The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by capitalising on processes that minimise, reduce, recover and recycle without exposing employees, the public and the environment to unmitigated impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams be it gaseous, liquid or solid waste.

A new European Project has been launched in April 2008 under the 7 th EURATOM Framework Programme (FP7-211333), with a duration of four years, addressing the ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)’. The objective of this project is the development of best practices in the retrieval, treatment and disposal of irradiated graphite & carbonaceous waste-like structural material e.g. non-graphitised carbon bricks and fuel coatings (pyrocarbon, silicon carbide). It addresses both legacy waste as well as waste from future generations of graphite-based nuclear fuel. After defining the various targets for an integrated waste management, comprehensive analysis of the key stages from in-reactor storage to final disposal will then be undertaken with regard to the most economic, environmental and sustainable options. This will be supported by a characterisation programme to localize the contamination in the microstructure of the irradiated graphite and so more to better understand their origin and the release mechanisms during treatment and disposal. It has been discovered that a significant part of the contamination (including 14 C) can be removed by thermal, chemical or even microbiological treatment. The feasibility of the associated processes will be experimentally investigated to determine and optimise the decontamination factors. Reuse of the purified material will also be addressed to close the ‘Graphite Cycle’ for future graphite moderated reactors. The disposal behaviour of graphite and carbonaceous wastes and the improvement of suitable waste packages will be another focus of the programme. The CARBOWASTE project is of major importance for the deployment of HTR as each HTR module generates (during a 60 years operational lifetime) about 5,000 to 10,000 metric tonnes of contaminated graphite containing some Peta-Becquerel of radiocarbon. It is strongly recommended to take decommissioning and waste management issues of graphite-moderated reactors already into account when designing new HTR concepts.

The molded block fuel element (FE) also called monolith is a molded body, consisting of a substantially isotropic highly crystalline graphite matrix, fuel regions within the same matrix and cooling channels. The fuel regions contain the fuel in the form of coated particles which are well bonded to the remaining graphite matrix, so that both parts of the block form a monolithic structure. The monolith meets the requirements for the very high temperature reactors attaining helium outlet temperatures above 1000°C. To fabricate the molded blocks FE demonstration plant was erected and put into operation. The equipment worked without malfunction. The produced block FEs meet the specifications of GA machined block FEs. All specimens and block segments irradiated at temperature up to 1600°C and max. fast fluence E > 0, 1 MeV of 11×10 21 n/cm 2 show perfect behaviour without any damage.

Following the fabrication technique originally developed for HTR-molded block fuel elements a process was introduced to fabricate nuclear graphite with the aim to attain the improved irradiation stability above 3 × 10 22 n/cm 2 , E > 0,1 MeV and to increase corrosion resistance. Nuclear highly crystalline natural graphite is used. A phenol formaldehyde resin with additives of silicon or zirconium oxide powder serves as binder. The mixture thus obtained is isostatically consolidated into spheres and spheres are crushed to granules from which the 0.3 – 3 mm fraction is obtained. The granulate is hot molded into graphite bodies. The green bodies are heated to about 800 °C to carbonise the resin and subsequently annealed at 1900°C in vacuum. The key feature of the proposed process is based on the chemical affinity of binder coke with the structure obtained by carbonisation of green bodies. Consequently it reacts selectively in situ with the added SiO 2 , or ZrO 2 to carbides in vacuum at 1900°C. Silicon carbides and zirconium carbides are characterised by high mechanical strength and very good resistance to corrosion. The properties of reactor graphite, such as density, mechanical properties and in particular stability to fast neutron irradiation are considerably improved.

South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

The Pebble Bed Modular Reactor is being developed in South Africa. Important for PBMR implementation is a viable strategy for waste management. Irradiated graphite from fuel and structural components is too voluminous for practical treatment with traditional higher level waste methods and too radioactive to recycle. To clean the graphite of radionuclides, a two-step process is being pursued: (1) non-carbon radionuclides (activation products, fission products and actinides) are removed on an elemental basis by a chemical or microbial process. (2) 14 C requires separation at an isotopic level, which would be impractical with established methods (gaseous diffusion or centrifuge). PBMR is investigating a method of isotope separation using biofractionation. Preliminary experiments indicate that microorganisms do separate radioactive 14 C from stable 12 C. An aqueous slurry of 14 C-spiked, powdered graphite was “fed” to the microbes for 15–18 hours. The microbes initially contained only background levels of 14 C, i.e. orders of magnitude less than the slurry. In post-experiment analyses, a sample of the microbes was found to contain approximately twice the amount of 14 C present in the bulk slurry material. Experiments are underway to further quantify and verify these results, which indicate distinct microbial processing mechanisms for 14 C and 12 C. The most current results will be presented.

A low decay heat (implying Spent Fuel (SF) pebbles older than 8–9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks’ vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading / unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell.

For various countries, the direct disposal of high level nuclear fuel wastes is a key option for the backend of the fuel cycle. For HTR/VHTR reactors this is assumed for the introductory phase of this reactor system. However, closed fuel cycles or a separation of spent coated-particles from the graphite moderator and specific treatment, conditioning and disposal of these waste streams are also possible. In the European Community project “RAPHAEL”, fuel waste performance is going to be studied in depth, including post-irradiation fuel characterization, analysis of the stability and failure mechanism of coatings and of fuel kernels and overall performance of waste packages with compact fuel and/or only with fuel particles in geological disposal environments. Different confinement matrices for separated fuel particles (vitrification, SiC, ZrO2) have been adapted to limit release of radionuclides into groundwater at low temperatures over geological time spans. The investigations are limited to Low-Enriched Uranium (LEU) fuel with uranium oxide and uranium oxycarbide kernels that will allow higher burn-up, but may be more susceptible to leaching.

Generation IV Very High Temperature Reactors (VHTRs) are well-known for their flexibility with respect to feasible fuel cycle options. In this paper, the LEU- and TRU-fueled VHTR configurations are analyzed accounting for their capabilities to attain an extended single-batch OTTO (Once-Through-Then – Out) mode of operation without intermediate refueling. The requirement of waste minimization is imposed as one of the design constraints defining possible system configurations and deployment strategies. The resulting “used fuel” vectors are examined considering anticipated disposal options as well as viability of fuel reprocessing. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The presented analysis is focused on prismatic block core concepts for VHTRs. It is being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations.