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The development of CFETR integration design platform is in progress to accelerate the CFETR design in consideration of both core physics and engineering practicability. Within the engineering framework of CFETR integration design platform, the neutronics interface module has been implemented. This work presents the parameter study on helium cooled ceramic breeder blanket neutronics with the neutronics interface module of the CFETR integration design platform. The parameterized 3-D neutronic model of CFETR was set in the first place. Afterwards, the workflow of neutronic analyses was established. The concerned geometry and composition parameters of breeder blanket were selected as input and the tritium breeding ratio was selected as target. And, then, the workflow was abstracted as the mathematical model in the form of the so called response surface based on sufficient MCNP simulations. With the workflow or response surface model, the estimation and optimization of target parameter could be achieved. Finally, a set of local optimal parameters was proposed for CFETR helium cooled ceramic breeder blanket. [ABSTRACT FROM AUTHOR]

• The detailed design of the back supporting structure (BSS) of CFETR WCCB blanket has been presented. • Heat transfer Thermal-hydraulic analysis of the BSS has been carried out based on a global model of BSS and the attachment. • Mechanical assessment of the BSS and the attachment has been performed and results can meet the design requirements. China Fusion Engineering Test Reactor (CFETR) is a key step to realize the future demonstration fusion power plant and plays an important role in the roadmap of magnetic confinement fusion development in China. As an important in-vessel component, CFETR blanket back supporting structure (BSS) bridges the blanket modules onto the vacuum vessel, withstanding the gravity and electromagnetic loads and distributing the coolant into blanket modules. This paper presents the preliminary design of BSS for the CFETR WCCB blanket and the detailed design of the attachments between BSS and Vacuum Vessel. Based on the proposed design, a global model of BSS and the attachment for one sector has been established with ANSYS code. The thermal-hydraulic analysis has been conducted to assess the heat exchange and temperature distribution. The analyses of individual load conditions and the coupling analysis have been carried out to obtain the stress distribution and deformation. Based on the analyses results, the corresponding design targets and criterion have been met, indicating the feasibility of current BSS and the attachment design. The design and analyses of the BSS together with the attachments can not only provide good reference for the VV design but also provide guidance for design of flexible support used in the blanket pipe system. [ABSTRACT FROM AUTHOR]

The China Fusion Engineering Test Reactor (CFETR) has been designed as steady-state operated tokamak device. The CFETR new engineering design has been initialized corresponding to the latest core parameters (major/minor radius is R = 7.2 m/a = 2.2 m). The nuclear analysis of CFETR was carried out to evaluate the neutron shielding capabilities and nuclear responses to toroidal field coils (TFC). The 3D neutronics model of CFETR was built with the support of McCad tool to convert the CAD model automatically. An extended variance reduction (VR) technique of combined MAGIC GVR with density reduction method has been involved to generate suitable weight windows to enhance the efficiency of particle transport. Highly detailed neutron and photon map has been obtained which meet the requirements for Monte Carlo nuclear analyses. The global neutron flux distribution of 3D CFETR with P f = 200 MW to 1.5 GW was obtained to evaluate its shielding performance. The nuclear responses to TFC such as the nuclear heating and fast neutron fluence have been evaluated. Initial results show the shielding of current divertor design is not sufficient and the optimizing work has been done. The final results satisfy the referred ITER design limit and it can provide the support for the future more detailed design of divertor and other shielding systems. [ABSTRACT FROM AUTHOR]

• Investigate and evaluate the ripple effect of WCCB blanket on TF ripple in CFETR. • CFETR TF ripple generated by WCCB blanket has been calculated. • Potential methods for reducing the TF ripple are studied in this paper. China Fusion Engineering Test Reactor (CFETR), the next fusion device in China, is proposed to bridge the gaps between ITER and the demonstration reactor (DEMO). The Water Cooled Ceramic Breeder (WCCB) blanket, one candidate option of breeding blanket conception for CFETR, impacts the toroidal field (TF) ripple in CFETR, as the reduced activation ferritic/martensitic (RAFM) steel was used as its structure material. The purpose of this paper is to investigate and evaluate the ripple effect of WCCB blanket on TF ripple in CFETR. An analysis is performed using three-dimensional finite element model with 16 TF coils and WCCB blanket modules. ANSYS code is employed for this study. The results indicated that the TF ripple was less than 0.5% at all points in target separatrix when it was estimated with TF coils only. We also calculated TF ripple in case of WCCB blanket modules and various port configuration schemes. It is found that TF ripple cannot meet the design requirements by the presence of WCCB blanket modules, even in difference port configuration. Discussions are made on the implication of the WCCB blanket modules and ports on TF ripple. Potential methods for reducing the TF ripple are studied in this paper. [ABSTRACT FROM AUTHOR]

