INTERNATIONAL NUCLEAR GRAPHITE SPECIALISTS' MEETING

INGSM-20 16th – 19th September 2019

at CROWNE PLAZA HOTEL, BRUGES, BELGIUM

PROGRAMME AND ABSTRACTS

Kindly hosted by SCK·CEN and organised under the auspices of The International Atomic Energy Agency and The British Carbon Group

Organising Committee Conference Chairman: Tony Wickham, Nuclear Technology Consultancy and The University of Manchester, UK • Elie Valcke, SCK•CEN • Luc Noynaert, SCK•CEN • Guido Vittiglio, SCK•CEN • Jean-Marie Noterdaeme, Max Planck Intstitut for Plasma Physics, Garching • Frederik Reitsma, IAEA • Gerd Haag, Nuclear Technology Consultancy, Linnich, Germany (formerly Forschungszentrum

Jülich) WELCOME TO BRUGES! Dear participant, This year is the 20th anniversary of the INGSM, the International Nuclear Graphite Specialist Meeting! It is not without proud that we, the Belgian Nuclear Research Center SCK•CEN, may warmly welcome you to this most important event where nuclear graphite specialists from all over the world meet to exchange their latest findings and to discuss on the way forward in the many fields where ‘nuclear graphite’ deserves or requires more in-depth studies. The committee has chosen the 20th INGSM to take place in the historical city of Bruges, Belgium. A UNESCO World Heritage City, Bruges will surely capture your heart, whether you are passionate about architectural and religious heritage, arts, gastronomy… or simply attracted by a modern multicultural city. As we firmly believe that combining pleasure and duty is key to success, we encourage you to take some time to discover and experience this marvelous town. On behalf of the SCK•CEN and of the organizing committee, I wish you a very fruitful meeting and an unforgettable stay in Bruges! Elie Valcke Unit Head, R&D Waste Packages, Institute for Environment, Health and Safety, SCK•CEN On behalf of The International Atomic Energy Agency and The British Carbon Group, we would like to offer our thanks to SCK·CEN for offering to host this meeting in Bruges. These meetings began in September 2000 at the suggestion of members of the technical steering committee of what was then the IAEA International Database of Nuclear Graphite Properties, with an assembly of just 20 persons at the Oak Ridge National Laboratory in Tennessee. The annual event has continued to widen its perspective to cover all aspects of the life cycle of nuclear graphite, and during this time many delightful venues have been explored in the USA, Japan, Republic of Korea, China, Germany, France, The Netherlands, The United Kingdom and Austria (HQ of the IAEA). We are most grateful to have this opportunity to visit Belgium for the first time and trust that our hosts will find the opportunity to share their experiences with others in the nuclear graphite field beneficial and a just reward for all the hard work that they have put in to make this event a success. We offer SCK·CEN our sincere thanks once again and look forward to a continued collaboration in the future. Frederik Reitsma, Tony Wickham

PRESENTATIONS & AUDIO/VIDEO FACILITIES Presentations will be 25 minutes inclusive of discussion. PLEASE KEEP WITHIN YOUR ALLOTTED TIME. Presenters are encouraged to bring their presentations on a USB memory stick and to upload them ahead of the session on to the conference computer. Whilst the latest PowerPoint software and computer facilities will be available, presenters should check their presentation prior to their session. To facilitate smooth running of the programme, presenters are also encouraged to introduce themselves to the session chairpersons prior to their session. IMPORTANT!! ALL presentation MUST be loaded by Wednesday morning so that USB drives can be prepared for presentation to delegates before the end of the conference.

INGSM-20 PROGRAMME AT A GLANCE Sunday September 15th 2019

18:00-20:00 Registration open at Crowne Plaza Hotel Monday September 16th 2019

08:00 Registration open: Coffee and Croissants 09:00 Opening of INGSM-20

Chair: A.J. Wickham 09:00 Opening Remarks - Conference Chairman 09:05 Welcome

Dr. Christophe Bruggeman, Deputy Director of the Institute for Environment, Health and Safety of the SCK•CEN, and Head of the Expert Group Waste and Disposal

09:10 SCK•CEN – 60 years of R&D for Peaceful Applications of Radioactivity Prof. Peter Baeten, Secretary-General of the SCK•CEN

09:30 Role of IAEA in Nuclear Graphite, and Status of IAEA Nuclear Graphite Knowledge Base N. Gallego

09:45 SESSION 1: Management and Characterisation of Irradiated Graphite Waste #1 Chair: L. Noynaert

09:45 Treatment of Thetis Reactor Graphite at Belgoprocess Diete Vanherck, Robin Tuerlinckx, Thomas Huys, Myriam Monsieurs and Isabelle Meirlaen

10:10 Management of the Irradiated Graphite Resulting from the Decommissioning of the WWR-S Research Reactor Cristian Dragolici, Mihaela Nicu, Laura Ionascu and Gheorghe Dogaru

10:35 Evolution of Irradiated PGA Graphite Microstructure under Molten-Salt Decontamination Conditions Tatiana Grebennikova, Iulia Ipatova, Robert N. Worth, Ben F. Spencer, Abbie N. Jones and Clint A. Sharrad

11:00 Coffee break 11:25 Radiological of Irradiated Graphite using Accelerator Mass Spectrometry (AMS)

M. Dewald, B.-A. Dittmann, A. Dewald, G. Hackenberg, S. Heinze, S. Herb, C. Müller-Gatermann, M. Schiffer, R. Spanier, A. Stolz, E. Strub, R. Margreiter and K. Eberhardt

11:50 The Release of Carbon-14 from Irradiated PGA Graphite by Thermal Treatment in Air Nassia Tzelepi, Martin Metcalfe and Glen Copeland

12:15 An Investigation into the Transport and Retention Mechanisms of Fission Products in Nuclear Graphite Alex Theodosiou, Abbie Jones and Barry Marsden

12:40 Ab-Initio Simulation of Fission Product Diffusion on Pure and Defective Graphene James McHugh, Kenny Jolley, Paul Mouratidis and Patrick Briddon

13:05 Lunch break 14:00 SESSION 2: Management and Characterisation of Irradiated Graphite Waste #2

Chair: E. Valcke 14:00 Graphite and Carbon Materials in ISOL Applications

Donald Houngbo, Marc Dierckx and Lucia Popescu 14:25 Modelling the Release of Wigner Energy during Irradiated Graphite Disposal: Are We Doing

it Right? P.C. Minshall and A.N. Jones

14:50 Monitoring Programme of the Wigner Energy at the BR1 Reactor Guido Vittiglio, Bart Van Houdt, Koen Van Aken, Elie Valcke and Frank Druyts

15:15 Irradiated Graphite Processing Approaches: Update on IAEA Project 'GRAPA' A.J. Wickham

15:40 Coffee break 16:05 SESSION 3: Irradiation Creep in Nuclear Graphite

Chair: W. Windes 16:05 High Dose Irradiation Creep of IG-110 and IG-430

Anne A. Campbell, Yutai Katoh, Richard H. Howard and Masatoshi Yamaji 16:30 Uncertainty Quantification in the Advanced Graphite Creep Experiments

Paul Humrickhouse, Jason Brookman, Vishal Patel, Jorge Navarro and Will Windes

16:55 General Discussion: Status of Knowledge of Irradiation Creep in Graphite, and Status of IAEA TECDOCS All

19:30 Welcome Reception and Tour: "De Halve Maan" "The Half Moon" Microbrewery Dress code: Informal Easy walking distance from most hotels in the old town

Tuesday September 17th 2019

08:30 SESSION 4: Graphite Fracture and Wear #1 Chair: N. Gallego

08:30 Safety and Operational Challenges of Graphite Moderator Component Cracking J. Reed

08:55 Cracking in Ageing Reactor Components – Understanding, Uncertainty and Forecasting M.R. Bradford

09:20 Notch Sensitivity Measurements of Gilsocarbon Graphite Small Specimens Matthew S. L. Jordan, Selim M. Barhli, Glen Copeland, John Dinsdale-Potter Athanasia Tzelepi, Alan G. Steer, David Nowell and T. James Marrow

09:45 Effect of AGR Fuel-Brick End-Face Features on Stress Predictions Muhammad Fahad, Emma Tan, Nick Warren, Abbie Jones and Graham Hall

10:10 In Situ Crack Growth in Unirradiated and Irradiated Graphite James Wade-Zhu, Joshua Taylor, Ram Krishna and Paul Mummery

10:35 Coffee break 11:00 Crack Initiation and Growth under Internal Loading in Scaled Fuel Bricks

William Bodel, Philippe Martinuzzi, Bruce Davies, Alan Steer and P. Mummery 11:25 Insights into the Fracture Behaviour of a Fine Nuclear Graphite Grade

Antoine Corneta, David S. Eastwood, Paul M. Mummery, Carl M. Cady and Neil K. Bourne 11:50 Electron Tomography of Nano-Cracks in Nuclear Graphite

José David Arregui-Mena, David Cullen, Robert N. Worth, Matthew Jordan, Michael Ward, Chad Parish, Cristian Contescu, Nidia Gallego, Timothy Burchell, Yutai Katoh, Philip D. Edmondson and Nassia Tzelepi

12:15 Failure from Notches in Irradiated and Virgin Graphite: Insights on Onset of Keyway Root Cracking Muhammad Treifi, Matthew Jordan, Nassia Tzelepi and Paul Mummery

12:40 Characterisation of Microcracks in Nuclear Graphites and their Relationship with the Coefficient of Thermal Expansion Xue Wang, Zhoutong He, Can Zhang, K.L. Tsang and Xingtai Zhou

13:05 Lunch break 14:00 SESSION 5: Graphite Fracture and Wear #2

Chair: J. Reed 14:00 Fatigue Behaviour of Nuclear Grade Graphite SNG742

Hui Yang, Hanchun Xu, Dai Huang and Houzheng Wua 14:25 Numerical Research and Experiment on Pneumatic Vertical Transportation Characteristics of

Pebble Fuel in HTGR Qi Sun, Wei Peng and Suyuan Yu

14:50 General Discussion on Fracture: Identification of Key Issues for All Reactor Types and in Fuel Pebbles All

15:15 Coffee break 15:40 SESSION 6: Underwriting Operational Performance and High-Fluence Irradiations

Chair: B.J. Marsden 15:40 Considerations for Characterising Reactor Graphite Prior to Operation

L. Chidwick, C. Li and M. Brown 16:05 Investigation into the Effect of Geometry, Pressure, Microstructure and Weight Loss on Gas-

Permeability Measurements for Irradiated and Unirradiated AGR Graphite Athanasia Tzelepi, John Dinsdale-Potter and Glen Copeland

16:30 Project Blackstone: The EDF Energy AGR Graphite Irradiation Programme C.Li, T.O. van Staveren, M. Joyce and B. Davies

16:55 On the Behaviour of Gilsocarbon Graphite Irradiated to Life-Time Dose Richard Gray, Bruce Davies, Tjark van Staveren, Chichi Li and Mark Joyce

17:20 RBMK - Russian Graphite Reactors - What Can We Learn From Them ? J. Reed

Wednesday September 18th 2019 08:30 SESSION 7: Structure and Properties #1

Chair: G. Vittiglio 08:30 Property Changes of Reactor Graphite due to Fast-Neutron Irradiation: Testing our

Knowledge G. Haag

08:55 Deriving Property-Averaging Equations for Polycrystalline Nuclear Graphite with and against the Grain Barry J Marsden, Antonio Fernandez-Caballero, James Wade, Abbie Jones and Graham Hall

09:20 Effects of Microstructure Change on Graphite Material Properties W. Windes, T. Yoder, A. Matthews and J. Kane

09:45 The Temperature Dependence of Defect Evolution in Irradiated Graphite Steve Johns, Joshua J. Kane, William E. Windes, Rick Ubic, K. Bustillo and Chinnathambi Karthik

10:10 Correlating Bulk Properties of Reactor-Extracted Gilsocarbon Graphite with Pore Distributions Measured by X-ray Tomography Matthew S.L. Jordan, Glen Copeland, Adam Qaisar, Tjark O. van Staveren, Joshua E. Taylor, John Dinsdale-Potter, Matt Brown and Athanasia Tzelepi

10:35 Coffee break 11:00 Ion-Irradiation Induced Damage of Graphite and its High-Temperature Recovery: Preliminary

Observations using in-situ Raman Spectroscopy Zhiyang Wang, Ondrej Muránsky, I. Karatchevtseva, Hanliang Zhu, Mihail Ionescu, Lyndon Edwards and W. Windes

11:25 Microstructural and Micro-Mechanical Characterisation of a Pair of ACCENT Graphite Samples Juan C. Luque Gutierrez, Adel El-Turke, Mark Davies, Jim Reed and Dong Liu

11:50 A Multi-Technique Image Library of Graphite Microstructures J. David Arregui-Mena, Cristian Contescu, D. V. Griffiths, Robert N. Worth, Lee Margetts, Paul M. Mummery, Anne A. Campbell, Nidia Gallego, Ercan Cakmak, Cory J. Hayes, Timothy Burchell, Yutai Katoh and Philip D. Edmondson

12:15 SESSION 8: Molten-Salt-Reactor Graphite and Composites Chair: D. Tsang

12:15 Qualifying Structural Graphite for Kairos Power’s Fluoride-Salt-Cooled, High-Temperature Reactor Gabriel Meric de Bellefon and Micah Hackett

12:40 Graphite Qualification Programme for IMSR T.O. van Staveren and M. Ivanova

13:05 Lunch break 14:00 Compatibility Studies of Graphite for Molten-Salt Reactors

Nidia C Gallego, Cristian Contescu, Tim Burchell, James Keiser, Stephen Raiman, Karol Putyera and Lou Qualls

14:25 The Chemical Interactions Among Graphite, Molten Salt, and Tritium Raluca O. Scarlat

14:50 U. S. NRC Research Activities to Address Technical Issues Related to the Application of Graphite Components for Licensing Advanced Non-Light Water Reactors R. Iyengar

15:15 Fracture Behaviour of Nuclear Graphite (MSR) Yantao Gao, Hui Tang, Zhoutong He, Derek K.L. Tsang and Xingtai Zhou

15:40 Coffee break 16:05 Irradiation Effects on Fibre, Matrix and Their Interfaces Induced by He+ Ion for C/C

Composites in TMSR Shanglei Feng and Yingguo Yang

16:30 SESSION 9: ASTM International Standards for Nuclear Graphite (Open Session). Chair: N. Tzelepi

16:30 Role of ASTM International and How Standards are Created, Approved and Maintained N. Tzelepi

16:50 Graphite Specific Standards: Recent Developments T. Van Staveren

17:00 New Standards for Molten Salt Reactors D. Tsang

17:10 General Discussion: Additional Requirements for Standards

Future ASTM Meetings during INGSM Q&A for Prospective Participants All

19:30 for 20:00

Conference Banquet: Concertgebouw Dress code: Business casual Walking (or taxi) from hotels

Thursday September 19th 2019

08:30 SESSION 10: Structure and Properties #2 Chair: G. Haag

08:30 Ab initio Study of Partial Basal Dislocations in Bilayer Graphene Pavlos Mouratidis, Malcolm Heggie, James McHugh, Kenny Jolley and Patrick Briddon

08:55 Assessment of Neutron Damage in Irradiated Graphite using Gas Adsorption Methods Cristian Contescu, James Spicer, Nidia Gallego, Anne Campbell, Jose D Arregui-Mena and Tim Burchell

09:20 Structural Effects on Synthetic Graphite Induced by 3 MeV Gold-Ion Irradiation at Elevated Temperatures Benjamin März, Zhizhen Shao and Houzheng Wu

09:45 Low-Dose Ion Irradiation-Induced Void Formation and Buckling in Highly-Oriented Pyrolytic Graphite Dong Liu, David Cherns, Joshua J. Kane and William E. Windes

