Tetsuo Fukasawa

GENERATION AND DECOMPOSITION BEHAVIOR OF NITROUS ACID DURING DISSOLUTION OF UO2 PELLETS BY NITRIC ACID

The generation and decomposition behavior of nitrous acid is experimentally investigated during dissolution of unirradiated uranium dioxide (UO{sub 2}) pellets by a nitric acid solution. The nitrous acid is generated by the dissolution of UO{sub 2} and it then decomposes to nitrogen oxides through the solution surface. The generation rate is equal to the dissolution rate of the uranium pellet and it depends on the nitric acid concentration, solution temperature, and effective pellet surface area. The decomposition rate depends on the solution surface area and temperature. These findings allow prediction of changes in nitrous acid concentration during and after dissolution.

Investigation of silver-impregnated alumina for removal of radioactive methyl iodide

The removal efficiency of methyl iodide for silver-impregnated alumina from gaseous waste has been experimentally evaluated as a function of atmospheric relative humidity. A new adsorbent has been developed for the iodine filter installed in the off-gas treatment system of a radioactive waste tank vent. To improve its removal efficiency under a highly humid atmosphere, the optimum average pore size of alumina was determined to be {approximately}60 nm, and the most effective chemical form of the impregnated silver was identified as silver nitrate. Holding capability of the impregnated silver was also improved by developing a double-pore-structure alumina.

A new concept for the nuclear fuel recycle system: Application of the fluoride volatility reprocessing

Abstract Hybrid Recycle System (HRS) is proposed as an advanced recycle system. The HRS consists of improved fluoride volatility reprocessing and vibration packing MOX fuel fabrication processing. For the former, a part of U is volatilized as hexafluoride with diluted F 2 gas, and then residual U and Pu are volatilized with concentrated F 2 gas. Plutonium content of the mixed fluoride gas can be adjusted as desired by controlling the U fluorination reaction in the first step. The U is highly decontaminated and the mixture gas of UF 6 and PuF 6 is not purified. The fluoride mixture is reacted with H 2 O and H 2 and directly converted to the mixed oxide grain for the vibration packing. The HRS can reduce the costs of reprocessing and fuel fabrication, the amounts of radioactive wastes and the probability of Pu proliferation.

Valence Control and Solvent Extraction of Americium in the Presence of Ammonium Phosphotungstate

Fundamental investigations on valence control and solvent extraction of americium were carried out to develop a method for americium separation from reprocessing solution. In order to adjust americium valency from III to IV and VI, (NH4)10P2W 17O61 synthesized was used as complexant stabilizing Am(IV). Oxidation behavior of americium was investigated as a function of (NH4)10P2W 17O61 americium ratio. Using 0.1M (NH4)2S2O8 and 0.01M AgNO3 as oxidation reagent, Am(IV) was obtained quantitatively at the ratio of 15. On decreasing the ratio to 0.6, 92% of americium was adjusted to Am(VI). The concentration of (NH4)2S2O8 could be reduced to 1/15 compared to the previously reported method in which no complexant was used. Americium(IV) was also prepared by reacting O3 and AgNO3 but no Am(VI) was obtained even at low (NH4)10P2W 17O61 to americium ratio. Americium(VI) could be extracted by tri-n-butyl phosphate stably without influence of (NH4)10P2W 17O61. The distribution coefficient of Am(VI) was 4 between 100% ...

Transition period fuel cycle from current to next generation reactors for Japan

Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named “Flexible Fuel Cycle Initiative (FFCI)” which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual “Recycle Material” is, according to FBRs introduction status, immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles.

Activities of Water and Nitric Acid in Simulated Reprocessing Waste Solutions

Simulations were carried out for thermodynamic activities of water and nitric acid in simulated reprocessing solutions under a high level liquid waste (HLLW) evaporation condition. A prediction method based on a “Hydration Model” was developed to evaluate the activity coefficients of nitric acid and water in nitric acid solutions containing fission products. Suitability of the “Hydration Model” was confirmed; the changes in activity coefficient were well explained. Models predictions agreed well with the reported experimental results for changes of nitric acid concentration in the HLLW solution during the evaporation concentration process. The calculated activities of water and nitric acid were constant during the concentration process, except for the initial transient stage. Therefore nitric acid concentration changes should not have significant effects on chemical reactions, such as precipitation or volatilization, in the solution during HLLW concentration.

Incineration of ion exchange resins using concentric burners

A new incineration method, using concentric burners, is studied to reduce the volume of spent ion exchange resins generated from nuclear power plants. Resins are ejected into the center of a propane-oxygen flame and burned within it. The flame length is theoretically evaluated by the diffusion-dominant model. By reforming the burner shape, flame length can be reduced by one-half. The decomposition ratio decreases with larger resin diameters due to the loss of unburned resin from the flame. A flame guide tube is adapted to increase resin holding time in the flame, which improves the decomposition ratio to over 98 wt%.

Radiolytic Oxidation Behavior of Neptunium in Sodium Chloride Solutions

Radiolytic oxidation behavior of neptunium was investigated in order to predict its migration behavior at the disposal site for radioactive waste. Neptunium, in 5 mol/dm3 sodium chloride solutions of several pH values, was irradiated by α-particles of 238Pu which had been placed in the solutions as dioxide powder. Solution neptunium redox behavior was compared with that of an unirradiated sample. Pentavalent neptunium, which was stable in the absence of 238Pu, was found to be oxidized to hexavalent and even to heptavalent neptunium. Oxidizing species would be chloride molecule anion (Cl− 2) and/or hypochlorite anion (ClO−) which were generated by the reaction between radiolytically generated hydroxide radical (OH) and chloride ion (Cl−). The oxidation rate of pentavalent neptunium was independent of its concentration, but dependent on solution pH. The measured rate constant was (19±4)[H] mol/dm3/d.

Americium Valence Control Experiments in the Dissolved Solution of Irradiated Nuclear Fuel

A method based on valence control and solvent extraction is presented for separating minor actinides (Am, Np) from nuclear fuel reprocessing solution. With this method, Am is adjusted from trivalen...

Reuse System for Powdered Ion-Exchange Resins

A reuse system has been developed for powdered ion-exchange resins generated from nuclear power plants in order to reduce their waste volume. The system consists of: 1. crud removal from resins; 2. decomposition of flocks (flocculated resins); 3. resin separation into cation and anion types; 4. regeneration of each type. The most important points in this system are items 2 and 3, because generally resins flocculate too tightly to separate easily. By combined usage of a strong electrolyte (15 wt% NaOH solution) and a dual-basket-type centrifuge, spent powdered resins can be separated with an efficiency of 95% and regenerated for another use. The waste volume can be reduced to one-half after four reuse cycles, with a decrease in the ion-exchange capacity of only 5%.