Application of an experimental design to optimize a segregation method of <Superscript>129</Superscript>I and <Superscript>14</Superscript>C

In this work, the optimization of a segregation method of ¹²⁹I and ¹⁴C, two long-living radionuclides, main constituents of nuclear radioactive waste, has been developed. To be able to carry out this project, a fractional factorial experimental design was applied using 5 factors and 2 levels by factor (2^5–2). Only 8 experiments were necessary to identify the variables affecting the process, and very good recoveries of both radionuclides were obtained: (94?±?2)% for ¹²⁹I, and (99?±?1)% for ¹⁴C. The segregation of ¹²⁹I was influenced by flow (Q), volume of H₂SO₄ (V_H+), and carriers (CR), while V_H+ and time (t) played a major role in the segregation of ¹⁴C.     

Determination of <Superscript>90</Superscript>Sr, <Superscript>129</Superscript>I and gross beta radioactivity concentration in some teas

In this study, ⁹⁰Sr (540 keVβ ⁻), ¹²⁹I (150 keVβ ⁻) and the gross beta radioactivity concentrations were determined for the samples of tea as the most leading consumed hot drink in the markets (processed and packaged for sale) in our country. Furthermore, the obtained data were statistically analyzed. For determination of ¹²⁹I (150 keVβ⁻), ⁹⁰Sr (540 keVβ⁻) and gross radioactivity concentrations in tea samples, a sensor system consisting of scintillation detector with BP4 probe sensitive to beta radiation and a radiation meter (ST7) configurable for windows at desired power was used.     

Experimental and FLUKA study relevant to the production of <Superscript>124</Superscript>I through <Superscript>nat</Superscript>Te(p,xn) reaction at a medium-sized cyclotron

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 L _D = 9.59 × 10^?4 Bq g^?1. For the spent resin an average result of 1.54 × 10² Bq g^?1 was obtained.     

New measurement of <Superscript>129</Superscript>I in IAEA-375 reference material

Summary Under an IAEA project for upgrading of reference materials, a new determination of the ¹²⁹I concentration in the IAEA-375 reference material was performed. A chemical procedure was set up for the preparation of the AgI samples. Measurement of ¹²⁹I was carried out using the IsoTrace Tandetron AMS facility at University of Toronto. To ensure the accuracy of the calibration, the tuning of the AMS system was iterated using not only the QC (quality control) samples but also all unknown samples. To minimize any possible current-dependent effects between ¹²⁹I and ¹²⁷I ions in the injection magnet, low Cs⁺ sputtering beam intensity (10 μA) was used. The reproducibility in determining the ¹²⁹I/¹²⁷I ratio in the IAEA-375 AgI samples was less than 1%. The activity concentration (C_A) of ¹²⁹I in the IAEA-375 reference material was determined to be 1.59±0.08 mBq ^. kg^-1 at 95% confidence level. The present value is about 7% lower than the IAEA recommended value (1.7 mBq ^. kg^-1) listed in 2000 or 20% lower than the recommended value (2 mBq ^. kg^-1) listed in the IAEA AQCS Reference Material Catalogue (2002-2003). Since the IAEA recommended values for IAEA-375 materials was issued about 10 years ago and error range of the recommended values were large, the results we obtained might be useful in upgrading the recommended value.     

Automated radiochemical analysis of total <Superscript>99</Superscript>Tc in aged nuclear waste processing streams

Summary A fully automated analysis procedure and instrument for the measurement of total ⁹⁹Tc in aged nuclear waste has been developed. The overall analysis approach is based on a fully automated wet radiochemical analysis method. Microwave-assisted sample oxidation is used prior to a chemical separation step in order to oxidize all of the non-pertechnetate species to pertechnetate. Separation of the pertechnetate from interfering radioactive and stable matrix species is carried out using an anion-exchange column. The separated ⁹⁹Tc is quantified using a flow-through solid cell scintillation detector. The instrument is capable of an analysis time of <13 minute per sample with a detection limit of 2000 dpm/ml. Nuclear waste samples from the Hanford site with a high content of non-pertechnetate species were successfully analyzed using this method.     

