Low Activity Radioactive Waste Research Articles

Rheology, Setting, Heat of Reaction, and Compressive Strength of a Geopolymer Radioactive Waste Form

This work contributes to our understanding of the properties of geopolymers as waste forms made of three pozzolans, to be used, e.g., to immobilize liquid low-activity radioactive waste. A binary blast furnace slag and metakaolin geopolymer composition was tailored to achieve flexible rheological properties and workability for in-can mixing or mixing/pumping applications by adding fly ash, a third pozzolan. We investigated quantitatively the early stages of geopolymerization (before and after setting) of alkali-activated slag–metakaolin–fly ash pozzolans. The effect of fly ash glass particle size on material properties was studied as well. Measurements include heat of reaction, compressive strength, yield stress, plastic viscosity, rheological setting, as well as initial and final setting following the Vicat method. A rheological time of initial setting is suggested for geopolymers, in analogy to Portland cement. The results of Vicat needle and rheological measurements are compared and quantitatively related to heat and set time data. Maximum heat flow and compressive strength are linearly correlated.

Comparative analysis of non-destructive methods to control fissile materials in large-size containers

The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one.

Development of a purification technology for treatment of medium- and low-activity radioactive waste of radiochemical production from Co-60 and Cs-137

The technological flowchart of purification of medium- and low-activity waste from Co-60 and Cs-137 is developed and introduced. The developed purification scheme has been successfully tested using genuine medium- and low-level liquid radioactive waste of radiochemical production containing complexing and colloid forming components complexons, surfactants. The optimal conditions of the presented method of purification ensure reduction of the residual specific activity of 60Co and 137Cs radionuclides in the solution to less than 0,9 Bq per litre.

Investigation of clay bricks for storage facilities of radioactive-wastage

Shielding is an integral part of any radiation facility, it is the method used for radiation protection. Safe storage of nuclear waste materials has become an important issue after Fukushima event. In Japan and France surface facilities are being considered for the storage of low-activity radioactive wastes. At these surface facilities, compacted clay liner obtained from clayey soil has been used as a cap barrier. The purpose of clay liner is to minimize the permeation of moisture, especially due to rainfall, through the barrier up to the radioactive material. Thereby it also reduces the emanation probability of undesirable radon gas and gamma-rays into the atmosphere. The brainchild behind the present investigation is to find the possibility of the burnt clay bricks to be used in these surface storage facilities. The present study aims to evaluate the gamma-ray shielding behaviour of burnt clay bricks in energy range 0.001–15MeV. The samples of bricks were collected from four local brick-factories located near Bathinda-Ferozpur road, Punjab (India). An existing computer programme, Gamma Ray Interaction Coefficients (GRIC)-toolkit was modified for theoretical calculations of some interaction parameters for selected sample. Mineralogical study using X-Ray Diffraction (XRD) followed by Rietveld refinement (XRD-R) has been found in good agreement with Wavelength Dispersive X-Ray Fluorescence (WDXRF) investigation of samples for chemical compositions. XRD-R could be an interesting tool to recognize the radiation attenuation characteristics of materials. It has been concluded that local burnt clay bricks own the moderate gamma-ray shielding abilities while brick of clay (M-0) shows comparatively superior shielding behaviour. Thus brick of clay is the suitable candidate in the construction of environmentally-safe storage facilities for radioactive waste.

Technetium Incorporation into Goethite (α-FeOOH): An Atomic-Scale Investigation.

During the processing of low-activity radioactive waste to generate solid waste forms (e.g., glass), technetium-99 (Tc) is of concern because of its volatility. A variety of materials are under consideration to capture Tc from waste streams, including the iron oxyhydroxide, goethite (α-FeOOH), which was experimentally shown to sequester Tc(IV). This material could ultimately be incorporated into glass or alternative low-temperature waste form matrices. However, questions remain regarding the incorporation mechanism for Tc(IV) in goethite, which has implications for predicting the long-term stability of Tc in waste forms under changing conditions. Here, quantum-mechanical calculations were used to evaluate the energy of five different charge-compensated Tc(IV) incorporation scenarios in goethite. The two most stable incorporation mechanisms involve direct substitution of Tc(IV) onto Fe(III) lattice sites and charge balancing either by removing one nearby H(+) (i.e., within 5 Å) or by creating an Fe(III) vacancy when substituting 3 Tc(IV) for 4 Fe(III), with the former being preferred over the latter relative to gas-phase ions. When corrections for hydrated references phases are applied, the Fe(III)-vacancy mechanism becomes more energetically competitive. Calculated incorporation energies and optimized bond lengths are presented. Proton movement is observed to satisfy undercoordinated bonds surrounding Fe(III)-vacancies in the goethite structure.

