Technetium-99 is considered as one of the most dangerous nuclear environmental pollutants due to its long half-life (210,000 y.) and high mobility in aqueous solutions under oxidizing conditions. Development of sorbents, which are capable of irreversible uptake of Tc and further direct conversion into durable ceramic waste forms, is an important field of research. Titanate ceramic doped with up to 10 wt. % Tc was successfully synthesized using Layered Hydrazinium Titanate, LHT-9 (PCT/EP2010/001864) as starting precursor. LHT-9 is a new advanced compound of general formula (N2H5)1/2Ti1.87O4xH2O containing 6-7 wt. % of hydrazine chemically incorporated into a TiO2-based matrix. It was demonstrated that LHT-9 (5g/l) can reductively adsorb up to 90.2 wt. % of Tc from aqueous solutions containing 0.5g Tc/l. The obtained adsorption products can be easily converted into stable titanate ceramic by one-step sintering in argon atmosphere at 1200°C. Phase and chemical composition of synthesized Tc-doped ceramic are discussed.

Nickel base alloys are considered among candidate materials for engineered barriers of nuclear repositories. The localized corrosion resistance is a determining factor in the materials selection for this application. This work compares the crevice corrosion resistance of selected nickel base alloys, namely 625, G-30, G-35, C-22, C-22HS and HYBRID-BC1. The crevice corrosion repassivation potential (ER,CREV) of the tested alloys was determined by the Potentiodynamic-Galvanostatic-Potentiodynamic (PD-GS-PD) method. The testing temperature was 60ºC and the chloride concentrations used were 0.1 M, 1 M and 10 M.

A linear relationship between ER,CREV and the logarithm of chloride concentration was found. ER,CREV increased linearly with PREN (Pitting Resistance Equivalent Number) in concentrated chloride solutions. ER,CREV is the sum of three contributions: ECORR*, η and ΔΦ. ECORR* and η increased linearly with PREN, while ΔΦ increased linearly with PREN for concentrated chloride solutions, not showing a definite trend with PREN for the less concentrated solutions.

Cementation is the dominant encapsulation technology for UK intermediate level waste (ILW) from nuclear operations. However, Portland cement (PC) based encapsulation systems encourage corrosion of reactive metals such as uranium, aluminium and magnesium. Thus, the development of alternative systems is required. A candidate alternative system is magnesium potassium phosphate cement (MKPC) and in this paper the results of a wide testing programme to establish a formulation envelope for MKPC is presented. This envelope determines a region where bleed production, fluidity and workability are within a range suitable for plant processing and encapsulation. Additionally, the evolution of the compressive strength with curing age for multiple formulations and curing temperatures was determined. Fly ash (PFA) inclusion was found to be highly beneficial to slurry workability and its addition also improved strength; high PFA content formulations reached compressive strength values of 34 ±2 MPa and 44 ± 2 MPa at 28 days and 180 days respectively. A decreased in achieved compressive strengths was observed on increasing the curing temperature from 47 °C to 72 °C.

The Tritium, Carbon-14 and Cobalt-60 content of a trepanned sample from one of the Wylfa Magnox reactor have been experimentally determined using beta liquid scintillation counting and gamma spectroscopy. The WIMS9a reactor code and FISPACT-2007 neutron activation software have also been used to calculate this inventory for the sample, considering only a model which is isolated from the reactor circuit. Comparison between experimental and calculated results has shown that the calculated values for 14C are within 26%, 60Co within 24% and 3H 120%. These results show that the original impurity levels are sufficient to explain the experimentally determined end of life activity, without additional consideration of contamination from other materials in the reactor circuit, in this type of simulation. Additionally the calculations show that the production of 14C from 14N is approximately equal to that produced from 13C. These results are only applicable to the isolated system models developed here, and do not explicitly model existing reactor conditions, where external operating conditions may interact with the graphite and the core environment

At INFN-LNS we have developed a new kind of gamma mini-sensor, which allowed to implement a prototype array of modular sensors for online monitoring of radioactive waste. The sensors are based on Silicon PhotoMultipliers and scintillating fibers sensitive to gamma radiation, that have to be placed in a repository in shape of a fine grid, around each single waste drum. We simulated the counting response of thin scintillating fibers and verified the results with measurements using radioactive sources. Front-end electronics and an FPGA-based counting system were developed to handle the field data, also implementing data transmission, a graphical user interface and a data storage system. Moreover a test of four sensors in a real radwaste storage site was performed with promising results.

