# **Proceedings of the 15th International Workshop on Beryllium Technology (BeWS-15)**

September, 14-15, 2022, Karlsruhe, Germany

Pavel Vladimirov, Christopher Dorn, Ramil Gaisin (eds.)

Pavel Vladimirov, Christopher Dorn, Ramil Gaisin (eds.)

### **Proceedings of the 15th International Workshop on Beryllium Technology (BeWS-15)**

September, 14-15, 2022, Karlsruhe, Germany

**Karlsruhe Institute of Technology KIT SCIENTIFIC REPORTS 7764**

# **Proceedings of the 15th International Workshop on Beryllium Technology (BeWS-15)**

September, 14-15, 2022, Karlsruhe, Germany

by Pavel Vladimirov, Christopher Dorn, Ramil Gaisin (eds.)

Report-Nr. KIT-SR 7764

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ISSN 1869-9669 ISBN 978-3-7315-1284-4 DOI 10.5445/KSP/1000156312

### Table of contents




#### Executive Summary

The 15*th* International Workshop on Beryllium Technology (BeWS-15) was initially planned to be held as a satellite meeting of the SOFT-32 in Dubrovnik, Croatia in 2021. Due to the global COVID pandemic the SOFT-32 conference was held as a virtual event only and the International Organizing Committee (IOC) of the BeWS-15 decided to postpone the workshop until in person participation will be possible.

Finally, the Workshop was held as a joint event combining both the BeWS-15 and an industrial forum BeYOND-IX during 14-15 September 2022 in Karlsruhe, Germany with a great success. The workshop was organized by the Karlsruhe Institute of Technology. A company Karlsruhe Beryllium Handling Facility (KBHF) headed by A. Goraieb and Dr. M. Lemmens was responsible for the industrial forum and helped us with the organization of joint social events as well as. Due to still difficult pandemic situation in Asia, it was decided to organize the workshop as a hybrid event, i.e., provide a possibility for the participants having difficulties with travel with online access.

Participants came mainly from Germany, the US, the UK, Kazakhstan, Latvia, Czech Republic, Japan, Sweden, France and China, totaling 55 persons, which was not expected immediately after the global pandemic.

The BeWS-15 program agenda comprised two keynote presentations, which were followed by five technical sessions.

The first session was devoted to the memory of our colleague and friend Prof. Dr. Glen Longhurst who passed away on May 12, 2021. The session was started with the presentation given by Dr. Milan Zmitko from F4E, who reported on the status of Be neutron multiplier materials for the ITER Test Blanket Module (TBM) program and overviewed developments and opened issues. The final keynote was presented by Dr. Jae-Hwan Kim, QST, Japan, who gave an overview on R&D on Neutron Multipliers in Japan.

The five technical sessions featured presentations on


The last day of the joint event was devoted to the industrial forum BeYOND-IX where contributions from Fusion-Start-Ups and Health & Safety aspects of beryllium fabrication and study were interchanged. This day has started from a Lab Tour visit to the Karlsruhe Beryllium Handling Facility situated in KIT Campus North.

The Prof. Mario Dalle Donne Memorial Award (MDDMA) was presented for the first time at the BeWS-11 in Barcelona to recognize researchers with outstanding achievements in beryllium-related research. This year, the IOC of BeWS awarded this prize to Mr. Aniceto Goraieb, the KBHF leader, based on his numerous technological achievements in beryllium-related research over the past decades. For him it was a special pleasure to receive this award as he was a student of Prof. M. Dalle Donne.

The IOC of BeYOND industrial forum decided to establish a Glen Longhurst Award for people working on the edge between industry and technological applications of beryllium materials. The Prof. Glen Longhurst Award was presented for the first time to Mr. Christopher Dorn for his long-term support for R&D on beryllium materials for their application for fusion technology working as one of the chief managers at Materion Inc. in USA, keeping beryllium safety standards and services at ITER in France and at JET in UK, and running his own consulting company on beryllium business.

### SESSION 1

# In memory of Glen Longhurst

#### **In memory of Glen Longhurst (1943-2022)**

#### Pavel V. Vladimirov<sup>1</sup> , Christopher K. Dorn<sup>2</sup>

*<sup>1</sup>Karlsruhe Institute of Technology, Institute for Applied Materials - Applied Materials Physics, 76344 Eggenstein-Leopoldshafen, Germany <sup>2</sup>Be4FUSION LLC, Upland, California, U.S.A.*

This presentation is devoted to the memory of our colleague and dear friend, Prof. Dr. Glen Longhurst, who passed away on May 12, 2021.

Together with Mario Dalle Donne, Hiroshi Kawamura and Vladimir Shestakov, Glen was one of the co-founders of the Beryllium Workshop in 1991. After that, he was a member of the International Organizing Committee (IOC) for many years, and he actively worked on further development of the Workshop Series. In 2005, Glen hosted the BeWS-7 in Santa Barbara, California, USA as a satellite meeting of the ICFRM-12 conference, and the workshop was a great success. In 2012 during the BeWS-10 in Karlsruhe, he suggested the establishment of the Mario Dalle Donne Memorial award to honor outstanding contributions by scientists working in beryllium area.

In the field of beryllium technology, Glen was well known for his investigations of beryllium behavior under irradiation, its handling, recycling, application for fusion and, especially, for his studies on the retention of accumulated tritium within beryllium. Glen developed the Tritium Migration Analysis Program (TMAP) for simulation of tritium transport and release, which has survived through seven major releases and is still regarded as the gold standard in the field today.

A special award for lifetime achievement, in particular, for the establishment and development of the BeWS series, was presented to Prof. Emeritus G. Longhurst during the previous BeWS-14 held in Long Beach, California, USA in 2019. The following slides share some memories from this event.

To honor Glen's achievements, the IOC of the BeYOND Industrial Forum has decided to establish a Glen Longhurst Memorial Award for exceptional persons working on the edge between industry and technological applications of beryllium materials.

Glen will be forever enshrined in our hearts as an outstanding scientist, a great personality, and a wonderful friend to us all.

#### **Corresponding Author:**

Dr. Pavel Vladimirov pavel.vladimirov@kit.edu Karlsruhe Institute of Technology P.O. Box 3640 76021 Karlsruhe Germany

In memory of Glen Longhurst Department of Metallic Alloys, Institute for Applied Materials - Applied Materials Physics

Dr. Glen Longhurst was one of the co-founders of the Beryllium Workshop in 1991. He also served on the IOC for many years and worked to develop the BeWS series.

Glen was well known for his investigations of beryllium behavior under irradiation and, especially, for his studies on the retention of accumulated tritium within beryllium. Glen developed the Tritium Migration Analysis Program (TMAP) for simulation of tritium release and transport, which has survived through seven major releases and is still regarded as the gold standard in the field today.

A special award for lifetime achievement, in particular, for the establishment and development of the BeWS series, was presented to Prof. Longhurst during the BeWS-14 Glen Longhurst: 1943-2021 held in Long Beach, California, USA in October 2019.

#### **The HCPB Test Blanket Module: Current Status in Development and Qualification of Beryllium Materials and an Overview of Open Issues**

Milan Zmitko<sup>1</sup> , P. Vladimirov<sup>2</sup> , V. Chakin<sup>2</sup> , A. Spagnuolo<sup>3</sup>

*<sup>1</sup>Fusion for Energy (F4E), TBMs and Materials Development, Barcelona, Spain 2 Institut fuer Materialforschung I, KIT, Karlsruhe, Germany <sup>3</sup>EUROfusion - Programme Management Unit, Boltzmannstrasse 2, Garching, Germany*

One of the reference tritium Breeder Blanket concepts developed in the Europe that will be tested in ITER machine under the form of Test Blanket Module (TBM) is the Helium-Cooled Pebble-Bed (HCPB) TBM concept in which lithiated ceramic pebbles are used as a tritium breeder and beryllium/beryllides pebbles as a neutron multiplier material. This concept uses the EUROFER97 reduced activation ferriticmartensitic (RAFM) steel as a structural material and pressurized helium for heat extraction (8 MPa, 300-500ºC).

The paper gives a brief general description of the HCPB TBM design and the main design requirements including the requirements to beryllium multiplier material. The ITER HCPB TBM development and qualification plan with identification of the main milestones will be presented, taking into account recently signed EU-KO TBM Partnership agreement between Fusion for Energy and ITER Korea.

The main part of the paper will be devoted to the presentation of beryllium materials development strategy, qualification plan and overview of the current status of research, development and characterization.

