Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment. (C) 2016 Elsevier B.V. All rights reserved.

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase. (C) 2015 Published by Elsevier B.V.

The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability.

Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown. (C) 2014 L.M. Giancarli. Published by Elsevier B.V. All rights reserved.

Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (#16, #18 and #02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (10) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed. (C) 2012 ITER Organization. Published by Elsevier B.V. All rights reserved.

The European Joint Undertaking for ITER and the Development of Fusion Energy ('Fusion for Energy'F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European Test Blanket Systems (TBS) in ITER. An overview of the ITER TBS program has been presented recently at ISFNT-10. Currently two EU TBS designs are in the phase of conceptual design Helium-Cooled Lithium-Lead (HCLL) and Helium-Cooled Pebble-Bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the Conceptual Design Review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years in the fields of update of the TBS safety demonstration file; TBS Safety approach, design principles, requirements, features and safety functions; detailed TBS components classifications; Radiation Shielding and Protection; and Selection of reference accidents scenarios and Accidents analyses. Finally the authors share the planned future EU TBS safety activities.

This paper gives the rationale for creating through International Standard Organization a standard for confinement systems in Tritium facilities under the frame of the working group dealing with confinement, ventilation and shielding issues in the radiation protection subcommittee of the nuclear energy technical committee (ISO TC85/SC2/WG23).

Up to six mock-ups of different tritium breeding blanket systems, referred to as Test Blanket Systems (TBS), will be tested in three equatorial ports of ITER. The paper describes the recent studies performed by the IO for the preparation of the most urgent interfaces required for the integration of the TBSs into the ITER device and Tokamak complex.Main addressed items concern the impact of the TBM ferromagnetic structural material on plasma performance, the preliminary design of the TBM frame, the main dimensions and location for the TBM cooling systems and for the Tritium circuits, and preliminary requirement for measurement systems.A large effort has also been devoted to the proposed maintenance/refurbishment procedure and to the space and equipment requirements in the hot cell facility, including provisions for the shipping of TBM sub-components outside the ITER site for post-irradiation examinations. (C) 2010 Published by Elsevier B.V. All rights reserved.

This paper presents an overview of the safety related operating feedback taken into account in the ITER baseline design, and of the previously completed and ongoing Research and Development (R&D) activities in support of ITER safety analyses. Operating feedback relevant to ITER mostly comes from previous and currently existing fusion devices, and from the operation of tritium laboratories. Regarding the safety related R&D, since the early times of the ITER project, an extensive program has been devoted to understanding the issues, gathering data on source terms, modeling underlying phenomena, and developing analytical tools for safety analysis.

On the 31 of January 2008, the General Director of ITER Organization sent the request for Authorization of Creation of the ITER Basic Nuclear Installation (INB), “Demande d’Autorisation de Création, DAC”. This paper presents the licensing process for ITER in application of “loi n 2006-686 relative aux installations nucléaires de base et au contrôle, en matière de sûreté nucléaire, du transport de substances radioactives”, the so-called “Transparency Law, TSN law” and the technical content of the DAC files with special emphasis on the issues that new laws and new orders published in France since the ITER site selection have introduced in the ITER design and in the safety analysis. This paper pays special attention to regulation application and licensing process which is new for the fusion community. Examples of practical implications for ITER as a nuclear facility such as operational domain and safety important components are given for the plasma current, the plasma facing materials and the test blanket modules.

Code validation activities have been promoted inside the European fusion development agreement (EFDA) to test the capability of codes in simulating accident phenomena in fusion facilities and, specifically, in the International thermonuclear experimental reactor (ITER). This work includes a comparison between three different computer codes (CONSEN, MAGS and MELCOR) and one analytical model (ITER Model) in simulating cryogenic helium releases into the vacuum vessel (W) which contains hot structures. The scope was the evaluation of the transient pressure inside the W. The results will be used to design a vent duct (equivalent diameter, length and roughness) to allow pressure relief for the protection of the VV, which has a maximum design pressure of 200 kPa. The model geometry is a simplified scheme preserving the main features of the ITER design. Based on the results of the simulations, a matrix of experiments was developed to validate the calculated results and to design the vent duct for the ITER W. The experiments are planned to be performed in the EVITA test facility, located in the CEA Cadarache research centre (France).

The Generic Site Safety Report (GSSR) was part of the ITER Final Design Report issued in July 2001. Changes in implementation of confinement and other design changes from the 2001 design necessitate an update of safety analyses for regulatory submissions. Discussions in preparation for site selection and initial informal regulatory discussions suggested that a more transparent approach was needed for the selection of events to be presented. In addition, a minimum set of such events necessary to demonstrate the safety case should be presented. The development and application of an approach to event selection to these requirements led to the selection of 10 events needing updating. Preliminary results indicate that the updated safety analysis for ITER licensing will continue to show an acceptable (and improving) level of safety.

Knowledge of the levels of tritium in the First Wall (FW) coolant and components of ITER is important for public and operator safety and waste management. To overcome the large uncertainty of plasma wall interaction and physical properties, a basic set of properties is theoretically calculated for the dissolved tritium atoms in a perfect Beryllium (Be) lattice. These properties are combined with models for tritium trapping by lattice imperfections including the equilibrium conditions between gaseous, dissolved and trapped hydrogen isotopes. The 3 models for trapping by impurities, radiation damage and surface defects are adjusted to experimental solubilities, to tritium release experiments from irradiated samples and to outgassing of hydrogen isotopes from the JET FW. An elastic lattice model evaluates the activation energy of diffusion. For the calculations, the code DIET (Diffusion, Implantation and Equilibrium Trapping) was developed, which includes tritium trapping with time-dependent trap concentrations of multiple trap sites. The sensitivity analysis, with the expected deviations from the basic properties provides confidence that tritium permeation is below one gram in ITER for a neutron load of 0.3MWa/m(2) within 10 years.

