Thermo-mechanical modelling of pebble bed-wall interfaces

In HCPB blankets, interfaces between pebble beds and structural material provide for an additional heat resistance, which depends on local mechanical stresses and temperature. The heat transfer coefficient of pebble bed-wall interfaces was investigated by modelling particle-wall contact, radiation effect, and interstitial gas. The predictions of the model were compared to the experimental data. Interfacial modelling as presented by this paper, which takes the coupled thermo-mechanical behaviour of the interface into account, opens up the possibility to implement these effects in a finite element simulation of a structure containing pebble beds.     

On the theoretical-numerical study of the HEXCALIBER mock-up thermo-mechanical behaviour

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facility.The paper presents the main results of the mock-up experimental tests and of their numerical simulations performed adopting the finite element method, which allowed the DIN constitutive model to be assessed and validated.     

Thermo-mechanical test rig for experimental evaluation of thermal conductivity of ceramic pebble beds

The experimental determination of mechanical and thermal properties of ceramic pebble beds, such as the lithium orthosilicate or lithium metatitanate, is a key issue in the framework of fusion power technology, for the reason that they are possible candidates in the design of breeder blankets.The paper deals with an experimental method for the evaluation of the thermal conductivity of ceramic pebble beds versus the temperature and compressive strain, based on a steady state heat flux through a material (alumina) of known conductivity. The alumina thermal conductivity is determined by means of the hot wire method. To assess the experimental method, a thermo-mechanical characterization of alumina pebble beds (a material largely available), having different diameters, considering a wide range of temperatures and compression forces has been carried out.Moreover preliminary tests have been performed on lithium orthosilicate and lithium metatitanate pebble beds.     

Thermal conductivity of highly neutron-irradiated beryllium in nuclear fusion reactors

Beryllium will be used as a neutron multiplier in Helium Cooled Pebble Bed (HCPB) DEMO blankets. The beryllium thermal conductivity is determining the maximum pebble bed temperature and, therefore, is very important for blanket design. Different grades of beryllium discs were neutron-irradiated at temperatures between 343 and 673 K and at fluences up to 1.6 × 10²³ cm⁻². At lower irradiation temperatures a significant drop of the beryllium thermal conductivity occurs even after small neutron fluences. With increasing neutron fluence, further moderate decreases of the conductivity are observed. With increasing irradiation temperature, the thermal conductivity further decreases. If the thermal conductivity of the irradiated beryllium is known, the conductivity of irradiated beryllium pebble beds can be assessed using the model suggested in this study.     

Preliminary Thermo-mechanical Analysis of Water Cooled Ceramic Breeder Blanket for China Fusion Engineering Test Reactor

Bridging from ITER to DEMO, China Fusion Engineering Test Reactor aims at tritium self-sufficiency which is one of the main functions of blanket. The structure and thermo-mechanical performance influences strongly the operation and tritium breeding of blanket. In this paper, Water Cooled Ceramic Breeder blanket was designed with multilayer mixed pebble beds. And preliminary thermo-mechanical analysis has been done by the coupling of ANSYS finite element (FE) model and self-developed finite difference (FD) code under normal steady state condition. The results showed that the temperature distribution of the FE model corresponds well to that of the FD code. The obtained equivalent stress of the blanket is presented and critically verified the compliance with the SDC-IC code as reference criteria. At last, possible improvements such as adding fillets and plug-in materials are proposed to ameliorate the structure.     

Thermo-mechanical analyses and assessment with respect to the design codes and standards of the HCPB-TBM Breeder Unit

