Results of measuring the content of actinide, neodymium, and cesium isotopes and the burn-up in a sample of high-burn-up WWER-1000 fuel by destructive methods

Experimental data on the content of U, Pu, Np, Am, Cm, Nd, and Cs isotopes in a sample of WWER-1000 (1000 MW water-cooled water-moderated energy reactor) spent nuclear fuel with the burn-up of up to 70 GW day (t U)^?1 are reported. The data were mainly obtained by isotope dilution followed by mass-spectrometric determination. The burn-up was determined from the weight fractions of ¹⁴⁸Nd and ^134,137Cs in U. The data obtained can be used for the development of low-waste cost-saving technolofy for reprocessing highburn-up SNF and for the improvement of calculation methods and programs intended for calculation of the burn-up and nuclide composition of SNF from WWER reactors.     

Destructive analysis of the nuclide composition of spent fuel of WWER-440, WWER-1000, and RBMK-1000 reactors

The results of analysis of spent nuclear fuel (SNF) by destructive methods, carried out systematically at the Khlopin Radium Institute for over 20 years till the mid-1990s, are presented. These data constitute the experimental base for the development of nondestructive methods, correction of calculation programs, substantiation of correlation techniques of determination of individual components of SNF, and for some other purposes. The isotope compositions of U, Pu, Am, and Cm and the fuel burn-up values are presented for 81 SNF samples from WWER-440, WWER-1000, and RBMK-1000 reactors. The burn-up values are determined with ¹⁴⁸Nd, ^{145 + 146}Nd, and ¹³⁷Cs monitors. The utilized methods, including ion-exchange and distribution chromatography, electromigration, and coprecipitation, as well as α-ray spectrometry, luminescence analysis, and mass spectrometry, are briefly discussed. The principal method utilized is isotope dilution α-ray spectrometry or isotope dilution mass spectrometry. A number of isotopes certified as reference samples of different categories, prepared at the Radium Institute, served as spikes. A combination of these methods allows the isotope composition to be estimated accurately to within ≤0.15% for U, ≤0.5% for Pu, and 3–5% for Am and Cm. The accumulated data set for the SNF from WWER and RBMK reactors is presented.     

Radiochemical analysis of a sample of mixed uranium–plutonium nitride fuel

Radiochemical analysis of a mixed uranium–plutonium nitride (MUPN) fuel sample irradiated in a BOR-60 reactor was performed. The study was made using the set of procedures developed at the Research Institute of Nuclear Reactors for determining the nuclide composition and gravimetric content of U, Pu, Am, Cm, Nd and other fission products, platinum group metals, and transition metals, including nuclear physical methods, atomic emission spectrum analysis, mass spectrometry for determining the nuclide composition, and isotope dilution mass spectrometry for determining the gravimetric content of nuclides with preliminary radio-chemical separation of fractions of elements by ion-exchange, extraction-chromatographic, precipitation, and distillation methods. The MUPN fuel burn-up was determined from the ratio of the number of atoms of the fission product selected as a burn-up monitor to the number of heavy atoms in the dissolved fuel sample (method of fission product accumulation, MFP). The ¹⁴⁵Nd + ¹⁴⁶Nd sum and ¹⁴⁸Nd were used as burn-up monitors.     

Radionuclide composition of fragments of lava-like fuel-containing materials from the CNPP fourth unit

Samples of fuel-containing materials taken inside the CNPP Fourth Unit were analyzed by γ- and α-ray spectrometry. The isotope ratios for Cs, Eu, Pu, Am, and Cm were measured, and the fuel burn-up in the samples was determined. Inconsistencies in theoretical estimations on the production of all the radionuclides over ²⁴¹Am were revealed. The burn-up values determined from data on the Cs isotopes systematically differ from those determined from data on the other radionuclides. The causes of these facts are discussed.     

