The current breeding blankets proposed in the different conceptual fusion power plants are based mainly on the use of Li4SiO4 and/or Li2TiO3 as tritium breeder and Be/Be12Ti as neutron multiplier or ...an eutectic Li17Pb83 for as a hybrid tritium and neutron multiplier. While these materials offer some tritium breeding capabilities, some recent studies show that the tritium self-sufficiency may not be ensured with these materials due to the strong reduction of blanket coverage after the integration of other in-vessel reactor systems (heating and current drive, limiters, large or double-null divertor systems, etc.). Also, some materials like Be raises several key feasibility concerns. The goal of this paper is to perform an update of the screening for tritium breeder and neutron multiplier materials and to assess the tritium breeding performance of the selected compounds in order to reveal new options. As for the neutron multiplier materials, a new subdivision between solid and liquid multipliers is proposed. For the selected compounds, detailed 3D heterogeneous neutronic analyses have been performed with MCNP5-1.60 assuming the architecture of the current EU DEMO Helium Cooled Pebble Bed (HCPB) as a benchmark breeding blanket. From the point of view of ceramic breeders, Li8ZrO6 has been found to outperform Li4SiO4 by more than 4% in terms of tritium breeding, having 6% higher melting point. From the point of view of solid neutron multipliers, Be12Cr, Be12V, Be13Zr and Be13Y show a similar performance as Be12Ti, while LaPb3, Zr5Pb4 and YPb2 offer a solution for a Be-free blanket. As for liquid multipliers, Pb in combination with a ceramic breeder shows a very promising option. Moreover, Pb compounds like Pb90Mn10 and Pb95Ba5 offer similar performance as Pb with a lower melting point (290 °C). Due to the significant advantages of molten Pb as neutron multiplier, future work will be conducted to define a design of a helium cooled Molten Lead Ceramic Breeder blanket, as simple, cost effective blanket concept.
•Discussion and specification of neutronic requirements for a DEMO power plant.•TBR uncertainties are reviewed/discussed and design margins are elaborated.•Limits are given for radiation loads to ...super-conducting magnets and steel structural components.•Available DEMO results are compared to recommended limits and TBR design target.
This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.
The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the 4 BB concepts being investigated in the EU for their possible implementation in DEMO. During 2014 the former “beer-box” BB ...concept based on the ITER’s HCPB Test Blanket Module suffered several design changes so as to meet the different counteracting nuclear, thermohydraulic and thermomechanical requirements. These studies evidenced that the concept is too rigid to meet the tight TBR requirements imposed for the EU DEMO (i.e. TBR≥1.10). Additionally, the complex manifold system with unbalanced helium mass flow in each of the 2 parallel cooling loops made the concept thermohydraulically complex. However, parametric studies during 2015 revealed that the HCPB concept have potential for a better nuclear performance, as well as margin for a significant simplification of the cooling internals by redefining the cooling plates and the architecture of the blanket, building a symmetric flow scheme.
This paper describes the new HCPB concept based on an integrated FW with the breeding zone thermohydraulics and helium manifold systems. The former complex manifold backplates have been compacted and integrated in the cooling plates, releasing ≈300mm of radial space that can be used now to increase breeder zone, the neutron shielding, to reinforce the Back Supporting Structure (BSS) or basically to reduce the reactor size. Detailed neutronic analyses have yielded a TBR of ∼1.20 for the baseline design. Initial analyses show a correct thermohydraulic behavior. Preliminary thermomechanical analyses also indicate that the design can potentially withstand an in-box LOCA at 9 MPa at a level C according to the RCC-MRx code. Future consolidation activities are described, which shall lead to a concept meeting the BB requirements.
•The latest improvements of the McCad conversion approach including decomposition and void filling algorithms is presented.•An advanced interface for the materials editing and assignment has been ...developed and added to the McCAD GUI.•These improvements have been tested and successfully applied to DEMO and ITER NBI (Neutral Beam Injector) applications.•The performance of the CAD model conversion process is shown to be significantly improved.
McCad is a geometry conversion tool developed at KIT to enable the automatic bi-directional conversions of CAD models into the Monte Carlo (MC) geometries utilized for neutronics calculations (CAD to MC) and, reversed (MC to CAD), for visualization purposes. The paper presents the latest improvements of the conversion algorithms including improved decomposition, void filling and an advanced interface for the materials editing and assignment. The new implementations and features were tested on fusion neutronics applications to the DEMO and ITER NBI (Neutral Beam Injector) models. The results demonstrate greater stability and enhanced efficiency of McCad conversion process.
•TBR requirement for tritium fuel cycle re-assessed and set to TBR ≥ 1.05.•Advances in fuel cycle concept, fuel cycle simulations and progress in DEMO design taken into account.•Requirement to ...generate tritium start-up inventory for a follow-up fusion power plant assumed.•New TBR design target specified for DEMO taking into account modelling and computational uncertainties: TBR ≥ 1.15.•TBR design target achievable with DEMO driver breeding blankets
This work presents an up-to-date assessment of the TBR (Tritium Breeding Ratio) requirements for the European DEMO which is under development in the EUROfusion PPPT (Power Plant Physics and Technology) programme. Previous requirements were updated in accordance with the recent advances in the fuel cycle concept, the progress achieved with the DEMO design, and the additional requirement to generate the start-up tritium inventory for a follow-up fusion power reactor. The assessment results in a requirement of TBR ≥ 1.05 which is needed to provide the amount of tritium to the fuel cycle required to attain self-sufficiency. The TBR design target, which includes margins for calculation uncertainties and incomplete models, was set to TBR ≥ 1.15. The attainability of this design target is to be demonstrated with a neutronics calculation using an appropriate simulation model, without considering auxiliary systems, limiters, extra ports, etc. With the latest blanket design versions, this requirement can be fulfilled by the driver blanket candidates considered for DEMO, the HCPB (Helium Cooled Pebble Bed) and the WCLL (Water Cooled Lithium Lead) breeding blankets.
