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.     

ITER纵场磁体过渡馈线结构力学分析

运用数值方法计算了不同等离子体运行时刻纵场磁体过渡馈线(CFT)超导母线上的电磁载荷,并确定了磁感应强度最大的时刻,采用增量有限元法对过渡馈线进行非线性力学分析,得到不同工况下结构上的应力分布及变形情况。分析结果表明,带有万向节的过渡馈线结构具有足够的强度来承受运行过程中的各种载荷,从而证明了结构设计的合理性。     

Safety and Environmental Activities for ITER

This paper will summarize highlights of the safety approach and discuss the ITER EDA safety activities. The ITER safety approach is driven by three major objectives: (1) Enhancement or improvement of fusion's intrinsic safety characteristics to the maximum extent feasible, which includes a minimization of the dependence on dedicated safety systems; (2) Selection of conservative design parameters and development of a robust design to accommodate uncertainties in plasma physics as well as the lack of operational experience and data; and (3) Integration of engineered mitigation systems to enhance the safety assurance against potentially hazardous inventories in the device by deploying well-established nuclear safety approaches and methodologies tailored as appropriate for ITER.     

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.     

ITER Safety Analyses with ISAS

Detailed analyses of accident sequences for the International Thermonuclear Experimental Reactor (ITER), from an initiating event to the environmental release of activity, have involved in the past the use of different types of computer codes in a sequential manner. Since these codes were developed at different time scales in different countries, there is no common computing structure to enable automatic data transfer from one code to the other, and no possibility exists to model or to quantify the effect of coupled physical phenomena. To solve this problem, the Integrated Safety Analysis System of codes (ISAS) is being developed, which allows users to integrate existing computer codes in a coherent manner. This approach is based on the utilization of a command language (GIBIANE) acting as a glue to integrate the various codes as modules of a common environment. The present version of ISAS allows comprehensive (coupled) calculations of a chain of codes such as ATHENA (thermal-hydraulic analysis of transients and accidents), INTRA (analysis of in-vessel chemical reactions, pressure built-up, and distribution of reaction products inside the vacuum vessel and adjacent rooms), and NAUA (transport of radiological species within buildings and to the environment). In the near future, the integration of S AFALY (simultaneous analysis of plasma dynamics and thermal behavior of in-vessel components) is also foreseen. The paper briefly describes the essential features of ISAS development and the associated software architecture. It gives first results of a typical ITER accident sequence, a loss of coolant accident (LOCA) in the divertor cooling loop inside the vacuum vessel, amply demonstrating ISAS capabilities.     

ITER Magnet Feeder:Design,Manufacturing and Integration

The International Thermonuclear Experimental Reactor(ITER) feeder procurement is now well underway.The feeder design has been improved by the feeder teams at the ITER Organization(IO) and the Institute of Plasma Physics,Chinese Academy of Sciences(ASIPP)in the last 2 years along with analyses and qualification activities.The feeder design is being progressively finalized.In addition,the preparation of qualification and manufacturing are well scheduled at ASIPP.This paper mainly presents the design,the overview of manufacturing and the status of integration on the ITER magnet feeders.     

Prototype Engineering Test Platform of ITER Magnet Gravity Support

ITER magnet gravity support (GS) has been redesigned as a structure of pre- assembled multi-flexible plates instead of the original welded structure. In the past several years, engineering tests of the new structure have been proposed. A prototype engineering test plat- form is being developed. In order to apply the loads/load combinations onto the test mock-up, seven hydraulic bolt tensioners in three directions have been applied to simulate various loads (forces and moments), through which the deformation of bolts, flexible plates and clamp blocks, the stress distribution in the flexible plates, the friction between the contact surface, etc. can be monitored/tested. The measurement and control system includes seven sets of synchronization controller, a 16-channel strain gauge, 25 sets of displacement sensors, etc. Principles of EDC220 digital controller and development of multi-channel control software are also demonstrated.     

Safety Analysis Results for Cryostat Ingress Accidents in ITER

Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.     

Safety Analysis Results for ITER In-Vessel Loss of Coolant Events

A number of postulated in-vessel loss of coolant accidents (LOCAs) associated with the first wall and baffle cooling systems of the ITER detailed design have been analyzed for the ITER Non-site Specific Safety Report (NSSR-1). A range of break sizes from one first wall tube break (1.57 × 10^–4 m²) to damage to all in-vessel components (0.6 m² break) have been examined. These events span the ITER event classification from likely events to extremely unlikely events. In addition, in-vessel pipe breaks in combination with bypass of the two confinement barriers through a generic penetration have been examined. In all cases, when the vacuum vessel pressure suppression system is activated, most of the radioactive inventory is carried to the suppression pool where it remains for the duration of the event. Releases in these events are well within ITER release limits.     

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.     

Evolution of the Design of Cold Mass Support for the ITER Magnet Feeder System

This paper presents the evolution of the design of cold mass support for the ITER magnet feeder system. The glass fibers in the cylinder and the flanges of the normal G10 support are discontinuous in the preliminary design. The heat load of this support from the analysis is only 4.86 W. However, the mechanical test of the prototype showed that it can only endure 9 kN lateral force, which is significantly less than the required 20 kN. So, the configuration of the glass fibers in the cylinders and flanges of this G10 support are modified by changing it to a continuous and knitted type to reinforce the support, and then a new improved prototype is manufactured and tested. It could endure 15 kN lateral forces this time, but still not meet the required 20 kN. Finally, the SS316LN material is chosen for the cold mass supports. The analysis results show that it is safe under 20 kN lateral forces with the heat load increased to 14.8 W. Considering the practical application, the requirements of strength is of primary importance. So, this SS316LN cold mass support is acceptable for the ITER magnet feeder system. On the other hand, the design idea of using continuous and knitted glass fibers to reinforce the strength of a G10 support is a good reference for the case with a lower heat load and not too high Lorentz force.     