The achieved results on fabrication technologies development, materials characterization and performance under TBM/DEMO relevant conditions, including the performance under neutron irradiation and thermo-mechanical performance will be briefly overviewed and a new neutron irradiation experiment, foreseen for the functional materials (i.e. beryllium materials and ceramic breeder pebbles), will be introduced.

A special attention will be focused on technical issues related to the HCPB TBM design and manufacturing, like for instance, Be/ceramic breeder pebbles filling procedure and realization of post-weld heat treatment operations during the course of TBM box manufacturing, and an assessment of their possible implications.

#### **Corresponding Author:**

Dr. Milan Zmitko milan.zmitko@f4e.europa.eu Fusion for Energy (F4E) c/ Josep Pla 2 – Torres Diagonal Litoral B3 08019 Barcelona, Spain

### **Test Blanket Systems testing at ITER**

Europe proposes **two Test Blanket Systems,** aiming to test **two types of tritium breeding materials and two coolants:**


**EUROFER97** structural material (RAFM steel)

**Port #16**

M. Zmitko et al., The HCPB TBM: Current Status of Beryllium Materials and Open Issues

• Easy reprocessing capability for the CB pebbles (in the view of DEMO/future FPR)

9

10

### **HIDOBE Post-Irradiation Examination Microstructure of Be pebbles**

### **Beryllium materials R&D activities: HIDOBE-01/02 PIE results**

### **Summary of the main PIE results:**

#### **Microstructure:**


#### **Tritium release:**


$$^{45}$$

17

Note: Following a new organization of the **European TBM Project** ! R&D support activities on Functional Materials (incl. Be materials for TBM) now under **EURO***fusion* **responsibility; Role of F4E/TBM – scope & objectives specification for ITER TBM activities**

The **main scope** of the activity is to define **a roadmap to secure the delivery of sufficient amount and requested quality of the pure Be pebble material** for the ITER-TBM. The following activities should be addressed (some of them have already been addressed):

### **Supply Capacity Development**


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#### **Functional Materials Neutron Irradiation**

	- o irradiation temperatures of the irradiated materials (irradiation temperatures for the ACB in the range of 500-900°C, of 350-650°C for pure Be pebble materials and of up 1000°C for beryllides)
	- o irradiation dose / neutron fluence (target irradiation dose for the ACB is ~2-3 dpa *in steel*, for Be pebbles ~2 dpa)
	- o type of the experiment (i.e. materials irradiation for subsequent PIE, in-situ on-line tritium release measurement for the ACB, compatibility test with structural material)
	- o Be pebbles microstructure and morphology with OM, SEM, TEM
	- o Dimensional stability (swelling), pebbles' size distribution, close and open porosity
	- o Mechanical properties (e.g. Vicker's hardness, creep behavior)
	- o Tritium and Helium release characteristics (with Temperature Programmed Desorption (TPD) process)
	- o Be pebbles compatibility with EUROFER97 structural material

M. Zmitko et al., The HCPB TBM: Current Status of Beryllium Materials and Open Issues

### **Summary**

25

### **Beryllium Multiplier Materials:**

	- Irradiation temperature most significantly affects the development of the material properties ! microstructure, tritium retention and release (mainly in HT form)

### **Overview of R&D activities on Neutron Multipliers in QST**

Jae-Hwan Kim, Taehyun Hwang, Yutaka Sugimoto, Suguru Nakano, Yoshiaki Akatsu, Masaru Nakamichi

### *National Institutes for Quantum Science and Technology, Aomori, Japan*

DEMO reactors require advanced neutron multipliers that have higher stability at high temperature. Beryllium intermetallic compounds (beryllides) are the most promising advanced neutron multipliers. Development of the advanced neutron multipliers has been started between Japan and the EU in the DEMO R&D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities. In Japan, beryllides fabrication R&D has been carried out in the DEMO R&D building at IFERC, Rokkasho.

Since beryllides are too brittle to produce the pebbles, establishing fabrication techniques for beryllides is a key issue for development of the advanced neutron multipliers. Conventional syntheses of the beryllides involve a powder metallurgy process involving a hot isostatic pressing method, a casting method, and an arcmelting method. However, beryllides synthesized conventionally are so brittle that it was not easy to fabricate the block or rod type by these methods.

On the other hand, a plasma sintering method has been proposed as a new technique for beryllides synthesis and joining because this method results in powder surface activation that enhances powder particle sinterability and reduces high temperature exposure. From the results of beryllide synthesis experiments, it was clarified that the not only disk type but rod type of beryllide has been successfully fabricated by the plasma sintering method.

To fabricate the beryllide pebbles using the plasma-sintered beryllide rod, a rotating electrode method (REM) was selected because the experience base for its use is broad for not only Be pebbles but also metallic pebbles in general industry. The result of beryllide granulation revealed that the prototypic beryllide pebbles with 0.5 mm to 2.5 mm in average diameter were successfully fabricated. In this study, the recent progress on R&Ds of beryllides as the advanced neutron multipliers in Japan will be presented.

#### **Corresponding Author:**

Dr. Jae-Hwan Kim kim.jaehwan@qst.go.jp 2-166 Omotedate, Obuchi, Rokkasho, Aomori 039-3212, JAPAN

### **Summary**


**• A joining process between Be and F82H by plasma sintering was tried. It was clear that the bonding strength of the joined materials was inversely proportional to the thickness of the reaction layer.** 

**Jae-Hwan Kim | BeWS-15, KIT| 14th 15th Sept. 2022 | page (23/24 )**

### SESSION 2

# News from Industry

### **Overview of the United States Beryllium Industry - 2022 Update**

### Keith Smith<sup>1</sup>

### *<sup>1</sup>Materion – Brush Inc, Elmore, Ohio USA*

Beryllium is a critical material of construction for the ITER First Wall Panels. The supply and fabrication of beryllium is key to the success of the ITER project. This presentation will provide an updated overview of the US beryllium industry including mining, manufacturing, and fabrication capabilities. Beryllium raw material availability, ITER relevant First Wall beryllium grades, engineering, and program management services provided in support to those products will be discussed. Additionally, activities related to process improvements within Materion's beryllium manufacturing plants including activities related to the manufacturing and purification of beryllium containing molten salt (FLIBE) be addressed.

### **Corresponding Author:**

Keith J. Smith Keith.smith@materion.com Materion Brush Inc. 14710 West Portage River South Rd Elmore, OH 43416 USA

8

### **Additive Manufacturing**


© 2021 Materion Corportation

23

### **Beryllides - experience of UMP JSC in development and testing**

Ye. Frants, M. Kolmakov, B. Zorin, M. Kylyshkanov, M Podoinikov, S. Udartsev, A. Vechkutov

### "*Ulba Metallurgical Plant" JSC, Ust-Kamenogorsk, Republic of Kazakhstan.*

Intermetallic compounds of beryllium (beryllides) have outstanding characteristics in terms of heat resistance, hardness, and resistance to oxidation. For this reason, beryllides have potential applications in nuclear and thermonuclear energy, aerospace, instrumentation and other industries.

To date, the process of producing billets and products from beryllides has not passed into the stage of stable industrial production.

In recent years, UMP JSC has been making efforts to develop and implement technologies for producing billets and products from various beryllides, as well as testing and studying the properties of beryllides.

This article presents information on the results of producing billets and products from tantalum, titanium and chromium beryllides, as well as resource thermocyclic tests under conditions simulating the thermal modes of operation of helium-cooled blanket modules of the DEMO/may be ITER reactor.

### **Corresponding Author:**

Ye. Frants FrantsEV@ulba.kz

102, Abay Avenue, Ust-Kamenogorsk, 070005, Republic of Kazakhstan

**0**

#### **Beryllium Additive Manufacturing**

#### FritzCarl Grensing1,2 , Andreas Frehn<sup>3</sup> , Carole Trybus<sup>2</sup> , Andrew Ruzek<sup>2</sup> , Rhea Christopherson<sup>2</sup> , Jacob Huxol<sup>2</sup>

*<sup>2</sup>Materion – Performance Materials, Elmore, Ohio USA <sup>3</sup>Materion,- Performance Materials, Stuttgart, Germany*

Beryllium components have historically been produced via powder metallurgy processes followed by subtractive machining operations. In the last 25 years, rapid advances in additive manufacturing have occurred in aluminum, titanium and many other materials while little to no work was done on beryllium. In 2021 Materion installed the first dedicated beryllium additive manufacturing laboratory with the goal of both developing additive technology for beryllium materials and understanding and mitigating the Environmental, Health and Safety issues associated with additively processing beryllium. Metallurgical results obtained will be described as well and the EHS data.

