L.V. Boccaccini

Demo blanket technology R&D results in EU

Abstract The European breeding blanket R&D for DEMO is focussed on the development of two concepts: the Water Cooled Lithium Lead (WCLL) and the Helium Cooled Pebble Bed (HCPB). Test Blanket Modules (TBMs) based on these two concepts will be tested in ITER during the first 10 years of operation. The EU R&D programme is mainly focussed on the development and characterisation of materials: structural (EUROFER), breeding (Pb–17Li, Li4SiO4, Li2TiO3) and neutron multiplier (beryllium), and manufacturing technologies (powder and solid HIP, joining, tritium permeation barrier). In the paper a short description of the two EU breeding blanket concepts, the main R&D results, the key issues still to be clarified and the TBM test programme in ITER are reported.

Test Strategy for the European HCPB Test Blanket Module in ITER

Abstract According to the European Blanket Programme two blanket concepts, the Helium Cooled Pebble Bed (HCPB) and a Helium Cooled Lithium Lead (HCLL) will be tested in ITER. During 2004 the test blanket modules (TBM) of both concepts were redesigned with the goal to use as much as possible similar design options and fabrication techniques for both types in order to reduce the European effort for TBM development. The result is a robust TBM box being able to withstand 8 MPa internal pressure in case of in-box LOCA; the TBM box consists of First wall (FW), caps, stiffening grid and manifolds. The box is filled with typically 18 and 24 breeding units (BU), for HCPB and HCLL respectively. A breeding unit has about 200 mm in poloidal and toroidal direction and about 400 mm in radial direction; the design is adapted to contain and cooling ceramic breeder/beryllium pebble beds for the HCPB and eutectic Lithium-Lead for the HCLL. The use of a new material, EUROFER, and the innovative design of these Helium Cooled components call for a large qualification programme before the installation in ITER; availability and safety of ITER should not be jeopardised by a failure of these components. Fabrication technologies especially in the welding processes (diffusion welding, EB, TIG, LASER) need to be tested in the manufacturing of large mock-ups; an extensive out-of-pile programme in Helium facility should be foreseen for the verification of the concept from basic helium cooling functions (uniformity of flow in parallel channels, heat transfer coefficient in FW, etc.) up to the verification of large portions of the TBM design under relevant ITER loading. In ITER the TBM will have the main objective to collect information that will contribute to the final design of DEMO blankets. A strategy has been proposed in 2001 that leads to the tests in ITER 4 different Test Blanket Modules (TBM’s) type during the first 10 years of ITER operation. For the new HCPB design this strategy is confirmed with some additional possibilities taking into account the modular design of the breeding zone.

EU blanket design activities and neutronics support efforts

Abstract An overview is provided of the design activities and the related neutronics support efforts conducted in the European Union for the development of breeder blankets for future fusion power reactors. The EU fusion programme considers two blanket lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles (Li4SiO4 or Li2TiO3) as breeder and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy as breeder and neutron multiplier. The blanket design and the related R&D efforts are based on the use of the same coolant and the same modular blanket structure to minimise the development costs as much as possible. The neutronic support efforts include design analyses for the layout and optimization of the modular HCPB/HCLL blankets based on detailed three-dimensional Monte Carlo calculations as well as underlying neutronics activities conducted in the frame of the European Fusion and Activation File (EFF/EAF) projects to develop qualified nuclear data and computational tools for reliable neutronics design calculations.

Design Update and Mock-Up Test Strategy for the Validation of the EU-HCPB-TBM Concept

In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group of ITER, the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) is developed to investigate DEMO relevant concepts for blanket modules. The main functions of a blanket module (heat removal, tritium breeding and sensitive components shielding) will be tested in DEMO relevant conditions during four different test campaigns in ITER. One TBM of the HCPB concept will be installed into the vacuum vessel connected to one equatorial port designed for vertical TBM orientation during each of the four test campaigns. This paper describes the FZK activities in order to verify the design of the HCPB TBM with regard to operational conditions in ITER and to prove the feasibility of the manufacturing techniques proposed. As the studies performed in FZK up to 2006 concerned a horizontal orientation of the HCPB, a review of the design was necessary to match with the new ITER specifications. Even if the general architecture of the horizontal TBM is maintained, nevertheless the change of configuration has significant impact on the design of the TBM sub-components. An overview of the new vertical HCPB design is presented, detailing the strategy adopted to assess the design and the thermal and fluid dynamic analyses performed for the TBM First Wall. In parallel to the TBM design and analysis, a large mock-up programme addresses the main issues of manufacturing and performances for single components and systems. The three medium-size mock-ups foreseen in FZK to validate the fabrication and mounting processes are presented detailing their purposes.

