Eiji Shikata

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.

Dry-Distillation of Iodine-131 from Several Tellurium Compounds

Several Te compounds including a new Te oxide Te5O11 were subjected to dry-distillation after reactor irradiation in order to overcome the difficulties accompanying the ordinary dry-distillation from TeO2. The release of 131I from Te5O11 was examined as function of distillation temperature and the thermal behavior of 131I released was studied by means of thermochromatography. For comparison, the 131I release from other Te compounds such as H5TeO6, (H2TeO4) n , and TeO3 was also studied. The compounds studied in this work were found to release 131I rapidly at temperatures lower than employed in the ordinary dry-distillation of 131I from TeO2, thus obviating all risk of contaminating the final products with Te vaporizing at higher temperatures. The rapid release of 131I was concluded to be caused by thermal decomposition of the compounds studied. In an attempt to prepare a “dry source” of 131I, 131I dry-distilled from Te5O11 was adsorbed on Molecular Sieve, and the release of 131I from the Molecular Sieve w...

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.

Research of Radioisotope Production with Fast Neutrons, (VI):Preparation of Cu-67

Copper-67, the longest-lived radioisotope of copper, was prepared by using the reaction 67Zn(n p,)67Cu. Zinc enriched in 67Zn up to 93.4% was irradiated in JRR-1 reactor for 5hr with a fast neutron flux of 1.5×1012n/cm2/sec. Copper-67 formed in the zinc target was separated by anion exchange separation using the chloro-complex of the metals. The yield of 67Cu and radiochemical impurity were investigated. Seventeen millimicrocuries of 67Cu was obtained from 1mg of zinc. No radioactive impurities other than 64Cu were not found. Even with the use of enriched target, a more MCu was produced than 67Cu. The proportion of 67Cu to total copper radioactivity was 26% immediately after the irradiation.

Release of Carbon-14 from Neutron-Irradiated Aluminum Nitride in the Dry Procedure

The dry procedure was applied to the separation of 14C from either finely powdered A1N or A1N in tablet form after neutron irradiation. The mechanism of the 14C release from the irradiated materials was studied, and the possibility of using A1N target in tablet form for routine production of 14C was examined. Powdery AlN, sintered powder tablet, molded powder tablet, and sintered mold tablet were irradiated for 12 days in the JRR-2 reactor. After irradiation, 14C was separated in the form either of as dioxide or of monoxide, from the irradiated material at various temperatures under oxygen stream, and the rate of combustion of A1N and that of the 14C release were determined as a function of temperature and reaction time. The release of 14C from neutron irradiated A1N was found to accompany the oxidation of A1N which took place in the form of tarnishing reaction. Above 1,090°C, the release of 14C obeyed the logarithmic rule, indicating that it is governed by the migration of 14C in the oxide layer. The che...

Studies on Fundamental Properties of Alumina Adsorbent for Radioisotope Production

Some fundamental properties of chromatographic alumina, used for the preparation of radioisotope generators and for the separation of fission products, were examined in connection with pretreatment procedures prescribed for the alumina to improve its properties in respect of alumina dissolution by certain eluants, its adsorption capacity for certain ions, and the effects of differences in the mode of storage. The commercially available brand of chromatographic alumina used in this work showed an increase in its adsorption capacity and a decrease in alumina dissolution, when it was ignited for 10 hr at 400°C and treated thereafter with boiling dilute hydrochloric acid. When treated by the above procedure, the properties of the alumina in respect of dissolution and adsorption capacity were found to be unaffected by extended periods of storage under conditions corresponding to normal operation. The pretreated alumina was made into column form. The dissolution of alumina with various kinds of eluants and the ...

Research of Radioisotope Production with Fast Neutrons, (VII):Production of Several Radioisotopes of High Specific Activity in JRR-1 Reactor

Production of radioisotopes of high specific activity was studied in the JRR-1 reactor using several (n,p) and (n,α) reactions, such as 24Mg(n,p)24Na, 27A1(n,α)24Na, 35Cl(n,p)35S, 35C1(n,α)32P, 58N...

PREPARATION OF FLUORINE-18. SEPARATION IN ALUMINA COLUMN, AND THE CHEMICAL FORM OF

Fluorine-18 was prepared by neutron irradiation of lithium carbonate in the reactor JRR-1. Five grams of lithium carbonate was irradiated for 2hr with a thermal neutron flux of 6×1011 n/cm2/sec. An...

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Abstract A program for producing high activities of tritium (of order TBq) in a nuclear reactor made it desirable to predict tritium yields conveniently, safely and with an accuracy within 10%. This paper describes how this was done by measuring the 60 Co activity from n 7 reactions with cobalt impurities in the aluminium cladding of the 6 Li-Al alloy target, from which the tritium was generated. The cobalt impurity concentration was measured by a neutron activation technique.

F OBTAINED

pharmaceuticals has increased so rapidly in the past few yearsl) that the production and the supply of 99Mo are one of the most important missions of an organization for radioisotope production. Molybdenum-99 having enough high specific activity for the medical application can be produced by nuclear fission of 235U. For the separation of 99Mo from fission products have been reported various methods such as adsorption on alumina column2), solvent extraction with di (2-ethylhexyl) phosphoric acid31 or the precipitation by a-benzoinoxime4). Those methods, however, include timeconsuming and complicated procedures or have an danger of radiation decomposition of organic materials. In this work, a new, simple and practical separation method was developed, which based on the selective precipitation of fission-produced 99Mo. Uranium dioxide having a high melting point