Beryllium Pebbles Research Articles

Neutronic design optimisation of modular HCPB blankets for fusion power reactors

Neutronic design optimisation analyses based on three-dimensional Monte Carlo calculations have been conducted with the MCNP code as part of the development work for the modular helium cooled pebble bed (HCPB) blanket for a fusion power reactor of the model B-type of the European power plant conceptual study (PPCS). Design variants with breeder pebble beds parallel and perpendicular to the first wall have been investigated with the objective to assess and optimise their tritium breeding efficiency. Although both configurations are feasible, the latter design solution was shown to provide a superior performance in terms of the required blanket thickness, the beryllium inventory and the 6Li enrichment due to a better utilisation of the space available for the breeder and the beryllium pebbles.

Inner structures of compressed pebble beds determined by X-ray tomography

In present ceramic breeder blankets, both the ceramic breeder and beryllium are used in form of pebbles. The thermal–mechanical behaviour of these beds strongly depends on the arrangement of the pebbles in the bed, their contacts with other pebbles and with walls and the related contact surfaces. The knowledge of the contact surfaces is of special importance for beryllium pebble beds because the thermal conductivity is sensitively dependent on these surfaces. Results from compression and microtomography experiments are reported where, first, pebble beds consisting of 3.5 mm aluminium spheres, (simulating the beryllium pebbles) were compressed at different pressures in uniaxial compression tests (UCTs). In a second stage, detailed three-dimensional (3D) microtomography experiments were performed. By post-processing the data, very valuable information was obtained on void fraction distributions, number of contacts between the particles, the corresponding contact surfaces and the angular dependence of these contacts. These data are very important for the interpretation of experiments performed to characterise the thermal–mechanical performances of pebble beds.

EU Blanket Design Activities and Neutronics Support Efforts

An overview is provided of the design activities and the related neutronics support efforts conducted in the European Union for the development of breeder blankets for future fusion power reactors. The EU fusion programme considers two blanket lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles (Li4SiO4 or Li2TiO3) as breeder and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy as breeder and neutron multiplier. The blanket design and the related R&D efforts are based on the use of the same coolant and the same modular blanket structure to minimise the development costs as much as possible. The neutronic support efforts include design analyses for the layout and optimization of the modular HCPB/HCLL blankets based on detailed three-dimensional Monte Carlo calculations as well as underlying neutronics activities conducted in the frame of the European Fusion and Activation File (EFF/EAF) projects to develop qualified nuclear data and computational tools for reliable neutronics design calculations.

Topology of compressed pebble beds determined by X-ray tomography

In currently developed ceramic breeder blankets for future nuclear fusion reactors, both the ceramic breeder and beryllium are in form of pebbles. The thermal-mechanical behaviour of pebble beds strongly depends on the arrangement of the pebbles in the bed, their contacts with other pebbles and with walls and the related contact surfaces. The quantitative assessment of contact areas is of special importance for beryllium pebble beds because the thermal conductivity is ruled by this quantity. Results from experiments are reported where, first, pebble beds consisting of 3.5mm aluminium spheres, (simulating the 1mm beryllium pebbles) were uniaxially compressed in the Forschungszentrum Karlsruhe at different pressure levels. Then, detailed three-dimensional microtomography experiments were performed at the European Synchrotron Radiation Facility in Grenoble. Three-dimensional views of the pebble arrangements and void fraction distributions were calculated. By further post-processing the data, the number of contacts between the particles, the corresponding contact areas and the angular dependence of these contacts were determined. It has been shown that in uniaxial compression tests the poloidal distribution of contact surfaces is non-homogeneous. This fact should be considered when using these tests as standard tests for the generation of thermal-mechanical pebble bed data.

