Scaling-up experience of actinides partitioning from intermediate level high salted alkaline waste streams of Purex process using Versatic-10 extractant

Partitioning of minor alpha-emitting actinides, especially U, Pu and Am from medium active alkaline waste is possible from intermediate level liquid wastes (ILLW) produced during spent fuel reprocessing following Purex process. This paper deals with the efficient removal of alpha-activity from ILLW by solvent extraction process. Counter current batch extraction with O/A ratio 2:1 as well as multistage mixer settler has demonstrated that most of the alpha-activity was removed from the alkaline effluents using 20% Versatic-10 (V-10) in dodecane after giving 3 to 4 contacts, thus converting alkaline waste as non-alpha waste. Under the optimum conditions (pH 9.0-9.5 and VA-10), both Pu(IV) and Am(III) are highly extractable whereas U(VI) is relatively poorly extracted. To assess the applicability of this process for regular treatment of the waste, a feasibility study on pilot plant scale using six stage mixer settler was operated to treat the ILLW. The results indicated that almost >99.90% alpha-emitting actinides are removed. Dilute nitric acid (0.5M HNO₃) served as the most efficient strippant for all these actinides. This facilitate an easy regeneration of the extractant which can be recycled. This method is useful in obtaining alpha-free wastes and had positive impact on ease and safety aspects during subsequent waste treatment and long term storage.     

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.     

What is wise in the production of 99Mo? A comparison of eight possible production routes

The present paper addresses eight possible routes of producing ⁹⁹Mo, and discusses both yield and ⁹⁹Mo specific activities (SA) in the context of anticipated worldwide demand. Target dimensions are modelled by considering both limits set by cooling and by inside-target radiation attenuation characteristics. Energy deposition profiles are set up by MCNP6, reaction probabilities are taken from TALYS/TENDL and JANIS codes, and both are used in arriving at the produced ⁹⁹Mo. The outcomes suggest that U neutron-fission may remain one of the most relevant and efficient means of producing ⁹⁹Mo at the world-demand level, but that within this domain new developments may surface, such as ADSR or AHR production modes. Accelerator-based ⁹⁹Mo production is discussed as asking for developments in both target cooling and new concepts in post-EOB upgrading of ⁹⁹Mo SA, and/or new concepts for ⁹⁹Mo/^99mTc-generators, the latter possibly in both volumes (mass) and ⁹⁹Mo capacities.     

Initial generation and separation of99Mo at Sandia National Laboratories

The radioisotope^99mTc, used in greater than 80% of nuclear medicine applications, has traditionally been produced and supplied to radiopharmaceutical companies in the form of its precursor⁹⁹Mo. Nuclear fission produced⁹⁹Mo had been supplied by Nordion International of Canada and Cintichem, Inc. of New York, USA. With the shutdown of Cintichem's reactor in 1989, a need was recognized for a US supply, and the US Department of Energy recently published a record of decision designating Sandia National Laboratories (SNL) to meet that need. A recent campaign was launched which utilized the SNL Annular Core Research Reactor to irradiate UO₂ coated targets fabricated by Los Alamos National Laboratory to produce⁹⁹Mo. The irradiated targets were chemically processed in the SNL Hot Cell Facility to separate and purify the⁹⁹Mo. The campaign also included final product quality analysis, and process waste handling. The campaign was accomplished with high⁹⁹Mo recovery. Final product quality was assessed at SNL, and samples were sent to an outside laboratory for independent verification. The campaign provided data and experience useful in pursuing US Food and Drug Administration and radiopharmaceutical company approval.     

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%.     

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%.     

Solid targets for 99mTc production on medical cyclotrons

Recent disruptions in the molybdenum-technetium generator supply chain prompted a review of non-reactor based production methods for both ⁹⁹Mo and ^99mTc. Small medical cyclotrons (E _p ~ 16–24 MeV) are capable of producing Curie quantities of ^99mTc from isotopically enriched ¹⁰⁰Mo using the ¹⁰⁰Mo(p,2n)^99mTc reaction. Unlike most other metallic target materials for routine production of medical radioisotopes, molybdenum cannot be deposited by reductive electroplating from aqueous salt solutions. To overcome this issue, we developed a new process for solid molybdenum targets based on the electrophoretic deposition of fine ¹⁰⁰Mo powder onto a tantalum plate, followed by high temperature sintering. The targets obtained were mechanically robust and thermally stable when irradiated with protons at high power density.     

Feasibility study for production of 99mTc by neutron irradiation of MoO3 in a 250 kW TRIGA Mark II reactor

The subject of this paper is to explore the possibility to obtain ^99mTc from activation of ⁹⁸Mo, using the TRIGA Mark II low flux research reactor (Vienna, Austria). Irradiation of both natural and enriched in ⁹⁸Mo molybdenum oxides was compared. Aims of this work included the determination of neutron fluxes and ⁹⁸Mo(n, γ)⁹⁹Mo reaction effective cross section in the TRIGA Mark II reactor irradiation channels, calculation of ⁹⁹Mo specific activities, determination of optimal irradiation conditions for the subsequent ^99mTc separation from MoO₃ targets using concentrating technologies.     

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.     

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.     

