Hideo Matsuzuru

Leaching behavior of cesium-137 in cement-waste composites.

To assess the safety for disposal of a radioactive waste-cement composite, the leaching of I3’Cs from a waste composite into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. The leachability was measured as functions of waste to cement ratio (Wa/C), concentration of the salt (Na,SO,), temperature of leachant and curing time of specimens. The fraction of 137Cs leached from a specimen of Portland cement is 0.3-0.4 at the leaching time of 100 days, whereas the corresponding value for a specimen of slag cement is 0.1-0.15. The leachability depends on the structure of cement composite which is influenced by such factors as Wa/C, salt concentration of the wastes, and curing time of the specimens. The leach coefficient of ‘37Cs increases with increasing porosity of the cement composite. At lower temperature, curing time of the specimens significantly affects the leachability, probably due to insufficient curing. A curing time of ca. 50 days before the leaching test is enough to eliminate the apparent effect of temperature.

Mobile and immobile plutonium in a groundwater environment

Abstract The mobility of plutonium was measured by laboratory experiments conducted under conditions simulating the groundwater environment. The results thus obtained are discussed in comparison with those obtained with an in situ study on fallout plutonium migration. Plutonium (IV) was found in a series of laboratory experiments to have large distribution coefficients ( K d ). On the other hand, the K d values of the fallout plutonium were about one-tenth as large as those of Pu (IV). This might be ascribed to either the different oxidation states or the difference in the chemical forms. Furthermore, a trace amount of mobile plutonium was confirmed to be produced even in the simulated groundwater environment. The production rate of mobile plutonium and the degree of change in chemical forms of plutonium during underground migration were found to be strongly influenced by the amount of suspended solids in the groundwater.

Leaching behaviour of strontium-90 in cement composites

The leaching of 90Sr from a cement composite into an aqueous phase has been studied by the method recommended by IAEA. The amount leached was measured as functions of waste to cement ratio (Wa/C), salt content of waste, temperature of leachant and curing time of specimens. The leach coefficient of 90Sr varies from ca. 6 × 10−8 to 4 × 10−7 cm2/day depending on the composition of specimen and the leaching conditions. The leachability depends on such factors as Wa/C, temperature of leachant and curing time. The Portland cement composite gives a higher leaching fraction than the slag cement one. Additives used have no significant effect on the leachability. The amount leached in deionized water as a leachant is higher than in synthetic sea water. On the basis of the results obtained, the amount leached of 90Sr from a composite of 2001 drum size for extended period was estimated.

Effect of Dimension of Specimen on Amounts of Cesium-137, Strontium-90 and Cobalt-60 Leached from Matrix of Hardened Cement Grout

The relationship between the fraction leached and the ratio of surface area and volume (S/V) of a specimen has been studied in order to evaluate the safety of a cement composite for disposal. The leaching fraction of radionuclides was measured for the different S/V ratios. In a case of either 137Cs, 90Sr or 60Co, the amount leached is directly proportional to the S/Vratio. It is in harmony with the empirical formula proposed here, which can be derived from the diffusion equation with a semi-infinite plane source model. The amounts of 137Cs, 90Sr and 60Co leached from a cement composite of 200l drum size are estimated on the basis of the above relation, referring to the data obtained with a cylindrical shape specimen (4.5cm in dia. and 4.4cm high).

Leaching of Radionuclides from a Cement Composite Incorporating Evaporator Concentrates Generated at a Pressurized Water Reactor Nuclear Power Plant

The low- and intermediate-level wastes such as evaporator concentrates have been routinely solidified with cement in a power reactor plant. The leaching behavior of a cement composite incorporating evaporator concentrates produced at a pressurized water reactor nuclear power plant has been studied for safety assessment of the final disposal of waste solids. Leaching tests were carried out in accordance with the method recommended by the International Atomic Energy Agency. Amounts leached were measured as functions of the waste-cement weight ratio (Wa/C), temperature of leachant, and curing time of specimens. 28 refs.

Leaching behaviour of tritium from a hardened cement paste

Abstract Leaching of tritium from a hardened cement paste into an aqueous phase has been studied to assess the safety of solidification of the tritiated liquid waste with cement. Leaching tests were carried out in accordance with the method recommended by the International Atomic Energy Agency (IAEA). The leaching fraction was measured as functions of the waste-cement wt ratio ( Wa C ), temperature of leachant and curing time. The tritium leachability of cements follows the order: alumina cement > Portland cement > slag cement. The fraction of tritium leached increases with increasing Wa C and temperature and decreasing curing period. A deionized water as a leachant gives a slightly higher leachability than the synthetic sea water. The coating of the specimen surface with bitumen reduces the leachability to about 5% of its value for the specimen without coating.

Kinetics of Ion Exchange in the Chelating Resin Dowex A-1

The kinetics of ion exchanges of Ag+, Zn2+ and Cr3+ at extremely low concentrations on the chelating resin Dowex A-1 has been studied by means of finite volume method. The rate of exchanges for both Ag+ and Zn2+ is dependent on the ionic strength, particle size of the resin and reaction temperature. At higher ionic strength (0.1–0.05) the kinetics is controlled by particle diffusion, whereas at lower ionic one (0.01–0.001) film diffusion is predominant. The apparent activation energy obtained is 3.84 kcal/mol for Ag+ and 3.91 kcal/mol for Zn2+. The exchange rate of Cr3+ obeys a first-order rate equation independent of the ionic strength and particle size of the resin. The apparent activation energy is 15.5 kcal/mol. These results support the view that the ratedetermining step of this reaction is chelate formation reaction.

Modeling of release of radionuclides from an engineered disposal facility for shallow-land disposal of low-level radioactive wastes

Abstract The computer code, ENBAR-1, for the simulation of radionuclide releases from an engineered disposal facility has been developed to evaluate the source term for subsequent migration of radionuclides in and through a natural barrier. The system considered here is that a waste package (waste form and container) is placed, together with backfill materials, into a concrete pit as a disposal unit for shallow-land disposal of low-level radioactive wastes. The code developed includes the following modules: water penetration into a concrete pit, corrosion of a drum as a container, leaching of radionuclides from a waste form, migration of radionuclides in backfill materials, release of radionuclides from the pit. The code has the advantage of its simplicity of operation and presentation while still allowing comprehensive evaluation of each element of an engineered disposal facility to be treated. The performance and source term of the facility might be readily estimated with a few key parameters to define the problem.

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.

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.