The safe management and disposition of used nuclear fuel and/or high level nuclear waste is a fundamental aspect of the nuclear fuel cycle. The United States currently utilizes a once-through fuel cycle where used nuclear fuel is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. However, a decision not to use the proposed Yucca Mountain Repository will result in longer interim storage at reactor sites than previously planned. In addition, alternatives to the once-through fuel cycle are being considered and a variety of options are being explored under the U.S. Department of Energy’s Fuel Cycle Technologies Program.

These two factors lead to the need to develop a credible strategy for managing radioactive wastes from any future nuclear fuel cycle in order to provide acceptable disposition pathways for all wastes regardless of transmutation system technology, fuel reprocessing scheme(s), and/or the selected fuel cycle. These disposition paths will involve both the storing of radioactive material for some period of time and the ultimate disposal of radioactive waste.

To address the challenges associated with waste management, the DOE Office of Nuclear Energy established the Used Fuel Disposition Campaign in the summer of 2009. The mission of the Used Fuel Disposition Campaign is to identify alternatives and conduct scientific research and technology development to enable storage, transportation, and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The near-and long-term objectives of the Fuel Cycle Technologies Program and its’ Used Fuel Disposition Campaign are presented.

In order to enhance the safety of geological disposal and the reliability of the safety assessment carried out for each stage of the geological disposal project, an assessment methodology focused on a sensitivity analysis and an evaluation of importance, which incorporates “system understanding” and “information feedback” into the existing assessment approach, has been developed in this study. In this paper, the assessment methodology and the assessment results as applied for the transuranic (TRU) waste disposal system in Japan will be described. In the sensitivity analysis, an approximate analytical solution was used in order to understand the response characteristics of the engineered barrier system (EBS). In the importance evaluation, important factors relating to the robustness of system safety were identified based on the response characteristics. Furthermore, important features, events, and processes (FEPs) related to such factors and high-impact scenarios were identified based on the information accumulated through “system understanding” and “information feedback”. Based on this approach, the robustness of the TRU waste disposal system was assessed and measures for improving the robustness were identified.

Using soft X-ray spectroscopy, we have studied the reactions between Cu2O and CuO with an aqueous solution of 1mM Na2S. Exposure to dissolved sulfide for several hours clearly reduces Cu(II) to Cu(I) compounds. However, there are still strong oxygen signals in the spectra, indicating a more complex surface composition than simple copper sulfides.

Ni-Cr-Mo alloys offer an outstanding corrosion resistance in a wide variety of highly corrosive environments. Alloys 625, C-22, C-22HS and HYBRID-BC1 are considered among candidates as engineered barriers of nuclear repositories. The objective of the present work was to assess the effect of temperature on the crevice corrosion resistance of these alloys. The crevice corrosion repassivation potential (ER,CREV) of the tested alloys was determined by the Potentiodynamic-Galvanostatic-Potentiodynamic (PD-GS-PD) method. Alloy HYBRID-BC1 was the most resistant to chloride-induced crevice corrosion, followed by alloys C-22HS, C-22 and 625. ER,CREV showed a linear decrease with temperature. There is a temperature above which ER,CREV does not decrease anymore, reaching a minimum value. This ER,CREV value is a strong parameter for assessing the localized corrosion susceptibility of a material in a long term timescale, since it is independent of temperature, chloride concentration and geometrical variables such as crevicing mechanism, crevice gap and type of crevice former.

Nuclear waste isolation programs both inside and outside the United States have provided evidence that there are many geologic options for a repository, but virtually all of them rely to some degree on an engineered barrier system (EBS) to isolate and/or retard the migration of radionuclides to the biosphere. At Yucca Mountain, the design of the EBS was unexpectedly challenging because of uncertainties in quantitatively determining the local environment of the EBS particularly during the thermal pulse. The EBS design for the Yucca Mountain site evolved from a thin-walled, limited-lifetime, corrosion-resistant canister through a corrosion-allowance canister, to the present design, which may have a lifetime of more than 106 years. The EBS proposed for the Yucca Mountain repository has many individual sub-barriers, beginning with the spent fuel and waste, the cladding of the spent fuel, the geometry of the package, etc. The anticipated modes of degradation of engineering materials, including corrosion of the fuel, of the canister, and of the drip shield proposed specifically for the Yucca Mountain project, and the consequences of the materials degradation on the performance of the repository are presented. The roles of conservative modeling and simplifying assumptions for radionuclide mobilization and transport in the EBS on characterization of the source term are addressed.