10:10 Ultra-High Temperature Neutron Irradiation Effects on Graphite Microstructure Anne A. Campbell, Ercan Cakmak, Cristian I. Contescu, Nidia C. Gallego and Timothy D. Burchell

10:35 Coffee break 11:00 High-Resolution Plasma-FIB Tomography of Gilsocarbon

Dan Bradshaw, Houzheng Wu, Nassia Tzelepi, Mark Davies and Jim Reed 11:25 Irradiation Lifetime Estimation of Nuclear Graphite based on Ion-Beam Irradiation

Zhoutong He, b, Yongqi Zhu, Andy Smith, Alex Theodosiou, Abbie Jones and Barry Marsden 11:50 An Empirical Relationship for Manufactured Nuclear Graphite

Derek Tsang 12:15 In situ High-Temperature Neutron Diffraction Characterisation of Several Grades of Fine-

Grain Graphite Dong Liu, Saurabh Kabra and Houzheng Wu

12:40 The Perspective of Thermodynamics on Why Some Carbons May or May not Graphitise and the Link to Irradiation Damage in Nuclear Graphites Philippe Ouzilleau, and Marc Monthioux

13:05 Lunch break (please note 40 mins only) 13:45 SESSION 11: Oxidation / Purification

Chair: J-M Noterdaeme 13:45 High-Temperature Purification of Natural Graphite for Nuclear Applications

Ke Shen, Suyuan Yu, Bing Liu and Feiyu Kang 14:10 Application of a Random Pore Model for Thermal Oxidation of Nuclear Graphite in the

Kinetic Regime Ryan Paul

14:35 Oxidation Effects on Graphite Material Properties A. Matthews, W.D. Swank, J. Kane, and W. Windes

15:00 INGSM-21 Announcement (2020: Chicago, USA) 15:10 Close of Conference

SCK·CEN / Conference Chair For information: Trains to Brussels: 15:58 Arr Midi: 16:53 change for Airport arr: 17:31 16:08 Arr Midi: 17:07 change for Airport arr: 17:35 16:25 Arr Midi: 17:24 direct to Airport arr: 17:57

DETAILED PROGRAMME AND ABSTRACTS

INTERNATIONAL NUCLEAR GRAPHITE SPECIALISTS MEETING

INGSM-20

A Conference at The Crowne Plaza Hotel, Bruges, Belgium

16th – 19th September 2019

By kind invitation of SCK·CEN Sunday 15th September 2019 Registration open at Crowne Plaza Hotel, 18h00 – 20h00 Monday 16th September 2019 08h00 Registration open. Coffee and croissants Opening Session Chairman: Prof A.J. Wickham 09h00 Welcome: Conference Chairman 09h05 Welcome on behalf of SCK·CEN

Dr. Christophe Bruggeman, Deputy Director of the Institute for Environment, Health

and Safety of the SCK•CEN, and Head of the Expert Group Waste and Disposal 09h10 SCK•CEN – 60 years of R&D for Peaceful Applications of Radioactivity

Prof. Peter Baeten, Secretary-General of the SCK•CEN 09h30 Role of IAEA in Nuclear Graphite, and Status of IAEA Nuclear Graphite

Knowledge Base N. Gallego, Oak Ridge National Laboratory, USA

SESSION 1: Management and Characterisation of Irradiated Graphite Waste (1)

Chair: L. Noynaert 09h45 Treatment of Thetis Reactor Graphite at Belgoprocess Diete Vanhercka, Robin Tuerlinckxa, Thomas Huysa, Myriam Monsieursb, Isabelle Meirlaenb a Belgoprocess, Gravenstraat 73, 2480 Dessel, Belgium bUniversity of Ghent, Dept. of Health Physics, Proeftuinstraat 86, 9000 Gent, Belgium

The Thetis research reactor, in service from 1967 until December 2003, was located on the site of the Nuclear Sciences Institute of the Ghent University. The pool-type reactor (LEU UO2 fuel type), with a maximum operating power of 250 kW, has been used as a neutron-source for various purposes e.g. the production of radioisotopes and activation analyses. In the reactor configuration, graphite was used as a neutron reflector. This graphite-reflector consists of 8 blocks, each with a mass of about 400 kg. The blocks were removed using an overhead crane, dried and transported in a IP2 container to Belgoprocess for further treatment.

Due to its specific characteristics, even limited amounts of irradiated graphite are considered to be a special waste stream for treatment, conditioning and disposal. Based on lab analysis of both radioactivity content (e.g. 3H, 14C and 36Cl) and stored Wigner energy, a suitable treatment scenario has been developed and approved by the regulator. The aluminum cladding has been removed mechanically, the graphite blocks were cut and granulated prior to processing in the CILVA-installation. In this campaign, all specific treatments concerning irradiated graphite have been managed thoroughly.

10h10 Management of the Irradiated Graphite Resulting from the Decommissioning of the WWR-S Research Reactor Cristian Dragolici, Mihaela Nicu, Laura Ionascu and Gheorghe Dogaru IFIN-HH, Reactorului 30, 077125 Magurele, Romania

The decommissioning activities of the nuclear research reactor WWR-S from IFIN-HH, Magurele site in Romania are almost done, generating an important amount of waste. Among the different type of wastes, the total quantity of irradiated graphite resulted from the decommissioning activities is about 4,611Kg coming mainly from the thermal column. Because there is no high activity of the resulted graphite, the management took the decision that the whole amount should be temporary stored in the old Spent Nuclear Fuel Storage (SNFS) pools which perfectly could fit almost the entire quantity of the graphite waste. Only 33% from the total amount of graphite was accommodated in a special concrete container because its bigger dimensions do not allow being stored in the former SNF basins. Before all the decommissioning operation ceased, a suitable technology for i-graphite has to be set up in order to be allowed for final disposal in the National Radioactive Waste Repository (NRWR) in Baita, Bihor County. In our case the best solution for i-graphite treatment is incorporation in a stable cement matrix at the Radioactive Waste Treatment Plant (RWTP) and final disposal in our NRWR for ILLW and LLW. Calculations have been performed to estimate if the total amount of graphite could be accomplished the Waste Acceptance Criteria (WAC) of the NRWR. All these items are technically feasible and cost effective. To accomplish this goal, research activities have been performed to establish the most stable and suitable concrete matrix to embed the irradiated graphite for final storage. In order to establish the compliance with the WAC radiological should be performed for all the temporary stored i-graphite.

10h35 Evolution of Irradiated PGA Graphite Microstructure under Molten-Salt Decontamination Conditions Tatiana Grebennikova a,b*, Iuliia Ipatova c, Robert N. Worth b, Ben F. Spencer a, Abbie N. Jones b and Clint A. Sharrad a

a School of Chemical Engineering and Analytical Science, The University of Manchester, Oxford Road, Manchester M13 9PL, United Kingdom b The Nuclear Graphite Research Group (NGRG), School of Mechanical, Aerospace and Civil Engineering, The University of Manchester, Oxford Road, Manchester M13 9PL, United Kingdom c School of Computer Science and Electronic Engineering, Bangor University, Dean Street, Bangor Gwynedd, LL57 1 UT, United Kingdom

The irradiated graphite waste stream represents a significant challenge for nuclear power plant decommissioning in the UK with an estimated 96,000 tonnes of graphite waste, arising from the shut-down of the UK’s gas-cooled reactors. In this study, a pre-disposal treatment of graphite to remove contamination from corrosion and fission products and therefore downgrade the category of waste has been investigated. An electrochemical decontamination approach in a high-temperature molten salt medium was applied to irradiated Pile Grade A graphite fixed on the working electrode immersed in LiCl-KCl at 723K. The primary assessment of material behaviour and structural changes under molten salt treatment conditions have been assessed using multi-technique characterisation (Brunauer–Emmett–Teller surface area, Scanning Electron Microscopy, X-ray photoelectron spectroscopy and X-ray Powder Diffraction). The research revealed up to 80% reduction of total initial activity for 60Co isotope without significant degradation of the graphite material. The limited degradation after electrochemical treatment was mainly associated with moderate changes to the binder and impregnant phases, leaving the filler particles intact even under extreme conditions of treatment (maximum current and cycle number). The assessment of the chemical state of the sample surface analysis showed significant differences in atomic concentrations of C 1s deconvoluted peaks, suggesting the mechanism involved both diffusion of pre-adsorbed oxygen and limited chlorination of the surface. The observed results have provided a foundation for the understanding of the mechanism behind graphite decontamination using the molten salt approach and supported the future potential of irradiated graphite reduction in a waste category.

11h00 Coffee break 11h25 Radiological Characterisation of Irradiated Graphite using Accelerator Mass Spectrometry (AMS)

M. Dewalda, B.-A.Dittmanna, A.Dewaldb, G.Hackenbergb, S.Heinzeb, S. Herbb, C. Müller-Gatermannb, M. Schifferb, R. Spanierb, A. Stolzb, E. Strubc, R. Margreiterc and K. Eberhardtd aGesellschaft für Anlagen- und Reaktorsicherheit (GRS) GmbH, Germany bInstitute for Nuclear Physics, University of Cologne, Germany cDepartment for Nuclear Chemistry, University of Cologne, Germany dInstitute for Nuclear Chemistry, Mainz, Germany The radiological characterisation of irradiated graphite plays an important role during the whole process of decommissioning of many nuclear facilities worldwide. It is essential for an estimation of waste streams and waste management planning, for the planning of dismantling steps, radiation protection measures and finally for the waste classification and release of material from regulatory control. Since there is no established disposal route for irradiated graphite available yet a worldwide debate on possible treatment and characterisation methods is ongoing. In the upcoming years many nuclear facilities will reach the end of their design operation and must be decommissioned. Therefore, there is a growing need of reliable and precise characterisation

methods. In the case of hard to measure nuclides (e. g. C-14, or Cl-36) common radiation detection techniques may reach their application boundaries. This is especially relevant in the regime of low radionuclide concentrations and therefore important for material supposed for clearance. Hence, a reliable radiological characterisation is necessary to improve estimations and avoid as much nuclear carbonaceous waste as possible. A common technique to determine C-14, which is the dominant radionuclide in irradiated graphite, is Liquid Scintillation Counting. This method can be disturbed by the activity from other radionuclides, like H-3 or Cl-36, and furthermore requires a relatively elaborate process of chemical sample preparation. The project which will be presented is a collaboration of GRS and the University of Cologne. It investigates the suitability of Accelerator Mass Spectrometry (AMS) as an alternative characterisation method for irradiated graphite. The project aims to determine the C-14, Cl-36 and H-3 amount from one graphite sample using an Elemental Analyser (EA) coupled to the AMS system. A gas system is under development, which connects the EA to the AMS system and allows an automated injection and dilution of gas originating from the combustion of the graphite sample into the AMS system. The research work was performed at the 6 MV TANDETRON AMS set-up, “CologneAMS”. Several samples of graphite were irradiated with thermal as well as epithermal neutrons with different doses at the Mainz TRIGA research reactor to obtain well defined samples. We will report on the results with respect to possibilities of using AMS for radiological characterisation of hard to measure nuclides in graphite samples, which are relevant for decommissioning. The project is financed by the Federal Ministry of Education and Research.

11h50 The Release of Carbon-14 from Irradiated PGA Graphite by Thermal Treatment in Air Nassia Tzelepi, Martin Metcalfe and Glen Copeland

National Nuclear Laboratory, Sellafield, UK The United Kingdom Magnox reactors have all ceased generation and await decommissioning. This fleet of reactors contains 50,000-60,000 tonnes of irradiated graphite categorised as Intermediate Level Waste, which will require long-term storage/disposal either in a proposed Geological Disposal Facility (England and Wales) or in an on-site near surface storage facility (Scotland). Carbon-14 is one of the principal long-lived radionuclides in the graphite that determines this waste categorisation. Graphite from some Magnox reactors is known to contain carbonaceous surface deposits. Recent studies have shown that the C-14 activities (expressed as Bq per gramme of material) in these deposits are significantly higher compared to those in the underlying graphite (Figure 1). The mechanism for this phenomenon has not yet been explained but the adsorption of the C-14 pre-cursor nitrogen points to a possible production pathway. Such C-14-rich deposits could potentially lead to some alleviation on waste categorisation through their removal by thermal treatment in air. In addition, historical aqueous leaching studies have further shown that there is a small but relatively rapid initial release of C-14 from irradiated graphite, leaving a much more significant non-leachable fraction. If this mobile C-14 fraction were associated with carbonaceous deposits or even graphite surfaces that had been exposed to nitrogen, thermal treatment in air could render the material more radiologically inert even if removal provided no alleviation on waste categorisation, thereby benefiting packaging and storage options. The study presented here broadens previous thermal treatment investigations of C-14 in carbonaceous deposits to a larger number of core samples from two Magnox reactors with differing operational histories. Thermal treatment in air has also been used to investigate C-14 distributions in the underlying graphite. In addition, a small test experiment is presented to investigate any possible link between the mobile fraction of C-14 observed in leaching

studies and C-14-rich carbonaceous deposits. Finally, some interesting observations regarding the C-14 release behaviour during bulk oxidation in air under conditions of chemical control have been recorded.

12h15 An Investigation into the Transport and Retention Mechanisms of Fission Products in Nuclear Graphite Alex Theodosiou, Abbie Jones and Barry Marsden

Nuclear Graphite Research Group, School of MACE, University of Manchester, M13 9LP, UK Interest in nuclear graphite physiochemical behaviour is undergoing a resurgence due to its potential use in new Gen IV reactor designs, particularly High-Temperature reactors (HTRs) and Molten Salt Reactors (MSRs). In order to fulfil design and licensing requirements for such reactor designs, a better understanding of fission product behaviour under accident conditions and in molten salt environments is required. A new US/UK collaborative project is underway to employ advanced experimental, analytical and computational techniques in order to better elucidate the retention and transport mechanisms of a range of potential fission products (FP) in various nuclear graphite grades. Seven elements that are likely to be present in isotopic form as FPs have been highlighted as of particular interest: I, Cs, Kr, Sr, Ru, Ag and Eu. The diffusivity behaviour, bonding behaviour and transport/retention mechanisms will be investigated for Gen IV candidate graphite grades, including POCO, NBG-18, IG-110 and PCEA with work also done on historic graphite grades such as PGA and Gilsocarbon for comparative purposes. Work carried out at the University of Manchester has involved pre-characterisation of various nuclear graphite samples using Raman spectroscopy, X-ray analysis and microscopy techniques followed by ion implantation experiments using Ag, Kr and Cs with a range of ion implantation energies and temperatures. Implantation experiments were designed to embed these ions into the lattice and mimic FP introduction. post-characterisation of the graphites was carried out in order to study its effects on the graphite matrix after implantation.

12h40 Ab-Initio Simulation of Fission Product Diffusion on Pure and Defective Graphene James McHugha, Kenny Jolleya, Paul Mouratidisa and Patrick Briddonb

aDept. of Chemistry, Loughborough University bSchool of Engineering, Newcastle University Graphite has been used for neutron moderation from the beginning of the nuclear reactor era. While research activities have abated over the years, there is renewed interest in graphite motivated by its use in Very High Temperature Reactors (VHTRs) and Molten Salt Reactors (MSRs). Retention of the activated fission products is paramount during normal operating and accident conditions, and a mechanistic understanding of their bonding and diffusion is imperative for predicting release rates and designing appropriate barriers.

Figure 1: Preferential adsorption sites of adatom on graphene (left) and calculated diffusion barrier profiles using the vdW-DF-cx functional (right).

While the simulation, via Density Functional Theory, of the adsorption of a number of elements on graphene has received some attention, to the best of our knowledge little work has been conducted on their bonding and diffusive properties. We have performed high-accuracy DFT simulations of nuclear fission products on graphene using LDA, GGA and van der Waals exchange-correlation functionals within the plane-wave Quantum Espresso code. Structural relaxation at symmetrical sites is performed and preferential sites are identified for a range of nuclear fission products (Ag, Ba, Ce, Cs, Eu, I, Kr, Ru, Sr). The bonding, structural and diffusive properties have been extracted using nudged elastic band (Figure 1) projected Density of States (Figure 2) and Bader charge analysis calculations, and the effect of the different functionals is compared.