RAMI analysis of the ITER Central Safety System

ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the main functions.     

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.     

Safety analysis of ITER failures and consequences during maintenance

This study has been a first attempt at identifying potential worker overexposure situations during machine maintenance operations. The results indicate potential areas, or situations, where worker overexposure may be possible [A. Natalizio, T. Pinna, Safety analysis of failures and consequences during maintenance, ENEA Report, FUS-TN-SA-SE-R-170, June 2007, Frascati, Italy].The key findings obtained are as follows. Firstly, we have found no machine maintenance operations where the risk of worker overexposure is considered significantly large that immediate design attention is needed.Secondly, the most significant risk of worker overexposure is due to airborne releases of radioactivity from cooling water pipes and tubes that may not have been fully drained and dried, when they are cut, or inadvertently opened, by workers (frequency of pipe-cutting activities could be significantly high).Thirdly, the risk of overexposure from human error could also be significant. This varies from mistaking the machine sector, to mistaking the component to be maintained. This is analogous to working on a live electrical circuit, when it is believed to be dead (disconnected from the power source) because the worker has mistakenly selected the wrong circuit—a look-alike one. Similarly, consider the situation of a worker mistakenly preparing to work on a cooling water circuit that is still at pressure and temperature, instead of the one that has been drained and dried. The more look-alike situations there are in the facility, the greater the probability of committing this type of error.Fourthly, when consideration is given to human error, we believe that the aggregation of different diagnostics in the same port enhances the probability of human error. At the moment, these risks cannot be quantified. The task of quantifying those risks in the future should be considered.Finally, the transport of activated in-vessel components, including components of plasma-heating and current-drive systems, in non-shielded casks, could carry with it a significant risk of worker overexposure. In the context of ALARA, this approach requires a specific study to justify its use.Concluding, it is important to note that by having identified the possibility of an overexposure situation does not mean that it is probable. The calculation of probability awaits further studies of this nature, when the design reaches a more detailed level.     

In-situ Calibration Techniques and Preliminary Assessment of Accuracy for NFM on ITER

Absolutely calibrated measurements of the neutron yields which need to cover both D-D and D-T phase of the international thermal-nuclear experimental reactor （ITER） are important for the evaluation of fusion power and fusion gain Q in D-D and D-T operations. This paper describes the in-situ calibration techniques and methods, the neutron sources including ^252Cf and neutron generator for calibration, the preliminary accuracy assessment and the error analyses. In addition, some difficult problems regarding the in situ calibration for the neutron flux monitor （NFM） on ITER are presented and discussed.     

Assessment of ITER PF coil quality from magnetic measurements

A feasibility has been demonstrated for numerical reconstruction on the base of magnetic measurements for geometrical displacements or deformations occurred in the manufacture and assembly of magnet coils. For validation of the proposed approach the test results of reconstruction of possible misalignments and deviations of the ITER PF1 coil are presented.     

Source Term and Mobilization Assessment in NSSR-1 of ITER

We describe the radioactive sources in the International Thermonuclear Experimental Reactor (ITER). The most important sources are co-deposited tritium, tritiated water, tokamak dust, and corrosion products. The co-deposited tritium is limited to 1 kg-T; the total on-site tritium inventory in the Basic Performance Phase (BPP) is 4 kg-T. Tritiated water concentrations are kept below 0.2 g-T/m³ in the divertor; other coolant loops have lower tritium concentrations. The in-vessel dust inventory is up to 100 kg-W, 100 kg-Be, and 200 kg-C. The activated corrosion product inventory is kept below 10 kg per loop.     

安全文化的评价

根据IAEASCART指南所提出的安全文化评价方法，较全面地介绍了安全文化评价的背景、基础和方法，并对安全文化评价工作在我国的开展提出了建议。     

Safety Analysis of ITER Test Blanket Module for Water Cooled Blanket with Pebble Bed Breeder

Performance test of test blanket modules in the fusion environment using the International Thermonuclear Experimental Reactor (ITER) is one of the most important mile-stone for the development of the breeding blanket of the fusion power plant. In the design of test blanket modules in the ITER, it is very important to show that test modules do not cause additional safety concern to the ITER. This work has been performed for the evaluation of the preliminary safety of the test blanket module of a water cooled solid blanket, which is the primary candidate of the breeding blanket in Japan currently. Major issues of the evaluation were, establishment of post-accident cooling in the test blanket module, hydrogen gas generation by Be/steam reaction, and pressure increase and spilled water amount by the event of coolant leakage. The analyses results showed that, suppression tank system is necessary to accommodate the over-pressure by the coolant water after pipe break in the box of the test module. Coolant water pipe break of the first wall of the test blanket module will result relatively small impact to the ITER safety because of the small inventory of the coolant water of the test module and large volume of the vacuum vessel of the ITER. However, it was clarified that the water cooled blanket with beryllium pebble as the multiplier will have the potential hazard of the hydrogen formation. Further investigation to maintain the safety on this aspect is required.