[en] The properties of materials for thermal and fast reactor core components applications are assessed for suitability. In-service, experimental and theoretical aspects of the behaviour of structural materials under irradiation are considered. The factors which influence the choice of core component materials for fuel pins etc. are considered. Austenitic and ferritic steels, nickel-based alloys zircaloy, magnox and some non-metallic materials have been tested for suitability. The first 7 papers are on helium embrittlement and creep deformation. The next 22 papers are on creep, microstructure and fracture mechanics of non-ferrous materials. Austenitic and high nickel alloys are studied in the next 12 papers, then ferritic steels (6 papers). Segregation and microstructural stability (7 papers) and environmental aspects (8 papers) are considered. All 53 papers are indexed separately. (U.K.)

[en] A programme of isothermal corrosion testing is being performed to investigate the effect on corrosion of Zircaloy-4 fuel cladding if hideout of lithia and boric acid occurs at high PWR fuel burn-ups. Corrosion rates in simulated start-of-cycle primary coolant were similar to those in pure water. Some acceleration of corrosion occurred in a high concentration environment simulating that which might be generated by hideout, but significantly less than expected due to lithia in the absence of boric acid. (author)

[en] Irradiation-induced solute redistribution has been investigated in a series of Fe-Cr-Ni ternary alloys neutron irradiated to a dose of 12.7 dpa at 400⁰C. Nickel enrichment and chromium and iron depletion have been observed at the internal point defect sinks such as grain boundaries, voids and dislocation loops and a detailed study of the effect of alloy composition on the degree of solute segregation to grain boundaries has been undertaken. (author)

[en] Tensile and creep test specimens of the austenitic steel 1.4981 (X 8 CrNiMoNb 16 16) in two microstructural conditions were irradiated at 673 K in five reactor experiments to accumulate thermal neutron fluences between 1.2 x 10²² and 3 x 10²⁵ m^-2. The postirradiation examinations demonstrated that by appropriate thermomechanical treatments microstructures can be obtained which retain good ductility at temperatures where high temperature embrittlement is predominant even after high neutron fluences. The variant optimized for high initial ductility-45% cold work followed by heat treatment at 1173 K for 2 hours which produced a recrystallized fine grained microstructure with homogeneously distributed carbide precipitates - retained total elongations of more than 10% in tensile and creep tests in the range 873 and 1123 K. In short-term creep tests a reduced secondary creep stage was observed. Both 0.2% proof stress and ultimate tensile stresses were increased. For the optimized variant those values were 20 to 40% higher. A normalization of the ductility data showed that the relative embrittlement of the two variants of the alloy was about the same for neutron fluences above 1 x 10²⁵ m^-2; therefore the higher initial ductility of the optimized variant was responsible for the desired higher total elongation after irradiation. (author)

[en] The aim of this work is to give a general comparison of the swelling behaviours of two Ti-stabilized austenitic steels (type 316 and 15 Cr-15 Ni). The irradiation of samples at 500⁰C shows that 15-15Ti SS swells less than 316 Ti SS only when irradiated in the cold-worked state. The irradiation of these two cold-worked alloys as fuel pin cladding shows that the higher swelling resistance of 15-15 Ti steels is only clearly established in the temperature range 450-550⁰C. Moreover this steel may exhibit large swelling variability from cast to cast. Transmission electron microscopy examinations are used to understand the origin of the swelling variability of cold worked 15-15 Ti steel and its difference of behaviour with cold worked 316 Ti steel in the upper range of irradiation temperature. (author)

[en] A high voltage electron microscope has been used to simulate a nuclear reactor in a study of radiation induced solute segregation in 20Cr/25Ni/Nb stabilised stainless steels. Subsequent monitoring of the solute redistribution around grain boundaries using STEM based EDX microanalysis has shown depletion of chromium and enrichment of nickel. The results are in qualitative agreement with profiles from neutron irradiated material and provide the basis for further investigation of the effect of prior cold work and minor solute additions on the radiation induced segregation. (Author)

[en] Gilsocarbon graphites were irradiated to high weight losses in three different CO₂ based coolants. The experimental data is tested against a model which interprets the gas phase chemistry and pore geometry and allows weight loss and gas flow properties to be calculated. The observed changes of oxidation rate with dose were successfully predicted from the model. An empirical relationship was also derived which was shown to fit data for moderator, sleeve and special pore structure graphites. Changes in graphite permeability and diffusivity were predicted by the model, and also by other simplified, more approximate methods. The model based upon the measured transport pore spectrum was shown to be the best with other methods proving adequate to moderate doses. (author)

[en] Fully instrumented creep and stress rupture tests have been performed at 873K for times up to 20,000h on a series of type 316 steel/17Cr 8Ni 2Mo weld metal specimens in the unirradiated and thermal neutron irradiated conditions. The specimens tested included all weld metal longitudinal and transverse composites in the as-welded condition and following a stress relief heat treatment of 10h at 1075K. Simulated heat affected zone (HAZ) specimens were also tested. Analysis of the creep results combined with metallography, autoradiography and TEM established that the decrease in properties of irradiated samples is caused by an increasing secondary strain rate due to enhanced helium induced grain boundary fracture of the simulated HAZ and enhanced interdendritic fracture in the weld metal. Implications of strength reductions on the design of welded structures subjected to thermal irradiation are briefly assessed. (author)

[en] 12Cr-15Ni based steels with additions of Si alone or (Si+Ti) have been irradiated to a maximum dose of 47 dpa at temperatures ranging from 399 to 649⁰C. Irradiation-induced transformation of austenite to ferrite and occasionally martensite was observed. The extent of transformation was increased by Si and Ti additions but reduced by cold working. Void swelling was generally reduced by Si additions, but Ti-modification resulted in enhanced swelling in solution treated alloys at high temperatures. (author)

[en] An evaluation of the effects of approximately 11 years of service on Rod Cluster Control Assemblies was performed. Some fretting and sliding wear of the rodlet cladding was observed. Intergranular cracks in the cladding were attributed to a combination of mechanical interference between the absorber and cladding and to irradiation effects in the stainless steel cladding. The information gained from this evaluation has been translated into design changes that enhance the Rod Cluster Control Assembly performance. These design changes include a wear resistant coating applied to a low impurity, crack resistant cladding and a slight increase in the absorber-to-cladding gap. (author)