Coal fly ash as a potential fixation reagent for radioactive wastes

Israel produces ∼1.3Mt/year of fly ash (FA), a byproduct of its coal-fired power plants. Due to increasing environmental regulations, these imported coals are processed to reduce the sulfur concentration (∼0.6%). These processing methods result in a material that has an enriched alkali/alkali earth component with pozzolanic and basic properties (pH>10.5).FAs are utilized worldwide, mainly as a cement additive for the construction industry. Recently, it was demonstrated that Class F FA can act as an excellent fixation reagent for acidic wastes from the phosphate or the oil regeneration industries. In the current work the potential utilization of Class F FAs as fixation reagents for low-activity radioactive waste from the nuclear industry was examined. Aqueous solutions containing radionuclide simulants: cesium (Cs+), strontium, (Sr2+), and cerium (Ce3+, Ce4+) were used as case studies with promising results. It is suggested that the primary fixation mechanism involves the aluminate/silicate anions at the FA surface. A novel experimental fixation approach utilizing the formation of carbonates is demonstrated and a new interaction mechanism is suggested based on the electrostatic interactions of the positively charged fine precipitates with the negatively charged FA surface.

Cast Stone Formulation for Nuclear Waste Immobilization at Higher Sodium Concentrations

Low activity radioactive waste at U.S. Department of Energy sites can be immobilized for permanent disposal using cementitious waste forms. This study evaluated waste forms produced with simulated wastes at concentrations up to twice that of currently operating processes. The simulated materials were evaluated for their fresh properties, which determine processability, and cured properties, which determine waste form performance. The results show potential for greatly reducing the volume of material. Fresh properties were sufficient to allow for processing via current practices. Cured properties such as compressive strength meet disposal requirements. Leachability indices provide an indication of expected long‐term performance.

Geochemical composition of natural waters near a storage site of low-activity radioactive wastes

Sampling data on subsoil and surface water near a storage site of low-activity liquid radioactive wastes, receiving neutralized uranium-containing pulps are given. Thermodynamic calculations are used to describe the behavior of major polluting and petrogenic components in the drainage system, allowing the occurrence forms of elements in solution and/or solid phase to be determined. The results suggest a positive effect of natural geochemical barriers hampering the process of uranium migration into natural waters. Over-all, the technogenic-natural system under consideration was found to effectively cope with the removal of uranium-containing compounds from industrial water.

POLICY DEVELOPMENT FROM THE INDUSTRY PERSPECTIVE

The responsibility and burden for implementation of any option for the disposition of low-activity radioactive waste will primarily reside with the industries that generate the waste and those that provide the necessary waste management and disposal services. Unfortunately, there is no general consensus among these industries on the multiple policy issues that are critical to the application of an effective management and disposal program. This paper discusses the various policy issues that affect these industries and the problems that will likely occur during the course of their activities. These policy issues include public and political concern, opposition, and reaction; multiple and sometimes conflicting government agency jurisdiction, regulation, and policy; issues relating to the transition to a new comprehensive system; conflicts between the interests of the various industry groups; and operational, radiation safety, and regulatory implementation issues. For these reasons, the acceptable options may be limited. Case studies of various waste streams are included to show the specific impact and problems that these issues can create.

IMPROVING THE REGULATION AND MANAGEMENT OF LOW-ACTIVITY RADIOACTIVE WASTES

This paper summarizes the first phase of a study in progress by a committee of the National Research Council's Board on Radioactive Waste Management. The Board initiated the study after observing that statutes and regulations administered by the federal and state agencies that control low-activity radioactive wastes have developed as a patchwork over almost 60 y. These controls usually reflect the enterprise or process that produced the waste rather than the waste's radiological hazard. Inconsistencies in the regulatory patchwork or its application may have led to overly restrictive controls for some low-activity wastes while others were neglected in comparison. In the first phase of this study, the committee reviewed current low-activity waste inventories, regulations, and management practices. This led the committee to develop five categories that encompass the spectrum of low-activity wastes and serve to illustrate gaps and inconsistencies in current regulations and management practices. The committee completed its first phase with four findings that will lead into the final phase of the study. This paper is excerpted from the committee's interim report that was issued in October 2003.

Influence of three ageing types (thermal oxidation, radiochemical and hydrolytic ageing) on the structure and gas transport properties of epoxy–amine networks

Three types of ageing (thermal oxidation, radiochemical ageing and hydrolytic ageing) have been applied to a series of epoxy networks which can be used for the embedding of low activity radioactive wastes. The effects on the chemical composition and structure were studied and related to the evolution of the gas diffusion properties of these networks. In the case of thermal oxidation, the detailed kinetic study of the degradation and the measurement of O 2 solubility and diffusivity allowed precise modelling of the oxidation profile.

