Yoshii Kobayashi

The role of suspended particles in the discharge of 210Pb and 7Be within the Kuji River watershed, Japan

Abstract Fluvial discharge of atmospherically-derived 210Pb and 7Be was studied in the Kuji River watershed in Japan. The daily load of 210Pb and 7Be, associated with suspended particles, greatly increased during high-flow conditions. However, the activity concentrations of 210Pb and 7Be in the suspended particles were found to be rather independent of the flow rate. Due to this feature, the daily load of the activities associated with the suspended particles was mainly controlled by that of the suspended particles. Compared with the other reported values, the estimated mean residence time of atmospherically-derived 210Pb in the present study was rather short (about 800 years), possibly reflecting the high availability of the suspended particles eroded from the top soil in the watershed. The annual load of 210Pb and 7Be was predominantly of particulate form (approx 90%). These results suggest that the fluvial discharge of atmospherically-derived, highly adsorptive (kd ≳ 104 mg l−1) radionuclides in a river watershed is primarily due to the discharge of suspended particles, which strongly depends on the flow rate of river water and also on the location of the watershed.

Coprecipitation of uranium and thorium with barium sulfate

The coprecipitation behaviour of uranium or thorium with barium sulfate is investigated from the variation of yields with uranium or thorium concentration, acid and acidity, and amount of sodium and/or potassium sulfate. Uranium or thorium in quantities less than 1.5 mg is quantitatively coprecipitated with barium (5.9 mg) sulfate when using an optimum conditions. The chemical form of uranium in barium sulfate precipitates is discussed by determination of mole ratio of potassium to uranium.

URANIUM AMALGAM FROM AQUEOUS SOLUTION.

Abstract Various factors which influence the extractibility of uranium from aqueous uranyl solution with sodium amalgam have been investigated. Conditions for the virtually complete extraction of uranium from aqueous solution are described. Uranium amalgam containing more than 50 mg U/ml Hg has been obtained in one extraction.

Autoxidation and metal surface catalysed oxidation of uranium (III) in acid solutions

Abstract The oxidation of uranium (III) in acid solutions has been studied by spectrophotometric measurement in an inert gas atmosphere. Uranium (III) obtained by reduction of uranyl salt with zinc amalgam was oxidized more rapidly than that prepared by dissolution of uranium amalgam or electrolytic reduction of uranyl ion. The oxidation is strongly catalysed by copper or lead ion contained as a trace impurity in zinc. The autoxidation and the metal surface catalysed oxidation of uranium (III) are discussed.

Praseodymium-143 from Neutron-Irradiated Uranium

From neutron-irradiated uranium cooled for about 2 days, 143Pr was separated indirectly by solvent extraction. After removing uranium and some extractable nuclides with bis(2-ethyl hexyl)orthophosphoric acid from 10 M nitric acid solution, 143Ce was extracted with bis(2-ethyl hexyl) orthophosphoric acid (n-heptane diluent) from a 10 M nitric acid solution containing potassium bromate. Then, 143Pr was purified after most of the 143Ce had decayed off. Detailed measurements indicated a half-life of 13.55 ±0.02 days and maximum β-ray energy of 0.93±0.05 MeV.

Neutron yields from actinide oxides

Abstract Neutron yields emitted from actinide oxides by spontaneous fission and by the (α, n) reaction of oxygen were measured separately. The measured neutron yields of spontaneous fission were consistent with the neutron yields calculated from ν - values and rates of spontaneous fission. The (α, n) neutron yields calculated from stopping power and thick target yield data were lower than the measured yields by about 10%.

A comparison of the acid digestion of spent ion exchange resins using H2SO4-HNO3 and H2SO4H2O2

An alternative acid digestion system, H2SO4H2O2, to a conventional reaction system, H2SO4HNO3, has been proposed to reduce the volume of spent ion exchange resins generated at nuclear power plants. A comparative study on both reaction systems has been carried out to obtain the relationship between the reaction conditions and the conversion of the resins, the coprecipitation behavior of 60Co and 137Cs, and the release rate of the radionuclides from a reaction vessel, and to elucidate the feasibility of the system proposed. The mole ratio of an oxidant and carbon contained in the resins, required to digest the resins, was comparable between both systems, while the H2SO4H2O2 system gave higher conversion than the H2SO4HNO3 system. Some degradation products of an anion exchange resin were relatively stable against the oxidation, and required excessive amounts of the oxidant and a higher reaction temperature to digest, especially in the H2SO4HNO3 system. The coprecipitation behavior of the radionuclides in both systems was almost identical; 60Co coprecipitated effectively with Fe(III) sulfate while 137Cs did not coprecipitate. The release rates of the radionuclides were on the order of 10−3%/hr.

Non-destructive determination of spontaneously fissioning nuclides by neutron coincidence counting using multichannel time spectroscopy

A non-destructive method for determining the amount of actinoids has been developed. The method is based on thermal neutron coincidence counting and employs a selective detection of neutrons resulting from the spontaneous fission of actinoids. The detection system is described in detail and the measurement results of244Cm as an example are presented. The results show that the measured fission rate of244Cm is consistent with the fission rate calculated from ENDF/B-V data and that the amount of244Cm can be determined within about 5% accuracy even in the presence of a large amount of actinoids, for example, up to 2.6·106, 3.6·104, or 1.6·103 times in the mass ratio of239Pu,241Am, or240Pu to244Cm, respectively.

Non-destructive determination of239Pu in wastes by application of γ-ray measurement

Amounts of239Pu in alpha-bearing wastes were determined by γ-ray measurement using a NaI/Tl/ scintillation detector. The deviations of measurement caused by the nonuniform distribution of239Pu in the wastes and by a volume factor were corrected by rotating and scanning the objects. Contribution from241Am coexisting with239Pu was also corrected by taking into account the difference of the counts between A /356–470 keV/ and B /475–620 keV/ intervals. Attenuation of γ-ray of239Pu was estimated from the matrix density. Thus the content of239Pu could be determined within an accuracy of about 25% at the 1 mg level in a cardboard carton of 17 1.

Separation of Pu-Am from the Leachant of a Deposit in an Acid Digestion Solution by Calcium Oxalate Coprecipitation

Abstract Pu and Am were recovered from the deposit in a synthetic acid digestion solution. After being leached out from the deposit with water, they were separated from the resultant solution by Ca-oxalate coprecipitation. The alpha radioactivity in the solution was reduced from about 4 mCi/L to 0.1 μCi/L by the technique. The precipitate obtained was dissolved in 7 M nitric acid solution, from which Pu-Am were separated by the use of anion and cation exchange resin columns, respectively. The coprecipitation technique was also utilized for the concentration of Am in the effluent of the anion column prior to its separation by a cation column. The overall recovery efficiencies of Pu and Am were about 80 and 85%, respectively.