The Helium Cooled Pebble Bed-Test Blanket Module (HCPB-TBM) is one of the two breeding blanket concepts currently under development in Europe. Key component of the HCPB-TBM, the Breeder Unit (BU), has entered the detailed engineering design phase. After establishing a base design, thermal and thermo-mechanical analyses have been performed under typical ITER operational conditions: the results are presented and discussed in this paper.A full scaled finite element model of the base design of the HCPB-TBM BU has been built to run thermal analyses of the beryllium and Li₄SO₄ pebble beds and thermo-mechanical analyses of the BU structure, both in steady state and in a typical transient regime during a pulse of the ITER D–T phase. The temperatures reached in the Li₄SO₄ and beryllium pebble beds in the BU base design are 930.8 °C and 712.9 °C, respectively, which are above the recommended values of 920 °C for Li₄SO₄ and 650 °C for the beryllium pebbles. The maximum temperature in the structural steel is 548.4 °C, which remains under the design limit defined for the TBM studies (550 °C). In order to decrease the temperature in the hot spots identified in the pebble beds, a reduction of the Li₄SO₄ and beryllium bed volumes has been adopted. As for the structural material, the thermo-mechanical analyses have been assessed with respect the RCC-MR design code (completed for irradiation damages with ITER SDC-IC). The results reveal some problematic points in the base design, concentrated in the coolant inlet and outlet pipes and in the connection region of the BU cooling plates with the BU backplate. Submodeling technique has been used to improve the design in these regions. An increase in the thickness of the coolant inlet and outlet pipes and a redesign of the BU backplate have led to a fulfillment of the codes and standards. The design modifications of pebble bed region and structural material have been implemented in the final design of the BU that is presently used as reference for the design and test of a BU mockup in KIT.     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

Construction of PREMUX and preliminary experimental results,as preparation for the HCPB breeder unit mock-up testing

One of the European blanket designs for ITER is the Helium Cooled Pebble Bed (HCPB) blanket. The core of the HCPB-TBM consists of so-called breeder units (BUs), which encloses beryllium as neutron multiplier and lithium orthosilicate (Li₄SiO₄) as tritium breeder in form of pebble beds. After the design phase of the HCPB-BU, a non-nuclear thermal and thermo-mechanical qualification program for this device is running at the Karlsruhe Institute of Technology.Before the complex full scale BU testing, a pre-test mock-up experiment (PREMUX) has been constructed, which consists of a slice of the BU containing the Li₄SiO₄ pebble bed. PREMUX is going to be operated under highly ITER-relevant conditions and has the following goals: (1) as a testing rig of new heater concept based on a matrix of wire heaters, (2) as benchmark for the existing finite element method (FEM) codes used for the thermo-mechanical assessment of the Li₄SiO₄ pebble bed, and (3) in situ measurement of thermal conductivity of the Li₄SiO₄ pebble bed during the tests.This paper describes the construction of PREMUX, its rationale and the experimental campaign planned with the device. Preliminary results testing the algorithm used for the temperature reconstruction of the pebble bed are reported and compared qualitatively with first analyses completed with the FEM codes.     

Development of the IFMIF Tritium Release Test Module in the EVEDA phase

This paper presents the engineering design of the IFMIF (International Fusion Materials Irradiation Facility) Tritium Release Test Module (TRTM). The objectives of the TRTM are: (i) in-situ measurements of the tritium released from lithium ceramics and beryllium pebble beds during irradiation, (ii) studying the chemical compatibility between lithium ceramics and structural materials under irradiation, and (iii) performing post irradiation examinations for the irradiated materials. The TRTM has eight rigs which are arranged in two rows (2 × 4) perpendicular to the beam axis and enclosed by a structural container. Each rig includes one capsule that contains lithium ceramic or beryllium pebbles for irradiation. Neutrons reflectors are implemented at different locations to reflect the scattered neutrons back to the active region aiming to improve the tritium production. The TRTM is required to provide irradiation temperature range of 400–900 °C for the capsules filled with lithium ceramics and 300–700 °C for the ones packed with beryllium. The engineering design of the TRTM components such as container, rigs, capsules, pebble beds, neutrons reflectors, and purge gas and coolant tubes are presented. In addition a test matrix for the irradiation campaign is proposed.     

