CuCrZr alloy is candidate heat sink material for the ITER blanket, first wall, and divertor. During ITER operation it will be exposed to a combination of elevated temperatures, heat flux, and intense fast neutron radiation. This environment will challenge the performance of components and joints based on CuCrZr. To address this issue, mechanical tests were performed with irradiated and reference specimens of CuCrZr and its joints with 316 L(N)-IG (ITER Grade) stainless steel made by hot isostatic pressing. The reactor exposure up to similar to 0.7 dpa was performed in the BR2 reactor at SCK center dot CEN, in water at a temperature of 257 degrees C A special design was used to allow irradiation of specimens axially pre-stressed at different strain levels. The post-irradiation examination included: (i) tensile test, (ii) measurements of plastic deformation of samples axially loaded during irradiation (in situ creep test); (iii) thermal creep tests on irradiated samples. The fracture surfaces were examined in a hot cell using a Scanning Electron Microscope (SEM). The results were compared with data obtained from mechanical tests and SEM/EDX fracture surface analysis on non-irradiated reference samples. The level of a possible creep under irradiation is below the experimental uncertainty.

Fusion for Energy (F4E), the European Union's Domestic Agency for ITER, is responsible for the procurement of about 50% of the ITER blanket first wall (FW), called normal heat flux FW. A procurement strategy has been implemented by the In-Vessel Project Team at F4E aimed at mitigating technical and commercial risks for the procurement of ITER blanket FW panels, promoting as far as possible competition among industrial partners. This procurement strategy has been supported by an extensive Research and Development (R&D) programme, implemented over more than 15 years in Europe, to develop various fabrication technologies. It includes in particular the manufacture and testing of small-scale, medium scale mock-ups and full-scale prototypes of blanket FW panels. In this R&D programme, significant efforts have been devoted to the development of a reliable materials joining technique. Hot Isostatic Pressing was selected for the manufacture of the FW panels made from beryllium, copper chromium zirconium alloy and 316L(N)-IG austenitic stainless steel.This paper presents the main outcome of the on-going R&D programme, the latest results of the FW qualification programme together with the procurement strategy implemented by F4E for the supply of the European contribution to the procurement of the ITER blanket FW. (C) 2016 Published by Elsevier B.V.

Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for inservice inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation. (C) 2015 Elsevier B.V. All rights reserved.

This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm x 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale "fingers" and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels).The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Julich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m(2) is described in detail. (C) 2015 Elsevier B.V. All rights reserved.

The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m(2) in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41-1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich. (C) 2015 Elsevier B.V. All rights reserved.

A typical NHF FW panel consists of a series of fingers, which represent the elementary plasma facing units and are designed to withstand 15,000 cycles at 2 MW/m(2). The fingers are mechanically joined and supported by a back structural element or "supporting beam". The structure of a finger is made of three different materials: stainless steel for the supporting structure, copper chromium zirconium for the heat sink, and beryllium as armour material. Due to their location and to the interfaces with other systems (e.g. Diagnostics, Remote Handling), the NHF FW panels are divided in different main and minor variants. The aim of this paper is to present the design work performed towards the PA signature. CAD detailed models have been created in CATIA for main and minor variants. Examples of local design solutions, as well as design work to achieve the global configuration of specific modules are provided. Finite Element (FE) analyses have been carried out, in order to simulate the operational scenario of ITER and assess the thermo-mechanical behaviour of the most important FW panels against the required design criteria. This design and analyses activity is required to progress towards the finalization of the detailed design of the NHF FW main and minor variants. (C) 2015 Elsevier B.V. All rights reserved.

The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a "semi-prototype" with a dimension of 305 mm x 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m(2)) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles.The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares.In addition, a "tile-size" mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time. (c) 2014 Elsevier B.V. All rights reserved.

This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements.

ITER first wall (FW) panels are a layered structure made of the three following materials: 316L(N) austenitic stainless steel, CuCrZr alloy and beryllium. Two hot isostatic pressing (HIP) cycles are included in the reference fabrication route to bond these materials together for the normal heat flux design supplied by the European Union (EU). This reference fabrication route ensures sufficiently good mechanical properties for the materials and joints, which fulfil the ITER mechanical specifications, but often results in a coarse grain size for the CuCrZr alloy, which is not favourable, especially, for the thermal creep properties of the FW panels. To limit the abnormal grain growth of CuCrZr and make the ITER FW fabrication route more reliable, a study began in 2010 in the EU in the frame of an ITER task agreement. Two material fabrication approaches have been investigated. The first one was dedicated to the fabrication of solid CuCrZr alloy in close collaboration with an industrial copper alloys manufacturer. The second approach investigated was the manufacturing of CuCrZr alloy using the powder metallurgy (PM) route and HIP consolidation. This paper presents the main mechanical and microstructural results associated with the two CuCrZr approaches mentioned above. The mechanical properties of solid CuCrZr, PM CuCrZr and joints (solid CuCrZr/solid CuCrZr and solid CuCrZr/316L(N) and PM CuCrZr/316L(N)) are also presented.

