Distribution of 14C and 3H in low level radioactive wastes generated by wet waste streams from pressurized water reactors

Summary It is impossible to detect ¹⁴C and ³H by direct methods such as γ-spectroscopy because they are pure b-emitters and thus they are classified as hard to measure nuclides (HTM). In this paper the analysis results of ¹⁴C and ³H in the low level radioactive wastes (LLWs), including spent ion exchange resin, evaporated bottom and sludge are presented. The LLWs were generated by three nuclear power plants (NPPs), in Korea all with pressurized water type reactors (PWRs). A simultaneous separation procedure for ¹⁴C and ³H in LLWs was established by wet oxidation-acid stripping. A liquid scintillation analyzer was used for the measurement of ¹⁴C and ³H. It was found that the recovery of ¹⁴C and ³H was 82-99 and 78-103%, respectively, by wet oxidation-acid stripping with diluted standard solutions. At the lowest injection of ¹⁴C and ³H, i.e., at 1.44 Bq for ¹⁴C and 1.22 Bq for ³H, the minimum detectable activity (MDA) of ¹⁴C and ³H was calculated as 0.88 and 0.78 Bq/g, respectively, for the minimum allowable sample weight, using wet oxidation and 16 wt% H₂SO₄ acid. By the wet oxidation-16 wt% H₂SO₄ stripping method no interfering nuclides were detected in the trapping solution of ¹⁴CO₂ and the distillate of ³H. The activity concentration range of ¹⁴C in the analyzed samples, i.e., spent ion exchange resin, evaporated bottom and sludge, was 0.17-110,000, 8.4-1380 and 0.1-10,006 Bq/g, respectively, and that of ³H in the same was from no detectable to 769, 134-14,383 and 0.7-4820 Bq/g, respectively.     

Gasification characteristics to <Superscript>14</Superscript>CO<Subscript>2</Subscript> of <Superscript>14</Superscript>C radionuclide desorbed from spent resin by phosphate solutions

The removal characteristics of H¹⁴CO₃ ⁻ ions from IRN-150 mixed resin contaminated with ¹⁴C radionuclide and the gasification effects of ¹⁴C radionuclide on ¹⁴CO₂ are investigated in this study. The stripping solutions used for the removal of ¹⁴C from spent resin are NaNO₃, Na₃PO₄, NH₄H₂PO₄, and H₃PO₄. The influence of the stripping solution concentration on the desorption characteristics of an inactive HCO₃ ion into a stripping solution from IRN-150 mixed resin and the gasification of this ion to CO₂ is analyzed. The gasification behavior to CO₂ using NaOH, HNO₃, and HCl was also compared to that of phosphate solution. Spent resin stored in Wolsong nuclear power plant is used to evaluate the gasification characteristics of ¹⁴C radionuclide to ¹⁴CO₂. Gamma radionuclides such as ¹³⁷Cs and ⁶⁰Co in residual striping solutions after desorption experiments are analyzed.     

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.     

Distribution of <Superscript>14</Superscript>C, <Superscript>90</Superscript>Sr and <Superscript>228</Superscript>Th in an elephant tusk

The period of date of death of an elephant can be assessed by analyzing four different radionuclides, ¹⁴C, ⁹⁰Sr, ²²⁸Th and ²³²Th in its ivory. These nuclides are supposed to have variing concentrations at different parts of a tusk. The reason is the procedure of growth which takes place at the butt-site of a tusk. Therefore the site of sampling could have a big influence on the assessed date of death. However, to find out if the position of sampling is important a complete tusk was analyzed regarding the distribution of these nuclides. Results show that the concentration activity of ¹⁴C and ²²⁸Th varies in different parts of a tusk. The activity concentration of ⁹⁰Sr is very similar in all analyzed parts. The conclusion is that sampling at the butt of a tusk is recommended for age assessment.     

Thermodynamics of ion exchange of H<Superscript>+</Superscript> by Na<Superscript>+</Superscript> or NH<Subscript>4</Subscript><Superscript>+</Superscript> on ion-exchange resins based on <Emphasis Type="Italic">C</Emphasis>-tetramethylcalix[4]resorcinarene

The corrected selectivity coefficients of the ion exchange H⁺-Na⁺ and H⁺-NH₄ ⁺ on ion-exchange resins based on C-tetramethylcalix[4]resorcinarene were calculated from the experimental data obtained from studying ion-exchange equilibria. The preference of the ion-exchange resins for cations increases in the sequence: Na⁺ < NH₄ ⁺ < < H⁺, and the ion-exchange resin based on (2-furyl)hydroxymethyltetramethylcalix[4]resorcinarene has a higher preference for ammonium cations. According to the results of microcalorimetric measurements, the exchange H⁺-Na⁺ on this ion-exchange resin is accompanied by the highest change in the differential enthalpy. It follows from the quantum-chemical calculations that the introduction of a (2-furyl)hydroxymethyl group into the structure of the polymer induces additional electrostatic interactions between an ammonium cation and an elementary unit of the ion-exchange resin.__________Published in Russian in Izvestiya Akademii Nauk. Seriya Khimicheskaya, No. 12, pp. 2560–2563, December, 2004.     

