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.     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.     

Stability analysis of Alfvén eigenmodes in China Fusion Engineering Test Reactor fully non-inductive and hybrid mode scenarios

In this paper, NOVA/NOVA-K codes are used to investigate the stability of Alfvén eigenmodes(AEs) in the China Fusion Engineering Test Reactor(CFETR). Firstly, the stability of AEs excited by energetic alpha particles is investigated. For the fully non-inductive scenario, it is found that all AEs are stable, and the least stable toroidal mode number is n= 8. However, for the hybrid mode scenario, it is found that many AEs are unstable, and the least stable toroidal mode numbers are n= 7, 8. Secondly, the effect of energetic alpha-particle parameters and beam ions on AE stability is also presented. The threshold of the least stable AE is about β_(crit,α) = 1.12%,crit,less than the value of alpha-particle beta(β_α=1.34%). The result demonstrates that the AEs excited by alpha particles are weakly unstable. The effect of the beam ions on AE stability is found to be very weak in CFETR.     

A new method for improving the tritium breeding and releasing performance of China Fusion Engineering Test Reactor phase II helium-cooled ceramic breeder blanket

China Fusion Engineering Test Reactor (CFETR) is a brand new Tokamak reactor which is currently being designed to fill the intervals between ITER and future DEMO fusion reactor. It has two operation phases: phase I with P_f = 200 MW is to demonstrate steady-state operation; phase II with P_f = 1000 MW to validate DEMO technology. Helium-cooled ceramic breeder (HCCB) blanket is one of the candidate blanket concepts for CFETR. Until now, there are many research institutes which have performed conceptual designs and comprehensive analyses works for phase I HCCB blanket in detail. However, lately, the operational stage of CFETR has transformed from phase I to phase II, and the latest core design parameters have been just determined (major radius equals to 7.2 m, minor radius equals to 2.2 m). Therefore, the design and analyses work for HCCB blanket should also be initialized. In this work, based on the original integrated optimization method, the radial structure layout optimization of the outboard equatorial phase II HCCB blanket module is conducted by NTCOC. However, the calculation results show that the original optimization method provides inadequately optimized tritium breeding performance and insufficient tritium releasing ability under the condition of high fusion power. To solve these problems, a dynamic feedback is incorporated into the original optimization method for the first time, and then the calculation results show that the problem has been solved well after revision. This work can supply worthy guidance and reference for the conceptual design and comprehensive analyses of CFETR phase II HCCB blanket.     

Thermal hydraulic analysis of the HECLIC blanket breeder unit for CFETR

China Fusion Engineering Test Reactor (CFETR) is proposed by the China National Integration Design Group. A helium‐cooled lithium ceramic (HECLIC) blanket for CFETR has been designed by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and University of Science and Technology of China (USTC). The HECLIC blanket includes the tungsten armor, first wall (FW), breeder units (BUs), caps, stiffening grids and back plates. The BU consists of the beryllium pebble beds, lithium ceramic pebble beds and cooling plates. The stiffening grid reinforces the blanket structure and also cools the BU together with cooling plates. Thermal hydraulic analysis has been performed to assess the heat transfer coefficient (HTC) and temperature distribution. The results indicate that the velocity is fairly uniform in the cooling loops, and the HTC is high enough to remove the heat timely. The maximum temperatures are within the design temperature limits. Besides, optimization has been done to improve the layout of the cooling channels. In addition, thermal mechanical analysis has been carried out. The maximum stress can satisfy design limits, and it proves the feasibility of the design. Copyright © 2014 John Wiley & Sons, Ltd.     

Modeling and multi-pipe manifolds optimization of the WCCB blanket sector for CFETR

In this paper, one standard water cooled ceramic breeder blanket sector has been modeled for the Chinese fusion engineering test reactor using RELAP5/MOD3.3 with details of anisotropic structures, positions and nuclear heat of the blanket modules. The multi-pipe manifolds of the current sector design scheme has been designed and analyzed. And an optimized scheme was proposed to further reduce the pressure drop, uniform the flow distribution, and prevent overheating. Also the fusion power excursion transients were simulated to evaluate the system heat removal and recovery ability. The results indicated that high-transient heat flux up to 0.8 MW/m2 can cause sub-cooled boiling of the coolant in the first wall area of certain modules. Coolant returns to single phase soon after the end of the transient. According to the analysis, it is suggested that the blanket modules surrounding plasma have as similar structure design features as possible and sizes of the modules should be kept relatively small so as to obtain a reasonable pressure drop.     

Effects of kinetic profiles on neutron wall loading distribution in CFETR

The China Fusion Engineering Test Reactor (CFETR) is under design,which aims to bridge the gaps between ITER and the future fusion power plant.The neutron wall loading (NWL) depends on the neutron source distribution,which depends on the density and temperature profiles.In this paper,we calculate the NWL of CFETR and study the effects of density and temperature profiles on the NWL distribution along the first wall.Our calculations show that for a 200 MW fusion power,the maximum NWL is at the outer midplane and the vaule is about 0.4 MW m-2.The density and temperature profiles have little effect on the NWL distribution.The value of NWL is determined by the total fusion power.     

Simulations of energetic alpha particle loss in the presence of toroidal field ripple in the CFETR tokamak

Energetic alpha particle losses with the toroidal field ripple and the Coulomb collision in the CFETR tokamak have been simulated by using the orbit-following code GYCAVA for the steady-state and hybrid scenarios. The effects of the outer boundary and the ripple amplitude on alpha particle losses have been investigated. The loss fractions and heat loads of alpha particles in the hybrid scenario are much smaller than those in the steady-state scenario for a significant ripple amplitude. Some alpha particles in the plasma core are lost due to the ripple stochastic transport for a large ripple amplitude parameter. The heat loads with the last closed flux surface boundary are different from those with the wall boundary for the CFETR tokamak, which can be explained by typical alpha particle orbits. Discrete heat load spots have been observed in alpha particle loss simulations, which is due to the ripple well loss. The transition of the lost alpha particle behavior from the ripple stochastic diffusion to the ripple well trapping has been identified in our CFETR simulations. The Coulomb collision effect is responsible for this transition.