•Description of selected architecture of the DEMO tokamak.•Description of shielding concept.•Rationale of the segmentation of the in-vessel components.•Description a and rationale of selected ...maintenance strategy.•Description of integrated design of the breeding blanket including an overview of technology aspects.
The EU fusion roadmap defines as one of its goals the development of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. This device shall be tritium self-sufficient, produce net electricity, and acts as a component test facility to demonstrate fusion power plant relevant technologies, e.g. those of the breeding blanket.
This article identifies the main DEMO requirements, introduces the rationales for the selected architecture of the DEMO tokamak, and describes how the configuration of the main tokamak components has been derived. This includes (i) the DEMO shielding concept, (ii) the segmentation of the in-vessel components and their maintenance strategy, (iii) an overview of the vessel and in-vessel component technologies, and (iv) a description of the integrated design of the breeding blanket.
•DTE2 is a unique opportunity to validate the numerical tools for ITER shutdown dose rate analysis.•Simulations are performed with the R2Smesh (Rigorous 2-step) approach at the end of DTE2 to predict ...dose rate levels.•The impact of TT and DTE2 campaigns on the dose rate is presented.
The EUROfusion Work Package JET3 programme, established to enable the technological exploitation of the future Deuterium-Tritium (DT) operations at JET over the next years, includes, within the NEXP subproject, a novel Shutdown Dose Rate (SDR) benchmark experiment. The measurement of the SDR due to neutron activation in a fusion machine operating with Deuterium and Tritium is of primary importance for planning its operation in respect of dose limits for the external radiation exposure. The next high-performance DT campaign at JET (DTE2) which will follow a Tritium-Tritium (TT) operation phase, is a unique opportunity to validate the numerical tools for ITER shutdown dose rate analysis, through the comparison between numerical predictions and measured quantities in terms of C/E (Calculation/Experiment) ratios. Within this framework, this work is a pre-analysis of the impact of the future campaigns on the shutdown dose rate. Three-dimensional simulations are performed with the R2Smesh (Rigorous 2-step) approach at the end of DTE2 to predict dose rate levels at different cooling times. The results are presented and discussed in the paper with the major objective to contribute to the optimization of the planned SDR benchmark experiments.
•18 cm blocks of hydrates like ZrH1.6 or TiH2 can potentially lead to less dpa accumulation in the VV compared to the WCLL blanket.•16 N and 17 N produced by coolant activation: cumulated doses after ...7 fpy in the UPC due to the γ 16 N decay have been calculated referring to the current end of life dose of 2 MGy of valves used in PHTS.•Two dedicated experimental rigs at TLK aim at quantifying co-permeation of Q2 species in He/He or He/water set-ups.•R&D efforts aim at characterizing the two anti-permeation coatings, PLD and ALD, in terms of fabrication and performances under relevant operational conditions.•A combination of anti-permeation coatings and CPS would allow to rely on existing technology for the water detritiation.
In the frame of the EUROfusion consortium activities the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL) concepts are being developed as possible candidates to become driver Breeding Blanket (BB) for the EU DEMO, which aims at the tritium self-sufficiency and net electricity production. The two BB design options encompass water or helium as coolants and solid ceramic with beryllium/beryllides or PbLi as tritium breeder and neutron multipliers. The BB segments have evolved towards a more stable conceptual design taking into account multiple feasibility aspects and requirements imposed by interfacing systems. Possible solutions to improve shielding capabilities of Helium cooled BB are investigated and the impact of water coolant activation is assessed by studying the spatial distribution of 16N and 17N isotopes dose rates, in particular in proximity of isolation valves. The reference and back-up technologies for the Tritium Extraction and Removal (TER) from the helium purge gas and the PbLi are developed addressing key feasibility aspects and implications on the tokamak layout and with considerable R&D efforts. As the BB internals offer an ideal environment (high temperatures, thin structural material) to promote the tritium permeation, studies are devoted to establish a tritium balance in the different systems during operation, with special care to the permeation rate and inventory in the coolant. Those are the key parameters for the feasibility assessment and technology selection for the Coolant Purification Systems (CPS).
McDeLicious is an extension to the MCNP Monte Carlo code with the ability to simulate the generation of source neutrons based on D-Li interaction processes. This paper presents an overview of the ...McDeLicious approach for the D-Li neutron source term modeling including latest code extensions, improvements, validation efforts, and applications to IFMIF neutronics analyses. The current McDeLicious capabilities include the generation of source neutrons and photons based on evaluated d +
6,7
Li cross-section data and the sampling of the impinging deuteron beam intensity distribution using fitting functions or probability tables. The McDeLicious software has been updated to comply with the standard of the state-of-the-art version 5 of the MCNP code. A message passing interface program version enables one to run McDeLicious in parallel mode on large computer clusters. McDeLicious is in routine use for IFMIF neutronics calculations employing different nuclear data including a test version of the recent FENDL-3 nuclear data.
Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled ...Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase.
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