#### **<sup>1</sup>Corresponding Author:**

Dr. FritzCarl Grensing Fritz.Grensing@materion.com Materion Brush Inc. 14710 West Portage River South Rd Elmore, OH 43416 USA

### **Supporting Capabilities and Heritage Systems**

Gas Atomization, Powder Processing and Characterization Hot Isostatic Pressing, Post Processing and Machining Chemical, Mechanical and Microscopic Analysis Metrology, NDT and Radiography Beryllium Metallurgical Expertise Labs and Testing Capabilities

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SESSION 3

# DEMO, ITER & JET

#### **Current design of the EU DEMO Helium Cooled Pebble Bed breeding blanket**

Guangming Zhou<sup>1</sup> , Francisco A. Hernández<sup>1</sup>

#### *<sup>1</sup>Karlsruhe Institute of Technology*

In the Work Package Breeding Blanket (WPBB) of the European DEMO program, the Helium Cooled Pebble Bed (HCPB) breeding blanket is one of the two driverblanket candidates for the European DEMO and to be tested as test blanket module (TBM) in ITER. In the Pre-Concept Design (PCD) phase (2014-2020), within the framework of the EUROfusion consortium in Europe, the design of the HCPB breeding blanket has been changed to address various challenges facing the HCPB blanket concept. One of the big challenges was the use of Beryllium pebbles as the neutron multiplier in the previous design. Irradiation campaign showed that the tritium retention in the Be pebbles could impose severe safety issues and exceed the tritium limit of EU DEMO. Beryllides, on the other hand, have better properties in terms of volumetric swelling, tritium retention, irradiation and melting temperature.

This talk will focus on the current design status of the European DEMO HCPB breeding blanket and conclude with future activities in the Concept Design phase (2021-2027).

### **Corresponding Author:**

Dr. Guangming Zhou guangming.zhou@kit.edu

Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

### Outline of content


Page 2

#### • Reference design Tritium Extraction and Removal (TER) system ➢ Two stages in series First the adsorption of Q2O on the Reactive Molecular Sieve Bed (RMSB), thereafter the adsorption of Q2 on the Cryogenic Molecular Sieve Bed (CMSB) at 77 K. ➢ Tritium recovered via isotope exchange on RMSB and by heating-up of the CMSB. ➢ Extrapolated to DEMO scale is realized with industry. RMSB CMSB ➢ 80 bar purge gas, introduced to improve reliability of BB, results show that TER operating at 80 bar not a issue. ➢ CMSB requires large amount of liquid N2, getter bed is explored as alternative. ➢ Wetted purge gas to have a higher isotopic exchange rate compared to H2 and oxidized Q2, reducing permeation. 12 • Outlook [5] I. Cristescu et al., Fusion Eng. Des. 158 (2020) 111558. Q = H, D, T

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### Outlook


14

### **Regulatory situation of Beryllium in EU and France – Update Beryllium Good Practices at the Workplace - Be Responsible Program - Update**

### Angélique RENIER1,2

#### *<sup>1</sup>NGK BERYLCO FRANCE <sup>2</sup>Beryllium Science and Technology (BeST) association*

Since the last Beyond event that I attended in Karlsruhe, Germany, in 2018, the regulatory situation of beryllium in EU has evolved.

At EU level, a harmonized Occupational Exposure Limit (OEL) for beryllium has been adopted in July 2019. Member States had 2 years to implement it in their national law. Furthermore, Beryllium has been assessed for an eventual restriction in the Electrical and Electronic Equipment (EEE) under the RoHS Directive (Restriction of Hazardous Substances in EEE). The assessment started in February 2018 and the conclusion has been published in March 2021 by the European Commission: it is not recommended to restrict beryllium. Another important point is that it was decided to keep beryllium in the list of Critical Raw Materials (CRM) for the EU (last list published in September 2020). Lastly, this presentation could be the opportunity to introduce new concepts currently discussed in the frame of the green deal and the Chemicals Strategy for Sustainability, including the ongoing revision of the main EU chemicals regulations (REACH and other sectorial directives). Essential Use Concept, Safe Use Concept or General Risk Approach are emerging new concepts which could have impacts on beryllium.

At French level, ITER being based in France, it is important to mention the implementation of the EU OEL in France since the 1st March 2022 (Decree No. 2021- 1849 of December 28, 2021).

In a second part of the presentation, I suggest to introduce the Be Responsible Program (www.berylliumsafety.eu) developed by the Beryllium Association (BeST) in which NGK is involved and our actions to promote these good practices to the users of our beryllium-containing products. I was honored to present our beryllium good practices in the frame of trainings for ITER beryllium workers organized by INSTN (Institut des Sciences et Techniques du Nucléaire), several sessions in French and English. As French SME processing articles in copper-beryllium alloys, we also were very honored to share our experience during the conference on occupational cancers organized as part of the French Presidency of the European Union on March 07 and 08, 2022 in Paris.

#### **Corresponding Author:**

Mrs. Angélique RENIER Chemicals Engineer (CPE Lyon, France, 1993) Communication and Environment Manager at NGK BERYLCO France & Secretary General of the Beryllium Association BeST renier@ngkbf.com NGK BERYLCO FRANCE, 103 Quai Jean-Pierre Fougerat CS 20017 44220 COUËRON, FRANCE


A. Renier et al., Regulatory Situation of Beryllium in EU and France – Update



#### **Beryllium in JET with the ITER-Like Wall: Fuel retention, oxidation, melt erosion, dust**

M. Rubel<sup>1</sup> , A. Widdowson<sup>2</sup> , I. Jepu<sup>2</sup> , L. Dittrich<sup>1</sup> , T.T. Tran<sup>3</sup> , J. Grzonka4,5, E. Fortuna-Zaleśna<sup>4</sup> , S. Moon<sup>1</sup> , P. Petersson<sup>1</sup> and JET Contributors\*

*<sup>1</sup>KTH Royal Institute of Technology, Fusion Plasma Physics, 100 44 Stockholm, Sweden*

*<sup>2</sup>Culham Science Centre for Fusion Energy, Abingdon OX14 3DB, United Kingdom <sup>3</sup>Uppsala University, Department of Physics and Astronomy, 751 20 Uppsala, Sweden*

*<sup>4</sup>Warsaw University of Technology, 02-507 Warsaw, Poland <sup>5</sup>University of Cadiz, 11003 Cadiz, Spain*

Joint European Torus (JET) functions with the ITER-Like Wall (JET-ILW): beryllium in the main chamber and tungsten in the divertor. Be is used in the form of castellated bulk metal limiter blocks and evaporated coatings on the inner wall cladding (IWC) tiles. The operation with Be plasma-facing components (PFC) involves a spectrum of nuclear safety issues arising from plasma-material interactions: erosion, fuel retention, co-deposition with oxygen, melt damage including metal splashing and melt layer propagation, dust generation and mobilization, behaviour under massive water (liquid, vapour) or air leak, impact on invessel diagnostic components.

All above mentioned aspects of Be erosion and migration have been studied in detail with a range of methods including high-resolution in-vessel photographic survey, over twenty exsitu ion-, electron-, photon-based material analysis techniques and dedicated laboratory experiments performed on materials retrieved from JET. Results are summarized as follows.


In summary, all these points answer specific questions formulated at JET and ITER. A comprehensive report will contain research details accompanied by a critical assessment of results with emphasis of their weight and possible impact on the use of Be in future devices.

### **Corresponding Author:**

Prof. Marek Rubel rubel@kth.se KTH Royal Institute of Technology, 100 44 Stockholm, Sweden

### **Concluding Remarks: specific to JET**


Marek Rubel| Workshop on Be Technology & BeYond-IX|Karlsruhe| September 2022| Page 23

#### **Thermal desorption of tritium from beryllium plasma-facing components of the JET ITER-like wall**

Aigars Vītiņš<sup>1</sup> , Elīna Pajuste1,2, Juris Jansons<sup>1</sup> , Gunta Ķizāne<sup>1</sup> and JET Contributors\*

*EUROfusion Consortium JET, Culham Science Centre, Abingdon, OX14 3DB,UK 1 Institute of Chemical Physics, University of Latvia, One Jelgavas Street, Riga, LV-1004, Latvia*

*<sup>2</sup>Faculty of Chemistry, University of Latvia, One Jelgavas Street, Riga, LV-1004, Latvia*

\* See the author list of E. Joffrin et al., 2019 Nucl. Fusion, 59, 112021.

