M.T. Porfiri

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.

STARDUST experimental campaign and numerical simulations: influence of obstacles and temperature on dust resuspension in a vacuum vessel under LOVA

Activated dust mobilization during a Loss of Vacuum Accident (LOVA) is one of the safety concerns for the International Thermonuclear Experimental Reactor (ITER). Intense thermal loads in fusion devices occur during plasma disruptions, edge localized modes and vertical displacement events. They will result in macroscopic erosion of the plasma facing materials and consequent accumulation of activated dust into the ITER vacuum vessel (VV). These kinds of events can cause dust leakage outside the VV that represents a high radiological risk for the workers and the population. A small facility, Small Tank for Aerosol Removal and Dust (STARDUST), was set up at the ENEA Frascati laboratories to perform experiments concerning the dust mobilization in a volume with the initial conditions similar to those existing in ITER VV. The aim of this work was to reproduce a low pressurization rate (300?Pa?s?1) LOVA event in a VV due to a small air leakage for two different positions of the leak, at the equatorial port level and at the divertor port level, in order to evaluate the influence of obstacles and walls temperature on dust resuspension during both maintenance (MC) and accident conditions (AC) (Twalls = 25??C MC, 110??C AC). The dusts used were tungsten (W), stainless steel 316 (SS316) and carbon (C), similar to those produced inside the vacuum chamber in a fusion reactor when the plasma facing materials vaporize due to the high energy deposition. The experimental campaign has been carried out by introducing inside STARDUST facility an obstacle to simulate the presence of objects, such as divertor. In the obstacle a slit was cut to simulate the limiter?divertor gap inside ITER VV. In this paper experimental campaign results are shown in order to investigate how the divertor and limiter?divertor gap influence dust mobilization into a VV. A two-dimensional (2D) modelling of STARDUST was made using the CFD commercial code FLUENT, in order to get a preliminary overview of the fluid dynamics behaviour during a LOVA event and to justify the mobilization data. In addition, a numerical model was developed to compare numerical results with experimental ones.

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.

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.

Simulation of cryogenic He spills as basis for planning of experimental campaign in the EVITA facility

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 (VV) which contains hot structures. The scope was the evaluation of the transient pressure inside the VV. 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 VV. The experiments are planned to be performed in the EVITA test facility, located in the CEA Cadarache research centre (France).

Ex-Vessel Break in ITER Divertor Cooling Loop Analysis with the ECART Code

A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal � /hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal � /hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed. # 2003 Elsevier Science B.V. All rights reserved.

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.

Fusion safety codes: international modeling with MELCOR and ATHENA-INTRA

Abstract For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA–INTRA codes and presents their modeling results for the following breaches of a water cooling line into the tokamak vacuum vessel, (1) water injection in vacuum with condensation of the flashed steam and (2) water injection in vacuum. The modeling predictions are compared with experimental results from the ingress-of-coolant event (ICE) facility in Japan. It is observed that the codes’ predictions exhibit good agreement with the experiment data, which suggests that the codes include the proper physical mechanisms associated with the chosen accident scenarios. It is thereby concluded that either of the codes can be used to model the defined transient.