Leach Rates and Thermal Properties of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste were studied on their leach rates and thermal properties.

The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/₀ is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15 w/₀ at 950°C. The leach rate of the vitrified waste glass was in the order of 10^?7 g/cm².d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35% enabled it to produce the glass form with the waste near 20 w/₀ at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed.

Removing Na ions from wastes improves considerably both the leach rate and the thermal stability of the LIP glass waste form.     

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.     

Effect of Condition of Recrystallization Heat Treatment on High-Temperature Mechanical Properties of an Irradiated AISI 316 Austenitic Steel

A study was made of the effects brought upon the radiation-induced high-temperature embrittlement of AISI 316 austenitic steel by different conditions to the specimens to produce recrystallization.

Cold worked specimens were recrystallized at temperatures 950°–1,100°C held for periods ranging 2~60 min. The specimens were then irradiated to 1.7×10²¹nvt(< MeV) at 55°C after which they were subjected to tensile testing at 650°C.

In the specimens recrystallized at 950°C, carbide precipitation was observed to have occurred, and these specimens were found less liable to show radiation-induced embrittlement. On the other hand, specimens recrystallized at 1,025° and 1,100°C became completely solution treated, and exhibited severe radiation-induced embrittlement. It is surmisted that the carbides precipitated on the grain-boundaries tend to inhibit propagation of grain-boundary cracks, and hence contribute to lessing radiation-induced embrittlement.     

Immobilization of Radioactive Wastes in Hydrothermal Synthetic Rock, (III)

The high-level radioactive waste form (21.8w/o waste, 47.7w/o α-quartz, 20.5w/o amorphous aluminosilicate, 10w/o Al(OH)₃) was produced with the addition of 10-N NaOH solution under hydrothermal hot-pressing conditions at 350°C and 66 MPa for 6 h. In order to characterize the waste form, following tests and measurements were performed: Soxhlet flow leach tests (97°C), static leach tests in deionized water at low temperatures (40, 90°C) and under hydrothermal conditions (100–350°C), crystalline phase determination, microstructure observation, compressive strength measurement, weight loss measurement by heat treatment and thermal conductivity measurement.

The waste form was mainly composed of α-quartz and had porous structure. As for leach rates determined by static low temperature leach tests, the leach rates of the waste form were much lower than a concrete form but slightly higher than a glass form. The waste form was stable under hydrothermal conditions in comparison with a glass form. It had high compressive strength and thermal resistance. Its thermal conductivity was higher at room temperature than that of a glass form.     

Initial dissolution rate of a Japanese simulated high-level waste glass P0798 as a function of pH and temperature measured by using micro-channel flow-through test method

Aqueous dissolution tests were performed for a Japanese type of simulated high-level waste (HLW) glass P0798 by using a newly developed test method of micro-channel flow-through (MCFT) method, and the initial dissolution rate of glass matrix, r ₀, was measured as a function of solution pH (3–11) and temperature (25–90°C) precisely and consistently for systematic evaluation of the dissolution kinetics. The MCFT method using a micro-channel reactor with a coupon shaped glass specimen has the following features to provide precise and consistent data on the glass dissolution rate: (1) any controlled constant solution condition can be provided over the test duration; (2) the glass surface area actually reacting with solution can be determined accurately; and (3) direct and totally quantitative analyses of the reacted glass surface can be performed for confirming consistency of the test results. The present test results indicated that the r ₀ shows a “V-shaped” pH dependence with a minimum at around pH 6 at 25°C, but it changes to a “U-shaped” one with a flat bottom at neutral pH at elevated temperatures of up to 90°C. The present results also indicated that the r ₀ increases with temperature according to an Arrhenius law at any pH, and the apparent activation energy evaluated from Arrhenius relation increases with pH from 54 kJ/mol at pH 3 to 76 kJ/mol at pH 10, which suggests that the dissolution mechanism changes depending on pH.     

