International Thermonuclear Experimental Reactor Design Research Articles

Exploring the limits of a very largeNb3Sn conductor: the 80 kA conductor of the International Thermonuclear Experimental Reactortoroidal field model coil

In a phase II experiment on the International Thermonuclear Experimental Reactor(ITER) toroidal field model coil (TFMC) the operation limits of its 80 kANb3Sn conductor were explored.To increase the magnetic field on the conductor, the TFMC was tested in the presence ofanother large coil: the Euratom LCT coil. Under these conditions the maximum fieldreached on the conductor was around 10 T.This exploration has been performed at constant current, by progressively increasing thecoil temperature and monitoring the coil voltage drop in the current sharing regime.Such an operation was made possible thanks to the very high stability of the conductor.The aim of these tests was to compare the critical properties of the conductor withexpectations and to assess the ITER TF conductor design. These expectations are based onthe documented critical field and temperature dependent properties of the 720superconducting strands which compose the conductor.In addition the conductor properties are highly dependent on the strain, due to the compression appearingon Nb3Sn during the heat treatment of the pancakes and related to the difference in thermal compression betweenNb3Sn and the stainless steel jacket. No precise model exists to predict this strain, which istherefore the main information which is expected from these tests.The method to deduce this strain from the different tests is presented, including athermohydraulic analysis to identify the temperature of the critical point and a carefulestimation of the field map across the conductor.The measured strain has been estimated in the range−0.75% to−0.79%. This information will be taken into account for ITER design and some adjustment of theITER conductor design is under examination.

Test results on the first 50 kA NbTi full size sample for the International ThermonuclearExperimental Reactor

Within the framework of the R&D studies for the International Thermonuclear ExperimentalReactor (ITER) project, the first full size NbTi conductor sample was fabricatedindustrially and tested in the SULTAN facility (Villigen, Switzerland). This sample(PF-FSJS), which is relevant to the poloidal field coils of ITER, is composed of two parallelstraight bars of conductor, connected at the bottom through a joint designedaccording to the CEA twin-box concept. The two conductor legs are identical exceptfor the use of different strands: a nickel plated NbTi strand with a pure coppermatrix in one leg, and a bare NbTi strand with a copper matrix and internal CuNibarrier in the other leg. The two conductors and the joint were extensively testedas regards DC and AC properties. This paper reports on the test results andanalysis, stressing the differences between the two conductor legs and discussing theimpact of the test results on the ITER design criteria for the conductor and joint.While joint DC resistance, conductors and joint AC losses fulfilled the ITERrequirements, neither conductor could reach its current sharing temperature at relevantITER currents, due to instabilities. Although the drop in temperature is slightfor the CuNi strand cable, it is more significant for the Ni plated strand cable.

ITER site selection studies in Spain

The studies carried out to evaluate and select a candidate site for International Thermonuclear Experimental Reactor (ITER) construction in Spain are presented in this paper. The ITER design, completed in July 2001, considered a number of technical requirements that must be fulfilled by the selected site. Several assumptions concerning the ITER site were made in order to carry on the design before final site selection. In the studies undertaken for ITER site selection in Spain, the referred technical requirements and assumptions were applied across the whole of Spain and two areas were identified as being preferential. These areas are on the Mediterranean coast and are situated in the Catalan and Valencian regions. A comparative evaluation based on technical characteristics for the concrete plots, proposed within the preferential areas, has been done. The result of these studies was the selection of a site that was deemed to be the most competitive—Vandellós (Tarragona)—and it was proposed to the European Commission for detailed studies in order to be considered as a possible European site for ITER construction. Another key factor for hosting ITER in Spain, is the licensing process. The present status is summarised in this paper.

