93.3 ± 1.2 %, 95.3 ± 1.9 %, and 92.6 ± 0.6 %, respectively. The separation of these radionuclides from sixteen radioactive waste samples required only seven person-days.

Determination of 126Sn in radioactive waste using TEVA® resin and gamma spectrometry

A new method employing TEVA® resin from Eichrom Technologies for determination of 126Sn in nuclear waste was developed. The influence of stable tin carrier and hydrochloric acid concentration on the separation and chemical yield of 113Sn was tested. The method is suitable for the separation of 126Sn isotope from hydrochloric acid solution. For the determination of 126Sn from radioactive waste samples, tin was first precipitated with ammonium sulfide in 0.5 mol L−1 HCl, dissolved in concentrated HCl, loaded in 1 mol L−1 HCl on TEVA® resin column and eluted with 2 mol L−1 HNO3 acid. Eluted tin fraction was directly measured by gamma spectrometry on HPGe detector.

Xenon Gamma-detector Applicability for Identification and Characterization of Radioactive Waste

In this paper described applicability of xenon gamma detector for identification and characterization of radioactive waste was researched. Standard calibration gamma ray sources were used to determine real physical and technical characteristics of xenon gamma spectrometer. Samples of radioactive waste were measured by xenon gamma detector for identification and characterization.

Determination of ultratrace-levels of 99Tc in low-level radioactive waste samples using ICP-QMS

This paper proposes a rapid, accurate method of measuring 99Tc in low-level radioactive waste using microwave digestion, TEVA resin for chemical separation, and inductively coupled plasma-quadrupole mass spectrometry (ICP-QMS). The minimum detectable activity of ICP-QMS for the determination of 99Tc was 8.5 mBq g−1 (13.6 pg g−1). The proposed method achieved chemical recovery of 88 % and with sensitivity superior to that of alternative radiochemical methods (e.g., liquid scintillation analyzer or LSA) when the data acquisition time for identical, low-concentration samples was taken into account.