Development of an online measurement system using an alpha liquid scintillation counter and a glass-based microfluidic solvent extraction device for plutonium analysis.

An online measurement system using an alpha liquid scintillation counter (α-LSC) coupled to microchip solvent extraction has been developed. A flow-through cell of α-LSC has been prepared by packing PTFE tube into glass tube to combine microchip. Two-phase flow in microchannel has been stabilized by using coiled tube. The Pu in organic phase has been mixed with scintillation cocktail by T-junction connectors. The system separates and detects Pu by online with detection limit of 6.5 Bq/mL, generating only µL-level wastes.

Quantitative determination of total cesium in highly active liquid waste by using liquid electrode plasma optical emission spectrometry.

A sensitive analytical method for determination of total cesium (Cs) in highly active liquid waste (HALW) by using modified liquid electrode plasma optical emission spectrometry (LEP-OES) is developed in this study. The instrument is modified to measure radioactive samples in a glove box. The effects of important factors, including pulsed voltage sequence and nitric acid concentration, on the emission of Cs are investigated. The limit of detection (LOD) and limit of quantification (LOQ) are 0.005 mg/L and 0.02 mg/L, respectively. The achieved LOD is one order lower than that of recently developed spectroscopic methods using liquid discharge plasma. The developed method is validated by subjecting a simulated HALW sample to inductively coupled plasma mass spectrometry (ICP-MS). The recoveries obtained from a spike-and-recovery test are 96-102%, implying good accuracy. The method is successfully applied to the quantification of Cs in a real HALW sample at the Tokai reprocessing plant in Japan. Apart from dilution and filtration of the HALW sample, no other pre-treatment process is required. The results agree well with the values obtained using gamma spectrometry. The developed method offers a reliable technique for rapid analysis of total Cs in HALW samples.

Evaluation of plutonium(IV) extraction rate between nitric acid and tri-n-butylphosphate solution using a glass chip microchannel.

Extraction of Pu(IV) with tri-n-butylphosphate is performed using a glass chip microchannel to evaluate the extraction rate. Two-phase flow forms in the microchannel by introducing a solution of Pu(IV) and tri-n-butylphosphate with flow rates above 5 µL/min. The Pu(IV) extraction reaction proceeds at the interface between the two phases. To evaluate the extraction rate, the contact time between the two phases is varied from 0.48 to 4.8 s by changing the confluent length of the microchannel and the flow rate. The Pu concentration of each phase collected from the microchannel is measured with an alpha liquid scintillation counter, and the contact time dependence of Pu(IV) extraction is obtained. An extraction model based on diffusion in the microchannel and the reaction at the interface is proposed and applied to determine the extraction rate. The extraction process is assumed to follow pseudo-first-order kinetics, and the extraction rate constant of Pu(IV) is determined to be 1.5 × 10(-2) cm/s. The investigation demonstrates that a microfluidic device can be a new tool to determine Pu(IV) extraction rates.

Automated gravimetric sample pretreatment using an industrial robot for the high-precision determination of plutonium by isotope dilution mass spectrometry.

A robotized sample-preparation method for the determination of Pu, which is recovered by extraction reprocessing of spent nuclear fuel, by isotope dilution mass spectrometry (IDMS) is described. The automated system uses a six-axis industrial robot, whose motility is very fast, accurate, and flexible, installed in a glove box. The automation of the weighing and dilution steps enables operator-unattended sample pretreatment for the high-precision analysis of Pu in aqueous solutions. Using the developed system, the Pu concentration in a HNO(3) medium was successfully determined using a set of subsequent mass spectrometric measurements. The relative uncertainty in determining the Pu concentration by IDMS using this system was estimated to be less than 0.1% (k = 2), which is equal to that expected of a talented analyst. The operation time required was the same as that for a skilled operator.

Application of laser-induced photoacoustic spectroscopy for determination of plutonium concentration in nuclear waste solutions.

Laser-induced photoacoustic spectroscopy was used in a quantitative analysis of Pu in HNO3 medium. Plutonium was quantitatively oxidized to Pu(VI) using Ce(IV). The photoacoustic measurement of Pu(VI) with maximum absorption at 830.5 nm was subsequently performed to determine the concentration. The photoacoustic signal was linearly proportional to the Pu(VI) ion concentration. The detection limit of Pu(VI) was estimated to be 0.5 microg mL(-1) (3sigma) in 3 M HNO3. By the proposed method, Pu concentration was successfully determined in a nuclear waste solution for use in nuclear materials management.

Spectrophotometric determination of plutonium in highly radioactive liquid waste using an internal standardization technique with neodymium(III).

A simple and rapid spectrophotometric method has been developed for the determination of Pu in highly radioactive liquid waste. This method uses Nd(III) as an internal standard, which enables us to determine the concentration of Pu and to authenticate the whole analytical scheme as well. A Nd(III) standard mixed with a sample solution and Pu was quantitatively oxidized to Pu(VI) with Ce(IV) in a nitric acid medium, having the maximum absorbance at 830 nm. A spectrophotometric measurement of Pu(VI) was subsequently performed to determine the concentration compared with the maximum absorbance of Nd(III) at 795 nm. It was estimated that the relative expanded uncertainty for a real sample is less than 10%. The limit of detection was calculated to be 1.8 mg/L (3 sigma). The proposed method was also validated through comparison experiments with isotope dilution mass spectrometry, and was successfully applied to analysis for nuclear waste management at spent nuclear fuel reprocessing plants.

Preparation, certification and validation of a stable solid spike of uranium and plutonium coated with a cellulose derivative for the measurement of uranium and plutonium content in dissolved nuclear fuel by isotope dilution mass spectrometry.

A stable solid spike for the measurement of uranium and plutonium content in nitric acid solutions of spent nuclear fuel by isotope dilution mass spectrometry has been prepared at the European Commission Institute for Reference Materials and Measurements in Belgium. The spike contains about 50 mg of uranium with a 19.838% (235)U enrichment and 2 mg of plutonium with a 97.766% (239)Pu abundance in each individual ampoule. The dried materials were covered with a thin film of cellulose acetate butyrate as a protective organic stabilizer to resist shocks encountered during transportation and to eliminate flaking-off during long-term storage. It was found that the cellulose acetate butyrate has good characteristics, maintaining a thin film for a long time, but readily dissolving on heating with nitric acid solution. The solid spike containing cellulose acetate butyrate was certified as a reference material with certified quantities: (235)U and (239)Pu amounts and uranium and plutonium amount ratios, and was validated by analyzing spent fuel dissolver solutions of the Tokai reprocessing plant in Japan. This paper describes the preparation, certification and validation of the solid spike coated with a cellulose derivative.