Miniature Neutron Source Reactors in medical research: achievements and challenges

Miniature neutron source reactors (MNSRs) and their challenges were investigated. Then, MNSRs structure and operation were discussed, and conflicting usages of high enriched uranium fuel and its technical challenges were explored. By using low enriched uranium it is possible to produce ⁹⁹Mo isotopes with similar quality, plus using low enriched uranium does not increase the potential risks and high does level. Furthermore, it is possible to produce several medical radioisotopes such as ⁹⁹Mo/^{99 m}Tc generator, ¹³¹I, ⁶⁵Zn, and ¹⁷⁷Lu. These reactors can be employed in Boron Neutron Capture Therapy. After conversion, these reactors can be used more frequently and effectively.     

Selective separation of99Mo from fission products in chloride media on activated alumina

The radioisotope⁹⁹Mo generated through the fission of neutron-irradiated uranium targets was separated in an extremely radiochemically pure state. The target dissolution mixture was run on a silica gel column whereby the⁹⁵Zr−⁹⁵Nb activity, the serious radiocontaminants of⁹⁹Mo, was adsorbed on the gel. The effluent from the gel column was processed and treated with sodium dihydrogen phosphate. The processed solution was run on an activated alumina column whereby⁹⁹Mo was adsorbed as phospho-molybdate complex ion, then desorbed from the column as the molybdate ion by eluting with 10% NaOH solution. By recycling the adsorption and desorption steps on alumnia,⁹⁹Mo was finally obtained totally free from all other radiocontaminants.     

Production of99Mo and its application in nuclear medicine

Molybdenum-99 is produced in large quantities as the parent radioisotopes of^99mTc, which has been used recently in nuclear medicine. The neutron capture reaction on molybdenum and the nuclear fission of uranium are used for the large scale production of⁹⁹Mo. The products by these methods are used properly according to the objects of diagnosis. In this paper, the production of⁹⁹Mo is reviewed and the development of the production in JAERI is outlined.     

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.     

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.     

Biosorption of 241Am by immobilized Saccharomyces cerevisiae

More than half of the world's annual production of radionuclides is used for medical purposes such as diagnostic imaging of diseases and patient therapy. Using aqueous homogeneous solution reactor technology, production quantities of medical radioisotopes ⁹⁹Mo and⁸⁹Sr, can be extracted from one reactor cycle. ⁹⁹Mo may be produced directly from UO₂SO₄ uranyl sulfate in an aqueous homogeneous solution nuclear reactor in a manner that produces high purity radionuclides, making efficient use of the reactor's uranium fuel solution. The process is relatively simple, economical, and waste free, eliminating uranium targets. The short-lived radioisotope ^99mTc is eluted from ⁹⁹Mo for diagnostic imaging. Radioisotope ⁸⁹Sr infusion is a therapeutic modality that reduces reliance on narcotic analgesia through palliation of metastatic bone pain caused by metastases of the cancer to the bone. Painful disseminated osseous metastases are common with carcinomas of the lung, prostate, and breast. Synergistic interleaving of two manufacturing processes, one producing ⁹⁹Mo and another producing ⁸⁹Sr in the same production cycle of an aqueous homogeneous solution reactor makes full and efficient use of the time for both the neutron irradiation stage and the extraction stage of each radionuclide. Interleaving the capture of ⁸⁹Sr radioisotope with production processing of ⁹⁹Mo radioisotope is achieved, since the extraction and subsequent elimination of radionuclide impurities occurs during separate parts of the reactor cycle. The process applies to either HEU or LEU nuclear fuels in an aqueous homogeneous solution reactor.     

Radiochemical neutron activation analysis of Mo and W in geochemical and cosmochemical samples with an emphasis on the correction of fission products

Radiochemical neutron activation analysis (RNAA) was applied to geochemical and cosmochemical samples to determine trace amounts of Mo and W. To determine the Mo concentration by NAA accurately, the contribution of the fission products of U should be corrected. For that reason, we developed a simple and effective method, where a contribution of fissiogenic ⁹⁹Mo was estimated by monitoring the ratio of uranium fission-product ⁹⁹Mo to ¹³³I. Mo concentrations corrected for fission with the W concentrations were consistent with the literature values, showing that ¹³³I was found to be an effective monitor for fission correction. Detection limits are estimated to be 10 ppb for Mo and W and 30 ppb for U under the present experimental conditions.     

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.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.     

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.     

Influence of the postreaction time in the concentration of141Ce anionic species in uranium tricarbonated solutions

An effective method to minimize the presence of¹⁴¹Ce in the final hexahydrated uranyl nitrate recovery product has been obtained. This condition is considered as one prepurification stage in the recovery process of residual nonfissioned uranium in the production of⁹⁹Mo of fission.     

Absolute yields of 99Mo and 111Ag in the reactor neutron induced fission of 238U

The absolute fission yields of ⁹⁹Mo and ¹¹¹Ag have been measured in the reactor neutron induced fission of depleted uranium employing a fission track-cum-radiochemical technique. The absolute disintegrations were determined from beta decay and Ge(Li) gamma spectrometry. The total number of fissions were measured by fission track registration on lexan solid state track detector immersed in fissile material solution. The yields for⁹⁹Mo and ¹¹¹Ag were found to be (6.27 ± 0.30)% and (0.046 ± 0.002)% respectively. The results are compared with the published data and the merits of the present method have been discussed.     