Effect of pressure on the radiation induced electrical conductivity of NBI insulator gases

A study of the effect of pressure on the radiation induced electrical conductivity (RIC) of candidate international thermonuclear experimental reactor (ITER) neutral beam injectors (NBI) insulation gases has been carried out. Large volume, pressure and voltage values similar to those for the ITER system are not readily achievable experimentally. To overcome this problem different experiments have been performed in a Van de Graaff electron accelerator at lower values and the results compared with a theoretical model. Excellent agreement is found between theoretical and experimental results allowing extrapolation to the higher voltage, gas volume and particularly higher gas pressure required in the ITER design. The results show that the use of high gas pressure to increase the voltage breakdown threshold and reduce size may also result in enhanced leakage current losses.

Neutron irradiation effect on the mechanical properties of type 316L SS welded joints

In the International Thermonuclear Experimental Reactor design activity, the vacuum vessel (VV) is designed as a double walled structure so that some parts are not qualified by the conventional design standards. JAERI has executed the preparation activity of the new design standards and obtained the technical data to support them. In this study, neutron irradiation effects on the mechanical properties of 316L SS welded joints were investigated. The tensile and Charpy-impact specimens were irradiated at 473 K. The results of post irradiation experiments indicate that sufficient ductility is still maintained for tungsten inert gas (TIG) and electron beam (EB) welded joints; whereas, the impact properties of metal inert gas (MAG) weld metal are extremely poor. Consequently, the soundness of the 316L SS base metal and its TIG and EB welded joints are retained after 0.2–0.5 dpa neutron irradiation. However, it is rather difficult to adopt MAG welding for the fabrication of the VV.

Comparison of athena/ relap results against ice experimental data

In order to demonstrate the adequacy of the International Thermonuclear Experimental Reactor design from a safety stand point as well as investigating the behavior of two-phase flow phenomena during an ingress of coolant event, an integrated ICE test facility was constructed in Japan. The data generated from the ICE facility offers a valuable means to validate computer codes such as athena/ relap5, which is one of the codes used at the Idaho National Engineering And Environmental Laboratory (INEEL) to evaluate the safety of various fusion reactor concepts. In this paper we compared numerical results generated by the athena code with corresponding test data from the ICE facility. Overall we found good agreement between the test data and the predicted results.

Evaluation of Shutdown Gamma-ray Dose Rates around the Duct Penetration by Three-Dimensional Monte Carlo Decay Gamma-ray Transport Calculation with Variance Reduction Method

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation locations, and forcibly terminate the gamma-ray transport calculation in the cells except for the identified cells. In order to validate the effectiveness of the method, shielding calculation for actual ITER (International Thermonuclear Experimental Reactor) configuration is performed, and small statistical errors below criteria are obtained. The effectiveness of the proposed method for ITER design analysis is demonstrated.

Collective dose at ITER feat

The safety study performed until December 2001 to assess the collective dose to the workers based on the International Thermonuclear Experimental Reactor (ITER) Fusion Energy Advanced Tokamak (FEAT) project, is presented in this work together with the relevant results. All systems located in the tokamak building of ITER FEAT facility important from the radiological point of view are considered. Radiological source terms with an important collective dose impact are airborne tritium and activated corrosion products (ACP) in the coolant of the water cooling system (WCS). A suitable computer code is used to assess the ACP inventory in the different WCS components. The dose rate is then assessed by considering walls attenuation with the 3D transport code MCNP, a Monte Carlo code that performs gamma ray transport calculation. Working activities needed to operate, maintain and replace each component of systems under study are partly provided by system designers and partly derived from the experience developed in operating nuclear fission reactors. In conclusion the collective dose result is shown and partly compared with that of previous ITER design stages considering the different systems and solutions and pointing out ALARA improvements.

Clearance and disposal of ITER radioactive waste components

Future fusion reactors will generate radioactive material. In the present paper results from studies on radioactive material from two different International Thermonuclear Experimental Reactor (ITER) designs are presented. Results on quantities of radioactive materials, clearance possibilities and waste management are given for the 1998 ITER design. Preliminary results on quantities of radioactive materials and clearance are also given for the Fusion Energy Advanced Tokamak, ITER-FEAT.

