A rapid chemical separation procedure for the determination of burn-up of nuclear fuel

A chemical separation procedure for the determination of burn-up is described. The isolation of uranium, plutonium and neodymium is based on two ion-exchange separations, the first one an anion exchange in hydrochloric acid—nitric acid medium and the second one a cation exchange under high pressure with α-hydroxyisobutyric acid. Several experimental facts like chemical yields, contaminations and cross contaminations are discussed.     

Some problems with determination of alpha-active nuclides during decommissioning of nuclear power plant A-1

The first step in decommissioning of NPP is connected with defueling and successful disposal of liquid radioactive wastes (LRW), both from previous production and all work connected with this step. These require precise knowledge about the concentration of radioactive nuclides presented, especially alpha-emitters in all material fluxes arising in sich work (namely shut-down after some accident). This paper reports, results of the determination of U, Pu, Am and Cm by means of both the commonly used and the newly developed radiochemical procedures in various types of materials, e.g., LRW, aerosol filters, ashes resulted from incineration technology, metal surfaces, decontamination solutions etc.     

Application of isotope dilution method for mass- and alpha-spectrometric determination of burn-up and uranium,plutonium and transplutonium element content in VVER-440 spent fuel

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.     

Determination of isotopic composition and concentration of uranium,plutonium and neodymium by mass-spectrometric isotope dilution in the irradiated fuel of the Czechoslovak atomic power station A-1

Determination of the isotopic composition and concentration of uranium, plutonium and neodymium by mass-spectrometric isotope dilution is described. Isotopes²³³U,²⁴²Pu and¹⁵⁰Nd were used as spikes. Isotopic composition was measured with a Varian-TH 5 mass spectrometer. Optimum amounts loaded onto the filament were 2–5 μg U, ∼0.1 μg Pu and <0.1 μg Nd. The accuracy and reproducibility of the isotopic ratio and concentration measurements were evaluated.     

Evaluation of Ge(Li) Compton suppression spectrometers for non-destructive radiochemical analysis

Ge(Li) Compton suppression spectrometers are evaluated relative to the specific needs of the analytical radiochemist. Their application to the analysis of radionuclides found in neutron activation, fission product, and environmental radionuclide analyses is discussed. This paper is based on work performed under United States Atomic Energy Commission Contract AT(45-1)-1830.     

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.     

High-pressure liquid chromatography of irradiated nuclear fuel: Separation of neodymium for burn-up determination

Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear chemistry. Preseparation of the rare earths and trivalent actinides using a di(2-ethylhexyl)phosphoric acid/kieselguhr column is described.     

Determination of oxidation states of uranium and plutonium in cooling solution of spent fuel element assemblies from NPP A-1

The knowledge of oxidation states of uranium and plutonium is necessary from the point of view of the proposed vitrification technology for the final liquidation of the cooling solution as a high radioactive waste containing alpha activity. The valence states of uranium and plutonium have a major influence on the subsequent leaching rate of these elements from vitrification matrices. Using the various chemical behaviours of these elements, in accordance with their valence states, we made an attempt to establish their oxidation state in the original solution.     

Gamma-spectrometric determination of106Ru,134Cs,137Cs and144Ce in the fuel from the Czechoslovak atomic power station A-1

Determination of¹⁰⁶Ru,¹³⁴Cs,¹³⁷Cs and¹⁴⁴Ce in samples of irradiated fuel from the Czechoslovak atomic power station Al is described. The determination is based on gamma-spectrum analysis. The analysis was performed using a Ge(Li) semiconductor detector; for the determination of the isotope mentioned the lines in the energy interval from 400 to 1300 keV were used. The analyses of both dissolved and non-dissolved samples of the fuel were performed. The results of the determinations and their comparison are given in detail.     

