Estimation of neptunium in a fuel reprocessing plant

The solutions at various stages of the purex process used at Trombay have been analysed for their²³⁷Np content to know the path of neptunium in the process. The results of these analyses, though not very quantitative, revealed that in the co-decontamination cycle, neptunium is extracted along with uranium and plutonium almost quantitatively while in the partitioning cycle the extraction is reduced to about 50%. Analysis of various oxidation states of neptunium in the feed of the first cycle showed that neptunium is predominantly present as Np(V).     

Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.     

Effects of the shutdown of a nuclear fuel reprocessing plant on nearby streams

In a study conducted in 1971, levels of tritium were found in Cattaraugus-Creek, a stream in Western New York State. This material was attributed to the operation in West Valley, New York of the world's first commercial nuclear fuels reprocessing plant. Several fission fragment isotopes in addition to tritium were also observed in Buttermilk Creek, one of the tributaries of Cattaraugus Creek that runs through the reprocessing plant grounds. The plant ceased processing nuclear fuel in December 1971, and a new set of measurements in these streams were made to assess the effect of the ending of the plant's operation. Substantially lower concentrations of tritium and no fission produced isotopes have been observed.     

Determination of long-lived radionuclides in environmental media around a nuclear fuel reprocessing plant

Methods for the analysis of¹²⁹I and²⁴¹Pu are described briefly. Neutron activation is necessary to achieve an adequate degree of sensitivity for the measurement of¹²⁹I, but otherwise all laboratory manipulations are straightforward and use commonly-found, well-tried techniques. With these methods, both radionuclides can be measured easily in the terrestrial environment around a nuclear fuel reprocessing plant;²⁴¹Pu is measureable elsewhere in integrating media such as undisturbed soil.     

Behavior of iodododecane in nuclear fuel reprocessing

By simulation experiments with a 10^–5 mol/l solution of iodododecane labeled with¹³¹I in n-dodecane the influence of various materials and conditions, which are possible in nuclear fuel reprocessing, has been investigated. The formation of decomposition products was detected via HPLC with a radioactivity monitor. By means of²⁵²Cf plasma-desorption mass spectrometry (PDMS) the decomposition products were identified. It was found that a temperature of 100°C favored the formation of iodoalkanes with chain lengths of C₁ to C₁₁. The presence of TBP(tri-n-butyl-phosphate) accelerated the decomposition of iodododecane. In pure TBP only iodobutane was formed as a decomposition product.     

Recovery and purification of plutonium in lean plutonium product stream obtained during reprocessing of fast reactor spent fuels

A method has been developed for final purification of plutonium from uranium and fission products of high beta gamma activity. This method involves selection of a suitable ion exchange resin for the purification of plutonium in order to deliver a quality PuO₂ product. The effect of the concentration of uranium and plutonium, effect of increased loading of uranium and number of bed volumes for effective washing, which are some of the parameters that generally affect the recovery and purification of plutonium were investigated. An excellent decontamination factor for fission products has been achieved by this anion exchange process which in turn delivered an excellent PuO₂ product quality in terms of purity and associated beta gamma activity with low personnel radiation exposure.     

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).     

Dry deposition velocity of iodine-129 onto grass as obtained by field measurements in the environment of a nuclear fuel reprocessing plant

The deposition velocity of gaseous organic¹²⁹I species from the exhaust air stack of the Karlsruhe reprocessing plant onto pasture grass was measured by a field experiment. By simultaneously measuring the amount of¹²⁹I deposited per unit area of pasture grass and the time integrated mean air concentration of¹²⁹I a deposition velocity of V_g=5.8×10^–1 /cm s^–1/ onto pasture grass was determined.     

Identification of iodoorganic compounds in kerosene from nuclear fuel reprocessing

In kerosene samples from nuclear fuel reprocessing, iodoalkanes with chain-lengths from C₄ to C₁₃ have been identified. The kerosene samples were purified by means of solid-phase extraction. By this method other fission products like¹²⁵Sb and¹⁰⁶Ru were quantitatively removed from the solution. The only remaining radioactive nuclide was thus¹²⁹I. The iodoorganic compounds in the kerosene from the solvent were enriched from 6000 Bq/L to 100 000 Bq/L¹²⁹I by vacuum distillation. Chromatographic separation by HPLC, fractionation, and -measurement of the fractions showed that at least one polar and one nonpolar iodoorganic compound were present. Derivatisation of the iodoorganic compounds with, 1,4-diazabicyclo-2,2,2-octane to quatermary ammonium salts and²⁵²Cf plasma desorption mass spectrometry of the products revealed that the main iodoorganic constituents in the kerosene were iodobutane as polar and iodododecane as nonpolar compound in approximately equal concentrations.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

