Breeding Blanket Concepts Research Articles

Engineering studies for integration of the test blanket module (TBM) systems inside an ITER equatorial port plug

The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak. The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port. This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system.

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments. In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation. The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.

A study of the potential influence of frame coolant distribution on the radiation-induced damage of HCLL-TBM structural material

Within the European Fusion Technology Programme, the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept is one of the two EU lines to be developed for a Long Term fusion reactor, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. The HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a steel-supporting frame, actively cooled by pressurized water. That supporting frame has been designed to house two different TBMs, providing two cavities separated by a dividing Plate 20 cm thick. As the nuclear response of HCLL-TBM might vary accordingly to the supporting frame configuration and composition, at the Department of Nuclear Engineering of the University of Palermo, a parametric study has been launched to investigate such an influence. Previous works dealt with the dependence of the nuclear response of HCLL-TBM on the configuration of a homogeneous frame, the present one has been focused on the investigation of the potential influence of coolant distribution within the frame on the radiation-induced damage of the HCLL-TBM structural material. To this purpose, a detailed parametric study of the HCLL-TBM nuclear response has been performed by means of 3D-Monte Carlo neutronic analyses to asses both the rates of displacements per atom and helium and hydrogen production within the structural material. A semi-heterogeneous model of the supporting frame, assuming a realistic coolant distribution, and a 3D heterogeneous model of the HCLL-TBM, taking into account EUROFER as structural material, has been set-up. Therefore, the models of the TBM and of its frame have been inserted into the existing 3D ITER one, simulating realistically the reactor lay-out up to the cryostat and providing for a proper D-T neutron source. The analyses have been performed by means of the MCNP-4C code, and the results obtained are reported and critically discussed.

Helium-cooled test blanket module box behaviour under accidental pressurisation

The helium-cooled lithium–lead (HCLL) breeder blanket concept is one of the two breeder blanket lines presently developed by the EU for DEMO reactor. In the short-term so-called DEMO relevant test blanket modules (TBMs) of these breeder blanket concepts shall be designed, manufactured, tested, installed, commissioned and operated in ITER for first tests in a fusion environment. For the purpose of licensing such test module in the ITER facility, a safety assessment of the different possible accidental conditions has to be performed. This paper presents the results of the thermo-mechanical calculations that have been obtained in case of internal module leak occurring between the 8.0 MPa pressurised helium circuit and the lithium–lead circuit. In a conservative manner, it is also assumed that helium cooling of the TBM box is stopped at the accident initiation. In such situation the TBM box will be entirely pressurised at 8.0 MPa and the different challenges are related to the thermo-mechanical behaviour, the plasma power hold up and the heat removal capacity of the module. An assessment to estimate the time when the TBM box can withstand the different loads induced by the situation is presented. Particularly the available time span to trigger an emergency plasma shutdown is estimated in regards of the maximal allowable stress for the EUROFER which is the test blanket module material. A 3D finite element model has been developed with CAST3M computer code taking into account the beryllium layer, the decay heat after shutdown and the thermal radiation phenomena. After a description of the model, the paper presents and explains the results and particularly the methodology followed to determine the maximal allowable stress location which must combine the thermal transient calculations and the pure mechanical calculation under the 8.0 MPa loading. The results show that the hottest spot of the first wall (FW) was the most challenged location and that the structure can withstand such accidental conditions without plasma shutdown up to 15 s. Afterwards, the structural design criteria defined by ITER for in-vessel components (SDC-IC) is not any more fulfilled. This does not necessary mean that a lithium–lead leakage can occur in the vacuum vessel. More research and development are needed to have a clear understanding of crack propagation and break size in a material like EUROFER where few data are available, particularly at the temperature reached in this accident. Finally as a conclusion of the studies presented in this paper, it can be stated that, although very conservative assumptions were taken, the time span is large enough to trigger a plasma shutdown in order to avoid more severe operational and safety consequences of this accidental situation.

Li 2TiO 3 pebble beds constitutive equations of using standard three-axial tests: Implementation of the model in an FEM code

During the ITER operation, various Test Breeding Modules (TBM) will be tested in the equatorial ports of the reactor. These TBMs are currently being developed in several research centres around in the world. The Helium-Cooled Pebble Bed (HCPB) blanket is one of two breeder blanket concepts developed in Europe. HCPB uses Beryllium pebbles act as a neutron multiplier and Li 4SiO 4 or Li 2TiO 3 pebbles act as breeder materials. The structural analysis of the breeding blanket is complex in terms of its geometry and loading conditions. Advanced computer codes are needed in order to determine the reliability of the various designs. In this context the mechanical characterization of pebble beds is important in order to simulate their behaviour. This paper illustrates three-axial experimental tests performed on Li 2TiO 3 pebble beds. In addition, preliminary results of the numerical simulations are shown. The numerical analyses were performed using an original model, based on the elastic–plastic behaviour of a single pebble, implemented in an FEM code. The paper highlights the importance of the confining pressure on the mechanical behaviour of pebble beds and the possibility of simulating them with homogeneous materials.

