Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.     

Overview of tritium safety technology at the Tritium Process Laboratory of JAERI

The Tritium Process Laboratory of the Japan Atomic Energy Research Institute is the only laboratory in Japan where grams of tritium can be handled to carry out R&D on tritium processing and tritium safety handling technologies for fusion reactors. The tritium inventory is approximately 13 grams. Since 1988, basic research has been performed using gram-level tritium quantities. During the past 5 years, approximately 1 kilogram of tritium has been handled in experimental apparatus. The total amount of tritium released through the stack of TPL was controlled to less than 1 Ci without any accidents. In order to establish more complete tritium safety for DT fusion reactors, main R&D areas on tritium safety technology at TPL were focused on a new compact tritium confinement system, reliable tritium accounting and inventory control, new tritium waste treatments, and tritium release behavior into a room.     

Preliminary Occupational Radiation Exposure evaluation related to NET/ITER tritium systems

This paper presents the criteria adopted to evaluate Occupational Radiation Exposure (ORE) during normal operation and maintenance of NET/ITER and some results concerning the fuel cycle systems located in the tokamak and tritium buildings. Prompt radiation, activity concentration, and intake situations as well as number of workers, number of events, and exposure time are considered. Many systems and components, whose location in the plant can affect radiological protection during maintenance and/or surveillance, are identified together with the operations needed for each activity. Accidental conditions and equipment failures have been considered in the special maintenance activity when they are due to events with a high probability of occurrence so that such events might be expected during the life of the plant. Some results are reported showing the ORE figures with reference to the main activities. The total man-Sv/y for the systems and activities considered is about 0.5. Such a result, even if very preliminary and incomplete, means that ORE for the tritium systems of a machine like NET/ITER is not negligible and has to be continuously controlled during the design phase.     

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.     

氦-3回路中氚的辐射安全分析

氚安全是确保燃料元件堆内功率瞬态试验的关键因素之一.本文首先分析了氚的来源和危害,提出了氦-3回路氚的防护和去污措施,然后讨论了氚在正常运行和事故时释放到包容箱和工艺间的量和处理措施,最后评价了氦-3系统发生不同安全措施失效的事故情况下工作人员的氚内照射剂量.结果表明:系统正常运行时工作人员所受最大剂量为1.27μSv...     

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.     

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.     

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.     

Main results of the analyses related to a few Reference Accident Sequences of NET/ITER tritium systems

Based on the existing conceptual design, a set of Reference Accident Sequences (RAS) has been defined by tritium system experts. This paper presents the main results related to a preliminary analysis of the following RAS for fuel cycle systems: hydrogen detonation in the isotopic separation system, rupture of a cryoline in a cryopump, fire in the long-term storage beds, and mechanical faulty operation in the pellet injector. The systems and the component mode of operation have been analyzed, as well as the characteristics of all streams, the inventory of T, D, and H, and the energy released under accident conditions. Failures that could give rise to the release of tritium and/or to the formation and ignition of oxygen/hydrogen mixtures and failures that could lead to containment pressure transients are discussed and examined.     

ITER safety studies: The effect of two simultaneous perturbations during a loss of plasma control transient

The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. Conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer are used to determine the effects of such transients on wall integrity from a thermal point of view.In this contribution, progress in a “two simultaneous perturbations over plasma” approach to the analysis of the loss of plasma control transients in ITER is presented. The effect of variation in confinement time is now considered, and the consequences of this variation are shown over a n–T diagram. The study has been done with the aid of AINA 3.0 code. This code implements the same 0D plasma-1D wall scheme used in previous LOPC studies.The rationale of this study is that, once the occurrence of a loss of plasma transient has been assumed, and due to the uncertainties in plasma physics, it does not seem so unlikely to assume the possibility of finding a new confinement mode during the transient.The cases selected are intended to answer to the question “what would happen if an unexpected change in plasma confinement conditions takes place during a loss of plasma control transient due to a simultaneous malfunction of heating, or fuelling systems?”Even taking into account the simple models used and the uncertainties in plasma physics and design data, the results obtained show that the methodology used in previous analyses could probably be improved from the point of view of safety.     

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.     

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.     

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.     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

In-port-plug transmission line design of the ITER plasma position reflectometer

This work describes the microwave design of the transmission line housed in the in-port-plug region of the ITER plasma position reflectometer(PPR). The design of the components of the inport-plug reflectometers(located in equatorial port-plug 10(EPP10) and in upper-port-plug 01(UPP01)) is presented. Using a 3 D ray tracing code, the spatial position and optimum orientation angles of each set of emission and detection antennas were determined. A feasible path was then created from the obtained antenna positions and orientations to the primary vacuum window.Oversized tall waveguides were chosen to reduce ohmic losses. Due to space constraints in the ITER crowded environment, bends in oversized waveguides were unavoidable, and thus mode conversion was produced. To keep mode conversion losses at bay, hyperbolic secant curvature bends had to be used whenever possible. However, E-plane bends in tall waveguides proved to be especially critical, making it necessary to employ other approaches when higher bending angles were needed. Mode conversion results were obtained by evaluating the mode coupling equations. Ohmic losses have also been computed and their results compared with commercial simulators, obtaining a perfect agreement.     

ITER中国液态锂铅实验包层模块第一壁等离子体注入氚滞留分析

对等离子体注入ITER中国液态锂铅实验包层模块第一壁滞留的氚进行了分析,考虑了第一壁温度梯度、材料表面清洁度、加挂Be瓦及结构材料内缺陷等因素对氚滞留量的影响。分析结果显示,滞留的氚主要存在于中子辐照引起的缺陷内;氚滞留量对第一壁面向等离子体侧的清洁度及加挂Be瓦很敏感;总的氚滞留量约0.58 mg,不会对ITER真空室内氚滞留造成显著影响。     

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.     

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.     

ITER实验包层吹洗气前处理系统的FMECA研究

实验包层(TBM)输出吹洗气前处理系统将安装在国际热核聚变实验堆(ITER)装卸TBM的通道内(Port Cell),它的功能是将TBM输出的含氚吹洗气进行过滤、除HTO、冷却、调流量等处理,处理后输出到氚提取系统。介绍了该系统的工艺过程和系统组件,以氚释放危险为判据,对该系统进行FMECA(故障模式、影响及危害性分析),并作出分析表。找出了几种故障模式或薄弱环节,进行了尝试性的风险优先数和故障模式危害度计算,提出了设计改进措施和使用补偿措施;最后确定了需要重点关注的4种需导致释非正常过量释放的潜在故障模式。这些故障分析为降低系统氚过量释放危险设计提供了依据,也为TBM其他附属氚系统的安全分析奠定了基础。     

Fusion safety and environmental R&D tasks in Japan

This paper aims at listing and evaluating the status of all the research and development (R&D) tasks necessary for the construction of a safe and environmentally benign fusion experimental reactor. At this time, it is not possible to define precisely the R&D tasks necessary for the licensing approval and those that are useful in improving safety but not necessarily required for licensing because the licensing procedure itself is still being discussed. Among the R&D tasks, the most important are considered to be those related to tritium safety, namely, those effective in reducing the uncertainty in tritium inventory in the plasma facing components and blanket, uncertainty in tritium permeation and leakage, and those to clarify tritium behavior in the containment and in the environment. The R&D tasks with second priority are judged to be those related to mobilization of the activation products such as activated erosion dust or the corrosion products. The volatilization of structural metal caused by the oxidation at high temperature seems to be highly unlikely but some experiments are needed to assure that this is the case.