Comex Nucleaire

Saint-Paul-lès-Romans, France

Comex Nucleaire

Saint-Paul-lès-Romans, France
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Giancarli L.M.,ITER Organization | Cortes P.,ITER Organization | Iseli M.,ITER Organization | Lepetit L.,ITER Organization | And 7 more authors.
Fusion Engineering and Design | Year: 2014

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. © 2014 L.M. Giancarli.

Giancarli L.M.,ITER Organization | Barabash V.,ITER Organization | Campbell D.J.,ITER Organization | Chiocchio S.,ITER Organization | And 23 more authors.
Fusion Engineering and Design | Year: 2015

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. © 2015 ITER Organization.

Abonneau E.,CEA | Le Coz P.,CEA | Settimo D.,EDF | Hamy J.-M.,AREVA | And 8 more authors.
International Congress on Advances in Nuclear Power Plants, ICAPP 2014 | Year: 2014

The Preconceptual Design phase (AVP1) of the ASTRID project started in 2010 and ended in late 2012. During this phase, the aim of the ASTRID project is to integrate innovative options to meet the objectives of the 4th generation reactors and comply with the related specifications. Therefore, ASTRID project is based on a significant research and development program started in 2006 with EDF and AREVA to supply the necessary elements for the qualification of the ASTRID options in due time. Option Selection Processes (RCOs) were performed during the AVP1 phase and were structuring stages to compare several options studied. RCOs were mainly scheduled for 2012 and the last ones during the first part of 2013. RCOs make it possible to determine, among the technical solutions left open at the time when the AVP1 was launched, the reference technical option for the conceptual phase studies (AVP2, from 2013 till the end of 2015) and, if applicable, to keep a backup alternative or an advanced solution. Main results of choices made in 2013 are presented in this paper. Cost reduction phase was mastered in order to reduce the overall project cost and establish a new configuration of the ASTRID plant. In the meantime, the Safety Orientation Report (DOrS) was issued and discussed with the Nuclear Safety authority (ASN). These two points are also developed in this paper.

Jadot F.,CEA Cadarache Center | Laffont G.,CEA Cadarache Center | Le Coz P.,CEA Cadarache Center | Sibilo J.,AREVA | And 2 more authors.
International Congress on Advances in Nuclear Power Plants, ICAPP 2013: Nuclear Power - A Safe and Sustainable Choice for Green Future, Held with the 28th KAIF/KNS Annual Conference | Year: 2013

On the basis of available feedback and the new safety requirements, the ISI&R for instrumentation, in service inspection and repair, has been identified as a major issue for next Sodium cooled Fast Reactor prototype ASTRID. Indeed, ISI&R is strongly involved in safety analysis, in economy reliability, and in investment protection. The safety objective for the ASTRID project is threefold: a) a level of safety that it is at least equivalent to the 3rd generation reactors, b) integration of the defence-in-depth principle and c) a better safety demonstration than previous SFR. For the instrumentation, the objectives are mainly the improvement of the in-operation continuous monitoring and the protection against accidents, for the whole plant. That means: better diversification and redundancy to take into account events which were not considered in the past?up to date technologies specific post-accidental instrumentation For ISI&R, the main axes remain: improvement of the primary system conceptual design, development of measurement and inspection techniques accessibility and associated robotics, development and validation of repair processes. The ASTRID strategy is supported by the French R&D program for ISI&R improvement and industrial developments in the frame of specific ASTRID agreements based on several aspects: ensuring a strong connection between the reactor designers and inspection specialists, and developing tools and techniques which must be operated in a sodium environment (monitoring up to 550°C and inspection at 200°C). The aims of an ambitious R&D program are: ?Validation for high temperature fission chambers manufacturing used in order to detect several incidents and monitor the reactor operation, Validation of ultrasonic transducers (US telemetry and NDE sensors) under sodium conditions, Definition of key components of the robotic equipment for operation in sodium, Preliminary validation of repair processes and techniques (cleaning, machining and welding). Associated ISI&R needs are being defined through an iterative method between designers and instrumentation specialists : adaptation of the Design to ISI&R requirements, validation of the ultrasonic and chemical transducers, of ultrasonic non destructive simulation, of acoustic surveillance, of laser repair intervention processes, of connected robotic equipment. This paper summarizes the main orientations of the project and gives a few examples of the work carried out by CEA and its industrial partners.

Pascal R.,ITER Organization | Cortes P.,ITER Organization | Friconneau J.-P.,ITER Organization | Giancarli L.M.,ITER Organization | And 10 more authors.
Fusion Engineering and Design | Year: 2013

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. © 2013 Elsevier B.V.