The JET ITER-like wall (ILW) has beryllium in the main plasma chamber and tungsten in the divertor as foreseen for the ITER vacuum vessel. Limiters and an upper dump plate of the JET vacuum vessel are bulk beryllium tiles. The goal of the ILW experimental campaigns is to characterize plasma-wall interactions and to compare plasma performance with the previous, carbon wall. One of the scientific objectives of the ILW experimental campaigns is to demonstrate sufficiently low fuel retention and, in particular, to demonstrate ITER-relevant tritium retention mitigation. The aim of the present study is to assess tritium accumulation in the beryllium tiles from the three ILW campaigns (2011-2012, 2013-2014, 2015-2016) and to investigate its thermal desorption patterns with respect to development of possible detritiation techniques. Tritium is a minor component of the fuel in the present study as tritium has not been introduced within the ILW campaigns 2011-2016. Tritium accumulation in plasmafacing tiles can occur as a result of its co-deposition with eroded material and implantation/diffusion in the bulk of the tiles.

Thermal desorption of tritium was performed in a He + 0.1% H<sup>2</sup> flow at 4.8 K/min from 290 to 1300 K. For inner wall samples, the surface concentration of tritium decreased in the three ILW campaigns from 83E11 (2011-2012) to 6E11 (2015-2016) atoms / cm<sup>2</sup> . The temperature of 50% detritiation was 697 K and 880 K for these two samples respectively. For outer wall samples, the maximum value of the surface concentration of tritium 13E13 atoms / cm<sup>2</sup> was found for a centre sample of the 2015- 2016 campaign, but the minimum value of that 0.52E13 atoms / cm<sup>2</sup> was for a righthand wing sample of the 2013-2014 campaign. The temperature of 50% detritiation was 966 K and 840 K for these two samples respectively. For the 2011-2012 and 2013-2014 campaigns, both inner and outer wall samples from a tile middle part had higher temperatures of their 50% detritiation than those of the respective wing samples. The different values of the surface concentration of tritium and dissimilarity of tritium desorption patterns of the samples investigated indicate both quantitative and qualitative differences in tritium accumulation in the samples.

**Acknowledgment:** This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.

#### **Corresponding Author:**

Aigars Vītiņš aigars.vitins@lu.lv Laboratory of Radiation Chemistry of Solids, Institute of Chemical Physics, University of Latvia, Four-storey building, Room 108, 32 Miera Street, Salaspils, LV-2169, Latvia

14 September 2022

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### BeWS-15, Karlsruhe, Germany, 14 September 2022 4 The JET ITER-like wall (ILW) • Limiters and an upper dump plate of the JET vacuum vessel are castellated bulk beryllium tiles. • One of the scientific objectives of the ILW experimental campaigns is to demonstrate sufficiently low fuel retention and, in particular, to demonstrate ITER-relevant tritium retention mitigation. • The aim of the present study is to assess tritium accumulation in the beryllium tiles from the three ILW campaigns (2011-2012, 2013-2014, 2015-2016) and to investigate its thermal desorption patterns with respect to development of possible detritiation techniques.


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BeWS-15, Karlsruhe, Germany, 14 September 2022


BeWS-15, Karlsruhe, Germany, 14 September 2022

7

### Gas exchange time of the tritium TPD setup • Quartz tube of 156 cm3: 72 cm3 - before the sample; 59 cm3 - after the sample; 25 cm3 – the zinc compartment. • Column of 74 cm3 for silica gel. • Cold trap of 66 cm3. • Gas filter of 20 cm3. • Volume of these items of the tritium TPD setup together is about 320 cm3. • Purge gas He + 0.1% H2 14-15 L/h or about 250 cm3/min. • The gas exchange time of the tritium TPD setup itself is about 1.3 min. • Operating volume of tritium detector DDH 32 is 300 cm3. • Purge gas He + 0.1% H2 14-15 L/h or about 250 cm3/min. • Counting gas Ar + 10% CH4 42-45 L/h or about 750 cm3/min. • The gas exchange time of the tritium detector itself is about 18 seconds. • According to the data of the manufacturer of DDH 32 ("MAB", Germany), there will be 62% of the fresh gas mixture in the detector after 18 s, but 94% will be after 54 s. • Therefore, the released tritium activity was measured repeatedly with the measuring time of 2 min.

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10

## Rationale of the tritium TPD experiments


BeWS-15, Karlsruhe, Germany, 14 September 2022

# Outline


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13

### TPD of tritium from quarters 1-3 of ILW1-3 Be sample 691 (2011-2016: WPL 4D15: R6-C2 of RH wing) (2/2)


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25

### BeWS-15, Karlsruhe, Germany, 14 September 2022 26 Conclusions about IWGL 2XR10 samples 1. For the inner wall samples investigated in the present study, the surface concentration of tritium gradually decreased in the three ILW campaigns. The maximum of 83E11 atoms / cm2 was found for the LH Wing sample of ILW1 (2011-2012), and the minimum of 6E11 atoms / cm2 was for the RH Wing sample of ILW3 (2015- 2016). The temperature of 50% detritiation was 697 K and 880 K for these two samples respectively. 2. The samples had quite different thermal desorption spectra of tritium indicating different tritium accumulation in the samples. The ILW2 (2013-2014) samples had by 17-43 K higher temperature of 50% detritiation than the respective ILW1 (2011-2012) samples. For ILW1 and ILW2 samples, the temperature of 50% detritiation gradually decreased from the centre towards the wings by 24-88 K. For all the samples, the temperature of 50% detritiation was in the range of 697-880 K.

## Conclusions about Upper Dump Plate 2BC2 samples


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27


# Acknowledgements


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29

### **A Study on Technician Variability in Wipe Sampling for Beryllium & Potential Contributions to Robotic Sampling Equipment**

Eilish McKEON<sup>1</sup> , Amy BLEASDALE<sup>1</sup> , Christopher DORN<sup>1</sup> , Maddy KEARNEY<sup>1</sup> , Jon VERDON<sup>1</sup> , and Heather LEWTAS<sup>1</sup>

*<sup>1</sup>UK Atomic Energy Authority, Culham Centre for Fusion Energy, Abingdon, U.K.*

The Joint European Torus (JET) is the largest operational magnetically confined fusion research reactor in the world, beginning operations in 1983. Beryllium (Be) was first installed in JET in 1989, which gives the UKAEA a high level of experience and expertise in the safe handling, storage, and disposal of this high-performance but toxic material.

The UKAEA's Health Physics Group (HPG) is comprised of a team of trained technicians and analysts who support an on-site Be handling facility and operate a Be analysis laboratory, in which accredited processes are certified to UKAS ISO 17025. The laboratory capabilities enable the UKAEA to carry out technical studies involving beryllium in a controlled, safe environment by using established air and surface sampling techniques to monitor conditions over a study's specified experiments and any related activities.

To date, surface sampling for beryllium, known on the Culham site as "smearing", has been performed by HP technicians trained in defined, manual processes. Others' studies in recent years have looked at collection efficiencies between the sampling media being used and the type of beryllium compound being sampled. Given the manual nature of the task, natural variations occur from person to person, and building on the aforementioned research, UKAEA has decided to investigate the impact of force exerted by a technician during the sampling process.

This presentation gives an overview on the history of beryllium use at JET, surface sampling processes used on site, a summary of the pilot study in progress, and how these manual studies are contributing to further research into robotic sampling equipment in challenging environments.

#### **Corresponding Author:**

Ms. Eilish McKEON eilish.mckeon@ukaea.uk Postal address: UK Atomic Energy Authority Culham Centre for Fusion Energy Abingdon, Oxfordshire OX14 3DB, United Kingdom

### Presentation Plan

	- ❖ *Presenter: Chris Dorn*
	- Overview of the Joint European Torus (JET)
	- History of Beryllium Use in JET
	- Sampling Programme at UKAEA

#### • **Part 2: Technician Variability Study**


#### • **Part 3: Robotic Surface Wiping**

	- Tool Design & Trials
	- Future Work

*Culham Centre for Fusion Energy, Abingdon, Oxfordshire, UK*

### Part 1: Introduction / Background

**Beryllium Wipe Sampling Variability Study – BeWS-15 & BeYOND-IX – Sep-2022**

Presenter: CHRIS DORN

#### • 1978: JET construction starts • 1983: First plasma in JET • 1989: First installation of beryllium in JET • 1991: First experiments with tritium Overview of the Joint European Torus (JET) 4 • 1997: High-performance, full deuterium-tritium experiments. JET achieves world record fusion power of 16.1MW. • 2009-10: JET installs ITER-Like Wall: beryllium & tungsten • 2019-20: Preparations well advanced for new deuteriumtritium experiments, designed to sustain high fusion performance for longer periods. *Above: Beryllium RF Antenna Tiles Production Sequence [2][3] Left: JET Reactor c.1982-83 Right: JET Reactor c.1990 with Be Belt Limiters Installed* [1]