Chemical and radiochemical analysis of corrosion products on BWR fuel surfaces

This paper aimed at the characterization of the metallic composition and surface analysis of fuel rod Chalk River Unidentified Deposit (crud) for a BWR-6 unit at a nuclear power plant. In this study, inductively coupled plasma-atomic emission spectroscopy and gamma spectrometry were carried out to analyze the corrosion product distributions and to determine the elemental compositions along the fuel rod under conditions of hydrogen water chemistry (HWC) switched from normal water chemistry (NWC) as the reactor coolant. Most of the crud consisted of flakes and irregular shapes via scanning electron microscope morphology. The loosely adherent oxide layer was mostly composed of hematite (α-Fe₂O₃) with amorphous iron oxides, as determined using X-ray diffractometer results. The deposited amount of crud was the order of 0.2 mg/cm², suggesting that the fuel surface of this plant under the HWC environment appeared to have lower crud deposition due to low feedwater iron levels. There was also no significant difference in comparison with the NWC condition.     

Impact of dissolution on the uncertainty of spent fuel analysis

One of the objectives of the French Alternative Energies and Atomic Energy Commission in the Marcoule Centre is to accurately quantify the composition of nuclear spent fuel, i.e. to determine the concentration of each isotope with suitable measurement uncertainty. These analysis results are essential for the validation of calculation codes used for the simulation of fuel behaviour in nuclear reactors and for nuclear matter accountancy. The different experimental steps are first the reception of a piece of spent fuel rod at the laboratory of dissolution studies, and then dissolution in a hot cell of a sample of the spent fuel rod received. Several steps are necessary to obtain a complete dissolution. Taking into account these process steps, and not only those of analysis for the evaluation of measurement uncertainties, is new, and is described in this paper. The uncertainty estimation incorporating the process has been developed following the approach proposed by the Guide to the Expression of Uncertainty in Measurement (GUM). The mathematical model of measurement was established by examining the dissolution process step by step. The law of propagation of uncertainty was applied to this model. A point by point examination of each step of the process permitted the identification of all sources of uncertainties considered in this propagation for each input variable. The measurement process presented involves the process and the analysis. The contents of this document show the importance of taking the process into account in order to give a more reliable uncertainty assessment to the result of a concentration or isotope ratio of two isotopes in spent fuel.     

Utilization of Slovak bentonites in deposition of high-level radioactive waste and spent nuclear fuel

The basic strategic aims in the field of managing high-level radioactive waste and liquidation of nuclear power plants are all contained in the Energy policy of the Slovak Republic. Its aim is to resolve the concept of the backside of the nuclear energetics fuel cycle??long-term deposition of high-level radioactive waste and spent nuclear fuel (SNF). The most important form of high-level radioactive waste and SNF long-term deposition is their deposition in deep geological formations created by natural as well as engineering barriers used to isolate the long-lived radionuclides from the biosphere. The basic components of these barriers are clays, of which bentonite is generally referred to as the most suitable clay material. There are a few significant bentonite deposits in the Slovak Republic: Jel?ový potok, Kopernica, Lastovce, Lieskovec, Dolná Ves. The review article summarizes the information on geotechnical properties of Slovak bentonites published up-to-date, which is inevitable to know for the intention of their use. It highlights the advantages and shows drawbacks of five Slovak deposits. It suggests further research direction, to draw a thorough hydraulical, microbial and radiation profile of Slovak bentonites.     

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 radiochemical neutron activation analysis method for the determination of total mercury in coal

A radiochemical neutron activation analysis method has been developed based on pyrolysis followed by double gold amalgamation for the determination of mercury in solid samples. Accurate results were obtained for mercury in six standard reference materials of varying matrices, including coal. Linearity was demonstrated up to mercury concentrations of 10,000 ng/g. The method is capable of yielding precise, reproducible values with a detection limit of 5 ng/g for mercury in coal.     

Improvements on radiochemical method for the determination of Am and Cm in typical waste streams from nuclear power plants

The determination of Am and Cm in typical waste streams from nuclear power plants using anion exchange chromatography has some drawbacks like the contamination by Pu and Po. This improved procedure solves these problems, and it has been applied with success to the analysis of Am and Cm in nuclear waste samples: ion exchange resins, ion exchange resins solidified with cement and evaporator concentrates.     

Simultaneous determination of trace uranium and thorium by radiochemical neutron activation analysis

A method for the simultaneous, radiochemical neutron activation analysis of uranium and thorium at trace levels in biological materials is described, based on a technique known as LICSIR, in which a double neutron irradiation is employed. In the first, long irradiation²³³Pa (27.0 d) is induced by neutron capture on²³²Th and then the sample is cooled for several weeks. A second short irradiation to induce²³⁹U (23.5 m) is followed by a rapid sequential radiochemical separation by solvent extraction of²³⁹U with TBP and²³³Pa with TOPO. Chemical yields of²³⁹U and²³³Pa were measured for each sample aliquot using added²³⁵U and²³¹Pa tracers from the -spectra of the separated fractions. The technique was validated by quality control analyses.     