The determination of molybdenum and tungsten in sea and surface water

A simple method for the determination of molybdenum and tungsten in sea and surface water is presented. Molybdenum and tungsten are concentrated on activated charcoal by adsorption as the ammonium pyrrolidine dithiocarbamate complex; the optimal pH for adsorption is 1.3. Mo and W are then determined by thermal neutron activation, forming ⁹⁹Mo (T_{$\text{1}\text{2}$} = 66.7 h) and ¹³⁷W (T_{$\text{1}\text{2}$} = 23.8 h), respectively. The ^99mTc daughter of ⁹⁹Mo is measured as soon as the equilibrium between ^99mTc(T_{$\text{1}\text{2}$}= 6 h) and ⁹⁹Mo is established. The detection limits are 0.05 μg Mo l^-1 and 0.05 μg W l^-1 (or 0.001 μg W l^-1 after a simple chemical separation).     

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.     

<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.     

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.     

Pharmaceutical grade sodium [99mTc] pertechnetate from low specific activity 99Mo using an automated 99Mo/99mTc-TCM-autosolex generator

Performance study of a computer controlled automated closed cyclic module for the separation and recovery of ^99mTc from low specific activity (n, γ) ⁹⁹Mo using methyl ethyl ketone (MEK) solvent extraction technique named ⁹⁹Mo/^99mTc-TCM-AUTOSOLEX (Technetium automated solvent extraction) Generator is described. The entire system is automated and controlled by a user-friendly PC based graphical user interface that actually supervises process via an embedded system based electronic controller. The average yield of separation of ^99mTc was above 85 % and ⁹⁹Mo breakthrough in ^99mTc pertechnetate was <0.002 %. The sodium pertechnetate obtained was a clear solution having pH 6–7, Radiochemical (RC). Purity >99 %, MEK content <0.1 % (v/v), Al and Mo content <10 µg/ml. R. C. Purity of ^99mTc-radiopharmaceuticals studied was not less than 96 %. Bio-Quality control studies confirm that sodium pertechnetate obtained was sterile and pyrogen free. Imaging studies in animals and humans with limited radiopharmaceuticals show that the quality of ^99mTc-pertechenate obtained in the present module was good enough to do clinical study.     

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.     

Separation of no-carrier-added <Superscript>90</Superscript>Nb from proton induced natural zirconium target

Amongst various radionuclides of molybdenum, ⁹⁰Mo and ⁹⁹Mo have suitable β energy for clinical uses. In this paper we report separation of ⁹⁹Mo from ⁹⁹Mo-^99mTc equilibrium mixture. The liquid–liquid extraction technique has been employed using trioctylamine (TOA) diluted in cyclohexane as organic phase and HCl as aqueous phase. At 10⁻⁵ M HCl and 0.5 M TOA concentration ^99mTc quantitatively transferred to the organic phase leaving ⁹⁹Mo in the aqueous phase. The developed separation method is efficient and provides very high separation factor.     

Combined determination of <Superscript>99</Superscript>Tc and <Superscript>108m</Superscript>Ag in L/ILW liquid wastes

The low- and intermediate-activity level liquid wastes produced by the Paks Nuclear Power Plant (NPP) contain routinely measureable gamma-emitting (e.g., ⁵⁴Mn, ⁶⁰Co, ^110mAg, and ¹³⁷Cs) as well as many so-called “difficult-to-measure” radionuclides. Despite of their low specific activity compared to the total, the reliable determination of these radionuclides is an important issue of nuclear waste management. The increasing amount of waste samples to be qualified yearly by our laboratory put a pressure on revising the existing procedure of ⁹⁹Tc separation applied. We have managed to halve the initial amount of the sample required to achieve the same level of detection of technetium. Furthermore, one of the new purifying steps introduced have proved to be able to separate ^108mAg (and ^110mAg) better than 99% keeping the ⁹⁹Tc content of the product almost intact. Means of separation of ⁹⁹Tc from ¹⁰⁶Ru and ^124+125Sb have also been successfully investigated. As intended, this new procedure has a major impact on the chemical reagent as well as the electricity requirement of the separation making it more cost-effective.     

A modified sorbent for99mTc generator

A modified sorbent for^99mTe generators of higher activities has been developed. The sorbent consists of two layers. The layer in which (F.P.)⁹⁹Mo is adsorbed contains alumina and silica gel mixture in the weight ratio 40∶60%. The underlaying layer contains 0.5% g of pure alumina. The performances of the columns filled with this sorbent are compared to these containing pure alumina with respect to the total elution efficiency of^99mTc and the elution efficiency ratio of subsequent elutions. Radiochemical and radionuclidic purities (⁹⁹Mo breakthrough) of eluates from both kinds of columns have also been determined and compared.     

Separation and purification of 99mTc from 99Mo produced by electron linear accelerator

The medical radionuclide ⁹⁹Mo was produced by the ¹⁰⁰Mo(γ,n) reaction using bremsstrahlung photons generated by an electron linear accelerator. The amount of ⁹⁹Mo produced was compared to that predicted by calculation using the particles and heavy ion transport code system. From the ⁹⁹Mo produced, highly pure ^99mTc was separated using the so-called technetium master milker, and the chemical yield of ^99mTc was 83–99 %. The installation of a new complex using this method and the electron linear accelerator with the preferable specification was suggested, and a possibility to supply the demand of ^99mTc was discussed and shown.