Design and thermal-hydraulic evaluation of helium-cooled ceramic breeder blanket for China fusion engineering test reactor

Lei, Mingzhun; Xu, Shuling; Guo, Chao; Liu, Sumei; Song, Yuntao; Lü, Kun; Pei, Kemei; Xu, Kun

doi:10.1002/er.3961

Cited by 16 publications

(5 citation statements)

References 14 publications

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“…One principle that needs to be followed in the blanket system integration is to be able to accommodate all the components located on the vacuum vessel. In addition, the integration is based on coherent design of all the components with reference to the HCCB blanket concept …”

Section: Design Requirements and Strategy Of Cfetr Blanket Systemmentioning

confidence: 99%

“…The nuclear heat generation of CFETR components is displayed in Table , and the nuclear energy multiplication factor is 1.35. The performed nuclear heating profiles in each component of the blanket were used for thermal‐hydraulic analysis and optimization for the HCCB blanket module and were used in the preliminary safety analyses of the typical HCCB blanket module.…”

Section: Shielding Evaluation Of the Integrated Blanket Systemmentioning

confidence: 99%

“…In addition, the integration is based on coherent design of all the components with reference to the HCCB blanket concept. 7 At the beginning of design and integration of the blanket system, the maintenance access has been considered for replacement and the subsequent repair activities. The maintenance access is very important for the prevention of the failure risk of the in-vessel components, which are unpredictable and cannot be excluded under varies and cyclic loads conditions besides the design complexity.…”

Section: Design Requirements and Strategy Of Cfetr Blanket Systemmentioning

confidence: 99%

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Conceptual design of the HCCB blanket system integration for CFETR

et al. 2019

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Summary In a fusion power plant, the integration of the blanket system in the design progress is of vital importance to address the fundamental function of sufficient tritium production, reliable nuclear heat extraction, and permanent components protection due to the complex assembling systems into the tokamak vessel. Some progress activities of the blanket system design and integration for China fusion engineering test reactor (CFETR) are developed. The integration work involves the design of the breeding blanket, back plate support, shielding blanket, and supporting structures. To guarantee normal operation of the reactor, the design and arrangement of cooling pipes are very critical. The layout of the complex cooling pipes inside the blanket system is designed and integrated. Interfaces between main connecting components are designed for blanket system integration. In this work, the U‐shaped HCCB blanket is utilized as the adaption breeding module into the integrated blanket system. The three‐dimensional (3D) neutronic analyses verified that the integrated design of the blanket system could well meet the requirements of tritium self‐sufficiency and neutron shielding. The nuclear heat generation of main components in the reactor is obtained with a nuclear energy multiplication factor of 1.35. The primary principle during the integration is to accommodate all the components allocated on the vacuum vessel. This work is the preliminary integration design and validation for CFETR blanket system, and further detailed design will be performed around these obtained references for better fusion feasibility and performance in the next‐step design stage.

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Section: Design Requirements and Strategy Of Cfetr Blanket Systemmentioning

confidence: 99%

Section: Shielding Evaluation Of the Integrated Blanket Systemmentioning

confidence: 99%

Section: Design Requirements and Strategy Of Cfetr Blanket Systemmentioning

confidence: 99%

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Conceptual design of the HCCB blanket system integration for CFETR

et al. 2019

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“…For these reasons, several countries (China, European Union, India, Japan, Korea, Russia, and United states) have conducted research on the International Thermonuclear Experimental Reactor (ITER), which is the first fusion device maintaining fusion for long periods, and the demonstration reactor (DEMO), designed to prove the generation of electricity over several years [1]. However, there are several scientific/technological problems in developing nuclear fusion reactors, related to plasma instability [2], superconducting systems [3], blankets [4,5], and materials [6,7]. In particular, the divertor, which has the role of removing exhausted particles from the plasma and securing the shape of the plasma in the tokamak [1], is one of the most challenging components of a fusion plant, whose function is complicated by the requirement to dissipate large quantities of heat (~10 MW/m 2 ).…”