This paper summarises the results of the ITER safety assessments during the course of the Engineering Design Activities (EDA). The key aspects of ITER safety are effluents and emissions from normal operation, including planned maintenance activities; occupational safety for workers at the facility; radioactive materials and wastes generated during operation and from decommissioning; and potential incidents and accidents and the resulting transients. The implementation of a generic safety approach, the safety aspects of the design, and the assessments of effluents, occupational safety, waste, and accidents are documented in the Generic Site Safety Report (GSSR). The analyses and assessments completed in collaboration with the Home Team experts and documented in the GSSR offer a well-developed technical basis for regulatory applications in potential host countries. This paper reviews the accomplishments in ITER safety during the course of the EDA. This work needed to integrate detailed analyses by a geographically dispersed safety team consisting of Joint Central Team (JCT) and Home Team experts who initially had different approaches and methods based on conceptual fusion reactor studies and fission power plant practices. Since an ITER site was not yet selected, a safety approach was developed for a generic site in such a way that compatibility with the Parties' regulatory frameworks can be expected. The work on safety has contributed to a design providing the technical basis for regulatory approval with the expectation that only minor changes would be needed to meet the host country's regulations. After siting, safety design and implementation will be finalised in accordance with host country regulations and practices.

Abstract A one-dimensional sublimation model for cometary nuclei is used to derive size limits for the nuclei of sungrazing comets and to estimate oxygen ion fluxes at 1 AU from their evaporation. Given that none of the ≈300 sungrazers detected by the SOlar and Heliospheric Observatory (SOHO) was observed after disappearing behind the sun, and that small nuclei with a radius of ≈3.5 m could be observed, it is assumed that all SOHO sungrazers were completely destroyed. For the case that sublimation alone is sufficient for destruction, the model yields an upper size limit as a function of nuclear density ϱ, albedo A , and perihelion distance q. If the density of the nuclei is that typical of porous ice (600 kg m −3 ), the maximum size is 63 m. These results confirm similar model calculations by Weissman (1983). An analytical expression is derived that approximates the model results well. We discuss possible modifications of our results by different disruption mechanisms. While disruption by thermal stress does not change the upper size limits significantly, they may be somewhat increased by tidal disruption (up to 100 m for a density of 600 kg m −3 ), dependent on the isotropy of the sublimation process and the tensile strength of the comet. Implications for the Kreutz family of sungrazers are discussed. Oxygen ions from the sublimation of sungrazing comets form a tail. Fluxes from this tail are sufficiently high to be measured at 1 AU by particle detectors on spacecraft, but the duration of a tail crossing is only about half an hour. Therefore, the probability of a spacecraft actually encountering a tail of an evaporating sungrazer is only of the order of 2% per year.

Safety has been an integral part of the design process for the ITER project since the conceptual design activities of the project. The safety approach implemented in the current ITER-FEAT design and the complementary assessments to be documented in the Generic Site Safety Report (GSSR) can provide valuable lessons for future fusion facilities. This paper summarises the ITER-FEAT safety approach and the assessments underway with a view to identifying key issues that have arisen and may have implications for such facilities. This paper discusses how these issues arose and are being addressed in the ITER project.

Hydrogen–air mixtures in the vacuum vessel of international thermonuclear experimental reactors (ITER) may be produced by accidents. The basic safety approach is to limit the amount of hydrogen to keep concentrations below flammability limit. For ITER, 5-kg hydrogen uniformly mixed in the vacuum vessel with air at 1.0 bar would lead to concentrations below the flammability limit of 4 vol.%. However, in the case of air ingress into a hydrogen-containing vessel, the hydrogen concentration unavoidably passes through the flammability and detonability range. Also the local formation of burnable clouds is of concern. The problem is to determine the amount of hydrogen allowed to deflagrate or detonate in the vessel without impairing the radioactive confinement function of the vessel and its secondary enclosures. First, the vacuum vessel has to withstand the static load of a deflagration, which pressurizes the vessel for a substantial time period. The final adiabatic pressure after detonation is practically equal to the deflagration pressure because there is only a small difference in burn completeness. Second, in case of a detonation, the vessel has to withstand the dynamic load of the high peak pressure of very short duration during reflection of the detonation wave at the vessel wall. To determine the static load, the maximum deflagration pressures during air ingress transients are evaluated for completely mixed gases and local clouds of flammable mixtures. To characterize the dynamic load, a model for the momentum transferred to the wall by a reflected detonation wave is developed. It results in the simple rule, that the momentum of the wave is at maximum 13% of the total theoretical momentum defined by the total mass of gas in the vessel moving with the Chapman–Jouguet (C–J) detonation front velocity [A.H. Shapiro, The Dynamics and Thermodynamics of Compressible Fluid Flow, vol. 1, Ronald Press, 1953, p. 210]. The momentum transferred to the wall is about twice the momentum of the impacting wave due to reflection on the wall. The results for the ITER limit of 5 kg show a maximum deflagration pressure below the 2 bar design limit of the vessel and a detonation momentum, which is within comparable design loads of the vessel and its support structure, such as for the dropping of a blanket module from top of the vessel to the bottom.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points. to withstand the high thermal loads expected during off-normal events, e.g.. disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., similar to 350g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER’s radiological hazards taking into account fusion’s favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway.