ITER International Organization Research Articles

The European master of science in nuclear fusion and engineering physics (FUSION-EP): 15 years of experience

The aim of the Erasmus Mundus Joint Master Degree (EMJMD) FUSION-EP is to provide a high-level research-oriented education and a well-integrated cultural experience within an international consortium of institutions leading the interdisciplinary field of magnetic fusion. The combined and harmonized teaching and research activities coordinated among eight partner universities in Belgium, Czech Republic, France, Germany and Spain offer a variety of competences in a field of crucial importance to the problem of world energy supply. Besides the ITER International Organization, there are 24 EU and non-EU academic and research associate partners. As the fusion community makes pace towards commercialization, the balance in research will shift from an academic-and-fundamental direction to a technology-and-industry-driven orientation. For this purpose, the programme offers two tracks: fusion science and fusion engineering. In addition, an advisory board keeps the curriculum in line with the research activities carried out by the associate and industrial partners. Internationalization, multidisciplinarity, and student mobility are inherent parts of the programme structure and philosophy. There exist several student and alumni initiatives leveraging meaningful connections within a global network. As a first step towards long-term sustainability, the consortium decided to transfer the programme coordination from Ghent University to Aix-Marseille University in 2019; the geographical, topical and professional proximity of ITER is expected to allow an optimal alignment of the programme with the professional needs of the emerging nuclear fusion industry.

ОШИБКИ В РАДИАЦИОННОМ НАГРЕВЕ УРАНА ПРИ РАСЧЕТАХ С ИСПОЛЬЗОВАНИЕМ БИБЛИОТЕКИ FENDL 3.1D

The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.

Creation of Thermonuclear Experimental Reactor ITER as an Example of International Scientific and Technical Cooperation in Energy Sector

The article analyses the prospects for the implementation of the largest international scientific and technical project (“mega-science” project) ITER of controlled thermonuclear fusion. With the depletion of fossil fuel reserves, the role of new energy sources, including those that have not yet been developed, is increasing. Due to the extremely high costs and ambitious scientific goals facing the project, its practical implementation and funding are possible in close international scientific and technical cooperation in the energy sector. The role of an international group of seven participants (Russia, the EU, Japan, China, India, the Republic of Korea and the USA) in the creation of laboratory fusion facilities is noted. The choice of a Tokamak-type design for the ITER international reactor being built in France is the recognition of Russian’s decisive role. The article reveals the main parameters of international partnership, the specification of which is carried out on a regular basis at ITER coordination meetings with the participation of representatives of national ITER agencies. The parameters of interaction with scientific institutions of countries that are not members of the consortium are given. The evolution of the project since its launch in 1985 as a joint Soviet-American initiative is shown. Four options for the initial placement of the reactor are considered: in Spain, France, Canada and Japan. The features of the organizational and legal regulation of the international consortium for the management of ITER are disclosed, including the establishment of the ITER International Organization for Thermonuclear Energy for the joint implementation of the ITER project in 2006. The role of the in-kind contribution of countries in international projects in the exchange of unique world-class technologies is discussed. The main goals set by the countries participating in the project and the time horizons for their practical achievement are listed. The parameters of Russia's participation in the project are given, the prospects for further cooperation in this area are evaluated. Particular attention is paid to the peculiarities of the political interaction of the participating countries, as well as to the intermediate results already achieved during the implementation of the ITER project.

Progress in the ITER neutral beam test facility

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER.This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

Integration of the Full Tokamak Reference Model with the Complex Model for ITER Neutronic Analysis

The ITER International Organization has developed a number of reference Monte Carlo N-Particle (MCNP) models including the tokamak machine C-model, the Tokamak Complex model, and the neutral beam injection (NBI) systems model. The Tokamak Complex model primarily describes building structures beyond the bioshield. Representation of the tokamak and its systems are not included in this model. The Oak Ridge National Laboratory Radiation Transport Group has conducted two ITER neutronic analysis model integrations: (1) integration of the tokamak C-model with the Tokamak Complex model for shutdown dose rate characterization in Port Cell 16 at level B1, and (2) integration of the NBI model with the Tokamak Complex model for estimating the spatial distribution of biological dose rate at levels L1, L2, and L3 of the Tokamak Complex. The integrated models were further extended to include models of system components that are essential to the neutronic analyses. This paper presents the approach and computer tools used to integrate existing reference models, describes the additional design details implemented in the integrated models, and provides representative neutronic calculations based on the extended models.

Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 “Blanket First Wall” of February 14, 2014, and 1.6.P3.RF.01 “Blanket Module Connections” of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m2. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015–2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

Advanced geometry navigation methods without cavity representation for fusion reactors

Particle transport simulations of models comprising large numbers of complex and irregular geometrical shapes is a great challenge in neutronics design and analysis of fusion reactors. All the space in the particle transport universe should be described including cavity in MCNP which is the most common deployed tool in fusion reactors. The quality of the cavity directly affects the calculation accuracy and efficiency. In view of the difficulty in describing the cavity and the instability of calculation efficiency, a new geometry representation method without cavity description was formed in Super Monte Carlo Program for Nuclear and Radiation Simulation (SuperMC). Advanced geometry navigation was developed in SuperMC to get high performance for particle transport in fusion reactors. The ITER benchmark model, a validation model released by ITER International Organization, was used to verify the accuracy and efficiency of geometry representation and navigation methods in SuperMC. The results in SuperMC showed consistency with MCNP. Besides, the computation speed was 10% faster than MCNP, while it brought more convenience without cavity representation for fusion reactors.