Review of accidental safety studies for the European HCPB test blanket system

This paper presents a review of safety studies for accidental sequences in the European solid breeder test blanket module (TBM) system. These studies are the starting point for the Preliminary Safety Analysis Report of ITER, under preparation to get the construction permit first and then later the operation licence. In general the reduced inventory of activation products and tritium associated with the TBM system makes the impact of this test system almost negligible on the overall safety risk of ITER. Nevertheless, the possibility of jeopardizing the ITER safety concept has been analysed in connection to the consequences of specific accident sequences, e.g. the pressurization of the vacuum vessel due to the He coolant blow-down, the hydrogen production from the Be-steam reaction, the possible interconnection between the port cell and the vacuum vessel causing air ingress and the necessity to assure heat removal in the short and long periods. In the frame of this assessment, three LOCA sequences have been selected as representative of accidents judged to cover all scenarios envisaged in Cat II to IV events involving the TBM, namely, in-vessel LOCA, ex-vessel LOCA and in-box LOCA.

Tritium permeation issues for helium-cooled breeding blankets

Tritium permeation through Breeding Blanket 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 in order to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modelling and parametric issues. In this study the main permeation issues are identified and possible solutions are discussed to face the modelling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled breeding blankets, i.e.: 1) Helium-Cooled Pebble Beds (HCPB) and 2) Helium-Cooled Lithium-Lead (HCLL). For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC in order 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).

Neutronic analyses of PPA reactor blanket concepts

Two variants of the Helium-cooled Pebble Bed (HCPB) blanket and an advanced version of the Dual Coolant Lithium Lead (DCLL) blanket have been investigated in the framework of the EU Power Plant Conceptual Study-Availability (PPA) with the main objective to explore their potential for a long lifetime and high power loading levels. This work presents the related neutronic analyses performed on the basis of three-dimensional Monte Carlo calculations for the PPA reactor to assess and optimise the nuclear performance of the considered blanket concepts.

Blankets – key element of a fusion reactor – functions, design and present state of development

Abstract Blankets are key elements of a future fusion power reactor, as they breed the fusion fuel tritium, extract the heat from the reactor for power generation and contribute to the nuclear shielding of the plasma confining magnetic field coils. On the way to the engineering implementation of fusion, in particular the blanket design approach has changed substantially. Novel blanket designs require, already from the beginning, incorporating close coupling of plasma physics with engineering physics to develop robust solutions coping with thermal, mechanical and also electrodynamic loads – not only during the stationary operating phase, but also during transients. Simultaneously, nuclear licensing feasibility as well as component failure safety must be part of the design approach. This article describes advanced blanket design approaches undertaken in the past years by the example of the helium cooled pebble bed blanket (HCPB), aiming at an efficient blanket engineering design, starting from the development of modular integral reactor analysis tools, via design analysis and engineering validation of fabrication and interface performance, towards safety analysis on the reactor level.

Power extraction and tritium self-sufficiency

Abstract Major requirements for a future fusion power plant are to ensure the extraction of the fusion power for electricity production and to provide the whole T necessary to sustain the thermonuclear reaction during the plant lifetime. Blanket first wall and divertor targets have to be designed to sustain high heat flux in small surface regions facing the plasma; thick structures have to surround the plasma to capture the neutron energy and protect vacuum vessel and magnets. All these components have to be efficiently cooled to respect strong additional requirements to ensure an efficient operational regime for the power conversion system. The neutrons generated by the plasma have to be used for T generation; for this scope blankets are filled by Li compounds in solid or liquid form. To reach the tritium self-sufficiency the blanket needs an accurate and clever design of the breeder zone and an accurate material selection. Furthermore, recovery time of T that has to become available for the fuel cycle have to be minimised, avoiding large inventories in reactor and in the fuel processing plant. The development of these technologies is the key issues for the development of a first generation of fusion devise after ITER.

Thermal-hydraulic and structural analysis of a helium-cooled first wall mock-up

Abstract The first 3D thermal-fluid-dynamic and structural analyses done for the design and pre-test assessment of the so-called Thermo-Cycle Mock-up (TCM), reproducing about 0.3 m2 of a flat first wall (FW) with relevant helium cooling channels, are presented, based also on previous computational and experimental activities conducted at KIT but limited so far to a single cooling channel with straight heated length. The TCM is the first of a series of FW mock-ups presently under construction, to be tested starting from 2015 in the large HELOKA facility at KIT. Here, the fluid dynamics in the 180° turns of the TCM cooling channels is investigated together with the effects of heat transfer between neighboring channels, when the plate is subject to steady-state heat fluxes in the range 0.3-0.5 MW/m2. Based on the computed temperature maps, the stresses in the TCM and the related damage figures for the main failure modes (i.e., ratcheting and creep/fatigue) are assessed. These are compared with allowable limits in code and standards for the qualification of the TCM design and related to the prediction of the behavior of the component in the actual fusion environment.