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.     

Elution of 235U fission products through a calcined hydrotalcite-packed column

Calcined hydrotalcite packed columns were utilized to sorb ²³⁵U fission products and their decay products. The elution behavior of some radionuclides was studied after washing the columns, either with distilled water or 0.5% NaCl solution. Afterwards, fission products and their decay products were eluted using 0.5% NaCl solution. It was found that no matter the washing process, ^99mTc, the b^--decay product of ⁹⁹Mo, was easily separated from ⁹⁹Mo which was strongly retained on the hydrotalcite. ¹³²I, the b^--decay product of ¹³²Te, was eluted slowly and was separated from ¹³²Te which was retained on the column. ¹³¹I and ¹⁴⁰Ba were eluted together with ^99mTc and ¹³²I, although in smaller proportions.     

Radiochemical separation of99Mo from natural uranium irradiated with thermal neutrons

⁹⁹Mo was separated from uranium and insoluble fission product hydroxides. More than 98% of⁹⁹Mo radioactivity was extracted with bis (2-ethylhexyl)phosphoric acid. The organic phase was washed and⁹⁹Mo was back-extracted from the organic phase with NH₄OH solution. The percent recovery from the organic phase was 91% and the purity of⁹⁹Mo was more than 99%. Pure^99mTc was also extracted from the organic phase with a saline solution. Reversed-phase partition chromatography was used for the purification of⁹⁹Mo from¹³¹I and other fission products (10% HDEHP on kieselguhr bed).¹³¹I and other isotopes were quantitatively eluted with 0.1M H₂SO₄,⁹⁹Mo was eluted using a mixture of 0.5 M HCl and 30% H₂O₂.     

Separation of trace impurities of131I from99Mo obtained by extraction from fission products of uranium

The radioisotope⁹⁹Mo was separated from a mixture of²³⁵U fission products in the presence of Hg²⁺ by sorption on a chromatographic column filled with Al₂O₃ and elution with 1M NH₄OH. Trace impurities of¹³¹I in the molybdenum fraction were eliminated by selective sorption on silver iodide or by repeated sorption of⁹⁹Mo on Al₂O₃.     

Fission yields of molybdenum in the Oklo natural reactor

The isotopic compositions of molybdenum in six uranium-rich samples from the Oklo Zone 9 natural reactor were accurately measured by thermal ionization mass spectrometry. The samples were subjected to an ion exchange separation process that removed the isobaric elements zirconium and ruthenium, with high efficiency and a low blank. Molybdenum possesses seven isotopes of which ^92,94,96Mo are unaffected by the fission process, enabling the raw data to be corrected for isotope fractionation by normalising to ⁹²Mo/⁹⁶Mo, and to use ⁹⁴Mo to correct for the primordial component in each of the fission-produced isotopes. This enables the relative fission yields of Mo to be calculated from the isotopic composition measurements, to give cumulative fission yields of 1:0.941:0.936:1.025 for ^95,97,98,100Mo, respectively. These data demonstrate that the most important nuclear process involved in reactor Zone 9 was the thermal neutron fission of ²³⁵U. The consistency of the relative cumulative fission yields of all six samples from different locations in the reactor, implies that Mo is a mobile element in the uraninite comprising Zone 9, and that a significant fraction of molybdenum was mobilized within the reactor zone and probably escaped from Zone 9, a conclusion in agreement with earlier published work.     

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.     

Improved processes of molybdenum-99 production

Two improved processes of⁹⁹Mo production have been developed on laboratory scale. The first one allows to purify Mo of natural isotopic composition from tungsten impurities from 64 to <10 ppm by using preferential adsorption of tungsten on hydrated tin(IV) oxide (SnO₂ nH₂O) before irradiation in a nuclear reactor. The second process deals with the separation of pure fission product⁹⁹Mo from²³⁵U irradiated in a reactor. Two versions of separation process for production of fission⁹⁹Mo have been developed. Both versions start with the dissolution of²³⁵U oxide target in nitric acid and are based on sequential use of alumina and anion exchange resin AG® 1-X8 columns. The yield of⁹⁹Mo in both versions is 80–89%.     

The determination of molybdenum in geological samples by Neutron Activation Analysis

A metal-silicate extraction technique combined with neutron activation analysis has been developed to determine molybdenum in geological samples. The samples are equilibrated with Femetal powder at high temperatures. Molybdenum is completely extracted into the metal phase because of very reducing conditions in the furnace. The metal spherule is separated from the silicates, irradiated and dissolved in an acid solution. The molybdenum is precipitated as a sulfide and the precipitate is dissolved in aqua regia and counted on a Ge/Li/detector. The radiochemical yield is obtained by irradiation of the solution. The method avoids production of⁹⁹Mo from induced fission of²³⁵U by performing the metal-silicate separation before irradiation. The precipitation step may be necessary to remove the high background from the decay of⁵⁹Fe. Mo concentrations down to 15 ng/g have been obtained using this method.     