Separation of U and Pu in spent nuclear fuel sample using anion-exchange-group-introduced porous polymer sheet for ICP-MS determination

Anion-exchange porous sheets were prepared by radiation-induced graft polymerization and subsequent chemical modifications. A diethylamino (DEA) group as an anion-exchange group was introduced into the polymer chain grafted onto a porous sheet. The DEA group-introduced porous sheet was cut into disks 13 mm in diameter and 3 mm in thickness to fit an empty cylindrical cartridge (DEA cartridge). The DEA sheet had a DEA group of 3.4 mol/kg of the DEA-group-containing porous sheet and a linear velocity of 46 m/h at a permeation pressure of 0.1 MPa at 298 K. The adsorption capacity of the DEA cartridge for FeCl₄⁻ as a model ion in equilibrium with 1 g-Fe(III)/L in 10 M HCl was 0.17 mmol-Fe(III)/DEA cartridge. No Pu leakage during the permeation of 5 mL of 10 M HCl-0.1 M HNO₃ containing Pu ionic species through the DEA cartridge was observed irrespective of the permeation rate ranging from 0.3 to 80 mL/min. A solution containing known amounts of ²³³U, ²⁴⁰Pu, and ²⁴¹Am in 10 M HCl-0.1 M HNO₃ was loaded onto the DEA cartridge. U and Pu were retained on the DEA cartridge, while Am was allowed to pass through the DEA cartridge. Subsequently, 7 M HNO₃ and 1 M HCl as eluents were permeated to elute U and Pu from the DEA cartridge, respectively. The decontamination factor of U in a Pu fraction, defined by dividing the activity of U in the feed solution by that of U in the Pu fraction, was 2.7 × 10⁵, which is desirable for the highly accurate ICP-MS determination of Pu for samples containing both U and Pu. The method using the DEA cartridge was validated by measuring isotopic compositions and quantities of U and Pu in a spent nuclear fuel sample by double-focusing magnetic sector ICP-MS.     

81mSe tracer for determination of the chemical yield in radiochemical neutron activation analysis of selenium

A radiotracer method is described for measurement of the chemical yield in radiochemical neutron activation analysis of selenium using the⁷⁵Se (120 d) induced nuclide. It is based on^81mSe (57 min) radioisotopic tracer, prepared immediately before its use in the radiochemical separation procedure, by neutron irradiation of highly enriched⁸⁰Se. The recovery of selenium is calculated from the 103 keV -peak of^81mSe in the separated selenium fraction used for quantitation of⁷⁵Se. The technique is illustrated by results for biological reference materials of good accuracy and reproducibility.     

Thermal analysis of reference repository for RBMK-1500 spent nuclear fuel in crystalline rocks

Total amount of RBMK-1500 type spent nuclear fuel (SNF) in Lithuania is approximately 22 thousands of fuel assemblies. All these assemblies should be stored for 50 years and then disposed of. International consensus prevails that SNF and long-lived high-level radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. Disposed nuclear waste induces a number of coupled thermo-hydro-mechanical processes around the repository. Thermal analysis of a deep geological repository could provide temperature distribution which is required for the repository’s design and for evaluation of thermal integrity of the engineered barriers. One of the most critical parameters for the repository is peak temperature at the outer surface of the canister. This temperature cannot exceed 100 °C; otherwise, unfavourable groundwater chemistry can adversely affect chemical stability of the engineered barriers. Thermal behaviour of the conceptual Lithuanian repository for RBMK-1500 type SNF in crystalline rocks was modelled using numerical codes ANSYS FLUENT and COMPASS. Very similar temperature distributions around the disposal canister were determined using both modelling tools. The modelling results revealed the importance of coupled heat and hydrodynamical processes for peak temperature in the engineered barriers, whereas the impact of mechanical processes evaluation was insignificant. It was also determined that peak temperature at the outer surface of the disposal canister does not exceed the permitted 100 °C.