Malate formation in guard cells of Vicia faba leaves is enhanced by light. The action spectrum for this effect was determined for epidermal strips of Vicia faba, and two different spectra were obtained under different light conditions and with and without background irradiation with high-irradiance red light (>600 nm, 3.0 mW cm(-2)) superimposed on monochromatic light of other wavelengths. The spectrum obtained at quantum fluxes of 1.7-2.2 nE cm(-2) s(-1) of monochromatic light without background red light showed a sharp peak at 433 nm with a shoulder around 475 nm and a lower peak at 670-680 nm; the spectrum obtained at much lower quantum fluxes of 0.05-0.07 nE cm(-2) s(-1) of monochromatic light with red-light background had two peaks of comoparable heights at 380 and 460 nm. The formation of malate with 430-nm blue light was saturated at a quantum flux of 3 nE cm(-2) s(-1) without the background red light but at a much lower quantum flux of 0.2 nE cm(-2) s(-1) with the background red light. At this low intensity, blue light was practically ineffective without background red light. A synergistic action of red light presumably absorbed by the chlorophylls, and blue light absorbed by a yellow pigment is thus demonstrated by these experiments. The action maxima at 380 and 460 nm for the blue-light effect in the presence of background red light agree with the absorption maxima of flavins.
An accelerated experiment using the actinide doping technique was performed to investigate the effects of alpha decay on the properties of actual nuclear waste glass at high radiation doses. A fully radioactive borosilicate waste glass, containing the actual high-level radioactive liquid waste generated from the Tokai Reprocessing Plant of PNC, was prepared by JAERI, and a powder mixture of the ground fully radioactive glass and 244CmO2 was melted at 1200°C for 2 hrs. The radioactivity concentration of 244Cm was 1.0 } 1010 Bq/g-glass at the date of preparation. The homogeneity of curium-doped glass samples was confirmed by the density measurement, heat load measurement and alpha autoradiography. The properties of the irradiated samples were investigated by the mass spectrometer for helium determination, the optical microscope, the electron probe microanalyzer, the densitometer, the Soxhlet and MCC-1 leach testing apparatus. By measuring the amount of helium released from the curium-doped glass samples, more than 99% of helium remained in the matrix at room temperature. The density of the sample slightly decreased with the increase of cumulative alpha decays and the decrease of 0.77% was observed at a dose of 1.55 } 1018 alpha decays/g, corresponding to an equivalent age of 150000 yrs. Optical and scanning electron micrographs showed that no cracks were observed on all samples having up to a dose of 1.5 } 1018 alpha decays/g. The leach rates, based on weight loss, in both the Soxhlet (100°C, 7 days) and MCC-1 (90°C, 28 days) tests did not significantly change with alpha decay dose.
Carbon steel is one of the candidate materials for overpacks for high-level radioactive waste disposal in Japan. Passivation behavior and corrosion rate of carbon steel were investigated by electrochemical measurements under simulated repository conditions. The results of the anodic polarization measurements showed that carbon steel was hard to passivate in highly compacted bentonite. Therefore, general corrosion seems to be most probable in repository conditions. In order to monitor the in-situ general corrosion rate in compacted bentonite, the AC impedance of carbon steel was measured under aerated conditions. It was confirmed that the corrosion rate in saturated bentonite decreased with time and it was almost the same as that observed in deaerated aqueous conditions. The corrosion rate did not increase in the presence of corrosion products formed by external current supply.
The corrosion rate of carbon steel in compacted bentonite was evaluated with regard to the test period length, temperature, chemicals of solution and bulk density of compacted bentonite.The average corrosion rate decreased gradually with increasing test period up to 180 days in immersion tests. The corrosion rate of carbon steel in compacted bentonite at a dry density of 1.32g/c was estimated to be about 0.01 mm/y which was one order of magnitude lower than that in bentonite slurry. No significant influence of temperature on corrosion rates was observed in compacted bentonite in the range of 50-180 *C . Variation of kinds and concentration of anion(chloride, floride, sulfate, and carbonate)in aqueous solution did not have much influence on the corrosion rate of carbon steel.Immersion tests of carbon steel in compacted bentonite at a dry density of 0.69 -1.32 g/cm 3 , which was mixed with an aqueous solution(synthetic sea water and distilled water), were carried out. The corrosion rate in compacted bentonite decreased from 0.04 to O.O05mm/y as the density of bentonite increased.This result suggests that the corrosion rate of carbon steel in compacted bentonite is governed by the diffusivity of corrosive materials. In general, oxygen is the dominate factor affecting corrosion rate, therefore prediction of the average corrosion rate of carbon steel was carried out on the basis of the diffusion behavior of dissolved oxygen in compacted bentonite.The prediction agreed with experimental results. INTRODUCTIONThe Japanese concept for high-level radioactive waste management is based on immobilization with borosilicate glass, folllowed by 30 to 50 years of storage for cooling, and ultimate disposal in underground formations deeper than several hundred meters (I ) . Candidate sites have not been selected yet, and site characteristics such as composition of groundwater have not been determined.According to the current engineered barrier system concept, waste glass packed in the waste container(overpack) is emplaced in a borehole. Buffer materials are emplaced between the overpack and host rock. These engineered barrier systems act as a primary barrier to the release of radionuclies from the radioactive waste to host rock.The primary role of the overpack is to confine the radionuclides for more than several hundred years. The life of the overpack will be mainly limited by corrosion Two types of materials were sellected as candidate overpack materials for achieving this long life. One is corrosion allowance type material and the other is high corrosion resistance type material. Carbon steel is the primary candidate of corrosion allowance type metal.Immersion tests in test solution have been conducted in order to estimate the influence of individual environmental factors on the corrosion behavior of carbon steel. Immersion tests in bentonite slurry have also been carried out to estimate
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