Leaching kinetics of <Superscript>54</Superscript>Mn and <Superscript>89</Superscript>Sr radionuclides fixed in cement­vermiculite composition

Summary Leach characteristics of ⁵⁴Mn and ⁸⁵Sr radionuclides from ordinary Portland cement have been studied using International Atomic Energy’s (IAEA) standard leach method. The retardation factors, K_F, and coefficients of distribution, k_d, have been determined using a simplified mathematical model for analyzing the migration of radionuclides. The lowest leaching values after 60 days were achieved in samples with 5% of vermiculite. Results presented in this paper are the examples of results obtained in a 10 year mortar and concrete testing project, which will influence the design of the engineered trench system for a future central Serbian radioactive waste storage center.     

Magnesium-Potassium Phosphate Matrix for Immobilization of <Superscript>14</Superscript>C

Possibility of using a low-temperature magnesium-potassium phosphate matrix to solve the problem of immobilizing the radioactive wastes containing radioactive carbon (¹⁴C) in the form of calcium carbonate was examined. The physicochemical characteristics of the compounds obtained were determined. Large values of the ultimate compression strength (22 ± 5 MPa), which satisfy the technical requirements for cemented radioactive wastes (no less than 4.9 MPa), were obtained. The minimum carryover of carbon dioxide into the atmosphere in the course of synthesis and in keeping of samples for 14 days was noted: not more than 3 wt % relative to the starting CaCO₃. The leaching rate of carbonate ions from magnesium-potassium compounds by 28th day of contact with air does not exceed 10?9 g cm^?2 day^?1, with this value for the rest of the compound components not exceeding 10^?4 g cm^?2 day^?1. Thus, it was found that the magnesium?potassium phosphate matrix is an alternative to the cementation for solidification of radioactive wastes containing ¹⁴C.     

Separation and determination of <Superscript>90</Superscript>Sr in low- and intermediate-level radioactive wastes using extraction chromatography and LSC

A methodology for the determination of ⁹⁰Sr in low- and intermediate-level radioactive wastes from nuclear power plants is presented in this work. It is a part of a methodology developed for the sequential radiochemical separation of radionuclides difficult-to-measure directly by gamma spectrometry in these radioactive wastes. The separation procedure was carried out using precipitation and extraction chromatography with Sr Resin, from Eichrom and the ⁹⁰Sr was measured by liquid scintillation counting (LSC). Optimum conditions for the pretreatment, separation and LSC measurements were determined using simulated samples, which were prepared using standard solutions and carriers. The procedure showed to be rapid and achieved a good chemical yield, in the range 60–90%, and a detection limit of 6.0 × 10⁻⁴ Bq g⁻¹. The method was also tested by participation in a national intercomparison program, with aqueous samples, with good agreement of results.     

<Superscript>239</Superscript>Np as a tracer of <Superscript>237</Superscript>Np in effluent samples and low level nuclear waste

In this paper a technique to separate and measure both isotopes (²³⁷Np and ²³⁹Np) together is presented. A combined shape pulse discrimination liquid scintillation measurement with gamma-spectrometry, permits a precise measurement after the radiochemical separation. This technique was carried out by using an Eichrom chromatographic column (TEVA) as the first step of a more complete method, applied in the Nuclear Regulatory Authority, to separate actinides in nuclear waste and liquid effluents. The MCA is 0.08 Bq/l by alpha-spectrometry and 0.22 Bq/l (2σ) by liquid scintillation counting (LSC) for 93.7% of measurement efficiency and 98.4% of chemical recovery.     

