Mo Yield Research Articles

A New Molybdenum Production Element for Implementation in TRIGA Reactors: Thermal-Hydraulic Characterization

Oregon State University (OSU) has developed and patented a technology that produces 99Mo within a standard TRIGA reactor core and does not negatively impact safety bases for the operations of such reactor designs. This new technology, referred to as the “molybdenum element,” is intended on being demonstrated within the OSU TRIGA Reactor (OSTR) with figures of merit including 99Mo yield and operation. A comprehensive design and thermal-hydraulic analysis has been conducted to characterize the safety-related traits of the molybdenum element to facilitate a license amendment through the U.S. Nuclear Regulatory Commission to insert such an experiment in the OSTR.This study details the thermal-hydraulic characteristics of the molybdenum element exhibited within the OSTR under the three sets of conditions necessary to demonstrate the element’s safety. The study leverages the lumped-parameter code RELAP5-3D Version 2.4.2 for conduct of the primary body of this work. The first condition analyzes the molybdenum element’s response under steady-state, full-power operation; the second condition assumes that the inner region of the annular molybdenum element is blocked while remaining at full power; and the last condition considers several loss-of-coolant-accident scenarios. Key thermal-hydraulic parameters that may impact the safety of the OSTR are identified, presented, and discussed herein. The result of this study provides objective evidence through use of RELAP5-3D that the molybdenum element remains in a safe state during the steady and abnormal conditions considered.

Computational investigation of 99Mo production yield via proton irradiation of natU and 232Th targets

Accelerators have some advantages such as safety and cheaper operating and decommissioning costs for 99Mo production. Yield theoretical calculation using computational codes can powerfully estimate usefulness of a proposed nuclear reaction for a routine manufacturing. In this work, Monte Carlo-based code was used to compute 99Mo yield in 232Th and natU proton-irradiated targets, as well as maximum applicable beam current. Results showed that the code well agrees with published experimental data. The targets can bear maximum beam current of 30µA. Targets from 232Th provides higher 99Mo yield.

Preparation and characterization of cerium(IV) tellurium molybdate gel and its application as a bed for chromatographic <sup>99</sup>Mo/<sup>99m</sup>Tc generator

Cerium(IV) tellurium molybdate was prepared as a new gel bed for chromatographic 99Mo/99mTc generator. The gel material prepared with the optimum conditions was characterized with IR spectroscopy, thermal analysis, XRD, EDX and FESEM. The highest precipitation yield of 99Mo within the formed gel was found to be 97.6 % at a Ce:Te:Mo molar ratio of 1:0.2:1. The prepared 99Mo/99mTc chromatographic column based on Ce(IV) tellurium molybdate gel bed showed a good performance. The 99mTc elution yield was 77.8±3.0 % with a radionuclidic purity of ≥ 99.99 %, radiochemical purity of 96.5±1.3 % (as 99mTcO4-) and pH-value in the range of 5-7. Molybdenum and cerium levels in 99mTc eluates did not exceed 1 and 0.3 ppm, respectively, while no tellurium was detected.

Preparation of 99Mo and 99m Tc by 100Mo(γ,n) photonuclear reaction on an electron accelerator, MT-25 microtron

Molybdenum-99 was prepared by 100Mo(γ,n) photonuclear reaction on an electron accelerator, MT-25 microtron (FLNR). Natural Mo foil (9.6% 100Mo) and 100MoO3 (100% 100Mo) were used as the targets. The radiation yield of 99Mo was 3.2 kBq/(μA h mg 100Mo). Technetium-99m was isolated from the target material and concomitant radioactive impurities by an ion-exchange procedure. The possibility of preparing 99mTc generator from molybdenum carbonyl was studied. The content of radioactive impurities in the final 99mTc preparations was no higher than 10−5 Bq Bq−1.

Uncertainty Analysis on Fission Molybdenum Production with a Nuclear Fuel Target in a Research Reactor

The use of a low-enriched uranium (LEU) fuel target was examined for the feasibility of 99Mo production in a High-flux Advanced Neutron Application Reactor (HANARO). Uncertainty analysis was done with respect to the 99Mo yield ratio, 239Pu yield ratio, annual production rate, and decontamination requirement. Validity of a coupled code system, MCNP/ORIGEN2, was evaluated to estimate reliable isotopic number densities after irradiation and cooling. An equilibrium core model for the MCNP fixed-source problem was found by the reactor design methodology known as WIMS/VENTURE. Optimized target design options were proposed for both the LEU and highly enriched uranium (HEU) targets. Variables related to the target fabrication process and reactor physics condition were considered as uncertainty-inducing parameters. The most important factor affecting the overall uncertainty of the LEU option was the engineering tolerances achievable in the fabrication process of fuel film. The LEU has twice the uncertainty of HEU under current technology, which makes the economics of LEU worse than HEU. It is acceptable, however, in view of the radioactive purity of the alpha emitter because the uncertainty of the impurity level of 239Pu is expected to be relatively small - only 6.5% with a 95% confidence level.