The ITER first wall (FW) panel consists of beryllium in the form of tiles covering its surface, high strength copper alloy as the heat sink material and stainless steel as the structural material. Small-scale normal heat flux FW mock-ups, provided by Fusion for Energy, are tested in the electron beam facility JUDITH 2 at Forschungszentrum Julich to determine the performance of this design under thermal fatigue. The mock-ups are loaded cyclically under a surface heat flux of 2MW m (2) with ITER relevant water coolant conditions. In this study, three-dimensional finite element method thermo-mechanical analyses are performed with ANSYS to simulate the thermal fatigue behaviour of the mock-ups. The temperature results indicate that the beryllium surface temperature is below the maximum allowed temperature (600 degrees C) of beryllium to be tested. The thermal mechanical results indicate that copper rupture and debonding between Be and copper are the drivers of the failure of a mock-up. In addition, the experimental data, e.g. the surface temperature measured using an infrared camera and the bulk temperature measured using thermocouples, are reported. A comparative study between experimental and simulation results is performed.

The ITER blanket is in the final stage of design completion. The issues raised during the 2007 ITER design review about the first wall (FW) heat loads and remote handling strategy have been addressed, while integrating the recently confirmed in-vessel coils. This paper focuses on the FW design, which is nearing completion. Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues.

A Research and Development program for the ITER Blanket-First Wall has been implemented in Europe to provide input data for the manufacture of the full-scale production components. In this frame, FW mock-ups have been fabricated according to ITER FW design requirements.In order to define acceptance criteria for non-destructive examination (NDE) for the series production. FW mock-ups (FWMU) representative of ITER FW are manufactured with calibrated defects to be validated by heat flux tests to assess the critical defect dimensions able to degrade fatigue performance and lifetime, when located at Be/CuCrZr joint corners and beryllium tile edges, and at the CuCrZr/CuCrZr and CuCrZr/316L SS joints.In order to create the defects of given dimensions, two techniques were studied: alumina and zirconia coating using a PVD technique in one hand; and on the another hand alumina and quartz thicker inserts.The paper describes the different approaches used to manufacture test samples with calibrated defects, before applying on FW mock-ups, and related non-destructive examination (NDE) by ultrasonic examination (UT). High heat flux (HHF) testing is not part of this work. (C) 2012 Elsevier B.V. All rights reserved.

Paper introduces thermal fatigue testing devices, developed and operated under fusion-related projects with focus on testing of ITER First Wall (FW) mock-ups. In frame of EFDA tasks, FW mock-up testing device was developed and put into operation in Centrum Výzkumu Řež s.r.o. (CV Rez) research centre, and one testing device was modified for thermal fatigue tests in Forsungszentrum Jülich (FZJ) research centre. The FW mock-ups were tested for several parties, where the key role was played by Fusion for Energy (F4E), the European Union's Joint Undertaking for ITER and the Development of Fusion Energy.

Beryllium is one of the most interesting materials to be used in nuclear fusion devices due to its promising properties. However, testing of beryllium components is quite challenging because of its toxicity and the required safety procedures.Within this paper an overview of high heat flux thermal fatigue tests carried out in the electron beam facility JUDITH 2 (Juelich Divertor Test Facility in the Hot Cells) at Forschungszentrum Juelich (FZJ) on actively cooled beryllium mock ups is given. The performance of the mock ups was documented by temperature measurements as well as optical inspections before and after testing. Special attention was paid to the temperature diagnostics during the tests for two different reasons. Firstly, local overheating of the beryllium tiles, which are the proof of bonding failures, is detected by temperature measurements. Secondly, temperature information is exceptionally important to meet the required safety standards because of the toxicity of beryllium. In order to avoid evaporation of beryllium, a temperature limit for an automatic safety shut down of the electron beam facility JUDITH 2 was set to a certain value. Finally examples of the temperature distribution under different loading conditions for the beryllium mock ups are presented within this paper. (C) 2011 Elsevier B.V. All rights reserved.

The dummy armour prototypes are identical to the reference components in terms of geometry, cooling circuit and material except for the armour material, which is replaced by an equivalent thickness of copper alloy. The main objectives of the dummy armour prototypes are the demonstration of the overall engineering concept of the Divertor, the integration in a 3° cassette together with components manufactured by the other ITER Home Teams and the successive thermo–hydraulic tests on the whole Divertor module. This paper describes the realization of both the wing and the vertical target dummy armour prototypes focusing on the critical aspects of the fabrication and their impact on a further industrialization of the components.

The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions.

In the frame of the ITER Engineering Design Activity (EDA), the European Home Team (EU HT), in collaboration with the Joint Central Team (JCT) and the three other HTs, has contributed to the development of a joint water-cooled solid breeder blanket design concept for the ITER Enhanced Performance Phase. In early 1997 the EU HT gradually arrived at a different basic layout characterized in particular by an enhanced degree of overall reactor relevance and then started a rigorous design activity for its validation. The main peculiarities of the EU HT proposal are (i) the use of beryllium in form of a pebble bed, (ii) the adoption of flat radial-poloidal cooling panels, and (iii) the confinement of the breeder pebbles in poloidally oriented circular elements. In addition to the design rationale, this paper describes the main results of the April-September 1997 design activity in the areas of neutronic, thermal, thermal-hydraulic, electromagnetic, structural, thermo-mechanical as well as tritium and safety analyses. A specific sub-section is devoted to the results of an industrial assessment of the blanket module manufacturing and assembly.