The ITER Safety Approach for External Events

ITER will be the first large-scale tokamak to be designed as a nuclear facility to provide public protection from external hazards such as earthquakes. The design approach for such events has been developed consistent with ITER's moderate hazards and overall safety approach on a basis of the ITER site assumptions. Seismic design is described including selection of ground motions for design purposes, seismic safety requirements, and the seismic classification scheme. The results of preliminary seismic assessments are summarized including the potential for seismically induced plasma vertical displacement events (VDE). Finally, potential facility modifications available to deal with site-specific external hazards are suggested. At the Detailed Design Report stage of the Engineering Design Activity (EDA), it is concluded that ITER has been designed to deal with the site design assumptions for earthquakes and can be designed to safety cope with a range of site-specific external hazards with modest changes to the facility.     

聚变实验堆ITER实验包层瞬态热工安全分析

在中国向ITER(International Thermonuclear Experiment Reactor)实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)设计分析的基础上,通过对DFLL-TBM系统相关的瞬态事故如真空室内部冷却剂泄漏、TBM(实验包层模块)内部冷却剂泄漏以及真空室外部冷却剂泄漏事故进行计算分析,评价DFLL-TBM对ITER在热工方面对安全的影响.结果表明:当发生瞬态事故时,DFLL-TBM有能力通过热辐射将余热排出,且包层结构不会熔化.DFLL-TBM可满足ITER在热工方面对安全的要求.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

Development of Safety Assessment Method for Plasma Anomaly Events in Fusion Reactors

The objective of the study is to provide a safety assessment method for plasma transients including thermal response of in-vessel components. We developed a plasma physics model for safety analysis which has been implemented in a safety analysis code (SAFALY). The SAFALY code consists of a 0-D plasma dynamics model and a 1-D thermal behavior model of in-vessel components in the thickness direction. The code can treat hydraulic accidents using the results from a hydraulic code and analyze a passive plasma shutdown due to the impurity release from the wall. The overpower events in International Thermonuclear Experimental Reactor (ITER) were investigated, when the fueling rate and confinement improvement changes. The results show no significant damage to the confinement boundary of ITER is expected, as long as the cooling system works normally.     

Architectural concept for the ITER Plasma Control System

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals.In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same command signals as their outputs. They consist of a collection of potentially conflicting control functions from which one at a time is exclusively activated by a mode selector signal.It can be shown that this architectural design is capable of implementing all of the presently known functional control requirements. Furthermore, this design takes already into account that the result of future experiments at ITER will create additional requirements on the functions or performance of ITER plasma control.     

Tritium Safety-Related Studies at TPL of JAERI

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.     

Updated safety analysis of ITER

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

Korean high heat flux test facility by using electron beam system for ITER first wall semi-prototype

A Korean high heat flux test facility for the semi-prototype (SP) qualification of an ITER first wall (FW) will be constructed to evaluate the fabrication technologies required for the ITER FW, and the acceptance of these developed technologies will be established for the ITER FW manufacturing procedure. Korea participated in this qualification program, and is responsible for suitable arrangements for the heat flux test of our fabricated SPs. Qualification testing can be started provided that adequate quality and control measures are implemented and validated by the ITER Organization (IO). The controlling measures required for all heat flux tests shall be concrete and demonstrate the satisfaction of the IO test programs. Each country shall provide a test plan covering the quality and controlling measures in the high heat flux test facility to be implemented throughout the test program. Korean high heat flux testing for these ITER plasma facing materials will be performed by using a 60 kV electron beam and a power supply system of 300 kW, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber. In addition, this facility needs a cooling system for a high-temperature target and decontamination system for beryllium filtration.     

ITER10kA电流引线高温超导段研究

国际热核实验反应堆(ITER)高温超导电流引线（HTSCL）的特点是不仅电流容量大,且安全性要求非常高,高温超导段是HTSCL的关键部件。本文论述了ITER10kA电流引线高温超导叠和超导组件的真空钎焊工艺,分析了高温超导段漏热,并对高温超导段漏热和电流引线在10kA下的安全性参数进行了测试。结果表明,电流引线不仅漏热小,且安全裕度大,满足ITER设计要求。     

Breaking down the requirements: Reliability in remote handling software

Software requirements have an important role in achieving reliability for operational systems like remote handling: requirements are the basis for architectural design decisions and also the main cause of defects in high quality software. We analyze related recommendations and requirements given in software safety standards, handbooks etc. and apply them to remote handling control systems, which typically have safety-critical functionality, but are not actual safety-systems?for example the safety-systems in ITER will be hardware-based.Based on the analysis, we develop a set of generic recommendations for control system software requirements, including quality attributes, software fault tolerance, and safety and as an example we analyze ITER remote handling system software requirements to identify and present dependability requirements in a useful manner. Based on the analysis, we divide a high-level control system into safety-critical and non-safety-critical subsystems, and give examples of requirements that support building a dependable system.     

