Leach Rates of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis.

The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C.

The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.     

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.     

Leaching Model of Nuclear Waste Glass

The three-dimensional spherical leaching model was developed on the basis of the experimental results for nuclear waste glass. Emphasis was placed on proposing a model for the growth of surface layers and for an immobilized reaction inside these layers. The diffusion equations derived from the modeling were solved numerically using the Crank-Nicholson implicit method and the resulting equations were implemented in a computer code named LEACH-2. To determine the location of the moving boundary, which was one of the most important parameters in this model, two simple examples were presented. One was the location of the moving boundary under the reaction controlled mechanisms and the other was that under the diffusion controlled mechanisms. The computed results using the latter example were in good agreement with the previous experimental data.     

Crystalline Phase in Surface Layers of Glass Waste Form

An integral test of γ-ray production data of iron in the latest version of Japanese Evaluated Nuclear Data Library (JENDL-3.2) has been performed by means of a shielding benchmark analysis of KfK leakage neutron and γ-ray spectrum measurements from iron spheres with a ²⁵²Cf source in the center. Two comprehensive systems which consist of a continuous-energy Monte Carlo method and a multi-group Sn transport method have been adopted in this benchmark analysis. For comparison, analyses with JENDL-3.1, FENDL-1 and ENDF/B-IV have been also carried out. The calculation using JENDL-3.2 showed a good agreement with the experiment. It has been concluded that the γ-ray production data of iron in JENDL-3.2 were applicable for use of shielding designs and analyses of the fission neutron source problem.     

Liquid Metal Extraction for Removal of Molybdenum from Molten Glass Containing Simulated Nuclear Waste Elements

Laboratory-scale experiments for removing Mo and MoO3 from molten borosilicate glass were performed using liquid Cu as an extractant. Removal of Mo from the simulated HLW glass containing oxides of Nd, Fe, Zr, Mo, Sn, Ni, Sr, Cd, Ru, and Se was also performed, and the fractions of these elements transferred into Cu were examined. Mixtures of Cu anda ternary SiO₂-B₂O₃-Na₂O glass containing metallic Mo or MoO₃ were heated in an alumina crucible at 1,673K in an Ar environment. The amounts of Mo and MoO₃ added to 10 g of the ternary glass were fixed at 0.1 and 0.15 g, respectively. As for the glass containing metallic Mo, more than 90% of Mo was extracted into liquid Cu. Spherical Cu metal buttons containing Mo formed on the bottom of the crucible when Cu was added at more than 10 times that of Mo on a mass basis. Removal of Mo from the glass containing MoO3 was also achieved by the addition of Si as a reducing agent for the reduction from MoO₃ to Mo. The fraction of Mo extracted into liquid Cu depended on the molar ratio of Si to Cu added to the glass. The fraction increased up to 84% with an increase in the molar ratio of Si/Cu. However, the excess addition of Si may enhance the chemical interaction between the metal phase and the glass phase, and some of the metal phase containing Mo remained in the glass phase without forming a metal button. The optimum molar ratio of Si/Cu that produces the highest removal fraction was found to be approximately 0.5. Almost the same removal fraction of 88% was obtained from the simulated HLW glass under the condition of Si/Cu = 0.5. Nearly 100% of Ru was extracted into Cu with Mo, while Sr, Zr, and Nd were hardly extracted and remained in the glass.     

典型贵金属元素在玻璃固化体中的溶解度及其对玻璃性质的影响

本文主要针对高放废液中贵金属(PGMs)的含量、存在形态,玻璃固化体中PGMs的溶解度、存在形态、颗粒大小等化学行为进行归纳,同时针对玻璃固化过程中PGMs对玻璃性质(流变、粘度、热导、电阻)的影响进行总结。该文为研究典型贵金属元素在玻璃中的溶解度及其变化规律、典型贵金属元素在玻璃中的化学行为及沉淀机理以及典型贵金属元素对玻璃性质的影响研究提供调研基础,可为我国玻璃固化技术的研发提供前期的理论依据。     

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.     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

Determination of Heat Capacity of Simulated Radioactive Waste Glasses by Drop Calorimetry

The enthalpies of two kinds of simulated radioactive waste glasses have been measured with an ice isothermal calorimeter at temperatures of 424~875 K by means of drop calorimetry. The fitting functions for the enthalpies per gram have been determined by the least squares fitting. Specific heat and average molar heat capacity have been obtained. It is likely that heat capacity of waste glass remarkably depends on its composition especially above its glass transition temperature.     

Three Year Experimental Study of Leaching Phenomena from Low Level Radioactive Cement-Based Waste Forms

Abstract

This study was carried out in order to demonstrate the safety of homogeneous cementbased waste forms (hereinafter called cement forms) for BWR's low level radioactive wastes as engineered barriers. Eighteen full scale simulated cement forms were manufactured with the addition of ¹³⁷Cs, ⁶⁶Co and ⁹⁰Sr.

