Sergio Ciattaglia

Preliminary Safety Analysis of ITER

Abstract In order to support the licensing application for the ITER facility at Cadarache, a preliminary safety case has been prepared and submitted to the French nuclear safety authorities. This paper provides an overview of technical aspects of this case, which is based on an evolution of the safety approach developed and applied in earlier phases of the ITER project. The basis of the safety of ITER derives from the fundamental safety characteristics of fusion. The potential radiological hazards that arise are related to the tritium fuel and material activated by neutrons. The confinement of these materials is therefore the principal safety function, and it is reliably provided by robust barriers inherent in the design together with filtering and detritiation as a secondary level of confinement provision. A Defense in Depth approach is taken to ensure that off-normal events are minimized in their frequency, and that the consequences of accidents, even though extremely unlikely, are limited. A comprehensive set of analyses of postulated event sequences provides the demonstration that the consequences of enveloping scenarios are well within acceptable limits, and that even for hypothetical events involving two or more independent failures, the public and environmental impacts remain limited. An ALARA approach is taken to minimizing occupational radiation exposure, as well as other potential impacts of normal operation such as routine releases. Other hazards arising from internal and external risks are also considered, with design provisions, for example the Tokamak building is built on seismic isolation pads to minimise the effect of an earthquake.

The European DEMO fusion reactor: Design status and challenges from balance of plant point of view

DEMO initial conceptual design studies are being conducted in Europe as part of the European Union Roadmap to Fusion Electricity, which aims to demonstrate the feasibility of electricity produced by nuclear fusion reactors around the middle of this century. The aim of this paper is to provide an overview of the DEMO project, highlighting its main characteristics and challenges in terms of design, integration, and operation. Particular emphasis is given on some important systems of the Balance of Plant (BoP), such as the primary heat transfer systems, the related power conversion systems, and the electrical power plant. The relevance of such systems is due to the need of a continuous reanalysis at any significant design change because of their huge dimensions, technical complexity, and strong impact on design integration, maintenance, and safety.

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.

Main safety issues at the transition from ITER to fusion power plants

In parallel to the ITER design process and in close cooperation with the designers a fusion-specific safety approach was developed and implemented. Detailed safety assessments have been performed and documented in the ITER Generic Site Safety Report (GSSR). Following the decision on ITER construction in France, results from the GSSR and from on-going safety-related activities tailored to the Cadarache site and the French licensing process are now being used to write the ITER Preliminary Safety Analysis Report.In the most recent European fusion power plant conceptual study (PPCS) inherent fusion favourable features have been exploited, by appropriate design and choice of materials, to provide major safety and environmental advantages. The study focused on five power plant models, which are illustrative of a wider spectrum of possibilities. These span a range from relatively near-term concepts, based on limited technology and plasma physics extrapolations, to a more advanced conception. All five PPCS plant models differ substantially in their plasma physics, blanket and divertor technology, size, fusion power and materials compositions, and these differences lead to differences in economic performance and in the details of safety and environmental impacts.This paper uses the quite detailed information available from ITER safety documents and highlights the differences between ITER and future fusion power plants. The main areas investigated are releases and doses during normal operation and under accidental conditions, occupational radiation exposure and optimization and waste management, including recycling and/or final disposal in repositories.Due to an error, an incorrect version of this paper was published in issue 7. For the convenience of the reader we have included the correct full article below rather than a list of changes.

Benchmark of the GETTHEM Vacuum Vessel Pressure Suppression System (VVPSS) model for a helium-cooled EU DEMO blanket

In the nuclear field, the correct evaluation of the effects of design-basis accidents is fundamental to correctly design the countermeasures needed to preserve the integrity of the containment barriers and to confine the ra-dioactive material. Therefore, both in fission and in fusion, notwithstanding the different amounts of radioac-tive materials, the availability of models that can predict the accidental transients is crucial. Here we describe the model recently developed to analyse an in-vessel Loss-Of-Coolant-Accident in the EU DEMO fusion reactor, and implemented in the GETTHEM code. In particular, we focus on the release of coolant inside the Vacuum Vessel (VV) following a break in the breeding blanket cooling loop, considering a helium-cooled blanket solution. The model of the VV pressure suppression system is calibrated and bench-marked exploiting results from the validated CONSEN code by ENEA.

R&D on in-vessel dust and tritium management in ITER

In a tokamak, plasma-wall interactions can result in production of dust. During operation, the tritium present in the Vacuum Vessel (VV) can then be trapped in the in-vessel materials but also in dust. The vacuum vessel represents the first confinement barrier to this radioactive material. In the event of a postulated accident involving ingress of steam into the VV, hydrogen could in principle be produced by chemical reaction with hot metal and dust. If the ingress of air into the VV is also postulated, reaction of air with hydrogen and/or dust cannot be completely excluded and could lead to a possible explosion which could challenge the VV tightness. In order to prevent such accidents and their radiological consequences, limitations on the accumulation of dust and tritium in the VV and on the air ingress are imposed. Correlatively, ITER has defined a strategy for the control of in-vessel dust and tritium inventories based on both measurement and removal techniques. In this context, this paper reports on the status of tasks under F4E responsibility aiming at developing some of the measurement systems and necessary R&D for the validation of the ITER strategy.

The Role of Operational Feedback and R&D in ITER Safety

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.

Safety related R&D for the ITER baseline design

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.