Separation of <Superscript>142</Superscript>La from fission products by SISAK technique

A rapid separation system based on SISAK technique was established to isolate ¹⁴²La successfully from fission products. SISAK technique is often applied in the separation of nuclides with the half-life of seconds or minutes. Here it was used to separate the parent nuclide of ¹⁴²La, which the half-life is in the magnitude of several seconds. According to the separation procedure designed in the paper, the activity of ¹⁴²La acquired is more than 10⁴ Bq and the decontamination factors for most γ-emitters are higher than 10³.     

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.     

Determination of14C in spent moderator ion-exchange resin from Bruce Nuclear Generating Station A

Spent ion-exchange resins are produced in the purification of coolant and moderator systems during the normal operation of CANDU (Canada deuterium uranium) nuclear reactors. Carbon-14 is a radionuclide of concern in disposal of ion-exchange resins because of its relatively long half-life, its potential high mobility and its ability to be easily incorporated into organisms. Only limited data are presently available on the¹⁴C concentrations of spent from CANDU reactors. To establish a more comprehensive datahase for this radionuclide, concentrations of¹⁴C were determined for two moderator resins from Bruce Nuclear Generating Station A. Mixed bed resins were separated into anion and cation fractions using a sugar solution, and the¹⁴C concentrations were determined for each fraction. Carbon-14-was located predominantly on the anion beads. Samples of anion resin were found to undergo an 81% loss in the¹⁴C concentration over a period of 160 d following the sugar separation procedure. Some evidence is given to suggest this loss in¹⁴C may result from microbial activity. Concentrations and distributions of other predominant radionuclides, such as⁶⁰Co and¹⁵³Gd, are discussed as well.     

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.     

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.     

Synthesis of <Superscript>14</Superscript>C analogue of 1,2-diaryl-[2-<Superscript>14</Superscript>C]-pyrroles

Three 1,2-diaryl pyrroles selective COX-2 inhibitors, 2-(4-fluorophenyl)-5-methyl-1-(4-methylsulfonyl-phenyl)-1H pyrrole, 2-(4-fluorophenyl)-1-[4-(methylsulfonyl) phenyl]-1H-pyrrole and 4-[2-(4-fluorophenyl)-1H-pyrrol-1-yl]benzenesulfon-amide, all three labeled with ¹⁴C in the 2-position were prepared from para-fluoro-benzaldehyde-[carbonyl-¹⁴C].     

Dietary intakes of eighteen elements and <Superscript>40</Superscript>K in eighteen food categories by Japanese subjects

Summary Dietary intakes of eighteen elements and ⁴⁰K were estimated by Japanese subjects using a market basket study. High concentrations of most nuclides were found in 4 categories among 18 categories (nuts and seeds, bean products, seaweeds, and fishes and shellfishes). The main contributors were rice, bean products, and fishes and shellfishes. Daily intakes were estimated (in mg) as follows: sodium 3.91 ^. 10³; potassium: 2.49 ^. 10³; phosphorus: 1.09 ^. 10³; calcium: 551; magnesium: 273; iron: 9.82; zinc: 9.41; manganese: 3.54; strontium: 2.52; rubidium: 2.34; copper: 1.61; barium: 0.543; chromium: 0.283; nickel: 0.172; lithium: 0.060; cadmium: 0.022; cesium: 0.0091; cobalt: 0.0095; and ⁴⁰K: 89 Bq.     

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.     

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.     

Direct low-energy measurement of <Superscript>37</Superscript>Ar and <Superscript>127</Superscript>Xe in a radiotracer gas using low-background proportional counters

A radiotracer gas with a blend of ³⁷Ar and ¹²⁷Xe was created for a gas migration experiment and was characterized at Pacific Northwest National Laboratory using ultra-low-background proportional counters. This paper describes the direct low-energy measurement of ³⁷Ar and ¹²⁷Xe in a dual-isotope sample. Using this low-energy technique, the dual-isotope radiotracer gas was determined to have activity concentrations of 483 Bq/cc and 1435 Bq/cc for ³⁷Ar and ¹²⁷Xe, respectively, and a ratio of 1:3 on the reference date of 7/11/2016.     

