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.     

Probabilistic cost-benefit seismic design criterion for a dedicated shutdown heat removal system

A probabilistic methodology is developed for assessing the risk reduction potential and cost-benefit tradeoff of a Dedicated Shutdown Heat Removal System (DSHRS) for a PWR as a function of its seismic design capability. The option of coping with a very small LOCA is included. The annual seismic risk of a plant and a similar hypothetical plant having a proposed DSHRS with various seismic strengths are computed. The difference in the annual seismic risks is the annual seismic risk reduction benefit for having the system. The present value of the future risk reduction benefit is then compared to the cost of building a DSHRS and the incremental seismic cost associated with building the system to withstand a stronger earthquake.A reactor like Zion was used for application of the method due to the availability of data. Studies were performed to investigate the sensitivity of the results to the assumed seismic hazard, probability of occurrence of seismic-induced accident initiating events, equipment seismic fragility, accident costs, and discount rate. The incremental seismic risk reduction benefit computed in these studies ranges from $207 million for a DSHRS with a median seismic capacity of 1.70g (i.e. 10 × SSE) in a new plant built at the site. The total cost of a DSHRS is crudely estimated to be $25 million or more, if it were built to withstand the current SSE of the plant (for which the system probably would have a median seismic capacity of 0.85g or more due to various design and construction conservatisms). The cost associated with the seismic design aspect of such a system is estimated to be approximately $2.5 million and it may be doubled if the seismic design capability of the system is tripled. The cost/benefit results and their inherent large uncertainties are not definitive but indicate that probabilistic seismic design of a DSHRS should be examined in further detail.     

Updating the design of the feeder components for the ITER magnet system

The ITER superconducting magnet system generates an average heat load of 23 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68 kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0 MPa and temperature between 4.3 and 4.7 K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components.     

Recent improvements to the ITER neutral beam system design

The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D⁰ or up to 870 keV H⁰) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3], [4], [5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6], [7], [8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community.     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

ITER软X射线检测系统及其电磁兼容设计

国际热核聚变实验反应堆是世界上在建的最大的磁约束聚变装置托克马克装置,通过对其中软X射线的测量,可实现等离子体辐射对锯齿、色骨模等磁流体现象的物理研究和成像反演。软X射线诊断系统就是用来检测软X射线的设备。由于热核聚变时恶劣电磁环境及远距离传输,在设计信号检测系统时必须进行电磁兼容设计,以降低系统噪声、提高检测精度。本文中使用的检测电路采用差分结构实现电流信号到电压信号的转换,重点研究检测电路的实现及其电磁兼容设计。从电磁抗干扰的三要素出发,结合实验测试,针对电磁干扰的特殊性,讨论了滤波电路设计、印制电路板(Printed Circuit Board,PCB)走线、电磁屏蔽及信号接地在系统中实现。本文采用32通道板卡集成设计;信号增益提高至107 V?A-1;放大器带宽达到120 k Hz。通过测试结果可以看出,信号噪声降至8 mV。通过优化设计提高了检测电路的集成度和放大电路的增益及带宽,同时降低了检测电路的噪声。     

Structural design features for commercial fusion power reactor magnet systems

The evolution of structural design features for commercial fusion power reactor magnet systems is discussed. Changing concepts in plasma physics and impurity control, new data on radiation damage in materials and developments in the maintainability and repairability of the magnet systems are the driving influences in this evolution.Generic problems in the magnet designs are discussed for several proposed magnetic confinement system configurations, including tokamaks, tandem mirrors, the Elmo Bumpy Torus, and the reversed field theta pinch. These systems are compared on the basis of how efficiently the magnets make use of structural materials.A measure of the effectiveness of a magnet system is found by determining the ratio of net electric power output from the reactor to the stored energy in the magnetic fields produced by the magnet coils in a given system. The stored energy in the magnetic field can then be used to establish a minimum structural volume and mass by use of the virial theorem. Experience with coil types such as solenoids, toroids, Yin-Yang, etc. has established factors by which the minima must be multiplied to yield anticipated volumes and masses of realistic magnet systems. These initial, admittedly approximate, calculations allow designers to estimate early in the process the contribution of the magnet systems to the overall cost of a fusion reactor. As work progresses these estimates can be used to indicate the degree to which the designer is making effective use of the structural material.Basic rules for effective placement of structure, common to all magnet systems, are also discussed in detail. Factors are presented which make it possible to compare structural savings to the cost of researching the parameters involved in the stability of superconductors.     

Nuclear design of a heliotron-H fusion power reactor

Nuclear analysis was carried out for the heliotron-H fusion power reactor employing anl=2 helical heliotron field. The neutronics aspects examined were (a) tritium breeding capability, (b) shielding effectiveness for the superconducting magnet (SCM), and (c) induced activity after shutdown. In this reactor design of the heliotron-H, the space available for the blanket and shield is limited due to the reactor geometry. Thus, some parametric survey calculations were performed to satisfy the design requirements. The nucleonic design features of the heliotron-H are as follows. An adequate tritium breeding ratio of 1.17 is obtained when a 10-cm thick Pb neutron multiplier and a 40-cm thick Li₂O breeding blanket are used. In this case, the total nuclear energy deposition is 16.10 MeV per 14.06 MeV incident neutron. The performance of the SCM is assured during 2 yr of continuous operation using a 20-cm thick tungsten shield. Biological dose rate behind the SCM at 1 day after shutdown is too high for hands-on maintenance.     

