[en] IAEA activities on high-temperature gas-cooled reactors are conducted with the review and support of member states, primarily through the Technical Working Group on Gas-cooled Reactors. This paper overviews current and planned activities through a review of Co-ordinated Research Projects, meetings, conferences and workshops. (author)

[en] This presentation will address the challenges involved in materials specification writing that involves acceptable processing methods, establishing minimum acceptable properties and properties ranges, establishing design codes that consider appropriately the inherent variability in properties of graphite, graphite failure criteria, and the inspection requirements for nuclear graphite before use and during service. The presentation will also include the status of activities related to the codes and standards development in the US and report on recent activities and directions in the ASTM and ASME standards and codes. (author)

[en] In 2001 the Nuclear Safety Division (NSD) of the UK Health and Safety Executive (HSE) decided to underwrite the Nuclear Graphite Research Group (NGRG) at the University of Manchester, UK with the aim of providing a source of independent research and advice to the HSE (NSD). Since then the group has rapidly expanded to 16 members and attracted considerable funding from the nuclear power industry and the regulator for a wide range of research and consultancy work. It is now also part of the Material Performance Centre within the BNFL Universities Research Alliance. Extensive collaboration exists between the group and other nuclear research institutes, both in the UK and overseas. This paper briefly describes some of the research programmes being carried out by the NGRG at Manchester. (author)

[en] Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite. As this behaviour compensates, to some extent, some other irradiation effects such as dimensional change or thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK as described by the UK Creep Law as well as other countries. It then offers an alternative analysis of irradiation creep, using the AGR moderator graphite as an example, to high values of neutron fluence and applied stress. The paper will demonstrate that the new analysis provides for a more physically satisfying explanation for the decrease in the apparent creep strain observed at high fluence. The prediction of creep strain for a range of graphites and applications, including HTR systems, will be discussed. (author)

[en] In response to growing interest in high-temperature, gas-cooled reactors (HTGRs) in many countries and the need for improved materials for nuclear applications in high-temperature environments, the NEA organised the Third Information Exchange Meeting on Basic Studies in the Field of High-temperature Engineering. The proceedings of this meeting provide an overview of high-temperature research currently under way, including studies on the behaviour of irradiated graphite and improvements in material properties under high-temperature irradiation. These proceedings also contain recommendations for further international work in the areas of high-temperature engineering. (author)

[en] A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k_eff with burn-up. Nitride fuel (¹⁵N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k_eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

[en] 6H silicon carbides (SiCs) were implanted with phosphorus (P) ions at energies ranging from 9-21 MeV between RT and 1200 deg C. To develop the evaluation techniques for P donors created in SiC by nuclear transmutation doping (NTD), P atoms implanted in SiC were studied using electron spin resonance (ESR). The ESR signal arising from residual defects decreased as the implantation temperature increased. Although the damage creation was strongly suppressed by implantation at 1200 deg C, signals relating to shallow P donors were not observed for samples implanted up to 1200 deg C. The signals arising from the isolated shallow P donors on the quasi-cubic silicon sites were observed for samples after annealing at 1650 deg C for 30 minutes. Neutron irradiation into 6H-SiC at 1200 deg C was carried out to develop an efficient process for NTD for SiC. The conduction type of the samples irradiated with neutrons was converted from p-type to n-type through annealing at 1800 deg C for one minute. (author)

[en] Three-point bend testing of sandwiched specimens has been applied to nuclear graphite. This test geometry gives stable crack propagation, which allows optical imaging of the crack path and the crack tip process zone. Direct measurement of specimen strains by electronic speckle pattern interferometry (ESPI) has been used to monitor the crack tip strain distribution. Image correlation techniques have also been employed to study the crack tip process zone. The sandwich bend test has been simulated using a continuum damage mechanics (CDM) model, which has been developed for nuclear graphite. The model has been implemented into the commercial finite element code ABAQUS via the user element subroutine (UEL). The model can be used to predict both the initiation and propagation of cracks in components under mechanical loading. The aim of this simulation is to verify the CDM model against the existing experimental data. (author)

[en] Two types of tailored ceramics were prepared by colloidal processing. First, colloidal processing was applied to obtain dense and homogeneous alumina- and zirconia-based ceramics with fine-grained microstructure. Large tensile elongation exceeding 550% was achieved for 10 Vol.% zirconia-dispersed Al₂O₃ at 1773 K and high-strain-rate (above 10^-2 s^-1) superplastic zirconia-based ceramics were, achieved by adding TiO₂ and MgO. In the second process, a high magnetic field was applied during slip casting of alumina suspension to produce textured microstructure. A green body with a slight degree of crystallographic orientation was obtained and the degree of texture increased as temperature was increased. After densification, alumina grains grow to yield ceramics with textured grains and high degrees of crystallographic orientation. This processing can be applied to other ceramics having, asymmetric unit cell, such as SiC, Si₃N₄, etc. (author)

[en] HTR projects have been launched within the 5. Framework Programme of the European Commission to consolidate and advance HTR technology in Europe. This paper considers the mid-term results from the project HTR-M, which looks at the materials requirements for the HTR reactor pressure vessel, the high-temperature resistant alloys for the internal structures and turbine, and graphite for the reactor core. The programme of work includes irradiation testing of vessel steel and graphite materials. and the main focus of the paper is on the progress of this test work. (author)