Figure 2: Sr (left) and Ru (right) spin-up projected Density of States at the energetically favourable Hollow adsorption site, calculated using the vdW-DF-cx functional. Shift in the Fermi level relative to Dirac point and partial occupation of the s-state are indicative of ionic bonding for the Sr adatom. Ru has a significant impact on the electronic structure, with evidence of strong hybridization due to the prominent peaks in the PDOS. For most elements we find very small (≤ 0.1 eV) barriers for activation on pure graphene, while experiment generally finds larger barriers and slower diffusion rates. Prior work on graphene has found that adatoms diffuse readily from non-defected regions, clustering around defects such as vacancies and holes in the lattice, and to this end we have conducted simulations of the interaction of the most important fission products with vacancy & di-vacancy structures on graphene. These simulations form a basis for the understanding of the diffusion and retention of fission products in nuclear graphite. Funding Acknowledgement: We would like to acknowledge the support of EPSRC grant number EP/R005745/1.

13h05 Lunch break SESSION 2: Management and Characterisation of Irradiated Graphite Waste (2) Chair: E. Valcke 14h00 Graphite and Carbon Materials in ISOL Applications Donald Houngbo, Marc Dierckx, Lucia Popescu SCK•CEN, Boeretang 200, 2400 Mol, Belgium

The Isotope Separation On Line (ISOL) is a method for the production and study of radio nuclei with very low or no natural abundance. To enable the exploration of the properties of these nuclei, the ISOL method relies on inducing nuclear reactions in a hot target (up to 2400°C) irradiated with a beam of protons or light ions.

Graphite and other carbon materials have over time found several applications in ISOL facilities. These range from the rather traditional applications:

• Graphite as a refractory target container insert • Graphite as a target material for producing neutron-deficient light isotopes • Exfoliated graphite foils as a high-conductivity substrate for carbide target-materials

to recent applications currently under study: • Graphite and Carbon Reinforced Carbon Composites as proton-beam-dump • Glassy carbon as ion source or target container.

The suitability of carbon materials for ISOL applications owes to a good combination of thermal, structural and nuclear properties. However, extending their usage to the new set of applications under study is faced with a range of challenges among which:

• Structural degradation of the materials during several years of proton irradiation • Isotopes-transport/retention-properties degradation under weeks of proton irradiation • Post-irradiation treatment of the materials for disposal • Safety of hot (2000 °C) carbon materials against oxygen ingress.

The different carbon materials applications in ISOL facilities will be presented with an emphasis on the applications currently under research. The research findings and challenges specific to these new applications will be communicated to this audience of specialists, along with proposed solutions, hoping to trigger discussions and/or material research in the direction of these proton irradiation applications of carbon materials.

14h25 Modeling the Release of Wigner Energy during Irradiated Graphite Disposal: Are We Doing it Right? Peter Minshalla,b and Abbie Jonesa aUniversity of Manchester, UK

bMagnox Ltd, South Gloucestershire, UK

The BEPO reactor (British Experimental Pile Zero) stated up on the Harwell site in 1948 and operated continuously until it was shut down in September 1968. The preferred option of decommissioning the reactor is to remove the graphite bricks into 6 m3 concrete boxes, pending final disposal in the Geological Disposal Final Repository (GDFR). A final decision on grouting the graphite has not yet been made. The core was annealed on three occasions to remove Wigner energy. Following the final anneal in 1968 the air inlet temperature was raised to 50°C, reducing the rate of accumulation of Wigner energy. Measurements on graphite samples removed after BEPO had shut down show significant quantities of Wigner energy in some regions of the core. The question for the decommissioning project is thus whether it is necessary to anneal the graphite before disposal: a complex and expensive process. To assess whether annealing is required, it is necessary to determine the graphite temperature at which sufficient Wigner energy is released to cause unacceptable safe heating. The rate of release is assumed to be described by a spectrum of thermally activated processes, each with an individual activation energy, 𝐸𝐸𝑖𝑖. Each group, 𝑖𝑖, releases energy at a rate given by 𝑑𝑑𝑆𝑆𝑖𝑖𝑑𝑑𝑑𝑑

= 𝜈𝜈𝑆𝑆𝑖𝑖𝑛𝑛𝑒𝑒𝑒𝑒𝑒𝑒(−𝐸𝐸𝑖𝑖 𝑘𝑘𝑘𝑘⁄ ) (1) where 𝑆𝑆𝑖𝑖 is the quantity of stored energy remaining in group 𝑖𝑖 at time 𝑡𝑡. It is apparent from the form of Equation (1) that the temperature at which Wigner energy is released at a significant rate is determined by the activation energy of a group and the initial quantity of energy, 𝑆𝑆𝑜𝑜,𝑖𝑖, associated with it. These parameters are determined from measurements on graphite samples taken from BEPO and subjected to a linear rate of

temperature increase. Several methods can be used to extract values of 𝑆𝑆𝑜𝑜,𝑖𝑖 and 𝐸𝐸𝑖𝑖 from the measurements, but all are based, in one form or another, on Equation (1). On applying the various methods to the measurements of energy release from BEPO graphite it was found that the activation energies obtained depended on the method of data analysis. As a result, one method predicted an unacceptable release rate at about 70°C, indicating that annealing was required. An alternative method, developed to assess the release of stored energy from graphite from the Windscale Piles, gave higher activation energies and indicated an onset of release at about 100°C. This would suggest that annealing was not necessary for safe disposal. Although Equation (1) is based on a plausible model for the behaviour of the defects that comprise Wigner energy, the analysis of the measurements is no more than the fitting of a numerical function to data obtained under specific and idealised conditions. There is no a priori reason why the parameters obtained from the relatively rapid temperature rise of the samples are transferrable to the slow rate of temperature change expected during decommissioning and disposal of the graphite from BEPO. This paper explores the possible reasons for the discrepancies between the various methods of analysis with the objective of providing assurance that graphite behaviour during disposal is being correctly described. Simple numerical modeling was undertaken to determine the sensitivity of the derived parameters both to uncertainties in the measurements and the form of the numerical models. An experimental programme is described which was designed to validate the numerical investigations.

14h50 Monitoring Programme of the Wigner Energy at the BR1 Reactor Guido Vittiglioa, Bart Van Houdta, Koen Van Akena, Valcke Eliea and Frank Druyts

aSCK•CEN, Boeretang 200, 2400 Mol, Belgium The BR1 reactor at SCK•CEN, Belgium, is a natural uranium-fueled, graphite-moderated, air-cooled research reactor that first became critical in 1956 and is still operational today. It contains 492 tons of graphite of pile grades A and B. The BR1 reactor first went critical on May 11th 1956 and was initially used for preliminary testing purposes before reaching full power the following year in 1957. During its early years, the BR1 reactor was run at the power of 4 MWth corresponding to a neutron flux of in de order 1012 cm-1.s-1. Since the end of the sixties the reactor run at an average power of 700 kW for 7 hours a day with peak power of 1 MWth. It is mainly used for activation, ageing test, instruments calibration and training. Being a low power reactor that operated at low temperature (graphite temperature is about 70°-90°) the monitor of the accumulation of Wigner energy is an important issue for the reactor safety. Indeed, since the start of the reactor operation, the Wigner energy of the BR1 graphite is systematically measured. In 1962, the graphite was annealed to release accumulated energy and since the new operating regime, with lower ventilation power, the temperature of the graphite has increased slightly, leading to a saturation and slow decrease of the stored Wigner energy over the years. In this paper, an overview of the monitoring program of the BR1 Wigner energy is presented.

15h15 Irradiated Graphite Processing Approaches: Update on IAEA Project 'GRAPA' A.J. Wickhama, W. Meyerb and P. O’Sullivanb

a Nuclear Technology Consultancy, Builth Wells, UK and The University of Manchester b International Atomic Energy Agency, Vienna

The International Atomic Energy Agency (IAEA) has paid a great deal of attention to the problems arising from the need to dispose of irradiated graphite competently and safely. Options for the disposal of the irradiated graphite worldwide were reviewed at a conference held in Manchester, UK (March 2007), on “Solutions for Graphite Waste: A Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors”. The collected submitted papers along with records of the discussion sessions were published as an IAEA Technical Document entitled Progress in Radioactive Graphite Waste Management (IAEA-TECDOC-1647) in 2010. The IAEA subsequently organised a coordinated research project entitled “Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal” (T21026), which was conducted from 2010 to 2014. The IAEA Technical Document Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal (IAEA-TECDOC-1790), which contains the results from that CRP, was published in 2016. In order to further support Member States in resolving irradiated graphite management issues up to the industrial implementation of processing technologies, the IAEA launched in 2016 the International Project on Irradiated Graphite Processing Approaches (GRAPA) to coordinate investigations in the 13 participating Member States, with Technical Meetings held in 2016 and 2017 in Vienna and 2018 in Vilnius Lithuania to report progress to Member States. The project considered ongoing activities in the areas of Characterisation, Removal/Retrieval, Treatment, and Packaging /Disposal, seeking to relate this work to, and learn from, ongoing decommissioning activities which include research reactors (particularly in Italy, India and Romania, production reactors in Russia, and the major efforts being undertaken on the Ignalina NPP in Lithuania and Chernobyl in Ukraine. The very different nature and scale of these operations led to the realisation that a possible international site for a modular ‘pilot-plant’ for investigating dismantling and disposal was not really practical, but the idea of using one reactor as a ‘pilot’ for future operations has been taken up in France whilst the concept of innovative treatments for the irradiated graphite continues to receive close attention especially in those Member states where no geological disposal facility currently exists. The GRAPA Project is now drawing to a close, with a final report in preparation and recommendations for future international activities intended to facilitate decommissioning activities in the future to be presented to Member States at a further Technical Meeting in October 2019.

15h40 Coffee break SESSION 3: Irradiation Creep in Nuclear Graphite Chair: W. Windes 16h05 High Dose Irradiation Creep of IG-110 and IG-430 Anne A. Campbella, Yutai Katoha, Richard H. Howarda, Masatoshi Yamajib

a Oak Ridge National Laboratory, Oak Ridge, TN 37830, United States of America b Toyo Tanso Co., Ltd., Osaka 555-0011, Japan The effects of neutron irradiation and the neutron irradiation-induced creep behaviour of IG-110 and IG-430 has been studied at the Oak Ridge National Laboratory (ORNL) in a program sponsored by Toyo Tanso Co., Ltd. This program studied the effects of neutron irradiation on the changes of the physical properties (volume, elastic properties, strength, thermal conductivity, and thermal expansion) of IG-110 and IG-430 (phase 1) and the neutron irradiation-induced creep behaviour (phase 2). The results from phase 1 were presented at the

INGSM meeting in 2017 in Baltimore, Maryland. The neutron irradiation-induced creep experiments were performed in the flux trap of the ORNL High Flux Isotope Reactor (HFIR). The irradiations were performed in three capsules that spanned the entire active core height of the HFIR. Each capsule contained both compressively stressed and unstressed reference samples, where the compressive stress was applied via a tungsten deadweight. Within each capsule were six temperature regions (three temperatures, with and without stress) with designed temperatures of 300°C, 450°C, and 600°C. The neutron fluence ranged from 10-42x1025 n/m2 [E>0.1 MeV] (~7-30 dpa). This presentation will focus on the capsule design, post-irradiation capsule disassembly, and post-irradiation measurement and analysis of the irradiation creep behaviour. The creep results will be presented and compared with results for other grades from historical irradiation creep campaigns.

16h30 Uncertainty Quantification in the Advanced Graphite Creep Experiments Paul Humrickhousea, Jason Brookmana, Vishal Patelb, Jorge Navarroc, Will Windesa

aIdaho National Laboratory, PO box 1625, Idaho Falls, ID 83415, USA bUltra Safe Nuclear Corporation cOak Ridge National Laboratory The fast fluence and radiation damage received by graphite specimens irradiated in the Advanced Graphite Creep experiments is presently estimated using spectral adjustment methods that are based on both flux wire activity measurements, and MCNP model predictions. This work describes an ongoing effort to quantify and propagate uncertainties in inputs to the spectral adjustment process, and thereby quantify the resultant error in radiation damage (dpa) estimates. The effort is multi-faceted, and we consider the impacts of both the set of flux wires selected, and the counting process. An expanded set of flux wires is identified that provides a more comprehensive data set on the fast spectrum. To address the counting process itself, a series of round-robin counts in different facilities is underway to identify differences in counting practices. To address the contribution of uncertainty in the MCNP model predictions, an uncertainty quantification (UQ) tool has been developed that statistically samples the model input parameters, runs a series of cases, and assimilates the results to provide an overall uncertainty. The impact of the MCNP UQ tool results is demonstrated by re-analysing previous AGC flux wire and irradiation data. While the expanded flux wire set obviously cannot be added to these experiments retroactively, plans for future graphite irradiations are outlined.

16h55 General Discussion: Status of Knowledge of Irradiation Creep in Graphite, and Status of IAEA TECDOCS

The objective of this discussion is to rationalize the published information and to determine what further work needs to be done to support safety cases for operational reactors and new designs

19h30 Welcome Reception and Tour: "De Halve Maan" ("The Half Moon") Microbrewery Dress code: Informal Easy walking distance from most hotels in the old town

Tuesday September 17th 2019 SESSION 4: Graphite Fracture and Wear (1) Chair: N. Gallego 08h30 Safety and Operational Challenges of Graphite Moderator Component Cracking

Jim Reed EDF Energy Nuclear Generation Ltd In 2014, the first examples of late-life cracking of AGR moderator fuel bricks were observed in Hunterston B. In 2018, the rate of cracking was observed to have increased above expectations, several components were observed to have more than one crack and “chains” of aligned cracks together with some examples of transportable fragments and debris were seen. This has led to an extensive work programme to demonstrate, firstly to ourselves and then to the public, that the behaviour of the AGR cores is understood and that it is safe to continue to operate the reactors at Hunterston B and at its sister station, Hinkley Point B, which is now showing indications of the same behaviour.

This presentation will give an overview of the cracking observations at the Hinkley Point B and Hunterston B power stations, the safety challenges that the observations present and how the safety case for operation is changing to ensure that these have been addressed in a manner that assures the continued safe operation of these plants.

08h55 Cracking in Ageing Reactor Components – Understanding, Uncertainty and Forecasting Mark Bradford EDF Energy Nuclear Generation Ltd

An overview is given of the late-life cracking of moderator fuel bricks observed in Hunterston B and Hinkley Point B and the approach that is being taken within the safety case to assure safe operation. Here, we present an overview of the underlying technical work programme. Forecasts of future cracking behaviour are built up from a combination of understanding informed by visual, dimensional and eddy current inspections, analyses on isolated components and studies of the post-cracking behaviour of bricks within a wider array. These

are used to formulate “event tree” models which are then conditioned on the inspection observations using Bayesian updating as part of probabilistic forecasting.

Uncertainty in projected behaviour arises from the limited extent of inspections that are currently practicable and alternative interpretations of the data which underlie the models. This is can be handled through a combination of conservative bias, diverse modelling analyses and sensitivity studies. This presentation will give a brief overview of the assessment route and how successive inspections and developments in computational capabilities has led to an evolution of the understanding of the progression of brick cracking and associated damage in the Hinkley Point B and Hunterston B AGRs.