Leaching tests on these forms were carried out for approximately three years. In order to study the relationship of leachability to environments at disposal sites, this Three Year Leaching Test was conducted for three kinds of environmental conditions, sea water, land water and soil. After the tests, all of these forms were cut to measure the distribution of the radionuclide's density within them. In case of the soil tests, the distribution of radionuclide in the soil was also measured.

The radionuclide leachability results reveal that ⁶⁰Co was almost completely retained in the cement forms and that ¹³⁷Cs leached from cement forms was mostly adsorbed by the soil. On the other hand, ⁹⁰Sr was not trapped in the forms and leaked through the soil around them in retard. This study also showed that simulated cement forms buried in the soil were more physically and chemically stable, and had longer term stable radionuclide containment capability than those which were submerged in sea or land water.     

放射性废物热等离子体处理熔融炉温度分布数值模拟及熔渣玻璃化配方初步研究

目前核电厂放射性废物处理工艺具有高增容性的特点,为最终废物处置带来很大的压力。针对高增容的处理现状,介绍了放射性废物减容处理技术的研究重点——热等离子体技术;通过数值模拟分析热等离子体熔融炉内的温度分布,给出固定床熔融炉关键部位的最高可能温度约为1 445℃,结合可选耐火材料探讨了炉体建造的可实现性。选取核电站3种典型的放射性技术废物进行模拟玻璃化配方实验,在限定的熔融温度条件下,得到符合我国核行业标准要求的玻璃固化体。     

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.     

Comparison of Nuclear Waste Glass Leached in Natural Groundwater at 14°C with That in Synthesized Groundwater at 70°C

The influence of fast neutron irradiation was studied for CdTe radiation detectors in the range of fluence 1.0 × 10⁸-1.8 × 10¹¹ n/cm². The effect of γ-ray heavy irradiation toward the detector was also investigated for the comparison. Americium-241 and ¹³⁷Cs γ-ray spectra measured before and after irradiation were compared to examine the change of the detector performance. A pulse height and a rise time of a radiation signal were simultaneously measured in order to evaluate the transport property of both elections and holes. Peak shift toward lower energies was observed in the ²⁴¹Am γ-ray spectrum after fast neutron irradiation. Electron μτ product showed large decrease with increasing the neutron fluence, while the one for holes remained unchanged. For the γ-ray irradiation, the spectrum shape of ¹³⁷Cs changed due to the influence of carrier detrapping. Experimental results support that the electron-trapping center generated by fast neutrons degrades the detector performance of the CdTe detector.     

Non-technical Considerations in Regulation of Very Low Level Radioactive Waste: Comparative Analysis of Japan and the United States

With presently over 400 commercial nuclear power reactors being operated worldwide, many of which will retire within the next 50 years, the future generation of world nuclear energy depends upon strategies for low level waste management and decommissioning of those reactors. These strategies must address issues such as: economical feasibility, environmental and health standards, post-decommissioning land and facility usage. This paper considers those issues in the context of the inherently intertwined social and technical characteristics, with an emphasis on the management of very low level wastes. Until now 70 commercial power reactors have been decommissioned, however, most have been relatively small in comparison to those that will be preparing for decommissioning in the next 50 years. The resulting materials will add to the already increasing amounts of waste and material from nuclear reactors. Since the move to harmonization of clearance level regulation may have critical impacts on the environment and health as well as decommissioning costs and priority setting this paper examines both the areas of consensus and uncertainties between countries regarding very low level waste regulations for recycling of materials arising from decommissioning, against the background of international discussions. In conclusion, we discuss the need for deliberation regarding the assumptions and cultural factors.     

Reduction of Cesium Leaching Ratio from Cementitious Resin Forms Using Natural Active Silica

Cesium adsorption behavior of active silica, which is a natural acid clay composed of cristobalite and quartz, was evaluated for its applicability as Cs adsorbent to be added to cementitious waste forms containing spent ion exchange resin. Since active silica carried the Cs exchangeable silanol group (—SiOH) originally, the Cs distribution coefficient was remarkably high (<10⁴). It increased in saturated Ca(OH)₂ solution, simulating the cement paste, due to formation of new silanol groups. With its addition to the cementitious forms with ¹³⁴Cs adsorbed ion exchange resin solidified by slag cement, the Cs leaching ratio was reduced to below 1/10 that without active silica.     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

Diffusion Behavior of Cesium,Strontium and Cobalt in Pellet Waste Package

A leaching model for a radioactive pellet package for nuclear power plant wastes was studied. Calculated results obtained in term of the model were compared with experimental ones, where volumetric averages of diffusion coefficients in the Na₂SO₄ pellet and in the cement-glass were used to solve the diffusion equation in the pellet package. Since Na₂SO₄ pellets dissolved quickly, the pellet part were simulated by saturated Na₂SO₄ solution. The leaching ratios were meaured with different volume ratios of cement-glass in the package to discuss the validity of the model. The cement glass diffusion coefficient and distribution factor, which were required to solve the equation, were also measured. The calculated results agreed with the experimental ones within a 20% error.