Determination of <Superscript>93</Superscript>Zr and <Superscript>237</Superscript>Np in nuclear power plant wastes

A combined radioanalytical method for determination of ⁹³Zr and ²³⁷Np (as well as other actinoids) in radioactive wastes has been developed. Analytes were co-precipitated on iron(II)-hydroxide, separated and purified on UTEVA columns, and detected by inductively coupled plasma mass spectrometry. According to Zr and Np, 65 and 75% yields were achieved, respectively.     

A selective separation method for <Superscript>93</Superscript>Zr in radiochemical analysis of low and intermediate level wastes from nuclear power plants

The zirconium isotope ⁹³Zr is a long-lived pure β-particle-emitting radionuclide produced from ²³⁵U fission and from neutron activation of the stable isotope ⁹²Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, ⁹³Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of ⁹³Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. A radiochemical procedure based on liquid–liquid extraction with 1-(2-thenoyl)-3,3,3-trifluoroacetone in xylene, ion exchange with Dowex resin and selective extraction using TRU resin has to be carried out in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique (LSC). To set up the radiochemical separation procedure for ⁹³Zr, a tracer solution of ⁹⁵Zr was used in order to follow the behavior of zirconium during the process by γ-ray spectrometry through measurement of the ⁹⁵Zr. Then, the protocol was applied to low level waste (LLW) and intermediate level waste (ILW) from nuclear power plants. The efficiency detection for ⁶³Ni was used to determination of ⁹³Zr activity in the matrices analyzed. The limit of detection of the 0.05 Bq l⁻¹ was obtained for ⁶³Ni standard solutions by using a sample:cocktail ratio of 3:17 mL for OptiPhase HiSafe 3 cocktail.     

Determination of <Superscript>237</Superscript>Np, <Superscript>93</Superscript>Zr and other long-lived radionuclides in medium and low level radioactive waste samples

The majority of long-lived radionuclides produced in the nuclear fuel cycle can be regarded as “difficult-to-measure” nuclides, hence chemical separation is needed before the nuclear measurement of them. A combined radiochemical procedure that enables the simultaneous determination of some “difficult-to-measure” nuclides in medium and low level radioactive wastes has been developed in our laboratory. Recently, this method has been extended for determination of ²³⁷Np and ⁹³Zr. ²³⁷Np and ⁹³Zr are pre-concentrated by co-precipitation on iron(II) hydroxide and zirconium oxide, separated by extraction chromatography using UTEVA, and measured by inductively coupled plasma mass spectrometry (ICP-MS). As even traces of polyatomic ions and isotopes at m/z 237 or 93 cause considerable interferences during ICP-MS detection, a purification step by extraction chromatography was needed. Analyzing real samples (evaporation concentrates of a nuclear power plant) 66–99% and 31–99% chemical yields were achieved for Np and Zr, respectively.     

Half-life of <Superscript>97</Superscript>Zr,re-measured

Summary The half-life of ⁹⁷Zr, used for the calculation of thermal/epithermal neutron flux ratio in k₀-NAA, is re-determined using three measurement systems with different pulse processing principles. The result of 16.755±0.013 hours clarifies the discrepancy between two widely used literature values, 16.744±0.011 and 16.90±0.05 hours. Different dead-time correction methods used on various measurement systems are evaluated. Factors influencing precise measurement of relative peak counting rates are discussed in time-series measurements over a dynamic range of 1000-fold radioactive intensities (10 half-lives).     

Quantification of <Superscript>241</Superscript>Pu in aerosols released from nuclear power plants

Two peptide ligands conjugated adenine, [9-N-(tritylmercapto acetyl diglycyl aminoethyl) adenine, Tr-MAG₂-Ade] and [9-N-(tritylmercapto acetyl triglycyl aminoethyl) adenine, Tr-MAG₃-Ade], are synthesized and labeled with ^99mTc by directly labeling method. The stability of ^99mTc-MAG₂-adenine and ^99mTc-MAG₃-adenine in vitro is measured. The uptake radios of tumor to muscle at 3h post-injection are 5.70 and 4.92, respectively. The biodistribution and scintigraphic imaging studies show that the two complexes have high localization in tumor and high contrasted tumor images can be obtained, which suggest their potential utility as tumor imaging agents. But the high radioactivity of abdomen could prevent the tumor imaging in this area.     

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%.     