A Semi-Theoretical Approach to a Correlation for the Thermal Conductivity of a Beryllium Pebble Bed

In the framework of the European Fusion Technology Programme, Lithium ceramics and Beryllium packed pebble beds are foreseen to be used as Tritium breeders and neutron multipliers, respectively, for the Helium Cooled Pebble Bed breeding blanket of a fusion power reactor operating with a D-T plasma. The present work is focused on the semi-theoretical investigation of the thermal conductivity of single size Beryllium pebble beds, starting from the main hypothesis that this conductivity depends linearly on pebble bed local temperature and mechanical volumetric strain and introducing a method to determine the coefficients of such dependence on the basis of the results obtained by the SUPER-PEHTRA experiments. It has been mainly assumed that the SUPER-PEHTRA Beryllium pebble bed can be considered as a homogeneous, isotropic, and linear elastic medium, and the analytical solution of the direct static problem of the thermo-elasticity for such a system has been used to fit the experimental thermal distributions, uncovering the best values for the thermal conductivity function coefficients. This thermal conductivity has been used together with a constitutive model, realistically taking into account the pebble bed mechanical behavior to reproduce the experimental tests. The results of the analyses agree quite well with the experimental ones, thus encouraging the use of the derived thermal conductivity correlation for Beryllium pebble beds undergoing low plastic volumetric strain.

Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release.

Elementary development for beryllide pebble fabrication by rotating electrode method

Beryllides (Be 12Ti, Be 12V, etc.) are expected to be promising candidates for advanced neutron multipliers for the DEMO blanket. However, it has been difficult to fabricate beryllide pebbles by rotating electrode method (REM) that is the standard fabrication process for the beryllium (Be) pebbles, because of the brittleness of beryllide electrode rod. In this study, the chemical composition of the materials and the fabrication process of the electrode were surveyed. Then the small-scale electrode was fabricated and preliminary trials of pebbles production were conducted by REM. The electrode of the improved material withstood the thermal stress under arc heating and a certain amount of pebbles were obtained. These pebbles by REM were dense and had fine structure. The feasibility of the production of beryllide pebbles by REM was confirmed.

HIP experiments on the first wall and cooling plate specimens for the EU HCPB blanket

First wall and cooling plates are considered the most important structural parts of the EU HCPB blanket concept which is based on the use of ferritic–martensitic steel as structural material, Li 4SiO 4 pebbles as breeder material, beryllium pebbles as neutron multiplier, and 8 MPa helium as coolant. Both the first wall and cooling plates contain complex arrays of internal He coolant channels. The favourite manufacturing technology is diffusion welding of two halves of plates applying the hot isostatic pressure (HIP) welding method that allows uniform distribution of the pressure acting on the outer surfaces of the welding objects. The HIP experiment was started with small MANET specimens with internal coolant channels. The objective of this work is to investigate the appropriate HIP technique, boundary conditions, and parameters in order to achieve good mechanical properties of the welding joints as well as to achieve a transition to test specimens of larger dimensions.

Helium and tritium kinetics in irradiated beryllium pebbles

High-precision measurements of the release rate of helium and tritium from weakly irradiated beryllium pebbles were carried out by means of a Knudsen-cell technique. The samples were heated in a high vacuum to temperatures above the melting point and the evolution of gas release was measured by a mass spectrometer. Different gas diffusion and release stages were recorded and the ruling mechanisms were analysed up to complete exhaustion of the gas inventory. A minor part of the helium content was released by atomic diffusion to free surfaces, starting at about 700 K with a temperature ramp of 10 K/min and about 850 K with 30 K/min. Most of the helium precipitated into bubbles and was released only during abrupt pore venting events, which started at approximately 1500 K. Tritium release was mostly diffusion-dominated. A small fraction of tritium was trapped by helium bubbles during their nucleation phase and was consequently released corresponding to the appearance of the helium peak.

Thermal conductivity of compressed ceramic breeder pebble beds

The effective thermal conductivity of pebble beds is an important parameter for the thermomechanical design of solid breeder blankets operating at temperatures of up to 900 °C in breeder pebble beds and up to 700 °C in beryllium pebble beds. Compressive stresses and creep might cause significant pebble deformations. The knowledge of the thermal conductivity as a function of bed deformation, therefore, is of prime importance. For strongly deformed pebble beds, only results for beryllium pebbles existed where the conductivity increased by a factor of about 5 for bed deformations of about 1%. For ceramic breeder beds, the increase of the bed conductivity with increasing bed deformation is expected to be much smaller. Quantitative results were missing. This paper presents results on the thermal conductivity of lithium orthosilicate and different types of lithium metatitanate pebble beds (monsized and binary beds) for bed deformations up to 4.5% and temperatures up to 800 °C using the pulsed hot wire technique. Most of the measurements at high temperatures were performed in air; at ambient temperature, helium and argon were also used. A distinct increase of the thermal conductivity with bed deformation was found: however, this effect is quite small compared to deformed beryllium beds and might be neglected at high temperatures. The results for zero bed deformation agree well with correlations from literature.

