Study on Multiphysics Coupling and Automatic Neutronic Optimization for Solid Tritium Breeding Blanket of Fusion Reactor

A tritium breeding blanket (TBB) is an essential component in a fusion reactor, which has functions of tritium breeding, energy generation and neutron shielding. Tritium breeding ratio (TBR) is a key parameter to evaluate whether the TBB could produce enough tritium to achieve tritium self-sufficiency (TBR > 1) for a fusion reactor. Current codes or software struggle to meet the requirements of high efficiency and high automation for neutronic optimization of the TBB. In this paper, the multiphysics coupling and automatic neutronic optimization method study for a solid breeder TBB is performed, and a corresponding code is developed. A typical module of China fusion engineering test reactor (CFETR) helium cooled ceramic breeder (HCCB) TBB was selected, and a 3D neutronics model of an initial scheme is developed. The automatic neutronic optimization was performed by using the developed code for verification. Results indicate that the TBR could increase from 1.219 to 1.282 (~5.17% improvement), and that the maximum temperature of each type of material in the optimized scheme is below the allowable temperature. It is of great scientific significance and engineering value to explore and study the algorithm for automatic neutronic optimization and the code development of the TBB.

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Neutronics Perturbation Calculation Method Study of Solid Breeder Tritium Breeding Blanket for TBR Enhancement

Frontiers in Energy Research ◽

10.3389/fenrg.2021.777286 ◽

2021 ◽

Vol 9 ◽

Author(s):

Shen Qu ◽

Qixiang Cao ◽

Fengchao Zhao ◽

Xueren Wang ◽

Xuru Duan ◽

...

Keyword(s):

High Efficiency ◽

Fusion Reactor ◽

Density Perturbation ◽

Tritium Breeding ◽

Comparison Analysis ◽

Perturbation Calculation ◽

Neutron Shielding ◽

Self Sufficiency ◽

Breeding Ratio ◽

Key Parameter

Tritium breeding blanket (TBB) is an essential component in a fusion reactor, which has functions of tritium breeding, energy generation, and neutron shielding. Tritium breeding ratio (TBR) is a key parameter to evaluate whether the TBB could produce enough tritium to achieve the tritium self-sufficiency (TBR >1) for fusion reactor. Current codes or software are hard to meet the requirements of high efficiency, high resolution, and high automation for neutronic optimization of TBB. In this article, the application of the density perturbation calculation on a solid breeder TBB was first performed. Then, the method of the geometry perturbation calculation based on the virtual density theory was studied. Results and comparison analysis indicate that the 1st + 2nd-order neutronic perturbation calculations (including the density perturbation and the geometry perturbation) results are consistent with the transport results under a perturbation of −15% to +15%. It is proven to be valid to use the perturbation calculation for rapid TBR enhancement study of the solid breeder TBB.

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Study on fusion blanket with ceramic solid as tritium breeding material

Journal of Physics Conference Series ◽

10.1088/1742-6596/2072/1/012004 ◽

2021 ◽

Vol 2072 (1) ◽

pp. 012004

Author(s):

I R Maemunah ◽

Z Su’ud ◽

A Waris ◽

D Irwanto

Keyword(s):

Fusion Reactor ◽

The Self ◽

Tritium Breeding ◽

Thermodynamic Behavior ◽

Breeding Material ◽

Self Sufficiency

Abstract Variation of solid ceramic breeding might be one of the excellent candidates in a fusion reactor. The LiAlO2, Li4SiO4, Li2O, and Li2ZrO3 show pretty good requirements in tritium breeding capability and thermodynamic behavior. Especially for LiAlO2 and Li2ZrO3, in which they could be possible to breed without neutron multiplying needed as blanket used generally in order to reach the self-sufficiency reactor. So that, it makes up the material could be possible as high-estimation breeder material.

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Neutronic Analyses for CFETR With Modular Helium Cooled Lithium Ceramic Blanket

Volume 5: Advanced and Next Generation Reactors, Fusion Technology; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues ◽

10.1115/icone25-67291 ◽

2017 ◽

Author(s):

Kun Xu ◽

Minyou Ye ◽

Yuntao Song ◽

Mingzhun Lei ◽

Shifeng Mao

Keyword(s):