Monitoring of the <Superscript>14</Superscript>C concentration in the stack air of the nuclear power plant VVER Jaslovske Bohunice

¹⁴C releases in the stack air of the NPPs V1 and V2, Jaslovske Bohunice was determined during the year 2004–2010. Radioactivity concentration of ¹⁴C in the stack air was determined in the forms of inorganic ¹⁴CO₂ and ¹⁴C _n H _m . The annual average activity concentration in the stacks air samples varies between 12 and 121 Bq m⁻³. NPP V1, starting with 45 Bq m⁻³ in 2005 is decreasing due to the shutting down of the reactors (the first reactor was shut down in December 2006 and the second reactor in December 2008). The average value of radioactivity concentration for power unit V2 was 32 Bq m⁻³ in 2004 and reached the value of 102 Bq m⁻³ in the first-quarter of the 2010. The average normalized yearly discharge rates were between 0.39 and 0.64 TBq GW_e⁻¹ year⁻¹ (2005–2008), NPP V1 and 0.19–0.61 TBq GW_e⁻¹ year⁻¹ (2004–first-quarter 2010) for NPP V2, Jaslovske Bohunice. Most of the discharged ¹⁴C is in a hydrocarbon form, (95% for Jaslovske Bohunice NPP V2), but the CO₂ fraction may reach 37% in the air stack for Jaslovske Bohunice V1.     

Determination of <Superscript>241</Superscript>Am/<Superscript>243</Superscript>Am ratios

Determination of ²⁴¹Am/²⁴³Am ratios is required for vanous purposes including assay of Am by isotope dilution techniques. Alpha-spectrometry on electrodeposited sources is a preferred technique for this determination. However, there is an inherent problem of tail contribution which necessitates the use of suitable algorithms to account for the same. Recently, in the frame of a Coordinated Research Program (CRP) of the International Atomic Energy Agency (IAEA), WinALPHA software has been developed which is a combination of an asymmetrical Gaussian for the main part of the peak and a low energy function. Therefore, it was of interest to compare the use of this algorithm with the routinely used method, in our laboratory, based on geometric progression (G. P.) decrease. Since, there are no reference materials available commercially for ²⁴¹Am/²⁴³Am ratios, synthetic mixtures covening a wide range (0.3 to 2.0) of ²⁴¹Am/²⁴³Am α-activity ratios were used and un-ignited electrodeposited sources were prepared for α-spectrometry. The α-spectra obtained using PIPS detector, were evaluated using the two algonthms The ²⁴¹Am/²⁴³Am α-activity ratios obtained were also compared with those determined by thermal ionization mass spectrometry (TIMS). An agreement of about 1% was obtained in the ²⁴¹Am/²⁴³Am ratios determined by the two methods and also by using the two algorithms for α-spectrum evaluation.     

<Superscript>1</Superscript>H and <Superscript>31</Superscript>P nuclear spin-lattice relaxation in C-phosphorylated oximes

The ¹H spin-lattice relaxation times of the proton-bearing groups and the ³¹P spin-lattice relaxation times in C-phosphorylated oximes R¹C(=NOH)P(=O)R²R³ (R¹ = Ph, R² = R³ = OMe; R¹ = Ph, R² = OMe, R³ = OCH²CH²Br; R¹ = PhCH², R² = R³ = OCHMe²) and dioxime R²P(=O)C(=NOH)(CH₂)₄C(=NOH)P(=O)R² (R = OMe) in DMSO-d₆ were measured. The characteristic reorientation times of the whole molecules were estimated using the measured values of the ¹H relaxation times and the results of semiempirical PM3 quantum chemical calculations of the molecular geometries. The reorientation times were used to identify the contributions of different relaxation mechanisms to the rate of ³¹P spin-lattice relaxation. The anisotropy of the chemical shielding of ³¹P nuclei was evaluated from the difference between the ³¹P relaxation rates measured at 101.27 and 161.92 MHz.     

Combined determination of <Superscript>99</Superscript>Tc and <Superscript>108m</Superscript>Ag in L/ILW liquid wastes

The low- and intermediate-activity level liquid wastes produced by the Paks Nuclear Power Plant (NPP) contain routinely measureable gamma-emitting (e.g., ⁵⁴Mn, ⁶⁰Co, ^110mAg, and ¹³⁷Cs) as well as many so-called “difficult-to-measure” radionuclides. Despite of their low specific activity compared to the total, the reliable determination of these radionuclides is an important issue of nuclear waste management. The increasing amount of waste samples to be qualified yearly by our laboratory put a pressure on revising the existing procedure of ⁹⁹Tc separation applied. We have managed to halve the initial amount of the sample required to achieve the same level of detection of technetium. Furthermore, one of the new purifying steps introduced have proved to be able to separate ^108mAg (and ^110mAg) better than 99% keeping the ⁹⁹Tc content of the product almost intact. Means of separation of ⁹⁹Tc from ¹⁰⁶Ru and ^124+125Sb have also been successfully investigated. As intended, this new procedure has a major impact on the chemical reagent as well as the electricity requirement of the separation making it more cost-effective.     

