Radiochemical separation methods have been applied for the neutron activation analysis of impurities in four high-purity refractory metals, Ta, Nb, W and Mo. Impurities in the metals of Ta, Nb and W can be separated into groups using anion exchange resin with HF and/or a mixture of HF acid and HCl, but those in Mo is done using both anion and cation exchange resins. The coprecipitation of U with Th in HF media is also investigated.
In this work, an easy, fast and reliable measurement technique for the quantitative determination of retained fission gases
in an irradiated oxide fuel was developed. Many experiments were conducted to determine the optimum conditions for fusion
of an oxide fuel, for the quantitative collection and measurements of the released gases. Ion implantation technology was
applied to make a krypton or xenon references in a solid matrix. A fragment of oxide fuel, about 0.1 g of an unirradiated
SIMFUEL, was completely fused with excess metallic fluxes, 1.0 g of nickel and 1.0 g of tin, in a graphite crucible of a helium
atmosphere for 120 s at 850 A as a mixture of metals and alloys. About 96 ± 3 to 98 ± 4% of the krypton and xenon that were
injected into the instrument using a standard gas mixture was reproducibly recovered by collecting the releasing gas through
the instrument for 120 s. Using the same fusion and collection conditions, it was possible to recover about 97 ± 3% of the
injected krypton and xenon by fusing a fragment of SIMFUEL which was wrapped with krypton or xenon implanted aluminum foils.
The recovery test results of krypton and xenon using ion planted aluminum foils gave encouraging results suggesting their
potential use as a reference specimen. It was confirmed that a fragment of irradiated oxide fuel, 0.051 g, with a code burn-up
of 56.9 MWd/MtU, was completely fused as the mixture of metals and alloys through the fusion conditions and more than 99%
of the retained fission gases were recovered during the first fusion. Since no cryogenic trap was needed, the collected gas
could be measured directly and thus the analysis time could be further reduced. Approximately 7 min was sufficient to finish
the measurement of retained fission gases in the irradiated oxide fuel using the developed procedure.
The relative ans single comparator methods have been applied to determine 7 rare-earth elements and U, Th in Korean Monazites by 14.5 MeV neutron activation analysis. The (n, 2n) nuclear reactions are used for all elements except La, for which (n, p) reaction is used. Al is used as a flux monitor for the relative method and as a singlle comparator for the single comparator method. The analytical results obtained by the two methods agree well within 3% deviation except for Sm and Gd. These results are also compared with the result obtained by a single comparator method using reactor neutron.
The determination of palladium, platinum and rhodium in industrial concentrates such as lead foam and raw lead by neutron
activation analysis is described. The noble elements are separated from the matrix by spontaneous deposition on amalgamated
copper powder prior to activation. After the determination of palladium and platinum, rhodium is coprecipitated on iron hydroxide,
and the precipitate irradiated for the determination of rhodium. The results are compared with those obtained by fire assay.
A systematic study has been made on the reducing power of amalgamated copper powder in hydrochloric acid solution for palladium,
platinum, rhodium, iridium, gold and silver. In order to apply this method to the activation analysis of palladium, platinum
and rhodium in industrial concentrates which contain a large amount of ‘base elements’, the behaviour of palladium, platinum
and rhodium in the presence of the ‘base elements’ has also to be considered.
New165Dy and166Ho macroaggregates(165Dy-MA,166Ho-MA) were prepared by reacting the aqueous solution of165Dy(NO3)3 and166Ho(NO3)3, respectively, with sodium borohydride solution in 0.2N NaOH.165Dy-MA and166Ho-MA are sterile suspension of macroaggregates consisted of Dy/Ho (50–56 wt.%) and boron (5–7%) in saline with mean particle size of 2.6 m (1–8 m range). Both MA could be prepared from the pre-made164Dy-MA and165Ho-MA followed by neutron irradiation. Even though the165Dy-MA and166Ho-MA suspension in saline were stored at 37°C for 24 hrs (Dy-MA), 10 days (Ho-MA) or autoclaved at 121°C for 30 minutes, there was no significant change in particle size and no leakage problem indicating the prepared165Dy-MA and166Ho-MA are sufficiently stable. The results of in-vivo retention studies using rabbits showed high retention (>99.5%) in knee joint even at 24 hrs (165Dy-MA) or at 10 days (166Ho-MA) after administration. Rabbits treated with intra-articular injections of164Dy-MA or165Ho-MA equivalent to 20–30 times the typical clinical dose showed no signs of any toxic effects at 1 month after administration. The ease with which the165Dy-MA and166Ho-MA can be made in the narrow size range and their high in-vitro and in-vivo stability make them attractive agents for radiation synovectomy.