Study of the microstructure of neutron irradiated beryllium for the validation of the ANFIBE code

The behaviour of beryllium under fast neutron irradiation is a key issue of the helium cooled pebble bed tritium breeding blanket, due to the production of large quantities of helium and of a non-negligible amount of tritium. To optimise the design, a reliable prediction of swelling due to helium bubbles and of tritium inventory during normal and off-normal operation of a fusion power reactor is needed. The ANFIBE code (ANalysis of Fusion Irradiated BEryllium) is being developed to meet this need. The code has to be applied in a range of irradiation conditions where no experimental data are available, therefore a detailed gas kinetics model, and a specific and particularly careful validation strategy are needed. The validation procedure of the first version of the code was based on macroscopic data of swelling and tritium release. This approach is, however, incomplete, since a verification of the microscopic behaviour of the gas in the metal is necessary to obtain a reliable description of swelling. This paper discusses a general strategy for a thorough validation of the gas kinetics models in ANFIBE. The microstructure characterisation of weakly irradiated beryllium pebbles, with different visual examination techniques, is then presented as an example of the application of this strategy. In particular, the advantage of developing 3D techniques, such as X-ray microtomography, is demonstrated.

Development of Fusion Nuclear Technologies at Japan Atomic Energy Research Institute

An overview of the present status of development of fusion nuclear technologies at Japan Atomic Energy Research Institute is presented. A tritium handling system for the ITER was designed, and the technology for each component of this system was demonstrated successfully. An ultraviolet laser with a wavelength of 193 nm was found quite effective for removing tritium from in-vessel components of D-T fusion reactors. Blanket technologies have been developed for the test blanket module of the ITER and for advanced blankets for DEMO reactors. This blanket is composed of ceramic Li2TiO3 breeder pebbles and neutron multiplier beryllium pebbles, whose diameter ranges from 0.2 to 2 mm, contained in a box structure made of a reduced-activation ferritic steel, F82H. Mechanical properties of F82H under a thermal neutron irradiation at up to 50 displacements per atom (dpa) were obtained in a temperature range from 200 to 500°C. Design of the International Fusion Materials Irradiation Facility (IFMIF) has been developed to obtain engineering data for candidate materials for DEMO reactors under a simulated fusion neutron irradiation up to 100 to 200 dpa, and basic development of the key technologies to construct the IFMIF is now under way as an International Energy Agency international collaboration.

Tritium and helium release from neutron irradiated beryllium pebbles from the EXOTIC-8 irradiation

For the Helium Cooled Pebble Bed Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier presently consists of a mixed bed of about 2 and 0.1–0.2 mm diameter beryllium pebbles. Beryllium pebbles with a diameter of 0.1–0.2 mm have been irradiated for 200 days in the High Flux Reactor (HFR) in Petten at temperatures ranging from 500 to 600 °C with a fast (>1 MeV) neutron fluence of about 5×10 24 m −2. Tritium inventories as well as release kinetics of both tritium and helium from the irradiated pebbles were investigated by out-of-pile annealing experiments, heating the samples with 7 °C/min up to 1000 °C and purging them with either He+0.1 vol.% H 2 or Ar+1 vol.% H 2. According to previous studies, the tritium release from the pebbles was found to be slow and the maximum release rate occurred at a temperature of about 900 °C. Although tritium is released, no helium release is observed at temperatures lower than 800 °C. However, at 900 °C the samples showed a concurrent release of tritium and helium leading to the conclusion that T and He partially reside in common bubbles in the irradiated material.

