Progress in Nuclear Energy 

About: Progress in Nuclear Energy is an academic journal published by Elsevier BV. The journal publishes majorly in the area(s): Nuclear reactor core & Nuclear reactor. It has an ISSN identifier of 0149-1970. Over the lifetime, 4226 publications have been published receiving 58316 citations.

The molten salt reactor (MSR) in generation IV: Overview and perspectives

Abstract: Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel. Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

Radiation effects in nuclear waste forms for high-level radioactive waste

Abstract: High-level nuclear waste in the United States comprises large volumes (tens of millions of cubic meters), high total activities (billions of Curies) and highly diverse and complex compositions. The principal sources of nuclear waste are: (i) spent nuclear fuel from commercial and research nuclear reactors; (ii) liquid waste produced during the reprocessing of commercial spent nuclear fuel; (iii) waste generated by the nuclear weapons and naval propulsion programs. The latter category now includes over 100 metric tons of plutonium and many hundreds of tons of highly enriched uranium from the dismantling of nuclear weapons. Most of these wastes will require chemical treatment, processing and solidification into waste forms for permanent disposal. The long-term effects of radiation on waste form solids is a critical concern in the performance assessment of the long-term containment strategy. In the case of spent nuclear fuel, the radiation dose due to the in-reactor neutron irradiation is already substantial, and additional damage accumulation during disposal is not anticipated to be significant; thus, this is not a subject addressed in this review paper. In contrast, the post-disposal radiation damage to waste form glasses and crystalline ceramics is significant. The cumulative α-decay doses which are projected for nuclear waste glasses reach values of 1016 α-decays g−1 in 100 yr. Similarly, crystalline waste forms, such as Synroc will reach values of > 1018 α-decay events g−1 in 1000 yr for a 20 wt% waste loading. These doses are well within the range for which important changes in the physical and chemical properties may occur, e.g. the transition from the crystalline-to-aperiodic state in ceramics. This paper provides a comprehensive review of radiation effects (due to γ-, β- and α-decay events, as well as from actinide doping experiments and particle irradiations) on nuclear waste form glasses and crystalline ceramics, particularly Synroc phases, zircon, apatite, monazite and titanite. The paper also includes recommendations for future research needs.

Fast iterative methods for discrete-ordinates particle transport calculations

Abstract: In discrete-ordinates (S N ) simulations of large problems involving linear interactions between radiation and matter, the underlying linear Boltzmann problem is discretized and the resulting system of algebraic equations is solved iteratively. If the physical system contains subregions that are optically thick with small absorption, the simplest iterative process, Source Iteration, is inefficient and costly. During the past 40 years, significant progress has been achieved in the development of acceleration methods that speed up the iterative convergence of these problems. This progress consists of ( i ) a theory to derive the acceleration strategies, ( ii ) a theory to predict the convergence properties of the new strategies, and ( iii ) the implementation of these concepts in production computer codes. In this Review we discuss the theoretical foundations of this work, the important results that have been accomplished, and remaining open questions.

Development of pyroprocessing technology

Abstract: A compact, efficient method for recycling IFR fuel is being developed. This method, known as pyroprocessing, capitalizes on the use of metal fuel in the IFR and provides separation of actinide elements from fission products by means of an electrorefining step. The process of electrorefining is based on well-understood electrochemical concepts, the applications of which are described in this chapter. With only the addition of head-end processing steps, the pyroprocess can be applied with equal success to fuel types other than metal, enabling a symbiotic system wherein the IFR can be used to fission the actinide elements in spent nuclear fuel from other types of reactor.

Assembly homogenization techniques for light water reactor analysis

Abstract: Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods. Applications of these techniques to PWR baffle/reflector homogenization and BWR bundle homogenization are discussed. Nodal solutions to realistic reactor problems are compared to fine-mesh PDQ calculations, and the accuracy of the advanced homogenization methods is established. Remaining problem areas are investigated, and directions for future research are suggested.