可控核聚变与国际热核实验堆（ITER）计划

介绍了我国能源的基本隋况,核聚变能和可控核聚变的基本原理,以及国际热核聚变实验堆ITER的历史与现状。对我国磁约束核聚变的研究发展历程做了简要的回顾。     

Physics of the conceptual design of the ITER plasma control system

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints.In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions.     

Towards the conceptual design of the ITER real-time plasma control system

ITER will be the world's largest magnetic confinement tokamak fusion device and is currently under construction in southern France. The ITER Plasma Control System (PCS) is a fundamental component of the ITER Control, Data Access and Communication system (CODAC). It will control the evolution of all plasma parameters that are necessary to operate ITER throughout all phases of the discharge. The design and implementation of the PCS poses a number of unique challenges. The timescales of phenomena to be controlled spans three orders of magnitude, ranging from a few milliseconds to seconds. Novel control schemes, which have not been implemented at present-day machines need to be developed, and control schemes that are only done as demonstration experiments today will have to become routine. In addition, advances in computing technology and available physics models make the implementation of real-time or faster-than-real-time predictive calculations to forecast and subsequently to avoid disruptions or undesired plasma regimes feasible. This requires the PCS design to be adaptable in real-time to the results of these forecasting algorithms. A further novel feature is a sophisticated event handling system, which provides a means to deal with plasma related events (such as MHD instabilities or L-H transitions) or component failure. Finally, the schedule for design and implementation poses another challenge. The beginning of ITER operation will be in late 2020, but the conceptual design activity of the PCS has already commenced as required by the on-going development of diagnostics and actuators in the domestic agencies and the need for integration and testing. This activity is presently underway as a collaboration of international experts and the results will be published as a subsequent publication. In this paper, an overview about the main areas of intervention of the plasma control system will be given as well as a summary of the interfaces and the integration into ITER CODAC (networks, other applications, etc.). The limited amount of commissioning time foreseen for plasma control will make extensive testing and validation necessary. This should be done in an environment that is as close to the PCS version running the machine as possible. Furthermore, the integration with an Integrated Modeling Framework will lead to a versatile tool that can also be employed for pulse validation, control system development and testing as well as the development and validation of physics models. An overview of the requirements and possible structure of such an environment will also be presented.     

Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.     

Reliability analysis of shutdown system

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 × 10⁻⁸/de for failure of shutdown function in case of global faults and 4.4 × 10⁻⁸/de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 × 10⁻⁶/ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is <1 × 10⁻³/ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability.     

ITER Upper Port Plug handling cask system assessment and design proposals

The current design of the ITER cask for Upper Port Plugs has been evaluated. Careful reduction of the number of mechanical degrees of freedom is an opportunity to relax the tolerances in the design, resulting in cost reduction and reliability increase. A new kinematical design for the tractor module has a higher stiffness to weight ratio, reduces actuator forces by a factor four and minimizes cross-talk between lift and rotation motion. Non-cantilevered handling is recommended to reduce wheel loads on the tractor by a factor six and to simplify guidance. At the system level the tubular guide (TG) is proposed, a semi-permanent 3.5 m long tube which is an extension of the Upper Port. Cask docking is simplified and the risk of the cask tilting is prevented. Redesigning the system concept is recommended and the TG looks promising. Since a system level redesign impacts the external interfaces, overall feasibility has to be investigated.     

ITER中国液态锂铅实验包层模块氚提取系统设计

ITER中国液态锂铅实验包层模块氚提取系统(TES)是通过含0.1%H2的低压氦吹洗气流,在鼓泡器中将液态锂铅内产生的氚交换和载带出来,进入同位素分离系统连接进行氚提取.给出了该系统总体参数、工艺流程、辅助设施等设计.     

Design of a high power millimeter wave launcher for EC H&CD system on ITER

An optimization procedure of the quasi-optical system for a millimeter wave launcher is developed for the ITER electron cyclotron heating and current drive (EC H&CD) launcher. In the launcher, the radiated RF beams from eight corrugated waveguides are reflected sequentially by two mirrors and injected into a plasma through a small aperture in the blanket shield module on the top of the launcher. Using a steepest decent method, the heat load on the mirrors is successfully reduced to the acceptable level by flattening the RF power profile on the mirrors keeping the scattering of the RF power to a minimum from the mirrors. It is found that 20 MW injection will be acceptable even when the resistivity of 2.64 × 10⁻⁸ Ωm for the surface of the mirror (dispersion strengthened copper, 151 °C assumed) is increased by a factor of ∼10 with a contamination.     

The status of the ITER design

In parallel with a rapid build up to almost 300 people within the International Organization at Cadarache, the project team, including many from the member countries represented by their domestic agencies (DA), has concentrated its effort on an overall design review of ITER. An updated technical baseline was presented to council at the end of 2007. Several additional improvements were included during spring 2008 and it is probable that the results of the review will be accepted by council. As a result, the ITER design today provides a robust basis for a technical design that allows operation over a wide range of physical parameters, a design that can operate stably with high gain and can exploit the full scientific potential of the device. In the technical area, design changes have been integrated to improve performance, provide more robust subsystems and to minimize technical or operational risks. All of the adaptations required to support the licensing process as a nuclear facility in France have been made. In parallel major components are already under construction within the DAs. A full overview of the status of ITER design and construction, including the detailed discussion of the 2007 ITER baseline, is given. In addition, the construction status and the overall project review is presented.