• The integration design platform (CIDP) is developed for the CFETR design. • The data and task management system are developed for the design data and work management. • The integrated design framework is developed to provide the unified design environment. • The GPU-based design cloud is built to provide the virtual workstation for geographically-distributed designers. The design of the China Fusion Engineering Test Reactor (CFETR) involves a complex system structure and intricate constraints, and the design work will be performed by a great number of geographically distributed groups. To support consistent design work effectively, the CFETR Integration Design Platform (CIDP) is being developed [IEEE Trans. Plasma Sci. 45 (2017) 512; Fusion Eng. Des. 123 (2017) 87]. CIDP is envisioned as a design center, where all design work can be performed and related design data and documents can be managed. The functions of CIDP are realized by hardware supporting a GPU-based design cloud and three software systems, which are a data management system for data and document management, a task management system for coordinating the design work and an integrated design framework to provide the unified environment for the implementation of detailed design work. The architecture and the detailed implementation of CIDP are given in this article. [ABSTRACT FROM AUTHOR]

The China Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak, and one mission is to demonstrate a full cycle of self-sustaining tritium. In this study, in order to validate the electromagnetic (EM) load distribution on the helium-cooled ceramic breeder (HCCB) blanket, a three-dimensional model of the CFETR sector was developed in the finite element software ANSYS. The system was analyzed considering a plasma major disruption scenario with an exponent current quench of 36 ms. The EM forces and the moment distribution on the blanket module as well as on the segment were calculated. Then, the EM loads were used as the input for the structural analysis of the HCCB blanket. Moreover, the results were compared to those of linear analyses to underline the effect of the non-linearity of the ferromagnetic materials on the EM load distribution of the HCCB blanket. These results indicated that the maximum equivalent stress and displacement satisfied the requirements of structural safety design. With respect to the design and optimization of CFETR, this investigation will provide technical reference and data support for the future design and optimization of the HCCB blanket. Highlights • A FE model of the CFETR sector was developed in the ANSYS to validate the HCCB blanket EM load. • The mechanical design of the HCCB blanket was analyzed considering a plasma major disruption. • The effect of the ferromagnetic materials on the EM load distribution of the HCCB blanket was studied. [ABSTRACT FROM AUTHOR]

Highlights • An integrated analysis of blanket and EDPP is carried out with the conventional FE method code (ANSYS). • The EDPP has larger influence on the EM loads distribution of the IB blanket compared to the OB blanket. • A diagnostic port plug with small front openings could be useful to reduce its effect on the EM loads of the blanket system. Abstract The Chinese Fusion Engineering Test Reactor (CFETR) is the next step in China's research into fusion energy. In phase I, the preliminarily design of diagnostic systems for the CFETR was completed; however, the material and spatial position of the diagnostic systems severely affect EM loads distribution in blanket modules. In order to evaluate the effect of a diagnostic port plug on the EM loads distribution in the blanket system of CFETR, a finite element (FE) model with an equatorial diagnostic port plug (EDPP) was established using ANSYS. The EM forces and moments are calculated and discussed, followed by a comparison of EM loads in the helium-cooled ceramic breeder (HCCB) blanket to verify the EM effect of the EDPP. Simulation results show that the EM loads in the inboard (IB) and outboard (OB) blanket modules and segments were changed with the EDPP, but the direction of the main loads on the blanket system did not change. The results also show that the EDPP has a greater influence on the EM loads of the IB blanket compared to the OB blanket, meaning it is necessary to take into account the effect of the EDPP during the evaluation of the mechanical design of the blanket system in further studies. The investigation presented here may also serve as a technical reference for the design and optimization of diagnostic port plugs in the future. [ABSTRACT FROM AUTHOR]

Chinese Fusion Engineering Testing Reactor (CFETR) is aimed to obtain the technologies to fill the gaps between ITER and DEMO. The helium cooled ceramic breeder (HCCB) blanket is one of the candidates for CFETR. Ceramics Li 4 SiO 4 , beryllium and helium of 8 MPa were selected as tritium breeding material, neutron multiplier and coolant, respectively. CLF steel, one of reduced activation ferritic/martensitic (RAFM) steels, was applied as structural materials of blanket. Neutron and gamma transport was simulated with MCNP code and FENDL/MC-2.1 neutron cross section data. Activation calculation was performed with FISPACT-2007 code and EAF-2007 activation file. In this study, neutronics characteristic of HCCB blanket such as tritium breeding ratio (TBR) and shielding were assessed. The arrangement of Be, Li 4 SiO 4 and cooling plate in blanket module was optimized to improve TBR. The sensitive study for First Wall on TBR was also performed. The radioactivity inventories of HCCB blanket were estimated. The dominant radionuclides in structural materials, Li 4 SiO 4 and Beryllium were discussed, respectively. The damage of plasma facing components such as First Wall and divertor due to DT neutron irradiation was analyzed in terms of DPA value and gaseous product. [ABSTRACT FROM AUTHOR]

