RAMI analysis for ITER fuel cycle system

ITER is the first fusion device designed to continuously process DT (Deuterium–Tritium) plasma exhaust and supply recycled fuel in a closed loop, using a FCS (Fuel Cycle System) which includes tritium plant, vacuum system, and fuelling and wall conditioning systems. In addition, as an important step towards the commercially viable power plant, ITER has an ambitious inherent availability target of 60% for plasma production. To be able to achieve this objective, the potential technical risk which may impact the machine operation was assessed by means of RAMI (Reliability, Availability, Maintainability and Inspectability).Firstly, a functional breakdown of the FCS was performed and critical components associated with its basic functions were identified. Secondly, FMECA (Failure Mode, Effects and Criticality Analysis) was performed to evaluate potential causes of failures and their consequences for the plasma operation. The analysis highlighted two critical functions in the tritium plant, namely storage and distribution performed by SDS (Storage and Distribution System) and isotopic separation performed by ISS (Isotope Separation System). Since the system involves various active components, inherent availability of the system has been finally estimated to be around 74% in DT phase.     

RAMI analysis of the ITER CIS

ITER is the first worldwide international project aiming to design a device that proves the physics and technological basis for fusion power plants to produce nuclear fusion energy. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to guide the design of components in preparation for operation and maintenance. RAMI analysis of the ITER central interlock system (CIS), which shall provide investment protection for the ITER systems was performed on the conceptual design. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 5 main functions and 7 sub-functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to estimate the reliability and availability of each function under stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CIS expected after implementation of mitigating actions was calculated to be 99.86% over 2 years, which is the typical interval of the scheduled maintenance cycles. A failure modes, effects and criticality analysis (FMECA) was performed to initiate risk mitigation action. Criticality matrices highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. It was assessed that the availability of the ITER CIS, with appropriate mitigating actions applied, meets the project availability requirement for the system.     

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.     

Preliminary Shielding Analysis for HCCB TBM Transport

A preliminary shielding analysis on the transport of the Chinese helium cooled ce?ramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during trans?port. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package con?taining low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future.     

Neutronic Calculation Analysis for CN HCCB TBM-Set

Using the Monte Carlo transport code MCNP.neutronic calculation analysis for China helium cooled ceramic breeder test blanket module(CN HCCB TBM) and the associated shield block(together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model.Key nuclear responses of HCCB TBM-set.such as the neutron flux,tritium production rate,nuclear heating and radiation damage,have been obtained and discussed.These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set,such as thermal-hydraulics,thermal-mechanics and safety analysis.     

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.     

Mechanical Analysis for ITER Lower CC Feeder

ITER correction coils (CCs) feeder is the important component of ITER feeder systems to supply the cryogens and electrical power for CCs. They should withstand the huge electromagnetic (EM) force and high thermal shrinkage. Considering the EM and thermal loads, mechanical analysis is performed to qualify the structural strength of the lower CC feeder. Results show that containment duct and cryopipe can meet the static criteria but busbar jacket cannot meet. It is proposed that more supports should be added at the corners for the busbar. Basically, the lower CC feeder design is valid and feasible.     

ITER校正场磁体馈线结构分析

基于ANSYS对ITER校正场磁体馈线的结构进行了分析。根据馈线结构特点,对有限元模型进行了简化。通过电磁分析,获取了超导母线的电磁力,研究了电磁力与超导母线支撑间距的关系,并根据结果提出了推荐的支撑间距。在进行馈线模型结构分析时,电磁力按支反力的形式施加,并施加不同的载荷工况,获取了不同工况下馈线的应力及变形数据。分析结果显示馈线结构设计是合适的,满足设计应力准则要求。     

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 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.     

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.     

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.     

Primary Design and Analysis of Feeder for ITER Poloidal Field

An electromagnetic （EM） analytic model for the PF feeder, applied to ITER and needed to convey the cryogenic supply and electrical power to the PF magnets, was ...     

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific and technological goal. The reactor has the equivalent scale compared with ITER, but has the complementary function to ITER. CFETR is a demonstration of long pulse or steady-state operation with duty cycle time not less than 0.3–0.5 and the full cycle of tritium self-sustained with TBR not less than 1.2. At the same time it will be exploring options for DEMO blanket and divertor with an easy changeable core by remote handling way. To be able to reach its scientific and technological objectives, as one of technical risks control methods, RAMI analysis need to be done during the hold lifetime of CFETR, from conception design to decommissioning. Base on stating of CFETR lifetime and preliminary operational programme, the RAMI analysis program and process are designed and discussed, it consists of five major steps: (1) functional analysis are performed, (2) calculating reliability block diagrams, (3) analyzing failure mode, effects and criticality analysis, (4) risk mitigation actions are taken to ensure every system is compatibility with RAMI objectives, (5) All the RAMI analysis are integrated as the final RAMI analysis reports to be reviewed in the system final design review. Along with the elements of the analysis the vacuum vessel (VV) system was performed to provide as examples, detailed showing how the CFETR RAMI analysis is carried out. CFETR RAMI analysis guidelines were designed and established, after constantly revised and improved these analysis criteria and programs will become the basis standards for CFETR RAMI analysis. Preliminary RAMI analysis of CFETR VV system was obtained, which will be updated with the VV system design progresses.     

Neutronic investigation and activation calculation for CFETR HCCB blankets

The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder(HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor(CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio(TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil.The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1?×?10-4 k W, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.