Proposal of an improved design of IFMIF Test Cell components for enhanced handling and reliability

An adequate design of components to be manipulated by remote handling is a key factor in the success of any activated facility, having a decisive impact on availability, prompt and safe maintenance, occupational exposures and flexibility of the facility. Such components should satisfy at least the basic remote handling requirements of simplicity, accessibility, modularity, standardization and assembling adequacy. Highly activated components in the IFMIF facility are found in the Test Cell, a pit closed by stepped shielding plugs. The Test Cell confines the Test Modules which contain the samples and experiments. The present reference design of the IFMIF Test Cell shows some drawbacks, in particular the jamming tendency of the shielding plugs, slow and complex access to the Backplate, a low lifetime component, and difficult positioning of the Test Modules. This paper summarises several modifications aiming at improving, under such remote handling requirements, the present reference design of the Test Cell shielding plugs and aspects of the geometrical structure of the Test Cell. A functional modularization of the present shielding plugs has been carried out and positioning guides for the Test Modules have been devised.     

New progresses in the IFMIF target and test cell design and a proposal for the specimen flow

The development of the conceptual design of the IFMIF target and test cell (TTC) is briefly summarized by outlining the previous reference TTC design and the current modular test cell (MTC) concept. Based on the MTC concept, the latest progresses of the preliminary engineering design of the key TTC components, including the TTC vessel, the top shielding plugs, the removable intermediate ring, and the test module interface heads, are described. A specimen flow, based on handling requirements of the high flux test module, between the TTC, access cell (AC), test module handling cell (TMHC) and the post irradiation examination facilities is proposed as well as the function of the AC and TMHC is preliminary defined. The TMHC is proposed to be divided into a component handling cell and a rig handling cell regarding the dimension differences of the components to be handled inside of the cells. Re-use of irradiated specimens for another campaign of irradiation is also considered in this specimen flow.     

Numerical simulations on natural convective heat transfer and active cooling of IFMIF Test Cell

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The Test Cell (TC) accommodates the Test Modules and the lithium target assembly. Due to the nuclear heat generation, all the Test Modules inside the TC will be actively cooled. Other components like supporting structures, pipelines, cables etc., will be passively cooled by natural convection. The heat will be removed from the steel liners surrounding the TC by active water cooling. This paper concerns the thermo-hydraulic simulations of the Test Cell using Ansys-CFX. The current simulation model includes the natural convection inside the TC, several forced convective water flows in the pipelines attached on the steel liners and the helium-cooled HFTM (High Flux Test Module). The simulations provide the only means for validating the design before the construction and operation.     

The European contribution to the development and validation activities for the design of IFMIF lithium facility

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility.The paper gives an overview of the status of the activities and of the main outcomes achieved so far.     

Overview of the main challenges for the engineering design of the test facilities system of IFMIF

High intense radiation fields were demanded to IFMIF to address the lack of information on effects in materials due to radiation fields with fusion reactor features. Such intense radiation fields will also produce a number of unwanted effects in exposed materials and components. The main difficulties to achieve a reliable engineering design of the Test Facilities System during the Engineering Validation and the Engineering Design phase of IFMIF now under development are reviewed in this paper. The most challenging activities will be the design of the high flux test module, the creep fatigue test module, the test cell and the remote handling system. The intense radiation fields in the irradiation area and the high availability required for IFMIF (70%) are the main reasons for these difficulties.     

IFMIF test cell design: Current status and key components

The IFMIF (International Fusion Material Irradiation Facility) test cell design has been further developed and optimized based on the existing modular test cell concept. Key features of the current test cell include actively cooled surrounding shielding walls with coverage of internal surfaces with stainless steel liner, independent two layer top shielding plugs for protecting the access cell from neutron and gamma radiation from the test cell, optimized piping and cabling plugs for accommodating pipe and cable penetrations and for minimizing neutron streaming, rearranged lithium quench tank to outside of the test cell, etc. According to preliminary neutronic calculation results, limited access to the quench tank area for maintenance after beam shut-off can be expected with the current arrangement. Maintenance of the lithium inlet and outlet pipes as well as the two beam ducts are also possible by introducing removable shielding plugs which can be removed and replaced in case of failure.     