Diffusion of Neptunium in Simulated High Level Waste Glass

Neptunium in high level radioactive wastes has to be retained in glasses through geological period from the point of biological toxicity for more than million years. Neptunium-237 diffusion in borosilicate glass with simulated wastes of 26.4% was investigated in the temperature range of 400–600°C by the use of α-degradation method. The energy loss rate dE/dx of α-particles, which is necessary in order to determine diffusion coefficients by the α-degradation method, was calculated for the waste glass.

The penetration depth of α-particle with 4.787MeV from ²³⁷Np was 17 μm, which gives a limit in applying the α-degradation method in the waste glass. The temperature dependence of the diffusion coefficient of Np in the waste glass was given by

D _Np=3.67 exp(-55,900/RT) (cm²/s), in which the activation energy of the diffusion was 55.9 kcal/mol. It was clarified that Np is one of the elements with the lowest mobility in waste glasses.     

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.     

Generation mechanism and prevention method of secondary molybdate phase during vitrification of PUREX wastes in liquid-fed ceramic melter

ABSTRACT

In order to develop a glass production method that inhibits formation of the Mo-bearing secondary phase, the so-called yellow phase (YP), and also to clarify its formation behavior, vitrification tests using a small-scale liquid-fed ceramic melter (LFCM) were conducted. Chemically simulated PUREX liquid waste containing Na, Mo and other fission elements was fed into a molten glass pool in the melter together with several sizes of feed glass. When the glass beads with a diameter of around 2 mm were fed, some YP was contained in the vitrified glass products at 25 wt% waste loading. While, almost no YP was detected in the glass products when glass powder with a diameter of less than 63 μm was used. The microscopic analysis of the cold-cap samples revealed that powderization of the feed glass contributed to the facilitation of Na dissolution to the feed glass in the cold-cap. This quick intake of Na to the feed glass prevented the formation of liquid Na₂MoO₄ aggregation and provided homogeneous Mo dispersion in the cold-cap with forms of alkali earth and/or rare earth molybdates. The homogeneous dispersion of Mo resulted in the fast and complete dissolution of Mo in the glass melt.     

Corrosion Behavior of a Powdered Simulated Nuclear Waste Glass in the Presence of Bentonite

Static corrosion tests of a powdered simulated waste glass were performed in deionized water with and without bentonite at 90°C for periods of up to 130 days. The glass irradiated with thermal neutron for activating Cs in the glass was used as a glass specimen in order to determine the sorption of Cs on bentonite. In the corrosion tests without bentonite, it was observed that normalized elemental mass loss: (NL) values for soluble elements (B, Li, Na and Mo) were larger than those for Si by a factor of three and continued to increase after saturation of Si. In the corrosion tests in the presence of bentonite, it was observed that the glass corrosion was enhanced, and a large amount of Cs was sorbed on bentonite.

The experimental results were analyzed by use of some corrosion models. The analysis showed that diffusion of the soluble elements is expected to be a dominant process for the glass corrosion, as well as the dissolution/precipitation reactions. In addition, it is expected that the glass corrosion in the presence of bentonite is largely affected by both ion-exchange equilibrium of the aqueous phase with Na-montmorillonite and precipitation of sepiolite from dissolution of dolomite.     

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.     

Temperature Effect on Plutonium Leach Rate of Nuclear Waste Glass

The two kinds of nuclear waste glass with similar composition, a ²³⁸Pu-doped and nonradioactive waste glass, were leached under the ISO-test conditions at temperature between 23 and 90°C. An activation energy of 22±10kJ/mole was obtained from the initial leach rates of Pu, which was much lower than the 78±9kJ/mole obtained from those of Si, Na, Sr and Cs, It is suggested that in the initial stages of leaching. Pu is not released from the waste glass with the same mechanisms as the releases of Si, Na, Sr and Cs, but the dissolution of hydrous plutonium dioxide PuO₂·_xH₂O formed on the glass surface becomes predominant. In the long duration tests (<32d), the release of Pu appears to be affected by the solubility of PuO₂·_xH₂O remaining in the leached surface layers.     

