J. Akatsu

Extraction chromatography in the DHDECMP—HNO3 system

The extraction behaviour of Ce(III) and Am(III) in extraction chromatography has been investigated on the basis of partition and infrared studies. The stationary phase was purified undiluted DHDECMP supported on Amberlite XAD-4 and the mobile phase was nitric acid. The results have shown that the equilibria for the extraction of Ce(III) and Am(III) by the DHDECMP/XAD-4 resin agreed very closely with those in solvent extraction

Extraction chromatography in the DHDECMP-HNO3 system. II : Characteristics of the DHDECMP/XAD-4 resin on separation of trivalent actinide elements

The extraction behaviour of Ce(III) and Am(III) in extraction chromatography has been investigated on the basis of partition and infrared studies. The stationary phase was purified undiluted DHDECMP supported on Amberlite XAD-4 and the mobile phase was nitric acid. The results have shown that the equilibria for the extraction of Ce(III) and Am(III) by the DHDECMP/XAD-4 resin agreed very closely with those in solvent extraction.

Extraction chromatography in the DHDECMP(XAD-4)-HNO3 system

An organic resin impregnated with DHDECMP was prepared for extraction chromatography and used for separation of the actinide elements. A known amount of inert support, Amberlite XAD-4, was contacted with a given amount of the extractant and water. By contacting the three phases for several hours at room temperature, the resin can be modified regardless of its mesh size. It is loaded with (1.13±0.03) g of DHDECMP per 1 g of the support. A given amount of the modified resin was contacted 50 times with fresh 3M nitric acid solution to remove the loosely bound extractant from the support. its amount was less than 1 weight%. As an application,241Am was recovered from 3.5M nitric acid waste solution on a column of the modified resin. Decontamination factor for the effluent was 105, while approximately 113 mg of241Am was obtained in 0.3M nitric acid eluting solution with 95% efficiency.

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.