Analyses of temperature transients in ITER design concepts following hypothetical loss of cooling accidents

Bounding calculations of thermal transients following a hypothetical (Category V) loss of coolant accident in two international thermonuclear experimental reactor (ITER) design options, EU-I and ITER-fusion energy advanced tokamak (FEAT), have been carried out for a variety of assumptions concerning heat-transfer to and heat rejection from, the cryostat. It is shown that, even when unrealistically conservative assumptions are made in illustrative calculations, the temperatures do not approach melting point for any part of the plant structure. In all cases, maximum temperatures are reached within 1 year. It is also concluded that for the baseline scenario for ITER-FEAT in which helium gas is assumed to be present in the cryostat, convectively transferring heat between components, the outboard first-wall temperature never exceeds 530°C. A significant improvement could still be made if the surface emissivity of the thermal shields could be increased by some means during the post-accident period.

ITER divertor maintenance L7 R&D project — results and perspectives

The availability of International Thermonuclear Experimental Reactor (ITER) (or an ‘ITER-like’) reactor will be strongly influenced by the effectiveness of the in-vessel components remote handling strategy. In the present reactor concept, the relevant key components are the divertor cassettes, located in the lower region of the vacuum vessel. Due to erosion and damage to the divertor plasma facing components, estimated scheduled replacement of the cassettes will be required eight times during the machine lifetime. Moreover, for such an experimental tokamak where completely new plasma regimes will be established, unscheduled interventions cannot be excluded a priori. To test and optimise the divertor maintenance scenario, the so called ITER L7 R&D project has been realised at the ENEA Brasimone laboratories, with a full scale simulation of the in-vessel cassette maintenance environment and of the hot cell refurbishment operations. The basic demonstration of the validity of the current scenario has been established, and now new activities are in progress to optimise many aspects of the operations (procedures, hardware improvements, reliability, etc.). Based on the background and the results of these activities, this paper discusses the lessons learned during the project implementation, and identifies key points of the current strategy that should be maintained for any new design of ITER or an ‘ITER-like’ reactor.

Investigation of neutron and photon flux spectra in a streaming mock-up for ITER

A neutronics mock-up of the International Thermonuclear Experimental Reactor (ITER) inboard shielding system with streaming path for the fusion neutrons was investigated in order to validate the design. Neutron and photon flux spectra were measured at several positions in the assembly. The experimentally determined flux spectra are analysed with the ITER design tools Monte Carlo code MCNP-4A and Fusion Evaluated Nuclear Data Library FENDL-1 and FENDL-2. Helium production, radiation damage and photon heating rates, which are design parameters of the shielding system are derived from both, the measured and the calculated spectral fluxes.

Modeling an Unmitigated Quench Event in an ITER Toroidal Field Magnet

An unmitigated quench of a toroidal field (TF) magnet has been proposed as an extremely unlikely event for International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity safety analysis. While the frequency of such an event is highly improbable (<1 × 10-6/yr), the public safety consequences of this event must be explored because the TF magnets are located midway between the two primary confinement barriers of the ITER design. These confinement barriers are the vacuum vessel (VV) and the cryostat. An unmitigated quench has the potential for producing melting of the magnet. If molten material from the magnet were to impinge on the walls of the VV and cryostat, these walls could fail, resulting in a pathway for release of radioactive material to the environment from the VV. A model has been developed at the Idaho National Engineering and Environmental Laboratory called MAGARC to investigate the consequences of this accident. This model is described in detail, and results from this model used in ITER safety analysis are presented.

Progress in U.S. Fusion Safety and Environmental Activities over the Last Decade

Magnetic fusion energy has the potential for superior safety and environmental (S&E) characteristics relative to other energy options, which is one of the main reasons for developing fusion power. Excellent progress has been made in understanding the nature of the S&E concerns associated with fusion power and in demonstrating the S&E potential of fusion. Over the past 10 yr, U.S. fusion S&E activities have been largely focused on the International Thermonuclear Experimental Reactor (ITER). The design of ITER is such that the hazards addressed are similar to those of a future fusion power plant; hence, many of the safety issues addressed by ITER are relevant to commercial fusion power plants. This paper reviews the progress and accomplishments in fusion S&E activities performed largely in support of ITER over the past decade and discusses future directions in fusion safety design criteria development and implementation; characterization of the radioactive and hazardous materials in fusion and the potential energy sources that could mobilize those materials during an accident; integrated state-of-the-art safety and risk analysis tools, methods, and results; and development of environmental design criteria for radioactive and hazardous fusion waste minimization as well as the evaluation of recycle/reuse potential of fusion materials.