Obtaining molybdenum-99 in the IRT-T research reactor using resonance neutrons

The yield of 99Mo from the 98Mo(n,γ) 99Mo reaction significantly depends of the energy spectrum of the neutron flux. It is well known that the cross-section for this reaction is about 130 mb, whereas the resonance integral of the reaction is 6.9 b. The aim of this work was to investigate the conditions that let to increase 99Mo yield from the targets with natural and enriched isotope composition under irradiation by resonance neutrons at the IRT-T research reactor. The calculations of integrated cross-sections of all Mo isotopes in the region of the 98Mo resonances showed that screening in the target with natural isotope composition by other isotopes is relatively small. So the 98Mo in the natural mixture can be activated by resonance neutrons approximately in the same manner as pure 98Mo. Experimental measurements of the 98Mo(n,γ) effective cross-section using the MoO 3 sample with natural and enriched composition in the reactor channels with the beryllium moderator with the thickness of 20 up to 90 mm showed that the effective cross-sections in these channels reach the value of 700 mb. The contribution of the epithermal neutrons into the 98Mo activity was 68% for the enriched targets and 78% for natural molybdenum, respectively. At that channel it is possible to produce 99Mo with specific activity up to 3.4 Cu/g with samples of natural isotope composition and up to 15 Cu/g with enriched samples on the base of reactors with neutron flux of (1.7 × 10 14 n/(cm 2 s)). Such 99Mo specific activity is enough not only to realize extraction technologies production of 99mTc, but to manufacture sorption generators of 99mTc without wastes.

Alice predictions on the accelerator production of molybdenum-99

The ALICE code was used to calculate the excitation functions for proton induced reactions on 100Mo that lead to the production of Nb, Mo and Tc isotopes. The ALICE code predicts a maximum cross section of the order of 100 mb for the 100Mo(p,pn) 99Mo nuclear reaction. ALICE predicts that the production of a few Curies of 99Mo is feasible if high intensity (mA) protons are degraded from 50→30 MeV in a target of highly isotopically enriched 100Mo. The ALICE code cannot be used to predict the cross sections for the isomeric states for the 100Mo(p, 2p) 99m+gNb→ 99Mo reactions. However, the predicted cross section for protons on 100Mo to yield 99Nb suggests that the (p,2p) reaction channel does not contribute significantly to the thick target yield of 99Mo. Several considerations suggest that accelerator production of 99Mo→ 99mTc generators is not a practical economic alternative to fission produced 99Mo. Considerations for the cyclotron production of 99mTc are discussed.

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%.

96/01586 Validation of CATHENA againts the CS28-1 high temperature experiment under an impaired cooling condition

99mTc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of 99Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of 99Mo is investigated. A modified 37-element fuel bundle, suitable for the production of 99Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of 99Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium.The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of 99Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated 99Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of 99Mo.

The occurrence of molybdenum-99 in natural and in depleted uranium salts and the spontaneous fission half-life of uranium-238

Measurements of the 99Mo/ 238U equilibrium ratios in aqueous solutions of non-irradiated uranium salts gave the following results: (4·4 ± 0·4) × 10 −4 and (4·4 ± 0·3) × 10 −4 disintegrations per second of 99Mo per gramme of 238U in natural uranium and in depleted uranium, respectively. A value of (8·0 ± 0·5) × 10 15 years, which agrees with Segrè's value, is obtained from the above data for the spontaneous fission half-life of 238U, if one assumes that the yield of 99Mo is 6·4 per cent and the spontaneous fission of 238U is the only source of 99Mo.

The mass-yield curve for fission of natural uranium by 14-MeV neutrons

Relative yields for the fission of natural uranium (effectively 238U fission) by 14-MeV neutrons have been measured for mass numbers 89, 91, 99, 111, 115, 140, 143, 144, 147, 153, and 156, and the general shape of the mass-yield curve is given. The peak-to-trough ratio is 9·1 ; the average number of prompt neutrons emitted per fission is found to be (4 ± 0·5); and a value for the absolute fission yield of 99Mo is found to be (5·58 ± 0·56) per cent. The light peak has moved 3–4 units towards the heavy peak when compared with the 14-MeV fission of 235U.