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.     

Manufacturing of Porous Boron Steels Potentially Useful as Nuclear Materials

B₄C is a good neutron absorber, commonly used together with light materials in panels. The objective of this work is to manufacture high boron steels, using B₄C additions, through mechanical alloying and sintering, to get a material potentially useful for nuclear waste management. The porosity of the material can help to the removal of helium bubbles. Iron and B₄C powders were mechanically alloyed for different times, following the process studying apparent density, morphology (SEM) and structure (XRD).

Powder was uniaxially compacted and sintered at different conditions. Specimens were analysed by SEM and physical and mechanical properties were evaluated (density, dimensional change and bending strength).

Microstructures are very different and therefore, they have different properties depending on sintering temperature. Although boride formation always takes place, only ferritic areas were found at 600°C, meanwhile ferritic and perlitic areas appeared at 900°C, and both of them disappeared at l,200°C.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

Accommodation of Cerium and Neodymium Oxides into Alumina

Alumina was studied on its applicability for a host material to incorporate lanthanide and actinide elements present in HLW. Alumina-based waste forms containing about 8 mol% in all of Ce₂0₃ and/or Nd₂0₃ were prepared by firing at 1,500°C for 25 h in a reducing atmosphere, after drying and denitrating the mixed solutions of aluminum, cerium and/or neodymium nitrates, and then pelletizing the powder mixtures under 36 MPa.

The obtained ceramic assemblages were analyzed by X-ray diffraction to determine crystalline phases produced. Tested also was the leachability of the assemblages in a nitric acid solution of pH=1 and deionized water at 150°C. These observations revealed that Ce prefers to be stabilized in magnetoplumbites, while Nd in perovskites. To obtain even denser assemblages, it seems necessary to suppress the formation of magnetoplumbites as much as possible. Leaching study showed that the relatively high dissolution rates of the forms in the solution of pH=1 g/10^?6g/cm²·d) were probably due to the innate reactivity of alumina with acid, and perovskites were more prone to leaching than magnetoplumbites. In the deionized water, the elemental release rates from the forms are lower than 10-⁷g/cm²·d, and not influenced practically by the presence of perovskites.     

Film Boiling Characteristics of Potassium Droplets on Heated Plate

For providing background information on the possible vapor explosion in the event of a core disruptive accident of LMFBRs, an experiment was conducted on the film boiling characteristics of liquid metal potassium in association with the Leidenfrost phenomenon. In a steel container filled with Ar gas, K droplets were put on a joule-heated plate of 316-SS or Ta. The behaviors of droplet were observed by a camera and a color VTR through viewports. The experimental conditions were the Ar pressure 1 bar, the initial K temperature 350~760°C, and the plate temperature 900~1,250°C for 316-SS and 1,100~1,600°C for Ta.

Stable film boiling known as Leidenfrost phenomenon was observed for a high temperature condition of the plate, whereas an instantaneous break-up of droplet with extensive vaporization occurred for a low temperature. The heat transfer characteristics of film and transition boiling regions were obtained by estimating the heat flux from the volumetric reducing rate of droplet due to vaporization. The results in the film boiling region showed an appreciably good agreement with the prediction based on Bromley's expression combined with the theory of Baumeister et al. The minimum film boiling temperature and heat flux were found to be about 1,300°C and 15 W/cm², respectively, for a droplet size of 0.15 cm³.     

