[en] The spinning cylinder experiment organised by the Network for the Evaluation of Steel Components (NESC) is designed to investigate the cleavage initiation behaviour of both surface breaking and subclad defects in simulated end of life RPV material, exposed to a pressurised thermal shock transient. Pre-test structural integrity assessments are performed by the NESC Structural Analysis Task Group (TG3). The results of these structural integrity assessments are used to determine the design of the experiment and especially the sizes of the introduced defects. In this report the results of the pre-test analyses performed by the group Applied Mechanics at ECN - Nuclear Energy are described. Elastic as well as elasto-plastic structural analyses are performed for a surface breaking and a subclad defect in a forged cylinder with a 4 mm cladding. The semi elliptical defects have a depth of 40 mm and an aspect ratio of 1:3. (orig.)

[en] In the framework of developing the helium cooled pebble bed (HCPB) blanket four pebble bed assemblies are to be irradiated in the HFR in Petten. The objective of these experiments is to study the thermo-mechanical behaviour of the lithium ceramic breeder and beryllium pebble beds during irradiation. The basic test elements for the pebble bed assemblies consist of EUROFER-97 cylinders with a horizontal lithium ceramic breeder pebble bed sandwiched between two beryllium pebble beds. The breeder and beryllium pebble beds are separated by EUROFER-97 steel plates, which can float along the cylinder walls. The plates are not connected to the cylinder walls, but are supported only by the adjacent pebble beds. The breeder pebble beds are either Lithium-Ortho-Silicate (Li₄SiO₄) or Lithium-Meta-Titanate (Li₂TiO₃). Under influence of stress, temperature, and time the breeder and beryllium pebble beds show non-linear elastic (f(σ,T)), compaction (f(σ,T)), and creep compaction (f(σ,T,t)) behaviour. Additionally, the thermal conductivity of the beryllium pebble bed depends on compaction level. All mentioned effects have to be taken into account for the evaluation of the temperatures, stresses, and strains that develop in the test elements during in-pile operation. This paper describes how the non-linear elasticity, compaction, and creep compaction have been incorporated into the general purpose finite element program MARC. The method applied to determine the model parameters from experimental oedometer data is also described. The pebble bed model has been applied to calculate the thermo-mechanical behaviour of complete pebble bed assemblies to be irradiated in the HFR. The results of these calculations are critical for a safety assessment of the in-pile operation of the experiment and will provide a better understanding of the in-pile behaviour. (orig.)

[en] To study the possibilities of modelling crust formation at the interface of an oxidic melt and a steel bottom plate with a solid element model a modified heat transfer coefficient was defined to simulate the heat transfer by convection in the liquid corium. Above the melting temperature of Al₂O₃ (2050 C) the true conductivity of the oxidic material is increased by a temperature dependent conductivity multiplier. In this study for this multiplier a power law of order p was chosen. With this modified conductivity model a number of FE-analyses was made for p = 1, 3, and 5. For reasons of comparison an additional analysis was made without correction for heat transport due to convection (multiplier is equal to 1) and also an analysis was made with a jump in the conductivity as was used in former analyses. From theses analyses it can be concluded that: Immediately after the molten corium comes in contact with the steel plate an oxidic crust will be built up and the melting of the bottom plate will initiate later on. So it is possible using a solid model to simulate crust formation taking into account the heat transfer by convection. The result of an increased conductivity in the liquid corium is that the crust will be thinner and that the melting process in the bottom plate will initiate later. The influence of the parameter p in the power law on the calculated temperature field in the corium and in the bottom plate is only very small for values of p in the range 1-5. (orig./DG)

[en] NRG has more than 35 years experience in the nuclear field. Amongst others, this includes safety and integrity assessments and operation of the 50 MWth High Flux Reactor (HFR) owned by the European Commission. At the request of the Dutch Nuclear Regulatory Body a safety re-evaluation was performed for the HFR facilities. Special attention was paid to seismic evaluation in view of recent gas and oil exploration in the vicinity of the HFR site and the potential for induced earthquakes. Initially, a Basic Design Earthquake spectrum was defined for the HFR site on the basis of available seismic data and regulatory requirements. For the various HFR components acceptance criteria and requirements were agreed upon with the Dutch Nuclear Regulatory Body. The earthquake resistance of the HFR components was evaluated using finite element analyses, with strengthening modifications proposed for those components that did not meet the acceptance criteria. (authors)

[en] The CORVIS (Corium Reactor Vessel Interaction Studies) programme investigated the possible failure modes of a Reactor Pressure Vessel (RPV) lower head induced by direct contact with molten core material following a severe accident. The investigations included experiments and analyses. This paper presents the analysis of lower head thermal behaviour. Topics covered are the heat transfer by convection in the melt, crust formation at the interface between melt and colder structures, and the heat resistance at the interface due to imperfect thermal contact and the development of a gap. (orig.)