### Sampling Programme at UKAEA

	- Transfer of items from one designated area to another
	- Clearance of items from designated areas entirely
	- Action levels: trigger the creation of designated areas

### Part 2: Technician Variability Study

Presenter: EILISH McKEON

### Background: Assumptions, Impetus & Prior Work

#### • **Original Baseline Assumption**

	- Long-term study with surface sampling as integral part
	- Due to staff changes over the duration of the study, several technicians were used to collect surface samples
	- Results did not always consistently follow expectations
	- Damjanovic, 2019: *Comparison of Collection Efficiencies of Whatman No. 1 Filter Paper and Ghost Wipes for Loose Beryllium Surface Contamination*  (presented at BeWS-14) [6] *Glass plates spiked with beryllium-containing solution in known*

*amounts were used in the previous study [6]*

#### Purpose & Scope

#### • **Purpose of the Study**

	- Procuring pressure-measuring equipment
	- Manufacturing a pressure rig (test fixture)
	- Spiking a defined surface with known concentrations of beryllium
	- Measuring pressure ranges exerted by a number of individual technicians
	- Measuring the amount of beryllium detected
	- Determining the results are correlated and if so, determining an uncertainty factor which can be applied to analytical results incorporating the uncertainty from the sampling and analytical processes. (Analytical uncertainty is already accounted for.) *Pressure rig (test fixture) built at UKAEA and used in the*
	- **Beryllium Wipe Sampling Variability Study BeWS-15 & BeYOND-IX Sep-2022** 9

*study (the force gauge is not installed in this view). Design credit: Materion Corp., Ohio, USA Shown here with kind permission [7]*

### Project Outline & Milestones

#### Deliverables have been defined for the project as follows:

#### • **Preparatory Stage**


#### • **Experimental Stage**


#### • **Reporting Results**

	- Consider Investigation of Other Variables

*temporary controlled area set up in the Analytical Laboratory at UKAEA expressly for the purposes of this study*


### SESSION 4

# Beryllides

#### **Overview of activities in Kazakhstan related to study of beryllium and beryllium compounds**

A. Shaimerdenov1,2, T. Kulsartov2,3 , I. Kenzhina1,2,4\*, Zh. Zaurbekova2,3 , Y. Kenzhin<sup>4</sup> , Y.V. Chikhray<sup>2</sup> , S. Udartsev<sup>5</sup>

*<sup>1</sup>The Institute of Nuclear Physics, Almaty, Kazakhstan - 1 <sup>2</sup>al-Farabi Kazakh National University, Almaty, Kazakhstan - 2 3 Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Kurchatov, Kazakhstan - 3 <sup>4</sup>Kazakh-British Technical University, Almaty, Kazakhstan - 4 <sup>5</sup>Ulba Metallurgical Plant, Ust-Kamenogorsk, Kazakhstan - 5*

Kazakhstan is one of the world leaders in the production of beryllium and beryllium products. Since recently, the Ulba Metallurgical Plant has set up production of beryllium-based intermetallic compounds; in particular, beryllides of titanium, chromium, molybdenum, etc. are produced. Beryllides are candidate materials for fusion plants as a neutron multiplier. In addition, beryllides are considered as a material for hydrogen storage. On the basis of research reactors in Kazakhstan, studies are being conducted on the effects of neutron irradiation on the properties of beryllium and beryllides. This paper provides an overview of research programs in Kazakhstan to study the radiation resistance of metallic beryllium of different grades and beryllides.

#### **Corresponding Author:**

Dr. Inesh Kenzhina kenzhina@physics.kz 99/1 Nauryzbay batyra str., 94 app., 050000, Almaty,

Kazakhstan.

### **IGR RESEARCH REACTOR**


### **IVG1.M RESEARCH REACTOR**


### **WWR-K RESEARCH REACTOR**


### **EXPERIMENTAL POSSIBILITIES OF WWR-K**

### **Two kind of hot cells:**

❑concrete shielding (5 cells) ❑steel shielding (4 cells)

### **Critical assembly**

Zero power reactor Maximum thermal power: 100 W Reflector: deionized water and/ or beryllium Moderator: deionized water Fuel composition: UO2+Al; Enrichment in U-235: 19.7 % (since 2012) Max. thermal neutron flux density: 109 cm-2 s-1 Diameters of experimental channels: 65, 96 and 140 mm.

**Gas-vacuum loop facility** 

Gas-vacuum Loop Facility is designed to create a vacuum or gas environment in a sample during in-pile test

### **Hydraulic transfer system**

Utilized to load and unload of irradiation ampoule to/from the core during reactor operation

#### **Pneumatic transfer system**

Installed at beam tube #3, allows transfer of ampoule with a sample from laboratory room to the core and return back for gamma spectrometry measurement (neutron activation analysis)


### **Gas release results (2)**

Release of tritium and helium for samples S-200F, I-220H - the main amount of gas (both helium and tritium) is released close to the melting temperature and during melting (while in previous experiments with samples irradiated to a lower fluence, noticeable amounts of released tritium were observed in the temperature range of 900-1150 C).

Beryllium sample of grade S-65H, a noticeable amount of gas is released up to the melting point (with a peak for tritium in the region of about 1100 C). (A similar picture was observed with samples irradiated to a lower fluence).

*Temperature dependencies of tritium flux from beryllium*

### **Plans for study of radiation resistance of titanium beryllide (1)**

In this regard, Kazakhstan has great opportunities, since on its territory Ulba Metallurgical Plant (UMP) is located (a unique enterprise engaged in the production of beryllium, uranium and tantalum products for the needs of nuclear energy). UMP employees have developed a technological method for obtaining titanium beryllide in large quantities by vacuum hot pressing→ **Mr.Udartsev was reported.**

Titanium beryllide is a candidate material for a neutron breeder of a fusion reactor blanket. This is due to its improved mechanical properties compared to metallic beryllium.

Project named as "Influence of reactor irradiation on physical and mechanical properties and gas generation in titanium beryllide" (grant #AP14871445) has been accepted by Kazakhstan government for funding

### **Beryllides as advanced materials for neutron multiplication**

#### R. Gaisin<sup>1</sup> , R. Rolli, V. Chakin, P. Vladimirov

#### *<sup>1</sup> Karlsruhe Institute of Technology, Institute for Applied Materials - Applied Materials Physics, 76344 Eggenstein-Leopoldshafen, Germany*

The neutron multiplier is an essential component of the blanket of future thermonuclear reactors, which should provide the tritium breeder with a sufficient amount of neutrons of a certain energy. Among all chemical elements, only beryllium and lead have an advantageous ratio of high neutron multiplication reaction at low neutron absorption rates. However, pure metals – beryllium and lead – for various reasons cannot be used in the harsh operating conditions of a fusion reactor blanket. Intermetallic compounds of beryllium – beryllides have a number of advantages over pure beryllium and are currently considered to be the reference neutron multiplication material for the Helium Cooled Pebble Bed (HCPB) breeding blanket concept of EU DEMO fusion reactor. Recently, a batch of full-size beryllide blocks has been manufactured on an industrial scale in cooperation with the Ulba Metallurgical Plant. The present work is devoted to the characterization and analysis of these beryllide blocks so that the material could be used for the manufacture of a blanket.

Titanium beryllide (TiBe12) blocks are hexagonal prisms with an internal hole, while chromium beryllide (CrBe12) blocks are solid prisms of complex shape. The resulting blocks have a single-phase structure of the corresponding beryllide with a small impurity in the form of beryllium oxide. One of the titanium beryllide blocks also has about 7% residual beryllium phase. Grains of titanium beryllide have an average size of about 7–8 µm, while grains of chromium beryllide are much larger and reach 40– 50 µm. Mechanical compression and bending tests of beryllides showed their very high strength, which is maintained up to 1000°C. In terms of specific compressive strength, the single-phase TiBe<sup>12</sup> surpasses all materials, except diamond, in the 700– 1000°C temperature range. Chromium beryllide and titanium beryllide with 7% of the beryllium phase have lower strength, but higher ductility. Corrosion tests were carried out in air and in He + 2% water vapor at 800–1200°C. Beryllides have high corrosion resistance similar to Ni-base superalloys and high temperature ceramics. Long-term thermal cycling tests with rapid heating and cooling, simulating operation in a fusion reactor, showed high resistance of beryllides to thermal shocks. The results obtained are also discussed from the point of view of the application of beryllides in other areas.