Fusion Energy ◽

Time Interval ◽

Vacuum Vessel ◽

Tritium Breeding ◽

Thermal Shield ◽

Sector Model ◽

Self Sufficiency ◽

Radial Profiles ◽

Single Standard ◽

Test Reactor

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by national integration design group for magnetic confinement fusion reactor of China to bridge the R&D gaps between ITER and DEMO. Since the launch of CFETR conceptual design, a modular helium cooled lithium ceramic blanket concept had been under development by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences, to complete CFETR in demonstrating its fusion energy production ability, tritium self-sufficiency and the remote maintenance strategy. To validate the feasibility, the neutronic analyses for CFETR with this modular helium cooled lithium ceramic blanket were performed. The 1-D neutronic study for CFETR was done in the first place to give a preliminary and quick demonstration of the overall neutronic performance. Meanwhile, the neutronic analyses for a single standard helium cooled lithium ceramic blanket module were done in several times to give more insight for the material and geometry parameters of intra-module structures. Therefore, the principles for neutronic design and the module level optimized parameters were produced, based on which the design of practical blanket modules planted in tokamak vacuum vessel was completed. In the end, the 3-D neutronic analysis for CFETR was done utilizing the MCNP code, in which the 11.25 degree sector model (consist of blanket modules, manifold, support plate, shield, divertor, vacuum vessel, thermal shield and TF coils) was generated with the McCad automated conversion tool from the reference CAD model for analysis, the bi-dimensional (radial and poloidal) neutron source map was plugged via general source definition card to stimulate the D-T fusion neutrons. The concerned neutronics parameters of CFETR, mainly including the tritium breeding ratio to characterize tritium self-sufficiency, the energy multiplication factor to characterize power generation, as well as, the inboard mid-plane radial profiles of neutron flux densities, helium production rate, displacement damage rate and the energy deposition to characterize the shielding performance, were produced. In principle, the neutronics performance of CFETR with modular helium cooled lithium ceramic blanket is promising. The tritium breeding capability meets the design target and, by referring to that for ITER and the EU DEMO fusion power plant, the inboard mid-plane shielding is effective to fulfill the radiation design requirement of the superconducting TF-coils, resulting in a compulsory warm-up time interval of ∼2 FPY for TF-coils. The nuclear heating loads to other CFETR components were generated. As an outcome of this work, the applicability of McCad on CFETR neutronic modeling is demonstrated.

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Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

Applied Sciences ◽

10.3390/app11115234 ◽

2021 ◽

Vol 11 (11) ◽

pp. 5234

Author(s):

Jin Hun Park ◽

Pavel Pereslavtsev ◽

Alexandre Konobeev ◽

Christian Wegmann

Keyword(s):

Monte Carlo ◽

Fusion Reactor ◽

Nuclear Data ◽

Data File ◽

Nuclear Model ◽

Model Parameters ◽

Tritium Breeding ◽

Tritium Breeding Ratio ◽

Nuclear Data Library ◽

Breeding Ratio

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

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A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

Fusion Engineering and Design ◽

10.1016/j.fusengdes.2014.04.061 ◽

2014 ◽

Vol 89 (7-8) ◽

pp. 1190-1194 ◽

Cited By ~ 1

Author(s):

L.W.G. Morgan ◽

L.W. Packer

Keyword(s):

Fusion Reactor ◽

Tritium Breeding ◽

Reactor Model ◽

Neutron Poison

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Optimal design of high efficiency double helical gear based on dynamics model

SN Applied Sciences ◽

10.1007/s42452-021-04696-0 ◽

2021 ◽

Vol 3 (8) ◽

Author(s):

Fengxia Lu ◽

Xuechen Cao ◽

Weiping Liu

Keyword(s):

Dynamic Model ◽

High Efficiency ◽

Bending Strength ◽

Sliding Friction ◽

Elastohydrodynamic Lubrication ◽

Optimization Method ◽

Transmission Efficiency ◽

Helical Gear ◽

Transmission System ◽

Vibration Displacement

AbstractA 16-degree-of-freedom dynamic model for the load contact analysis of a double helical gear considering sliding friction is established. The dynamic equation is solved by the Runge–Kutta method to obtain the vibration displacement. The method combines the friction coefficient model based on the elastohydrodynamic lubrication theory with the dynamic model, which provides a theoretical basis for the calculation of the power loss of the transmission system. Moreover, the sensitivity analysis of the parameters that affect the transmission efficiency is carried out, and an optimization method of meshing efficiency is proposed without reducing the bending strength of the gears. This method can directly guide the design of the double helical gear transmission system.

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Updated Design and Development Route for CH HCCB TBM and its Mockup

Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering ◽

10.1115/icone21-16046 ◽

2013 ◽

Author(s):

Gang Hu ◽

Kaiming Feng ◽

Zhou Zhao ◽

Guoshu Zhang ◽

Qijie Wang ◽

...

Keyword(s):

Fusion Reactor ◽

Surface Heat ◽

Maximum Temperature ◽

First Wall ◽

Total Stress ◽

Design And Development ◽

Tritium Production ◽

Radial Length ◽

Development Route ◽

Purge Gas

Chinese helium-cooled ceramics breeder test blanket module (CH HCCB TBM) is determined to be tested in ITER machine to get data for fusion reactor design and development in future. Chinese TBM is designed to occupy half of port C with 484mm in torroidal and 1660mm in poloidal. Radial length is 675mm. TBM is composed of box, 12 submodules and independent backplate. Box formed by first wall, grids and caps have 12 caivities to hold submodules. Box and submodules are supported by backplate by welding. Backplate distribute helium with flow rate 1.36kg/s to cool first wall and then part of it go out of TBM by bypass. The rest 0.77kg/s go on to cool caps and girds first and then cool submodules. Submodules with dimensions 250mm×202mm×318mm have independent cooling and purging systems connected to backplate manifold systems. In a submodule, two U-shaped structures hold breeding material Li4SiO4 pebbles. Out of the structure filled beryllium pebbles. Neutronics results show that tritium production is ∼64mg/FPD. Maximum temperature 538°C of structure material occurs in the front of first wall with surface heat flux 0.5MW/m2. Maximum total stress at first wall is 471MPa at 394°C; that in submodules is 426MPa at 400°C; that in backplate is 526MPa at 410°C, In order to explore development technologies for the TBM, a mockup with dimensions 484mm (torroidal)×592mm (poloidal)×675mm (radial) has been designed. The mockup with similar structure ignores bypass and purge gas systems. In the mockup, there’s only one submodule and the other three are replaced by submodule replacements. By discussions and investigations, development route has been decided and the mockup is being fabricated.

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