Europe, Japan Finalizing Reduced ITER Design

MUNICH, GERMANY-- European and Japanese fusion researchers have drawn up what they hope will be a winning design for a scaled-down version of the International Thermonuclear Experimental Reactor (ITER). Last week ITER scientists described key elements of the smaller and cheaper design at a seminar here for policy-makers, industrialists, and journalists. Details will be revealed next month, in time to influence political decisions to be made starting next summer in Europe and Japan, the two major ITER partners.

ITER achievements by July 1998 and future prospects

The four parties to the Agreement on Co-operation in the Engineering Design Activities for the International Thermonuclear Experimental Reactor (ITER) have successfully completed 6 years of collaborative work, the results of which are encapsulated in the ITER Final Design Report, Cost Review and Safety Analysis, (FDR) and supporting technical information, including the output from ongoing major technology R&D projects. The FDR provides a detailed, complete and fully integrated engineering design of ITER, including the latest assessment of ITER plasma performance and safety characteristics and of its cost which remains as originally agreed. The technical results provided should have been sufficient to allow a construction decision by the parties who have now found the investment too large to be affordable. Further development of the project, aimed at enabling a future possible decision on construction, includes design work to achieve cost reduction by reducing performance in ways that will still allow to meet ITER overall programmatic objectives; to adapt the design to specific site characterisations; and to prepare for regulatory procedures.

Application of High-Temperature Superconductors to Enhance Nuclear Fusion Reactors

Fusion plasma engineers have made remarkable progress in designing a tokamak type of experimental reactor, as evidenced by the International Thermonuclear Experimental Reactor (ITER), which produces fusion energy of 1.5 GW(thermal) for 1000 s at least. However, the ITER design is more expensive and requires more advanced technology than earlier machines. With these concerns in mind, extending design options by using a high-temperature superconductor (HTSC) to improve plasma positional instability by placing HTSC ring coils inside the vacuum vessel would be desirable. Here, improving the plasma instability with HTSC coils is discussed, and a possible design of a smaller machine using the coils based on supporting experiments with HTSC tapes is given.

ITER: an integrated approcah to ignited plasmas

The programmatic objective of the International Thermonuclear Experimental Reactor (ITER) calls for it to demonstrate controlled ignition and extended burn of deuterium–tritium plasmas, to demonstrate essential fusion reactor technologies in an integrated system, and to perform integrated testing of high–heat flux and nuclear components. The ITER design, as embodied in the Final Design Report, and its physics basis are presented and the projections of plasma performance are summarized together with the assessment of ITER9s engineering feasibility and of the progress in validating technology R and D. The future prospects for ITER and its potential central role as an integrated physics and engineering experiment in the overall development of controlled fusion as a source of useful energy are discussed.

In-vessel tritium retention and removal in ITER

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER – and more generically in any other next-step experimental fusion facility fuelled with tritium – the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Panel Report to FESAC: Review of the 1996 ITER Detailed Design Report

This report presents the results of a U.S. review of the 1996 Detailed Design Report (DDR) of the International Thermonuclear Experimental Reactor (ITER) project. It was prepared by a panel established by the U.S. Department of Energy (USDOE) Fusion Energy Sciences Advisory Committee (FESAC) and was subsequently endorsed by FESAC and provided to the USDOE. Copies of the charge and transmittal letters are incorporated at the end of this paper. Also incorporated in this paper are the reports of several subpanels established to provide detailed review and recommendations on specific topics. The authors of those subpanel reports are acknowledged in the text. The ITER was subsequently reduced in size and scope; this review refers to the full-size ITER design as it was completed in 1996.