Nuclear power currently supplies some 17% of world electricity production from over 400 power stations. In addition, advanced nuclear systems are under development with the potential to make a significant contribution to future energy demands in an environmentally acceptable manner.
Materials R&D in The Nuclear Field
The safe, reliable and economic operation of such plant is critically dependent on good materials performance and, in particular, on understanding and mitigating specific environmental degradation processes (e.g. mechanical, corrosion and radiation effects). Materials R & D effort in the nuclear field has spanned some 40 years but, interestingly, has resulted in much detailed understanding of many generic aspects of materials behaviour, in areas such as crystal defects, diffusion and solute segregation, phase evolution and deformation and fracture processes etc outside the nuclear field. Such advances have been of both direct and indirect benefit to many other industries, including fossil fuel power generation, chemical plant, aerospace etc., and extending into such diverse areas as novel welding techniques, tribology, liquid metal technology, high purity alloy specification and production, structural integrity etc.
Materials and nuclear Power – Euromat ‘96
This is the background against which The Institute of Materials organised Euromat ‘96, under the conference title ‘Materials and Nuclear Power’ on behalf of The Federation of European Materials Societies (FEMS), and (for the first time) in association with the American Society of Metals (ASM). This international conference, one of a continuing series promoted by FEMS under the Euromat title, was held at the Bournemouth International Conference Centre on 21-23 October 1996, attracting over 130 delegates and speakers from 20 countries. The purpose of the meeting was not only to address the current status of materials for nuclear plant but also to highlight the potential for technology transfer to other industries. The content of the conference embraced a whole range of materials topics, from the design and construction of advanced systems through to aspects of nuclear fuels and finally to backend issues of waste management and decommissioning.
Materials For Light Water Reactors and Boiling Water Reactors
Light water reactors (the pressurised water reactor or PWR and the boiling water reactor or BWR) account for over 80% of an installed nuclear capacity, and thus materials aspects of these system were the dominant theme of Sessions 1 and 2. Manfred Erve (Siemens AG, Germany) reviewed materials choices for PWRs. Low alloy ferritic steels are selected for heavy section components principally the reactor pressure vessel (RPV), the steam generator (SG), the pressuriser and the reactor containment. For example, the RPV in modern plant is constructed of monoblock ring forging of A508 Class 3 - a high toughness Mn-Mo-Ni bainitic steel. Austenitic stainless steels of the 18Cr-10Ni type are used for core internals, small diameter primary loop systems and, in some designs, for main coolant pump casings. In some cases Nb-stabilised AISI 347 or other stabilised grades are selected whereas non-stabilised grades are specified with low (< 0.02%) carbon to minimise problems from intergranular stress corrosion cracking (SCC). Finally, high nickel alloys are used in heat exchanger tubing for the SC, and for small components in core internals. SCC has been problem in Inconel 600, but selection Inconel 690 and 800 has proved beneficial in this respect.
Development of High Grade A508 Ring Forgings for Reactor Pressure Vessels
G A Honeyman (Forgemasters Steel and Engineering Ltd, UK) traced the development of high grade A508 ring forgings for RPVs and, in particular, the specifications needed to confer high toughness and radiation resistance. The former is achieved by tailoring C, Ni, Mn levels and keeping sulphur levels low - 0.003% is now routinely obtained. Radiation embrittlement is associated with the presence of residual Cu in the steel, and is a problem at levels above 0.1%; again levels at or below 0.05% are now routinely obtained; in addition, any thermal ageing embrittlement due to P is also minimised by ensuring levels are below 0.005%.
Other contributions in the first two sessions reviewed aspects of environmental degradation in support of lifetime management (L Valibus, EdF, France) and described the degradation mechanisms caused by the neutron irradiation environment. The role of Cu in the radiation embrittlement of RPV steels at the coolant inlet temperature of 290°C is now well understood; although Cu retained in solid solution is thermally immobile at this temperature, radiation induced formation of fine epsilon-Cu particles occurs. These retain a bcc structure and act as potent matrix strengtheners. The routine specification of low-Cu A508 steel, however, has essentially solved the RPV embrittlement problem in new plant. Furthermore, in the latest on-going designs, such as the European pressurised water reactor (EPR) being undertaken by a FrancoGerman (Framatome/Siemens) consortium, the lifetime neutron dose even for a proposed extended 60 year life is significantly reduced compared with standard plant by specifying a larger than usual gap between the RPV and the core barrel.