Comparison of analytical methods used to determine <Superscript>235</Superscript>U, <Superscript>238</Superscript>U and <Superscript>210</Superscript>Pb from sediment samples by alpha,beta and gamma spectrometry

An intercomparison of the methodology (alpha, beta and gamma spectrometry) used for ²³⁸U, ²³⁵U and ²¹⁰Pb determination was carried out based on 38 sediment samples. The activity range of the samples varied from 10–700 Bq/kg for ²¹⁰Pb, 1–35 Bq/kg for ²³⁵U and 10–800 Bq/kg for ²³⁸U. Results obtained using the three methods were not statistically different at high activity levels, but agreement between the results decreased at lower sample activity levels. For ²¹⁰Pb, the smallest difference was found between alpha and gamma spectrometry. A good correlation between results from alpha and gamma spectrometry was observed over the whole activity range. In beta spectrometry, the results were slightly higher than those obtained by alpha or gamma spectrometry due to the impurity of ²²⁸Ra. In ²³⁸U analysis, good correspondence was observed between ²³⁸U determined by gamma and alpha spectrometry, particularly at higher ²³⁸U activity concentrations over 100 Bq/kg. In ²³⁵U analysis, attention needs to be paid to interference from ²²⁶Ra and its reduction.     

Isotopic fractionation during pretreatment for accelerator mass spetrometer measurement of (D<Subscript>3</Subscript>C)<Subscript>2</Subscript>O containing <Superscript>14</Superscript>C produced by nuclear reaction

A (D₃C)₂O (d₆-acetone) target was irradiated with semi-monoenergetic neutrons generated from ⁹Be(p,n)⁹B reaction with 20 MeV protons to convert ¹³C and oxygen nuclides in the target into ¹⁴C. With both liquid scintillation counting (LSC) and accelerator mass spectrometry (AMS) we measured the (D₃C)₂O (d₆-acetone) liquid targets, which were combustible and easy to afford CO₂ for the AMS measurements. The ¹⁴C yield measured by the LSC method turned out to be 80 times larger than that by the AMS method. This large discrepancy may be attributed to the loss of ¹⁴C atoms during the sample pretreatment in the AMS method such as combustion and cryogenic trapping of CO₂. It means that ¹⁴C newly produced by nuclear reactions can exist in various chemical forms, i.e., C₃D₆O, CO, CO₂, hydrocarbons, etc., and a simple sample pretreatment right after production can cause serious isotopic fractionation. Therefore, using the AMS method, extreme caution in sample pretreatment should be exercised when the ¹⁴C yield produced immediately by nuclear reaction is measured.     

Separation and purification of <Superscript>249</Superscript>Cf and <Superscript>245</Superscript>Cm

There is a need to provide radioactivity standards of the higher actinides in support of both decommissioning and remediation activities as well as routine environmental analysis. In the case ²⁴⁹Cf, this will provide a useful calibration nuclide for both α-and γ-spectrometry as well as improving knowledge of the decay scheme for this nuclide. There is anecdotal evidence to suggest that the chemical yield of americium and curium may differ in radiochemical analysis. Thus, a chemical yield tracer of ²⁴⁵Cm may help to resolve this issue and will be suitable for both, suitable for use as a chemical yield tracer for both α-particle spectrometry and mass spectrometry. An aged source of ²⁴⁹Cf was used as the source material for the separation of these two nuclides by cation-exchange, using 2-hydroxy-2-methyl-propanoic acid at controlled pH as an eluant, ²⁴⁹Cf being eluted before the ²⁴⁵Cm daughter. The purity of both nuclides was measured by γ-ray spectrometry.     

Measurements of <Superscript>210</Superscript>Po and <Superscript>210</Superscript>Pb in total diet samples: Estimate of dietary intakes of <Superscript>210</Superscript>Po and <Superscript>210</Superscript>Pb for Japanese

To estimate the dietary intakes of ²¹⁰Pb and ²¹⁰Po for the Japanese adults and their annual effective doses, ²¹⁰Pb and ²¹⁰Po were measured for 240 daily diet samples collected at two locations of Ishikawa Prefecture in Japan over three years by duplicate portion studies. No appreciable differences in intake rates of ²¹⁰Pb and ²¹⁰Po and their ²¹⁰Po/²¹⁰Pb ratios were seen among the years in each district, and between the two districts. The intake rates evaluated using 240 diet samples were 0.20 Bq/d/p for ²¹⁰Pb and 0.61 Bq/d/p for ²¹⁰Po as a median, respectively. Annual effective doses of ²¹⁰Pb and ²¹⁰Po for Japanese adults were estimated to be 0.050 and 0.053 mSv/y, respectively.     

High-Resolution <Superscript>1</Superscript>H and <Superscript>13</Superscript>C NMR of Glycyrrhizic Acid and Its Esters

Resonances for protons and C atoms in the ¹H and ¹³C NMR spectra of glycyrrhizic acid and its esters were assigned using high-resolution ¹H (600 MHz) and ¹³C (150 MHz) NMR methods. __________ Translated from Khimiya Prirodnykh Soedinenii, No. 4, pp. 347–350, July–August, 2005.