Isao Yamagishi

Radionuclide release to stagnant water in the Fukushima-1 nuclear power plant1

After the severe accident at the Fukushima-1 nuclear power plant, large amounts of contaminated stagnant water have accumulated in turbine buildings and their surroundings. This rapid communication reports calculation of the radionuclide inventory in the core, collection of measured inventory in the stagnant water, and estimation of radionuclide release ratios from the core to the stagnant water. This evaluation is based on data obtained before 3 June 2011. The release ratios of tritium, iodine, and cesium were several tens of percent, whereas those of strontium and barium were smaller by one or two orders of magnitude. The release ratios in the Fukushima accident were equivalent to those in the accident of the Three Mile Island, Unit 2 (TMI-2).

Hydrogen production in gamma radiolysis of the mixture of mordenite and seawater

Hydrogen production by γ-radiolysis of the mixture of mordenite, a zeolite mineral, and seawater was studied in order to provide basic points of view for the influences of zeolite minerals, of the salts in seawater, and of rise in temperature on the hydrogen production by the radiolysis of water. These influences are required to be considered in the evaluation of the hydrogen production from residual water in the waste zeolite adsorbents generated in Fukushima Dai-ichi Nuclear Power Station. As the influence of the mordenite, an additional production of hydrogen besides the hydrogen production by the radiolysis of water was observed. The additional hydrogen can be interpreted as the hydrogen production induced by the absorbed energy of the mordenite at the yield of 2.3×10−8 mol/J. The influence of the salts was observed as increase of the hydrogen production. The influence of the salts can be attributed to the reactions of bromide and chloride ions inhibiting the reaction of hydrogen with hydroxyl radical. The influence of the rise in temperature was not significantly observed up to 60°C in the mixture with seawater. The results show that the additional production of hydrogen due to the mordenite had little temperature dependence.

Characterization and storage of radioactive zeolite waste

For the safe storage of zeolite wastes generated by the treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, this study investigated the fundamental properties of herschelite adsorbent and evaluated its adsorption vessel for hydrogen production and corrosion. The hydrogen produced by the herschelite sample is oxidized by radicals as it diffuses to the water surface and thus depends on the samples water level and dissolved species. The hydrogen production rate of herschelite submerged in seawater or pure water may be evaluated by accounting for the water depth. From the obtained fundamental properties, the hydrogen concentration of a reference vessel (decay heat = 504 W) with or without residual pure water was evaluated by thermal–hydraulic analysis. The maximum hydrogen concentration was below the lower explosive limit (4%). The steady-state corrosion potential of a stainless steel 316L increased with the absorbed dose rate, but the increase was repressed in the presence of herschelite. The temperature and absorbed dose at the bottom of the 504 W vessel were determined as 60 °C and 750 Gy/h, respectively. Under these conditions, localized corrosion of a herschelite-contacted 316L vessel would not immediately occur at Cl− concentrations of 20,000 ppm.

Advanced ORIENT Cycle: Progress on Fission Product Separation and Utilization

Fission reaction of U-235 and/or plutonium generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 elements are categorized as rare metals. In a conventional fuel cycle U and Pu are reused but others are vitrified for disposal. Adv.-ORIENT (Adv anced O ptimization by R ecycling I nstructive E lement s) Cycle strategy was drawn up for the minimization of radio-toxicity and volume of radioactive waste as well as the utilization of valuable elements/nuclides in the waste. The present paper describes the progress on Fission Products (FP) separation in this Cycle. Highly functional inorganic adsorbent (AMP-SG, silica gel loaded with ammonium molybdophosphate) and organic microcapsule (CE-ALG, alginate gel polymer enclosed with crown ether D18C6) were developed for separation of heat-generating Cs and Sr nuclides, respectively. The AMP-SG adsorbed more than 99% of Cs selectively from a simulated High-level Liquid Waste (HLLW). The ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its selectivity; Ba > Sr > Pd >> Ru > Rb > Ag. The electrodeposition is advantageous for both recovery and utilization of PGMs (Ru, Rh, Pd) and Tc because PGMs are recovered as metal on Pt electrode. Among PGMs, Pd was easily deposited on the Pt electrode. In the presence of Pd or Rh the reduction of Ru and Tc was accelerated more in hydrochloric acid media than in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition of elements except for Pd. The deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen production than the original Pt electrode. The reactivity of deposits prepared from the simulated HLLW was higher than that from solution containing only PGM.Copyright

Selective Separation and Recovery of Cesium by Ammonium Tungstophosphate-Alginate Microcapsules

A fine crystalline ammonium tungstophosphate (AWP) exchanger with high selectivity toward Cs+ was encapsulated in biopolymer matrices (calcium alginate, CaALG). The characterization of the AWP-CaALG microcapsule was examined using SEM/WDS, IR and DTA/TG analyses, and the selective separation and recovery of 137 Cs were examined by the batch and column methods using simulated and real high-level liquid waste (HLLW). The free energy (ΔG0 ) of the ion exchange (NH4+ ↔ Cs+ ) for fine AWP crystals was determined at −13.2 kJ/mol, indicating the high selectivity of AWP towards Cs+ . Spherical and elastic AWP-CaALG microcapsules (∼700 μm in diameter) were obtained and fine AWP crystals were uniformly immobilized in alginate matrices. Relatively large Kd values of Cs+ above 105 cm3 /g were obtained in the presence of 10−3 ∼1 M Ca(NO3 )2 , resulting in a separation factor of Cs/Rb exceeding 102 . The irradiated samples (60 Co, 17.6 kGy) also exhibited large Kd values exceeding 105 cm3 /g in the presence of 2.5 M HNO3 . The Kd values in the presence of 0.1–9 M HNO3 for 67 elements were determined and the order of Kd value was Cs+ ≫ Rb+ > Ag+ . The breakthrough curve of Cs+ had an S-shaped profile, and the breakpoint increased with decreasing flow rate; the breakpoint and breakthrough capacity at a flow rate of 0.35 cm3 /min for the column (0.7 g AWP-CaALG) were estimated at 25.2 cm3 and 0.068 mmol/g, respectively. Good breakthrough and elution properties were retained even after thrice-repeated runs. The uptake (%) of Cs+ in simulated HLLW (28 metal components-1.92 M HNO3 , SW-11, JAEA) was estimated at 97%, and the distribution of Cs+ and Zr/Ru into the AWP and alginate phases, respectively, were observed by WDS analysis. Further, the selective uptake of 137 Cs exceeding 99% was confirmed by using real HLLW (FBR “JOYO”, JAEA). The AWP-CaALG microcapsules are thus effective for the selective separation and recovery of Cs+ from HLLWs.Copyright