Removal of iodooraganic compounds from kerosene in nuclear fuel reprocessing

A 10^–5 mol 1^–1 solutiopn of idododecane in n-dodecane was used to simulate a kerosene sample from nuclear fuel reporcessing. Several methods were developed for the quantitative removal of iodododecane from the n-dodecane solution. Decomposition to elemental iodine was achieved either by washing with hyperazeotropic nitric acid or by exposure to a high-intensity UV-light. Quantitative removal of iodododecane from n-dodecane was achieved by absorption on silver nitrate impregnated materila or on activated charcoal, which was impregnated with potassium thiocyanate or 1,4-diazabicyclo-2,2,2-octance. The reaction could be accelerated by stirring or heating. Thus a quantitative absorption of idododecane could be achieved within a few minutes. The results of the experiments were confirmed by absorption of iodoorganic compounds from kerosene of the Karl sruhe nuclear fuel reprocessing plant (WAK) on the tested material.     

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.     

Analysis of plutonium in biological and environmental materials

A procedure for separation of plutonium from some biological and environmental materials has been tested in model and real conditions. The procedure involves a commonly used way of conversion of plutonium to oxidation state (IV) in nitric acid medium and sorption of Pu(IV) on a strongly basic anion exchanger from hydrochloric acid medium thus eliminating interference of²²⁸Th with the²³⁸Pu analysis.     

129I in soil and grass samples around a nuclear reprocessing plant

Neutron activation analysis was used to determine¹²⁹I in soil and grass samples around a reprocessing plant. The method involved wet oxidation of samples, using chromic acid, followed by distillation, collection of iodine in alkaline solution, loading on Dowex-1, irradiation and post-irradiation purification steps. The -activity of¹³⁰I isotope of the purified samples was measured for quantitative determination of¹²⁹I. The experimental results showed that¹²⁹I and the¹²⁹I/¹²⁷I atomic ratio in soil samples varied from 1.09×10^–4 to 5.33×10^–3 pCi g^–1 and 0.10×10^–6 to 6.12×10^–6, respectively. Further, the geometric mean of soil-to-plant transfer factor (Bv) for¹²⁹I was found to be 0.16 which was comparable with other published values.     

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,...     

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

Summary 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, respectively, can be determined. The K-lines of Pu are excited by external -radiation (¹⁹²Ir) and measured by a high-purity Ge detector. A calibration curve is presented and the detection limits are plotted as function of the time of measuring.

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Zerstörungsfreie bestimmung von plutonium in kernbrennstoff-stäben und -pellets

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Dedicated to Prof. Dr. G. Tölg on the occasion of his 60th birthday     

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.     

Iodine-129 and natural iodine-127 in deer thyroids from the environment of a small nuclear fuel reprocessing plant and from sites remote from nuclear facilities

Concentrations of the fission product¹²⁹I and natural¹²⁷I were determined in deer thyroids collected in the environment of the small Karlsruhe nuclear fuel reprocessing plant (WAK) and in a region remote from¹²⁹I sources of nuclear facilities. The isotopic ratio¹²⁹I/¹²⁷I in thyroids from the environment of WAK varies from 1.0×10^–6 to 12.9×10^–6, which is about one order of magnitude higher than the¹²⁹I/¹²⁷I ratios in thyroids from deer in a region remote from nuclear facilities. These ratios were between 0.2×10^–6 and 0.7×10^–6.     

Studies on the age determination of trace plutonium

An isotope dilution multicollector inductive coupled plasma mass spectrometry (ID-MC-ICP-MS) method for determining age of trace Pu through measuring ²⁴¹Pu/²⁴¹Am, ²⁴⁰Pu/²³⁶U ratio was established. At the same time, other two methods-α-spectrometry combined with MC-ICP-MS and liquid scintillator combined with α-spectrometry through measuring ²⁴¹Pu/²⁴¹Am ratio to determine the age of trace Pu were also studied. The techniques were explored for the age determination of nanogram grade Pu sample on the basis of Pu/Am, Pu/U separation. The ages of two Pu samples—one with known and the other with unknown age—were determined by the three methods. The determined ages by the three methods were all in agreement with the reference value. The established methods for determining the age of trace Pu could be adopted in the verification activities of nuclear safeguards and nuclear arms control.