### **Corresponding Author:**

Dr. Ramil Gaisin ramil.gaisin@kit.edu Herrmann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen


### **19** 14.09.2022 Dr. Ramil Gaisin Institute for Applied Materials – Applied Materials Physics **Requirements for beryllides as neutron multiplier materials** • High tritium breeding ratio – maximum possible content of Be • Minimum content of impurities that form long-lived isotopes under irradiation (e.g. Uranium) • Fine grain structure for easy tritium release and reasonable mechanical properties • Beryllide blocks must retain their shape and not fracture during operation. • Low corrosion in the air and purge gas atmosphere • Low interaction with structural materials (e.g. EUROFER steel) • No fracture or cracks during rapid heating/cooling due to pulsed operation of DEMO **/44**

152

**/44**

**42** 14.09.2022 Dr. Ramil Gaisin Institute for Applied Materials – Applied Materials Physics

### **Mechanical properties of titanium beryllium intermetallic compounds**

Taehyun HWANG<sup>1</sup> , Jae-Hwan KIM<sup>1</sup> , Yutaka SUGIMOTO<sup>1</sup> , Suguru NAKANO<sup>1</sup> , Yoshiaski AKATSU<sup>1</sup> , Msaru NAKAMICHI<sup>1</sup> ,

*<sup>1</sup>Fusion Energy Research and Development Directorate, National Institutes for Quantum Science and Technology, Rokkasho, Aomori 2-166, Japan* 

Beryllium (Be) has been a promising functional material for the breeding blanket of a fusion reactor. However, Be has disadvantages such as the sharp increase of swelling above 600°C and deterioration for mechanical properties and thermal conductivity due to irradiation damage. Be intermetallic compounds (beryllide) such as Be12Ti and Be12V are considered as promising advanced neutron multipliers for DEMO instead of Be, because of its excellent high-temperature stability and low reactivity with water vapor. Besides, recently, there has been a growing interest in the usage of beryllide block instead of beryllide pebble. However, the evaluation of the mechanical properties of beryllide has not been sufficiently performed. Thus, the aim of this study is to investigate the tensile properties of beryllide.

A mixed powder of Be-7.7at.% Ti, which is the stoichiometric value of Be12Ti, was subjected to a single-phase treatment at 1200°C for 24 hours in an argon atmosphere. Also, Be12Ti powder which has been crushed using a planetary ball mill at 300 rpm for 5 hours was prepared for comparison of the effect of refinement. For the fabricating samples, plasma sintering (KE-PAS II, manufactured by Kaken) was conducted using beryllide single-phase powder. Tensile tests were performed on various temperatures with sub-sized tensile specimens, which fabricated by electric discharge machining.

Under the above-mentioned sintering conditions, Be12Ti single-phase berylide was obtained, and the sintering density of Be12Ti was almost 100% above 1000°C. Detail of tensile properties of beryllides will be reported.

### **Corresponding Author:**

Dr. HWANG Taehyun hwang.taehyun@qst.go.jp *Fusion Energy Research and Development Directorate, National Institutes for Quantum Science and Technology, Rokkasho, Aomori 2-166, Japan*

1

2

# **Mechanical properties of titanium beryllium intermetallic compounds**

Taehyun HWANG1 , Jae-Hwan KIM1, Yutaka SUGIMOTO1, Suguru NAKANO1 , Yoshiaski AKATSU1, Msaru NAKAMICHI1 ,

National Institutes for Quantum Science and Technology

In this presentation ,unfortunately, only published and opened materials will be presented because of export classification's aspect.

*establishment of the material database for DEMO design.*

*T. Hwang et al., Nuclear Materials and Energy 30 (2022) 101117*

*T. Hwang et al., Nuclear Materials and Energy 30 (2022) 101117*

5

# Summary

The mechanical properties of Be12Ti fabricated by plasma sintering were investigated to provide important properties data for the material database for beryllide blocks as advanced neutron multipliers.

From the experimental results, the following conclusions can be drawn:


### SESSION 5

# Modeling & experimental validation

### **Investigation of radiation damage effects in beryllium: updates on recent results obtained on proton, neutron and He-ions irradiated samples**

#### Slava Kuksenko

#### *United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, OX14 3DB, UK*

Beryllium is an essential material for a wide variety of application such as operating and future nuclear facilities including material testing nuclear fission reactors, fusion energy experimental and future commercial reactors, target component materials in currently running and near-future multi-megawatt particle accelerator sources. Therefore, beryllium and beryllium-based alloys are under extensive investigation by nuclear facility communities.

The paper gives an overview of the recent results obtained in the UK Atomic Energy Authority on beryllium samples, particularly during collaborative studies with the Karlsruhe Institute of Technology (Germany), the University of Horsfield (UK) and the international Radiation Damage In Accelerator Target Environments (RaDIATE) collaboration and include:


The paper will also give an overview of the future plans.

### **Corresponding Author:**

Dr. Viacheslav Kuksenko slava.kuksenko@ukaea.uk United Kingdom Atomic Energy Authority Culham Science Centre, Abingdon, OX14 3DB, UK


### **First principles simulation of resistivity recovery in irradiated beryllium**

#### Christopher Stihl<sup>1</sup> , Pavel Vladimirov<sup>1</sup>

#### *<sup>1</sup>Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics*

Future fusion devices like ITER or DEMO require closed fuel cycles. These vitally depend on neutron multiplying materials as part of a breeding blanket module like beryllium pebbles in the European Helium-Cooled Pebble-Bed. During operation the beryllium pebbles will accumulate point defects, tritium, and helium due to inevitable exposure to highly energetic neutron irradiation as emitted by the fusion plasma. A detailed knowledge of the characteristics of point defects is decisive for reliable simulations of microstructure evolution in irradiated beryllium. Such models are a prerequisite for predicting tritium inventory during operation as well as after the blanket´s end of life since tritium retention and release is the paramount safety concern.

A well-established experimental approach to assess the dynamics of relevant atomic defects consists in measuring electrical resistivity recovery (ERR) after irradiation during annealing. Within this approach, temperatures corresponding to electrical recovery steps are correlated with activation energies which are associated with different type of reactions between defects.

In this work, results of our ongoing efforts to model and understand the ERR of

beryllium are presented. To that end, we introduce a rate equation-based approach to model ERR spectra (see picture below) utilizing density functional theory results as input. Within this approach, electrical resistivity recovery models comprising the volume of spontaneous recombination of monovacancies and self-interstitial atoms in beryllium as well as various additional defects are considered. As a result, an intricate interplay between different defect dynamics is

uncovered, suggesting a clear route for further research to obtain systematically improved electrical resistivity recovery models.

#### **Corresponding Author:**

Dr. Christopher Stihl Christopher.stihl@kit.edu Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen

### **Radiation induced formation gas bubbles in beryllium after neutron irradiation up to 6000 appm helium production**

Michael Klimenkov, Ute Jäntsch, Vladimir Chakin, Nikolai Zimber and Pavel Vladimirov

### *Karlsruhe Institut of Technology, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen, Germany*

The current interest in mechanical properties and microstructure of neutron irradiated beryllium refers to its planned application in the Helium-Cooled Pebble Bed (HCPB) European concept of a breeding blanket of DEMO. Irradiation experiments in highneutron flux nuclear research reactors yield information about microstructural evolution of beryllium under conditions relevant to fusion (temperature, damage dose, helium and tritium productions) excluding 14 MeV neutrons impact which is not present in the neutron spectra of fission reactors. The HIDOBE-02 irradiation campaign accomplished at the HFR, Petten corresponds to 1246.5 Full Power Days at a reactor power level of 45 MW in the temperature range from 410°C to 680 °C. Transmission electron microscopy (TEM) has been to study the evolution of voids during neutron irradiation at different temperatures. The target preparation of specimens was performed using focused ion beam (FIB).

TEM study shows the formation of radiation induced hexagonal flat gas bubbles inside the grains, however at the lowest irradiation temperature of 410° the pebbles show the uniform shape. The diameters of the bubbles increase from a few nanometers for 410°C to more than hundred nanometers for 680 °C. The number density of bubbles decreases, correspondingly, by more than two orders of magnitude. The preferable formation of bubbles along the grain boundary and dislocation lines was observed. Analytical investigations using electron energy loss spectroscopy show the presence of He and H23 inside bubbles. Also the Si and Fe segregation on the voids was detected [2].

EDX mapping shows that the precipitates inside the grains and on the GBs have increased iron and aluminum content, indicating the formation of an Fe-Al-Be phase. In the material irradiated at 440°C, most of the precipitates also have Fe-Al-Be composition, while several other single- and multiphase precipitates were found. The Fe-Al-Be phase is observed as 10-15 nm precipitates within the grains and as 200 nm particles bound to a gas bubble at the GB. The present study shows detailed microstructural changes induced by neutron irradiation in beryllium.