Determination of <Superscript>3</Superscript>H, <Superscript>226</Superscript>Ra, <Superscript>222</Superscript>Rn and <Superscript>238</Superscript>U in Austrian ground- and drinking water

Screening measurements for ³H, ²²⁶Ra, ²²²Rn and ²³⁸U in ground water were performed within a ground- and drinking water project in Austria. The aim of this project is to get an overview of the distribution of natural radionuclide activity concentration levels in ground water bodies. In some cases this water is used for drinking water abstraction. In this paper methods and results of the screening measurements are presented. Regions with high activity concentrations were identified and in these regions further investigation for ²²⁸Ra, ²¹⁰Pb and ²¹⁰Po will be conducted.     

A novel method for the cyclotron production no-carrier-added <Superscript>93m</Superscript>Mo for nuclear medicine

Amongst the various radioisotopes of molybdenum, ^93mMo in its no-carrier added form might be a potential candidate radionuclide in the field of nuclear medicine due to its suitable half-life (T _1/2 = 6.85 h) and for its intense gamma line at 263.14 keV. Aim of the presented study is to compare the calculated cross sections for the production of ^93mMo incident proton energy up to 30 MeV, theoretical calculation of production yield and calculation of target thickness requirement. Also, deposition of ⁹³NbO₂ on Cu substrate was carried out via two special sedimentation methods for the production of ^93mMo and a simple and selective extraction of molybdenum was described.     

Determination of <Superscript>228</Superscript>Ra in human bone ash containing significant quantities of <Superscript>40</Superscript>K and Ca<Superscript>2+</Superscript>

The determination of ²²⁸Ra by means of γ-spectrometry, in material containing significant quantities of ⁴⁰K and Ca²⁺ such as bone ash results in increased values of counting uncertainty and lower limit of detection (LLD) because of a significant contribution from the Compton continuum of ⁴⁰K. However, ⁴⁰K is widely removed from bone ash if ²²⁸Ra is coprecipitated with barium sulfate. As a result, the counting uncertainty and LLD are significantly reduced. A method is presented for determination of very low activity concentrations of ²²⁸Ra. Impurities introduced by precipitation are negligible when applying high resolution γ-spectrometry.     

Determination of activity ratios of <Superscript>238,239+240,241</Superscript>Pu, <Superscript>241</Superscript>Am, <Superscript>134,137</Superscript>Cs,and <Superscript>90</Superscript>Sr in Bulgarian soils

The origins of different artificial radionuclides found in soils from Northern and Southern Bulgaria was determined by measurements of their actual concentrations and respective ratios. On the basis of the measured mobility and concentrations of the investigated radionuclides in soils, it was estimated that after the Chernobyl accident the mean depositions of fresh ¹³⁷Cs were 3.0 ± 2.5 kBq/m² for Northern Bulgaria and 15 ± 7 kBq/m² for Southern Bulgaria. As a result of global fallout following atmospheric nuclear weapon tests in the 1950s, mean depositions (corrected to 1965) were calculated for Northern and Southern Bulgaria as follows: for ⁹⁰Sr—1.0 ± 0.5 and 2.3 ± 1.3 kBq/m², ²³⁸Pu—1.3 ± 0.8 and 2.8 ± 1.6 Bq/m², ^239+240Pu—15 ± 14 and 47 ± 38 Bq/m², and ²⁴¹Pu—520 ± 200 and 760 ± 260 Bq/m².     

Determination of <Superscript>59</Superscript>Ni in radioactive waste

Composite material PAN-DMG, containing chelating agent dimethylglyoxime (DMG) immobilized in porous matrix of binding polymer polyacrylonitrile (PAN), was used for nickel separation and concentration. Method for preparation of ⁵⁹Ni source for low energy photon spectrometry was developed using homogeneous precipitation of nickel with DMG. The proposed method was tested with two types of real radioactive waste (boric acid concentrate from nuclear power plant (NPP) evaporator and spent ion exchanger from NPP).     

Conversion of endohedral <Superscript>133</Superscript>Xe-fullerene to endohedral <Superscript>133</Superscript>Xe-fullerenol to be used in nuclear medicine

Summary Hydrophilic endohedral ¹³³Xe-fullerenols, [¹³³Xe@C₆₀(OH)_xand ¹³³Xe@C₇₀(OH)_x], were synthesized from hydrophobic endohedral ¹³³Xe-fullerenes. The yield of endohedral ¹³³Xe-fullerenols extracted in water was about 40% and 23% for C₆₀and C₇₀, respectively. The products stored in 0.9% NaCl solution at 20 °C were stable enough to be used in nuclear medicine.     

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.     

<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.