09h20 Notch Sensitivity Measurements of Gilsocarbon Graphite Small Specimens

Matthew S. L. Jordana, Selim M. Barhlib, Glen Copelandc, John Dinsdale-Potterc Athanasia Tzelepic, Alan G. Steerd, David Nowelle and T. James Marrowf aNational Nuclear Laboratory, NNL Stonehouse, UK bSafran Analytics, France cNational Nuclear Laboratory, NNL Central Laboratory, UK dEDF Energy Generation Limited, UK eDepartment of Mechanical Engineering, Imperial College, UK fDepartment of Materials, University of Oxford, UK

The graphite bricks made of Gilsocarbon grade graphite used to moderate the UK’s Advanced Gas-cooled Reactors (AGRs) can develop axial cracks due to combined fast neutron irradiation and radiolytic oxidation. The formation of axial cracks at sharp corners of keyways around the brick periphery (known as “keyway root cracking” or KWRC) is of particular interest and accurate prediction of the cracking rates is necessary for confidence in estimations of their operational life. Gilsocarbon graphite is classified as medium-grained with approximately spherical filler particles of 0.6 mm nominal diameters held in a binder phase. Both the filler particles and binder are porous with highly connected networks of pores with dimensions that range from the nanometre to the millimetre scale. In the as-manufactured (“virgin”) state Gilsocarbon graphite is known to be relatively insensitive to the stress concentrating effects of surface features such as notches, which has been related to the size of the microstructural features. The effect of notches on strength been measured using component scale “feature tests”, in addition to recognised geometries for laboratory tests, e.g. single edge notched beams, circumferentially notched tensile bars and compact test specimens, where the minimum dimensions were typically above 12 mm. Tests of this size are significantly greater than the microstructure scale and produce relatively reproducible results. These tests use geometries unsuitable for reactor monitoring of graphite properties by post irradiation examination (PIE), as only samples machined from 19 mm diameter trepanned cores from the reactor’s fuel channel bores are available of measurement. The available data for the evolution of notch sensitivity of irradiated and oxidised material is, therefore, very limited. In this paper, the design constraints for a potential notch sensitivity PIE test will be considered. Any such technique must be demonstrated to measure notch sensitivity behaviour of virgin material in agreement with the large-scale results. A single edge notched four-point bend geometry has been evaluated with gross dimensions of 6 x 6 x 19 mm, and with 7 unique notch geometries. Measurements of the crack initiation behaviour of over 100 beams have been analysed using digital image correlation of multi-camera in situ imaging combined with finite element simulations. Strain softening of the uniaxial tensile modulus with increasing deflection has been confirmed. The calculated notch strengthening factors and notch sensitivity demonstrate notch insensitivity at this length scale, similar to that seen in historical studies on larger specimens. The factors calculated using a non-linear material model have also been considered, and the deformations ahead of blunt features during crack initiation were measured by digital volume correlation of X-ray tomography. The notch insensitivity may then be explained in terms of the concerted evolution of damage in the microstructure. The further effects of fast neutron irradiation and radiolytic oxidation have been investigated, through measurements of reactor-extracted material made using notched beam geometries. These tests also included a size effect study with repeat measurements of the notch sensitivity made at different scales. The results of this study will be presented and the implications for PIE measurements considered.

09h45 Effect of AGR Fuel-Brick End-Face Features on Stress Predictions Muhammad Fahada, Emma Tanb, Nick Warrenb, Abbie Jonesa, Graham Halla

a Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, UK M13 9PL b HSE Science and Research Centre, Harpur Hill, Buxton, UK SK17 9JN During a life of an advanced gas-cooled reactor, graphite components in a reactor core, specifically fuel bricks, experience higher stresses. The fuel bricks are directly exposed to fast neutron and are highly prone to radiolytic oxidation. These phenomena lead to deformation and material properties change which give rise to stresses across the fuel bricks. The magnitude of stresses may be altered further by the geometry of the fuel bricks which leads to cracking. Peripheral boundaries of the fuel bricks are designed in such a way as to accommodate a keying system (for interlocking of fuel bricks). Furthermore, the end-faces features accommodate the non-uniform deformation during the life of a reactor. These features are an important part of the fuel bricks; however they may act as stress concentrators. Fuel bricks in AGRs are designed with different end face orientations to accommodate the overall non-uniform deformation of the core. Recently the pattern of cracking has been observed to vary by end-face features orientation. In this work, three different fuel bricks end face feature designs have been modeled using finite element (FE) modelling. A sensitivity analysis is carried out to determine the parameters that are influential in determining stress and the effects of those parameters over time. The ultimate aim of this work is to use the results of the FE analysis in a probabilistic stress analysis via Monte Carlo simulation, to take account of the variability and uncertainty in the graphite material and field variables, and how these influence stresses. These stress predictions may then be compared against the predicted strength of the graphite bricks to predict the timing of keyway root cracking.

10h10 In Situ Crack Growth in Unirradiated and Irradiated Graphite

James Wade-Zhu‡, Joshua Taylor*, Ram Krishna, and Paul Mummery University of Manchester, Manchester, United Kingdom ‡Now at University of Birmingham, Birmingham, United Kingdom * Now at National Nuclear Laboratory, Sellafield, United Kingdom It is well known that fast neutron irradiation and radiolytic oxidation introduces defects and expands the open pore structures of graphite respectively. This, in turn, significantly affects the material’s mechanical behaviour. This can be confirmed by performing mechanical tests, with a measure of the work-of-fracture given by the stress/strain response. In this study, the effect of irradiation on the crack growth behaviour is studied in situ in 3D for the first time.

A novel specimen geometry (Figure 1), has been developed to allow tomographic imaging of toxic and irradiated specimens while under load and at elevated temperatures. The stress state in the specimen is such that a crack is always moving into a compressive field and a stable

configuration is attained, enabling 3D imaging. In addition, by measuring the crack length as a function of applied load, the stress intensity factor at the crack tip may be calculated. Using the aforementioned loading geometry, unirradiated and irradiated Gilsocarbon specimens, machined from installed sets at Hinkley Point B, were loaded and imaged at beamline I13 at the Diamond Light Source, UK. The load was incremented until a crack was initiated from the central hole when tomographic imaging was performed. The load was incremented again, and the process repeated until specimen failure.

Figure 1: Specimen geometry and stable crack growth

Crack growth was found to differ dramatically between the two samples, cracking in the irradiated graphite being more abrupt and requiring a significantly higher stressing condition for sequential stages of propagation (Figure 2). Analysis of the stress-strain curves and COD analysis of the unloaded irradiated specimens suggests that the abrupt crack growth in the irradiated material likely results from reduced strain release due to micro-cracking ahead of the crack tip. Expansion of the macro-pore structure resulting from radiolytic oxidation is also believed to play a key role. Such results suggest that microstructural mechanisms that govern crack growth, propagation and arrest in irradiated graphite are considerably different from those in unirradiated graphite. Therefore, the use of feature tests on large, unirradiated specimens to predict the onset of in operando keyway root cracking is unlikely to be representative and should be considered carefully.

10h35 Coffee break 11h00 Crack Initiation and Growth under Internal Loading in Scaled Fuel Bricks

William Bodel, Philippe Martinuzzi, Bruce Davies*, Alan Steer*, and Paul Mummery University of Manchester, Manchester, United Kingdom *EDF NG Ltd, Barnwood, United Kingdom The graphite core in nuclear Advanced Gas-cooled Reactors (AGRs) provides channels for fuel cooling and shut-down/control rod insertion. Therefore, damage tolerance assessments need to establish that adequate margins remain for the integrity of the fuel and proper behaviour of the safety systems. For UK existing plants, graphite brick cracking is key to making lifetime management decisions. For the next generation plants, it has an influence on expected lifetimes and on investment plans. The presence of keyway root cracks in the main population of fuel bricks at Hunterston increases the need to determine the criteria for crack initiation and the conditions under which cracks will grow under internal stress conditions, similar to those found in core. Recent work at University of Manchester, funded by Engineering and Physical Sciences Research Council and in partnership with EDF NG, has developed a technique for generating controlled internal stresses through bromine intercalation which mimic those formed by irradiation-induced dimensional change. One-tenth and one-quarter scale fuel bricks of designs for Hunterston/Hinkley Point and Hartlepool/Heysham were machined from gilsocarbon from Hinkley Point B. Controlled bromine gas concentrations was introduced to the internal bore, generating compression at the bore and tensile stresses at the keyways. As shown in figure 1, full height, axial cracks were induced to grow from keyways.

Figure 1: Full length axial cracks formed at keyways

Figure 2: Difference in crack initiation behaviour

Irradiated Unirradiated

In situ high resolution X-ray tomographic imaging was performed during bromination to both image the crack path and enable digital image correlation (DIC) to be used. DIC enabled the strain state as a function of local microstructure to be determined at crack initiation and the influence of variations in microstructure on crack growth. The influence of keyway geometry and length was investigated. In particular, the half-height keyways present in the Hartlepool/Heysham brick design affected the crack path significantly, leading to changes in crack direction and local arrest. In addition, extended bromination (beyond initiation of first crack) was used to assess the both the brick geometry on continued strain and develop criteria for secondary crack formation. These data were used to validate models developed for crack initiation and growth in related programmes.

11h25 Insights into the Fracture Behaviour of a Fine Nuclear Graphite Grade

Antoine Corneta, David S. Eastwooda, Paul M. Mummerya, Carl M. Cadyb, Neil K. Bournea aUniversity of Manchester bLos Alamos National Laboratory In the framework of the development of new nuclear graphite grades to act as neutron moderator, reflector, and structural material in the fourth generations of nuclear reactors1, we monitored though x-ray tomography the behaviour of a fine graphite grade during failure under mode I loading. The target material, the Mersen 2020 graphite, possesses a similar structure as the commonly used medium grades, i.e. a matrix consisting of filler particles embedded in a binder phase, but with a major difference in the particles size.

Figure 1 - Left: polarized micrograph of Gilsocarbon2. Right: Slice of 3D reconstructed volume of Mersen 2020 graphite. The figure 1 contains a polarized micrograph of a Gilsocarbon graphite2, a common medium grade, and a slice of a 3D volume from tomography of the fine graphite, on a similar scale. The differences in porosity and filler particles are striking. In Gilsocarbon, these filler particles play an important role in the crack development, because of their internal porosity3. The removal of such porosity in the fine graphite grade should therefore affect the fracture behaviour of the material, comparatively to Gilsocarbon. To study this fracture behaviour, we generated crack in Mersen 2020 graphite under mode I loading in a compression test as shown in figure 2. Using x-ray tomography and Digital Volume Correlation, we monitored the size of the process zone, i.e. the area ahead of the crack tip where the energy dissipation related to plasticity occurs, see figure 2.

Figure 2 - Left: Sample test geometry, sample in grey, anvils in black. Scheme of the stress field: an applied compression field C generates a tensile stress T close to the central cavity. Right: Evolution of plastic tensile strains (blue) and crack (red) with increasing load. We will discuss firstly the effect of the reduced filler size on the fracture behaviour, with a comparative study between Gilsocarbon and Mersen 2020 graphite, focusing on plasticity through the process zone extension. This process zone is known to be of great influence on fracture initiation and strength, especially for small samples4. Therefore, in a second time and in addition to temperature effects, we will present the effect of the specimen size in Mersen graphite. 1 B. Marsden et. al., Graphite as a core material for generation IV nuclear reactor, Structural Materials for Generation IV Nuclear Reactor, Woodhead Publishing, 2017, pp. 495-532 2 B. Marsden et. al., Dimensional change, irradiation creep and thermal/mechanical property changes in nuclear graphite, International Materials Reviews 61, 2016 3 D. Liu et. al., Damage tolerance of nuclear graphite at elevated temperature, Nature Communications 8, 2017 4 H. Li et. al., Fracture behaviour of nuclear graphite graphite NBG-18, Carbon 60, 2013

11h50 Electron Tomography of Nano-Cracks in Nuclear Graphite

José David Arregui-Mena a, David Cullen a, Robert N. Worth b, Matthew Jordan c, Michael Ward c, Chad Parish a, Cristian Contescu a, Nidia Gallego a, Timothy Burchell a, Yutai Katoh a, Philip D. Edmondson a, Nassia Tzelepi c a Oak Ridge National Laboratory b The University of Manchester c Nuclear National Laboratory The mechanical performance of nuclear graphite before and after irradiation is highly dependent on the porosity structure of graphite. Larger pores mainly dominate the mechanical response and fractures mechanics of graphite, whereas smaller pores control some of the dimensional change produced by irradiation. Detailed 3D models of the microstructure of nano-cracks in graphite can provide significant insight on the performance of nuclear graphite that cannot be normally obtained with standard 2D TEM characterisation techniques. Furthermore, detailed examination and quantification of nano-cracks remains difficult by other 3D characterisation techniques (x-ray computed tomography) due to the resolution limitations of conventional equipment. In order to characterise the shape of nano-cracks a technique known as electron tomography was adapted in this research to study the 3D structure of these types of pores in nuclear graphite. This technique consists of taking a series of Scanning Transmission Electron Microscopy (STEM) micrographs at different specimen tilting angles of a needle shape sample (Figure 1). These types of sample can be prepared and shaped with a focused ion beam.

Figure 1. Electron tomography specimen of neutron irradiated sample of IG-110 After the preparation of the samples they were loaded into ORNL’s Talos F200X S/TEM system with the on-axis rotation tomography holder (Fischione Instruments model 2050), that is capable of taking multiple STEM images from a rod-shaped specimen. The individual micrographs were taken at an acceleration voltage of 200 kV. Projections in HAADF-STEM mode were taken at every

5˚ until reaching 180˚ using a STEM cavity imaging technique developed at ORNL (Figure 2). On-axis rotation of the samples guarantees the maximum amount of data of the sample and homogeneous imaging conditions as well as having total access to the specimen. By using this technique, it is possible to capture nano-cracks and pores generated by the graphitization process and their evolution under neutron irradiation. As was mentioned before, these types of pores are believed to control the irradiation-induced dimensional change and coefficient of thermal expansion (CTE). The 3D characterisation of nano-pores will improve the understanding of their role during irradiation and experimental data for molecular dynamics simulations. Moreover, electron tomography will allow for the first time to quantify the pore content, orientation, connectivity and size in a 3D space. This type analysis will allow to improve the understanding of microstructural changes produce by irradiation and irradiation creep.

Figure 2. Electron tomography of IG-110 crack – a) 3D Reconstruction of solid material and crack (crack shown in purple), b) Individual HAADF images (micrographs taken with a tilt separation angle of 1º) This research was funded by the US Department of Energy, Office of Nuclear Energy, through the Advanced Reactor Technology program and by a Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) award. A portion of this research used the resources of the Low Activation Materials Development and Analysis Laboratory (LAMDA) operated by Oak Ridge National Laboratory for the US Department of Energy.

12h15 Failure from Notches in Irradiated and Virgin Graphite: Insights on the Onset of Keyway Root Cracking

Muhammad Treifi‡, Matthew Jordan*, Nassia Tzelepi*, and Paul Mummery University of Manchester, United Kingdom ‡Now at ONR, Bootle, United Kingdom *National Nuclear Laboratory, Sellafield, United Kingdom In order to predict the onset of cracking at keyway roots of irradiated graphite bricks, large scale mechanical tests have been performed on whole or sections of virgin graphite bricks. This allows the effect of the notch geometry on the failure load to be determined, which has been termed the notch strengthening factor. This is then conflated with the change in material strength due to irradiation and radiolytic oxidation to produce a value of the crack initiation stress at the corner. For this analysis to be robust, it must be assumed that the fracture behaviour of virgin graphite is the same as irradiated graphite. As the mechanical behaviour of graphite is affected by irradiation, this assumption should be tested; this study is such an attempt. The strength of small (6x6x18 mm) plain-sided and notched beams of virgin and irradiated graphite was measured in 3- or 4-point bending at the National Nuclear Laboratory. The irradiated material was sourced from installed sets at Hinkley Point B. Finite element analysis of the mechanical tests was performed to determine the notch strengthening factors (Figure 1). The notch strengthening factors of irradiated and virgin graphites were different. The virgin graphite notch strengthening factor was larger than that of the irradiated graphite, with the virgin notch strengthening factor about 1.9, and the irradiated notch strengthening factor about 1.4. This is probably due to the limited plastic deformation that is accessible to virgin graphite that allows energy from the deformation process to be absorbed by non-crack growth mechanisms. This is consistent with other studies comparing the effect of irradiation on

fracture behaviour of graphite. This casts doubt on the validity of the assumption. Indeed, it may bring the prediction of the onset of keyway root cracking forward by three to five years with concomitant consequences for the predicted lifetimes of the reactor fleet.