Determination of Mo,W and Zr in molybdates and tungstates of zirconium and titanium

Molybdate and tungstate of zirconium and titanium gels, used as matrices of ⁹⁹Mo/^99mTc and ¹⁸⁸W/¹⁸⁸Re generators, were synthesized under different conditions, in order to establish their performance and to choose the most appropriate gel to produce commercially. This type of generators demands a high content of Mo or W (>25%) in the matrices, since they use ⁹⁹Mo and ¹⁸⁸W of low specific activities. Therefore, it is of vital importance, to know the concentration of W and Mo in these gels, to determine their viability as matrices of the ⁹⁹Mo/^99mTc and ¹⁸⁸W/¹⁸⁸Re generators. There are different analytical methods to determine Mo and W, however, the presence of Zr and Ti in these gels, in many occasions, interfere in the analysis, imposing the previous separation of both metals before their determination. Therefore, the preparation time of the sample, the cost and the generation of chemical waste of these analyses are increasing. In order to eliminate these difficulties, the concentration of Mo, W and Zr of approximately 43 gels of molybdate and tungstate of zirconium and titanium, were evaluated by NAA without preparation of the samples. The results of this study reveal that the conditions of preparation of the gels influence directly their Mo and W content. In general, the titanium molybdate gels possess, on the average, a larger content of Mo (37%) than the zirconium molybdate gels (30%), while the titanium tungstenate gels contain only 8.5%.     

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.     

Application of neutron activation analysis in a fission molybdenum separation study

The separation of⁹⁹Mo from low-enriched uranium (LEU, 19.5%²³⁵U) targets was evaluated using natural uranium (NU) and non-radioactive tracers. Neutron activation analysis was used to determine (1) the efficiency of molybdenum recovery and (2) the decontamination factor of numerous fission product elements from the molybdenum product. Using NU and non-radioactive elements simplified procedures and allowed tests to be completed in a fume hood instead of a shielded cell. During activation of the non-radioactive tracers, uranium fission occurs, which can interfere with subsequent gamma-ray analysis. A comparison was made of the interferences caused by these fission products from both NU and LEU.     

An alternative route for the preparation of the medical isotope 99Mo from the 238U(γ, f) and 100Mo(γ, n) reactions

The radionuclide ⁹⁹Mo, which has a half-life of 65.94 h was produced from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions using a 10 MeV electron linac at EBC, Kharghar Navi-Mumbai, India. This has been investigated since the daughter product ^99mTc is very important from a medical point of view and can be produced in a generator from the parent ⁹⁹Mo. The activity of ⁹⁹Mo was analyzed by a γ-ray spectrometric technique using a HPGe detector. From the detected γ-rays activity of 140.5 and 739.8 keV, the amount of ⁹⁹Mo produced was determined. For comparison, the amount of ⁹⁹Mo from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions was also estimated using the experimental photon flux from ¹⁹⁷Au(γ, n)¹⁹⁶Au reaction. The amount of ⁹⁹Mo from the detected γ-lines is in agreement with the estimated value for ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions. The production of ⁹⁹Mo activity from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions is a relevant and novel approach, which provides alternative routes to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions, circumventing the need for a reactor. The viability and practicality of the ⁹⁹Mo production from the ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions alternative to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions has been emphasize. An estimate has been also arrived based on the experimental data of present work to fulfill the requirement of DOE.     

Preparation of99Mo,132Te isotopes and99mTc,132I generators

Adsorption and desorption of⁹⁵Zr−⁹⁵Nb,⁹⁹Mo,¹⁰³Ru,¹³²Te and²³⁹Np in a HCl-alumina system were studied in order to purify⁹⁹Mo and¹³²Te obtained by the cation-exchange separation of fission products and to prepare highly pure^99mTc and¹³²I generators.⁹⁹Mo and¹³²Te, of which radionuclidic purity was over 99.99% and 99.999%, respectively, could be obtained by passing the cation-exchange separated Mo and Te fractions through alumina columns, by washing with HCl and finally by eluting⁹⁹Mo with 1M NH₄OH and¹³²Te with 3M NaOH. In order to raise the recovery of⁹⁹Mo and¹³²Te from the alumina columns, they should be eluted as quickly as possible after the adsorption. The direct use of the alumina column containing⁹⁹Mo or¹³²Te as the generator allowed milking of^99mTc or¹³²I, of which radionuclidic purity was over 99.999%. Milking yields of^99mTc with 0.1M HCl and¹³²I with 0.01M NH₄OH were 77% and 90%, respectively. The latter value was much higher than that in usual performance of the generator.     

Radiochemical determination of fission products of uranium

An analysis has been elaborated to determine the long-living γ-emitting fission products of uranium. It consists of a sodium bisulphate melt of the fission product solution or the U-fuel, followed by liquid-liquid extractions. Afterwards the isotopes are absolutely counted with a standardized 3″×3″ NaI crystal. The total γ-spectrum of the original fission product solution, taken with a NaI crystal or a Ge−Li detector, is also analyzed mathematically by mixed γ-spectrometry. From a short post-irradiation of the fission product solution the concentrations of both²³⁵U and²³⁸U are determined. The absolute amount of fission products related to the U-concentration allows the calculation of the percent atomic burn-up, the irradiation time, the cooling period, the flux of the reactor and the original degree of enrichment of the uranium. Research associate of the I. I. K. W.     