Current status of design and analysis of Korean Helium-Cooled Solid Breeder Test Blanket Module

It is known that the test of the capability of tritium self-sufficiency and the extraction of high-grade heat using tritium breeding blanket concepts is one of the major ITER missions. This requires the development of test blanket modules based on a corresponding DEMO blanket design, despite the differences in operating conditions between DEMO and ITER. Two blanket concepts such as a Helium-Cooled Solid Breeder (HCSB) blanket and a Helium-Cooled Molten Lithium (HCML) blanket are being considered as Korean Test Blanket Module (TBM) for ITER with the aim for verifying the capability of the breeding blanket design and manufacturability. In this paper, the current status of the design and analysis for the HCSB TBM is described. The key features of the design include the use of graphite neutron reflectors to reduce the amount of beryllium multiplier. Major analyses have been executed in order to optimize the design. The results of nuclear and thermo-hydraulic analysis show that the design parameters and requirements are satisfied. Also, the safety analysis results show that the design of the TBM system is sufficient to endure several accident events.

Overview of Recent Japanese Activities and Plans in Fusion Technology

For ITER, trial fabrications of a superconducting cable for a Toroidal Field coil conductor and jacket sections for a central solenoid conductor were successfully performed. In the radio-frequency heating technology, development of 170 GHz ITER gyrotron has progressed to achieve a 600 kW for 3600 sec operation in JAEA RF test stand. In LHD, long-pulse operations near 1 hour with 1.6 GJ inputs have been achieved with ICH and ECH, and subcooled operations of SC magnets to improve cryogenic stability are scheduled soon. In the ITER Test Blanket Modules project, Japan has a plan to lead a water cooled ceramic breeder blanket concept, and to cooperate with other parties on other blanket concepts, such as, a helium cooled ceramic breeder blanket, a LiPb blanket, a Li/V blanket, and a molten salt blanket. As the Broader Approach activities between Europe and Japan, ITER satellite tokamak, JT-60SA, IFMIF-EVEDA and the International Fusion Energy Research Center programs will start soon in Naka and Rokkasho, respectively. Basic researches on broad areas have been enhanced with collaborations in the Fusion Engineering Network among Japanese universities and research institutes.

The test blanket modules project in Europe: From the strategy to the technical plan over next 10 years

The testing of breeding blanket concepts in ITER is recognized as an essential milestone in the development of a future reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeding blankets for DEMO reactor specifications that will be tested in ITER under the form of test blanket modules (TBMs): the helium-cooled lithium-lead (HCLL) and the helium-cooled pebble-bed (HCPB) blanket concepts. Following the decision on ITER construction, the EU has reviewed and detailed the fundamental elements for implementing the future TBMs project aimed at delivering TBMs systems to ITER under suitable schedule and acceptance standards. The time schedule up to ITER operation being tight, the technical plan of the TBMs project over the next 10 years has been detailed.

A study of the potential influence of frame coolant on HCLL-TBM nuclear response

Within the European Fusion Technology Programme, the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept is one of the two EU lines to be developed for a long term fusion reactor, in particular with the aim of manufacturing a test blanket module (TBM) to be implemented in ITER. The HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a steel-supporting frame, actively cooled by pressurized water. This supporting frame has been designed to house two different TBMs providing two cavities separated by a dividing plate 20 cm thick. As the nuclear response of HCLL-TBM could vary with the supporting frame configuration and composition, a parametric study has been launched to investigate such an influence. Previous works dealt with the dependence of the nuclear response of HCLL-TBM on the configuration of an homogeneous frame while the present one has been focussed on the investigation of the influence of coolant distribution within the frame. A detailed parametric study of HCLL-TBM nuclear response has been performed by means of 3D-Monte Carlo neutronic and photonic analyses, simulating the frame in a semi-heterogeneous way. Three-dimensional heterogeneous models of HCLL-TBM and of the supporting frame have been set-up considering both the usual poloidal lay-out and a toroidal one and taking into account 9% Cr martensitic steel (EUROFER) as structural material. The models have been inserted into an existing 3D semi-heterogeneous ITER-FEAT one, simulating realistically the reactor lay-out up to the cryostat and providing for a proper D-T neutron source. The analyses have been performed by means of the MCNP-4C code, running a large number of histories (2·10 8) in such a way that results obtained are affected by statistical uncertainties lower than 1%. The results obtained are reported and critically discussed.

