Masakazu Tanase

A 99mTc generator using a new inorganic polymer adsorbent for (n, γ) 99Mo

Abstract A 99mTc generator has been prepared using an inorganic polymer adsorbent newly developed for low specific activity 99Mo. The adsorbent (PZC) was synthesized from zirconium chloride and isopropyl alcohol on heating. To adsorb 99Mo, PZC was added to the solution of 99Mo prepared from neutron-irradiated natural MoO3. The amount of 99Mo (Mo) adsorbed to PZC reached about 200 mg/g-PZC. Technetium-99m was eluted with a small volume of saline solution in the yield of about 80%. The breakthrough of 99Mo was less than 0.5% in good conditions.

Successful labeling of 99mTc-MDP using 99mTc separated from 99Mo produced by 100Mo(n,2n)99Mo

We have for the first time succeeded in separating 99m Tc from a MoO 3 sample irradiated with accelerator neutrons free from any radioactive impurities and in formulating 99m Tc-methylene diphosphonate ( 99m Tc-MDP). 99 Mo, the mother nuclide of 99m Tc, was produced by the 100 Mo( n ,2 n ) 99 Mo reaction using about 14 MeV neutrons provided by the 3 H( d , n ) 4 He reaction at the Fusion Neutronics Source of Japan Atomic Energy Agency. The 99m Tc was separated from 99 Mo by sublimation and its radionuclide purity was confirmed to be higher than 99.99% by γ-spectroscopy. The labeling efficiency of 99m Tc-MDP was shown to be higher than 99% by thin-layer chromatography. These values exceed the United States Pharmacopeia requirements for a fission product, 99 Mo. Consequently, a 99m Tc radiopharmaceutical preparation formed by using the mentioned 99 Mo can be a promising substitute for the fission product 99 Mo, which is currently produced using a highly enriched uranium target in aging research reactors. A ...

Test Production of Tritium in 3 TBq Level from Neutron-Irradiated 6Li-AL Alloy Targets

Equipments were constructed for the development of tritium production technology. Gaseous tritium in a level of 3 TBq (∼80 Ci) was extracted from neutron-irradiated 6Li-Al alloy by heating at 1,073 K under vacuum and collected in an activated-uranium getter. The recovery of the tritium was 90% or more and the isotopic purity of the product was about 50%. Throughout the production experiment, no release of tritium out of the equipments to the environment was observed.

Trapping and recovery of radioiodine compounds by copper metal

Abstract The performance of Cu metal as an adsorbent for volatile radioiodine compounds, I2∗, HI∗ and CH3I∗ has been examined. Both I2∗ and HI∗ are effectively trapped. However, for the complete trapping of organic iodide, a process converting it into inorganic form must precede the trapping process. The iodine, trapped on the reduced Cu, is easily released as HI∗ by stream of H2 gas above 400°C. Compared to other conventional procedures, this trapping procedure has the advantage that Cu is regenerated during the recovery process to be used again.

Enrichment of tritium by a combined gaschromatography—isotopic decomposition process

Abstract A novel method for tritium enrichment, based on a combined process of gaschromatographic separation of H 2 , HT and T 2 with MnCl 2 -coated alumina and decomposition of HT to H 2 and T 2 on U turnings, has been studied. A mixture of H 2 , HT and T 2 was separated with an alumina column dipped in liquid N 2 , after which the HT fraction was passed over U turnings heated at 1073 K to be decomposed according to the equilibrium reaction of 2HT⇌H 2 + T 2 . The decomposed gas was again subjected to gaschromatography. This cycle process was repeated several times. After preliminary experiments with HD mixed gas, tritium was successfully recovered as T 2 with a yield higher than 80% from HT mixed gas including only 5.0 × 10 −2 atom% of tritium.