ITER articulated inspection arm (AIA): R&d progress on vacuum and temperature technology for remote handling

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R&D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER.It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs.To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 °C).The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility.As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system.This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.     

Development of a RF-driven ion source for the ITER NBI system

Extensive R&D work on RF-driven negative hydrogen ion sources carried out at IPP Garching led to the decision of ITER to select this type of source as the new reference source for the ITER NBI system. The principle suitability of the RF source has been demonstrated in a small scale, short pulse length experiment: accelerated current densities, co-extracted electron currents at a source operation pressure, all well inside the range of the ITER requirements have been achieved simultaneously. In subsequent experiments, pulse lengths up to 1 h and the possibility of modularly extending the source to ITER source dimensions were demonstrated. The results achieved at the various IPP test beds, the lessons learnt during optimising the source for negative ion production and extraction as well as the problems still to be solved are summarized. As the next step in support of the NBI development for ITER, IPP plans to build a new test facility for beam extraction from a source of half the size for ITER.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

High heat flux capabilities of the Magnum-PSI linear plasma device

Magnum-PSI is an advanced linear plasma device uniquely capable of producing plasma conditions similar to those expected in the divertor of ITER both steady-state and transients. The machine is designed both for fundamental studies of plasma–surface interactions under high heat and particle fluxes, and as a high-heat flux facility for the tests of plasma-facing components under realistic plasma conditions. To study the effects of transient heat loads on a plasma-facing surface, a novel pulsed plasma source system as well as a high power laser is available. In this article, we will describe the capabilities of Magnum-PSI for high-heat flux tests of plasma-facing materials.     

Thermal and Hydraulic Analysis of the ITER Upper Vertical Stabilization Coil

The ITER upper vertical stabilization（VS） coil is a part of the ITER in-vessel coil（IVC） system, which has the abilities of restraining edge localized modes（ELMs） and maintaining plasma vertical stability. Preliminary structural analysis of the coil has revealed serious thermal stress problems. Due to the very restricted geometry space, it is necessary to perform detailed analysis on thermal and hydraulic characteristics to help optimal design of the coil. It will focus on the temperature distribution and energy balance, as well as some key factors, such as the coolant flow state and surface emissivity, which have influences on the coil performance. The APDL code and some hand calculations are employed in the analysis. The results show that the coolant convection can effectively take away the heat deposited in the coil. But improving the coolant flow state can hardly mitigate the peak temperature occurring at the edges of coil attachments, which are located far away from the coolant. Thermal radiation was expected to be a good method of cooling down these parts. But the reality is not so optimistic since it usually contributes little in the whole energy balance. However, the effect of thermal radiation will become remarkable when bad scenarios or accidents take place. Poor radiation performance of the coil will result in a potential safety hazard.     

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.     

Plasma cleaning technique for the removal of mixed materials

Erosion of materials by physical sputtering is the most fundamental of plasma–surface interactions in tokamaks. Carbon and tungsten materials planned to be used in ITER divertor are subjected to erosion, which produces local redeposition of mixed layers. Tritium retention in mixed materials is the major concern due to the limits imposed for safety reason by nuclear licensing. No technique has yet been proven capable of remove trapped tritium in ITER operating environment. In-situ oxidative cleaning techniques have been developed to remove co-deposits. Since oxygen can be incompatible with Be first wall (getter effect), a plasma cleaning using hydrogen–argon gas mixture has been proposed in this work. C–W mixed materials were produced for this purpose by capacitively coupled RF plasma sputtering. The process has been investigated by optical emission spectroscopy and Langmuir probe.     

铀浓缩设施核应急实时评价系统开发

本文介绍了我国首次集成开发的铀浓缩设施核应急实时评价系统。系统针对铀浓缩设施可以实时评价核临界事故和UF₆泄漏事故的辐射影响。系统可根据事故γ报警仪剂量率读数估算核临界裂变次数,并自动评价核临界事故后果。针对UF₆烟羽的重气特性,将重气模型与高斯烟羽模型相结合,可更准确地模拟UF₆的扩散过程。系统的开发解决了铀浓缩设施应急准备和响应一直缺乏的技术支持能力。