Recent progress in DEMO fusion core engineering: Improved segmentation, maintenance and blanket concepts

After finalization of the European Power Plant Conceptual Study [D. Maisonnier, et al., A conceptual study of commercial fusion power plants, Final Report of the European Fusion Power Plant Conceptual Study (PPCS), EFDA-RP-RE-5.0, 2005 [1]] a new European effort was started to develop an optimized power core for a He-cooled DEMO device. The candidate breeder blanket concepts considered are the HCPB (helium cooled ceramic breeder) concept, the HCLL (helium cooled lithium lead) concept and the DCLL (dual coolant) concept. One basic segmentation and maintenance configuration involving so-called ‘multi-module segments’ (MMS) was identified to be particularly promising. In this paper, an overview on the concept is given and key design features are described. In addition, the design adjustment of the HCPB blanket and some new proposals in the DCLL blanket design are briefly discussed.

On the effects of the supporting frame on the radiation-induced damage of HCLL-TBM structural material

Within the European Fusion Technology Programme, research activities have been conducted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is planned to be located in an ITER equatorial port, housed inside an AISI 316 stainless steel-supporting frame. Since that frame has been designed to provide two positions separated by a dividing plate and the HCLL-TBM is expected to fill one of them, its nuclear response could vary depending on the filling status of the other position and on the plate thickness. A parametric study has been carried out to investigate the potential effects on the radiation-induced damage of the HCLL-TBM structural material, due to the dividing plate thickness and to the presence of either a void, a water-cooled steel plug, or a Helium-Cooled Pebble Bed TBM in the adjacent frame position.

Overview of recent Russian materials and technologies R&D activities related to ITER and DEMO constructions

An overview is given of the activities and major achievements within recent R&D performed in Russia on materials and technologies for ITER and DEMO. In Russia, the basic materials manufacturing and technologies have been selected for ITER, for two reference DEMO breeding blanket concepts and for the related long term R&D. The review on the recent results of investigations on low activation materials (V–Ti–Cr alloys, Fe–12Cr–2W–V–Ta steel EK-181), beryllium and superconducting materials is presented. The fabrication of tubes, sheet and other forms from low activation materials is mentioned. The activity in beryllium materials both in the domestic studies and international cooperation is outlined. The progress in enhancement of the properties of superconducting materials for the ITER magnet system is presented, and the prospect of further developments in superconducting materials for DEMO magnet system is analyzed.

ITER-Test blanket module functional materials

Solid breeder and liquid breeder blanket concepts are being developed for testing in ITER. In addition to fundamental investigations on the use of reduced activation ferritic/ferritic–martensitic steels (RAFMS) or V-alloys as the structural material, tritium breeder and neutron multiplier materials, and corresponding corrosion issues, there are three classes of functional materials being developed. Performance of these functional materials will significantly impact the performance of respective blanket concepts. Two classes of materials are related to the self-cooled liquid breeder design concepts. The first class is MHD coatings for the self-cooled Li breeder concepts. The second class is for flow channel inserts (FCI) for the dual coolant PbLi concepts. The third class of functional material is tritium permeation barrier material commonly needed for solid and liquid breeder blanket concepts. A description and the development status of these functional materials are described in this paper.

An overview of dual coolant Pb–17Li breeder first wall and blanket concept development for the US ITER-TBM design

An attractive blanket concept for the fusion reactor is the dual coolant Pb–17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb–17Li is circulated for power conversion and for tritium breeding. A SiC f/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb–17Li and as the thermal insulator to separate the high temperature Pb–17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb–17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb–17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

An overview of US ITER test blanket module program

A testing strategy and corresponding test plan have been presented for the two proposed US candidate breeder blankets: (1) a helium-cooled solid breeder concept with ferritic steel structure and Be neutron multiplier, but without a fully independent TBM and (2) a dual-coolant helium-cooled ferritic steel structure with self-cooled LiPb breeding zone that uses a flow channel insert as MHD and thermal insulator. Example test module designs and configuration choices for each line of ITER TBM are shown and discussed in the paper. In addition, near-term R&D items for decision-making on testing of both solid breeder and dual-coolant PbLi liquid breeder blanket concepts in ITER are identified.