Qualification Tests of Packages used for Transport and Storage of Low Activity Radioactive Wastes In INR Pitesti

AbstractRadioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international rules and standards. The technology for packaging and treatment of radioactive wastes can also be used at the Nuclear Power Plant Cernavoda. The qualification tests for the package used for transport and storage of radioactive wastes (low activity, up to 6·07 GBq (0·164 Ci) per drum) are described. The package used is a drum manufactured from 1 mm thick mild steel with the dimensions: height 915 ± 10 mm; diameter 600 ± 5 mm; volume 220 litres. To achieve adequate safety in the transport of radioactive wastes strict precautions must be taken to protect transport workers, the general public and the environment by ensuring, both in normal and in accident conditions, adequate containment and shielding of the materials, according to the IAEA Regulations for the Safe Transport of Radioactive Materials. The adequacy of the package design is therefore of primary importance, the design re...

Development and Evaluation of Containers for Radioactive Waste Disposal

The design and safety requirements for containers to be used for the final disposal and the interim storage of non-heat-generating waste are based in Germany on the criteria which are defined for the Konrad repository and for representative interim storage facilities, respectively. These requirements are compared with those defined by the transport regulations for shipping casks used for the transport of low activity radioactive waste. A proposal is submitted for package design requirements related to categories that cover all aspects of transport, interim storage and final disposal. Results of tests with a cubic waste disposal container (Konrad Type VI) carried out in 1991 and 1993 are presented in this context. Because of the considerable administrative differences for waste disposal containers on the one hand and for transport containers on the other hand and in order to apply quality assurance properly, a competent authority design approval or, at least, a manufacturer's design aproval to be registered by the competent authority that is responsible for quality assurance and control is proposed for IP-2, IP-3 and Type A packages. Such a requirement should be implemented in the IAEA Regulations for the Safe Transport of Radioactive Material.

Performance of Cement-Basedseal Systemcomponents Inawaste Disposal Environment

AbstractA grout based on portland cement, Class F fly ash, and bentonite clay was developed as part of the closure system of shallow subsurface structures for disposal of low-activity radioactive wastes. Heat output, volume change, and compressive strength of the sealing grout were monitored with time, at elevated temperature, and in physical models, to determine if this closure grout could maintain adequate volume stability and other required physical properties in the internal environment of the disposal structure.To determine if contact with an alkaline liquid waste would cause chemical deterioration of the sealing grout, cured specimens were immersed in a liquid waste simulant containing high concentrations of sodium and aluminum salts. Over a period of 21 days at 60 °C, specimens increased in mass without significant changes in volume. X-ray diffraction of reacted specimens revealed crystallization of sodium aluminum silicate hydrate.The new phase has an X-ray diffraction pattern similar to that of the commercial syntheticzeolite, Losod. Scanning electron microscopy used with X-ray fluorescence showed that clusters of this phase had formed in grout pores, to increase grout density and decrease its effective porosity. The testing was repeated at 100 °C for 5 days using a simulant containing sodium hydroxide and aluminum nitrate and the results were similar. Physical and chemical tests collectively indicate acceptable performance of this grout as a seal-system component.

ENVIRONMENTAL MONITORING ASSOCIATED WITH THE DISCHARGE OF LOW ACTIVITY RADIOACTIVE WASTE FROM WINDSCALE WORKS TO THE IRISH SEA.

Since 1952, low activity radioactive waste has been discharged by pipeline from Windscale Works to the Irish Sea off Sellafield. Environmental monitoring studies have been carried out throughout this period in order to assess the resultant radiation dose to man, and hence to estimate the discharge rate which would correspond to the maximum permissible radiation doses recommended by the International Commission on Radiological Protection. These studies have been concerned principally with those materials most likely to contribute substantially to the dose to man, and they have been supplemented by more widespread marine surveys and by laboratory studies aimed at gaining information about the movement and behaviour of radioactive nuclides in the marine environment. It has been concluded from these studies that the rate of discharge from Windscale Works to the Irish Sea cannot be permitted to exceed about 100,000 c/month of mixed fission products, and that specific restrictions are needed on a number of individual nuclides, namely, strontium-90 (about 2000 c/month), ruthenium-106 (about 5000 c/month) and cerium-144 (about 20,000 c/month). The present discharge rates are about 4000 c/month of mixed fission products, of which about half is ruthenium-106 and about 2 per cent strontium-90.