РH process as a technology for reprocessing mixed uranium–plutonium fuel from BREST-OD-300 reactor

A technology for reprocessing mixed uranium–plutonium nitride fiel (MUPN) from BREST reactor is considered. The technology should ensure reprocessing of spent nuclear fuel with the storage time after irradiation of no more than 1 year, 10–15% content of fissile materials (FM), and burn-up of 10% of heavy atoms. The target product of the technology is a mixture of actinide oxides separated from fission products with the separation factor of ~10⁶. A PH (Pyro–Hydro) process was suggested for MUPN SNF reprocessing. It involves pyroelectrochemical fuel reprocessing with separation of U, Np, and Pu from the major fraction of fission products responsible for the heat release from the fuel and for the radiation load on process media, a series of hydrometallurgical operations for final purification of the target products (U–Pu–Np–Am), and radioactive waste (RW) management. The PH process is being developed since 2011 by the teams from the Bochvar High-Tech Research Institute of Inorganic Materials, Khlopin Radium Institute, and Research Institute of Atomic Reactors with active participation of the Leading Research Institute of Chemical Technology, Siberian Chemical Combine, and institutes of the Russian Academy of Sciences, primarily Frumkin Institute of Physical Chemistry and Electrochemistry and Vernadsky Institute of Geochemistry and Analytical Chemistry.     

Determination of the weight and composition of the precipitates formed in the course of dissolution of irradiated WWER oxide fuel

The amount and composition of insoluble precipitates formed in the course of dissolution of spent fuel samples with the burn-up from 15 to 54 MW day (kg U)⁻¹ were examined. The weight of the insoluble precipitates was from 0.03 to 0.44% of the fuel weight. The major elements determining the composition of the precipitates were platinum group metals (Pd, Rh, Ru), Zr, and Mo. The specific β- and α-activity of the precipitates obtained was 30–840 and 0.01–8 Ci kg⁻¹, respectively. The major factor determining the concentration of α-emitting nuclides is the fuel burn-up. Depending on the dissolution conditions, the U content was 0.2–4, and the Pu content, 0.1–3%. The weight of secondary precipitates was from 0.005 to 0.3% of the irradiated fuel weight, or 11–50% of the total weight of the precipitates obtained in the experiments. The specific β-activity of the secondary precipitates obtained varies from 5 to ∼300 Ci kg⁻¹ and is determined by the same radionuclides as in the primary precipitates. The α-activity of the secondary precipitates increases with the burn-up and amounts to 0.1–30 Ci kg⁻¹. The values obtained vary only slightly depending on the dissolution conditions and on the time of solution keeping before control filtration.     

Treatment of water from spent nuclear fuel storage basins with ion-exchange resins modified with transition metal hexacyanoferrates

Ion-exchange resins modified with hexacyanoferrates were suggested as a material for charging filters for the treatment of water from spent nuclear fuel (SNF) storage basins. Modification of ion-exchange resins in the filter increases the degree of removal of ¹³⁷Cs radionuclides by a factor of more than 10, with high performance with respect to strontium radionuclides preserved.     

Radiography and local microanalysis for detection and examination of actinide-containing microparticles

A procedure was proposed for examination of the spatial microdistribution of fissile actinide isotopes (²³⁵U, ²³⁹Pu) in various environmental objects (soils, bottom sediments, aerosols, colloid material). Complex analysis of natural actinide-containing microparticles involves α-track radiography (ATR) and neutron-induced fission track radiography (NITR), highly sensitive nondestructive methods for determining the spatial microdistribution of α-emitting and fissile actinides, respectively, and scanning electron microscopy with energy-dispersive spectrometry (SEM-EDS) and secondary ion mass spectrometry (SIMS) for local microanalysis of the general elemental and isotope composition of the particles and determination of the morphology and size of the particles. The actinide-containing particles were localized using an SEM microgrid with electrodeposited ²³⁹Pu. Particles characterized by a high concentration of actinides were detected by SEM and SIMS, while those with a low content of actinides, by track radiography followed by fragmentation of the sample for subsequent analysis (γ-ray spectrometry etc.).     

