[en] The ITER project aims at building a fusion device with the general goal of demonstrating the scientific and technical feasibility of fusion power. The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The strategy for the development of the instrumentation of the HCLL and HCPB Test Blanket Systems, which include the TBMs and their Ancillary Systems, is briefly recalled in this paper, along with the overview of the requirements coming from the harsh operational environment and the main challenges related to the integration with the complex design of the TBS components. (authors)

[en] Highlights: • This work defined options and methods to instrument ITER TBSs based on functional categories: safety, interlock and control and scientific exploitation based on the ITER research program. • Presented the general architecture of the HCLL and HCPB Test Blanket System Instrumentation and Control. • Defined safety and interlock sensors count and technology selection based on preliminary safety analysis. • Discussed the development status of scientific instrumentation, with focus on integration with design and fulfillment of TBM research program. - Abstract: Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives

[en] The assessment of the different components of the Helium Cooled Pebble Bed (HCPB) and Helium Cooled Lithium Lead (HCLL) Test Blanket Modules (TBMs) based on elastic FE analyses has been performed considering main failure modes and respective structural design criteria. This study has been carried out for the operation at full power and for the so-called Approach 2 cooling that is set up at reduced heat extraction capability for compliance with temperature targets to be relevant to DEMOnstration power plant blanket system. Consequently this situation is the worst condition in terms of cooling performance as a result it reveals that many locations in the First Wall (FW) become critical in the sense that certain criteria are not fulfilled, particularly those for immediate plastic flow localization, progressive deformation (ratcheting) and creep-fatigue interaction. Of course one can better reach code compliance by improving the cooling performance (so called Approach 1) but it will degrade the DEMO relevancy. The critical failure modes are re-assessed performing non-linear analysis using an elasto-viscoplatic model, particularly developed for RAFM steels such as the structural material EUROFER97 by Aktaa & Schmitt, and considering the appropriate design criteria of the inelastic route of RCC-MRx and SDC-IC. Despite the conservatism in these criteria, the results show all-clear signal with respect to immediate plastic flow localization and at least much smaller breaches of the limits with respect to the other failure modes. Reporting details and results of the non-linear failure assessments demonstrates their potential and capability in supporting specific and efficient improvements to the design of highly loaded components. The conservatism in the considered design criteria and ways for its reduction are discussed in addition.

[en] Highlights: • New methodology to analyse complex systems under non-periodic dynamic loading. • Joints of HCLL TBM will be subjected to higher loads than those of HCPB TBM. • Dynamic amplification factor under electromagnetic loads ranges from 1.25 to 2.0. - Abstract: European test blanket modules (TBMs) are attached to radiation shields through mechanical joints. The structural integrity of the joints must be ensured under non-periodic dynamic loading caused by accidental electromagnetic events. This requires the transient dynamic analysis of the TBM-Joint-Shield under multiple loading scenarios. As temperatures and dynamic characteristics of the TBM-Joint-Shield system are updated under the course of the ITER project, the design process of the joints can become very time-consuming. This work presents a methodology based on a response spectrum to simplify the design process of the joint under electromagnetic loads. The mechanical system TBM-Joint-Shield was modeled as a single degree of freedom lumped spring-mass-dashpot system. The dynamic amplification was determined for a wide range of frequencies using the Finite Element Method. Eventually, a design envelope for each loading category was obtained, which can be readily used for determining the dynamic amplification factor of any potential new design of the TBM + Joint + Shield system. Results showed that helium cooled lithium lead (HCLL) undergoes higher loads than helium cooled pebble bed (HCPB), resulting in a dynamic amplification factor which varies from 1.25 to 2.0 for a damping of 4%.

[en] Within the framework of the European fusion strategy, two tritium Breeder Blankets concepts are developed to be tested in ITER as Test Blanket Modules (TBM): the Water-Cooled Lithium-Lead (WCLL) which uses the liquid Pb-16Li as a breeder, neutron multiplier and tritium carrier, and the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic and beryllium pebbles as breeder and neutron multiplier, respectively, and helium purge gas as tritium carrier. Both concepts consider as structural material reduced activation ferritic martensitic steel, the EUROFER97 (X10CrWVTa9-1). Pressurized water (15.5 MPa, 295−328 °C for WCLL) and pressurized helium (8 MPa, 300−500 °C for HCPB) are foreseen for heat removal. The TBM is constituted of a BOX (side caps, first wall), stiffened by stiffening plates (SP) and closed on its back, in the manifold area, with back plates with passing-through elements, like stiffening/tie rods, SP nozzles and inlet/outlet water/He pipes. Internally, breeder units, delimited by SP, are cooled-down by cooling plates for the HCPB and double-wall tubes for the WCLL. This paper describes manufacturing technologies for the TBM and deals with the definition of the preliminary Welding Procedure Specifications (pWPS) and feasibility mock-ups for the manifold area and TBM BOX assembly. The transfer and validation of the pWPS developed for the TBM, from P91 to EUROFER97 steel, is demonstrated. Consolidation of the pWPS for the TBM BOX assembly made with HIP process and development for the HCPB TBM manifold area elements, based on GTAW technique, are discussed. Manufacturing tests are performed using the EUROFER97 steel (RCC-MRx, AFCEN Code rules [1]). As part of the WCLL conceptual design development activities, an assembly and manufacturing welding sequence is detailed, focusing on design changes and the impact on the selection of manufacturing technologies. The article will present as well the future fabrication developments and the associated design proposals envisaged on TBM.

[en] Highlights: • The nuclear responses of the European TBMs have been updated with significant improvements. • The nuclear heat in the TBMs has been evaluated to be about a 10% higher than previous evaluation. • The tritium production in the TBMs has been evaluated to be about a 10% higher than previous evaluation. • The shutdown dose rate is found to be compatible with the planned in-situ maintenance operation for the first time. - Abstract: The depiction of the nuclear responses of the ITER European Test Blanket Modules (TBMs), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebbles Bed (HCPB) is presented in this work. Following important components update, and important methodological advances, the nuclear heat and the tritium production have been revisited, giving new estimations 10% higher than the previous evaluation for nuclear heat in both TBMs and to 15% higher for HCPB T production. This has an impact on the thermo-mechanical design of the TBM and the tritium handling. In addition, the Shutdown Dose Rates in the respective port interspace have been characterized in local approach. It shows a performance that could imply compatibility with planned in-situ maintenance activities when analysed in global approach, an improvement with respect to previous evaluations.

[en] Highlights: • Nuclear analysis for European TBMs and shields, in ITER Equatorial Port #16, has been conducted in support of the ‘Concept Design Review’ from ITER. • The objective of the work is the characterization of the Shutdown Dose Rates at Equatorial Port #16 interspace. • The role played by the TBM and TBM shields, the equatorial port gaps and the vacuum vessel permeation, in terms of neutron flux transmission is assessed. • The role played by the TBM, TBM shields, Port Plug Frame, Pipe Forest and the machine in terms of activation is also investigated. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5 × 10"6 n/cm"2 s at port interspace, while the EPP gaps and BSM gaps are responsible for 5 × 10"7 n/cm"2 s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121 μSv/h averaged near the port closure flange was obtained, out of which, only 4 μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range of 300–350 μSv/h were observed in the interspace. Thus, although the SDDR are found above the project target the good performance of the TBM shields design was demonstrated. Two main reasons for high SDDR were identified in this work—the neutron streaming through the equatorial PP gaps and the neutron penetration through the blanket shield modules and vacuum vessel.

[en] Fuel is one of the essential components in a nuclear reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behaviour, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy, and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This monograph surveys current knowledge regarding in-reactor behaviour, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved