An overview of the EU breeding blanket design strategy as an integral part of the DEMO design effort

Federici, G.; Boccaccini, L.V.; Cismondi, F.; Gasparotto, M.; Poitevin, Y.; Ricapito, I.

doi:10.1016/j.fusengdes.2019.01.141

Cited by 207 publications

(132 citation statements)

References 48 publications

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“…For such reasons, DEMO must have a closed fuel-cycle to recover, process and reuse either the unburnt tritium or the tritium opportunely produced inside the breeding blanket. Particularly the breeding blanket is a region surrounding the tokamak composed of a Li-based material (there are several options under investigation for the most appropriate blanket configuration, see Reference [15] where the neutrons interact with lithium for producing tritium, see Equation 2.…”

Section: The Problem Of Tritium Impurities Inside the Helium Primary mentioning

confidence: 99%

“…An effective breeding blanket, coupled with an efficient fuel cycle, will allow the DEMO reactor to achieve the ambitious goal of self-sufficiency, meaning that all the reacted tritium is reproduced on site. An accurate tritium control and management is fundamental also for safety aspects; it is worth to mention that the allowable tritium release into environment is currently set at 1 gr per year while the tritium produced inside the breeding blanket is about 320 gr per day -the indicated tritium production rate is referred to a DEMO design configuration having a fusion power of 1998 MW and a tritium breeding ratio of 1.05 [15]. Being the blanket the region where tritium is produced and the fusion heat is removed through the primary coolant loops, it is characterized by high tritium concentration, high temperature and large metallic surface areas with reduced wall thickness.…”

Section: The Problem Of Tritium Impurities Inside the Helium Primary mentioning

confidence: 99%

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Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery

et al. 2020

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Non-evaporable getters (NEGs) are metallic compounds of the IV group, particularly titanium and/or zirconium-based alloys and are usually used as pumps in vacuum technologies since they are able to sorb, by chemical reactions, most of the active gas molecules, with particular efficacy towards hydrogen isotopes. This work suggests an alternative application of these materials to fusion nuclear reactors, where there is the need to recover small amount of tritium from the large helium flow rate composing the primary coolant loop. Starting from the tritium mass balance inside the primary coolant loop, the amount of coolant to be routed inside the coolant purification system (CPS) is identified. Then a feasibility study, based on the bulk getter theory, is presented by considering three different commercial alloys, named ST707, ST101 and ZAO. The results provide the mass, the area and the regeneration parameters of the three different alloys necessary to fulfill the requirements of the CPS unit. By comparing the features of the three alloys, the ZAO material appears the most promising for the proposed application because it requires the lower amount of material and a lower number of regeneration cycles.

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Section: The Problem Of Tritium Impurities Inside the Helium Primary mentioning

confidence: 99%

Section: The Problem Of Tritium Impurities Inside the Helium Primary mentioning

confidence: 99%

Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery

et al. 2020

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“…In the European DEMO (EU-DEMO) fusion reactor, the Breeding Blanket (BB) component accomplishes several functions [1,2]. Firstly, it acts as a cooling device.…”

Section: Introductionmentioning

confidence: 99%

“…In the framework of the EUROfusion Programme, two breeding blanket concepts were selected for the EU-DEMO R&D strategy: Water-Cooled Lithium-Lead (WCLL) and Helium-Cooled Pebble Bed (HCPB) [1]. These two technologies will also be tested in the ITER fusion reactor, according to the goals of the ITER Test Blanket Module (TBM) programme [1]. The main outcome of this experimental campaign will be the Return of Experience for the EU-DEMO Breeding Blanket Programme [3].…”

Section: Introductionmentioning

confidence: 99%

“…The computational activity presented in this paper deals with the WCLL option. It foresees the usage of water at typical Pressurized Water Reactor (PWR) thermodynamic conditions (295-328 • C and 15.5 MPa) as coolant [1,2]. The blanket relies on liquid lithium-lead as breeder, neutron multiplier and tritium carrier and on Eurofer as structural material.…”

Section: Introductionmentioning

confidence: 99%

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Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category

et al. 2021

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The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

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Development and experiments of LiPb coolant technologies for fusion blanket in China

Qunying

Zhu

Team³

2020

Int J Energy Res

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Summary The liquid lithium lead (LiPb) blanket concept has become a promising design for fusion reactor, for example, international thermonuclear experimental reactor (ITER) and the demonstration (DEMO) reactor. To promote the successful application of fusion energy, detailed studies on key issues of the LiPb blanket were performed, such as corrosion behavior, thermal hydraulics and coolant safety etc. Some experimental facilities needed for the key issue studies were designed and developed under the research program. The series LiPb experimental loops, named DRAGON, were developed in China, including the thermal convection loops DRAGON‐I/II/III and the multi‐functional experimental loop DRAGON‐IV. To perform the integrated experiments under the multi‐physical field conditions for DEMO blanket, a dual‐coolant thermal integrated experimental loop DRAGON‐V was built in 2017. It is a multi‐functional test platform for the research and development (R&D) of key issues of liquid LiPb blanket to provide the necessary database for ITER and DEMO. Up to now, a lot of experiments have been performed to investigate the LiPb technologies, such as corrosion behaviors of structural materials both with and without magnetic field, corrosion resistance technology, magneto‐hydrodynamics effect, and interaction between typical coolants during in‐box loss of coolant accident etc. The results will support the R&D of the key techniques and the engineering design for ITER test blanket module and further DEMO.

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An overview of the EU breeding blanket design strategy as an integral part of the DEMO design effort

Cited by 207 publications

References 48 publications

Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery

Novel Non-Evaporable Getter Materials and Their Possible Use in Fusion Application for Tritium Recovery

Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category

Development and experiments of LiPb coolant technologies for fusion blanket in China

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