12h40 Characterisation of Microcracks in Nuclear Graphites and their Relationship with the Coefficient of Thermal Expansion Xue Wang, Zhoutong He, Can Zhang, K.L. Tsang and Xingtai Zhou Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, PR China

Microcracking is one of the fundamental factors for thermal expansion and irradiation-induced dimensional variability for nuclear graphite. This work statistically characterised the microcrack in filler in several grades of graphite by using transmission electron microscopy. The research includes graphites with different grain size from medium to ultra-fine. The microcracks, parallel to the basal planes in filler particles, exhibit lenticular in bright field images with 30 nm to 2 μm in length and 0.5 nm to 300 nm in width. The size distribution of the microcracks in these graphites was evaluated. Interestingly, both the length and width distributions conform to the lognormal distribution in every researched grades of graphite, and the average microcrack size can be representative of the microcrack in each graphite. The coefficient of thermal expansion (CTE) for these graphites was obtained by the dilatometer.

Figure 1: Test Geometry and FE mesh

The result shows that the CTE increases as the average microcrack size decreases. This conclusion further indicating it is the microcrack that accommodation the expansion of the graphite crystals and modify the macroscopic CTE values of graphite.

13h05 Lunch break SESSION 5: Graphite Fracture and Wear (2) Chair: J. Reed 14h00 Fatigue Behaviour of Nuclear Grade Graphite SNG742 Hui Yanga, Hanchun Xua, Dai Huanga, Houzheng Wua, b aSinosteel Advanced Materials (Zhejiang) Co., Ltd, Changxing, Zhejiang, 313100, China bDepartment of Materials, Loughborough University, Leicestershire LE11 3TU, UK

In order to meet the requirements of commercial high temperature reactor (HTR) in China, Sinosteel AMC has developed multiple grades of isostatically moulded nuclear graphite for structural components in the core of a reactor. Neutron irradiation qualification testing of these grades shows typical patterns of shrinking/expansion versus neutron fluence as a nuclear graphite should have. When a new graphite grade is chosen for a structural component of a nuclear reactor, it is necessary to have enough understanding of its physical/chemical and mechanical properties, including those under normal service conditions of a specific environment. As one of the exercises in licensing a nuclear reactor design, it is necessary to have detailed strength analysis under both static and dynamic loading conditions. Hence, in this study, we focused on the fatigue analysis of a nuclear grade graphite SNG742, a candidate chosen for HTR application in China. Detailed procedures on fatigue testing will be reported, and testing data will be reported in different format. Discussion on the testing method and data patterns will be shared in this paper.

14h25 Numerical Research and Experiment on Pneumatic Vertical Transportation Characteristics of Pebble Fuel in High Temperature Gas Cooled Reactor Qi Sun1, Wei Peng1,Suyuan Yu2*,

1.Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084, China; 2.Center for Combustion Energy, Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Department of Energy and Power Engineering, Tsinghua University, Beijing 100084, China) The collision and wear of fuel spheres in high temperature gas cooled reactor during the pneumatic transport in the lifting pipe is an important source of graphite dust. Therefore, the previous graphite pebble lifting test platform is improved to study the aerodynamic characteristics and collision behaviour of graphite pebble in the lifting pipe. The vibration spectrum of pipe wall during pneumatic transportation was measured by acceleration sensor to extract collision number. In addition, combining hydrodynamic theory with dynamic model, the aerodynamic characteristics and trajectories of graphite spheres in lifting pipes were calculated numerically. The results show that the aerodynamic force of graphite pebble increases with the increase of relative velocity and offset. The motion of graphite pebble in the pipeline presents a spiral upward trajectory by applying aerodynamic force. The collision number of graphite pebble decreases first and then increases with the increase of helium pressure due to the balance of contact force and velocity, which is in good agreement with the experimental results. It shows that the present method can provide a reference for estimating the impact wear in the real lifting pipe in HTGR.

This project was supported by the National Natural Science Foundation of China (NSFC), No. 51676112, the National Key R&D Program of China (Grant No. 2016YFC0202700), the National S&T Major Project (Grant No. ZX06901) and the National High Technology Research and Development Program of China (863) (2014AA052701). We also thank Prof. David Christopher for editing the English.

14h50 General Discussion on Fracture: Identification of Key Issues for All Reactor Types and in Fuel Pebbles

Fracture and friability are clearly key issues for the operational safety of nuclear reactors. This short discussion is intended to share experience and highlight the key areas for improving the knowledge.

15h15 Coffee break SESSION 6: Underwriting Operational Performance and High-Fluence Irradiations Chair: B.J. Marsden 15h40 Considerations for Characterising Reactor Graphite Prior to Operation L. Chidwick (Frazer-Nash), C. Li (NRG) and M. Brown (EDF Energy) 1Frazer-Nash Consultancy, UK 2 NRG Petten, The Netherlands 3EdF Energy Generation Ltd, UK

It is accepted that characterisation of graphite material properties prior to operation is essential. This paper demonstrates how some further steps can enable more complete predictions for the evolution of graphite component condition which enable better lifetime and safety assessments. A particular focus is given to the assessment of the variability and covariance of virgin graphite properties and their role in determining the likely component property variability observed through a lifetime of operation. This study presents a review of the AGR manufacturing records, complemented by more recent measurements carried out during the BLACKSTONE material test reactor programme. It highlights some of the key issues and provides examples of how the virgin graphite information can be used in predicting the behaviour of AGR moderator bricks. Moreover, it is considered that these issues are equally relevant for components in novel reactor designs.

16h05 Investigation into the Effect of Geometry, Pressure, Microstructure and Weight Loss on Gas-Permeability Measurements for Irradiated and Unirradiated AGR Graphite Athanasia Tzelepi, John Dinsdale-Potter and Glen Copeland National Nuclear Laboratory, Central laboratory, Sellafield, Seascale, Cumbria, CA20

Gas permeability is an important property for Advanced Gas-cooled Reactor (AGR) graphite weight loss predictions and forecasts, characterising how the coolant permeates through the graphite bricks in the reactor core structure. This presentation shows the results of an investigative study into the effects of geometry, pressure, microstructure and weight loss on gas permeability for virgin and irradiated AGR graphite. The method used, known as the active transient or ‘vacuum leak’ method, is an established method utilised in the historical AGR oxidation studies and in the Magnox inspection and sampling campaigns. Historical work suggests that permeability measurements have a dependency on the size of the sample. To assess the effect of sample geometries, the permeability test data from AGR

graphite samples with variable lengths and geometries were compared. The effect of applied differential pressure was investigated by measuring the permeability of two Gilsocarbon samples over differential pressures between 0.1 MPa and 1Mpa. As the graphite of the lead AGR reactors (e.g. Hinkley Point B and Hunterston B) and the newer reactors (e.g. Heysham B and Torness) was supplied by different graphite manufacturers, an investigation into the effect of virgin graphite microstructure on measured permeability coefficients was carried out; specifically, ten samples from Hunterston B and ten samples from Heysham B virgin graphite were measured. The final component of this study, the effect of radiolytic weight loss, was conducted on Hinkley Point B reactor graphite which included installed sets and recently trepanned samples in order to cover a range of weight losses between 3% and 20%. The results from this work highlight the importance of sample geometry for the active transient permeability measurements and provide insight into the differences in permeability between different AGR graphite cores, as well as showing the evolution of permeability coefficients under reactor conditions.

16h30 Project Blackstone: The EDF Energy AGR Graphite Irradiation Programme C.Lia, T.O. van Staverena, M. Joyceb, B. Daviesb

aNuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE, Petten, The Netherlands bFrazer-Nash Consultancy, Warrington, United Kingdom, WA4 6HL cEDF Energy, Graphite Branch, EDF Energy Generation Barnett Way, Barnwood, Gloucester, GL4 3RS, UK Project Blackstone at NRG Petten is a materials test reactor experiment that provides data on graphite property changes due to fast neutron irradiation and radiolytic oxidation ahead of the current AGR operating conditions. The primary objective of this project is to characterize Gilsocarbon graphite up to AGR end of life conditions. To achieve this, the HFR (High Flux Reactor) in Petten is used to accelerate graphite aging mechanisms under conditions similar to those in an AGR. In Phase I and Phase II of the Blackstone program, 4 irradiation campaigns were successfully completed and provided valuable data for the lifetime extension of the leading AGRs in the UK. Phase III is mainly aiming at providing graphite irradiation property data for two youngest AGRs, Heysham 2 and Torness. The whole irradiation is composed of two successive irradiation stages, with interim post irradiation examination (PIE) in-between. The first irradiation stage (Capsule05) has been successfully completed in the HFR end of 2018 and dismantled in the NRG Hot cell laboratories. Destructive and non-destructive PIE has been performed on the extracted graphite samples. The samples after PIE campaign will be reloaded in Capsule06 for the second irradiation stage. This presentation will provide an update on the project, highlighting key results with regard to irradiation and PIE process.

16h55 On the Behaviour of Gilsocarbon Graphite Irradiated to Life-Time Dose Richard Graya, Bruce Daviesb, Tjark van Staverenc, Chichi Lic and Mark Joycea

a Frazer-Nash Consultancy, Warrington UK b EDF Energy, Barnwood, UK c NRG, Petten, Netherlands

The continued safe operation of the UK Advanced Gas-cooled Reactor (AGR) fleet requires a robust understanding of the evolution of the graphite core. Whilst reactor core components are regularly inspected, and the resulting measurements assessed against expectations, it is also desirable to generate graphite material in a condition in advance of the operating reactor fleet. The Blackstone Material Test Reactor (MTR) programme provides such material and subsequent material properties data for all stations, and has now successfully completed five

irradiation campaigns. The programme is currently entering its final phase and is on track to deliver the last data towards the end of 2022. Due to changes in supplier and production method, the Gilsocarbon graphite used in the manufacture of the AGR moderator bricks may show some variation in its response to irradiation and oxidation between the station pairs. This presentation provides an overview of the findings from the Blackstone MTR programme thus far, and contrasts these across the AGR reactor fleet. A qualitative assessment of the evolution of the graphite microstructure from each station pair is compared with the corresponding evolution in material properties. This provides a high-level insight into the relative differences in irradiation behaviour across the fleet and forms the basis for further quantitative studies.

17h20 RBMK - Russian Graphite Reactors - What Can We Learn From Them ? J. Reed

There are currently 11 operating RBMK reactors in Russia. They are water cooled graphite moderated reactors. Their temperatures are higher than the UK AGRs. As a consequence, their irradiation dimensional change rate is accelerated compared to UK experience. The degradation due to neutron irradiation has resulted in graphite core shrinking in height, bowing channels in excess of 100mm, and a core expansion that it contacts the containment. All of these mechanisms have been addressed by an aggressive modification programme that includes taking a chainsaw to the graphite bricks. This is in contrast to the UK programme which is directed more towards tolerance by analysis at the moment. Does this experience suggest a change is required for AGRs?

Wednesday September 18th 2019 SESSION 7: Structure and Properties #1 Chair: G. Vittiglio 08h30 Property Changes of Reactor Graphite Due to Fast Neutron Irradiation - Testing Our Knowledge G. Haag Nuclear Technology Consultancy, Linnich, Germany

According to the classical models in use, graphite for reactor application is consisting of two constituents, crystalline carbon (crystallites) and voids, both with a large variety of sizes. Changes in the crystallites due to irradiation with fast neutrons are a matter of solid state physics resp. crystallography. Changes in the void structure are caused by changes of the features of the crystallites. The understanding of the effects of fast neutron irradiation on polycrystalline graphite is based upon data from irradiation tests with Highly-Oriented Pyrolytic Graphite (HOPG) representing crystalline graphite, and with graphite as manufactured. The analysis of the two sets of data should lead a model of fast neutron irradiation damage which in the best case does not leave open questions, but should at least not contain discrepancies. Such discrepancies will be addressed. In particular, the influence of the irradiation temperature and of the graphitisation temperature on the dimensional changes, the irradiation induced thermal expansivity changes, and the structure of iso-moulded graphite will be reviewed to check our knowledge.

08h55 Deriving Property-Averaging Equations for Polycrystalline Nuclear Graphite - with and against the Grain Barry J Marsden, Antonio Fernandez-Caballero, James Wade, Abbie Jones, Graham Hall Nuclear Graphite Research Group, University of Manchester, UK

To try and explain the properties of virgin and irradiated polycrystalline graphite, in the past, many authors have put forward relationships relating the properties of individual graphite crystals at the nano scale and micro scale to the bulk properties measured at the component level. These relationships take account of the crystal orientation, obtained using reflective or transmission diffraction (XRD or beam) patterns, using Reuss (constant stress) or Voigt (constant strain) assumptions. Often empirical factors are applied to try and account for the influence of lenticular porosity on the graphite structure. Of particular interest is the difference in graphite properties in both the component with-grain direction and against-grain direction. Recently the authors have revisited these relationships and found many inconsistences in the publish work, particularly in relation to the against-grain relationship. To address this shortcoming, this presentation derives the averaging relationships for Young’s modulus and coefficient of thermal expansion from first principles.

09h20 Effects of Microstructure Change on Graphite Material Properties W. Windes, T. Yoder, A. Matthews and J. Kane Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415, USA

The material property effects resulting from changes to the microstructure within nuclear graphite grades is explored. Irradiation induced material property changes can arise from irradiation damage at the atomic or crystallite length-scale but also due to changes at the mesoscale length-scale (i.e. the internal graphite microstructure). Under irradiation conditions, the anisotropic physical response of the irradiation damaged graphitic crystal structure

imposes significant physical change to the surrounding microstructure elements – particularly the inherent pore microstructure. This makes it difficult to ascertain which material property changes are affected by irradiation damage, physical alteration of the microstructure, or from both mechanisms. As an example, irradiation dimensional change is the primary life-limiting effect for determining the safe operation of a graphite component and is purely a physical effect dependent upon the total received dose. However, changes to thermal conductivity within an irradiated graphite are assumed to result from purely atomic damage at the crystallographic length-scale. By differentiating the purely physical effects from irradiation to the atomic level damage it is speculated that greater insight into irradiation damage effects may be possible. To achieve this insight, analysis of the material property changes after thermal and irradiation creep is used to indirectly distinguish between atomic and mesoscale damage effects. Material property change for both irradiated and unirradiated graphite specimens with similar plastic strains are analysed and compared. The effects of grain size, fabrication method, and creep temperature are included in the analysis to assist in the determination of atomic and mesoscale effects.

09h45 The Temperature Dependence of Defect Evolution in Irradiated Graphite Steve Johnsa, Joshua J. Kaneb, William E. Windesb, Rick Ubica, K. Bustilloc and Chinnathambi Karthika

a Micron School of Materials Science and Engineering, Boise State University, Boise, ID 83725, USA b Idaho National Laboratory, Idaho Falls, ID 83402, USA c National Center for Electron Microscopy, Molecular Foundry, Lawrence Berkeley National Laboratory, Berkeley, CA 94720, USA Graphite is commonly used as a moderator in many reactor designs and is a leading candidate core material for the envisioned Generation IV reactor concepts like the very high temperature reactor (VHTR). In reactors graphite is exposed to high-temperature neutron irradiation leading to the creation of lattice defects. These lattice defects cause changes in the mechanical properties of graphite which may adversely affect the performance of reactor components. When initially subjected to neutron irradiation, volumetric shrinkage is observed in nuclear graphite. With prolonged irradiation a turnaround point is reached where the volume then begins to increase. The turnaround point has been shown in literature to be a strong function of temperature [1], which suggests that there exists a temperature dependence to the atomistic nature of accumulated irradiation damage; however, the atomic mechanisms governing this defect evolution are not well understood. This knowledge gap is partly due to a lack of experimental data and difficulties in observing the dynamic nature of the defects. In-situ transmission electron microscopy (TEM) offers a way to monitor the atomic response of the graphite lattice during irradiation in which electrons are used as a substitute for neutrons. In this work, electron irradiation studies were conducted on IG-110 nuclear graphite and natural graphite. Graphite specimens were irradiated with a 200 kV electron beam within the temperature range of 25-800°C. Electron energy loss spectroscopy (EELS) was conducted on irradiated areas to analyze bonding character and density. Figure 1(a) shows a high-resolution transmission electron micrograph (HRTEM) of IG-110 image at 800°C near zero electron irradiation where exposed basal plane edges have curled and recombined. At temperatures above 700°C, basal planes often curl and close, which behaviour was not observed previously at lower temperatures [2]. Figure 1(b) shows the same crystallite in (a) post-irradiation conducted at 800°C where the curling of a basal plane around a prismatic dislocation results in expansion along the c axis. Figure 1(c) shows low loss EELS spectra conducted on natural graphite irradiated at equivalent doses between 25-800°C. Assuming a quasi-free electron model, the energy shift of the π+σ plasmon peak may be correlated to density, in which case a trend of decreasing density as irradiation temperature increases is observed.

