Susumu Muraoka

Effects of humic acid on the sorption of neptunium(V) on kaolinite

Abstract The sorption coefficient of Np(V), K d , on kaolinite was measured at pH 6 to 11 and in humic acid concentrations of 0 to 40 mg dm −3 at the ionic strength of 0.1 M. The K d value increased with pH both with and without humic acid. The K d value increased slightly with increasing concentration of humic acid in the pH range below 8 and decreased not more than an order of magnitude with increasing concentrations of humic acid in the pH region above 8. Below pH 8, Np(V) sorption is considered to be enhanced by the sorption of humic acid on the kaolinite to form Np(V)-humate complexes. Above 8, it is probable that the desorption of humic acid and formation of Np(V)-humate in solution result in the decrease of K d . The behavior of Np(V) sorption on kaolinite with humic acid is described by a simple model.

Development of ceramic waste forms for actinide-rich waste : Radiation stability of perovskite and phase and chemical stabilities of Zr- and Al-based ceramics

Abstract Ceramics are considered as most promising materials for conditioning of long-lived radionuclides because of their outstanding durability for long term. The Japan Atomic Energy Research Institute (JAERI) has developed ceramic waste forms, e.g. Synroc and zirconia-based ceramics, for the actinide-rich wastes arising from partitioning and transmutation processes. In the present study, α-decay damage effects on the density and leaching behavior of perovskite (one of three main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reached 1.3 % at a dose of 9 × 10 17 α-decays·g −1 . The leach rates (MCC-1 leach test inpH ∼ 2 solution at 90°C for 2 months) of perovskite specimens with accumulated doses of 1.6 × 10 17 , 4.0 × 10 17 and 8.3 × 10 17 α-decays·g −1 were 1.7, 2.3 and 3.0 μ·m −2 ·day −1 , respectively. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by the experiments using non-radioactive elements (Ce and Nd) with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia was stable crystallographically in the wide ranges of Ce and/ or Nd content and had excellent chemical durability.

Reaction of several iron and nickel based alloys with sintered Li2O Pellets

Abstract The reaction of type 316 stainless steel, Incoloy 800, Hastelloy X-R, Inconel 600 and pure Ni with sintered Li 2 O pellets has been studied between 800 and 1100°C under dynamic vacuum. The reaction products were analyzed by means of metallographic, microprobe and X-ray diffraction methods. The reactions proceeded measurably between 800 and 950°C and appreciably at 1000°C, being greatest with Incoloy 800 and least with Hastelloy X-R. Among the primary alloy constituents, chromium was exclusively attacked by lithium and oxygen diffusing from the Li 2 O into the alloys to form LiCrO 2 . This phase grew into a reaction zone (subscale) of uniform thickness beneath the surface of each alloy. Preferential growth of LiCrO 2 along the grain boundaries was observed only in the case of Inconel 600 below 950°C. On the other hand, iron diffused toward the Li 2 O pellets to form volatile Li 5 FeO 4 . However, any reaction product associated with Ni was not detected and Ni metal was little attacked by the Li 2 O pellet over the whole range of reaction temperature.

Numerical Model of Radionuclide Migration in Geologic Media

A three-dimensional radionuclide migration model has been developed by using of the direct-simulation method. The phenomena taken into account are radioactive decay, convection and dispersion in the ground water and sorption and desorption in the geologic media. Decay chain is represented by particles character change using decay probability. Smoothing method is proposed to make an even distribution of particles and to obtain the exact radioactive inventory. Numerical calculations of 245Cm decay chain and 234U decay chain in the single layered geologic media have been carried out, and reasonable results have been obtained in comparison with INTRACOIN study. Nuclide migration of 237Np decay chain in the three-layered geologic media were examined and shown graphically.

Carburization of Hastelloy alloy X

Abstract The carburization of Hastelloy alloy X has been studied by the standard tracer techniques using 14 C over the temperature range 700 to 1100°C in vacuo and helium environment. The 14 C activity in the specimen decreases exponentially with the depth of penetration and increases with carburization time. It is observed that carbon diffusion in the alloy is dominated by grain-boundary diffusion. The effect of oxygen in helium gas on carburization is observed. Oxygen has two opposite types of action on carburization of the alloy. One is inhibitive action and the other is accelerative action.

