Distribution and Fixation of Cesium and Strontium in Zeolite A and Chabazite

The selective removal and fixation of Cs and Sr have been studied in zeolite A and chabazite. Cesium ion was preferentially distributed into chabazite with a high distribution coefficient (K _Cs>10³ cm³·g^?1) in the presence of NaCl (10^?1 mol·dm^?3). The K _Sr values for zeolite A attained about 10³ cm³·g^?1 in the pH range of 8～10, and they gradually decreased with an decrease in pH.

The initial rate of Cs adsorption was fairly fast in chabazite, and the adsorption ratio reached almost 100% within a few hours. The adsorption ratio of Sr in binderless A zeolite reached almost 100% after 15 h. The adsorption of Cs and Sr on these zeolites was followed by Langmuir-type isotherm. Cesium forms of these zeolites recrystallized to pollucite (CsAlSi₂O₆) above 900°C for zeolite A and above 1,200°C for chabazite. As for Sr forms, these zeolites changed to SrAl₂Si₂O₆ above 900°C.

These recrystallized phases were suitable hosts for the immobilization of Cs and Sr in the nuclear waste solutions.     

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

Glasses developed for the treatment of low- and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90°C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (i.e., pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m²·d, when the temperatures were between 40 and 90°C and the leachant condition was pH 4–11. Except for the DG2 glass, the minimum forward dissolution rate (0.01–1 g/m²·d) was obtained at approximately pH 7–8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.     

Tritium diffusion in zircaloy-2 in the temperature range −78 to 204° C

Tritium diffusion measurements in Zircaloy-2 were carried out over the temperature range ?78 to 204 °C by direct measurement of tritium diffusion gradients. The ⁶Li (n, α)³H reaction was used to inject tritium into the specimens and to produce initial tritium concentration in the range 0.0065 ppm to 0.013 ppm ³H by weight. Two diffusion components were identified from the concentration profiles: a surface trapping region approximately 5 μm thick and a normal diffusion profile characteristics of bulk diffusion. Surface release measurements of tritium verified the existence of a surface trapping layer. The bulk diffusion component was consistent with classical diffusion solutions and was given by: D = 0.00021_?0.00018^+0.005 exp?(8500 ± 200 cal/RT) cm² · sec^?1.The surface trapping was attributed to oxide films formed on the Zircaloy-2 at room temperature. The apparent diffusion coefficients for the surface region were consistent with: D = 4.0_?3.3^+19.7 × 10^?14 exp?(7200 ± 1500 cal/RT) cm² · sec^?1 over the temperature range 25 to 411°C.     

Thermal Conductivity of ThO2-UO2 System

The thermal diffusivity of the ThO₂-UO₂ system in solid solution was measured by laser pulse method at temperatures ranging from 20° to 500°C. Reproducibility of the data was confirmed to be within 3%. The compositions of the samples were ThO₂, ThO₂-1%UO₂, ThO₂-5%UO₂ and ThO₂-10%UO₂.

The thermal conductivity was calculated from measured thermal diffusivity data and the specific heat data available in literature and corrected to zero porosity by using Loeb's equation, in which the shape factor is unity.

The values on ThO₂ thus obtained agreed very well with the data found in literature, throughout the range of temperature of the experiments. The thermal conductivity of ThO₂, ThO₂-1%UO₂, ThO₂-5%UO₂ and ThO₂-10%UO₂, at 20°C, were 0.0312, 0.0288, 0.247 and 0.0184 cal·cm/sec· °C·cm², respectively.     

A Method of In-Pile Calorimetry

In order to immobilize Cs into pollucite structure, hydrothermal reactions of Cs with amorphous aluminosilicate (siliceous sinter) and with mixtures of low-quartz (silicastone) and Al(OH)₃, were performed. In both cases, pollucite was formed in 5-N NaOH solution above 200°C for 10 min. Synthesized pollucite was solid solution between pollucite and analcite. Waste forms in which Cs was immobilized in pollucite structure were successfully produced by the hydrothermal hot-pressing method using the silica matrix (mixture of low-quartz and amorphous aluminosilicate) containing Al(OH)₃ and NaAlO₂. Effect of hot-pressing conditions on leachability of the waste forms was investigated. The waste form containing 10w/o CsOH produced at 300°C and 49 MPa for 24 h, had Cs leach rate of 3.15 g/m²·d by Soxhlet leach tests for 7 d. The waste form was porous, but had high compressive strength and thermal stability.     

Preparation of Uranium Carbonitride by Reaction of UC with Ammonia

With the view to establishing a method of directly converting uranium carbide into uranium carbonitride and hydrocarbons, an attempt has been made to induce reaction between UC and ammonia under various temperatures from 25° to 600°C and pressures from 1 to 1,500 kg/cm². The reaction aimed at was realized to the extent of practical significance under pressures exceeding 500 kg/cm² at 450°C and exceeding 250 kg/cm² at 500°C. The hydrocarbons produced thereby were found to be mainly methane, indicating that the formula of the predominant reaction was

• UC+NH₃→UN_2-x+ CH₄+H₂.

