W. Gulden

Thermohydraulic Analysis of Accident Scenarios of a Fusion DEMO Reactor Based on Water-Cooled Ceramic Breeder Blanket: Analysis of LOCAs and LOVA

Thermohydraulic analysis of postulated accidents will identify system responses to accident scenarios and aid in developing design of safety systems and strategies to prevent or mitigate accident propagation. This paper reports analyses of four accident scenarios of a fusion DEMO reactor based on water-cooled ceramic-pebble breeder blanket. The accidents analyzed, which were selected based on the previous logical accident analysis, are ex-vessel loss of coolant of the primary cooling system, in-vessel loss of coolant of the first wall (FW) cooling pipes, loss of coolant in blanket modules, and loss of vacuum. The analyses have identified thermohydraulic responses of the DEMO systems to these accidents, pressure loads to confinement barriers for radioactive materials. Effectiveness of the safety systems and the integrity of the primary and final (secondary) confinement barriers against the accidents are discussed. As for the final confinement barrier, we show for the first time that implementation of a pressure suppression system (PSS) to the cooling system vault and a vacuum breaker to the tokamak pit is effective in significantly keeping the integrity of the final confinement barrier against the ex-vessel loss-of-coolant and loss-of-vacuum accidents, respectively. As for the primary confinement barrier, we show for the first time that limitation of the number of blankets from which a helium purge gas line collects the bred tritium will be a key technical issue to prevent propagation of loss of coolant in a blanket box through the purge gas line and then suppress the pressurization of the vacuum vessel (VV). For the in-vessel loss of coolant of the FW cooling pipes, further optimization of the PSS or design solutions regarding in-vessel components and plasma control will be necessary to decrease the pressure load to the VV and ensure the integrity of the primary confinement barrier.

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

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.

Progress in thermohydraulic analysis of accident scenarios of a water-cooled fusion DEMO reactor

Thermohydraulic analysis of postulated accidents will identify systems responses to accident scenarios and aid to develop design of safety systems and strategies to prevent and mitigate accident propagation. This paper reports recent progress in analysis of some accident scenarios of a water-cooled fusion DEMO reactor. Four types of accidents are particularly considered: ex-vessel loss-of-coolant of the primary cooling system, in-vessel loss-of-coolant of the first wall cooling pipes, loss-of-coolant in blanket modules and loss-of-vacuum. The analysis has identified responses of the DEMO systems to these accidents, pressure loads to confinement barriers for radioactive materials and radiological consequences. Effectiveness of the safety systems and integrity of the confinement barriers against the accidents are discussed.

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.