Songlin Liu

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

ABSTRACT 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.

Preliminary safety analysis for the Chinese ITER Dual Functional Lithium–Lead Test Blanket Module

Safety analysis is part of the ITER Test Blanket Module (TBM) design process ensuring that the TBM does not adversely affect the safety of ITER. To get the licence for TBM as a whole with ITER, relevant safety analysis is required for each TBM system proposed by each party. The safety analysis for the Chinese Dual Functional Lithium–Lead Test Blanket Module (DFLL-TBM) has been performed based on the latest DFLL-TBM design. In this paper, the following safety considerations, such as source terms, operational releases, accident sequence analyses and waste assessment, were analysed. Both the deterministic approach and the complementary systematic approach starting with failure mode and effects analysis studies were adopted in the accidental analysis. The preliminary results showed that the DFLL-TBM system at normal operating conditions and under accident scenarios did not add additional safety hazards to the ITER machine and could meet the ITER safety requirements and additional safety requirements for the TBM system.

Evaluation of ACPs in China Fusion Engineering Test Reactor Using CATE 2.1 Code

Activated corrosion products (ACPs) are the dominant radiation hazard in water-cooled fusion reactor under normal operation conditions and directly determine the occupational radiation exposure during operation and maintenance. Recently, the preliminary design of China Fusion Engineering Test Reactor (CFETR) has been just completed. Evaluation of ACPs is an important work for the safety of CFETR. In this paper, the ACPs analysis code CATE 2.1 was used to simulate the spatial distribution of ACPs along the blanket cooling loop of CFETR, in which the influence of adopting different pulse handling methods was researched. At last, the dose rate caused by ACPs around the blanket cooling loop was calculated using the point kernel code ARShield. The results showed that the dose rate under normal operation for 1.2 years at contact is 1.02 mSv/h and at 1 m away from pipe is 0.45 mSv/h. And after shutting down the reactor, there will be a rapid decrease of dose rate, because of the rapid decay of short-lived ACPs.

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

ABSTRACT 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 obstacle position and number on the overpressure of hydrogen combustion in a semi-confined compartment

ABSTRACT During a severe accident in a pressurized water reactor, large quantities of hydrogen can be produced and released into the containment. The hydrogen mixed with air may be ignited. Rapid pressure rise is likely to occur during hydrogen combustion in a semi-confined compartment because the length–diameter ratio of the compartment is relatively large and there generally are some devices and components in the compartment. Obstacle position and number have a significant and complex effect on combustion pressure. In this paper, a numerical analysis of the effects of obstacle position and number on the overpressure of hydrogen combustion in a semi-confined compartment was carried out to identify the arrangement of obstacles that can decrease the pressure loads generated by hydrogen combustion. The overpressure in the compartment with a single obstacle was found first increases to a maximum value and then decreases as the obstacle moving from the closed end to the open end. When there are multiple obstacles existing in the compartment, minimize the obstacle spacing and number is likely to reduce the maximum overpressure of hydrogen combustion. In addition, intentionally adjust the length of obstacle array may help for reducing the maximum overpressure.

The Experimental Consideration for TBM Mock-Up Effect on Plasma Performance Based on MAPES Platform in EAST

EAST can provide better opportunities to contribute development of ITER-relevant plasma physics and engineering because it has ITER-like configuration, and has achieved 10s H-mode plasma, and aims steady-state operation of DD high performance plasma. The impact of Test blanket module (TBM) using RAFM (reduced activation ferritic/martensitic) steels on tokomak plasma is a major concern in ITER operations. In order to assess this effect due to TBM local ripple, an experiment plan of TBM mockup using RAFM steel is being planned on MAPES (Material and Plasma Evaluation System) in EAST. This paper reports experimental consideration on MAPES based on magnetic analysis and ripple calculation at separatrix point. The relevant experiments strategy and plan in EAST are also proposed.Copyright

The Thermal-Mechanical Analysis of the BIT HCCB Blanket for CFETR

CFETR is an “ITER-like” China Fusion Engineering Test Reactor. The BIT (Breeder insider tube) HCCB (Helium cooled ceramic breeder) blanket has been designed as one option for CFETR. Its unique feature is that a BIT structure is adopted for the blanket tritium breeder unit. The breeder unit is an assembly of three coaxial coil-pipes, internal coil-pipe is filled with the tritium-breeding material, and the coaxial coil-pipe is embeded in the beryllium pebble. In order assess the feasibility of the BIT HCCB blanket design, a 3D finite element model slice of the BIT HCCB blanket is developed, and the thermo-hydraulic calculation was performed, and thermo-mechanical computations were carried out. And then the primary stress and secondary stress were evaluated according to Structural Design Criteria for ITER In-vessel Components (SDC-IC). This paper presents relevant analyses and results. The preliminary results shows current design satisfies with allowable limit of material and requirement of SDC-IC code.© 2013 ASME

Preliminary neutronics design and analysis of the BIT helium cooling ceramics blanket for CFETR

CFETR is a Tokamak fusion engineering test reactor whose concept design is being developed in China. It is a key issue for breeding blanket design to attain tritium self-sufficiency as one of important missions of CFETR. This paper presents a preliminary neutronics design and analysis employing a BIT (breeder inside tube) type helium cooling ceramics blanket (HCCB) design concept as one of CFETR blanket design candidates. Firstly, 1D reactor model was designed using ceramic breeder Li4SiO4 and beryllium in pebble for multiplier. The primary blanket parameters were optimized to yield the higher tritium breeder ratio (TBR), including the thickness of outboard breeder blanket, enrichment of Li-6 and ratio of Li4SiO4 to Be. Secondly, based on the optimized blanket parameters and plasma parameters, a detailed 3D neutronics calculation model of 22.5° reactor sector was developed, including blanket modules, shield, divertor, vacuum vessel and TF coil. The gap between blanket modules had been taken into account. Finally, a set of nuclear analyses were carried out addressing the key neutronics issues by Monte Carlo neutron-photon transport code MCNP version 5 and the FENDL-2.1 data library.The preliminary analysis results showed that the global TBR could achieve 1.21 which satisfied the tritium self-sufficiency demand. Nuclear heat, neutronic flux, and distribution of neutron wall loading (NWL) were also analyzed as source terms of the blanket thermal-hydraulics design and reactor nuclear response.Copyright

Numerical analysis of forced convection heat transfer in first wall rib-roughened channels for liquid lithium lead blanket

The first wall as the key component of the liquid metal blanket must withstand and remove the maximum heat flux from the plasma chamber and high power density LiPb breeder zone. A suitable countermeasure would be artificial roughening of the coolant channel wall facing the plasma for intensifying heat exchange. The rib-roughened coolant channels with high pressure helium gas for FDS series liquid lithium lead blanket were designed to enhance the turbulence heat transfer in comparison with smooth coolant channels. Heat transfer coefficients and friction factors in rib-roughened coolant channels of the first wall are investigated using the computational fluid dynamics code FLUENT??based on the two dimensional physical model and periodic condition. From the results of the numerical analysis, the turbulence heat transfer Nusselt number (Nu) in the rectangular channels of rib-roughened face is higher 200% than those in the smooth channels, and the fraction coefficient (ƒ) between the helium gas and CLAM steel structure is 20% more than those in the smooth channels. The numerically optimized parameters are also presented considering different convexity part height and width as well.