Tadatoshi Yamamoto

Radiation Damage of Fluorocarbon by Krypton-85 Beta-Rays, (II): Radiolysis of Some Fluorocarbons by Cobalt-60 Gamma-Rays

Ampoule irradiations using 60Co γ-rays were undertaken on CC12F2, C2C14F2 and C2C13F3 to gain information concerning the radiation damage of fluorocarbon, used as solvent in the recovery of 85Kr from the off-gas of fuel reprocessing plants operating on the solvent absorption process. The dose rate was varied in the range of 2.6×104~8.0×105 rad/h and the irradiation temperature in the range of −100~70°C. Identified among the numerous degradation products were six compounds deriving from CC12F2, eleven from C2C14F2, and eleven also from C2C13F3. The decomposition yield of each of these fluorocarbons increased with irradiation temperature, and decreased with dose rate. It rose roughly in proportion to the absorbed dose. Comparison of decomposition behavior between the four fluorocarbons (including CC13F covered in the preceding report) was made in reference to the irradiation temperatures corresponding to those at which these fluorocarbons would respectively be applied in practice as solvent for 85Kr absorpt...

Retardation Factor of a Radionuclide for Undisturbed and Disturbed Sandy Soil

The relationship of the retardation factor of a radionuclide for undisturbed soil and that for disturbed soil is investigated. The migration model is based on the assumption that both reactive (dynamic) and nonreactive (stagnant) sites exist in the soil column. The retardation factor for undisturbed soil is represented by the following equation: Rf{sub u} = S{sub 2}/S{sub d}Rf{sub d}, where Rf{sub u} and Rf{sub d} are retardation factors for undisturbed and disturbed soil, respectively, and S{sub u} and S{sub d} are degrees of water saturation in undisturbed and disturbed soils, respectively. The migration experiments for both the undisturbed and disturbed soil columns were carried out using {sup 85}Sr. The average retardation factor for the undisturbed soil is smaller than that for disturbed soil, and the degree of water saturation in the undisturbed soil column is smaller than that in the disturbed soil column. The retardation factor for the undisturbed soil estimated, based on the above equation, is approximately the same as the measured retardation factor for the undisturbed soil.

Development of migration prediction system(MIGSTEM) for cationic species of radionuclides through soil layers.

The migration prediction system (MIGSTEM) has been developed for estimating the migration of cationic species of radionuclides through soil layers systematically. The MIGSTEM consists of the migration experiments, the one-dimensional fitting code (inverse analysis code) for determining retardation factor and dispersivity (migration factors) and the three-dimensional differential code (prediction code) for estimating the migration of the radionuclides. The migration experiments are carried out for obtaining the concentration profiles of the radionuclides in unsaturated and saturated soil layers. Using the inverse analysis code, the migration factors are obtained at one time by fitting the concentration profiles calculated to those observed. The prediction code can give the contours of concentration and the one-dimensional concentration profiles at selected time, as well as the changing in the concentration at a selected position with time. The validity of the MIGSTEM was obtained by the benchmark test on t...

Analysis of a hypothetical criticality accident in a waste supercompactor

A hypothetical nuclear criticality accident in a waste supercompactor is examined. The material being compressed in the compactor is a homogeneous mixture of beryllium and {sup 239}Pu. The point-kinetics equations with simple thermal-hydraulic feedback are used to model the transient behavior of the system. A computer code has been developed to solve the model equations. The computer code calculates the fission power history, fission yield, bulk temperature of the system, and several other thermal-hydraulic parameters of interest. Calculations have been performed for the waste supercompactor for various material misloading configurations. The peak power for the various accident scenarios varies from 1.04 {times} 10{sup 17} to 4.85 {times} 10{sup 20} fissions per second (fps). The total yield varies from 8.21 {times} 10{sup 17} to 7.73 {times} 10{sup 18} fissions, and the bulk temperature of the system varies from 412 to >912 K.

Validation of KENO V.a with ENDF/B-V cross sections for 233U systems

The multigroup Monte Carlo code KENO V.a and the 238- and 44-energy-group ENDF/B-V cross-section libraries were validated for {sup 233}U systems. Fifty-one critical experiments involving {sup 233}UO{sub 2}(NO{sub 3}){sub 2}, {sup 233}UO{sub 2}F{sub 2}, or {sup 233}U metal were selected for the validation. The H/{sup 233}U ratios for the experiments range from 0 to 1986. Each experiment was modeled with KENO V.a, and the effective multiplication factor k{sub eff} was calculated for each system using the 44- and 238-group ENDF/B-V, the 27- and 218-group ENDF/B-IV, and the 16-group Hansen-Roach cross-section libraries. The mean calculated k{sub eff} for all experiments using the 44- and 238-group libraries is 1.0090 {+-} 0.0021 and 1.0064 {+-} 0.0020, respectively. For comparison, the mean calculated k{sub eff} using the 27-, 218-, and 16-group libraries is 1.0142 {+-} 0.0038, 1.0125 {+-} 0.0038, and 0.9991 {+-} 0.0019, respectively. In general, an improvement exists in the agreement between the calculated k{sub eff}`s and the experimental results (i.e., k{sub eff} = 1.0) obtained with the newer ENDF/B-V libraries relative to ENDF/B-IV. This study is pertinent to {sup 233}U storage outside of the reactor.