Determination of <Superscript>3</Superscript>H and <Superscript>14</Superscript>C in the Dry Active Wastes (DAW) generated from the Light Water Reactors (LWR) by a wet oxidation method

³H and ¹⁴C Measurements of the dry active waste (DAW), such as the cotton, paper, and vinyl, generated from a nuclear power plant (NPP) were conducted with wet oxidation using open vessel equipment based on simulation results. The recovery efficiency with the simulated samples was around 93% with a relative standard deviation (RSD) of 1–3%. A liquid scintillation counter (LSC) was used for counting and adjusted to a quenching correction curve. The counting value was evaluated for the minimum detectable activity (MDA), which was found to be about 4 × 10⁻¹ Bq/g for ³H and 2 × 10⁻² for ¹⁴C when approximately 5 g of the samples were measured. The measured DAW samples for the cotton, paper, and vinyl generated from NPP achieved of RSD values of 25, 25, and 60%, respectively, for ³H and 0–50% for ¹⁴C.     

Sequential separation of <Superscript>99</Superscript>Tc, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe, <Superscript>90</Superscript>Sr and <Superscript>59/63</Superscript>Ni from radioactive wastes

A sequential separation procedure has been developed for the determination of ⁹⁹Tc, ⁹⁴Nb, ⁵⁵Fe, ⁹⁰Sr and ^59/63Ni in various radioactive wastes generated from nuclear power plants. Ion exchange and extraction chromatography were adopted for individual separation of the radionuclides. Precipitation was supplementarily utilized for both purification of the individual radionuclides and preparation of the radionuclide sources for use in a radioactivity measurement. The chromatographic separation behavior of the radionuclides both from the sample matrix metals and from one another was investigated using stable metals, Re (as a surrogate of ⁹⁹Tc), Nb, Fe, Sr and Ni. The validity of the procedure for reliability and applicability was evaluated by measuring the recovery of the metal carriers added to synthetic radioactive waste solutions. The recoveries by the chromatographic separation were in the range of 84.8 to 102.2% with 2s of less than 8.6%, the recoveries by the precipitation being in the range of 84.3 to 97.3% with 2s of less than 10.9%.     

Advances in <Superscript>79</Superscript>Se analyses on Savannah river site radioactive waste matrices

Waste cleanup efforts underway at the United States Department of Energy’s (DOE) Savannah River Site (SRS) in South Carolina, as well as other DOE nuclear sites, have created a need to characterize ⁷⁹Se in radioactive waste inventories. Successful analysis of ⁷⁹Se in high activity waste matrices is challenging for a variety of reasons. As a result of these unique challenges, the successful quantification of ⁷⁹Se in the types of matrices present at SRS requires an extremely efficient and selective separation of ⁷⁹Se from high levels of interfering radionuclides. A robust ⁷⁹Se radiochemical separation method has been developed at the Savannah River National Laboratory (SRNL) which is routinely capable of successfully purifying ⁷⁹Se from a wide range of interfering radioactive species. In addition to dramatic improvements in the K_d, ease, and reproducibility of the analysis, the laboratory time has been reduced from several days to only 6 h.     

Reduction of <Superscript>68</Superscript>Ge activity containing liquid waste from <Superscript>68</Superscript>Ga PET chemistry in nuclear medicine and radiopharmacy by solidification

PET with ⁶⁸Ga from the TiO₂- or SnO₂- based ⁶⁸Ge/⁶⁸Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity (⁶⁸Ge vs. ⁶⁸Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of ⁶⁸Ge activity is produced by eluting the ⁶⁸Ge/⁶⁸Ga generators and residues from PET chemistry. Since clearance level of ⁶⁸Ge activity in waste may not exceed 10 Bq/g, as stated by European Directive 96/29/EURATOM, our purpose was to reduce ⁶⁸Ge activity in solution from >10 kBq/g to <10 Bq/g; which implies the solution can be discarded as regular waste. Most efficient method to reduce the ⁶⁸Ge activity is by sorption of TiO₂ or Fe₂O₃ and subsequent centrifugation. The required 10 Bq per mL level of ⁶⁸Ge activity in waste was reached by Fe₂O₃ logarithmically, whereas with TiO₂ asymptotically. The procedure with Fe₂O₃ eliminates ≥90% of the ⁶⁸Ge activity per treatment. Eventually, to simplify the processing a recirculation system was used to investigate ⁶⁸Ge activity sorption on TiO₂, Fe₂O₃ or Zeolite. Zeolite was introduced for its high sorption at low pH, therefore ⁶⁸Ge activity containing waste could directly be used without further interventions. ⁶⁸Ge activity containing liquid waste at different HCl concentrations (0.05–1.0 M HCl), was recirculated at 1 mL/min. With Zeolite in the recirculation system, ⁶⁸Ge activity showed highest sorption.     