Determination of concentrations of fission products by ICP–AES in solutions from spent nuclear fuel reprocessing

A procedure was developed for determining the gravimetric concentration of U and fission products (Ba, Mo, Zr, Sr, Rb, Pd, and Cd) in solutions from spent nuclear fuel (SNF) reprocessing, using inductively coupled plasma atomic emission spectrometry (ICP–AES). The optimum analysis parameters were found. The analytical lines that are not significantly affected by the matrix element were chosen for the elements being determined. The total margin of error in measurements of the metal concentrations (0.1–2500 mg L^–1) by the suggested procedure is in the interval 1–20% (for the minimal concentration). The detection thresholds of uranium fission products in model solutions at optimum parameters vary in the interval from 0.020 to 0.1 mg L^–1.     

Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of ²³⁵U, ²³⁸U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.     

Radionuclides in ground waters from observation holes in the Shelter local area

The volume activity of ³H, ⁹⁰Sr, ¹³⁷Cs, ²³⁴U, ²³⁵U, ²³⁸U, ²³⁸Pu, ^239+240Pu, and ²⁴¹Am in ground waters from observation holes 1-G-6-G in the north section of the Shelter local area of the Chernobyl Nuclear Power Plant (CNPP) was measured. The distribution of radionuclides in the suspension fractions of the ground waters was evaluated. The main contribution to the pollution of ground waters with uranium is due to natural uranium isotopes: ^234,235,238U. The activity ratios of ²³⁸Pu, ^239+240Pu, and ²⁴¹Am in ground waters are similar to those in the spent fuel of 4th CNPP block.     

Dissolution of WWER-1000 spent nuclear fuel in a weakly acidic solution of iron nitrate and recovery of actinides and rare earth elements with TBP solutions

It is demonstrated on real solutions of samples of spent nuclear fuel (SNF) from WWER-1000 reactors (1000-MW_el water-cooled water-moderated energy reactors) that weakly acidic solutions of iron(III) nitrate at the molar ratio Fe(III): U ≥ 2.0 dissolve SNF with quantitative transfer of U and Pu into the solution. In the process, Fe partially precipitates in the form of a basic salt precipitate together with a part of the fission products (>90% of Ru, ~90% of Мо, >60% of Tc, and 40% of Zr) already in the step of the fuel dissolution. Cs, Eu, and Am pass into the solution together with U and Pu. With the required conditions followed, U and Pu can be separated from the solution by precipitation of their peroxides or quantitatively extracted from this solution with 30% TBP in Isopar L. The presence of ≥1 M Fe(NO₃)₃ in the solution considerably increases the distribution ratios of TPE and REE, which allows their recovery from a weakly acidic nitrate solution to be also performed with 30% TBP in a diluent. This process can serve in the future as a basis for the development of a new integrated technology combining the PUREX process with TPE partitioning using a common extractant.     

Recycling of Transmutation Targets in the Presence of Impurity Isotopes of Starting Nuclides

Relations were obtained for calculating the concentrations of impurity isotopes of starting nuclides ⁹⁸Tc, ¹²⁷I, ^133'134'137Cs in recycling ⁹⁹Tc, ¹²⁹I, ¹³⁵Cs targets, respectively. Conditions of removing ⁹⁸Tc from the transmutation cycle in a single irradiation run were found. The content of ¹²⁷I in the targetattains an equilibrium value, thus making unnecessary isotopic separation of iodine. For isotopic separation of cesium, a decrease of the initial content of ¹³⁷Cs in the ¹³⁵Cs target to 1% is sufficient. The ¹³³Cs and ¹³⁴Cs nuclides preserve steady-state concentrations for a long period, and for this reason the activity grows insignificantly, depending on the ¹³⁷Cs accumulation, which makes unnecessary isotopic separation of cesium from irradiated targets.     