[en] Four pebble-bed assemblies are to be irradiated in the HFR in Petten with the objective to study the thermo-mechanical behaviour of the breeder ceramic pebble beds during irradiation. The thermo-mechanical behaviour of the pebble bed assemblies was calculated in a 2D axi-symmetric model in MARC. In this approach there could not be accounted for the influence of thermocouple tubes on the temperature distribution in the assembly, because these are distributed in the assembly in a non axi-symmetric manner. The solution for this problem was to expand the model to a 3D model used for thermal computations only. For safety reasons the tritium production in the breeder and permeation through the first and second containment must be estimated before the in-pile experimentation begins. In order to do so, the calculated thermal distribution is used as input for the enhanced two-dimensional finite element model in MARC. Adaptations are made in the 2D model by adding the capability of performing mass flux calculations. This paper describes the finite element models used for computation of the temperature distribution and the tritium flux through the pebble bed assembly. The results of these calculations are critical for a safety assessment of the in-pile operation of the experiment and will give a better understanding of the in-pile behaviour on temperature and tritium management in advance. (orig.)

[en] In the framework of developing the helium cooled pebble-bed (HCPB) blanket an irradiation test of pebble-bed assemblies is prepared at the HFR Petten. The test objective is to concentrate on the effect of neutron irradiation on the thermal-mechanical behaviour of the HCPB breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. The paper reports on the project status, and presents the results of pre-tests, material characteristics, the manufacturing of the pebble-bed assemblies, and the nuclear and thermo-mechanical loading parameters

[en] In the framework of developing the European Helium Cooled Pebble-Bed (HCPB) blanket an irradiation test of pebble-bed assemblies is performed in the HFR Petten. The experiment is focused on the thermo-mechanical behavior of the HCPB type breeder pebble-bed at DEMO representative levels of temperature and defined thermal-mechanical loads. To achieve representative conditions a section of the HCPB is simulated by EUROFER-97 cylinders with a horizontal bed of ceramic breeder pebbles sandwiched between two beryllium beds. Floating Eurofer-97 steel plates separate the pebble-beds. The structural integrity of the ceramic breeder materials is an issue for the design of the Helium Cooled Pebble Bed concept. Therefore the objective of the post irradiation examination is to study deformation of pebbles and the pebble beds and to investigate the microstructure of the ceramic pebbles from the Pebble Bed Assemblies. This paper concentrates on the Post Irradiation Examination (PIE) of the four ceramic pebble beds that have been irradiated in the Pebble Bed Assembly experiment for the HCPB blanket concept. Two assemblies with Li₄SiO₄ pebble-beds are operated at different maximum temperatures of approximately 600 ^oC and 800 ^oC. Post irradiation computational analysis has shown that both have different creep deformation. Two other assemblies have been loaded with a ceramic breeder bed of two types of Li₂TiO₃ beds having different sintering temperatures and consequently different creep behavior. The irradiation maximum temperature of the Li₂TiO₃ was 800^_oC. To support the first PIE result, the post irradiation thermal analysis will be discussed because thermal gradients have influence on the pebble-bed thermo-mechanical behavior and as a result it may have impact on the structural integrity of the ceramic breeder materials. (author)

[en] Full size module systems of the candidate DEMO blanket concepts selected for the European Blanket Project (EBP) will be tested in ITER, presently called Test Blanket modules (TBM). The Helium Cooled Pebble Bed (HCPB) is one of the two concepts developed in the European Union (EU). This development path consists of several scaling steps, including integral testing of a subsized module at realistic operation conditions. As part of the HCPB development work for the TBMs, ECN performed jointly with JRC/IAM at Petten a feasibility study for irradiation tests of subsized HCPB modules. The first stage of the study was concluded with a report on the conceptual design of an in-pile test of a single submodule with a helium cooling loop. Such test was considered technically feasible, but would require significant project duration and expenditures. Further development of detailed objectives for in-pile tests was recommended, in particular in view of the different parameters for the HCPB-ITM and DEMO-concept. This objective has been modified by the EBP in fall 1997. For the final stage of the study the test objective has been: the in-pile testing of the thermal/mechanical behaviour of the HCPB ceramic breeder beds, while giving lower priority to tritium transport issues (release, permeation). Several configuration options in the High Flux Reactor (HFR) in Petten, Netherlands, to perform in-pile test of HCPB submodules have been considered. Neutronics analyses along with thermal and structural analyses have been made for selected options and several HFR peripheral in-tank positions. These pre-design analyses show that the whole range of breeder bed power densities and temperature levels, which are relevant to the HCPB DEMO and BTM designs, can be reached with the options presented. The options are all cooled by the reactor coolant flow. The use of an helium loop is not compulsory and is considered as an unnecessary complication with regard to the present test objectives. Relevant lithium burnups can be reached in a reasonable time scale, due to a high availability of the HFR. The proposed type of irradiation experiment can be readily detailed and realized. 21 refs

[en] An in-pile test of helium cooled pebble-bed (HCPB) typical pebble-bed assemblies will be performed in the high flux reactor (HFR) Petten. Its major objective is to study the thermo-mechanical behaviour of the ceramic breeder pebble-beds and neutron irradiation effects. In order to arrive at the final design of test elements, results have been obtained from out-of-pile pre-tests and modelling activities. Preliminary thermo-mechanical analyses of the behaviour of the in-pile test-elements have been performed