Damage assessment in structural metallic materials for advanced nuclear plants

Future advanced nuclear plants are considered to operate as cogeneration plants for electricity and heat. Metals and alloys will be the main portion of structural materials employed (including fuel claddings). Due to the operating conditions these materials are exposed to damaging conditions like creep, fatigue, irradiation and its combinations. The paper uses the most important alloys: ferritic-martensitic steels, superalloys, oxide dispersion strengthened steels and to some extent titanium aluminides to discuss its responses to these exposure conditions. Extrapolation of stress rupture data, creep strain, swelling, irradiation creep and creep-fatigue interactions are considered. Although the stress rupture- and the creep behavior seem to meet expectations, the long design lives of 60years are really challenging for extrapolations and particularly questions like negligible creep or occurrence of diffusion creep need special attention. Ferritic matrices (including oxide dispersion strengthened (ODS), steels) have better irradiation swelling behavior than austenites. Presence and size of dispersoids having a strong influence on high-temperature strength bring only insignificant improvements in irradiation creep. A strain-range-separation based approach for creep-fatigue interactions is presented which allows a real prediction of creep-fatigue lives. An assessment of capabilities and limitations of advanced materials modeling tools with respect to damage development is give

Abnormal subgrain growth in a dislocation-based model of recovery

Simulation of subgrain growth during recovery is carried out using two-dimensional discrete dislocation dynamics on a hexagonal crystal lattice having three symmetric slip planes. To account for elevated temperature (i) dislocation climb was allowed and (ii) a Langevin type thermal noise was added to the force acting on the dislocations. During the simulation, a random ensemble of dislocations develop into subgrains and power-law type growth kinetics are observed. The growth exponent is found to be independent of the climb mobility, but dependent on the temperature introduced by the thermal noise. The in-depth statistical analysis of the subgrain structure shows that the coarsening is abnormal, i.e. larger cells grow faster than the small ones, while the average misorientation between the adjacent subgrains remains nearly constant. During the coarsening Holt's relation is found not to be fulfilled, such that the average subgrain size is not proportional to the average dislocation spacing. These findings are consistent with recent high precision experiments on recovery.Comment: 17 pages, 11 figure

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10×2×0.2mm3 were investigated in a temperature range of 300°C to 500°C under irradiation, and significant creep strains were detected. At temperatures above 500°C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900°C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900°C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloy

Synchrotron X-Rays for Microstructural Investigations of Advanced Reactor Materials

X-rays from synchrotron beamlines provide a powerful tool for materials analysis in circumstances where long-term materials degradation under complex loading conditions (e.g., temperature, irradiation, and stress) becomes important. This may occur for advanced gas cooled reactors. Synchrotron X-rays can help to improve lifetime assessments by providing a more in-depth understanding of microstructural damage. This article summarizes results of X-ray absorption fine spectrum analysis and X-ray magnetic circular dichroism synchrotron techniques. They were employed to evaluate various microstructural features, which are important in understanding the lifetime of materials exposed to extreme conditions. Dispersoid strengthening by yttria particles, conditions that produce nanocrystal Zircaloy, and the role of magnetism on the stability of ferritic steels were taken as example

Scientific Assessment in support of the Materials Roadmap enabling Low Carbon Energy Technologies: Technology Nuclear Energy

This scientific assessment serves as the basis for a materials research roadmap for the nuclear fission technology, itself an integral element of an overall "Materials Roadmap Enabling Low Carbon Technologies", a Commission Staff Working Document published in December 2011. The Materials Roadmap aims at contributing to strategic decisions on materials research funding at European and Member State levels and is aligned with the priorities of the Strategic Energy Technology Plan (SET-Plan). It is intended to serve as a guide for developing specific research and development activities in the field of materials for energy applications over the next 10 years. This report provides an in-depth analysis of the state-of-the-art and future challenges for energy technology-related materials and the needs for research activities to support the development of nuclear fission technology both for the 2020 and the 2050 market horizons. It has been produced by independent and renowned European materials scientists and energy technology experts, drawn from academia, research institutes and industry, under the coordination the SET-Plan Information System (SETIS), which is managed by the Joint Research Centre (JRC) of the European Commission. The contents were presented and discussed at a dedicated hearing in which a wide pool of stakeholders participated, including representatives of the relevant technology platforms, industry associations and the Joint Programmes of the European Energy Research Associations.JRC.F.4-Safety of future nuclear reactor

Materials for the Very High Temperature Reactor (VHTR): A Versatile Nuclear Power Station for Combined Cycle Electricity and Heat Production

The International Generation IV Initiative provides a research platform for the development of advanced nuclear plants which are able to produce electricity and heat in a combined cycle. Very high-temperature gas-cooled reactors are considered as near-term deployable plants meeting these requirements. They build on high-temperature gas-cooled reactors which are already in operation. The main parts of such an advanced plant are: reactor pressure vessel, core and close-to-core components, gas turbine, intermediate heat exchanger, and hydrogen production unit. The paper discusses the VHTR concept, materials, fuel and hydrogen production based on discussions on research and development projects addressed within the generation IV community. It is shown that material limitations might restrict the outlet temperature of near-term deployable VHTRs to about 950 °C. The impact of the high temperatures on fuel development is also discussed. Current status of combined cycle hydrogen production is elaborated on

Materials for Nuclear Plants: From Safe Design to Residual Life Assessments

The clamor for non-carbon dioxide emitting energy production has directly  impacted on the development of nuclear energy. As new nuclear plants are built, plans and designs are continually being developed to manage the range of challenging requirement and problems that nuclear plants face especially when managing the greatly increased operating temperatures, irradiation doses and extended design life spans. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments  provides a comprehensive treatment of the structural materials for nuclear power plants with emphasis on advanced design concepts.   Materials for Nuclear Plants: From Safe Design to Residual Life Assessments approaches structural materials with a systemic approach. Important components and materials currently in use as well as those which can be considered in future designs are detailed, whilst the damage mechanisms responsible for plant ageing are discussed and explained. Methodologies for materials characterization, materials modeling and advanced materials testing will be described including design code considerations and non-destructive evaluation concepts.   Including models for simple system dynamic problems and knowledge of current nuclear power plants in operation, Materials for Nuclear Plants: From Safe Design to Residual Life Assessments is ideal for students studying postgraduate courses in Nuclear Engineering. Designers on courses for code development, such as ASME or ISO and nuclear authorities will also find this a useful reference

APPLICABILITY OF GROUPWARE FOR COMMUNICATION IN DIFFERENT PROJECT ENVIRONMENTS - A CASE STUDY Wolfgang Hoffelner, Roswitha Hoffelner ([email protected])

this article as a measure for number of different file types, necessity of actualisation of documents, necessity of comments, necessity of asynchronous and synchronous communication etc. Intensity of communication is understood in this article as a measure of how often an interaction with BSCW is necessary for the user to participate in the project. Although statistical evaluations what kind of groupware functionality is usually employed (Appelt 2001) exist, there is currently not enough statistical information available to put these two terms on an absolute scale. We therefore based our rating on a semi-quantitative relative scale which we however do not see as a restriction to our interpretation. More investigations would be necessary for an objective classification of the two terms "complexity of information" and "intensity of communication". Figure 2 summarizes the results in a two dimensional plot. It can be seen that the three successful BSCW projects 3,5,6 are located relatively close together in the upper right quadrant of the diagram, whereas the un-successful projects 1 and 2 show a different nature. Project 4 is something in between which is also compatible with the experience. Figure 2: Mapping of the investigated projects in terms of complexity of information and intensity of communicatio