Section: Introductionmentioning

confidence: 99%

Heat-Absorbing Capacity of High-Heat-Flux Components in Nuclear Fusion Reactors

Lee

Kim²,

Moon³

et al. 2019

Energies

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Nuclear fusion energy is a solution to the substitution of fossil fuels and the global energy deficit. However, among the several problems encountered for realizing a nuclear fusion reactor, the divertor presents difficulties due to the tremendous heat flux (~10 MW/m2) from high-temperature plasma. Also, neutrons produce additional heat (~17.5 MW/m3) from collisions with the materials’ atoms. This may lead to unexpected effects such as thermal failure. Thus, a comprehensive investigation on the divertor module is needed to determine the heat-absorbing capacity of the divertor module so to maintain the effect of incident heat flux. In this study, using an analytical approach and a simulation, the quantitative effect of heat generation on the thermophysical behavior, such as temperature and thermal stress, was analyzed while maintaining the incident heat flux. Then, a correlated equation was derived from the thermal design criteria, namely, the maximum thimble temperature and the safety factor at the vulnerable point. Finally, on the basis of the thermal design criteria, the heat-absorbing capacity of a nuclear fusion reactor in operating conditions was determined. This study contributes to the understanding of the divertor’s effects in nuclear fusion reactors for high-heat-flux and high-temperature applications.

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“…The PFCs are generally composed of plasma facing materials (PFMs), interlayer and heat sink materials 1,2 . With the development of fusion reactors, candidate PFMs have undergone a transition from low-Z carbon-based or beryllium materials to high-Z tungsten 3,4 . As tungsten has many unique properties such as low sputtering erosion and tritium retention, high melting point and moderate thermal expansion 5 , it has been choosing as the main divertor PFMs in ITER and has been foreseen as the most suitable candidate for the first wall in demonstration fusion reactor (DEMO) or future fusion reactors [6][7][8] .…”

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confidence: 99%

Thermal damage of tungsten-armored plasma-facing components under high heat flux loads

Wang

et al. 2020

Sci Rep

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Fusion energy is expected as a promising candidate for alternative next generation energy. For fusion reactor, the plasma facing components (PFCs) are the most critical components to achieve this goal. PFCs will suffer severe thermal shock due to repective cyclic high heat flux (HHF) loads. This paper investigates the effects of thermal shock and damage behavior of tungsten armored PFCs under steady, transient and combined thermal loads. The distribution of stress field is analyzed, and crack initiation is predicted using the extended finite element method (XFEM). The unique features of thermalmechanical behavior of tungsten armored PFCs under simulated service condition are discussed. The dominant factor of the cracking of the tungsten armor is the brittleness of tungsten below ductileto-brittle transition temperature (DBTT). Under the steady loads, the cracking position is apt to near the interface of tungsten armor and the interlayer, and the threshold of cracking is between 14 MW/ m 2 and 16 MW/m 2 . With 6 MW/m 2 steady loads, applying 1 ms duration of transient load, the cracking threshold is between 0.2 GW/m 2 to 0.4 GW/m 2 . The depth of cracking increases from 100 um to 500 um with the transient load increasing from 0.4 GW/m 2 to 1.0 GW/m 2 . Researches are useful for the design and structural optimization of tungsten-armored PFCs, and the long-term stable operation of further reactor. FE ModelGeometry, FE mesh and materials. We design the monoblock and build the model as shown in Fig. 1 8,10-12 .Both width and height of the selected monoblock are 28 mm, the axial length is 12 mm and the armor thickness is 6 mm. The CuCrZr alloy has been chosen as the heat sink material because of its good irradiation resistance and high thermal conductivity, and its tube diameters are 12/15 mm (ID/OD). The oxygen free high conductivity copper (OFHC-Cu)

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Design and thermal-hydraulic evaluation of helium-cooled ceramic breeder blanket for China fusion engineering test reactor

Cited by 16 publications

References 14 publications

Conceptual design of the HCCB blanket system integration for CFETR

Conceptual design of the HCCB blanket system integration for CFETR

Heat-Absorbing Capacity of High-Heat-Flux Components in Nuclear Fusion Reactors

Thermal damage of tungsten-armored plasma-facing components under high heat flux loads

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