Upon heating the powdery fine black product to 1,800°C in vacuo, unexpectedly marked sintering was found to occur, resulting in dense uranium monocarbonitride without any compaction pretreatment.     

Thermal conductivity of neutron-irradiated pyrolytic β-silicon carbide

Samples of pyrolytic β-silicon carbide deposited at 1400 °C (grain size $\text{～ 1 μ}\text{m}$) and at 1750 °C (grain size $\text{～ 3 μ}\text{m}$) were irradiated with fast neutrons to 2.7-7.7 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$) at 550 °–1100 °C. Irradiation reduced the room-temperature thermal conductivity from $\text{～0.15}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ to $\text{～ 0.02}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ after irradiation at 550 °C and to $\text{～ 0.05}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ for an irradiation temperature of 1100 °C. The thermal conductivity of unirradiated samples decreased with increasing measurement temperature, while that of the irradiated samples was much less temperature dependent. No difference in behaviour was found between the samples with $\text{～ 1 μ}\text{m}$ grain size and the samples with $\text{～ 3 μ}\text{m}$ grain size.     

Alkali Hydrothermal Synthesis of Zeolites from Coal Fly Ash and Their Uptake Properties of Cesium Ion

Zeolites were synthesized from coal fly ash by hydrothermal treatment with KOH solutions. K-H zeolite (K₂Al₂Si₄O₁₂-nH₂O) was produced under optimum conditions of 160°C, 3 d, 1 M (=mol/dm³) KOH and liquid-solid ratio of 15cm³/g. The uptake behavior of radioactive cesium for the products was investigated by batch and column methods. The uptake equilibrium of Cs+ for the above product was attained within 2h yielding the distribution coefficient of above 10⁴ cm³/g. The uptake of Cs⁺ was followed by a Langmuir adsorption isotherm and the maximum uptake capacity was estimated to be 3.34 mmol/g. The successive removal of Cs⁺ was accomplished through the column packed with granular composites of product-alginate gel polymer.     

Recoil Release of Fission-Gas from UO2

The fission-gas released from in-core UO₂ pellets was investigated with the helium-swept fuel irradiation facility installed in the Hitachi Training Reactor. It was found that the release rate of fission-gas increased gradually as irradiation continued for 5hr under a constant thermal neutron flux of 1.0×10¹² n/cm²·sec and at a fuel temperature of about 150°C. This observation indicates that a recoil-activated release mechanism may be more effective than direct-recoil release in the case of clad fuel in the region of temperature where fission-gas release rate is temperature independent.     

Accelerated Leach Tests of SRL-165 High-Level Waste Glass in Deionized Water

Accelerated short-term leach tests in a laboratory are neccessary in order to estimate, with reasonable accuracy, the long-term leaching behavior of high-level waste glass. In the present study, static leach tests of an SRL-165 high-level waste glass were carried out in deionized water at two different glass-surface-area to solution-volume ratios (SA/V-ratio), namely 0.85 and 0.079 cm^?1 at 90°C, and 0.85 cm^?1 at 40°C.

First, an equation was examined which related Si-concentrations with time, temperature and SA/V-ratio under the present static conditions. The parameter determined at 90°C, 0.85 cm^?1 can be used to calculate the Si-concentration at 40°C, 0.85 cm^?1. Second, at the low SA/V- ratio of 0.079 cm^?1, the concentrations of Ca and Mg in the leachates peaked and then decreased a little. The equation used above does not explain the variation of the concentrations of Ca and Mg at a low SA/V-ratio. The precipitation of Ca and Mg onto the glass surface is probably caused by the adsorption efficiency of the surface layer or the formation of crystalline materials at the low SA/V-ratio of 0.079 cm^?1. Third, the in-depth profiles of some elements obtained by secondary ion mass spectroscopy (SIMS) were qualitatively in agreement with the results of solution analyses. This indicates the particular usefulness of SIMS for analyzing leaching behavior of the glass in in-situ burial studies where solution analyses are often impractical.     

Molten Salt Extraction for Removing Iodine from Liquid Sodium

The initial burst and the diffusion of fission produced Xe were investigated on UC₂ irradiated to thermal neutron doses between 2.1 x l0¹⁴ and 1.3 x 10¹⁶ fission/cm⁸. In heating-runs with 5 and 10°C/min and below 1,000°C, the burst revealed three steps at around 200, 400 and 600°C (in 5°C/min heating run), to which activation energies were obtained. It was found that the temperatures and the activation energies for each burst were close to those appearing in recovery processes of fission induced defects. In diffusion process above 1,000°C, on the other hand, the activation energy of Xe in UC₂ increased with increasing fission dose. In the specimens once heated up to 1,500°C, however, the diffusion coefficient could be expressed as, D?10^-1 exp (— 344±50 kJ/RT) (cm²/s), which was independent of the level of the fission dose. An enhancement and a suppression of the diffusion were suggested also in the dicarbide as was done previously for the monocarbide of uranium.     