Measurement of <Superscript>36</Superscript>Cl in nuclear wastes and effluents: Validation of a radiochemical protocol with an in-house reference sample

Summary ³⁶Cl is a beta-emitter with a very low specific activity. It is produced during the irradiation of nuclear fuel, in the reactor core of power plants, from neutron capture by stable ³⁵Cl that may be present at trace level in any part of the irradiated material. Due to its long half-life (T_1/2 = 3.01 ^. 10⁵ y), ³⁶Cl may be significant in impact assessment studies of disposal sites of nuclear wastes. Considering these different elements, the National Radioactive Waste Management Agency (Andra-France) requests information on the ³⁶Cl content of the waste packages destined to be stored at Andra sites. As for other halogens, the measurement of ³⁶Cl is a difficult analytical task in view of its potential losses during the different chemical steps and also because of the lack of international certified reference material needed to validate the chemical and measurement procedures. This paper describes the methodology processed to constitute an in-house solid reference sample with a known content of stable and radioactive chlorine. The original radiochemistry developed to extract ³⁶Cl from solid samples and purify it before a liquid scintillation counting is explained. The comparison of the results given by this radiochemical protocol and other methods allow its validation. The replication of the measurements on the constituted reference materials gives a repeatability around 8% at a confidence level of 95% that is very close to the calculated combined uncertainty value.     

Determination of <Superscript>151</Superscript>Sm and <Superscript>147</Superscript>Pm using liquid scintillation tracer methods

The long-lived rare earth isotopes ¹⁵¹Sm (90 years, β _max = 76.3 keV) and ¹⁴⁷Pm (2.62 years, β _max = 224.6 keV) are low-yield fission products that generally require lengthy separation procedures to isolate and count by their beta emissions. We will describe novel liquid scintillation counting techniques using radioactive tracers to determine radiochemical yields from an environmental matrix. The recovery of ¹⁵¹Sm is determined from the alpha decay (2.25 MeV) of ¹⁴⁷Sm in the natural Sm carrier and is in excellent agreement with the gravimetric recovery. The ¹⁴⁷Pm recovery is determined by the use of ¹⁴⁵Pm (17.7 years, EC) tracer, custom-produced at LANL using an isotopically enriched target of ¹⁴⁴Sm. We have determined the ¹⁴⁵Pm recovery both from the 37.4 keV k_α1 X-ray, and the electron-capture emissions by LSC. A comparison of these recovery methods is presented.     

Improvements on the method for determining of <Superscript>210</Superscript>Pb and <Superscript>210</Superscript>Po in lead

An improved method is proposed to determine the content of ²¹⁰Pb in lead using ²¹⁰Po measured by alpha-ray spectrometry. This improved method, which is based on radiochemical separation by DDTC–toluene extraction, employs EDTA and citrate as masking reagents for the lead ions. To selectively extract polonium from an alkaline solution, the pH dependency was examined using a liquid scintillation counting method. And pH 9 was chosen as an extraction condition. Then ²¹⁰Po was electrodeposited on a stainless steel disk, and the chemical recovery was followed by ²⁰⁹Po tracer. The effectiveness of the new method was validated by the agreement with the analytical results from five samples as determined by gamma-ray spectrometry.     

