Application of GEMSAFE to ITER CDA and Its comparison with FER

A safety analysis for the design of International Thermonuclear Experimental Reactor (ITER) in the Conceptual Design Activity stage was performed by the GEMSAFE methodology, and its results were compared with those of Fusion Experimental Reactor (FER), a Japan's facility planned next to JT-60. The objectives of this study are to confirm the applicability of GEMSAFE to ITER and to select design basis events of ITER and identify R&D items with comparison to FER. Function-Based Safety Analyses (FBSA) were carred out to select 19 and 25 design basis events for FER and ITER, respectively. The major reason for the difference is that ITER has a class-2 RI source, e.g., tritium of 7.5 × 10⁵ Ci in mobile form, in the coolant for the first wall and blankets as well as a class-3 RI source, e.g., the immobile tritium of 2.2×10⁷ Ci absorbed in first wall and dust.     

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.     

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.     

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.     

ITER operating limit definition criteria

The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction.OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent.This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.     

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.     

Design and Structural Analysis of Support Structure for ITER Vacuum Vessel

The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.     

Recommendations on the Nature and Level of U.S. Participation in the International Thermonuclear Experimental Reactor (ITER) Extension of the Engineering Design Activities

The Department of Energy (DOE) Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to address the topic of U.S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction. Given that there is expected to be a transition period of 3 to 5 years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to include the U.S. role in an interim period between the EDA and construction.This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirely. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.     

Fusion Specific Features in ITER Accident Analysis

Fusion specific features like inherent plasma shutdown, low decay heat densities, cryogenic temperatures, and limited source terms were considered during the safety design process of ITER. Uncertainties in plasma disruptions motivates a robust design to cope with multiple failures of in-vessel cooling piping. A vacuum vessel pressure suppression system mitigates pressure transients and effectively captures mobilized radioactivity. In case of pump trips or ex-vessel coolant losses in the divertor the plasma needs to be actively terminated in a few seconds. Failure to do so might damage the divertor but radiological consequences will be minor due to the intact first confinement barrier. Tritium plant inventories are protected by several layers of confinement. Uncontrolled release of magnet energy will be prevented by design. Postulated damage from magnets to confinement barriers causes fluid ingress (air, water, helium) into the cryostat. The cold environment limits pressurization. Most tritium and dust is captured by condensation.     

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.     

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.     

Structural Analysis of ITER Lower In-Vessel-Viewing Port

A finite element model of the International Thermonuclear Experimental Reactor （ITER） in-vessel viewing port was developed by the ANSYS code in order to evaluate the stress level of this structure. The thermal, elastic and modal analyses were made in succession based on the loads designated by the ITER International team. The designed loads include electromagnetic loads, seismic loads, pressure, temperature and gravity. The preliminary results of the finite element analysis （FEA） show that the stress intensity exceeded the allowable stress and the maximum stress was concentrated in the geometric discontinuous region of the shroud stub extension （SSE）. Therefore, the SSE has been modified recently. For the modified structure, we found that the stresses do not exceed the allowable value for all load combinations. In addition the modal analysis results show that the natural frequencies of the IVV port structure are located in the typical diapason of seismic excitation.     

Preliminary Design and Analysis of ITER In-Wall Shielding

ITER in-wall shielding （IIS） is situated between the doubled shells of the ITER Vacuum Vessel （IVV）. Its main functions are applied in shielding neutron, gamma-ray and toroidal field ripple reduction. The structure of IIS has been modelled according to the IVV design criteria which has been updated by the ITER team （IT）. Static analysis and thermal expansion analysis were performed for the structure. Thermal-hydraulic analysis verified the heat removal capability and resulting temperature, pressure, and velocity changes in the coolant flow. Consequently, our design work is possibly suitable as a reference for IT＇s updated or final design in its next step.     

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

Structural Analysis of the ITER VV Lower Port Region

A structural analysis of the International Thermonuclear Experimental Reactor （ITER） vacuum vessel＇s lower port region was presented by means of a finite element analysis method. The purpose is to evaluate the stress and displacement level on this structure under various combinations of five designed loads, including the gravity of the vacuum vessel, seismic loads, electromagnetic loads, and possible pressure loads to ensure structural safety. The cyclic symmetry finite element model of this structure was developed by using ANSYS code. The re- sults showed that the maximum stress does not exceed the allowable value for any of the load combinations according to ASME code and the nine vacuum vessel （VV） supports have the ability to sustain the entire VV and in vessel-components and withstand load combinations under both normal as well as off-normal operation conditions. Stress mainly concentrates on the connecting region of the VV support and lower port stub extension.     

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S.Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.     

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.     

聚变堆安全特性评价研究

确保核安全是未来聚变堆设计、建造和运行过程中必须坚持的最高原则,是聚变堆获得建造和运行许可的前提条件,也是聚变能得以吸引公众的主要理由之一。聚变堆具有高能中子、大量放射性氚、复杂结构、极端服役环境等特点,具有独特的潜在安全问题,因而必须开展针对性研究。本文将从聚变中子与放射性源项、热流体与能量传输、氚安全与环境影响、可靠性与风险管理、安全理念与公众接受度五个方面分别总结其安全特性,系统梳理其关键技术挑战,为建立聚变安全评价体系提供技术支持,进而服务于未来聚变堆的设计与建造。     

Conceptual Design and Analysis of Cold Mass Support of the CS3U Feeder for the ITER

In the International Thermonuclear Experimental Reactor(ITER) project,the feeders are one of the most important and critical systems.To convey the power supply and the coolant for the central solenoid(CS) magnet,6 sets of CS feeders are employed,which consist mainly of an in-cryostat feeder(ICF),a cryostat feed-through(CFT),an S-bend box(SBB),and a coil terminal box(CTB).To compensate the displacements of the internal components of the CS feeders during operation,sliding cold mass supports consisting of a sled plate,a cylindrical support,a thermal shield,and an external ring are developed.To check the strength of the developed cold mass supports of the CS3U feeder,electromagnetic analysis of the two superconducting busbars is performed by using the CATIA V5 and ANSYS codes based on parametric technology.Furthermore, the thermal-structural coupling analysis is performed based on the obtained results,except for the stress concentration,and the max.stress intensity is lower than the allowable stress of the selected material.It is found that the conceptual design of the cold mass support can satisfy the required functions under the worst case of normal working conditions.All these performed activities will provide a firm technical basis for the engineering design and development of cold mass supports.     

Preliminary Thermal Mechanical Analysis of the Equatorial Thermal Shield for ITER

Thermal analysis of the equatorial thermal shield for ITER is conducted in order to confirm that the cooling tube design was reasonable under both the plasma operational and the baking operational conditions. The structural performance was analyzed by means of the finite element software ANSYS. A comparison of the results with design requirements shows that the results of the simulation are within allowable design requirements, which indicates the feasibility and reliability of the equatorial thermal shield structure.