The container is the only absolute barrier in the multi-barrier system that forms the basis of all nuclear waste disposal strategies. The selection of an appropriate container material is therefore of utmost importance. Some of the factors that underlay the choice of container material are discussed, ranging from the properties of the near-field and host rock to the desired or expected containment period.

There has been a trend towards the specification of container materials that will corrode actively under repository conditions, such as copper and carbon steel. Passive materials, such as titanium and nickel alloys and the various stainless steels, have found less acceptance and the reasons for this emphasis on active materials are also discussed.

In selecting an appropriate container material, it is essential to understand the nature of the corrosive environment and how it evolves over time. The evolution of environmental conditions will also cause the corrosion behaviour of the container to change with time. For repositories in saturated environments, it will be argued that passive alloys can provide long container lifetimes without some of the disadvantages of some active materials such as gas generation and other adverse impacts on other barriers.

Finally, areas of future development and areas requiring additional study will be discussed.

In many countries, regulations for the management of nuclear waste require a performance (safety/risk) assessment to demonstrate the safety asserted to be provided by the sites/facilities proposed for handling, storing, and disposing of the wastes. However performance assessment can play a bigger role than solely demonstration of compliance with applicable safety standards in support of a regulatory decision (i.e., licensing of a waste management facility). Performance assessment can be an effective management tool during all phases of a waste management program: from development of national nuclear waste management policies; to programmatic environmental impact assessments associated with design and siting evaluations, site selection, and site characterization; to licensing and operation of facilities.

International experience has demonstrated that nuclear waste management programs are long-term efforts, lasting at least two to three decades from initial policy development to licensing and commencement of waste management and disposal operations. This experience has also demonstrated that consistent attention to, and integration of, initial component studies are necessary to provide a comprehensive total system analysis for programmatic environmental impact assessments and for licensing.

For nearly 40 years, Sandia National Laboratories has developed and applied a performance assessment methodology in numerous national and international nuclear waste management programs. These applications range from development and feasibility testing of environmental health standards to preliminary evaluation of waste disposal sites; to establishing the basis for demonstration of compliance; to informing licensing (compliance demonstration) decisions. In many of these applications the performance assessment methodology has also served as a management tool for confirming the added value of research and development investments.

This paper presents examples to illustrate how performance assessment has been used as an effective management tool through multiple phases of a nuclear waste management program.

The geological disposal of radioactive waste, based on a multi-barrier concept wherein the first barrier consists of the metal waste container and the final barrier the host rock, is widely considered the only viable solution.

Following disposal the risk will remain of the formation of gases due to corrosion and other processes. Research being carried out at the Czech Technical University in Prague (CTU), Centre of Experimental Geotechnics (CEG), the Josef underground laboratory as part of FORGE and other projects focuses on gas migration in underground areas and especially within the EDZ.

The research consists of several stages including the design of gas conductivity equipment destined for in-situ testing and the gas conductivity tests proper. An important part of the research involves the evaluation of a potential correlation between rock mass classification parameters and gas conductivity; a certain degree of correlation was identified in earlier research projects. The discovery of such a correlation would greatly assist in the future design of underground gas storage and deep radioactive waste facilities.

Classical molecular dynamics simulations were used to compare the fracture behavior of pristine and disordered specimens of a simplified nuclear glass. The disordered specimen is prepared in order to mimic the effects of accumulating displacement cascades. It is characterized by a decreasing Boron coordination and an increasing Na concentration in a modifying role. We observe an enhancement of the plasticity of the disordered glass and a decrease of the elastic limit, resulting in greater fracture toughness. The simulation findings are consistent with experimental results.

Some human activities can focus radioactive natural materials on their waste and products through technological processes by increasing the local background radioactivity at values that may compromise the health and safety of persons and the environment. The oil industry waste containing technologically enhanced naturally occurring radioactive materials (TENORM) are high volume and low levels activity cannot be classified and arranged according to the Brazilian standard for classification of waste generating a stalemate in its management in Brazil. This work has identified technological alternatives for the disposal of waste containing TENORM available in the world and examined its use in the oil industry in Brazil, considering the Brazilian legislation.