Reaction of molybdenum and molybdenum-base alloy with sintered Li2O pellets

Abstract The reaction of sintered lithium oxide (Li 2 O) pellets with molybdenum and molybdenum-base alloy TZM has been studied in the temperature range 800–1100 °C for a constant reaction period of 100 h under a dynamic vacuum. The reaction proceeded measurably at about 950°C and appreciably above 1000°C. Metallographic and X-ray diffraction analyses indicated that a reaction product scale was formed at the interface and the scale consisted mainly of Li 4 MoO 5 . TZM showed slightly higher reactivity compared to molybdenum.

Leaching behavior of carbon-14 contained in portland cement

Abstract MCC-1 static leaching experiments were carried out for a cementitious waste form in distilled water for up to 64 days at 5°C and 20°C in order to examine the leaching behavior of carbon-14. The complicated leaching behavior of carbon-14, meaning that the leached carbon-14 activity did not increase with (time)0.5, was attributable to the precipitation of calcite and the formation of colloidal particles in leachates, which were mainly dependent on the pH value and calcium concentration of leachate. The normalized elemental mass loss of carbon-14 was about 7.5 × 10−4 g/cm2 at 20°C for 64 days, which was lower than those of cement constitute elements such as calcium, sodium and aluminum. Especially, the leach rate of aqueous carbon-14 was lower than that of carbon-14 in the suspended leachate by a factor of about 10.

Three Year Experimental Study of Leaching Phenomena from Low Level Radioactive Cement-Based Waste Forms

Abstract This study was carried out in order to demonstrate the safety of homogeneous cementbased waste forms (hereinafter called cement forms) for BWRs low level radioactive wastes as engineered barriers. Eighteen full scale simulated cement forms were manufactured with the addition of 137Cs, 66Co and 90Sr. Leaching tests on these forms were carried out for approximately three years. In order to study the relationship of leachability to environments at disposal sites, this Three Year Leaching Test was conducted for three kinds of environmental conditions, sea water, land water and soil. After the tests, all of these forms were cut to measure the distribution of the radionuclides density within them. In case of the soil tests, the distribution of radionuclide in the soil was also measured. The radionuclide leachability results reveal that 60Co was almost completely retained in the cement forms and that 137Cs leached from cement forms was mostly adsorbed by the soil. On the other hand, 90Sr was not trapp...

Corrosion of stainless steel for HLW containers under gamma irradiation

The corrosion behavior of type 304 stainless steel was studied under gamma irradiation as part of the evaluation for the long-term durability of high-level radioactive waste (HLW) disposal containers. Gamma rays, generated from fission products in high-level radioactive waste, are considered to change the environment around the canisters and overpacks. The redox potentials for NaCl solutions and corrosion potentials of stainless steel were measured to consider the effects of gamma irradiation, by using an electrochemical method. The pitting potentials of stainless steel for NaCl solutions were also measured to examine the pitting corrosion under gamma irradiation. As a result of this experiment, it is concluded that the oxidizing properties as a result of the formation of H{sub 2}O{sub 2} and H{sub 2} produced by gamma irradiation depended on the concentration of Cl{sup -}, and that the strength of oxidizing properties of 1M (mol{center_dot}dm{sup -3}) NaCl solution was particularly high, and the pitting corrosion as found for 1M NaCl solution under gamma irradiation at the dose rate of 2.6{times}10{sup 2} C/kg{center_dot}h (1.0{times}10{sup 6} R/h) at 60{degrees}C, by using an electrochemical method.

Effects of some glass additives on nuclear waste glass durability in water

Borosilicate glass is a candidate material for immobilizing high-level waste (HLW) stemming from the reprocessing of spent fuels used for nuclear power generation. The borosilicate glass containing HLW, which is called HLW glass, will be stored on the land surface for several tens of years and then disposed of in deep geologic formations. The contact of HLW glass with ground water in deep geologic disposal sites is the primary concern of this paper, because it is a first step for radionuclides in the glass to be released into the biosphere. The durability of HLW glass in ground water is fairly good for the purpose of safe disposal. However, if the leach rate of the HLW glass is reduced by four orders of magnitude to 10 .9 g cm -2 day -1, the glass can hold its integrity even after a considerable amount of 237Np with a half-life of 2.14 × 106 years decays in the glass. In this study, we try to improve the durability of the glass as much as possible. For this purpose, the effects of some glass additives on nuclear waste glass durability in water are examined. For example, Fe203 is used as a glass additive, since it has been