Improved dose estimates for nuclear criticality accidents

Slide rules are improved for estimating doses and dose rates resulting from nuclear criticality accidents. The original slide rules were created for highly enriched uranium solutions and metals using hand calculations along with the decades old Way-Wigner radioactive decay relationship and the inverse square law. This work uses state-of-the-art methods and better data to improve the original slide rules and also to extend the slide rule concept to three additional systems; i.e., highly enriched (93.2 wt%) uranium damp (H/{sup 235}U = 10) powder (U{sub 3}O{sub 8}) and low-enriched (5 wt%) uranium mixtures (UO{sub 2}F{sub 2}) with a H/{sup 235}U ratio of 200 and 500. Although the improved slide rules differ only slightly from the original slide rules, the improved slide rules and also the new slide rules can be used with greater confidence since they are based on more rigorous methods and better nuclear data.

Adsorption and Desorption Properties of Radionuclides on Source Materials Used for Ground Migration Test.

In order to select an available inlet source material for a migration test of radionuclides in a sedimentary layer, 8 different materials; carbonated cement, coastal sand, chromosorb, sandstone, cation-exchange resin, zeolite, silica gel and Toyoura standard sand, were tested in terms of adsorption ability and leaching rate with time for 60Co, 85Sr and 137Cs. The adsorption terms of adsorption ability and leaching rate with time for 60Co, 85Sr and 137Cs. The adsorption abilities obtained as distribution coefficients (Kds) were tested by batch-type experiments, and the leaching rates were examined using column-type experiments. Among the materials, silica gel had the smallest Kd of approximately 60ml/g for the three radionuclides, and its Kd values did not fluctuate with time. Cement, sandstone and silica gel had a constant leaching rate for 60Co, 85Sr and 137Cs, ranging from 50 to 300ml inflow of groundwater. Approximately 70% of the three radionuclides were leached out from silica gel by 500ml inflow of groundwater. These results suggest that the silica gel is a suitable inlet source material for constantly supplying a solution with various radioactivities into a geologic layer.

Migration Behavior of 85Sr under Discontinuous Flow Condition.

自然環境の通気層中に生じる不連続な水の流れの条件下における85Srの移行挙動を明らかにするため, 85Srで汚染させた砂質土壌層へ脱イオン水を断続的に流下するカラム移行実験を行った。流れの停止回数が増すに従って, 陽イオン性85Srの土壌層中移行速度は増大する傾向を示した。この原因は, 土壌から間隙水中へ溶出するCa2+の濃度が流れの停止期間に増加することにより, 85Srの分配係数が減少するためであるとみなされた。土壌層流出液のCa2+濃度から推定した85Srの分配係数と土壌層における85Srの移行速度から得た分配係数とは, 流れの停止回数に伴う減少傾向がよく一致した。

Analysis of radionuclide migration behavior in loess medium

A five-year cooperative research program has been carried out by China Institute for Radiation Protection and Japan Atomic Energy Research Institute to develop safety assessment methodology for disposal of low level radioactive waste. Migration behavior of radionuclides {sup 3}H and {sup 85}Sr through field tests, simulation tests in the laboratory and determination of distribution coefficients is discussed and analyzed in this paper. The results show that the retardation coefficients, R{sub d}, from field tests are about 0.08--3.3 times those from simulation tests and R{sub d} from batch tests are 1.1--44 times those from field tests for {sup 85}Sr and loess medium. It was observed from field tests that radionuclides moved mainly downward under artificial sprinkling and a part of them moved up besides downward under natural rain condition. In addition, it was discovered that the retardation coefficient, R{sub d}, increases with velocity of unsaturated water flow, u, in the analysis.

Release of dissolved 85Kr by standing.

The experiments on the release of dissolved 85Kr by standing at room temperature were carried out to examine the influence of liquid level in a sampler and properties of solvent on the release efficiency. Six kinds of organic solvents as well as water were taken as solvents.The half-life period in case of the decrease in concentration of the dissolved 85Kr which was used as an index of release efficiency, was proportional to the liquid level in the sampler and was inversely proportional to the diffusion coefficient of Kr gas in solvent. For organic solvents belonging to homologous series, the half-life period became longer with increasing the carbon number of solvent molecule. From the relationship between the half-life period and the carbon number, the release efficiency in the dissolved 85Kr can be predicted for any commonly used solvent as a practical application. This method was found to be an effective means of removing the dissolved 85Kr of low level though it takes rather long time.