Figure 1. (a) specimens of IG-110 imaged at 800°C showing the curling and closure of exposed basal plane edges pre-electron-irradiation. (b) Post electron-irradiation of the crystallite in (a) showing basal planes which curl and recombine around a prismatic dislocation. (c) Low loss EELS spectra of natural graphite irradiated at equivalent doses between 25-800°C. References: [1] B.J. Marsden et al., International Materials Reviews 61 (2016), p.155. [2] S. Johns et al., Carbon 143 (2019), p.908. The authors acknowledge funding from U. S. Department of Energy’s EPSCoR-State/National Laboratory Partnership Program (Award # DE-SC0016427). The authors acknowledge, work at the Molecular Foundry was supported by the Office of Science, Office of Basic Energy Sciences, of the U.S. Department of Energy under Contract No. DE-AC02-05CH11231. This document was prepared by Steve Johns as a result of the use of facilities of the U.S. Department of Energy (DOE), which are managed by The Regents of the University of California, acting under Contract No.DE-AC02-05CH11231. Neither The Regents of the University of California DOE, the U.S. Government, nor any person acting on their behalf: (a) make any warranty or representation, express or implied, with respect to the information contained in this document; or (b) assume any liabilities with respect to the use of, or damages resulting from the use of any information contained in the document.

10h10 Correlating Bulk Properties of Reactor-Extracted Gilsocarbon Graphite with Pore Distributions Measured by X-Ray Tomography Matthew S.L. Jordana, Glen Copelandb, Adam Qaisarb, Tjark O. van Staverenc, Joshua E. Taylorb, John Dinsdale-Potterb, Matt Brownd and Athanasia Tzelepib

aNational Nuclear Laboratory, NNL Stonehouse, UK bNational Nuclear Laboratory, NNL Central Laboratory, UK cNuclear Research and Consultancy Group, Netherlands dEDF Energy Nuclear Generation Limited, UK

The graphite grades used to moderate the UK’s CO2-cooled nuclear reactors contain pores with sizes that range from the nanometre to the millimetre scales and form highly connected networks. The interactions of the pore networks and the macroscopic phases of the material, the filler particles and binder, are expected to control certain bulk properties. As manufactured, the material contains both open (surface connected) and closed pore networks, which regulate gas flow through the bulk. For Gilsocarbon graphite, the grade used in Advanced Gas-cooled Reactors (AGRs), gas evolution during manufacture results in an open pore network throughout the binder that is not typically connected to the closed pore networks within the filler particles. Above the micron length scale, the binder pores appear globular and equiaxed, while the spherical filler particles contain elongated, often lenticular pores. When subjected to fast neutron irradiation and radiolytic oxidation the microstructure is changed dramatically. The oxidation results in progressive weight loss and the bulk properties

are significantly modified. Microstructurally the surfaces of the open pore networks are eroded, leading to increased binder porosity, while the filler particles are relatively unchanged. Eventually “break-through” into the filler particles occurs and the whole material is (more) uniformly oxidised.

10h35 Coffee break 11h00 Ion-Radiation-Induced Damage of Graphite and its High-Temperature Recovery: Preliminary Observations using in-situ Raman Spectroscopy Z. Wang 1, O. Muránsky 1, I. Karatchevtseva 1, H. Zhu 1, M. Ionescu 1, L. Edwards 1

and W.E. Windes 2 1 Australian Nuclear Science and Technology Organisation (ANSTO), Sydney, NSW, Australia 2 Idaho National Laboratory, Idaho Falls, Idaho, USA

Understanding the radiation-induced damage of graphite and its recovery (annealing) is of technological importance for development and deployment of Generation IV graphite-moderated nuclear reactors such as Very High Temperature Reactors (VHTR), and Molten Salt Reactors (MSR). Graphite radiation-induced dimensional changes are of major concern in these novel reactor systems affecting the life of the reactor in service. In the present study we investigate PCIB graphite, which was ion-irradiated using 35 MeV Carbon ions: fluence: 4.5 × 1021 ion/m2 (60,000 appm) reaching the peak damage of about 30 dpa at depth of about 37 µm. We employed the high spatial resolution of Raman spectroscopy technique (~ 1µm) enabling us the observation of ion stopping damage region – it was found that this agrees with the SRIM (the Stopping and Range of Ions in Matter) calculation. The Raman spectroscopy data shows that the radiation damage area is characterised by strong asymmetrically broadened D and G bands. The intensity ratio of Raman D and G bands (ID/IG), a measure of the defect quantity in graphite, shows a maximum of ~5 in the peak damaged location. This implies the defect accumulation and the refinement of the coherent crystalline domain size (La down to ~3.5 nm) - as supported by the high-resolution TEM measurements. Furthermore the present study shows that there is a significant recovery of ion radiation-induced damage starting at 600 °C.

SRIM

11h25 Microstructural and Micro-Mechanical Characterisation of a Pair of ACCENT Graphite Samples Juan C. Luque Gutierreza, Adel El-Turkea, Mark Daviesb, Jim Reedc and Dong Liua aUniversity of Bristol, UK bUSNC, Seattle, USA cEdF Energy, UK

The microstructure of Gilsocarbon Graphite and the changes in its properties due to irradiation have been topics of interest in the nuclear research world for many years. The purpose of this project is to characterize a pair of samples (5M71 and 5M73) irradiated in the High Flux Reactor (HFR) extracted from the ACCENT project. A set of unirradiated Gilsocarbon Graphite is also used for comparison. Both samples have a cuboid geometry before irradiation. The weights of the samples are 0.7459 g (5M71) and 0.7464 g (5M73), respectively. A constant compressive pressure (10 MPa) was applied on sample 5M71 while sample 5M73 was irradiated with no load applied. The irradiation temperature for both samples was 424°C and the end dose was 47.43x1020n/cm2 EDN. No end weight loss was experienced in the samples after irradiation. Raman Spectroscopy was used to study the samples by analyzing the characteristic bands obtained using an Ar+ laser source with a wavelength of 514.5 nm (2.41 eV). The laser power at the specimen surface was of the order 8 mW and an acquisition time of 10 s was used for each spectrum over the wavenumber range 100-4000 cm-1. The strain distribution in both samples was analyzed by evaluating the G band peak shift. For each sample, three area maps each containing 121 measurements were undertaken; the results were compared using ANOVA method to identify the difference in average values as well as statistical variance between groups. In addition, average crystalline size was estimated by measuring the intensity ratio of the D and G peak [1,2]. The formation of disorder is also discussed by studying the D (1355.6 cm-1) and 2D (2710.9 cm-1) in both radiated and unirradiated samples. In order to characterize the microstructural features, such as pores, 3D X-Ray Computed Tomographic datasets using Zeiss Xradia 520 Versa were acquired in both unirradiated and irradiated samples. During each scan, the same setup was used: 80 KeV with a LE2 filter to eliminate low energy X-rays to avoid beam hardening. Two regions, one from the surface and the other from the center of the samples, were scanned in each sample with a voxel size of 0.8 x 0.8 x 0.8 µm. Due to the limitation in the XCT pixel size, sub-micrometer pores cannot be accurately resolved. Therefore, focused Ga+ ion beam serial milling tomography with a FEI Helios NanoLab 600i Dualbeam was used to capture those smaller pores ( >20 nm). A series of 100 slices were milled across the samples surface with a cross-section area of 25 x 25 µm as shown in Fig.1a. For irradiated samples, protection procedures were adopted such as using an Al dish to collect all the debris. Post analysis of the 3D reconstruction was carried out in ImageJ and Avizo software. The data collected from both XCT and FIB-SEM Tomography for all the samples will be compared and discussed.

Peak damage

Last but not the least, in situ micro-cantilever tests were performed on the samples using a Kleindiek force measurement system installed in Helios NanoLab 600i Dual Beam workstation [1]. The mechanical properties of the filler particles and matrix were measured separately, and results will be discussed with respect to unirradiated samples.

Fig. 1. Images taken from NanoLab 600i Dualbeam on un-irradiated samples: (a) the final cut of serial sectioning tomography; (b) the surface of unirradiated sample showing filler and binder particles. References [1] F. Tuinstra and J. L. Koenig, "Raman Spectrum of Graphite," The Journal

of Chemical Physics, vol. 53, pp. 1126-1130, 1969. [2] D. Liu and P. Flewitt, "Deformation and Fracture of Carbonaceous

Materials using in situ Micro-Mechanical Testing," Carbon, vol. 114, pp. 261-274, 2017.

11h50 A Multi-Technique Image Library of Graphite Microstructures J. David Arregui-Mena a, Cristian Contescu a, D. V. Griffiths b, Robert N. Worth c, Lee Margettsc, Paul M. Mummeryc, Anne A. Campbella, Nidia Galleg a, Ercan Cakmak a, Cory J. Hayes a, Timothy Burchell a, Yutai Katoh a and Philip D. Edmondson a a Oak Ridge National Laboratory b Colorado School of Mines c The University of Manchester

Nuclear graphite is a composite material consisting of filler, binder and pores. The content of these phases and manufacturing process, determine the physical and mechanical properties of graphite. In order to understand and characterize the different phases in graphite, a variety of microscopy techniques such as Optical Microscopy (OM) (Figure 1a), Scanning Electron Microscopy (SEM) (Figure 1b), FIB-SEM tomography (Figure 1c), X-ray Computed Tomography (XCT) (Figure 1d) and Transmission Electron Microscopy (TEM) (Figure 1e) were used in order to create a comprehensive library of microstructures. These techniques are required as certain features can only be observed at a particular length scale. Similarities between the microstructures of graphite grades and different types of filler particles were identified with all these microscopy techniques. The identification of different features and porosity at different length scales will allow researchers in this field to improve the understanding of neutron irradiation effects in graphite as well as inform microstructural continuum mechanics-based models.

(a) (b)

Figure 1. Examples of micrographs in the library. a) Optical micrograph of CBG graphite, b) SEM micrograph of CBG graphite, c) Rendering of FIB-SEM tomography data of CBG graphite, d) XCT model of NBG-17, e) TEM micrograph of NBG-18

This research was funded by the US Department of Energy, Office of Nuclear Energy, through the Advanced Reactor Technology program and by a Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) award. A portion of this research used the resources of the Low Activation Materials Development and Analysis Laboratory (LAMDA) operated by Oak Ridge National Laboratory for the US Department of Energy. The work was also supported in the UK by EPSRC grant EP/N026136/1

SESSION 8: Molten-Salt-Reactor Graphite and Composites Chair: D. Tsang 12h15 Qualifying Structural Graphite for Kairos Power’s Fluoride-Salt-Cooled, High- Temperature Reactor Gabriel Meric de Bellefon and Micah Hackett Kairos Power, LLC, USA

Kairos Power’s fluoride-salt-cooled, high-temperature reactor (KP-FHR) technology uses a novel combination of existing technologies to achieve unique levels of economy, safety, flexibility, modularity and security for nuclear power production. The heart of the KP-FHR is the reactor system, which contains the reactor pebble bed core surrounded by a graphite reflector assembly. The graphite reflector assembly provides thermal inertia, neutron moderation and shielding, and coolant flow control. It operates at temperatures between 550°C and 650°C, with hotter temperature (coolant outlet) at the top, and lower temperature at the bottom and the periphery. The overall assembly consists of the top reflector, side reflector, and bottom reflector cylindrical subassemblies, that are each made of stacked blocks of graphite. The blocks are large to minimise the number of mating surfaces and potential coolant or neutron leakage paths. This talk will present the ongoing efforts aiming at qualifying structural graphite for the KP-FHR design. The qualification process is guided by the code requirements defined by the

American Society of Mechanical Engineer (ASME). Relevant degradation phenomena to be accounted for include irradiation-induced changes in dimension and properties and irradiation creep. A significant amount of irradiation test data for nuclear graphite has been generated in the past 50 years. Additional irradiation data needs related to the KP-FHR will be presented. Other key phenomena include graphite-fluoride salt interactions, which encompass molten salt infiltration and its impact on properties, molten salt wetting, and molten salt corrosion. The data requirements for graphite-fluoride salt interactions and corresponding testing strategy will be presented.

12h40 Graphite Qualification Programme for IMSR T.O. van Staverena and M. Ivanovab

aNRG, Petten, The Netherlands bTerrestrial Energy, mivanova@terrestrialenergy.com Terrestrial Energy is developing the Integral Molten Salt Reactor (IMSR®, see Figure 1) which uses molten salt nuclear fuel in combination with a graphite moderator. The design integrates the primary reactor components, including the graphite moderator, into a sealed and replaceable reactor core and can generate 195 megawatts of electricity per unit. As part of the design and licensing process a graphite qualification program is undertaken by Terrestrial Energy in collaboration with NRG. Within this qualification program, different graphite grades will be subjected to neutron irradiation at elevated temperatures and characterised on a range of material properties. The qualification program is defined such that a set of data is created that covers the life span of the graphite components in the IMSR. As part of the program, activities will be undertaken to characterise the graphite on mechanical and physical properties. In addition, microstructural investigations are performed to assess graphite resilience to salt impregnation. This presentation will give an overview of the activities undertaken for the graphite qualification program in support of the development of the IMSR reactor.

Figure 3: Schematic representation of the Terrestrial Energy IMSR reactor. 13h05 Lunch break

14h00 Compatibility Studies of Graphite for Molten-Salt Reactors (MSR) Nidia C Gallegoa, Cristian Contescua, Tim Burchella, James Keiser, Stephen Raiman, Karol Putyera and Lou Qualls

aOak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN, USA dEurofins EAG Materials Science, Liverpools, NY, USA

The new molten salt reactor (MSR) projects, currently under development, will need large amounts of nuclear graphite. The irradiation behaviour of many graphite grades, with fine and medium grains is being characterized in the U.S. through a vast irradiation campaign. In addition to irradiation and creep resistance, the graphite grades, candidates for MSRs should satisfy a few additional requirements related to their long-time operation in a more aggressive environment. Ideally, the graphite selected for MSR should be chemically and electrochemically inert, impermeable to gaseous fission products, and should exhibit little or no penetration by molten salts. This talk will outline the graphite testing and compatibility research activities at ORNL and first experimental results. These activities are currently organized in two main directions: (i) of chemical compatibility of various grades of graphite with fluoride salts, and (ii) measurement of graphite impregnation by molten salts under selected pressure and temperature. For the chemical compatibility studies, specimens of various grades of graphite were exposed in molten salts for at least 500 hours. Changes in surface compositions and microstructure of the graphite samples are being characterized by microscopy, depth profiling elemental analysis using glow discharge mass spectroscopy (GDMS), and laser-induced breakdown spectrometry (LIBS). For the impregnation studies, a pressurized cell was built for simultaneous exposure of multiple specimens. The cell can be operated up to 10 bar and 750 ºC in dry, ultra-pure argon. The extent of penetration will be calculated from the weight gain, salt density, and the open pore volume measured by helium pycnometry, per ASTM D8091-16 “Standard Guide for Impregnation of Graphite with Molten Salt”.