The Behavior of Iodine in Adsorption and Desorption by Graphite

Using tracer technique, studies were undertaken on the adsorption of iodine on natural graphite at temperatures ranging from 27° to 900°C under pressures from 0.03 to 1.4 mmHg, and on the desorption of adsorbed iodine from the same specimens into either vacuum or Ar stream. The specimens were heated up either isothermally, stepwise, or at constant rate of temperature rise, up to 1,100°C. The resulting data accorded well with a theoretical treatment presuming desorption from heterogeneous surface, considering the desorption as an n-th order irreversible reaction having distributed activation energies. The iodine is considered to be held on the graphite by physical adsorption below about 450° C, and above that temperature, by chemisorption through bond formation between iodine and carbon atoms. The heat of adsorption and the activation energy of desorption obtained was found to be a function of surface coverage. The presence of Ar gas was observed to inhibit chemisorption of iodine and to decrease the activation energy of desorption of physically adsorbed iodine.     

Behavior of Helium Release from Simulated Radioactive Waste Glasses

Rates of He release from two kinds of simulated high-level radioactive waste glasses were measured in the temperature range from 573 to 753K by a mass spectrometric method. Diffusion coefficients of He in the glasses were determined from the fractional release data. The diffusion coefficients were given as a function of temperature by the equations:

D/(m²·s^?1) = 2.3 × 10^?6 exp (?71.1±2.1 kJ/RT) for P0798 glass

=9.8 × 10^?7 exp (?60.2±2.1 kJ/RT) for P0504 glass.

In addition, solubilities of He in the glasses were determined from the measurement of the total amount of released He and found to be

S/(atoms/m³) = 2.0 × 10²² for P0798 glass

=2.2 × 10²² for P0504 glass.

at 693 K and 1.7 × 10⁵ Pa He.

The diffusivities and solubilities of He in the glasses were, respectively, several and an order of magnitude lower than silica, alkali silicate and Pyrex glasses, which properties are notable for the radioactive waste glasses.     

Angular Distribution of Gadolinium Vapor Produced by Electron Beam Heating

Static corrosion tests were performed for the glass phase of a simulated waste form of non-combustible radioactive low-level waste to study a basic aqueous corrosion behavior. The waste form, which was fabricated from simulated waste sample by use of in-can type induction-heated melting, consists of two separated phases; a glass phase and a metal phase. Tests were performed for the glass phase from two types of the waste form with different chemical composition at 35°C and S/V ratio of 2,600 m^?1. The glass phase with both types showed an incongruent dissolution similar to conventional high-level radioactive waste (HLW) glasses, i.e., the normalized elemental mass loss (NL_i) for soluble elements such as B and Na continued to increase after the saturation of insoluble elements such as Si, A1 and Ca. The NL_i for B increased in proportion to the square root of time except for early stage, which suggests that the rate of the long-term dissolution or alteration may be controlled by a diffusion process. Potential secondary phases forming as the results of incongruent dissolution were estimated to be kaolinite and calcite by comparison of the measured solution data with the thermodynamically calculated phase stability relationships. These results suggest that the glass phase has a potential chemical durability not so different from conventional HLW glasses.     

Separation of Isotope Mixture in a Free Molecular Jet

With the aim of finding the most desirable set of conditions for powder metallurgical preparation of graphite matrix fuels containing thorium carbide, the properties and structure of fuel samples prepared by compacting various mixtures of powdered material followed by heating were investigated.

It was found that of the resulting compacts those consisting of (1) purified and milled natural graphite powder, (2) metallic thorium powder and (3) furfuryl alcohol as binder showed the best properties after heating to 600°C. These compacts possessed 1.90–2.40 g/cc density, 3.5–10.0 kg/mm² Vickers hardness, a thermal expansion coefficient at 900°C of about 14×10^?6/°C in the direction parallel to, and 2.5times;10^?6/°C normal to pressing direction, and a thermal conductivity, in the same temperature range, of 0.02 (parallel) and 0.07cal/cm-sec°C (normal).

Since these favorable properties and structural features are not impaired by heating to about 2,000°C subsequent to treatment at 600°C, it can be said that a new, simple and effective set of conditions for fuel preparation has been found.