Non-destructive determination of plutonium in nuclear fuel rods and pellets

Analytical and Bioanalytical Chemistry - A non-destructive method and an experimental set-up are described by which the Pu content in UO2/PuO2 mixed oxide (MOX) pellets and in fuel rods,...     

Some possibilities for the plutonium content determination in an irradiated nuclear fuel

This paper describes the results of the methods used for the plutonium content determination in an irradiated nuclear fuel from the first Czechoslovak atomic power station A1. The main attention was paid to the following methods: mass-spectrometric isotope dilution method, radiometric method and correlation dependence method based on the analysis of the burnt fuel. The principle and the accuracy of the individual methods are discussed.     

Non-destructive determination of nuclear fuel burn-up by measuring the gamma radiation of fission products in irradiated uranium oxide

The burn-up of²³⁵U was determined in two uranium oxide samples (0.713 and 89.9%²³⁵U in mixture) irradiated simultaneously with a cobalt monitor, from the amounts of⁹⁵Zr,¹⁰³Ru,¹³⁷Cs,¹⁴⁰Ba and¹⁴⁴Ce obtained by measuring the intensities of the corresponding gamma radiations. The samples were irradiated for 23 days, and the fission products were measured after cooling for 100 days, nondestructively, by means of a Ge(Li) spectrometer. The integrated neutron flux was determined by measuring the produced⁶⁰Co in the cobalt monitor. The burn-up in both samples was determined by measuring the intensity of eight gamma energies (0.5–1.6 MeV). The determined values are in good agreement. The standard deviation of the mean value ( ) is 5%. The atom per cent fission of²³⁵U in both samples, calculated according to , differs by 1%. The measured σ _f for²³⁵U is in good agreement with the data reported in the literature.     

A modified radiometric method for plutonium determination in irradiated fuel

A modified radiometric method for the determination of plutonium in irradiated nuclear fuel is described. The analysis consists of total plutonium alpha-activity, activity ratio²³⁸Pu:(²³⁹Pu+²⁴⁰Pu) and burnup determination. The error of plutonium determination is practically identical with that of a classical radiometric method. The method suggested is suitable for large sample series of the same type of irradiated nuclear fuel.     

A spectrophotometric method for the determination of plutonium in fuel solutions

A spectrophotometric method is described in which microgram amounts of plutonium can be determined in the presence of uranium, thorium, fission products and cladding materials. Plutonium is extracted with TTA in xylene and reextracted into a solution of Arsenazo III. Zirconium is masked by a Fe(III)-EDTA mixture, fluoride ions by Al(III). 2 to 40 μg of plutonium are required for one analysis. The standard deviation is 1.3% at 15 μg plutonium.     

Age determination of plutonium material in nuclear forensics by thermal ionisation mass spectrometry

Age is a key parameter when deducing the history of plutonium material, i.e. the plutonium produced in the nuclear reactors. This is of vital importance, when a smuggled plutonium sample has been seized and the origin has to be determined. A methodology is described which allows accurately to determine the age of plutonium material by thermal ionisation mass spectrometry using independent parent/daughter relations. This has been demonstrated for Reference Materials of known ages as well as for real samples. The already established method using gamma spectrometry is compared to this.     

Age determination of plutonium material in nuclear forensics by thermal ionisation mass spectrometry

Age is a key parameter when deducing the history of plutonium material, i.e. the plutonium produced in the nuclear reactors. This is of vital importance, when a smuggled plutonium sample has been seized and the origin has to be determined. A methodology is described which allows accurately to determine the age of plutonium material by thermal ionisation mass spectrometry using independent parent/daughter relations. This has been demonstrated for Reference Materials of known ages as well as for real samples. The already established method using gamma spectrometry is compared to this.     

Application of neutron well coincidence counting for plutonium determination in mixed oxide fuel fabrication plant

Mixed oxide (MOX) fuel is an alternative to conventional enriched uranium oxide fuel in thermal reactors. Indian interest in plutonium recycle in thermal reactors is primarily due to the need to develop alternative indigenous fuel for two boiling water reactors (BWR) at Tarapur, which are designed to use imported light enriched uranium fuel. A few MOX assemblies have been fabricated and loaded into the reactors. Neutron well coincidence counting (NWCC) system has been successfully employed to check the enrichments of PuO₂ in MOX blends. NWCC has also been successfully applied in developing dry recycling process of clean rejected oxide (CRO) and dirty rejected oxide (DRO).     

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.     

Problems of plutonium and uranium determination

The experience with the determination of plutonium and uranium in the samples of nuclear fuel obtained during a 10-years period of the activity of the Central Control Laboratory of the Nuclear Research Institute at e is evaluated. The paper describes the principles of the methods used and the reproducibilities and accuracies of the results obtained with the aid of the titrimetric methods used for the determination of plutonium and uranium.     

Electromigration method for destructive determination of spent nuclear fuel burnup

In this report we discuss the methods and results of VVER spent fuel burnup determination by¹⁴⁶Nd content and the correlation with accumulation of some transplutonium nuclides. For separation of trivalent rare earths and transplutonium elements the method of paper electrophoresis is used. For the quantitative determination of americum and curium isotopes a modification of α-spectrometric analysis is proposed with the chemical yield control of isolated elements using²⁴⁴Cm. The amount of¹⁴³Nd is determined by the isotopic dilution method combined with mass spectrometry with¹⁴²Nd as a tracer.     