Interference in neutron activation analysis of rocks by uranium fission

Interferences by uranium fission for⁹⁵Zr,⁹⁹Mo,¹⁰³Ru,¹⁴⁰La,¹⁴¹Ce and¹⁴⁷Nd have been studied using a single comparator method with two monitors. The effect of the neutron energy spectrum on the interference factor was examined by using the effective activation cross section. All the activities of¹⁴⁰La produced during neutron irradiation of uranium were included in the calculation of the factor for lanthanum. The calculated and experimental interference factors are in good agreement within 10% deviation. The results have been applied for the analysis of several rock samples containing uranium in a wide concentration range.     

Analytical methods for the characterization and the leachability evaluation of a solid waste generated in a phosphoric acid production plant

Naturally Occurring Radioactive Materials (NORM) are present in the environment and can be concentrated by technical activities, particularly those involving natural resources. These NORM deposits are highly stable and very insoluble under environmental conditions at the earth's surface. However reducing or oxidant conditions or pH changes may enable a fraction of naturally occurring radionuclides to eventually be released to the environment. In this paper the analytical methods to determine the chemical and radiochemical composition of a solid waste generated in an acid phosphoric production plant was shown. The sample, analysed by gamma spectrometry and radiochemical methods, contained ²¹⁰Pb and ²¹⁰Po and a high activity concentration of uranium isotopes but it shows a very low quantity of ²²⁶Ra. Also a sequential extraction method consisting of five operationally-defined fractions was used to evaluate the leachability of uranium isotopes, ²¹⁰Pb and ²¹⁰Po. The average leaching potential observed in the sample is 97.6% for ²³⁸U, 93.2% for ²¹⁰Pb and 82.4% for ²¹⁰Po. Moreover the results show that ²¹⁰Pb and ²¹⁰Po are leachable under only extreme conditions, whereas uranium is very more soluble.     

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.     

Transfer of210Pb and210Po to the environment during mining and processing of uranium ores

Transfer of²¹⁰Pb and²¹⁰Po from an uranium mine and mill to the environment was studied by measuring their concentrations in different stages of uranium extraction technology, in waste products from the mill and in environmental waters. A slightly modified radiochemical method was used for²¹⁰Po and²¹⁰Pb determination.     

Developing a Chromatographic 99mTc Generator Based on Mesoporous Alumina for Industrial Radiotracer Applications: A Potential New Generation Sorbent for Using Low-Specific-Activity 99Mo

The commercial low-pressure column chromatographic ⁹⁹Mo/^99mTc generator represents a reliable source of onsite, ready-to-use ^99mTc for industrial applications. These generators use fission-produced ⁹⁹Mo of high specific activity, posing serious production challenges and raising proliferation concerns. Therefore, many concepts are aimed at using low-specific-activity (LSA) ⁹⁹Mo. Nonetheless, the main roadblock is the low sorption capacity of the used alumina (Al₂O₃). This study investigates the feasibility of using commercial alumina incorporated with LSA ⁹⁹Mo to develop a useful ⁹⁹Mo/^99mTc generator for industrial radiotracer applications. First, the adsorption profiles of some commercial alumina sorbents for LSA ⁹⁹Mo were tested under different experimental conditions. Then, the potential materials to develop a ⁹⁹Mo/^99mTc generator were selected and evaluated regarding elution yield of ^99mTc and purity. Among the sorbents investigated in this study, mesoporous alumina (SA-517747) presented a unique sorption-elution profile. It demonstrated a high equilibrium and dynamic sorption capacity of 148 ± 8 and 108 ± 6 mg Mo/g. Furthermore, ^99mTc was eluted with high yield and adequate chemical, radiochemical, and radionuclidic purity. Therefore, this approach provides an efficient and cost-effective way to supply onsite ^99mTc for radiotracer applications independent of fission-produced ⁹⁹Mo technology.     

A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

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.     

99mTc generator with hydrated MnO2 as adsorbent of 99Mo

An ^99mTc generator with MnO₂ as adsorbent of ⁹⁹Mo was investigated. Through batch experiments the retention of ⁹⁹Mo onto MnO₂was studied as a function of the shaking time and the pH value of the ⁹⁹Mo solution. It was found that ⁹⁹Mo showed a retention of 100% onto MnO₂ at the pH range from 3 to 11, and the equilibrium distribution was reached in less than 10 minutes. In column experiments the retention of ⁹⁹Mo onto MnO₂ was also high: 99.72%. In this case, the pH of the utilized ⁹⁹Mo solution was 5. The ^99mTc could be eluted from the MnO₂-⁹⁹Mo column by using either distilled water with a pH of 5 or an aqueous solution of 0.9% NaCl. With the saline solution, the ^99mTc elution yields were higher than 80%, and only one aliquot of 5 ml was needed to get these yields. The best results were obtained when the column was packed with 1 g of MnO₂. The water and the saline solution were passed through the column at a rate of 1.25 ml/min.