[1] M. Klimenkov, et al. Journal of Nuclear Materials 455 (2014) 660–664 [2] M. Dürrschnabel, et al. Scientific Reports, 11, 7572 (2021)

### **Corresponding Author:**

Dr. Michael Klimenkov michael.klimenkov@kit.edu Karlsruhe Institut of Technology, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen, Germany

### **Effect of impurities on microstructural evolution under irradiation in beryllium**

P.V. Vladimirov<sup>1</sup> , D.V. Bachurin<sup>1</sup> , C. Stihl<sup>1</sup> and N. Zimber<sup>1</sup>

#### *<sup>1</sup>Karlsruhe Institute of Technology, Institute for Applied Materials - Applied Materials Physics, 76344 Eggenstein-Leopoldshafen, Germany*

Impurities are known to affect mechanical properties of beryllium, but their effect on development of irradiation induced microstructure is still unknown. In this contribution we are making further attempt to reveal behavior of impurities in neutron irradiated beryllium pebbles by using both analytical transmission electron microscope (TEM) and first principles computer simulations.

TEM studies have revealed Al-Fe-Be precipitates, complex multiple phase precipitates, homogeneous segregations of elements to grain boundaries as well as abundant precipitation along dislocations. All precipitates are richly decorated with helium bubbles which are smaller in size than typical bubbles inside grains. Precipitate-free and helium-bubble-free zones were observed along grain boundaries.

Using density functional theory approach, we have calculated interaction of typical solutes found in beryllium, namely, Al, Fe, Cr, Mg and Si with vacancies, interstitials and free surfaces which can simulate a surface of helium bubbles. Interesting correlation has been revealed: an impurity which has attractive binding with a vacancy has also positive affinity to free surface. In particular, Al, Mg and Si are strongly bound with vacancies and also attracted by the free surfaces. This result is supported by the EDX measurements, (see Fig. above) which reveal decoration of He bubbles with Al, Si and Mg, while Fe is homogeneously distributed. Those impurities which repulse vacancies are attracted by self-interstitials, however, no correlation with the formation volume of respective substitutional atoms was found in this case.

### **Corresponding Author:**

Dr. Pavel Vladimirov, pavel.vladimirov@kit.edu Karlsruhe Institut of Technology, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen, Germany

### **Ab initio study of hydrogen behavior in titanium beryllides**

D.V. Bachurin<sup>1</sup> , C. Stihl<sup>1</sup> , P.V. Vladimirov<sup>1</sup> *<sup>1</sup>Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany*

An interest in titanium beryllides as candidate materials for advanced neutron multiplier for the Helium Cooled Pebble Bed breeding blanket of European DEMO fusion reactor is related to their lower tritium retention, lower swelling and higher oxidation resistance in comparison with pure beryllium. The latter was initially suggested as neutron multiplier in the International Thermonuclear Experimental Reactor (ITER) and for the above reasons has a number of limitations compared to beryllides.

One of the most important questions is how much weaker tritium, which is formed as a result of the interaction of high-energy neutrons with the pebbles is bound in titanium beryllides in contrast to pure beryllium. Such an interaction awakes formation of helium bubbles and degradation of the material properties. One of the main promising methods for studying the behavior of hydrogen in titanium beryllides is firstprinciples modeling technique based on density functional theory.

The present work is devoted to *ab initio* study of hydrogen (isotope effects were neglected and hydrogen was considered instead of tritium) behavior in three titanium beryllides (Be2Ti, Be17Ti2, Be12Ti). All of them have different crystal structure and contain a different number of crystallographically non-equivalent interstitial hydrogen sites.

Both the hydrogen solution energy in defect-free lattice and binding energy with a vacancy are important characteristics in terms of tritium dissolution, retention and release. Static *ab initio* calculations demonstrate that hydrogen solution energy in all interstitial non-equivalent sites is noticeably lower as compared with pure beryllium suggesting an easier dissolution of hydrogen atoms in titanium beryllides. Computation of binding energy of single hydrogen atom with all non-equivalent monovacancies reveals that hydrogen might be trapped by a vacancy without being inside it. The obtained results sheds light on the understanding of earlier tritium release in different titanium beryllides during thermo-desorption experiments and expand our knowledge of their properties.

#### **Corresponding Author:**

Dr. Dmitry Bachurin dmitry.bachurin@kit.edu Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

#### **Nanoscale characterization of beryllide materials**

#### M. Duerrschnabel<sup>1</sup> , R. Gaisin<sup>1</sup> , P. Valdimirov<sup>1</sup> , M. Rieth<sup>1</sup>

#### *1 Institute of Applied Materials- Applied Material Physics, Herrmann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen*

The most recent version of the Helium Cooled Pebble Bed (HCPB) foreseen for the European DEMO blanket considers solid blocks of titanium beryllide as neutron multiplicator material. The advantage of beryllide materials over pure beryllium is their higher operating temperature, higher corrosion resistance, lower swelling, and retention of tritium under neutron irradiation. Understanding the micro- and nanostructure especially after neutron irradiation is of crucial importance for the qualification process of the material.

The focus of this work will lie on the transmission electron microscopy (TEM) characterization of a titanium beryllide/beryllium composite material irradiated at two different temperatures during the HIDOBE neutron irradiation campaign. In particular, the structure and chemistry of the nanosized cavities in the pure beryllium region and the beryllide region was analyzed and is compared to each other. Apart from the cavities, structural defects were observed in the beryllide region that are not known from irradiated pure beryllium.

The presented results can be used for understanding and quantifying for example tritium retention in beryllide materials and to further optimize the material synthesis and the breeding blanked design in general.

#### **Corresponding Author:**

Dr. Michael Duerrschnabel michael.duerrschnabel@kit.edu

Institute of Applied Materials Applied Material Physics Herrmann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen


### SESSION 6

# Mechanical properties & irradiation damage

### **Mechanical compression behaviors and microstructure change under He ion irradiation of single phase Be and binary Be12Ti pebbles**

Pingping Liu1,\*, Qian Zhan<sup>1</sup> , Wen Hu<sup>1</sup> , Yumei Jia<sup>1</sup> , Farong Wan<sup>1</sup>

*<sup>1</sup> School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083, China*

The world urgently needs a carbon-free, safe, clean and limitless energy source. Fusion energy has the potential to meet this need. International thermonuclear fusion experimental reactor (ITER) and future commercial power demonstration reactor (DEMO) were designed and developed by scientists around the world. However, one key problem of "tritium fuel self-sustaining" needs to be solved firstly. Thus, tritium breeding blanket module (TBM) was developed. Beryllium (Be) pebbles with a diameter of 1 mm are planned to be used as a neutron multiplier in the helium-cooled ceramic breeder (HCCB) test TBM of ITER, which is also the primary option of the Chinese TBM program. Meanwhile, beryllium intermetallic compounds (beryllides) such as Be12Ti are the most promising advanced neutron multipliers in advanced DEMO fusion reactors because of high melting point and high stability at high temperatures.

In this study, in the prospect of establishing a database of advanced neutron multipliers for fusion reactor, the preparation, mechanical properties test and irradiation damage of Be and beryllides pebbles were carried out. The process of producing pure Be and Be-Ti alloy pebbles can be roughly divided into two stages: fabrication of Be and beryllide electrode rods and granulation of Be and beryllide pebbles. The Be-Ti beryllides synthesized by hot isostatic pressing (HIP) with a pressure of 230 MPa at 1073 K for 2 h. The Be and Be12Ti granulation process by rotating electrode process (REP). Then, surface analysis, mechanical compression and irradiation were performed to evaluate surface microstructure, mechanical properties and irradiation resistance.

Be pebbles with diameter of 1 mm and Be12Ti pebbles with diameter of 0.7 mm were successfully fabricated by combining HIP and REP methods. According to the XRD results, the phase composition of the Be pebbles was identified as Be with little BeO. The phase composition of the as-granulated Be-7.7 at.%Ti pebbles was identified as Be12Ti with little Be and BeO. The AFM and SEM results revealed that Be and Be12Ti pebble were well shaped with small surface fluctuates. During mechanical compression tests, Be pebbles exhibited very good ductility (no fracture at 50% deformation), which is better than that of Be12Ti pebbles. Rupture of Be12Ti pebbles occurred at the deformation beyond 10%. After high dose helium ion irradiation, the surface blistering and internal damage structure were examined and analysed by SEM and TEM, which may provide experimental basis for good understanding of irradiation damage mechanism and design optimization of advanced neutron multipliers for the fusion reactor.