Measurement and Calculation of Neutron Diffusion Parameters in Water

Neutron diffusion parameters in water at a room temperature of 10°C have been measured by the pulsed neutron method for the range of geometrical buckling from 0.093 to 1.36 cm^-2. The results are 205±4 μ60 for the neutron mean life time due to absorption, 34,120±610 cm²·sec^-1 for the diffusion coefficient and 3,350±560 cm⁴·sec^-1 for the diffusion cooling coefficient.

The decay constant has been calculated as a function of buckling for the Nelkin and the Rad-kowsky scattering models of water on the assumption of linear anisotropic scattering. The calculated diffusion coefficients, 36,290 cm²·sec^-1 for the Nelkin model and 37.610 cm²·sec^-1 for the Radkowsky model, are somewhat higher than the experimental result.

It is shown that the calculated diffusion coefficient approaches the experimental value if we use μ-(E), the mean value of cosine of scattering angle, obtained from the Beyster's experiment instead of that for the Nelkin model.     

In-reactor investigations of the creep of structural materials

Conclusions 1. A series of in-reactor tests was performed on a sample used to study radiation creep in 00X16H15M3B steel, XHM1 chrome-nickel alloy, the zirconium based alloys é110 and é635, and the vanadium-based alloy BTX8. The radiation creep modulus (in units of Pa⁻¹·(displacements/atom)⁻¹ equals 1.7·10⁻¹¹ for 00X16H15M3B steel, 4.6·10⁻¹¹ for XHM alloy with fluence up to 2.3·10²⁰ cm⁻² and 1.6·10⁻¹¹ for a fluence above 1·10²¹ cm⁻², (4.6–4.9)·10⁻¹¹ for é110 alloy, and 1.8·10⁻¹¹ for é635 alloy. For the alloy BTX8, at stresses below half the yield point and t=450°C, the modulus equals 3.3·10⁻¹² Pa⁻¹·(displacements/atom)⁻¹. At a higher stress, the deformation rate of the alloy increases progressively. 2. In the investigation of the temperature dependence of in-reactor creep of the alloy é110, it was found that at 350–370°C and higher, the thermal creep makes the predominant contribution to deformation. In the experimental range 370–455°C, the thermal activation energy of in-reactor creep was determined to be 36 ± 8 kcal/(g·atom). At temperatures below 350°C the creep of the alloy é110 is a temperature-independent radiation-stimulated process. 3. In the case of tests of zirconium alloys, a previously unobserved phenomenon of periodic rapid deformation of the material against the background of creep at stresses even well below the yield point of the irradiated material was discovered. The effect was manifested at a temperature of about 230°C. As the temperature increases up to 290°C and higher, no plastic movements are observed. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 386–391, May, 1996.     

Release of Neptunium from Neptunium-Doped Borosilicate Waste Glass

The release of neptunium from a neptunium-doped borosilicate waste glass was studied at 90°C in deionized water and silicate water. The standard MCC-1 static leach method was applied to the tests for durations up to 91 days with the SA/V ratio of 10 m^?1.

The normalized elemental mass loss obtained for neptunium was about 5 g/m² for both the deionized and the silicate water leachates. This value is similar to those for currently typical borosilicate waste glasses. That is, the studied glass is comparable with the typical glasses in terms of the ability to immobilize neptunium.

The time dependence of the release of neptunium from the glass was different from those of soluble glass components such as sodium, boron and cesium, but similar to that of strontium. A part of neptunium, like strontium, probably remained in the surface layer formed on the leached glass. The neptunium species in the surface layer was predicted to be NpO₂.xH₂O(am) based on available solubility data.     

Diffusion Coefficient of Fission Gas in Uranium Dioxide Powder Formed by Carbon Dioxide Oxidation of Uranium

In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO₂ oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized at 600° and 700°C, and in both cases the O/U ratio of the oxides, measured gravimetrically, was 2.0. The diffussion coefficients in the oxide powder formed by oxidation at 600°C were found to be 1.4× 10-²⁰, 1.3×10-¹⁹, 1.1×10¹⁸ and 1.0×10-¹⁷, cm²sec-¹, respectively at 450°, 550°, 650° and 750°C, and in the oxide powder formed at 700°C, 7.4×10-¹⁹ and 3.6×10-¹⁶cm²sec-¹ at 600° and 700°C, respectively. Activation energies calculated for the two oxide powders were comparatively low.     