Control systems for ITER diagnostics, heating and current drive

The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170GHz gyrotrons and 5 steerable launchers to deliver 20MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55MHz sources and two multi-element launchers to deliver 20MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5MW of 1MeV beams to the plasma.Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design.The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing the control designs, including on which is preparing detailed diagnostic use-cases to help designers out in the field. The ITER International Organization will reinforce the methodology by holding regular User Meetings, in which successes and problems can be shared. Also, the distribution of CODAC Cores Systems and the holding of Hands-On Workshops is intended to help users in the Domestic Agencies and their suppliers to buy in to the standardization and to build compatible I&C systems, which will integrate seamlessly into CODAC when they arrive on the ITER site.

Performance analysis of the Nb−Ti conductor qualification samples for the ITER project

The ITER machine will require approximately 275 tons of Nb−Ti strands that will be used in poloidal field (PF) coils, correction coils (CC) and feeder busbars. The performance of all these conductors for the ITER machine is qualified by a short full-size sample (4 m) current sharing temperature (Tcs) test in the SULTAN facility at CRPP in Villigen, Switzerland, at the design operating current and peak field. Three ITER domestic agencies participated in PF conductor fabrication (China, the European Union, Russia) while the conductors for feeder busbars and correction coils are entirely produced by the Chinese domestic agency. Each conductor type was qualified by the ITER International Organization after reaching Tcs values in excess of ITER specifications. This qualification enabled the launch of procurement and industrial production of the Nb−Ti cable-in-conduit conductors in each domestic agency. In this paper, we summarize the performance of the qualified Nb−Ti samples of the ITER Project, comparing strand performance with conductor performance. The details of the test results will be discussed in terms of dc performance, ac losses and minimum quench energies of each conductor type.

Benchmarking of CAD-based SuperMC with ITER benchmark model

Neutronics design and analysis of fusion reactors is significantly complex mainly on geometry and physical process of neutron. The great challenges brought by advanced nuclear energy system promote the development of Super Monte Carlo Calculation Program for Nuclear and Radiation Process (SuperMC). The ITER benchmark model, a verification model created by ITER International Organization, was used for benchmarking the latest SuperMC which can perform CAD-based neutron and photon transport calculation. The calculation results of SuperMC for the first wall, divertor cassettes, inboard toroidal field coils and equatorial port were compared with the results of MCNP and the results were coincident. The intelligence and advantage of SuperMC on automatic conversion from complicated CAD model to full format calculation model, complex source construction and geometry description method was demonstrated. The correctness of neutron and photon transport in energy range corresponding to fusion reactors was also demonstrated.

Overview of Conductor Production for ITER Toroidal Field Magnet in Korea

The ITER toroidal field (TF) conductor is made up of superconducting Nb3Sn and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded stainless steel jacket sections. For the ITER Project, the Korean Domestic Agency (KODA) took the responsibility of the procurement of 27 superconducting conductors for the ITER TF magnets. After concluding the Procurement Arrangement (PA) with the ITER International Organization in May 2008, KODA has been implementing the PA through four major industrial contracts: (1) Nb3Sn strand, (2) cable, (3) stainless steel jacket sections, and (4) jacketing. Prior to the production of conductors required for the TF coils, one 760-m-long copper dummy conductor and one 100-m-long superconducting conductor were fabricated for manufacturing process qualification. As of June 2013, 16 TF conductors were successfully manufactured. The full-size conductor performance tests in the SULTAN facility yielded very high performance. This article describes the technical requirements of the TF conductor and how KODA has been manufacturing the conductors with a high-level quality assurance/quality control system. It also presents the results of acceptance tests, including those of the SULTAN test.

Preliminary Performance Test Results of First CICC From Korea Destined for an ITER TF Magnet

The Republic of Korea is participating in the ITER project to construct and operate the ITER tokamak for the purpose of demonstrating the feasibility of fusion power. ITER Korea, the implementing agency for the Republic of Korea that is procuring items for the ITER project on behalf of the Korean government, has established domestic and international contracts for the procurement of cable-in-conduit conductor (CICC) that will be used in the ITER toroidal field (TF) magnets. The CICC for the ITER TF magnets is made of superconducting and copper strand cable inside a cylindrical stainless steel jacket, and is designed to operate at a nominal peak field of 11.8 T at 4.5 K with 68 kA of nominal operating current. Recently, the first CICC from Korea, which will be installed in an ITER TF magnet, has been manufactured and tested including testing performed on a 4 m sample near ITER operating conditions at the CRPP-EPFL SULTAN facility in Villigen, Switzerland under the coordination of the ITER International Organization. This paper provides a brief description of the CICC along with preliminary results of the tests and the conductor performance characteristics derived from the results.