Research funded by the Advanced Reactors Technology program of U.S. Department of Energy, Office of Nuclear Energy.

14h25 The Chemical Interactions Among Graphite, Molten Salt, and Tritium Raluca O. Scarlat Department of Nuclear Engineering, University of California Berkeley, USA

Understanding the physics and chemistry of graphite exposed to molten salts coolants and solvents can enable the development of advanced molten salt technology, including high temperature fluoride-salt cooled reactors (FHR) and molten salt reactors (MSR). This talk will demonstrate the relationship between chemical behaviour and microstructure of nuclear graphite upon its exposure to molten 2LiF-BeF2 salt (FLiBe). It will discuss the role of these interactions on tritium uptake into graphite. This talk also presents an electrochemical study of hydrogen thermal desorption from graphite, performed in FLiBe as the electrochemical medium. Electro-analytical techniques are extremely versatile and can be used to expand our understanding of chemistry and transport at high temperature, and specifically of radioisotope behaviour in graphite. This study demonstrates a proof of principle application of electro-analytical techniques to the study of hydrogen transport in graphite, in the temperature range of 600 to 700oC.

14h50 U. S. NRC Research Activities to Address Technical Issues Related to the Application of Graphite Components for Licensing Advanced Non-Light Water Reactors

R. Iyengar US Nuclear Regulatory Commission, Rockville, MD

The NRC has established several ongoing programs to better understand the technical gaps in materials and component integrity for advanced non-light water reactors (ANLWRs). This paper will provide a summary of two recent reports on the operating experience for sodium-fast reactors (SFRs) and high-temperature gas reactors (HTGRs) and technical gaps related to molten-salt reactors (MSRs), with reference to graphitic materials and components. In addition, the paper will detail some of the ongoing research activities in to better understand and resolve technical issues related to the application of graphite components. The purpose of one of the reports is to collect and document domestic and international OpE for SFRs and HTGRs with regard to material and component integrity performance. Specific operational experience in the application of graphite components will be provided. This compendium will be used to identify consensus code gaps and summarize additional work to be performed in future companion reports. The second report summarizes the most important materials issues that must be considered for licensing MSRs, as outlined in a technical gap assessment document prepared for the nuclear regulatory commission. Materials for both fluoride and chloride applications will be discussed, with both fast and thermal neutron spectra. Knowledge gained on graphite aging, fluorination of graphite, and compatibility with molten salt, will be provided. In this presentation, an overview of NRC research activities to address regulatory challenges related to the source dependency of graphite performance, molten salt compatibility of graphite, and graphite degradation will be provided.

15h15 Fracture Behaviour of Nuclear Graphite Yantao Gao, Hui Tang, Zhoutong He, Derek K.L. Tsan and, Xingtai Zhou Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China

In the molten salt reactor (MSR), the nuclear graphite is in direct contact with molten salt. It could be infiltrated by molten salt in some cases. This report will present the effect of molten salt infiltration on the strength and fracture behaviour of nuclear graphite after molten salt infiltration. The result shows the strength of nuclear graphite will decrease with the increase of molten salt infiltration. The possible mechanism of the strength reduction was investigated by microstructure characterisation. The fracture behaviour of nuclear graphite was studied by in-situ tensile test with DIC system. The result indicated that molten-salt infiltration can affect the crack propagation of nuclear graphite.

15h40 Coffee break 16h05 Irradiation Effects of Fibre, Matrix and Their Interfaces Induced by He+ Ions for C/C Composites in TMSR Shanglei Fenga, Yingguo Yanga,b

aShanghai Institute of Applied Physics, Chinese Academy of Sciences, 2019 Jialuo Road, Shanghai 201800, China bShanghai Synchrotron Radiation Facility, Zhangjiang Lab, Shanghai Advanced Research Institute, Chinese Academy of Sciences, 239 Zhangheng Road, Shanghai 201204, China

Carbon-fibre-reinforced carbon matrix (C/C) composite has been considered as one of the promising candidates in the next generation nuclear plant (NGNP) to replace the metallic alloy to ensure the higher outlet temperature and more freedom in the reactor scram procedure

because of its superior material properties of low density, low coefficient of thermal expansion, low neutron absorption cross-section, high specific strength, high thermal conductivity, etc. To optimise the performance of C/C composites by controlling the microstructure for more reliable and safety application in thorium molten salt reactor (TMSR), we have investigated the irradiation effects of fibre, matrix, and their interfaces in C/C composite induced by He+ ions and further reveal their corresponding micromechanism. Compared with fibres, an obviously fragmented surface morphology in matrix appears and then gradually becomes widespread around the surface of C/C composite with increasing dose of irradiation. This found is attributed to the breakage of crystallites observed by synchrotron-based grazing incidence X-ray diffraction (GIXRD) and the increase of defect state density revealed by X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy, respectively. Three different microstructure evolutions in fibre, matrix and fibre-matrix interface induced by irradiation damage have been further revealed in detail by transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). It is found that the layered structure gradually loses its initial ordering and the nanostructural degradation in carbon matrix is much more serious than that of the fibre, resulting in breaks and bends in the lattice with increasing dose. Observed by nano-indentation experiment, the enhancement of the hardness and modulus of the matrix is more significant than that in fibre, which can be attributed to the more obviously pinning of basal plane dislocations in the matrix due to lattice defects induced by He+ irradiation. These discoveries are properly contributing to improve the performance of the C/C composites by regulatory microstructure composition, such as fibre and matrix.

SESSION 9: ASTM International Standards for Nuclear Graphite Chair: N. Tzelepi N.B: All delegates are invited to attend this session, intended to define the future activities required in respect of nuclear graphite standards 16h30 Role of ASTM International and How Standards are Created, Approved and Maintained N. Tzelepi National Nuclear Laboratory, Sellafield, UK 16h50 Graphite-Specific Standards: Recent Developments T. Van Staveren NRG Petten, The Netherlands 17h00 New Standards for Molten Salt Reactors D. Tsang Shanghai Institute of Applied Physics, P.R. China 17h10 General Discussion: Additional Requirements for Standards Future ASTM Meetings during INGSM Q&A for Prospective Participants Any other relevant business! 19h30 Conference Banquet: Concertgebouw, Bruges (for 20h00) Dress code: Business casual Walking (or taxi) from hotels

Thursday 19th September 2019 SESSION 10: Structure and Properties (2) Chair: G. Haag 08h30 Ab initio Study of Partial Basal Dislocations in Bilayer Graphene Pavlos Mouratidis1, Malcolm Heggie†1, James McHugh1, Kenny Jolley1 and Patrick Briddon2

1Loughborough University – Department of Chemistry 2Newcastle University – School of Engineering

Graphite has been the material of choice in construction of nuclear reactors for many years due to its low neutron absorption cross-section and high scattering cross-section. The physical properties of a graphite moderator can greatly influence the cost, safety and lifespan of a reactor. Neutron collision damage in graphite results in the formation of basal dislocations. The subsequent interaction of basal dislocations with each other and the surrounding lattice causes severe dimensional changes along the basal direction. There has been a lot of interest recently in AB and AC stacking grain boundaries in bilayer graphene. Transition from AB to AC stacking can be described by the glide of partial basal dislocations resulting in expansion of dislocation cores and buckling of the bilayer. Herein we present full ab initio and molecular dynamics calculations of basal dislocation network structures in bilayer graphene in large supercells of up to 100 nm.

Figure 1. Partial basal dislocations in bilayer graphene. Pure Edge, Pure Screw, Mixed 600, and Mixed 300 dissociation into corresponding partials Acknowledgements: EDF Energy Generation Ltd, Loughborough University, United Kingdom

08h55 Assessment of Neutron Damage in Irradiated Graphite using Gas Adsorption Methods Cristian Contescua, James Spicerb, Nidia Gallegoa, Anne Campbella, Jose D Arregui-Menaa and Tim Burchella

aOak Ridge National Laboratory, Oak Ridge, USA bJohns Hopkins University, Baltimore, USA

The structural damage caused by neutron irradiation of nuclear graphite is reflected in variation of dimensional, thermal, and mechanical properties, and changes in porosity and texture. These changes must be evaluated for every new grade of nuclear graphite in order to understand the underlying mechanism. We examine irradiated graphite from a new angle, focusing on surface properties revealed by high resolution measurements of nitrogen adsorption at 77K. Gas adsorption offers bulk-averaged structural information covering a broad range of characteristic dimensions, between about 0.3 and 300 nm, which complement local structure information at nanometer-range obtainable from SEM and TEM techniques. Adsorption is useful not only for quantification of surface area in open pores, but also for identification of imperfections caused by surface roughening, and development of porosity caused by crystallite fracturing and splitting upon irradiation. Moreover, gas adsorption is easily available and used as a standard characterisation technique for porous materials. We measured high resolution gas adsorption isotherms for several graphite grades neutron-irradiated at fluences before and after the volume turn-around dose. The results demonstrate the possibility of using sub-monolayer adsorption data to identify and evaluate the fraction of atomically ordered graphite basal planes in the total surface area calculated by the BET (Brunauer-Emmett-Teller) method. Adsorption data showed that irradiation causes gradual shrinking of uniform, atomically smooth basal plane surfaces present in the un-irradiated graphite. These changes are accompanied by multiplication of surface defects and development of mesopores (5-20 nm in width) at high irradiation doses. Additional electron microscopy and X-ray tomography results support this interpretation and provide more interesting details on local microstructural changes. The results illustrate new, and so far not reported, effects of neutron irradiation on graphite surface atomic disorder. Surface roughening quantified by gas adsorption may possibly be related with changes in the chemical reactivity of irradiated graphite, including lower oxidation resistance in HTGR reactors and higher tendency for fluorination and tritium chemisorption in FHR/MSR reactors. This research was funded by the US Department of Energy, Office of Nuclear Energy, through the Advanced Reactor Technology program and by a Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) award. A portion of this research used the resources of the Low Activation Materials Development and Analysis Laboratory (LAMDA) operated by Oak Ridge National Laboratory for the US Department of Energy.

09h20 Structural Effects on Synthetic Graphite induced by 3 MeV Gold-Ion Irradiation at Elevated Temperatures Benjamin März, Zhizhen Shao and Houzheng Wu Department of Materials, Loughborough University, UK

Nuclear iso-graphite, produced by isostatic pressure moulding of petroleum pitch and finely ground coke, are likely to play an important role in Generation IV nuclear power stations, allowing prolonged reactor operation periods at higher operation temperatures of 700-1000°C. The thermo-mechanical properties of these iso-graphite grades were optimised with regards to resistance against neutron radiation at these temperatures. In order for the material to qualify as a nuclear grade graphite, its irradiation performance is analysed after being subjected to fast neutrons in a material test reactor. This is an expensive and time-consuming procedure. Using

ion irradiation instead is often discussed as a potential alternative. It allows much higher damage rates, more control over experimental parameters and is much cheaper. Moreover, it has the advantage of easier sample handling, since no activation is induced. Despite the distinct nature of both neutron and ion radiation, conducting experiments using ion fluences that can generate similar structural damage as induced by fast neutrons in terms of displacements per atom (dpa) can still provide valuable insights of resulting structural changes. Irradiation of graphite samples with gold ions was conducted at the Ion Beam Centre of the University of Surrey in Guildford, UK, using the High Energy Implanter facility. Irradiation time and temperature of 750°C during the implantation process were selected according to the irradiation campaign conducted at Oak Ridge National Laboratory earlier [1]. We combined angular darkfield scanning transmission electron microscopy (ADF-STEM) and Raman micro-spectroscopy to explore structural changes in synthetic polygranular iso-graphite grade SNG623, manufactured by Sinosteel Advanced Materials Co. Ltd. Our analyses imply that radiation defects in the SNG623 crystal structure remained stable mostly within the graphite basal planes. The low ID1/IG Raman band amplitude ratio determined on a non-irradiated sample side reflects values typical for this polygranular graphite grade [2]. As a result of mechanical damage an increased ID1/IG ratio was determined on a mechanically processed side which was facing opposite to the ion beam. A clear change to an increased ratio was found after ion irradiation showing almost similar values on both fracture and mechanically deformed surfaces, respectively, denoting additional damage introduced by ion irradiation (Figure 1). Although applying highly-resolved ADF-STEM imaging the amount of crystal defects anticipated through the results obtained by our Raman analysis was not very obvious.

Figure 1 Ion irradiation effect on Raman band amplitudes, ID1 and IG, of SNG623 in different surface conditions.

References: [1] J.W. Geringer, A.A. Campbell, J.D. Arregui-Mena, Y. Katoh, D. Huang, H. Wu, H. Yang, H. Li, Y. Lee, C. Contescu, Sinosteel AMC Graphite Irradiation Program at ORNL, in: Int. Nucl. Graph. Spec. Meet. INGSM-2017, Baltimore, MD., 2017. [2] B. März, K. Jolley, T.J. Marrow, Z. Zhou, M. Heggie, R. Smith, H. Wu, Mesoscopic structure features in synthetic graphite, Mater. Des. 142 (2018) 268–278. doi:10.1016/j.matdes.2018.01.038.

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09h45 Low Dose Ion-Irradiation-Induced Void Formation and Buckling in Highly- Oriented Pyrolytic Graphite Dong Liua, David Chernsa, Joshua J. Kaneb, William E. Windesb

aSchool of Physics, University of Bristol, Bristol BS8 1TL, UK bIdaho National Lab, Idaho Falls ID 83402, USA

It is of particular interest to understand the damage process caused by low dose displacing irradiation in nuclear graphite. To simplify the problem, we have carried out ex situ ion implantation experiment using 3MeV single charged carbon ions at Surrey Ion Beam Implantation Centre on high quality HOPG samples (~40 µm thick and ~3mm in diameter foils) at three doses: 0.005, 0.037 and 0.15 dpa at 100°C. For each irradiation condition, a reference sample was placed outside the aperture for comparison. Ion injection direction is along the c-axis with a flux of 1.6x1012 ions/cm2/s; the conversion from the total fluence to dpa was calculated using SRIM software based on the average dose at the first 300 nm depth. As it is low mass ions with high accelerating voltage, the stopping depth is relatively high and was estimated to be 1.8 µm. As such, there is a gradient of ion damage from the surface of the sample to about 2 µm depth according to the SRIM calculations. As the a-plane tends to shrink with displacing irradiation according to conventional theory on carbon atom displacement induced dimensional change, a gradient of strain or crystal lattice mismatch along the thickness of the sample was potentially formed. Three aspects of the samples were studied to gain understanding of the irradiation damage: (1) Surface inspection using scanning electron microscope and Electron backscatter diffraction. A formation of vein type structure was found on the sample surface at dose as low as 0.037 dpa and was enhanced at higher dose of 0.15 dpa. These veins were found to be independent of the crystalline structure/orientation to distinguish it from nano-scale kink band formation observed in other work1. Yet no satisfactory explanation was found in literature for either this micro-scale buckling2,3 or nano-scale kinking1; (2) Raman spectroscopy measurement. A gas laser radiation with 532 nm wavelength was used as it is suitable for distinguish different forms of carbon. The change of the in-plane size of all three samples was derived based on its correlation with the disorder induced D-band intensity and the excitation laser energy. In addition to the E2g vibration mode, the evolution of double-resonance Raman process induced D, D’ and G’ peaks involving phonons within the first Brillouin zone was discussed as a function of the irradiation dose; (3) Focused ion beam cross-sectioning and transmission electron microscopic (TEM) analysis of the cross-sectional nanostructures of edge-on basal planes. The gradient of ion damage was readily visible in the TEM analysis. Large voids due to buckling were formed close to the surface of the sample and are correlated to the surface vein structure. Contrary to basal slip theory aided by in-plane dislocation gliding4, it is proposed here that as the gradient of ion damage enhanced the mismatch between adjacent graphite layers, basal plane slip ({1000}<11-20>) was inhibited and as such one of the secondary slip systems with a high Schmid factor was activated which is the primary contributing factor to micro-buckling hence void formation.; The evolution of dislocations with ion implantation that is responsible for such buckling is under further investigated using high resolution TEM. It worth noting that the gradient in ion damage was not the sole factor for the buckling, as we also observed nano-scale buckling in thin foils (<50 nm) without such stress gradient. The details of the experimental evidences for both the micro-scale and nano-scale buckling will be discussed at the conference. The mechanisms concluded from this work will be compared with conventional dislocation pinning-unpinning model4, buckling, ruck & tuck model5 and ripplocation theory6. The understanding gained from the void formation in HOPG will be extended to the fundamental aspect of dimensional change in polycrystalline nuclear grade graphite.