Microwave dielectric heating was utilized to synthesize potential host phases for halide radioisotopes. Synthesis of Pb5(VO4)3I was successfully achieved, without the use of a sealed container, using a modified domestic microwave oven (DMO) operating at 2.45 GHz. Rapid synthesis of Pb5(VO4)3X (X = F, Cl, Br) was also achieved using a commercially available microwave muffle furnace, also operating at 2.45 GHz. The combination of rapid heating rate and inverse temperature profile characteristic of microwave heating was found to promote formation of the target phases whilst retarding the volatilization of the halide species. Pb5(VO4)3I ceramic bodies produced in the DMO exhibited a heterogeneous zoned microstructure, whereas Pb5(VO4)3X phases with X = F, Cl, and Br fabricated in the microwave muffle furnace were dimensionally uniform suggesting this could be a promising route to fabricating single phase, dense halide containing ceramics.

In the frame of the Safety and Feasibility Case 1 development, the Belgian agency for radioactive waste (ONDRAF/NIRAS) has to demonstrate that the geological disposal of spent nuclear fuel in the Boom Clay host rock following the Supercontainer design provides sufficient radiological safety in the long term. This safety relies on the containment of the radionuclides by the overpack, the limitation of radionuclide release from the waste matrix, the diffusion controlled transport of the radionuclides, and their low solubility and sorption on the Boom Clay host rock. This paper presents an evaluation of the main processes considered for the limitation of radionuclide release by the Spent Fuel in Supercontainer conditions, characterized by the presence of a concrete buffer and hyperalkaline pore water. We present a description of the main expected processes with their potential impact on the fuel alteration mechanisms, and potential reference fuel matrix alteration rates and effective fuel surface areas for use in safety and performance assessment models.

It is well discussed about biological effect to high-level radioactive waste (HLW) disposal and known that the biofilm is considered to be the uncertain factor to estimate for migration of radioactive elements. The objective of this research is to estimate the microbial effect of Cs migration in groundwater interacted with rock surface. Specially, we focus on Cs behavior at the rock surface surrounded by biofilm. The most important factor is the Cs sorption and diffusion to the microbe and/or their biofilm. Generation of bio-colloid absorbed with Cs and retardation of Cs by their matrix diffusion in rock will be influenced by these phenomena. We introduce about scenario analysis for biofilm and a simple Cs diffusion test with and without sulfur reducing bacteria (SRB) which is well known as easy to produce biofilm on the rock surface in order to clarify the existence effect of the bacteria at the rock surface. The Cs diffusion experiment, using Desulfovivrio desullfuricans as SRB, indicated that microbial effect was less to through their biofilm in the experimental condition. We consider that Cs is easy to contact the rock surface even if surrounded biofilm and not effect to retardation by matrix diffusion scenario.

The behavior of Eurobitum bituminized radioactive waste under geological disposal conditions is studied in water uptake tests at the Belgian Nuclear Research Centre SCK•CEN to assess the feasibility of geological disposal for the long-term management of this waste. The hydro-mechanical behavior of this waste is affected by the continuous evolution of the rheological properties of bitumen due to radio-oxidation (i.e. ageing).

The effect of the bitumen ageing degree on the kinetics of the water uptake, swelling and NaNO3 leaching is investigated in water uptake tests with ∼30 years old radioactive samples and inactive samples that were artificially aged. The first results of swelling and NaNO3 leaching of (i) radioactive samples that have been hydrated for more than 2 years at a constant total stress of 2.2 MPa, and (ii) a thermally aged sample that has been hydrated for ∼1.5 years under nearly zero effective stress conditions, revealed lower swelling and higher leach rates for these samples compared to non-aged samples. The effect of ageing on the osmotic efficiency of bitumen as a semi-permeable membrane is less pronounced when swelling of the samples is limited, and changes in time, probably because of the formation of low porosity layers, which seems to mask the difference in bitumen membrane efficiency of aged and non-aged bituminized waste.

A process has been developed at AWE for the immobilisation of halide containing wastes arising from the reprocessing of plutonium. Initially, the wastes are calcined with a calcium phosphate to form a number of target host phases, β-tricalcium phosphate (β-TCP) and apatite for the immobilisation of actinides, and apatite and spodiosite for halide.

These mineral phases are then mixed with a glass binder, cold pressed and sintered to form a monolithic waste form. Two glass binders GTI/168, a sodium alumino phosphate glass and GTI/206, a sodium calcium phosphate glass were compared to optimise the halide retention in the waste-form. Analysis from powder X-ray Diffraction (PXRD) and scanning electron microscopy (SEM) showed that neither glass stabilises spodiosite. However, GTI/206 glass retains 10 wt.% more apatite and results in a much smaller proportion of the non-chloride bearing whitlockite phase than GTI/168.

In all compositions where GTI/168 glass was used as a sintering aid, the sodium deficient and calcium enriched glass was present as an amorphous matrix phase.