• 3D full neutronic model of CFETR ECRH was built to deliver the nuclear analysis. • The maximum neutron damage on ECRH antenna appear at the steering mirrors. • Irradiation of VV and TFC can meet the requirements after shielding optimization. • TBR of the whole reactor reduced to 1.137 after 3 ECRH antenna introduced. Electron Cyclotron Resonance Heating (ECRH) is one of the auxiliary heating systems designed for CFETR. The antenna of ECRH system is planned to be plugged into CFETR through 3 upper ports and 2 blanket modules of each port will be penetrated for the transmission of wave. In this study, three dimensional neutronic analysis has been performed to evaluate the shielding effect and conduct the optimization of ECRH antenna design. The simulation has been done on the integration of ECRH antenna system with 22.5-degree CFETR model. The particle transport has been done by MCNP with the support of FENDL-2.1 nuclear data library. An accurate neutron flux distribution has been obtained by employing the 'on-the-fly' global variance reduction technique. The neutron damage of TFC, VV and port was analysed to assess the shielding performance of ECRH antenna system. Initial results demonstrated that the fast neutron fluence and nuclear heating density of TFC and helium production rate of VV port were beyond the design limit. By adding shielding blocks, these parameters didn't exceed the limits. The irradiation damage on antenna components were calculated for estimating its life after neutron irradiation and providing the support of cooling system design. The TBR decreased due to the blanket opening from original 1.149 to 1.137 in the whole reactor. [ABSTRACT FROM AUTHOR]

• Preliminary thermal-hydraulic analyses have been conducted to evaluate the design feasibility. • The cooling configuration has been optimized referring to preliminary flow distribution results. • Further detailed thermal-hydraulic behavior has been obtained based on a global module FE model with an acceptable result. • Thermo-mechanical performance of the WCCB blanket has been investigated under normal operation and two accident events. The blanket system is a key component to fulfill tritium breeding and nuclear energy conversion in the fusion reactor. This paper reports and presents the design and analysis of an equatorial inboard water-cooled ceramic breeder (WCCB) blanket module for Chinese Fusion Engineering Test Reactor (CFETR). The detailed structure design and the schematics of coolant flow scheme optimization has been presented. Based on the design model, a global model of the equatorial inboard WCCB blanket module has been established by the ANSYS code. Preliminary thermal-hydraulic performance of the blanket module is investigated which suggests that the cooling system inside the blanket module should be optimized. Then, optimization design has been done, the analysis results shows that the temperature distribution is reasonable and within the allowable temperature. Further thorough thermal-hydraulic and mechanical analysis have been studied to validate the feasibility of the design. Three scenarios (normal operation/LOCA occurs during plasma startup/LOCA occurs during normal operation) were postulated to evaluate the design rationality, and the results were assessed with ITER SDC-IC. Optimization results imply that the design of the equatorial inboard WCCB blanket module can meet the design requirements. [ABSTRACT FROM AUTHOR]

• 3D full neutronic model of CFETR ECRH was built to deliver the nuclear analysis • Operation dose of CFETR in the building above ECRH port is about 100 μSv/h • Shutdown dose rate at position C of ECRH is about 160 μSv/h 1 month after shutdown Electron Cyclotron Resonance Heating (ECRH) is one of auxiliary heating systems to be installed in CFETR. For heating purpose of ECRH, a penetration space has been designed to transmit the wave to plasma. This will significantly increase the neutrons streaming through the wave channel and result in heavy irradiation of ECRH port structures. For nuclear safety analysis, the irradiation dose during operation of CFETR should be limited according to the regulation of radiological zone. In case of failure, the ECRH is planned to be maintained by workers after shutdown and waiting time of one month until the dose rate decreases to low level of 100 μSv/h. The shielding capabilities of ECRH port has been evaluated in view of dose effect both during operation and after shutdown of CFETR. For that purpose, a 3-D neutronics model has been built to launch the nuclear analysis. The operation dose rate was directly transferred from the neutron and photon flux by using of ICRP-74 flux to dose conversion factors. The shutdown dose rate was obtained with NASCA code system by coupling the Monte Carlo particle transport and FISPACT activation calculation. The obtained operation and shutdown dose data can be provided for the shielding optimization of ECRH wave system. [ABSTRACT FROM AUTHOR]