Adsorption of fission products on a metal surface in nitric acid solutions: Radiochemical study using a multitracer

Adsorption behavior of fission products in nitric acid solution on various alloys and metals was studied by using a multitracer produced by neutron irradiation of UO₂. The adsorption behavior of the fission products ⁹⁹Mo, ¹³¹I, ¹³²Te, ¹⁴⁰La, and ¹⁴³Ce, and ²³⁹Np was simultaneously studied. Some chemical decontamination tests were also examined. Clear adsorption of ⁹⁹Mo, ¹³¹I, and ¹³²Te was observed, whereas adsorption of ¹⁴⁰La, ¹⁴³Ce, and ²³⁹Np was not. The adsorption characteristics were discussed by considering anion-exchange reaction and surface complexation.     

Separation of heavier rare earths from neutron irradiated uranium targets

A radiochemical method is described for the separation of heavier rare earths from the fission of uranium. The method is particularly suitable for the separation of low yield (10⁻⁵%–10⁻⁷%), highly asymmetric rare earth fission products viz.^179,177Lu,¹⁷⁵Yb,¹⁷³Tm,^172,171Er,¹⁶⁷Ho and^161,160Tb in the neutron induced fission of natural and depleted uranium targets. Additional separation steps have been incorporated for decontamination from²³⁹Np (an activation product) and^93-90Y (a high fission-yield product) which show similar chemical behaviour to rare earths. Separation of individual rare earths is achieved by a cation exchange method performed at 80°C by elution with α-hydroxyisobutyric acid (α-HIBA).     

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.     

Separation of gram quantities of uranium from fission products by extraction chromatography

The separation of gram quantities of uranium from fission products has been investigated by extraction chromatography. The separation which is based on the difference in distribution coefficients between uranium and the fission products on a tributyl phosphate (TBP) resin in nitric acid medium, was carried out by means of high acidity feed and stepwise elution on a TBP chromatography column. The results show that this technique is capable to separate 5 g of uranium from a large quantity of fission products. The recovery of uranium is more than 99%. The decontamination factors of g- and b-activities were 2.1^.10³ and 2.3^.10³, respectively.     

A novel Np–Pu separation procedure based on extraction with di-(chlorophenyl)-dithiophosphinic acid in nitric solution

The extraction behavior of Pu(III), Pu(IV), Np(IV) and Np(V) with di(chlorophenyl)-dithiophosphinic acid (DCPDTPA) in toluene from nitric acid solutions was studied systematically. In aqueous solution with high nitric acid concentration, the extraction capability (represented by distribution ratio D) for Pu and Np in different valences with DCPDTPA comes as D _Np(IV) > D _Pu(IV) > D _Np(V) > D _Pu(III). A new radiochemical procedure for Np/Pu separation based on DCPDTPA extraction was proposed and tested with simulated samples. The recoveries of Np and Pu are as high as 80 % after the whole separation procedure, with the decontamination factor of trivalent lanthanide fission product element (e.g. Eu) greater than 1.5 × 10⁴. The decontamination factor of Pu–Np is 2.0 × 10³, while the decontamination factor of Np–Pu is greater than 4.8 × 10³ after additional purification.     

A rapid separation of radioactive cesium by nitrobenzene extraction using the chromium complex [Cr(aniline)2(NCS)4]−

A rapid separation of radioactive cesium by the solvent extraction method was investigated. Cesium ions are quantitatively extracted with [Cr(NH₂C₆H₅)₂(NCS)₄]⁻ into nitrobenzene. EDTA is an effective masking agent for other polyvalent cations. The extracted cesium can be back-extracted into the aqueous phase by shaking with 6N HCl. The method was applied to samples of a natural mixture of fission products and a reactor coolant. The decontamination factors for other predominant isotopes in fission products were 10²∼10⁴. The separation of^137m Ba from a mixture of¹³⁷Cs and^137m Ba is also described.     

Trennung von Spurenelementaktivitäten aus Neutronenbestrahltem Selen mit Hilfe der Gegenstrom-Ionenwanderung

For the determination of trace elements in neutron irradiated, selenium by gamma-ray spectrometry the separation of the matrix activity is often necessary. In model experiments the decontamination of cation and anion impurities from the matrix solution was investigated by the counter current ion migration. After a processing of 3 hrs the trace activities of Co, Cr, Ga, Na and Zn were decontaminated from Se with a factor of >10³. For the trace elements As and Te representing the anionic constituents, a decontamination factor of 3·10² was obtained.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

Nuclear charge distribution in the spontaneous fission of252Cf: Determination of fractional cumulative yields of133mTe and133gTe

The fractional cumulative yields (FCY) of^133mTe and^133gTe in the spontaneous fission of²⁵²Cf were measured for the first time by a radiochemical method. The values ofFCY are 0.533±0.014 and 0.291±0.042 for^133mTe and^133gTe, respectively. The isomeric state to ground state fractional independent yield (FIY) ratio of¹³³Te,R, was found to be 3.5. The root-mean-square angular momentum of the primary fragment corresponding to the fission product¹³³Te, Jr.m.s.=8.8h, was estimated according to a simple one-parameter statistical model. The fractional cumulative yields from this work together with other literature data in the mass region A=131–141 are compared with the normal yields given by the empiricalZ _p model by Whhl. It suggests that both theN=82 neutrons shell and nucleus pairing effects are not apparent for the spontaneous fission of²⁵²Cf.     

A new very sensitive LSC procedure for determination of Tc-99 in environmental samples

A new method for the determination of Tc-99 in different environmental samples has been developed. The sample is carefully ashed in a muffle furnace and then fused with Na₂CO₃ and K₂CO₃. The first step is an enrichment and purification of TcO ₄ ^– on an anion exchange resin. The Tc is desorbed as a cationic thiourea complex, which is held on a cation exchange resin. The complex is destroyed by oxidation to TcO ₄ ^– with (NH₄)₂S₂O₈ in sulfuric acid. From this solution TcO ₄ ^– is extracted into TBP/toluene and the organic phase is mixed with a scintillation cocktail and counted in an anticoincidence shielded LSC. Tc-99m is used as a chemical yield tracer. The decontamination factors for all important fission and activation products and naturally occurring radionuclides are in the range between > 10⁵ and > 10⁸. The detection limit is about 5 mBq per sample at a counting time of 1000 minutes. The maximum sample amount of plants is 500 g dry weight and therefore the lowest detection limit achievable is 10 mBq/kg. Ashing and dissolution of the samples takes 24 h and 4 analyses are performed by one technician in 8 hours. The chemical yield ranges from 50 to 80%.     

Measurement of absolute intensity of 1001 keV gamma-ray of 234mPa

The sorption behavior of ²³⁵U fission fission products ⁹⁹Mo and ¹³²Te was studied through batch and dynamic experiments when they were dissolved in 1 to 7M HNO₃ solutions. It was found that ⁹⁹Mo is always totally adsorbed on hydrated SnO₂, while ¹³²Te is rather weakly adsorbed, therefore they can be separated from each other although ¹³²Te in the solution still remains contaminated with other radionuclides as well as ⁹⁹Mo does in the solid.     

Separation of <Superscript>142</Superscript>La from fission products by SISAK technique

A rapid separation system based on SISAK technique was established to isolate ¹⁴²La successfully from fission products. SISAK technique is often applied in the separation of nuclides with the half-life of seconds or minutes. Here it was used to separate the parent nuclide of ¹⁴²La, which the half-life is in the magnitude of several seconds. According to the separation procedure designed in the paper, the activity of ¹⁴²La acquired is more than 10⁴ Bq and the decontamination factors for most γ-emitters are higher than 10³.     

The determination of tellurium in geological materials using radiochemical neutron activation analysis with a low energy photon detector

A method is described for the trace level determination of Te in geological materials with a detection limit of 5–10 ppb. Destructive thermal neutron activation analysis is used with relatively simple radiochemistry employing efficient precipitation and ion exchange techniques. A germanium Low Energy Photon Detector (LEPD) is used for radioassaying which allows the relatively aboundant X-rays from^123mTe to be measured. This radioactive isotope emits Te Kα and Kβ X-rays at 27–31 keV which are readily resolved by the LEPD and therefore allows interference effects from fission product Te to be minimised giving reliable trace level data of high accuracy. The validity of the method is demonstrated by reporting analytical data for Te in a range of USGS Standard Rocks.     

A fast separation for antimony produced in heavy-ion induced fission

Fission fragments from heavy ion induced fission were stopped in thin magnesium foils. A fast procedure based on evolution of stibine was developed to separate the antimony isotopes embedded in the foil. A separation system, and a glass pressure filtration system was constructed for this purpose. The chemical yield measured by three independent methods was 80–90%. The degree of decontamination from other fission products was >10². The whole separation took eight minutes.     