Review of the activity in Russia in the area of fusion nuclear technology

The Russian Federation (RF) fusion nuclear technology program is mainly focused on the activities for the preparation of the ITER construction, the development of the DEMO project and the implementation of the DEMO breeding blanket in the ITER test blanket modules (TBM). Approaching the construction of ITER, the activity in Russia is concentrated on preparing Russian industry for the procurement of components in accordance with the ITER agreement. These components include vacuum vessel ports, first wall, divertor private region, limiter and blanket supporting systems. The main directions is the improvement of manufacturing technologies (e.g. industrial production of the materials in needed semi-finished forms, and the optimization of joining technologies) with the goal of reducing construction cost and, increasing reliability and quality. To support this activity and to provide the justification for the design and performance of the ITER in-vessel components, some activity is also ongoing in the area of materials studies, mainly on neutron irradiation effects on copper alloys and joints. Short description of the RF test blanket modules development program strategy, the ongoing R&D activities and the main results obtained so far are also reported. The Russian TBM program is focused on the development of two concepts: the helium-cooled pebble bed and liquid lithium/vanadium. Currently, the main activities are focused on the integration of the DEMO breeding blanket concepts into the design of the ITER test blanket modules.

On the influence of the supporting frame on the nuclear response of the Helium-Cooled Lithium Lead Test Blanket Module for ITER

Within the European Fusion Technology Programme, very intense research activities have been promoted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the specific aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a proper steel-supporting frame. In particular, since that frame has been designed to provide two cavities separated by a dividing plate and HCLL-TBM is foreseen to fill just one of them, its nuclear response could vary accordingly to the filling status of the other one, unless the dividing plate is thick enough to isolate the components housed in the two cavities, avoiding any significant nuclear interaction among them. At the Department of Nuclear Engineering of the University of Palermo a research campaign has been launched to investigate the potential impact on the HCLL-TBM nuclear response of either the dividing plate thickness or the filling status of the frame cavities. A computational approach based on the Monte Carlo method has been followed and a detailed parametric study has been carried out considering thicknesses of the dividing plate ranging from 0 to the reference value (20 cm) and assuming, for each case, the presence of either void, or a water-cooled steel plug, or a HCLL-TBM or a Helium-Cooled Pebble Bed (HCPB) TBM in the other frame cavity. The analyses have been carried out by means of the MCNP-4C code, running a large number of histories (2 × 10 7) so that the results obtained are affected by statistical uncertainties lower than 1%. The results obtained are herewith reported and critically discussed.

Carbon tiles as spectral-shifter for long-life liquid blanket in LHD-type reactor FFHR

In terms of engineering feasibility for long-life Flibe blanket in LHD-type reactor FFHR, the Spectral-shifter and Tritium breeder Blanket (STB) concept is evaluated by taking neutron irradiation effects into account under system integration such as Flibe cooling and components replacement. FEM calculations for the neutron wall loading of 1.5 MW/m 2 show that the temperature of the STB armor tile can be kept below 2000 K by optimizing the first metal wall thickness. The heat load experiment on the STB armor mockup confirms feasibility of the temperature control and mechanical joining. Degradation of STB armor tiles due to neutron irradiation requires replacement of them every few years by means of remote handling “screw coasters” using helical winding, where the replaced tiles are low level wastes. Although the STB concept is feasible within nuclear and thermal properties, more detailed structural optimization is needed including the mechanical and chemical properties.

The helium-cooled lithium lead blanket test proposal in ITER and requirements on test blanket modules instrumentation

In this paper, a preliminary testing program is proposed to test and validate the helium-cooled lithium lead breeder blanket concept in the ITER machine. The proposal envisages four different test mock-ups or modules, adapted for qualifying single or combined effects and whose design makes large use of engineering scaling for compensating the differences between the testing conditions and those expected in DEMO (e.g., neutron wall load, heat flux, pulsed operating conditions).

Present development status of EUROFER and ODS-EUROFER for application in blanket concepts

Within the European Union, the two major breeding blanket concepts presently being developed are the helium cooled pebble bed (HCPB), and the helium cooled lithium lead (HCLL) blankets. For both concepts, different conceptual designs are being discussed with temperature windows in the range 250–550 °C for conservative approaches based on reduced activation ferritic–martensitic (RAFM) steels, and in the range 250–650 °C for more advanced versions, taking into account oxide dispersion strengthened (ODS) steels. As a final result of a systematic development of RAFM-steels in Europe, the 9% CrWVTa alloy EUROFER was specified and produced in an industrial scale with a variety of product forms. A large characterisation program is being performed including irradiation in materials test reactors between 60 and 450 °C (≤15 dpa), and in a fast breeder reactor at 330 °C up to 30 dpa. EUROFER is resistant to high temperature ageing, and the existing creep-rupture data (∼30,000 h, 450–600 °C) indicate long-term stability and predictability. The ODS variant of EUROFER shows superior tensile and creep properties compared to EUROFER. Applying a new production route has diminished the problem of lower ductility and inferior impact properties. A reliable joining technique for ODS and RAFM steels employing diffusion welding was successfully developed.