T. Pinna

The accomplishment of the Engineering Design Activities of IFMIF/EVEDA: The European-Japanese project towards a Li(d,xn) fusion relevant neutron source

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European–Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.

Failure mode and effect analysis on ITER heat transfer systems

The complexity of the ITER (International Thermonuclear Experimental Reactor) plant and the inventories of radioactive materials involved in its operation require a systematic approach to perform detailed safety analyses during the various stages of the project in order to demonstrate compliance with the safety requirements. The failure mode and effect analysis (FMEA) methodology has been chosen to perform the safety analysis at system level for ITER. The main purposes of the work are: to identify important accident initiators, to find out the possible consequences for the plant deriving from component failures, identify individual possible causes, identify mitigating features and systems, classify accident initiators in postulated initiating events (PIEs), define the deterministic analyses which allow the possible accident sequences to be quantified, both in terms of expected frequency and radiological consequences, and consequently, to ascertain the fulfillment of ITER safety requirements. This paper summarises the FMEA performed for the heat transfer systems (HTSs).

Validation and benchmarking in support of ITER-FEAT safety analysis

This paper briefly describes the codes used for International Thermonuclear Experimental Reactor (ITER) safety analysis, including some information on their validation status, and summarizes some examples of validation and verification (V&V). V&V information is provided for the codes involved in accident analyses dealing with water coolant ingress into the vacuum vessel. Results obtained by the MELCOR and INTRA codes, and the ISAS system, are compared with the test results of the integrated ICE facility simulating water coolant ingress into the vacuum vessel. Benchmark calculation results of the MELCOR, INTRA, and ISAS were compared with the results from other codes like PAX, CONSEN, TRAC-BF1, CATHARE.

Modelling of two-phase flow under accidental conditions fusion codes benchmark

The scope of this benchmark is to assess the capabilities of the best estimate thermal hydraulic codes to simulate the main physical phenomena occurring during an in-vessel break transient within a water-cooled fusion-type reactor: pressurisation of a volume at low initial pressure, critical flow, counter pressure effect, relief into an expansion volume. The results, which are given by the code are of the same order of magnitude. Discrepancies are observed which are mainly due to the different ways the codes simulate the break mass flow rate and the pressurisation of the vacuum vessel.

Collection and analysis of component failure data from jet systems: neutral beam injectors and power supply

The objective of this ongoing activity is to develop a fusion specific component failure database with data coming from operating experiences gained in various fusion laboratories. The activity began in 2001 with the study of the Joint European Torus (JET) Vacuum and Active Gas Handling Systems. Two years later the neutral beam injectors (NBI) and the power supply (PS) systems were considered and since last year the ion cyclotron resonant heating system is under evaluation. The number of failures/malfunctions that have occurred during the years of operations, failure modes and, where possible, causes and consequences of the failures were identified, as well as the sets of components under analysis. Components were classified and counted in order to determine component counts by type of component, related total operating hours and related demands to operate (for components operating in an intermittent manner). Main reliability parameters (such as the failure rate and corresponding standard errors and confidence intervals) for the component types were also estimated. In this paper the NBI and PS results are presented.

Materials-related issues in the safety and licensing of nuclear fusion facilities

Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

Safety assessment for ITER-FEAT tritium systems

Abstract The design of the equipment and confinement barriers of ITER-FEAT should be consistent with the basic safety requirement that no emergency plan involving evacuation of the nearby population is required in case of the worst credible accident. Extensive probabilistic and deterministic analyses have been done to select abnormal event sequences, and to ensure that all potential consequences are within project guidelines. The paper deals with the work done for the tritium systems. A Bottom–Up methodology based on component level Failure Mode and Effect Analysis has been applied to point out accident initiators. Once possible accident sequences have been identified, detailed deterministic analyses on bounding events confirmed that the accidents in tritium plant are not a concern from a safety point of view. The no-evacuation goal of ITER-FEAT is attained also for accidents where ‘ultimate safety margin’ are challenged, as in case of hydrogen-air reactions in cold box or in hard shell, enclosing the process equipment of the isotope separation system.

European contribution to the ITER licensing

Abstract The DAC file (Demande d’Autorisation de Création) is the principal document supporting the application for the licensing of ITER. It includes the Preliminary Safety Report (RPrS - Rapport Préliminaire de Sûreté) and the “Impact Study”. On January 2008, the DAC was officially submitted to the French Nuclear Authority (ASN). To cope with the requests and recommendations given by the ASN to the earlier ITER Safety Options Report (DOS), CEA had taken commitments dealing with complementary information to be integrated into the RPrS. The necessary work had been implemented by EFDA (European Fusion Development Agreement) and, since its existence, by F4E (Fusion for Energy), in the EISS activities (European ITER Site Study) and in the European Safety Technology Work Programs. The executants of the work have been CEA-AIF (Commissariat à l’Énergie Atomique – Agence ITER France), several European Associations (CEA, CIEMAT, ENEA, FZK and VR/Studsvik) and industry. All of them have been working in full cooperation with ITER Organization (IO). In addition some long term R&D tasks, which will have to be performed in parallel to ITER construction, have been defined and their implementation started. Typical examples are dust management (production, mobilization, diagnostic and removal), combined hydrogen/dust explosion models development and validation, demonstration of the feasibility of prevention/mitigation of in-vessel hydrogen/dust explosions and studies on magnet arcing behaviour and consequences. The final writing of the DAC and the related studies has involved the equivalent to 35 man-years of effort. Most of the resources have been focused on the fulfillment of the supporting files for the RPrS: lessons learnt from fusion experiments, R&D status of the fusion technology, safety operational limits, definition of the safety control system, development of a maintenance program, analysis of occupational radiation exposures, incorporation of human factors into the design, categorization of incidents and accidents, definition of design and safety codes and standards, criteria for design reviews, configuration control, waste management dismantling and demolition, safety analysis of internal and external hazards and security concerns. The remaining licensing effort has been dedicated to the environmental impact of the project and the coordination, preparation and presentation of documentation to the Safety Authorities. This paper summarizes the main outcomes of the European contribution to the ITER licensing process and the related ongoing and planned supporting R&D activities.

Power Supply Reliability Estimates for Experimental Fusion Facilities

Abstract This paper presents the results of a task to analyze the operating experience data for large, pulsed power supplies used at the DIII-D tokamak. This activity supports the International Thermonuclear Experimental Reactor (ITER) project by giving fusion-specific reliability values for large power supplies that energize neutral beams and magnets. These failure rate data are necessary to perform system availability calculations and to make estimates of the frequency of safety-significant events (e.g., power supply arcs or fires) that might occur in other fusion facilities such as ITER. The analysis shows that the DIII-D data results compare well with the results of similar data analysis work that the Italian National Agency for New Technologies, Energy and the Environment (ENEA) has performed on the JET tokamak and compare fairly with data from two accelerators.

Summary of the 1st International Workshop on Environmental, Safety and Economic Aspects of Fusion Power

The 1st International workshop on Environmental, Safety and Economic Aspects of Fusion Power (ESEFP) was held on 13 September 2015 at Jeju Island, South Korea. The workshop was initiated by the International Energy Agency Implementing Agreement on a Co-operative Program on ESEFP. The workshop was well attended with about forty participants representing twelve institutions in ten countries. The presentations covered safety issues and environmental impacts, availability improvement and risk control and socio-economic aspects of fusion power. Safety and licensing gaps between DEMO and ITER were discussed in depth with the consensus output presented as a plenary presentation at the 12th International Symposium on Fusion Nuclear Technology (ISFNT-12). The next workshop is planned to be held in conjunction with the ISFNT-13 in 2017.