Separation of99Mo from132Te using thiourea as complexing agent. Application to the separation of99Mo from fission products

A radiochemical method to isolate⁹⁹Mo from¹³²Te, both produced in the fission of²³⁵U, has been developed. The method is based on the formation of a cationic complex of tellurium with thiourea in acid medium which is retained (98.7±0.5)% on a cation exchange resin (Dowex 50W-X8, 100–200 mesh), while (99.8±0.05)%⁹⁹Mo passes through it, due to the non-formation of such complex in the same experimental conditions. The radionuclidic purity of⁹⁹Mo was found to be suitable for the preparation of⁹⁹Mo–^99mTc generators. The retention of⁹⁹Mo on an alumina column as a function of pH was investigated and the best pH range for this purpose was found to be 4.0–4.5.     

Kathodenstrahlpolarographische Bestimmung von Spurenelementen in Tetrachlorogolds?ure nach ionenaustausch-chromatographischer Trennung

Zusammenfassung Die kathodenstrahlpolarographische Bestimmung von Ni, Pb, Cu, Co, Fe, Mo, Te, Sn, Zn, Bi, Cd, Tl in Tetrachlorogoldsäure wird beschrieben. Die Nachweisgrenzen liegen im 10^–4%- bzw. 10^–5%-Bereich (Ni, Co, Mo). Die Spuren werden über einen Ionenaustauscher AG 1X8 untereinander und von der Matrix getrennt.

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Cathode-ray polarographic determination of trace elements in tetrachloroauric acid after separation by ion exchange

A method is described for the cathode-ray polarographic determination of traces of Ni, Pb, Cu, Co, Fe, Mo, Te, Sn, Zn, Bi, Cd, Tl in H(AuCl₄). Gold is separated from the trace metals by absorption on the strongly basic anion-exchange resin AG 1X8. The detection limits are within the ranges of 10^–4% and 10^–5% ((Ni, Co, Mo).

     

Liquid-liquid extraction methods for the production of carrier-free 115cd and 89,90sr from fission products

The extraction of fission product elements with 1-phenyl-3-methyl-4-caprylpyrazolone-5 at various pH values has been investigated. The quantitative extraction of cadmium at pH 5.4 and that of strontium at pH 9.0 is utilised in devising procedures for the recovery of ¹¹⁵Cd and ^89,90Sr from the fission products. Good decontamination factors and more than 90% ¹¹⁵Cd and 80% ^89,90Sr activities were recovered.     

Chromatographic isolation of144Ce and144Pr from the wastes of irradiated uranium treatment

A two-step chromatographic technique was elaborated to isolate¹⁴⁴Ce,¹⁴⁴Pr from a solution of uranium fission products in 6M HNO₃. The oxidation to Ce(III) by bromate and selective adsorption of¹⁴⁴Ce(IV) on anion exchange column were used to concentrate and purify¹⁴⁴Ce. Some impurities of uranium,⁹⁵Zr,⁹⁵Nb,¹⁰⁶Ru remain in¹⁴⁴Ce solution after the first step of its isolation. The final purification is achieved by passing the 6M HNO₃ solution of¹⁴⁴Ce(IV) through the HDEHP-coated teflon column. The decontamination factors of¹⁴⁴Ce from main fission products are given. 7.2 mCi of (¹⁴⁴Ce+¹⁴⁴Pr) are recovered from each gram of irradiated uranium trioxide with the yield greater than 99%. An improvement of known generator was carried out to elute a purer¹⁴⁴Pr from maternal¹⁴⁴Ce(IV) adsorbed on the anion exchange column.     

A rapid radiochemical procedure for indium

A rapid radiochcmical procedure was developed for the separation of indium radionuclides from a mixed fission-product solution. An alcoholic pyridine solution is added to a uranium solution containing indium and tin carriers. The resulting tin precipitate is separated from the indium-containing solution by filtering through a cellulose membrane filter. The decontamination factor for tin is 2·10³. Other fission products are only partially removed. The chemical yield of indium is about 44%, and the time required for the separation is about 10 sec. After the tin-separated indium has decayed, the tin daughters of indium are removed from all the other fission products at a specified time and measured, so that the amount of indium present at the time of the tin precipitation is determined.     

Separation of iodine produced from fission using silver-coated alumina

A method developed for the preparation of silver-coated alumina, a new material for retention of iodine from alkaline solution is described. Experiments were carried out, oriented to the purification step of ⁹⁹Mo produced by uranium fission, based on the retention of radioiodine in this material. Iodine retention, as well as molybdenum non-retention, were found, both with excellent results. Further tests showed that the incorporation of this material has no influence on the subsequent ⁹⁹Mo retention in ion exchange resins. The elution of the radioiodines retained was tested with satisfactory results. This new material can be used not only for improving the ⁹⁹Mo purification and working conditions, but also as the basis of a method for recovering the fission produced ¹³¹I.