Evaluation of proton induced reactions on 100Mo: New cross sections for production of 99 mTc and 99Mo

The use of the ⁹⁹Mo^99mTc generator in nuclear medicine is well established world wide. The production of the ⁹⁹Mo (T_1/2 = 66 h) parent as a fission product of ²³⁵U is largely based on the use of reactor technology. From the early 1990's accelerator based production methods to provide either direct produced ^99mTc or the parent ⁹⁹Mo, were studied and suggested as potential alternatives to the reactor based production of ⁹⁹Mo. A possible pathway for the charged particle production of ^99mTc and ⁹⁹Mo is irradiation of molybdenum metal with protons via the reaction ¹⁰⁰Mo(p,2n)^99mTc and ¹⁰⁰Mo(p,pn)⁹⁹Mo, respectively. The earlier published excitation functions show large differences in their maximum that result in large differences in the calculated yields. We therefore decided to study the excitation function for these proton-induced reactions. In this work the newly measured excitation functions as well as an evaluation of earlier measured data and a discussion of the observed disagreements are presented.     

Uranium determination by133Xe in mineral separates from igneous rocks comparative study with fission track and thermal release characters of133Xe

Samples of an andesite (Asama-yama, Japan) and a basalt (Kilauea, Hawaii) were finely fractioned by density and the U distributions among the separates were determined by homogenized fission track method. Groundmass was found to be enriched with U; one half of the andesite U and almost all of the basalt U are accounted for by the groundmass U. Cerium and other REE are correlated with U. In the andesite separates Na is accompanied by these elements, but Fe and Co are rather anticorrelated. Uranium concentrations in the bulk samples and some of the separates were then compared with those determined by¹³³Xe extracted by heating at 1600°C. Fissiogenic¹³³Xe seems to become labile in groundmass, especially of the andesite, during reactor irradiations for 5 h or more. Stepwise heating experiments suggested that¹³³Xe in U-rich groundmass tends to escape at low temperatures and the remaining¹³³Xe is mostly retained in highly refractory sites in phenocryst rather poor in U.     

Sensitivity study on modeling radioxenon signals from radiopharmaceutical production facilities

As part of the Comprehensive Nuclear Test-Ban Treaty (CTBT), the International Monitoring System (IMS) was established to monitor the world for nuclear weapon explosions. As part of this network, systems are in place to monitor the atmosphere for radioxenon. The IMS routinely detects radioxenon from sources other than nuclear explosions. One of these radioxenon sources is radiopharmaceutical production facilities. This is a sensitivity study on the nuclear forensic signals possible from such facilities. A fission process model was produced to calculate the activity of ^131mXe, ^133mXe, ¹³³Xe and ¹³⁵Xe in the process utilized to produce ⁹⁹Mo and ¹³¹I for medical applications through high enriched uranium fission. The computer model accounts for fractionation of radionuclides within a decay chain that may result from filtering or chemical procedures. Ratios of the radioxenon isotopes are calculated as a function of decay time after the release. The ratios are then compared to those expected from nuclear explosions. The main conclusion from this work is that the two main factors that affect the nuclear forensic signal from radiopharmaceutical production facilities are the sample irradiation time and the use of emission gas storage tanks.     

Spontaneous fission yields for 238U in molybdenum measured in Archaean zircons

The relative cumulative fission yields of ^95,97,98,100Mo produced by spontaneous fission of ²³⁸U contained in Archaean zircons, were measured by sensitive thermal ionization mass spectrometry (TIMS). The relative yields for ^95,97,98,100Mo are 0.58 : 1.08 : 1.04 : 1.0, respectively. Combined with mass spectrometrically-determined ^{99,101,102,104}Ru fission yields,¹ the mass distribution from 95￡A￡104 can be delineated. Assuming an “absolute” fission yield of 6.1±0.4% for the cumulative fission yield at mass 97, it is possible to express the Mo and Ru relative spontaneous fission yields for ²³⁸U as “absolute” values. There is no evidence for a significant isotope anomaly at mass 98. This revised version was published online in July 2006 with corrections to the Cover Date.     

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.     

Determination of uranium fission interference factors for INAA

Instrumental neutron activation analysis (INAA) is a very suitable technique for the determination of several elements in different kinds of matrices. However, when the sample contains high uranium concentration this method presents interference problems of uranium fission products. The same radioisotopes used in INAA are formed in uranium fission. Among these radioisotopes are ¹⁴¹Ce, ¹⁴³Ce, ¹⁴⁰La, ⁹⁹Mo, ¹⁴⁷Nd, ¹⁵³Sm and ⁹⁵Zr. The purpose of this study was to evaluate uranium fission interference factors to be used in the INAA of environmental and geological samples containing high levels of U. The obtained interference factors agreed with literature reported values. The results point to the viability of using these experimentally determined interference factors for the correction of uranium fission products.