EVIDENCE FOR TWO TELLURIUM SITES IN DILUTE LIQUID Te-Tl ALLOYS.

Recently, experimental evidence was presented which suggests that as the stoichiometric composition CuTe, NiTe, Tl//2Te and MnTe are approached from pure Te in the liquid state, substantial charge transfer takes place and Te exists in the form Te**y**31 ions with y close to 2. The system studied (Te-Tl) is one in which charge transfer localizes electrons on the tellurium and leads to semiconducting behavior at the stoichiometric composition Tl//2Te.

Near-complete transuranic waste incineration in a thorium fuelled pressurised water reactor

The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Incineration, and virtual elimination, of waste stockpiles is possible in a thorium (Th) fuelled critical or subcritical fast reactor. Fuel cycles producing a net decrease in TRUs are possible in conventional pressurised water reactors (PWRs). However, minor actinides (MAs) have a detrimental effect on reactivity and stability, ultimately limiting the quality and quantity of waste that can be incinerated. In this paper, we propose using a thorium-retained-actinides fuel cycle in PWRs, where the reactor is fuelled with a mixture of thorium and TRU waste, and after discharge all actinides are reprocessed and returned to the reactor. To investigate the feasibility and performance of this fuel cycle an assembly-level analysis for a one-batch reloading strategy was completed over 125 years of operation using WIMS 9. This one-batch analysis was performed for simplicity, but allowed an indicative assessment of the performance of a four-batch fuel management strategy. The build-up of 233U in the reactor allowed continued reactive and stable operation, until all significant actinide populations had reached pseudo-equilibrium in the reactor. It was therefore possible to achieve near-complete transuranic waste incineration, even for fuels with significant MA content. The average incineration rate was initially around 330 kg per GW th year and tended towards 250 kg per GW th year over several decades: a performance comparable to that achieved in a fast reactor. Using multiple batch fuel management, competitive or improved end-of-cycle burn-up appears achievable. The void coefficient (VC), moderator temperature coefficient (MTC) and Doppler coefficient remained negative. The quantity of soluble boron required for a fixed fuel cycle length was comparable to that for enriched uranium fuel, and acceptable amounts can be added without causing a positive VC or MTC. This analysis is limited by the consideration of a single fuel assembly, and it will be necessary to perform a full core coupled neutronic-thermal-hydraulic analysis to determine if the design in its current form is feasible. In particular, the potential for positive VCs if the core is highly or locally voided is a cause for concern. However, these results provide a compelling case for further work on concept feasibility and fuel management, which is in progress. © 2011 Elsevier Ltd. All rights reserved.

Relaxation and fracture characteristics in Co-based amorphous alloys

In this article a study of the fracture characteristics of Co66Fe4Mo2Si16B12 amorphous ribbon in the as-quenched state and after relaxation is presented. In the as-quenched state, the morphology of the crack surface shows a 'vein pattern' structure that corresponds to a large amount of plastic flow. After relaxation the surface morphology of the crack shows that when the temperature of the thermal annealing increases the plastic flow involved in the crack decreases. In the as-quenched state dynamic fracture characteristics (crack branching and stress wave induced crack) have been observed. These dynamic characteristics have not been observed in the relaxed samples but in the samples annealed at 250 °C for 20 min apart from the main crack, a crack along the width of the ribbon has been observed. © 2006 Elsevier B.V. All rights reserved.

The mechanical integrity of fuel pin cladding in a pulsed-beam accelerator driven subcritical reactor

The Accelerator Driven Subcritical Reactor (ADSR) is one of the reactor designs proposed for future nuclear energy production. Interest in the ADSR arises from its enhanced and intrinsic safety characteristics, as well as its potential ability to utilize the large global reserves of thorium and to burn legacy actinide waste from other reactors and decommissioned nuclear weapons. The ADSR concept is based on the coupling of a particle accelerator and a subcritical core by means of a neutron spallation target interface. One of the candidate accelerator technologies receiving increasing attention, the Fixed Field Alternating Gradient (FFAG) accelerator, generates a pulsed proton beam. This paper investigates the impact of pulsed proton beam operation on the mechanical integrity of the fuel pin cladding. A pulsed beam induces repetitive temperature changes in the reactor core which lead to cyclic thermal stresses in the cladding. To perform the thermal analysis aspects of this study a code that couples the neutron kinetics of a subcritical core to a cylindrical geometry heat transfer model was developed. This code, named PTS-ADS, enables temperature variations in the cladding to be calculated. These results are then used to perform thermal fatigue analysis and to predict the stress-life behaviour of the cladding. © 2011 Elsevier Ltd. All rights reserved.