The use of poly(acrylamide-acrylic acid)-montmorillonite composite for preparing low-activity 60Co, 65Zn, 134,137Cs mixed sealed source for radiometric calibration purposes

Poly(acrylamide-acrylic acid)-montmorillonite composite [P(AAm-MA)-M] was synthesized and used for preparing ⁶⁰Co, ⁶⁵Zn, ^134,137Cs mixed sealed source. The synthesis procedure was based on ??-induced polymerization of polyacrylamide (PAAm) in the presence of acrylic acid (AA), clay minerals (montmorillonite, M), and N,N??-methylenebisacrylamide (NMBA) as a cross-linker. The adsorbent functionality was assayed using FTIR spectroscopy and thermal analysis. The distribution coefficients of ⁶⁰Co(II), ⁶⁵Zn(II), and ^134,137Cs(I) ions on P(AAm-MA)-M were determined in relation to HCl concentration. P(AAm-MA)-M was loaded with ⁶⁰Co, ⁶⁵Zn, and ^134,137Cs by equilibrating it (0.5 g) with a mixed solution of ⁶⁰Co, ⁶⁵Zn, and ^134,137Cs at pH 3 (HCl) for 24 h at 25 ± 1°C. Mixed sealed source containing 5.3, 0.03, 0.05, and 1.05 ??Ci of ⁶⁰Co, ⁶⁵Zn, ¹³⁴Cs, and ¹³⁷Cs, respectively, was prepared by packing 100 mg of the loaded matrix in the cylindrical cavity of a Chinese Artelone capsule. The sealed source was subjected to quality control tests.     

Recovery of Pd from spent fuel: 5. Recovery of Pd from high-level PUREX raffinate

Laboratory experiments were carried out on sorption and electrochemical recovery of Pd from actual high-level PUREX raffinate. The initial solutions were high-level liquid wastes after extraction processing of WWER-1000 spent fuel with a burn-up of about 40 MW (kg U)^?1, cooled for 5–7 years. In sorption recovery with VP-1AP anion-exchange resin, the Pd yield into the eluent was about 90%. After additional refinement by precipitation, the activity of the resulting palladium black sample was no more than 0.8 mCi (g Pd)^?1, and the total decontamination factor was 1.9 × 10⁴.     

A Method for Determining the Isotope Composition of U, Pu, Am, and Cm in "Hot" Particles and Spent Nuclear Fuel

A method was proposed for determining the content of ^234-238U, ^238-242Pu, ^241-243Am, and ^242-244Cm in "hot" fuel particles and spent nuclear fuel. The method is based on high-precision measurement of the -activity in the sample and calculation of the relative contributions of individual nuclides or radionuclide groups to the total activity. Partitioning of U, Pu, Am, and Cm was carried out by ion-exchange chromatography. The contents of ²³⁴U, ²³⁶U, ²³⁸U, ²³⁸Pu, ^239+240Pu, ²⁴²Pu, ²⁴¹Am, ^242mAm, ²⁴³Am, ²⁴²Cm, and ²⁴⁴Cm in "hot" particles sampled in the Chernobyl area were reported. The applicability of the method proposed to determining the radionuclide composition of spent nuclear fuel was discussed.     

Extraction of long-lived radionuclides from nitric acid solutions using an extractant based on dibutyl hydrogen phosphate and chlorinated cobalt dicarbollide

Chlorinated cobalt dicarbollide (CCD), when added to concentrations of 0.25 M to a solution of dibutyl hydrogen phosphate (HDBP) in m-nitrobenzotrifluoride (MNBTF), increases the distribution ratios of trace amounts of Eu and Am without changing the slope (tan α ～ 2) of their dependences on HDBP concentration in the 0–1.5 M range. At [CCD]/([CCD] + DBPA]) = 0.2–0.22, the synergistic effect is observed in the entire range of HDBP concentrations in extraction of these elements from 1.0 and 2.5 M HNO₃. In this case, HDBP suppresses the extraction of Cs with CCD in the area below the synergistic maximum, where antagonism is observed in the extraction of Cs. Polyethylene glycol (PEG, Slovafol-909) was added to the extraction mixture to improve the extraction of Sr. The extremum is attained at its concentration in the solvent with HDBP of ～0.033–0.065 M, which is smaller than that in the absence of HDBP by a factor of 1.5–2.5. With increasing concentration of HDBP in the HDBP-CCD-PEG-MNBTF extraction system, the slopes for Eu and Am are 1.3 and 0.6, whereas the slopes for Cs and Sr decrease nonlinearly and amount to ?1.8 and ?1.3, respectively. With increasing concentration of HNO₃, D for Eu, Am, and Cm decreases in proportion to the HNO₃ concentration to the power of ?3 irrespective of the PEG concentration, and for Cs and Sr, to the power of ?2 in the presence of PEG, whereas in the PEG-free systems the dependences are nonlinear. The synergistic extractant is characterized by higher (by an order of magnitude) solubility of metal solvates as compared to the HDBP-MNBTF system (concentration of Eu in the extractant >0.163 M). The extractant containing HDBP (1.1 M), CCD (0.23 M), and Slovafol-90 (0.065 M) in MNBTF is suggested for combined recovery of rareearth (REE) and transplutonium elements (TPE) and of Cs and Sr from high-level waste (HLW) after reprocessing of spent nuclear fuel (SNF) with high burn-up.     

Behavior of fission products in the course of dissolution of simulated spent nuclear fuel in iron nitrate solutions and of recovery of uranium and plutonium from the resulting solutions

Experiments aimed to examine the spent nuclear fuel dissolution in iron(III) nitrate solutions and to elucidate the behavior of fission products in the process were performed with simulated fuel corresponding to spent nuclear fuel of a WWER-1000 reactor. In Fe(III) nitrate solutions, U is quantitatively transferred from the fuel together with Cs, Sr, Ba, Y, La, and Ce, whereas Mo, Tc, and Ru remain in the insoluble precipitate and do not pass into the solution, and Nd, Zr, and Pd pass into the solution to approximately 50%. The recovery of U or jointly U + Pu from the solution after the dissolution of oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Ionic Composition of Hot Particles Containing U,Cs, and Sr,Evaluated from the Parameters of X-ray Photoelectron Spectra of Outer U5f and Inner U4f,Cs3d, 4d,and Sr 3d Electrons

The elemental and ionic composition of pellets produced from reactor fuel (UO₂) containing 0.1 wt % Cs and 0.5 wt % Sr relative to U and also of hot particles generated by heating of the fuel to 2000°C and then subjected to additional heating to 900°C in air or argon and condensed on aluminum support was analyzed by the X-ray photoelectron spectroscopy. It was shown that within the first 20 s of heating U and Cs sublime predominantly. In the subsequent 300 s of heating U, Cs, and Sr sublime. For example, it was found that hot particles collected in the first 20 s of heating and subjected to additional heating at 900°C in air flow contain 68% U and 32% Cs, whereas particles collected in the subsequent 360 s and subjected to the same additional heating contain 51% U, 13% Cs, and 36% Sr. It is assumed that these hot particles incorporate uranyl compounds of the following types: UO₂CO₃, Cs₂UO₄, Cs₄UO₂(CO₃)₃, CsUO₂(OH)₃, SrUO₄, Sr₃UO₆, and SrUO₂CO₃(OH)₂. Treatment of the surface of hot particles with Ar⁺ ions produces changes in their composition.     

Thermal Expansion of Simulated Fuels with Dissolved Fission Products in a UO2 Matrix

As a part of the DUPIC (direct use of spent PWR fuel in CANDU reactors) fuel development program, the thermal expansion of simulated spent fuel pellets with dissolved fission products has been studied by using a thermo-mechanical analyzer (TMA) in the temperature range from 298 K to 1773 K to investigate the effects of fission products forming solid solutions in a UO₂ matrix on the thermal expansions. Simulated fuels with an equivalent burn-up of (30 to 120) GWd/tU were used in this study. The linear thermal expansions of the simulated fuel pellets were higher than that of UO₂, and the difference between these fuel pellets and UO₂ increased monotonically with temperature. For the temperature range from 298 K to 1773 K, the values of the average linear thermal expansion coefficients for UO₂ and simulated fuels with an equivalent burn-up of (30, 60, and 120) GWd/tU are 1.19 × 10⁻⁵ K⁻¹, 1.22 × 10⁻⁵ K⁻¹, 1.26 × 10⁻⁵ K⁻¹, and 1.32 × 10⁻⁵ K⁻¹, respectively.