Material problems and requirements related to the development of fusion blankets: The designer point of view

Donne, M. Dalle; Harries, D.R.; Kalinin, G.; Mattas, R.F.; Mori, Seiji

doi:10.1016/0022-3115(94)90035-3

Cited by 17 publications

(3 citation statements)

References 22 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…[67][68][69] For the designing and licensing of ITER-TBM, design criteria compatible with the irradiation-induced property changes of the materials should be prepared.…”

Section: Toward the Iter Test Blanket Modulementioning

confidence: 99%

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Kimura

2005

Mater. Trans.

View full text Add to dashboard Cite

Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM).Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement.A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency.

show abstract

“…[67][68][69] For the designing and licensing of ITER-TBM, design criteria compatible with the irradiation-induced property changes of the materials should be prepared.…”

Section: Toward the Iter Test Blanket Modulementioning

confidence: 99%

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Kimura

2005

Mater. Trans.

View full text Add to dashboard Cite

show abstract

“…However, excessive, exothermic interactions with steam have been observed for dense Be at T > 800°C and for porous Be at T >675°C. Chemical interaction between Be layers and lithium-based ceramics initiates at~700°C [15]. For Be spheres in contact with ternary ceramic spheres, chemical interaction layers and diffusion zones have been observed to form on the Be in the temperature range of 750 -950°C [ 16].…”

Section: Compatibilitymentioning

confidence: 99%

Status of beryllium development for fusion applications

Billone¹

1995

Fusion Engineering and Design

View full text Add to dashboard Cite

Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma facing component of first wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold isostatic pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design.Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation.In this current work, the range of anticipated fusion operating conditions is reviewed. The thermal, mechanical, chemical compatibility, tritium retention/ release, and helium retention/swelling databases are then reviewed for fabrication methods and fusion operating conditions of interest. Properties correlations and uncertainty ranges are also discussed. In the case of the more complex phenomena of tritium retention/release and helium-induced swelling, fundamental mechanisms and models are reviewed in more detail. Areas in which additional data are needed are highlighted, along with some trends which suggest ways of optimizing the performance of beryllium for fusion applications. DISCLAIMERThis report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

show abstract

“…Beryllium-based materials, both pure beryllium and intermetallic compounds such as beryllides, possess unique nuclear physical properties, which currently allows them to be widely used as reflectors and moderators [1][2][3][4] in material testing nuclear reactors (beryllium) and as neutron multipliers in the future projects of European Helium Cooled Pebble Bed (HCPB) concept of ITER (beryllium and titanium beryllide as pebbles or blocks) and DEMO (titanium beryllide) blankets [5][6][7][8][9][10]. Beryllium transmutes under neutron irradiation into helium and tritium [11] that causes dimensional instability of beryllium products due to swelling.…”

Section: Introductionmentioning

confidence: 99%

Swelling of Highly Neutron Irradiated Beryllium and Titanium Beryllide

Chakin

Fedorov

Gaisin

et al. 2022

JNE

View full text Add to dashboard Cite

The swelling of beryllium and titanium beryllide after irradiation at 70–750 °C to neutron fluences of (0.25–8) · 1022 cm−2 (E > 1 MeV) was measured using methods of immersion, dimension, and helium pycnometry. Dependences of the swelling on the irradiation temperature and neutron dose were plotted and analyzed. The dose dependences show linear dependences of the swelling for all irradiation temperatures except 70 °C, where the swelling rate varies, depending on increasing neutron dose. Be-7Ti shows much less swelling than pure Be. Irradiation at 430–750 °C to neutron fluence of 1.82 · 1022 cm−2 (E > 1 MeV) leads to swelling of Be at about 50%; for Be-7Ti, it is 2.7%. The microstructure study shows that the formation of bubbles and pores in beryllium occurs much more intense than in titanium beryllide.

show abstract

Material problems and requirements related to the development of fusion blankets: The designer point of view

Cited by 17 publications

References 22 publications

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Status of beryllium development for fusion applications

Swelling of Highly Neutron Irradiated Beryllium and Titanium Beryllide

Contact Info

Product

Resources

About