Operability of the ITER Hot Cell Facility

The maintenance operation of the ITER in-vessel components are executed in the Hot Cell Facility (HCF). Therefore maintenance requirements and the operability analysis are essential inputs for the design. To fulfill the maintenance strategy, the HCF shall be capable of refurbishing and replacing vacuum vessel components and processing radwaste simultaneously. Based on the top level project requirements, the design scenario giving the most stringent constraints is based on the following tasks during the main shutdown: replacement of 54 cassettes, 3 Test Blanket Modules Port Plugs, the In Vessel Viewing System, the repair of 2 Equatorial and 2 Upper Port Plugs and the use of the Multi-Purpose Deployer. The operability between shutdowns needs also to be compliant with scheduled and unscheduled maintenance tasks. Furthermore, regular maintenance of the facility itself needs to be considered. This paper will outline the methodology, the basic tasks considered and it will present the schedule during shutdown and between shutdown periods. The goal is also to check the use rate of each workstation.     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

The remote handling compatibility analysis of the ITER generic upper port plug structure

The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final design review. The analysis focuses mainly on the insertion and extraction procedure of the diagnostic shield module (DSM), a removable cassette that contains the diagnostic in-vessel components. It is foreseen that the DSM is a replaceable component – the procedure of which is to be performed inside the ITER hot cell facility (HCF), where the GUPP can be oriented in a vertical position.The DSM removal procedure in the HCF consists of removing locking pins, an M30 sized shoulder bolt and two electrical straps through the use of a dexterous manipulator, after which the DSM is lifted out of the GUPP by an overhead crane. For optimum access to its internals, the DSM is mounted in a handling device. The insertion of a new or refurbished DSM follows the reverse procedure.The RHCA shows that the GUPP design requires a moderate amount of changes to become fully compatible with RH maintenance requirements.     

Combined application of Product Lifecycle and Software Configuration Management systems for ITER remote handling

The advantages of Product Lifecycle Management (PLM) systems are widely understood among the industry and hence a PLM system is already in use by International Thermonuclear Experimental Reactor (ITER) Organization (IO). However, with the increasing involvement of software in the development, the role of Software Configuration Management (SCM) systems have become equally important. The SCM systems can be useful to meet the higher demands on Safety Engineering (SE), Quality Assurance (QA), Validation and Verification (V&V) and Requirements Management (RM) of the developed software tools. In an experimental environment, such as ITER, the new remote handling requirements emerge frequently. This means the development of new tools or the modification of existing tools and the development of new remote handling procedures or the modification of existing remote handling procedures. PLM and SCM systems together can be of great advantage in the development and maintenance of such remote handling system. In this paper, we discuss how PLM and SCM systems can be integrated together and play their role during the development and maintenance of ITER remote handling system. We discuss the possibility to investigate such setup at DTP2 (Divertor Test Platform 2), which is the full scale mock-up facility to verify the ITER divertor remote handling and maintenance concepts.     

Conceptual design of the ECH upper launcher system for ITER

The challenge of developing the conceptual design of the ECH Upper Launcher system for MHD control in the ITER plasmas has been tackled by team of European Associations together with the European Domestic Agency (“F4E”). The launcher system has to meet the following requirements: (a) a mm-wave system extending from the interface to the transmission line up to the target absorption zone in the plasma and performing as an intelligent antenna; (b) a structural system integrating the mm-wave system and ensuring sufficient thermal and nuclear shielding; (c) port plug remote handling and testing capability ensuring high port plug system availability. The paper describes the reference launcher design. The mm-wave system is composed of waveguide and quasi-optical sections with a front steering system. An automated feedback control system is developed as a concept based on an assimilation procedure between predicted and diagnosed absorption location. The structural system consists of the blanket shield module, the port plug frame, and the internal shield for appropriate neutron shielding towards the launcher back-end. The specific advantages of a double walled structure are discussed with respect to adequate baking, to rigidity towards launcher deflection under plasma-generated loads and to removal of thermal loads, including nuclear ones. Basic studies of remote handling (RH) to validate design development are initiated using a virtual reality simulation backed by experimental validation, for which a launcher handling test facility (LHT) is set up as a full scale experimental site allowing furthermore thermohydraulic studies with ITER blanket water parameters.     

Results of an integration study of a diagnostics port plug in ITER

Diagnostics in ITER are mandatory to characterize the parameters of plasma and study its interactions with plasma-facing components. Diagnostics components in the vicinity of the plasma are supported by metallic structures called port plugs. At the tokamak mid-plane, these components are installed in port plugs through intermediate structures called drawers. Apart from hosting the diagnostics, the port plugs act as shielding against neutrons and gammas, in order to limit the nuclear loads in crucial components (such as diagnostics and superconducting coils) as well as the dose levels in the controlled zones of the tokamak. The radiation shielding function of the port plugs is ensured through an optimized mixture of heavy metallic materials and water, forming shielding blocks surrounding the diagnostics and called Diagnostic Shield Modules (DSMs). These DSMs constitute the rear part of the drawers (the front part being composed of the Diagnostic First Wall). This paper presents the main results of a study performed in Europe on the integration of a particular diagnostics port plug, the Equatorial Port Plug 1 (EPP1). The paper first provides the results of the EPP1 diagnostics integration analysis. In a second step it focuses on the design of the EPP1 DSMs and summarizes the major results of a thorough set of analyses aiming at studying the DSMs behaviour under different loads, suggesting recommendations to improve their current design.     