Irradiation Assisted Stress Corrosion Cracking
Austenitic core internals also experience loss of fracture toughness during irradiation to high doses at the coolant outlet temperature of 325°C, but the toughness saturates at levels compatible with the relatively low stresses experienced by these components; however, there is concern for advanced PWRs with lifetimes extending to 60 years for which further data is required. Nickel-base alloys also exhibit more severe radiation embrittlement and their use may need to be re-evaluated. Irradiation-assisted stress corrosion cracking (IASCC) may also be a problem in PWR core internals and certainly more so in BWRs. IASCC in austenitic steels is a relatively newly understood phenomena in which grain boundary depletion of Cr together with enrichment of minor elements (e.g. Si) occurs by a process of radiation-induced non-equilibrium segregation (RIS) in which solute element fluxes are driven by coupling to the point defect fluxes. The Cr-depleted boundary is susceptible to anodic dissolution in reactor water and, because the boundary is also weakened by segregants, a form of SCC occurs. Several instances of both IASCC and SCC in core components were reported, whilst the theoretical basis of RIS was covered by the presentations of G Martin (CEA, France) and R G Faulkner (Loughborough University, UK).
Advanced Fuel Cycles
A number of important aspects of advanced fuel cycles were covered in Session 3. The use of MOX (i.e. mixed oxide or (Pu,U)O2) fuel for PWRs as a means of recycling plutonium was reviewed by H Bernard (CEA, France) and fast reactor fuels including high-Pu U-free non-conventional variants were also discussed. The development of zircalloy fuel cladding for LWRs was highlighted by A Seibold (Siemens AG, Germany) and the benefits of restricted compositional specifications and/or exceeding ASTM elemental specifications to improve corrosion resistance were described. The elements with the largest effect on corrosion are Sn, Fe and Cr.
Radioactive Waste Management
The development of technology for safe and responsible radioactive waste management is a key issue for on-going confidence in nuclear power, and this aspect was considered in Session 4. Encapsulation of high level waste by vitrification was a central theme in the presentation by C Scales (BNFL, UK) while investigations of the potential of glass‑ceramics (K M Garrett, University of Warwick) and cementitious systems (E J Butcher, BNFL) for immobilisation of certain waste forms was described. From a different viewpoint, Henderson (Swedish Institute of Metals Research) presented creep test data on various grades of copper used in Sweden for radwaste canisters.
Structural Materials for Fast Breeder Reactors and Fusion Reactors
The final session covered structural materials for fast breeder reactor (FBR) and fusion reactor (FR) systems, and opened with a survey by W Dietz (Lindlar, Germany). In sodium-cooled FBRs, operating temperatures are in the range 350-550°C (excluding transients) and austenitic steels have been developed for primary system components (reactor vessel, core support, sodium coolant piping). Around the world there has been convergence to a low-C grade with added N, i.e. Type 316L(N), for resistance to intergranular attack during fabrication and improved ductility and strength during operation. There are trends away from Alloy 800 towards ferritic steels (initially 2.25Cr1Mo and now modified 9Cr1Mo) for FBR steam generators. Materials selection for first wall applications in FRs is still broad-based due to the conceptual nature of fusion systems, which thus permits a wide range of options for operating conditions. Martensitic stainless steels with 9-12%Cr, vanadium alloys and even fibre-reinforced ceramic composites based on SiC are under consideration. Current research is centred on low induced radioactivity (LA) materials for safe waste disposal purposes.
Despite the current downturn in new nuclear plant construction, the buoyant atmosphere evident at this conference indicated considerable optimism for the future. Attention is now focussed on plant life extension, which clearly requires a detailed understanding of material degradation processes, and on decommissioning of existing plant and radwaste issues. The continuing research effort on materials for advanced LWRs for near-term applications and for fusion reactor systems designed to operate early in the next century clearly implies that nuclear power will continue to play a key role in future electricity generation.