Non-destructive determination of nickel in iron meteorites

Application of a portable radioisotope X-ray fluorescence analyser for the determination of nickel in iron meteorites is described. The error of the XRF method was ±0.5% Ni at the 95% confidence level, which was confirmed by comparison with the results of other authors. In spite of limited accuracy this version of the radioisotope XRF method may find a wide application in meteorite studies, for rapid and non-destructive verification and classification of collected meteorite specimens.     

Comparison of alpha-spectrometry and alpha/gamma ratio method for the determination of americium in plutonium bearing fuel materials

Determination of americium is one of the requirements of chemical quality assurance of plutonium bearing fuel materials. Alpha-spectrometry is generally used for the determination of ²⁴¹Am in Pu bearing fuels since the efficiency of semiconductor detector used for alpha-spectrometry is independent of the alpha-particle energy in the 4 to 8 MeV range. However, this method has limitations for Pu samples containing extremely small or very large amounts of ²⁴¹Am. Thus an alternative methodology based on alpha/gamma (α/γ) activity ratio was developed and tested using different samples. The method is based on the determination of total γ-activity (of 60 keV peak) of an aliquot of the solution and the total α-activity present in the aliquot. The method is fast as it does not involve chemical separation of Pu and Am as required in the alpha-spectrometric method. Data obtained on synthetic and real life samples demonstrates the usefulness of the developed alpha/gamma ratio method for the determination of ²⁴¹Am in Pu bearing fuel samples.     

Luminescent determination of microamounts of plutonium

Determination of microamounts of plutonium isotopes by luminescent radiation in the infra-red region of PbMoO₄ crystallophosphors activated by plutonium is proposed. The determination is possible against the background of subtantial amounts of uranium, iron, rare earths and other elements. The detection limit of plutonium is 10^–9 g in the samples tested. The accuracy is 0.1 for a confidence level of 0.95.     

Voltamperometric determination of uranium and plutonium in alkaline solutions

The simultaneous determination of U(VI), Pu(VI), Pu(V) in 0.5–4.0 M NaOH has been elaborated by means of classical and differential pulse voltamperometry. U(VI) is determined with a dropping mercury electrode (DME) at the half-wave potential of E_1/2=–0.89 V vs. Ag/AgCl reference electrode due to reduction to U(V). The limiting current or peak heights are proportional to uranium(VI) concentration in the range of 1.3.10^–7–3·10^–4 M U(VI). Deviation from proportionality is observed for higher concentrations due to polymerization of uranates. Pu(VI) and Pu(V) are determined with a platinum rotating electrode at E_1/2=–0.02 V due to the reaction Pu(VI)+e^–»Pu(V) and with DME at E_1/2=–1.1 V due to the reduction to Pu(III). The limiting currents of both Pu(VI) and Pu(V) are proportional to their concentrations in the range of 4·10^–6–1.2·10^–3 M Pu. The determination of U(VI), Pu(VI), Pu(V) is not interfered by the presence of the following salts: 2M NaNO₃, 2M NaNO₂, 1.5M NaAlO₂, 0.5M NaF and ions of Mo(VI), W(VI), V(V), Cu(II). The presence of CrO ₄ ^2– and FeO ₂ ^– ions disturbs the determination of U(VI) in 1–4M NaOH, however, contribution of the reaction Fe(III)+e^–»Fe(II) to uranium reduction peak can be calculated from the height of the second peak Fe(II)+2 e^–»Fe(0).     

Luminescent determination of neptunium and plutonium in the environment

The preparation of^99mTc-complexes with nine derivatives of (4R)-1,3-thiazolidine-4-carboxylic acid over a broad pH range is described. The labeling efficiency of all compounds was checked by paper- and thin layer chromatography. Paper electrophoresis and gel chromatography indicated the presence of two types of^99mTc-complexes differing in charge and molecular size. Preliminary results of the biodistribution of anionic complexes in mice are given. The structure of both types of the Tc-complexes is also discussed.     

Redox determination of uranium in presence of plutonium in low phosphoric acid medium and the recovery of plutonium

Quantitative determination of uranium in (U, Pu)O₂ fuels is usually done by the DAVIES-GRAY method. High concentrations of phosphoric acid in the analytical waste generated by this method make the revocery of plutonium rather complex. Studies on the recovery of plutonium from nitric acid medium containing different concentrations of H₃PO₄ by conventional anion-exchange procedure reveal that more than 90% of the plutonium can be easily recovered when the phosphoric acid concentration is less than 0.5 M in the solution. A method was developed for the determination of uranium in the presence of plutonium, which involves the reduction of U(VI) to U(IV) by Fe(II) in a medium of 3.5M H₃PO₄ +4.5M H₂SO₄ instead of 10–11M H₃PO₄ so as to have the H₃PO₄ concentration 0.6M in the waste. A number of determinations of uranium in UO₂(NO₃)₂ working standard solutions and (U, Pu) synthetic solutions with uranium at the 3–7 mg level were carried out by this method. The precision obtained was better than ±0.2% and the accuracy was also within the precision limits. The resulting analytical waste generated was directly subjected to anion exchange separation for the recovery of plutonium which was found to be more than 90%.