The Korea Atomic Energy Research Institute (KAERI) completed the High-flux Advanced Neutron Application Reactor (HANARO) in 1995 and the radioisotope production facilities(RIPF) in 1997. Many devices and handling tools were developed and applied for the production of radioisotopes. Emphasis on RI production plan was placed on the development of new radiopharmaceuticals, the development of new radiation sources for industrial use and the steady production of selected radioisotopes. The selected items are 166Ho-based pharmaceuticals, fission 99Mo/99mTc generators, and products of 131I and 192Ir and 60Co sources for industrial use. Now KAERI regularly produces radioisotopes (131I, 99mTc, 166Ho, 192Ir, 60Co etc.) and labeled compounds including 99mTc cold kits. Newly developed therapeutic agents are a 166Ho-chitosan complex for liver cancer treatment, a 166Ho patch for skin cancer treatment and devices such as the stent and balloon for the prevention against restenosis of the coronary artery. Feasibility studies on the installation of a 99mTc generator loading facility and on 60Co production for food irradiation were finished. The 192Ir sealed source assembly for NDT has been supplied to domestic users since May 2001. The fission moly process, separation process of non-sealed sources (125I, 33P, 89Sr, 153Sm, 188Re) and fabrication process of sealed sources (169Yb, 75Se) are also under development. For the quality assurance of our final products, we obtained ISO certification in 2000. We are carrying out a feasibility study on a new research reactor for the stable supply of radioisotopes in Korea.
A neutron induced prompt γ -ray spectrometry (NIPS) facility has been developed at the Nuclear Chemistry Research Division,
of the Korea Atomic Energy Research Institute (KAERI) with the aim of analyzing the major components of various elements in
aqueous samples. The facility is equipped with a 252Cf neutron source and a γ-γ coincidence setup with two n-type coaxial HPGe detectors based on NIM spectrometric modules in
association with data acquisition and spectral analysis systems. The development of the system, its set-up and the calibration
of detection efficiency up to 8 MeV using a set of radionuclides and the (n,γ) reactions of chlorine are described in the
In this work the analysis procedures of fission gas compositions and their isotopic distributions using a gas chromatography
(GC) system and/or a quadrupole mass spectrometer (QMS) system were established, and their analysis results were reviewed
in order to evaluate their analytical performance. Also, the accumulated data, up to now, regarding fission gas measurement
were reviewed to discern any irradiation histories of the punctured fuel rods. A simple gas injection apparatus was designed
and fabricated for the quantitative injection of a small volume of fission gas into the GC and the QMS system. With an appropriate
temperature controlling of a molecular sieve 5A column, nitrogen, krypton and xenon of a mixture gas was clearly separated
within 7 min. According to the analysis results, the relative standard deviation in the determination of fission gas compositions,
krypton and xenon, by the GC analysis or by the QMS analysis was about 1%. Based on the review results of the isotopic ratios
of krypton and xenon of the released fission gas, it is likely that no abnormally irradiated rods, i.e. defected rods, were
included among the punctured rods.
It is regarded that the spent resins from the water purification systems of moderator (MOD) and the primary coolant of the
Canada deuterium uranium-pressurized heavy water reactor (CANDU-PHWR) are a unique waste, owing to their high 14C and gamma-emitting nuclides. In this work, 14C and 3H contents, anion and cation fractions and the predominant gamma-emitting nuclides of the spent resins from 4 units of CANDUPHWRs,
were investigated. Also the chemical species of 14C of the spent resins were determined. For a simultaneous separation of 14C and 3H from the spent resins, the wet oxidation-16 wt% H2SO4 stripping process was utilized. The 14C and 3H activity concentration range of the spent resins of the nuclear power plant (NPP), 4 units of all CANDU-PHWR types, was
2.48E5 Bq/g ∼5.33E6 Bq/g, 1.29E5 Bq/g and ∼2.33E5 Bq/g, respectively. Among the analyzed spent resins, the highest 14C and 3H activity concentration was detected in units 4 and 3, respectively. It was found that more than 92% of the 14C activity concentration was retained on the anion resin and the predominant chemical species was inorganic 14C. It was revealed that the anion resin fraction of the spent resins from unit 1 and unit 2, was about 40% and that of unit
3 and unit 4 was around 60%. More than 80% of the total gamma-radioactivity concentration was associated with the cation fraction
of the spent resin. The predominant gamma-emitting nuclide of the spent resin for unit 2 was 137Cs, a fission product, and that for unit 4 was 60Co, a corrosion product.