Design of a hydrogen/tritium permeation sensor for Gen-IV sodium fast reactors

Candido, Luigi; Alberghi, Ciro; Utili, Marco

doi:10.1016/j.vacuum.2021.110414

Cited by 4 publications

(2 citation statements)

References 28 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The D 2 partial pressure is 139 Pa. Assuming equal solubility of hydrogen isotopes (hydrogen, deuterium and tritium) in LiPb, the Sieverts' constants are related to each other by means of the following relationship [28,29]:…”

Section: Preliminary Results Obtainedmentioning

confidence: 99%

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Utili¹,

Alberghi²,

Candido³

et al. 2022

Nucl. Fusion

Self Cite

View full text Add to dashboard Cite

The experimental qualiﬁcation of the Tritium Extraction Unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the Water-Cooled Lithium-Lead (WCLL) breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction eﬃciency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone research centre, Italy. The facility is able to qualify Gas-Liquid Contactor (GLC), Permeator Against Vacuum (PAV) and Liquid-Vacuum Contactor (LVC) technologies at diﬀerent temperatures, lithium-lead mass ﬂow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.

show abstract

Section: Preliminary Results Obtainedmentioning

confidence: 99%

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Utili¹,

Alberghi²,

Candido³

et al. 2022

Nucl. Fusion

Self Cite

View full text Add to dashboard Cite

show abstract

“…One of the promising trends in the design of Generation IV reactors is the creation of high-temperature gas-cooled reactors, whose main purpose is not only to produce technological heat, but also to produce hydrogen [ 24 , 25 , 26 ]. According to the development concept and roadmap for Generation IV power systems, the main questions to be answered before commissioning high-temperature gas-cooled reactors are to determine their stability, safety, reliability, and physical protection against radiation damage [ 27 , 28 , 29 , 30 ].…”

Section: Introductionmentioning

confidence: 99%

Study of Morphological, Structural, and Strength Properties of Model Prototypes of New Generation TRISO Fuels

et al. 2022

View full text Add to dashboard Cite

The purpose of this work is to characterize the morphological, structural, and strength properties of model prototypes of new-generation TRi-structural ISOtropic particle fuel (TRISO) designed for Generation IV high-temperature gas reactors (HTGR-type). The choice of model structures consisting of inner pyrolytic carbon (I-PyC), silicon carbide (SiC), and outer pyrolytic carbon (O-PyC) as objects of research is motivated by their potential use in creating a new generation of fuel for high-temperature nuclear reactors. To fully assess their full functional value, it is necessary to understand the mechanisms of resistance to external influences, including mechanical, as in the process of operation there may be external factors associated with deformation and leading to the destruction of the surface of fuel structures, which will critically affect the service life. The objective of these studies is to obtain new data on the fuel properties, as well as their resistance to external influences arising from mechanical friction. Such studies are necessary for further tests of this fuel on corrosion and irradiation resistance, as closely as possible to real conditions in the reactor. The research revealed that the study samples have a high degree of resistance to external mechanical influences, due to the high strength of the upper layer consisting of pyrolytic carbon. The presented results of the radiation resistance of TRISO fuel testify to the high resistance of the near-surface layer to high-dose irradiation.

show abstract

Atomic and ionic hydrogen flux probe for quantitative in-situ monitoring of hydrogen recycling

Kuzmin

Miura

Kobayashi

et al. 2023

Fusion Engineering and Design

View full text Add to dashboard Cite

Design of a hydrogen/tritium permeation sensor for Gen-IV sodium fast reactors

Cited by 4 publications

References 28 publications

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Study of Morphological, Structural, and Strength Properties of Model Prototypes of New Generation TRISO Fuels

Atomic and ionic hydrogen flux probe for quantitative in-situ monitoring of hydrogen recycling

Contact Info

Product

Resources

About