Preparation and measurement of the half-life of <Superscript>101</Superscript>Tc

¹⁰¹Tc is a very important nuclide as fuel burn up monitor, and its half-life value has been measured many times, however, they were so different from each other. In this work, ¹⁰¹Tc liquid samples were prepared by irradiating analytical pure (NH₄)₆Mo₇O₂₄·4H₂O solution in Miniature Neutron Source Reactor (MNSR) of China. A rapid procedure which takes only 5 min was developed to separate ¹⁰¹Tc samples. The final samples were analyzed by γ-ray spectrum using high purity Germanium (HPGe) multi-channel analysis system. The results showed that there were no peaks of other nuclides except ¹⁰¹Tc. The half-life of ¹⁰¹Tc was accurately measured with HPGe γ-detector following 306.8 keV γ-ray for about 140 min, and three methods R-value method, iterative method and translation method were adopted to process the data. Finally, a more precise and accurate value 14.02 ± 0.01 min was given and compared with former measured data.     

<Superscript>40</Superscript>K activities and potassium concentrations in tobacco samples of Mexican cigarettes

Nine brands of tobacco cigarettes manufactured and distributed in the Mexican market were analyzed by γ-spectrometry to certify their non-artificial radioactive contamination. Since natural occurring radioactive materials (NORM) ⁴⁰K, ²³²Th, ²³⁵U, and ²³⁹U (and decay products from the latter three nuclides) are the main sources for human radiation exposure, the aim of this work was to determine the activity of ⁴⁰K and potassium concentration. Averages of ⁴⁰K and potassium concentration were of 1.29±0.18 Bq·g⁻¹, and 4.0±0.57%. The annual dose equivalents to the whole body from ingestion and inhalation of 26 Bq ⁴⁰K were 0.23 μSv and 15.8 μSv, respectively. The corresponding 50 years committed dose equivalents was 0.23 μSv. The total committed dose to the lungs due to inhalation of ⁴⁰K in tobacco was 16 μSv. Potassium concentrations obtained in this work were in the same range of those obtained by INAA, so showing that the used technique is acute, reproducible, and accessible to laboratories equipped with low background scintillation detectors.     

Systematic radiochemical separation for the determination of <Superscript>99</Superscript>Tc, <Superscript>90</Superscript>Sr, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe and <Superscript>59,63</Superscript>Ni in low and intermediate radioactive waste samples

A simple and rapid separation procedure was systemized for the determination of ⁹⁹Tc, ⁹⁰Sr, ⁹⁴Nb, ⁵⁵Fe and ^59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of ⁹⁹Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of ^134,137Cs and ^110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.     

Migration and sorption of <Superscript>137</Superscript>Cs and <Superscript>152,154</Superscript>Eu in crushed crystalline rocks under dynamic conditions

In migration experiments, sorption of ¹³⁷Cs and ^152,154Eu in the columns of crushed crystalline rocks of 0.25–0.8 mm grain size under dynamic flow conditions from the synthetic groundwater (SGW) has been studied. Five samples of crystalline rocks from Cavernous Gas Reservoir near Příbram were taken. Plastic syringes of 8.8 cm length and 2.1 cm in diameter were used as columns. The water phase was pumped downward through the columns, using a multi-head peristaltic pump, with a seepage velocity of about 0.2 cm/min. The radioactive nuclides, containing chemical carriers, were added into the water stream individually in the form of a short pulse. Desorption experiments were carried out with 2:1 (v/v) mixture of H₂SO₄ and HNO₃. In the columns the longitudinal distribution of the residual ¹³⁷Cs and ^152,154Eu activities was also determined. By the evaluation of respective breakthrough and displacement curves, the experimental and theoretical retardation factors, distribution coefficients and hydrodynamic dispersion coefficients were determined using the integrated analytical form of a simple advection-dispersion equation (ADE). Dynamic sorption experiments were also compared with the results of static sorption experiments. The paper was presented in part as a poster No. PB1-1 at the 11th International Conference Migration’ 07, held in Munich, Germany, August 26–31, 2007, Abstracts, p. 212.