Preliminary study on sublimation separation of 99Mo from neutron- irradiated UO2

Abstract For the purpose of developing the method of separating 99 Mo from irradiated UO 2 by sublimation, fundamental studies have been made on the temperature dependencies of sublimation of 99 Mo, 132 Te and 103 Ru and also on the behavior of 131 I, which is an important nuclidee because of the hazardness. The irradiated UO 2 is converted to U 3 O 8 by heating at about 500°C in an oxygen atmosphere of 1 2 kg/cm 2 and then the nuclides are separated from U 3 O 8 by heating at 1200°C under vacuum. The sublimated 131 I is identified to be elemental iodine and iodide, and it is trapped on copper metal grains. The trapped iodine is recovered as HI by heating in H 2 gas at 550°C; it is radiochemically pure.

Continuous enrichment of hydrogen isotopes by two-dimensional gas chromatography

Abstract A two-dimensional gas chromatogrpahy was proposed for continuous purification of a large quantity of tritium gas. With a rotating circular colunn, H 2 and D 2 were mutually separated under the following conditions; flow rates of the feed gas, a mixture of H 2 and D 2 (1:1) and He carrier gas, were 100 mL/min and 5 L/min, respectively, the column rotation rate 6 π rad/h, and the column temperature 77K. The peak positions and the widths at half height calculated for the circular column based on the results with one-dimensional gas chromatograms were in fair agreement with those obtained from the experiments with the circular column. Applicability of the column to the enrichment of tritium is shown by computation of the peak positions and resolutions of H 2 , HT and T 2 .

Measurement of the Effective Neutron Capture Cross Section of Cesium-134 by Triple Neutron Capture Reaction Method

The measurement of effective neutron capture cross section( ) of 134Cs was carried out with a method using a triple neutron capture reaction,133Cs(n, γ)134Cs(n,γ)135Cs(n, γ)136Cs. The target used for the experiment was natural cesium(133Cs), and was irradiated for 23 days and 17 hours in the reactor JRR-3 at Japan Atomic Energy Research fnstitute(JAERI). The neutron field of the reactor was monitored by Au/Al and Co/Al alloy wires. A high purity Ge detector was employed for the measurement of γ-rays from the irradiated target and monitor wires. Gamma-rays emitted in the decay of 134Cs and 136Cs nuclides induced by single and triple neutron capture reactions of 133Cs were seen at the energies of 563, 569. 605, 796, 802, 819, 1,039, 1,048, 1,168 and 1,235 keV. Decays of the 819 and the 1,048 keV γ-ray intensities were followed; it was confirmed that these γ-rays originated in the 136Ba, the decay product of 136Cs. Data obtained were analyzed with values of neutron capture cross sections of 133Cs and 135Cs. ...

Separation of 99Mo from neutron-irradiated UO2 by vacuum sublimation: Apparatus and recovery of 99Mo

Abstract With a view of constructing an apparatus for separating 99Mo on a practical and commercial scale from irradiated UO2 by sublimation, studies have been made to decide on the construction of each part of the reaction vessel and also on the recovery of 99Mo with high yields. Fused quartz is chosen as the material for the reaction vessel because of its resistance against high temperatures and thermal shock. Platinum is a suitable material for the 99Mo collector, because 99Mo deposited on the Pt surface is easily dissolved in NH4OH. The apparatus constructed as a trial proved to be satisfactory. Molybdenum-99 is successfully recovered from the irradiated UO2 with a yield > 80% in a simple operation.

Separation of Molybdenum-99 from Neutron-Irradiated Uranium-235 with Sulfur as Collector

Abstract The separation of 99Mo from other fission products has been successfully attempted by isolating S in a HNO3 solution of neutron-irradiated UO2 and collecting 99Mo on it. Influence of various factors on the separation of 99Mo was examined. When 5% aqueous solution of K2S was added to the solution of fission products of high HNO3 concentration and the mixture of S stood for 10 min, over 80% of the 99Mo could be separated with the precipitate. Besides 99Mo, 132Te, 103Ru and other nuclides also accompanied the precipitate. The presence of Mo carrier and U affected greatly the separation of 99Mo, but that of Zr or Te carrier did not.