Chemical reactivity of beryllium pebbles in air

In the design of the European Helium-Cooled Pebble Bed (HCPB) Blanket, the use of beryllium as a neutron multiplier in the form of a pebble bed is foreseen. In the development of the pebble bed, kinetic data are needed for the reaction between the beryllium pebbles and air. Therefore, thermogravimetric experiments were conducted to study the pebbles' behaviour when exposed to air in the temperature region 400–900 °C. The material tested consisted of beryllium pebbles produced by the Rotating Electrode Method. At temperatures up to 700 °C, the reaction kinetics is approximately parabolic, although a deviation from parabolic behaviour is observed after approximately 6000 s of air exposure. Above 700 °C, a combination of paralinear and accelerating behaviour is observed. From the experimental results the appropriate rate constants were determined by curve fitting.

Safety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.

Optimization of the Filling for the Improvement of the Performance of Reference ITER/DEMO Ceramic and Beryllium Pebble Beds

Both the EU Long Term Programmes for DEMO and the ITER R&D foresee the thermal-mechanical qualification of the beryllium, as neutron multiplier, and lithium orthosilicate or lithium metatitanate as breeder ceramics pebble beds. FZK has performed measurements on the pebble bed thermal-mechanical properties using cylindrical test sections. Using an alternative approach, ENEA, has launched similar testing on the SMARTS mock-up, reproducing on a small scale the reactor reference plane geometry1 instead. The tests have shown that the pebble bed thermal behaviour is strongly affected by the initial filling Packing Factor (PF). In fact, the higher the PF, the higher the thermal conductivity of the bed. Therefore, if the neutron multiplication needs an increase in the pebble PF, the only possibility is to adopt binary pebble beds (small pebbles infiltrating between larger ones) as an alternative to the mono-sized lattice. Using binary pebble beds, the filling quality should be guaranteed against the occurrence of de-mixing or swimming of the larger pebbles over the smaller ones during the thermal transients. A possible solution is to optimise the filling procedure, to improve the PF and its relevant thermal performance, and also to achieve a stable bed lattice during the cycling loads. In this case, the mechanical characteristics of the pebble beds would also be heavily affected, thus requiring a new tests campaign to determine the actual mechanical properties of an “optimised” pebble bed. This paper presents a new filling optimisation method and the experimental results from the compression tests of optimised pebble beds.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water–graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm 2) are considered. At the present time, development and fabrication of lithium orthosilicate–beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350–700°C. Technical problems associated with manufacturing of the modules are discussed.

Effects of helium production and radiation damage on tritium release behavior of neutron-irradiated beryllium pebbles

The tritium release from neutron-irradiated beryllium pebbles, irradiated under different helium production (0.5–1.0 × 10 3 appm He) and dpa (4.2–8.6) conditions, was studied. From these results, it was clear that the apparent diffusion coefficient at 600°C was significantly affected by irradiation conditions, but returned to normal values at 900°C, apparently due to thermal annealing. Multiple peaks in the tritium release curve at 900°C were observed.

Experimental Investigations on the Thermal and Mechanical Behavior of Single Size Beryllium Pebble Beds

For the next generation fusion reactors with a ceramic breeder blanket the use, as a neutron multiplier, of either a binary bed of large (≈ 2 mm) and small (≈ 0.1–0.2 mm) beryllium pebbles or a single size bed made of 1 mm or 2 mm pebbles is foreseen. The heat transfer parameters of such a binary pebble bed, namely the thermal conductivity and the heat transfer coefficient to the containing wall, have been investigated in the experimental devices PEHTRA and SUPERPEHTRA and have been reported in the companion paper at this workshop.1 In the present paper the results of analogous investigations for the single size beryllium pebble bed are shown and equations are given to correlate the heat transfer parameters.

Compression properties of neutron irradiated beryllium pebbles

Compression properties of neutron irradiated beryllium pebbles manufactured by the rotating electrode method and the Mg reduction method were evaluated from compression load tests at 10% deformation. Neutron irradiation was carried out up to a total fast neutron fluence of about 4×10 25 n/m 2 ( E>1 MeV, about 8 dpa) at 400, 500 and 600°C. There was no significant difference for the two kinds of beryllium pebbles. Additionally, it became clear that not only helium concentration, but also dpa value is an important factor which might influence the compression properties, because the compression load at 10% deformation for the present specimens irradiated at high dpa was smaller than that of the specimens, with similar helium concentration, irradiated at low dpa.