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Abstract  

In order to evaluate radionuclide inventories as an essential item for the permanent disposal of spent fuel storage racks, chemical conditions for a sample pretreatment of a spent fuel storage rack were studied. Especially, the surface microstructure and the radionuclide distributions for the spent fuel storage rack were investigated by using a SEM–EDX and γ-spectrometer for minimizing the matrix effect which could affect a chemical separation process of some β-emitting radionuclides. The samples were pretreated with a mixed solution of 5 M HCl and 2 M HNO3 by an ultrasonic surface leaching method. Some radionuclides in the raw racks showed the radioactivity of 102–103 Bq for about 10 g of sample weight. From the sample pretreatment, it was confirmed that almost all radionuclides in the rack were completely extracted from the rack when the dissolved thickness of the rack became a maximum 15 μm by the ultrasonic surface leaching method. The established pretreatment method was applied for all spent fuel storage rack generated from Korean NPPs to determine the scaling factor. The radioactivities of 60Co and 137Cs radionuclides in the pretreated solutions were in the range of 4.9E−1~1.5E+2 and 1.2E−1~9.0E+0 Bq/g, respectively.

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Abstract  

A system made up by a Zymark robot and a separation automate preteats spent fuel samples and monitors a tri-n-octylphosphine oxide column extraction chromatographic procedure in order to isolate and purify uranium and plutonium present in the samples, prior to the spectrometric measurements. Up to 16 subsamples of spent fuel in dried or solution form are handled simultaneously in a completely unattended mode. The throughput of the robotized analytical procedure has increased by a factor of 3 compared to the earlier manual procedure without loss in the quality of the chemical treatment and of the mass- and -spectrometric measurements.

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Abstract  

The method of differential spectrophotometry with the use of Arsenazo III for uranium determination with masking Zr and Pu by 1,2-diaminecyclohexanetetraacetic acid in acetate buffer and carboxyarsenazo for determination of plutonium without its separation from uranium is applied for analysis of the spent fuel of VVER type reactors.

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Journal of Radioanalytical and Nuclear Chemistry
Authors: D. Pant, G. Chaugule, K. Gupta, P. Kulkarni, P. Gurba, P. Janardan, R. Changrani, P. Dey, P. Pathak, D. Prabhu, A. Kanekar, and V. Manchanda

Abstract  

This paper deals with the optimization of experimental conditions for the estimation of Np in spent fuel dissolver solution using 2-thenoyltrifluoroacetone (HTTA) as extractant. The quantitative extraction of Np from the dissolver solution employing 0.5 M HTTA/xylene was followed by its estimation by Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) after stripping it from the organic phase with 8 M HNO3. The reliability of the method was checked by standard addition technique. The method is precise and accurate yielding Np analytical recovery of 99 ± 1%.

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Abstract  

A method for analyzing the content of237Np in spent fuel has been developed using inherent239Np as a chemical yield monitor. After ion-exchange separations for the dissolved fuel solution, the237Np content in the neptunium fraction was determined from the activity of237Np or of233Pa, which is in radioactive equilibrium with237Np. The chemical yield in the separations was determined both from the content of243Am which is in radioactive equilibrium with239Np before the separations and from the239Np content in the neptunium fraction after the separations by alpha- and gamma-ray spectrometry.

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Abstract  

During the burning-up fuel in a nuclear reactor, changes in the isotopic composition of the fuel as well as in the reactor power generation take place. To explain these effects, destructive radiochemical analyses of the spent fuel were carried out. The results obtained have been used for the calibration of non-destructive burn-up determination and for the evaluation of simple computer codes. The main results of the codes as well as the methods applied for the separation of radionuclides studied are presented.

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Abstract  

A direct simple and fast method was established, to overcome the influence of low and high level impurities on the measurement of 235U/238U isotopic ratio in nuclear spent fuel safeguard by thermal ionization mass spectrometry (TIMS), by using refractory metal oxide. The addition of refractory metal oxides forming solution (RMOFS), in certain proportions alongside with the spent fuel solution on the sample filaments were found to be useful during the analysis of uranium isotopic ratio by TIMS. RMOFS (with oxide melting point exceeding 2,000 °C), and particularly that of magnesium, were found to be very effective in improving the quality of the ion signal of 235U and 238U, when added without the need for prior purification. Solutions of chromium, cerium, thorium, and magnesium were investigated, to select the more convenient one, and it was found that magnesium was very useful to start with. The method was very simple, improve both the accuracy and precision of the collected data, reduce the time required to achieve steady uranium pilot signal, and hence the over all time of the analysis, regardless of the level of impurities present.

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Abstract  

In this report the procedures and the methodology of our versions of alpha- and mass-spectrometric techniques for destructive analysis of VVER spent fuel are discussed. These techniques allow the determination of the content of americium and curium isotopes with relative error 3–5%, that of plutonium isotopes with error ≤1% and of uranium isotopes −0.3–0.4%. They allow one to determine the fuel burn-up using148Nd monitor with relative error not exceeding 2% at confidence level P=0.95. The investigation was directed at the increase of sensitivity of analysis to ensure that the amount of analysed material should be equivalent to ∼1 mg of irradiated uranium at mean burn-up values. These techniques are based on the isotope dilution method.

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Abstract  

N,N-dimethylhydroxylamine (DMHA) is a novel salt-free reducing reagent used in the separation U from Pu and Np in the reprocessing of power spent fuel. This paper reports on the radiolysis of aqueous DMHA solution and its radiolytic liquid organics. Results show that the main organics in irradiated DMHA solution are N-methyl hydroxylamine, formaldehyde and formic acid. The analysis of DMHA and N-methyl hydroxylamine were performed by gas chromatography, and that of formaldehyde was performed by ultraviolet–visible spectrophotometry. The analysis of formic acid was performed by ion chromatography. For 0.1–0.5 mol L−1 DMHA irradiated to 5–25 kGy, the residual DMHA concentration is (0.07–0.47) mol L−1, the degradation rate of DMHA at 25 kGy is 10.1–30.1%. The concentrations of N-methylhydroxylamine, formaldehyde and formic acid are (8.25–19.36) × 10−3, (4.20–36.36) × 10−3 and (1.35–10.9) × 10−4 mol L−1, respectively. The residual DMHA concentration decreases with the increasing dose. The concentrations of N-methylhydroxylamine and formaldehyde increase with the dose and initial DMHA concentration, and that of formic acid increases with the dose, but the relationship between the concentration of formic acid and initial DMHA concentration is not obvious.

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Abstract  

The resistance against radiation of the tertiary pyridine resins synthesized for the treatment of spent nuclear fuels and high level radioactive waste was evaluated. After irradiation at 10 MGy, only approximately 10% or less of the exchange groups were lost in HCl solutions regardless of their concentrations, while 3040% were lost in HNO3. The pyridine resin has shown remarkable resistance against radiation particularly in HCl solution. It has been revealed that the decomposition of pyridine type resins results from the scission of the principal chains. An irradiation study was conducted also on the quaternary ammonium resins. Quatemization ratio was found to be reduced in HNO3 solutions at 10 MGy irradiation.

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