Assessment of materials data for blanket materials within the European contribution to ITER

In recent years, material development for ITER blanket components has been fervent; yet, a lack of materials data has been identified by the ITER International Organization (IO) and other ITER parties. Therefore, extensive work on assessment of materials data and qualification of the materials with all the relevant interfaces (joints) has been performed to ensure that requirements are fulfilled for the ITER operational conditions with acceptable margins for the foreseen lifetime of the ITER project. This paper will provide an overview of this qualification program with examples of recent results within the scope of the European contribution of blanket components to ITER. Ongoing actions to achieve a comprehensive understanding of the three main materials, beryllium, CuCrZr and stainless steel grade 316L(N)-IG, and their joints included in the blankets are discussed.

Development of ITER HTS Current Lead at ASIPP

In the past 4 years, Institute of Plasma Physics, Chinese Academy of Science (ASIPP) had developed and tested three-type (TF-type 68 kA, PF/CS-type 52kA, CC-type 10 kA) trial HTS current leads for ITER. According to the Procurement Agreement (PA) between ITER international organization (IO) and China Domestic Agency (CNDA), a large scale of feeder system including the key components, HTS current leads, will step into the qualification of fabrication soon in this year. Series products of the HTS current leads will commence on manufacturing from next year. The paper mainly describes the ITER HTS current lead development and tests at ASIPP.

Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and is the largest component of the ITER device.

RAMI Analysis for Designing and Optimizing ITER Tokamak Cooling Water System

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment.TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER’s success, and it interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements.RAMI analysis indicates appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. Results from the study indicate that pump and heat exchanger reliability are key issues.The ITER International Organization (ITER IO) RAMI group has proposed standardization for the pumps and heat exchangers to reduce fabrication risks of one-of-akind components. An aggressive maintenance/repair program should be an integral part of ITER operations. Preventive maintenance should be part of the maintenance system.For equipment that is infrequently operated, it is suggested that the key components be tested on a regular basis. The analysis has indicated that the RAMI requirements have been exceeded at this point, allowing concentration on cost optimization.

ITER’s Tokamak Cooling Water System and the Use of ASME Codes to Comply with French Regulations for Nuclear Pressure Equipment

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition - a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown.Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements.TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system.U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER–International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance.This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.

Design and Testing of ITER TF Coil Insulations

The design of the ITER Toroidal Field (TF) coils requires an insulation system that is amenable to the very large scale vacuum impregnation processes planned for the construction of these devices, and that will provide reliable electro-mechanical performance after radiation exposure. To address this need, CTD has developed an epoxy/cyanate ester resin system designated CTD-425. This material meets the processing requirements for use in the TF coils, and cyclic mechanical testing of conductor assemblies has demonstrated its electro-mechanical strength after 60,000 mechanical cycles. In addition, this product recently passed radiation exposure tests coordinated by the ITER International Organization and is now qualified for use in constructing the TF coils. This paper summarizes the performance characterization and qualification test results for this insulation.

The results of the second Chinese TF conductor sample

In the framework of the ITER qualification tests the second Chinese sample (CNTF2) was tested in the SULTAN facility. The sample was made of two identical ITER TF type cable-in-conduit conductor sections. Internal Sn strands offered by Western Superconducting Technologies Company (WST) were used for the conductor fabrication. The main characteristics of the conductor and the strand are presented as well as the sample assembly details. The test was performed in accordance with the ITER International Organization specified program. In order to evaluate the conductor performance, the current sharing temperature was measured at specified electromagnetic load cycling steps. Few voltage–current characteristics were measured with the aim of n-value estimation. Before cycling load, both conductor sections of CNTF2 showed different performance, but after 1200 cycles and warmed up and cool down, they showed identical performance: the current sharing temperature ( T cs ) was 6.36 K, which exceeds ITER requirement of 5.7 K.

The Results of Russian ITER TF Conductor Sample Test in SULTAN

In the framework of the ITER Qualification Tests the Russian sample RFTF2 was tested in SULTAN facility. The sample is made of two identical ITER TF-type cable-in-conduit conductor sections. Bronze route strands were used for the conductor fabrication. The main characteristics of the conductor and the strand are presented as well as the sample assembly details. The test was performed in accordance with the ITER International Organization specified program. In order to evaluate the conductor performance, the current sharing temperature was measured at specified electromagnetic load cycling steps. Few voltage-current characteristics were measured with the aim of n-value estimation. Both conductor sections of the RFTF2 sample showed identical performance: the current sharing temperature is 6.4 K, that not only satisfies, but substantially exceeds ITER requirement of 5.7 K. The tests of RFTF-2 conductor sample showed that electromagnetic cyclic load as well as warm-up had negligible effect on the conductor performance. The results of the SULTAN sample test are compared with strand performance.