Start-up phase of the HELOKA-LP low pressure helium test facility for IFMIF irradiation modules

As part of the Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) [1], it is foreseen to design and test a 1:1 scale prototype of the IFMIF High Flux Test Module (HFTM) [2]. The module has been designed to be cooled by a low pressure helium gas flowing through minichannels to remove the nuclear heat.The Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been designed to provide coolant at 1:1 HFTM operational conditions: massflow 12–120 g/s, inlet pressure 0.3–0.6 MPa, inlet temperature RT – 250 °C. A secondary objective is to use the experience gained with HELOKA-LP for the planning of the IFMIF helium cooling system.The facility has been put into operation in 2009, and has since then been in a test and optimization phase. It was proven, that the above mentioned requirements for the facility are achieved. The paper describes the process layout and components of the facility. The performance is characterized by the results of several steady state and transient benchmark tests. Typical start-up and transition times relevant for the operation mode in the IFMIF irradiation campaigns are obtained. Additionally first results on the impurity ingress and the cooling gas chemistry are described.     

Technical analysis of an early fusion neutron source based on the enhancement of the IFMIF/EVEDA accelerator prototype

In the framework of the Engineering Design and Engineering Validation Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA), three major prototypes have been designed and are being manufactured, commissioned and operated which are firstly the Accelerator Prototype (LIPAc) at Rokkasho, fully representative of the IFMIF low energy (9 MeV) accelerator stage, secondly the EVEDA Lithium Test Loop (ELTL) at Oarai, and thirdly critical components of the High Flux Test Modules to be tested in the helium cooling loop (HELOKA-LP) at Karlsruhe. The present paper analyses possibilities from a technical point of view, for combining, modifying, and enhancing, at limited cost, selected components of the prototypes towards the realisation of an early reduced-flux neutron source, able nonetheless to start the testing of candidate DEMO materials and realising by this a first step towards the construction and operation of a complete IFMIF plant.Various options of deuteron beam parameters, such as energy, current and shape are analysed with respect to their technical challenges and the neutron yield resulting from the nuclear reaction with the Li target. Related requirements for the liquid Li target with respect to jet parameters are evaluated and the neutron mapping in the high flux region is presented underlying an analysis of the available volume for testing reduced activation ferritic martensitic (RAFM) steels at relevant structural damage levels.     

Global dose rate assessment in ITER diagnostics ports based on the 3-D FEM ATTILA code

The USITER, through the Princeton Plasma Physics Lab (PPPL), is responsible for the delivery of several fully integrated upper, equatorial and lower port plugs dedicated for the diagnostics in ITER. Each port plug package consists of a generic port plug structure and a set of diagnostics and diagnostic housings. The shielding design of the integrated port plugs calls for maintaining a dose level not to exceed 100 μSv/h inside the interspace of each port; the room behind the port plug where maintenance personnel access the rear of the port. This is set as an upper target design in order to perform routine maintenance 1E6 sec (～two weeks) following shutdown. Expensive remote handling robots and tooling are required otherwise. In this paper we present results from a parametric study aimed at providing initial assessment of the attainable dose rates in the diagnostics ports and their extension areas in order to properly address the duration time and frequency for the workers to perform the scheduled maintenance. The nuclear analysis is performed using both the serial version and the distributed memory parallel (DMP) version of the ATTILA-7.1.0, 3-D FEM Discrete Ordinates code, along with the FENDL2.1/FORNAX and ANSI/ANS-6.1.1-1977 data bases.     

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments.In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation.The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.     

Current status of the engineering design of the test modules for the IFMIF

Under Broader Approach (BA) Agreement between EURATOM and Japan, IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) has been performed since the middle of 2007. IFMIF presents three main facilities (the Accelerator Facility, Li Target Facility and Test Facilities). A previous design of IFMIF was summarized in the comprehensive design report [1]. The present EVEDA phase aims at producing a detailed, complete and fully integrated engineering design of IFMIF. The delivery of the “Intermediate IFMIF Engineering Design Report” is foreseen mid-2013. The goal of IFMIF is to obtain the indispensable design database to allow the design and licensing of DEMO and ensuring commercial reactors thanks to the materials data set obtained from planned evaluation tests such irradiations in high flux test modules (HFTM-vertical rig, HFTM-horizontal rig), medium flux test modules (creep fatigue test module, tritium release test module, liquid breeder validation module) and low flux test modules of IFMIF. In addition, the Startup Monitoring Module will be used for IFMIF commissioning. This paper is summarizing the overall current progress status of the engineering and conceptual design of the test modules in IFMIF/EVEDA.