TEM investigation of long-term annealed highly irradiated beryllium

Renterghem, W. Van; Leenaers, A.; denBerghe, S. Van

doi:10.1016/j.jnucmat.2007.07.007

Cited by 9 publications

(5 citation statements)

References 13 publications

(26 reference statements)

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“…The type of microstructure formed in beryllium under irradiation strongly depends on the irradiation temperature. Below 400-500 °C, interstitial loops nucleate first and then grow with increasing fluence, finally building together into dislocation network 24,25 . No bubble observations were reported in these works.…”

Section: Discussionmentioning

confidence: 99%

New insights into microstructure of irradiated beryllium based on experiments and computer simulations

Klimenkov

Vladimirov

Jäntsch

et al. 2020

Sci Rep

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The microstructural response of beryllium after neutron irradiation at various temperatures (643-923 K) was systematically studied using analytical transmission electron microscope that together with outcomes from advanced atomistic modelling provides new insights in the mechanisms of microstructural changes in this material. The most prominent feature of microstructural modification is the formation of gas bubbles, which is revealed at all studied irradiation temperatures. except for the lowest irradiation temperature, gas bubbles have the shape of thin hexagonal prisms with average height and diameter increasing with temperature. A high number density of small bubbles is observed within grains, while significantly larger bubbles are formed along high-angle grain boundaries (GB). Denuded zones (DZ) nearly free from bubbles are found along both high-and low-angle grain boundaries. Precipitations of secondary phases (mainly intermetallic Al-Fe-Be) were observed inside grains, along dislocation lines and at GBs. EDX analysis has revealed homogeneous segregation of chromium and iron along GBs. the observed features are discussed with respect to the available atomistic modelling results. in particular, we present a plausible reasoning for the abundant formation of gas bubbles on intermetallic precipitates, observation of various thickness of zones denuded in gas bubbles and precipitates, and their relation to the atomic scale diffusion mechanisms of solute-vacancy clusters.

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Section: Discussionmentioning

confidence: 99%

New insights into microstructure of irradiated beryllium based on experiments and computer simulations

Klimenkov

Vladimirov

Jäntsch

et al. 2020

Sci Rep

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“…The beryllium matrix was subjected to irradiation for 15 years at 323 K up to a fission neutron fluence of 5.32 × 10 26 m −2 , which can be converted to a fast neutron fluence of 4.67 × 10 26 m −2 (E > 1 MeV) [1]. Consequently, the estimated production levels of helium and tritium in the irradiated beryllium matrix were 22,000 appm of 4 He and 2000 appm of 3 H. Due to the radioactive decay of tritium from 1995 to 2022, the content of tritium in beryllium to the time of the thermal desorption tests decreased to 440 appm.…”

Section: Methodsmentioning

confidence: 99%

“…The BR2 material testing reactor utilizes a beryllium matrix that has undergone several replacements since 1963. The pieces of the irradiated beryllium matrix, replaced in 1995, were used for a study of beryllium swelling and creep behavior, as well as microstructural evolution [1][2][3][4]. These findings are crucial not only for ensuring the safe operation of the BR2 reactor and other research nuclear reactors that use beryllium reflectors and moderators [5][6][7][8] but also for fusion applications.…”

Section: Introductionmentioning

confidence: 99%

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Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor

Chakin

Rolli

Gaisin

et al. 2023

JNE

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The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.

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“…swelling, embrittlement and microhardness [42][43][44]. These three degradations are caused by the formation of radiogenic helium and insufficient gas atom diffusion in the microstructure of the berylllium [40,[52][53][54].…”

Section: Berylliummentioning

confidence: 99%

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Purba

2017

Jurnal Pengembangan Energi Nuklir

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PHYSICAL AGEING OF THE RESEARCH REACTOR CORE STRUCTURAL MATERIALS DUE TO NEUTRON IRRADIATION EXPOSURE: A REVIEW.A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptable limits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures. ABSTRAK PENUAAN MATERIAL STRUKTUR TERAS REAKTOR RISET SECARA FISIK YANG DISEBABKAN OLEH PAPARAN RADIASI NEUTRON: SEBUAH KAJIAN.Reaktor riset adalah reaktor nuklir yang difungsikan untuk pembangkitan dan pemanfaatan fluks neutron dan ionisasi radiasi untuk tujuan penelitian dan aplikasi industri. Lebih dari 60% reaktor riset yang ada saat ini telah beroperasi selama 30 tahun atau lebih. Seiring dengan berjalannya waktu, kemampuan struktur, sistem dan komponen (SSK) tentunya mengalami penurunan yang dikenal dengan penuaan fisik. Penuaan fisik ini dapat disebabkan oleh karena adanya paparan radiasi neutron seperti irradiasi yang menyebabkan terjadinya dislokasi dan perubahan mikrostruktur material. Untuk memperpanjang umur reaktor dan untuk mengurangi terjadinya pemadaman yang tidak diinginkan maka perlu dilakukan manajemen penuaan SSK, khususnya yang terkait dengan keselamatan operasi. Tujuan dari penelitian ini adalah untuk mengkaji penuaan fisik yang terjadi pada material struktur teras reaktor riset yang disebabkan oleh paparan iradiasi neutron. Untuk mencapai tujuan ini maka dilakukan kajian pada berbagai macam pustaka yang terkait. Pembahasan yang komprehensif tentang sifat-sifat iradiasi dibatasi hanya pada bejana reaktor dan material struktur pendukung teras yang terbuat dari zirconium dan beryllium dan senyawanya yang umum digunakan oleh reaktor riset. Dari kajian yang telah dilakukan, ditemukan bahwa stabilitas sifat-sifat mekanik zirconium dan beryllium serta senyawanya sangat dipengaruhi oleh fluent dan temperatur neutron. Kata kunci: Reaktor riset, p...

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TEM investigation of long-term annealed highly irradiated beryllium

Cited by 9 publications

References 13 publications

New insights into microstructure of irradiated beryllium based on experiments and computer simulations

New insights into microstructure of irradiated beryllium based on experiments and computer simulations

Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

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