Effects of heat-treatment and neutron-irradiation on ultrasonic attenuation and young's modulus of carbon

This paper reports the changes in ultrasonic attenuation and Young's modulus due to heat-treatment and neutronirradiation of polycrystalline carbon. The measurements were made over a frequency range from 2 to 10 MHz. With increase of heat-treatment temperature (HTT) from 1300 to 2800 °C, Young's modulus decreased and the coefficient of ultrasonic attenuation increased, whereas at HTT = 3100°C the ultrasonic attenuation decreased, with a slight attendant decrease in Young's modulus. When the 3100°C-treated specimens were irradiated by fast neutrons, the coefficient of ultrasonic attenuation decreased and Young's modulus increased. These results were discussed in terms of motion of dislocations. The damping constant and dislocation density were roughly estimated to be about 10⁻⁵ dyne · s/cm² and 10¹¹/cm², respectively.     

Fission-enhanced self-diffusion of uranium in UO2 and UC

In-pile self-diffusion measurements in stoichiometric UO₂ sinters and single crystals and in arc-cast stoichiometric UC have been performed using the thin layer condition and ²³³U as tracer. The nominal irradiation temperature was 900°C. The resulting diffusion coefficients $\text{D1}$ of $\text{1.5 × 10}{}^{\mathrm{?16}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UO₂ and $\text{2.2 × 10}{}^{\mathrm{?17}}\text{cm}{}^2\text{· sec}{}^{\mathrm{?1}}$ for UC for a fission rate $\text{S}$ of 1 × 10¹³f/cm³ · sec represent radiation enhanced diffusion and are higher by factors of 10³ to 10⁴ than (extrapolated) coefficients of thermal diffusion. The data are of immediate relevance for understanding and predicting such important quantities as in-pile sintering and densification, diffusion controlled creep and fission gas behavior in the outer zones of the fuel. They are at the upper limit of expected values.     

Effects of Annealing under High Static Pressure on Precipitation Behavior of Fission Gas Bubbles in Irradiated UO2

Annealing experiments were carried out on irradiated UO₂ in argon gas under high pressure (600 and 1,000 kg/cm²) as well as atmospheric, at temperatures of 1,400°–1,600°C. The effects of high external pressure on the behavior of fission gas bubbles in the irradiated UO₂ were studied by comparing replica electron micrographs of fractured surfaces of specimens annealed under different temperatures and pressures. The results indicate that high pressures such as above 600 kg/cm² can be effective in surpressing the growth of fission gas bubbles in both intergranular and intragranular zones, and in inhibiting the joining together of intergranular bubbles to form direct passages for fission gas release.     

Homogeneously size distributed Ge nanoclusters embedded in SiO2 layers produced by ion beam synthesis

500 nm SiO₂ layers were implanted with 450 keV (F=3 × 10¹⁶ at./cm²) and 230 keV (F=1.8 × 10¹⁶ at./cm²) Ge ions at room temperature to obtain an almost constant Ge concentration of about 2.5 at.% in the insulating layer. Subsequently, the specimens were annealed at temperatures between 500°C and 1200°C for 30 min in a dry N₂ ambient atmosphere. Cross-sectional TEM analysis reveal homogeneously distributed Ge nanoclusters arranged in a broad band within the SiO₂ layer. Their mean cluster size varies between 2.0 and 6.5 nm depending on the annealing conditions. Cluster-free regions are always observed close to the surface of the specimens independent of the annealing process, whereas a narrow Ge nanocluster band appears at the SiO₂/Si interface at high annealing temperatures, e.g. ⩾1000°C. The atomic Ge redistribution due to the annealing treatment was investigated with a scanning TEM energy dispersive X-ray system and Rutherford back scattering (RBS).     

Self-Limiting Power Excursion Characteristics of Light Water Reactor, (IV)

A strain gage was used for the measurement, of fuel cladding strain generated during pulse operation tests on the Hitachi Training Reactor. In the analysis of the measured strain, two kinds of correction were called for: (1) the fadiation effect on the strain gage and lead wires, and (2) the temperature effect due to the lag of the gage filament temperature behind the true fuel cladding temperature. The experimental axial strain after the two corrections were applied was 781 × 10^?6 cm/cm for the hottest fuel rod in the pulse operation test with an inserted reactivity of 1.20%δk/k. This maximum strain corresponded to 2,169 kg/cm² of thermal stress and 111 cal/cm²·sec of heat flux. These results were obtained under the condition of maximum temperature in the fuel center of 1,200°C and a fuel cladding temperature of 140°C. When the axial strain was calculated with consideration given to the gap or contact conductance between the fuel and its cladding, a reasonable agreement was obtained between the calculation and the experimental results.