Lorenzo V. Boccaccini

Overview of the HCPB Research Activities in EUROfusion

In the framework of the EUROfusion’s Power Plant Physics and Technology, the working package breeding blanket (BB) aims at investigating four different BB concepts for an EU demonstration fusion reactor (DEMO). One of these concepts is the helium-cooled pebble bed (HCPB) BB, which is based on the use of pebble beds of lithiated ternary compounds and Be or beryllides as tritium breeder and multiplier materials, respectively, EUROFER97 as structural steel and He as coolant. This paper aims at giving an overview of the EU HCPB BB Research and Development (R&D) being developed at KIT, in collaboration with Wigner-RCP, BUTE-INT, and CIEMAT. The paper gives an outline of the HCPB BB design evolution, state-of-the-art basic functionalities, requirements and performances, and the associated R&D activities in the areas of design, functional materials, manufacturing, and testing. In addition, attention is given also to the activities dedicated to the development of heat transfer augmentation techniques for the first wall and the corresponding testing. Due to their nature as design drivers, a brief overview in the R&D of key HCPB interfacing areas is given as well, namely, the tritium extraction and recovery system, the primary heat transfer and power conversion systems, and safety topics, as well as some specific activities regarding the integration of in-vessel systems through the BB. As concluding remarks, an outline of the standing challenges and future R&D plans is summarized.

LOCA Accident for the DEMO Helium Cooled Blanket

The exploitation of Fusion as energy source requires also the demonstration of a limited impact in terms of risk to the staff, to the public, and to the environment, well below the limits established by international committees and national safety authorities. Therefore, a systematic safety analysis has to follow the design development to demonstrate that the safety objectives are met for each proposed solution. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. This analysis points out the dominant accident sequences and outlines the possible prevention, protection and mitigation actions and their associated systems. One of the most challenging accidents is a large break Loss of Coolant Accident (LOCA) of the Primary Heat Transfer System (PHTS) outside the Vacuum Vessel (VV), due to the possible consequences in terms of radiological releases to the environment. However, because of the relative small radiological inventory and to the lower decay heat density, the risk associated with a break of the primary cooling loop in a fusion reactor is lower than the risk of the same event in a fission reactor. Nevertheless the consequent peak of pressure in the Expansion Volume located within the Tokamak Building could severely impact the confinement function, hence the overall safety of the plant. For this purpose a numerical assessment of a blanket PHTS ex-vessel LOCA has been carried out considering two possible layout solutions. This analysis has been performed employing MELCOR 1.8.2 and aims to support the design of the Blanket and its PHTS with some safety-related considerations.

Tritium Transport Issues for Helium-Cooled Breeding Blankets

Tritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d).

Effect of thermal loads on different modules of DEMO PFCs

Abstract The thermal performance of different modules of plasma-facing components (PFCs) is analyzed for the DEMO reactor conditions in steady-state operation with the inclusion of the transient edge-localized modes (ELMs) for mitigated and unmitigated cases. As an example, the effect of these loads is considered for the tungsten (W) alloy mono-block design with a Cu OFHC/EUROFER water coolant tube first proposed in the framework of the Power Plant Physics and Technology (PPP&T) divertor study. A variant of this design with a EUROFER tube connected to the W block with a diamond/copper composite (DCC) used in the diagnostic windows is also analyzed. A design goal is to find the optimal thicknesses of material layers that allow one to keep the maximum temperatures within the allowable design limits under ITER water cooling conditions. Heat transfer and armor erosion due to the plasma impact has been modeled by using the MEMOS code.

Impact of pulsed operation on lifetime of DEMO blanket

The impact of pulsed operation on thermo-mechanical fatigue and creep-fatigue lifetimes of DEMO blanket is evaluated. Non-linear finite element (FE) analyses are performed for the current design of the helium cooled pebble bed test blanket module (HCPB-TBM), and assessed predicting the cyclic lifetime by evaluating the damage evolution obtained in the component for the candidate structural material. Therewith the influence of the pulse duration on the cyclic lifetime and hence full operation lifetime of the HCPB-TBM is investigated and evaluated considering the gain in operation time due to pulse durations longer than that specified for ITER. For very long pulse durations the gain in operation time is reduced due to creep-fatigue and consequently optimal pulse duration exists.

Effect of transient thermal loads on tungsten monoblock module in demo

Abstract Thermo-hydraulic analyses of the tungsten mono-block divertor module with a water cooling tube made from a diamond/copper composite (DCC) as a laminate and a martensitic steel EUROFER against the power loadings expecting in DEMO operation is presented. Thermal analysis is carried out by using the code MEMOS, which simulates W armor damage under the repetitive edge localized modes (ELM) heat impact. Heat transfer to the water coolant is studied for various coolant conditions which allow one to keep the material temperatures within the allowable design limits under neutron irradiation. The thermal performance is analyzed for the DEMO I and DEMO II reactor conditions for un-mitigated and mitigated ELMs. The importance of W vapor shielding effect is discussed.

Thermo-mechanical analysis of the DEMO FW module

Thermomechanical performance of the first wall (FW) W/EUROFER sandwich type module is analyzed under DEMO reactor conditions. Engineering heat loads to the FW panels are estimated for steady state operation with the edge localized modes (ELMs). Calculations carried out by MEMOS code show the inhomogeneity of the material temperature due to discrete location of the water cooling tubes embedded into EUROFER. The hot spots are formed in the W armor and EUROFER between the cooling sectors and depend on the distance of their mutual locations. The bending stress due to vertical temperature gradients in W and EUROFER layers is calculated and remains smaller than the ultimate tensile stress for expected temperatures. Calculations show that under the Type I ELMs expected in DEMO the W surface melts at the ELMs peak positions and solidifies between ELMs. There is no temperature difference found between hot and cool spots during ELMs.