Experimental investigation on the possible techniques of pebbles packing for the HCPB Test Blanket Module

The Helium Cooled Pebble Bed Test Blanket Module (TBM) features a structural box that consists of the first wall, two caps and a stiffening grid. Inside the stiffening grid the breeding units (BUs), consisting of the beryllium and lithium ceramic pebble beds and cooling plates, are accommodated. The BUs are closed by the BU back plates and several structural plates of the manifold system as well as the TBM back plate consequently the BUs may not be accessed directly after the assembly of the TBM box; however, access is possible through dedicated penetrations in the TBM caps. According to the current manufacturing strategy, the assembly of the TBM structural sub-components is based on several welding processes which require post-welding heat treatments (PWHT) at temperatures which exceed the temperature limit of the beryllium pebbles. For that reason the beryllium pebble beds will be packed after the TBM box is assembled and heat treated. The packing of the BUs will be performed using a small-diameter (5 mm) tube that will be inserted into some penetrations in the TBM caps. It is expected that the lithium ceramic pebbles can withstand the high temperatures of the PWHT (this assumption needs to be verified) therefore the current strategy is to pack the ceramic pebble beds during the TBM box assembly. This study experimentally demonstrates the packing procedures for the beryllium beds using a full-scale Plexiglas mock-up as well as the optimization of the packing process by dedicated measures such as vibrating and tilting of the mock-up. In addition the impacts of the experimental results on the TBM design are summarized and the paper is concluded by proposing a packing strategy that can be used to achieve a packing factor of 63.6%.     

Study of ceramic pebble beds in Post Irradiation Examination of the Pebble Bed Assemblies irradiation experiment

The irradiation experiment Pebble Bed Assemblies (PBA) consists of four mock-up representations (test elements) of the EU Helium Cooled Pebble Bed (HCPB) concept. The four test elements contain a ceramic breeder pebble bed sandwiched between two beryllium pebble beds and are regarded as one of the first DEMO representative HCPB blanket irradiation tests, with respect to temperatures and power densities. The design value of the PBA were to irradiate pebble beds at a power density of 20–26 W/cc in the ceramic breeder, to a maximum temperature of 800 °C.Two test elements contain lithium orthosilicate pebbles (Li₄SiO₄; FZK/KIT) and were irradiated with target temperatures of 600 and 800 °C, respectively. The other test elements have lithium metatitanate (Li₂TiO₃; CEA) with different grain sizes and were both irradiated with a target temperature of 800 °C. The PBA have been irradiated for 294 Full Power Days (12 cycles) in the High Flux Reactor (HFR) in Petten to a total neutron dose of 2–3 dpa in Eurofer, and an estimated (total) lithium burnup of 2–3% in the ceramic pebbles.This work presents results of Post Irradiation Examinations (PIE) on the four HCPB test elements. Using e.g. SEM, the evolution of compressed pebble beds and pebble interactions like swelling, creep, sintering, etc., under irradiation and thermal loads are studied for the candidate pebble materials Li₂TiO₃ and Li₄SiO₄. (Chemical) interactions between ceramic pebbles and Eurofer (e.g. chrome diffusion) are observed. Looking at different sections of the pebble beds, correlations between temperatures and thermal–mechanical behaviour are clearly observed.     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).     

Examination of C/C flat tile mock-ups with hypervapotron cooling after high heat flux testing

Two C/C flat tile mock-ups with a hypervapotron cooling concept, have been successfully tested beyond ITER specification (3000 cycles at 15 MW/m², 300 cycles at 20 MW/m² and 800-1000 cycles at 25 MW/m²) in two electron beam testing facilities [F. Escourbiac, et al., Experimental simulation of cascade failure effect on tungsten and CFC flat tile armoured HHF components, Fusion Eng. Des., submitted for publication; F. Escourbiac, et al., A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology, Fusion Eng. Des. 75-79 (2005) 387-390]. Both mock-ups provide a SNECMA SEPCARB^® NS31 armour, which has been joined onto the CuCrZr heat sink by active metal casting (AMC) and electron beam welding (EBW). No tile detachment or sudden loss of single tiles has been observed; a cascade-like failure of flat tile armours was impossible to generate. At the maximum cyclic heat flux load of 25 MW/m² all tested tiles performed well except one, which revealed already a clear indication in the thermographic examination at the end of the manufacture. Visual examination and analysis of metallographic cuts of the remaining tiles demonstrated that the interface has not been altered. In addition, the shear strength of the C/C to copper joints measured after the high heat flux (HHF) test has been found to be still above the interlamellar shear strength of the used C/C material. The high resistance of the interface is explained by a modification of the C/C to copper joint interface due to silicon originating from the used C/C material.     

R&D progress of Korean HCSB TBM

Several R&Ds are being performed for Korean helium cooled solid breeder (HCSB) test blanket module (TBM) in the field of hydrogen isotopes permeation characteristics measurement in the helium purge line, joining technologies of structural materials, breeder pebble materials development, and the measurement of pebble bed characteristics. Electron beam welding for reduced activated ferritic–martensitic (RAFM) steel is evaluated to find optimal welding conditions. Also, a hydrogen permeation measurement apparatus is newly installed for the evaluation of the permeation barrier characteristics of stainless steel and RAFM steels. Two fabrication methods of lithium orthosilicate pebbles are investigated using slurry droplet methods. As methods of silicon carbide coating on the graphite pebble, microwave coating and chemical vapor deposition coating are evaluated. Two apparatuses are established to assess the thermo-mechanical properties of graphite and breeder pebble beds. The current status of R&D activities on these areas is introduced and the main progresses are addressed in this paper.     

A Measurement Platform Scheme and Data Post-processing Method for Thermal Conductivity of Li<Subscript>4</Subscript>SiO<Subscript>4</Subscript> Pebble Bed

The effective thermal conductivity of tritium breeder pebble bed is an important thermal parameter and must be known for the thermo-mechanical design of solid tritium breeder blankets. In order to obtain the parameter, experimental measurement is an effective method. A measurement platform was designed by University of Science and Technology of China for CFETR solid blanket scheme to measure the immediate thermal conductivity data and study the effect of pebble bed temperature, the purge gas pressure and pebble deformation on the thermal conductivity of pebble bed. Measurements were performed based on about 1 mm diameter Li₄SiO₄ pebbles in the temperature range between 100 and 800 °C, with purge gas pressure ranging from 0.1 to 0.3 MPa. This paper described a measurement platform scheme by thermal probe method. On the other hand, for the sake of increasing the precision of thermal conductivity data transformed from temperature data, some improvements for the data post-processing using Monte Carlo inversion method were made in this paper too.     

Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code.This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported.     

Investigation of the packing structure of pebble beds by DEM for CFETR WCCB

Li₂TiO₃/Be₁₂Ti mixed pebble beds with multi-sized particles are one of the potential candidates for the WCCB (water-cooled ceramic breeder blanket) of the CFETR (China Fusion Engineering Test Reactor). To meet the neutronics requirements of a WCCB, a study of the packing structure of the concerned pebble bed is necessary. In this paper, the discrete element method (DEM) is applied to produce a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. According to the current simulation, the packing factor of a mono-sized pebble bed is 0.62–0.64, while the value will become more than 0.75 for Li₂TiO₃/Be₁₂Ti mixed breeding pebble bed with a diameter ratio of not less than 5 as well as an appropriate mixed volume ratio, and thus can meet the neutronics requirements.     

Development of a synthesis method of beryllides as advanced neutron multiplier for DEMO reactors

Advanced neutron multipliers with low swelling and high stability at high temperature are desired for pebble bed blankets, which will impact on the DEMO design features, especially blanket operating temperature. Among candidates for multiplier, beryllium intermetallic compounds (beryllides) are one of the most promising advanced neutron multipliers. In order to fabricate the beryllide pebble, beryllide with the shape of rod are necessary if a melting granulation method is applied. A plasma sintering method and a high-frequency heating method have been proposed as a new technique, because these methods are simple process and easy to control. A plasma sintering method for beryllide synthesis has been proposed as a new technique which uses a non-conventional consolidation process. The present paper describes beryllide synthesis results using this method for development of fabrication technique for a beryllide rod. It was clarified that the beryllide could be both synthesized and jointed with no variation of phase and hardness by the plasma sintering method. Additionally, trial synthesis of beryllide by a melting process using high-frequency heating was carried out. Although there were some pores in a melted ingot, beryllides could be also synthesized by this method.     

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li₂TiO₃ pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.     

Neutronic investigation and activation calculation for CFETR HCCB blankets

The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder(HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor(CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio(TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil.The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1?×?10-4 k W, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.