References: 1. Hinks, J. A. et al. Dynamic microstructural evolution of graphite under displacing irradiation. Carbon N. Y. 68, 273–284 (2014). 2. Muto, S. et al. TEM analyses of surface ridge network in an ion-irradiated graphite thin film. J. Nucl. Mater. 271–272, 285–289 (1999). 3. Niwase, K. & Tanabe, T. Modification of graphite structure by D+ and He+ bombardment — II. J. Nucl. Mater. 179–181, 218–222 (1991). 4. Kelly, B. T. & Foreman, A. J. E. The theory of irradiation creep in reactor graphite—The dislocation pinning-unpinning model. Carbon N. Y. 12, 151–158 (1974). 5. Heggie, M. I., Suarez-Martinez, I., Davidson, C. & Haffenden, G. Buckle, ruck and tuck: A proposed new model for the response of graphite to neutron irradiation. J. Nucl. Mater. 413, 150–155 (2011). 6. Barsoum, M. W. et al. Ripplocations: A universal deformation mechanism in layered solids. Phys. Rev. Mater. 3, 013602 (2019).

10h10 Ultra-High Temperature Neutron Irradiation Effects on Graphite Microstructure Anne A. Campbell, Ercan Cakmak, Cristian I. Contescu, Nidia C. Gallego and Timothy D. Burchell Oak Ridge National Laboratory, Oak Ridge, TN 37830, United States of America

The High Temperature Vessel (HTV) irradiation was performed at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR). This irradiation was performed to determine the turn-around fluences at high (900°C and 1200°C) and ultra-high (1500°C) temperatures, to final exposures of 1.8-3.3 dpa, to assist in the determination of the maximum irradiation fluences for the Advanced Graphite Creep (AGC) capsules AGC-5 and AGC-6 that were originally planned for irradiation at 1100°C. Six graphite grades were irradiated in the HTV capsule: NBG-17, NBG-18, H-451, PCEA, IG-110, and 2114. The microstructure features of these grades are being investigated to provide an understanding of how irradiation modifies the graphite structure at these extreme temperatures, and whether these structural changes are similar or different that the changes observed at lower irradiation temperatures. This presentation will discuss the results of the irradiation effects on the graphite crystallinity (measured via X-ray diffraction), while a second presentation will discuss the development and change of pores smaller than 300 nm (via N2 adsorption). These results will be compared to the results from the same grades that were irradiated at a lower temperature in the AGC-1 creep capsule that were reported at a previous INGSM. The comparison will include discussions about the similarities and differences of the irradiation temperature effects on the microstructure and the postulation of mechanisms that could be the source of any differences in response. This research was funded by the Advanced Reactor Technologies program of U.S. Department of Energy. A portion of this research used resources at the High Flux Isotope Reactor, a DOE Office of Science User Facility operated by the Oak Ridge National Laboratory. Oak Ridge National Laboratory is managed by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 for the U.S. Department of Energy.

10h35 Coffee break 11h00 High-Resolution Plasma-FIB Tomography of Gilsocarbon Dan Bradshaw a, Houzheng Wu a, Nassia Tzelepi b, Mark Davies c, Jim Reed d a Loughborough University b National Nuclear Laboratory c MARAD Limited d EDF Energy Generation Ltd

A reactor’s irreplaceable graphite bricks are currently viewed as being the life-limiting component and, as such, there is a drive to further knowledge of the material. To do so, it is essential to quantify the key microstructural features of Gilsocarbon across all length-scales.

Whilst many different techniques have been utilised to map the porosity and microcracks present, in nuclear graphites, these practices are limited by either their spatial resolution (e.g. X-μCT) or the volume of material which can be analysed. Recent developments in Plasma-Focused Ion Beam (FIB) and SEM technologies have significantly increased the capabilities of the Serial Sectioning Tomography technique (SST). Greater volumes of material can be excavated, with the exposed microstructures being imaged at high resolutions, bridging the length-scale gap between more established characterisation methods. Full knowledge of the porosity/microcracks will give insight into the true mechanisms governing material damage/deformation experienced during reactor operation. It can also improve experimentally informed models of porosity/microcracks & property changes at multiple length scales. A Helios G4 UXe Dualbeam Plasma-FIB (FEI, Oregon) was utilised to extract a 75 μm x 40 μm x 40 μm section from a fracture face and conducted an automated high-resolution SST programme. An in-plane imaging resolution approaching 10nm was achieved (micrograph HFW of 75 μm), as seen in fig.1, with individual slice thicknesses of approximately 20 nm; this allowed us to resolve thousands of previously unaccounted pores/microcracks. AVIZO® (Thermo Fisher Scientific Inc.) was used for both image processing and three-dimensional reconstructions, shown in fig.2; pre-segmentation processing is required to remove characteristic artefacts of P-FIB milling carbonaceous materials. The highly automated system allowed all slices to be simultaneously processed, guaranteeing parity across a dataset, and ensured that all pores/microcracks present were successfully identified and segmented with minimal user input. Post-reconstruction, the pore-microcrack 3D model can either be viewed in unison and quantified as a percentage porosity of the microstructural volume or as separate entities; for each pore/microcrack, a wide range of parameters were measured including volume, orientation, tortuosity, and elongation. There is capacity to skeletonise the model and quantify open-pore connectivity but at this scale we are predominantly concerned with closed porosity.

This presentation will share the learnings gained from extensive hours of conducting large-scale PFIB projects on Gilsocarbon, from sample preparation through to post-processing analysis. The section lift-out FIB SST technique will be briefly outlined: the advantages, limitations, and position within a correlative tomography framework, before in-depth discussions surrounding some of the key hurdles faced. Most of the difficulties arise due to the need for the acquisition and processing to be as automated as possible, something that will only be exaggerated with increasing dataset size. Finally, this presentation will discuss how we see the technique contributing to the wider nuclear graphite field and the future work we hope to undertake.

11h25 Irradiation Lifetime Estimation of Nuclear Graphite based on Ion-Beam Irradiation Zhoutong Hea, b, Yongqi Zhub, Andy Smithc, Alex Theodosioua, Abbie Jonesa and Barry Marsdena

aNuclear Graphite Research Group, School of MACE, University of Manchester, M13 9LP, UK bShanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, P. R. China cDalton Cumbrian Facility, University of Manchester, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA, UK

The development of Generation IV reactors requires development and understanding of new nuclear graphite grades and as such, their dimensional change under fast neutron irradiation is critical to their application. This has previously been obtained by using materials test reactors (MTR). However, MTR irradiation is difficult due to access, expensive and time. Ion beam irradiation shares the comparable damage mechanisms with fast neutron irradiation and is frequently used as a surrogate for fast neutron irradiation without resulting in radioactivity. Limited by ion projection range in material, it is difficult to measure the dimensional change induced by the ion beam irradiation. Here we present a new method which was designed to estimate the dimensional change behaviour of graphite based on measuring the bending of thin graphite foil caused by ion beam irradiation. With this method, the difference in the dimensional change behaviour of graphite grades can be compared. We used nickel ion beam to irradiate IG110 and an ultra-fine grained graphite grade to demonstrate the effectiveness of the method. The bending of the graphite foils with the irradiation dose shows that the method is very promising. Also, a model was developed to transduce the curvature of the graphite foil induced by ion beam irradiation to the dimensional change. In order to compare the difference in the changes of graphite under ion beam irradiation and fast neutron irradiation, the microstructure of the Ni ion beam irradiated graphite are characterized by micro-Raman mapping, nanoindenter and Nano-CT.

11h50 An Empirical Relationship for Manufactured Nuclear Graphite Derek Tsang Shanghai Institute of Applied Physics, Shanghai,P.R. China

Nuclear graphite is a key material of nuclear reactors where it is used as a moderator, reflector and for other structural components. In general mechanical and physical properties of nuclear graphite are strongly affected by methods of production, pore shape and porosity. Figure 1 (overleaf) shows pore-shape ratio against porosity for different grade of graphite. The data suggests that there is a general relationship between manufactured nuclear graphite. In this talk an empirical relationship for graphite will be presented. The general relationship will be used to predict uniform oxidation effect on Young’s modulus and CTE.

12h15 In situ High-Temperature Neutron-Diffraction Characterisation of Several Grades of Fine-Grain Graphite Dong Liua, Saurabh Kabrab and Houzheng Wuc

aSchool of Physics, University of Bristol, UK bISIS Neutron and Muon Source, Rutherford Appleton Laboratory, UK cDepartment of Materials, Loughborough University, UK

In the last five years, a range of nuclear graphite grades were studied using elastic neutron scattering including medium grained Gilsocarbon graphite1,2 and fine grain grades represented by T220, SNG623, SNG742 and SNG7223. These experiments used thermal neutrons from a pulsed spallation source (ISIS, RAL) and the intensity (diffraction maximum) were collected as a function of the time of flight where the d-spacings of the basal, prismatic and pyramidal planes in graphite can be derived spontaneously. We have conducted experiments with in situ deformation, including tension, compression and bending, at room temperature up to 850 oC to simulate the possible service temperature and loading conditions as a core component in a high temperature reactor (HTR). By synchronizing neutron diffraction, in situ observation using digital cameras has been set up to capture the total dimensional change on the sample surface at all temperatures. All samples were loaded up to fracture. We have acquired data for each graphite grade with data including elastic deformation in graphite crystal quantified from diffraction measurements, the corresponding total deformation measured through digital image correlation, and thermal expansion coefficient (CTE) at different temperatures for both graphite crystal and bulk graphite. In this work, we will summarise all the published and unpublished data so far and give a side-to-side comparison of the measured lattice spacings, CTE values (crystal and macro-scale), elastic strains to failure, and total inelastic deformations among the studied grades. An ‘inelastic energy’ term will be introduced by extracting the elastic deformation from the total stress-strain curve, and the ‘damage tolerance’ of two representative grades of fine grain graphite with respect to medium grained Gilsocarbon will be discussed. This is one of our first steps to gauge the deformation behaviour of fine grain graphite based on our understanding in Gilsocarbon material using neutron diffraction under in situ loading conditions.

References: 1. Marrow. T. J. and D. Liu et al, In situ measurement of the strains within a mechanically loaded polygranular graphite. Carbon, 96, 285–302 (2016). 2. D. Liu et al, In situ measurement of elastic and inelastic strains during high temperature deformation of a polygranular graphite, in preparation for publication. 3. D. Liu et al, In situ characterisation of the elastic-inelastic deformation in fine grain graphite using neutron diffraction at elevated temperatures, in preparation for publication.

12h40 The Perspective of Thermodynamics on Why some Carbons may or may not Graphitise, and the Link to Irradiation Damage in Nuclear Graphites Philippe Ouzilleau1 and Marc Monthioux2 1Université de Sherbrooke, Canada 2Université de Toulouse, France

Graphitisation is the heat treatment process which produces graphitic carbons from graphitisable precursors. However, not all precursors are equally graphitisable. Nuclear graphites are graphitisable carbons obtained through the thermal annealing of local graphenic defects, a process which increases order. On the other hand, irradiation of nuclear graphites generates graphenic defects which introduce disorder in the material. For this reason, some suggested that irradiation damage could be interpreted as some sort of de-graphitisation phenomena. The present work will discuss how thermodynamics can predict the graphitisability spectrum and thus provide some knowledge on the nature of generated defects in irradiated nuclear graphite.

13h05 Lunch break Please note: 40 minutes only!! SESSION 11: Oxidation / Purification Chair: J-M Noterdaeme 13h45 High-Temperature Purification of Natural Graphite for Nuclear Applications Ke Shena, Suyuan Yub, Bing Liub and Feiyu Kangb a Hunan University, China bTsinghua University, China

The natural graphite, including flake graphite and microcrystalline graphite, plays an important role in the production of nuclear graphite in high-temperature gas-cooled reactor. The flake graphite is a key raw material of the graphite pebble, which is made from 64% natural flake graphite, 16% artificial graphite and 20% phenolic resin. On the other hand, the microcrystalline graphite based isotropic graphite has a high graphitisation degree, a low coefficient of thermal expansion, so as to show a considerable potential application as nuclear graphite. High temperature treatment up to 2500 ºC is an effective approach to purify both the natural flake graphite and microcrystalline graphite. Ten different natural graphite ores from China are purified by high temperature treatment in the presence of halogen. By optimisation of the temperature curve, nuclear grade flake graphite and microcrystalline graphite samples were obtained. By heat treated at 2800 ºC with halogen, both the ash content and Equivalent Boron Content readily meet the requirement of HTR-PM graphite powder. And the most favorable flake graphite ores were selected for future HTR-PM pebbles. The main impurity in natural graphite include Si, Al, Fe, etc. The impurity evolution behaviour in the range of 200-2500 ºC was examined to determine how these impurities are removed at evaluated temperatures. For safety concerns, measures should be taken to avoid the massive evolution of impurities in a narrow temperature range during an industry scale operation. High temperature treatment causes the evolution of graphite microstructure as well. The change of d002, Lc and La were determined by XRD, the change of pore structure was detected by N2 adsorption. This study shows a comprehensive description of the purity and microstructure of natural graphite during high temperature heat treatment.

14h10 Application of a Random Pore Model for Thermal Oxidation of Nuclear Graphite in the Kinetic Regime Ryan Paul GrafTech International Holdings Inc

A persistent concern for high-temperature applications of synthetic graphite material is loss of graphite due to thermal oxidation gasification reactions with O2, H2O or CO2. In certain conditions, oxidation enlarges the inherent pores in graphite, which rapidly degrades microstructure and properties. Oxidation weight loss behaviour can vary significantly among grades of nuclear graphite, for example, which have different microstructures due to differences in raw materials and processing. However, there are currently no validated oxidation models in use that predict oxidation rates from initial microstructure and inherent material reactivity. This work presents a new model for isothermal kinetically-controlled oxidation of high-purity synthetic graphite. The model assumes graphite is a binary structure composed of solid graphite and pores, in which the pores are randomly-placed equisised spheres. Spheres growth at a temperature-dependent rate that can be constant throughout the sample or a function of depth. Samples are assumed to be cubes or cylinders, accordingly, the model predicts the effect of changing sample geometry or size on apparent weight loss behaviour.

14h35 Oxidation Effects on Graphite Material Properties A. Matthews, W.D. Swank, J. Kane, and W. Windes Idaho National Laboratory, USA

Changes to several graphite material properties after exposure to air oxidation at different temperatures and similar mass loss levels are measured for several nuclear graphite grades. Low temperature (550°C) air oxidation demonstrates a lower failure stress for fine grain grades than when oxidized at higher temperatures (750°C). Low temperature (485°C) oxidation demonstrates significant lowering of thermal properties, such as CTE and diffusivity, in contrast to previous studies showing only limited effects from air oxidation. It is surmised that the extremely low reactivity rate present for these low temperatures induces oxidation pores that are small enough to interrupt these thermal properties unlike the larger pore structures expected during higher temperature. Finally, the elastic modulus of graphite for low temperature (485°C) oxidation shows decreasing trends.

15h00 INGSM-21 Announcement (2020: Chicago, USA) 15h10 Close of Conference SCK·CEN / Conference Chair See you in Chicago!

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