Le Coz P.,French Atomic Energy Commission | Sauvage J-F.,Électricité de France | Hamy J.-M.,AREVA | Jourdain V.,Alstom | And 6 more authors.
International Congress on Advances in Nuclear Power Plants, ICAPP 2013: Nuclear Power - A Safe and Sustainable Choice for Green Future, Held with the 28th KAIF/KNS Annual Conference | Year: 2013

The pre-conceptual design of the ASTRID project has been launched in 2010 by CEA. The objectives of this first phase are to consider innovative options to improve the safety level with progress made in SFR-specific fields. A few examples of these innovations are: a core with an overall negative sodium void effect, specific features to prevent and mitigate severe accidents, power conversion system decreasing drastically the sodium-water reaction risk, improvements in In-Service Inspection and Repair, etc. ASTRID will also be designed to pursue the R&D on sodium fast reactors and demonstrate the feasibility of transmutation of minor actinides. CEA has concluded partnerships with industrial partners (EDF, AREVA NP, ALSTOM, BOUYGUES, COMEX NUCLEAIRE, TOSHIBA, JACOBS, ROLLS-ROYCE and ASTRIUM) and the total staff involved in the project is now about 500 people. The paper describes the organization and the current status of the project, the mains results obtained during the pre-conceptual design and the objective of the conceptual design with the associated milestones.

Van der Laan J.G.,ITER Organization | Cuquel B.,Airbus | Demange D.,Karlsruhe Institute of Technology | Ghidersa B.-E.,Karlsruhe Institute of Technology | And 6 more authors.
Fusion Engineering and Design | Year: 2015

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. © 2015.

Kim B.-Y.,ITER Organization | Marconi M.,LTCalcoli | Neviere J.-C.,Comex Nucleaire | Merola M.,ITER Organization | And 3 more authors.
Fusion Engineering and Design | Year: 2015

In three of the ITER equatorial ports, tritium breeding blanket concepts will be validated and tested using mock-up breeding blankets called test blanket modules (TBM). In these ports, two TBM-Sets are mechanically attached in a TBM Frame to form a TBM port plug (TBM PP). The ITER Organization is responsible for the design and manufacture of both this TBM Frame and the Dummy TBMs which will fill them. As a part of this development, in 2013, a conceptual design review (CDR) of TBM PP with two dummy TBMs revealed the need for improvement of design performance, interfaces and maintainability. This paper presents the main design improvements after the CDR as well as associated feasibility analysis of the improved design focusing on the attachment. Finally the work plan for the Preliminary Design phase is summarized. © 2015 Elsevier B.V.

Shuff R.,United Technologies | Locke D.,Fusion for Energy F4E | Roulet D.,COMEX Nucleaire
Fusion Engineering and Design | Year: 2011

This paper reports on the use of Discrete Event Simulation tools for operability analysis of the Iter Hot Cell (HC) design. A simulation model representing the operation of the ITER Hot Cell has been created. The model incorporates the process logic for ITER components that are required to be refurbished, maintained and disposed of within the Hot Cell. This paper presents some results of the simulation indicating the performance of the Hot Cell for Divertor and Port Plug refurbishment. The results show that based on the established task durations a full 54 cassette Divertor refurbishment takes 4536 h, longer than the planned 6-month shutdown scheduled for the task. The simulation provides a platform to accurately size the capacity of process equipment with respect to given budgetary constraints and to identify opportunities to smooth the process flow. Effects of parameters such as human resource shift patterns, equipment mean time between failure and random variability in process times on overall Hot Cell productivity have been studied. The simulation model is flexible, capable of evolving in parallel with the Hot Cell design as more detailed input data becomes available thereby providing a valuable decision making and design optimisation tool throughout the development of the Hot Cell and beyond that into operation. © 2011 Published by Elsevier B.V.

Comex Nucleaire and French Atomic Energy Commission | Date: 2016-04-14

The present invention relates to a device (1) for handling an absorbent rod (11) for controlling a nuclear reactor, comprising (a) an upper motor compartment (2) positioned on the closing slab (10) of the reactor vat (13), (b) a rod control stem (3), extending in said motor compartment (2) and in a guide sheath (3a) extending inside said vat, characterized in that it comprises (1) a sealed static confinement chamber (5) made of non-magnetic material arranged inside said upper compartment, (2) a first synchronous magnetic coupling system (6a) for transmitting linear translational movement without mechanical contact comprising a first outer component (6a-1) arranged outside said chamber (5), and able to be vertically translated, and a first inner component (6a-2) arranged inside said chamber (5), integral with said rod control stem (3), a magnetic coupling force of said first outer component (6a-1) and said first inner component (6a-2) making it possible, when the first outer component (6a-1) is displaced in vertical translation, for said first inner component (6a-2) and the rod control stem (3) follow a displacement in vertical translation, and (3) said motor compartment (2) comprising first motorized mechanical means for transmitting translational displacements (2a) of said first outer component (6a-1) of said first magnetic coupling system (6a).

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