#### **Corresponding Author:**

Dr. Pingping Liu ppliu@ustb.edu.cn University of Science and Technology Beijing, No. 30 Xueyuan Road, Haidian District, Beijing

### **Outline**


### **Background**


### **Neutron Multiplier of DEMO --High beryllium content --Low activation --High melting point**


### **Background**

### **Initial Motivation͵**

### **1、Fabrication of high-melting-point neutron multiplier pebbles.**

### **2、Evaluation of irradiation resistance of the pebbles.**

D-T fusion produces an extremely energetic (14.1-MeV) neutron, which is potentially useful for breeding more tritium but also creates challenges for the materials (**Irradiation damage**).

### **Outline**


### **Fabrication of neutron multiplier pebbles**


### **Outline**


### • **Summary**

**K. Tsuchiya, E. Ishitsuka, H. Kawamura, et al. Contact Strength Evaluation Of Irradiation Beryllium Pebbles, 7th IEA International Workshop on Beryllium Technology, 2008.**

### **Outline**



### **Summary**

**1) Be, BeTi and BeW pebbles could be fabricated by REP.**

### **2) Compression properties of Be and Be12Ti pebbles were evaluated by mechanical compression.**


**3) Be, BeTi and BeW were irradiated by high dose He-ion beam.** ----The surface swelling-resistance of BeTi and BeW is better than beryllium.

#### **Creep of beryllium pebbles after neutron irradiation to 6000 appm helium production**

Vladimir Chakin<sup>1</sup> , Rolf Rolli<sup>1</sup> , Milan Zmitko<sup>2</sup>

*1 Institute for Applied Materials - Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany*

*<sup>2</sup>Fusion for Energy, c/ Josep Pla, nº 2, Torres Diagonal Litoral Edificio B3, 08019 Barcelona, Spain*

Beryllium pebbles with 1 mm diameter are the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) blanket of ITER. High energy fusion neutrons cause swelling of the beryllium pebbles at the HCPB operation temperatures to 923 K. The radiation-induced swelling of beryllium as well as different thermal expansions of the beryllium pebbles and the structural material can cause the high thermal stresses in the pebble bed. Thermal creep of the pebbles should reduce the stresses because the relaxation. Neutron irradiation leads to degradation of mechanical properties, what expresses in the hardening and the embrittlement. This radiation effect hinders the effect of the relaxation.

In this study, creep properties of beryllium pebbles with 1 mm diameter produced by Rotating Electrode Method (REM) at NGK, Japan were studied. These beryllium pebbles were irradiated in the HFR, Petten, the Netherlands, at temperatures of 643, 723, 833, 923 K to 6000 appm helium production. The creep tests of individual pebbles were performed at temperatures which were equal to the irradiation temperatures by using of three different loadings per each temperature. For two lowest irradiation temperatures of 643 and 723 K, no creep effect was observed. The radiation hardening only occurs that manifests itself in significant reduction of the pebble deformation under loading. At higher irradiation temperatures of 833 and 923 K, the creep rates have significant values. The creep rates strongly depend on the testing temperatures and the loadings. At high irradiation temperatures the ability of beryllium pebbles to the significant deformation under applied loadings should provide the complete relaxation of the internal stresses in the beryllium pebbles.

### **Corresponding Author:**

Dr. Vladimir Chakin vladimir.chakin@kit.edu

Institute for Applied Materials - Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

### **Irradiation parameters of Be pebbles in HIDOBE-02 compaign**

=


**Institute for Applied Materials – Applied Materials Physics**


### **Placement of Ø 1 mm Be pebbles in HIDOBE-02**

Unconstrained Be pebbles are placed in capsule with Ø 3.2 mm. The pebbles were filled free in the capsule.

Constrained Be pebble beds are placed in capsule with Ø 14.9 mm. It was filled by pebbles and then, screwed on with screw cap with torque wrench. In particular, after contact of the cap with the pebbles, another forced turn of the cap along the thread was done. In this position, the cap was spot welded to the capsule. In this way, additional internal stresses were created in the pebbles.

This allows to simulate internal state of the pebbles swelled after irradiation to higher neutron dose than that in HIDOBE-02 experiment.

**Institute for Applied Materials – Applied Materials Physics**


# **Conclusions**

Creep tests of Ø 1 mm Be pebbles irradiated to 6000 appm He production show difference of creep behavior of unconstrained (free filled) and constrained pebbles.

There are points:


This means that constrained pebbles have comparatively better creep behavior than unconstrained pebbles especially for irradiation at higher temperatures than 0.5Tm.

**Institute for Applied Materials – Applied Materials Physics**

**Institute for Applied Materials – Applied Materials Physics**

### **Thermo-mechanical behavior of titanium beryllide pebble beds at elevated temperatures**

J. Reimann<sup>1</sup> , E. Cilingir<sup>2</sup> , R. Gaisin<sup>1</sup> , A. Goraieb<sup>2</sup> , M. Nakamichi<sup>3</sup> , P. Vladimirov<sup>1</sup>

*1 Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany <sup>2</sup> Karlsruhe Beryllium Handling Facility (KBHF GmbH), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany <sup>3</sup> National Institutes for Quantum and Radiological Science and Technology, 2-166 Omotedate, Obuchi, Rokkasho, Aomori 039-3212, Japan*

The thermomechanical behavior of titanium beryllide pebble beds was investigated experimentally at temperatures between 200 and 500°C in helium atmosphere at atmospheric pressure. The pebbles consist of a mixture of TiBe<sup>12</sup> and Ti2Be17 titanate beryllide phases and a small residual amount of Be phase, denominated as Be-7.7Ti.

Like previous experiments at ambient temperature [1], the pebble beds were compressed uniaxially up to 4.5MPa and the effective thermal conductivity k was measured using the hot wire technique.

Compared to ambient temperature, the stress-strain curves do not differ significantly in investigated temperature range. Because the thermal conductivity of solid TiBe<sup>12</sup> is fairly constant in a wide temperature range [2], k increases moderately with increasing temperature because of the increasing thermal conductivity of helium.

Compared to beryllium pebble beds, the k of the Be-7.7Ti pebble beds increases again much lesser because of the significantly smaller thermal conductivity of the solid material and the mechanically harder behavior resulting in smaller contact surfaces.

[1] J. Reimann et al, Fusion Eng. Des. 165 (2021) 112249

[2] M. Uchida, E. Ishitsuka, H. Kawamura, Fusion Eng. Des. 69 (2003) 499-503.

### **Corresponding Author:**

Dr. Joerg Reimann reimann-langhans@web.de

### APPENDIX A

# Workshop photos

Figure 1: Group picture of the BeWS-15 participants (courtesy of J. Reimann). From left to right:

1st row: S. Udartsev, A. Goraieb, M. Klimenkov, E. McKeon, M. Kearney, J.-H. Kim, R. Pearson, P. Vladimirov, J. Reimann, I. Kenzhina, D. Radloff, M. Ionescu-Bujor, K. Hesch, A. Renier, H. Huang;

2nd row: S. Kuksenko, S. Kovalskiy, J. Verdon, C. Baus, G. Kizane, K. Ashley, P. Sp¨ah, P. Dvor´akov´a-Ruskayov´a, M. Rubel, T. Knudson, K. Smith, R. Gaisin; 3rd row: M. D¨urrschnabel, A. M¨oslang, M. Rieth, R. Stieglitz, D. Bachurin, A. Frehn, A. Shaimerdenov, Y. Frants, P. M¨ahlmann, M. Voß, K. Zenkov, D. Sioui, A. Vitins, T. Scherer, C. Dorn, L. Whalen, L. Vandermark, R. Jilek, L. Toupal.

Figure 2: Grouppicture of the BeWS-15 participants without J. Reimann, who was taking this photo (courtesy of J. Reimann)

Figure 3: Workshop dinner in Hoepfner Schalander Hall

Figure 5: Awarding the Prof. Glen Longhurst Award to Christopher Dorn during the Workshop dinner in Hoepfner Schalander Hall

Figure 6: Guidedtour of the Hoepfner Burghof Brewery

Figure 7: Groupphoto (from left to right): R. Gaisin, K. Zenkov, S. Udartsev, Ye. Frants, P. Vladimirov

### APPENDIX B

# BeWS-15 Technical Program

**Title: Location: Date:** The 15th International Workshop on Beryllium Technology (BeWS-15) Karlsruhe Institute of Technology, Karlsruhe, Germany 14-16.09.2022

# **BeWS-15: 14.09.22**



**BeWS-15: 15.09.22**

7:00 PM Conference Dinner at Hoepfner Burghof Restaurant

### APPENDIX C

# The list of participants

#### Appendix C THE LIST OF PARTICIPANTS



### APPENDIX D

# Author Index

#### Appendix D AUTHOR INDEX


ISSN 1869-9669 ISBN 978-3-7315-1284-4