Tungsten is foreseen as a plasma-facing material in a fusion reactor, which provides an intense neutron radiation environment. The Coordinated Research Project (CRP) on "Plasma-Wall Interaction with Irradiated Tungsten and Tungsten Alloys in Fusion Devices" (2013-2018) will enhance the knowledge base on effects of neutron and surrogate irradiation upon tungsten and tungsten alloy microstructure and thereby upon surface erosion and upon trapping and transport of tritium in tungsten-based plasma-facing materials.
First RCM: Tuesday-Thursday 26-28 November 2013 at IAEA Headquarters in Vienna. Agenda and presentations.
Mini meeting at PSI-21, Kanazawa, Japan: Thursday May 29, 2014 Report repared by H.-K. Chung and Y. Hatano
Second RCM: Tuesday-Friday 08-11 September 2015 hosted by the Nuclear Engineering Department of Seoul National University, Republic of Korea. Meeting report and Presentations.
Mini meeting at PSI-22, The Pontifical Urbaniana University, Rome, Italy: Thursday June 2, 2016 Report repared by H.T. Lee and W. Jacob
Third RCM: June 27-30, 2017. Agenda and presentations.
Fusion energy production relies on the reaction of hydrogen isotopes deuterium (D) and tritium (T) forming helium and releasing 14-MeV neutrons. In the magnetic confinement approach to fusion, D-T plasma at a temperature of around 15 keV (about 150 million K) is trapped in a toroidal magnetic field inside a vacuum vessel. The confinement is not perfect and the plasma interacts with the vacuum vessel wall. This may cause erosion of the surface and may also cause tritium to become trapped in the wall material, making it unavailable for fusion. The plasma-wall interaction issues of erosion and tritium retention are perhaps the most serious impediment to the realization of fusion energy production.
The choice of wall material for a fusion demonstration experiment involves a difficult compromise between the demands of low erosion (favouring heavy elements such as Mo or W), low radiation loss as a plasma impurity (favouring light elements such as Be and C), high melting point and high thermal conductivity (favouring C and W), low nuclear activation (generally favouring lighter elements) and low propensity to absorb tritium (favouring tungsten most of all, steel and beryllium less, and all but ruling out carbon-based materials). It should be noted that the tritium retention issue is one of safety and nuclear licensing as well as cost; in ITER and in any successor facility there will be strict limits on the amount of tritium that may be trapped in the wall material. For ITER in its nuclear phase these considerations have led to the choice of tungsten as the wall material for the regions of highest heat load and beryllium for the larger main wall surface area that is exposed to less severe heat load. For a fusion demonstration experiment, DEMO, or a reactor the same considerations point to tungsten for the regions of highest heat load and some kind of reduced-activation steel for the main wall, but there are major unresolved issues still.
The most serious complication for these considerations, and the one that leads to the topic of the present CRP, is the very high neutron fluence in DEMO or in a reactor. For DEMO estimates of neutron fluence are around 30 dpa (displacements per atom) per year of full power operation and for a reactor the expected fluence is even higher. For tungsten this level of neutron irradiation will cause many dislocations and will change the composition to an alloy of (primarily) tungsten, rhenium and osmium. Pure crystalline tungsten has an extremely low affinity for tritium, but this good property will be impacted the wrong way by the neutron fluence in DEMO or in a fusion reactor and erosion properties will also be affected. These issues are critically important for fusion energy development beyond ITER, but investigations into properties of irradiated fusion materials are hampered by the unavailability of an adequate neutron source and by the great difficulty of relevant first principles computations. Therefore the material properties, the resistance to sputtering and ablation, and the behaviour of trapped tritium in tungsten-based materials after neutron irradiation are still poorly known.
The CRP on Plasma-Wall Interaction with Irradiated Tungsten and Tungsten Alloys in Fusion Devices was established to improve the knowledge base and databases on properties of irradiated tungsten. The most important topic is to understand how tritium retention, tritium migration and ways to extract trapped tritium are affected by radiation damage. The relevant level of neutron damage is so high that it cannot be simulated in present experiments, and a proposed intense fusion neutron source (the International Fusion Material Irradiation Facility, IFMIF) is still only a design effort with no commitment to build it. Experiments therefore employ only low levels of neutron irradiation or (more importantly) they employ surrogate irradiation by charged particle beams (H, D, T, He, Fe, W). The difference between neutron irradiation and surrogate irradiation in their effect on plasma-material interaction properties needs to be much better understood. This leads to the need to characterize the effects upon microstructure due to different kinds of irradiation and to characterize how changes in microstructure influence the tritium retention and tritium transport properties.
Tungsten microstructure occupies central place in the CRP; we want to understand the effect of neutron and surrogate (charged particle) irradiation upon microstructure and through that the effect upon plasma-material interaction properties, especially tritium retention and transport. Microstructure is also where experiment meets computation for quantitative comparisons that provide physical understanding. Experiment and computation are both used to study the effect of surrogate irradiation upon microstructure and that of microstructure upon plasma-material interaction. Experiments are crucial to validate the very difficult computations, but without IFMIF only computations are available to extrapolate to a fusion neutron environment.
The overall objective of the CRP is to support fusion plasma and fusion materials modelling and planning and design efforts towards DEMO and a Fusion Power Plant through the enhancement of the knowledge base on properties of tungsten as a plasma-facing material in a fusion nuclear environment, and thereby to contribute to the development of fusion energy generation. More specific objectives are:
In order to achieve these objectives the CRP will rely on experiment and computation. It is understood that experiments cannot use the relevant neutron fluence, hence the interest in surrogate irradiation, and it is understood that many experiments will work with hydrogen or deuterium instead of tritium. Computations will help to bridge the gaps.
General information about participation in IAEA CRPs may be found at the Coordinated Research Activities web site. We expect 19 or 20 participants in CRP F43021 on Irradiated Tungsten. The names, institutes and a brief indication of their projects follows.
Guang-Hong Lu et al., Computational Materials Physics Group, School of Physics and Nuclear Engineering, Beihang University, Beijing. Modeling and simulation of behaviours of hydrogen, helium and their synergy in irradiated tungsten. Project devoted to fundamental computations of microstructure: defects, defect clusters, H and He bubbles and their interaction and migration. The results of the fundamental computation are to be incorporated into models for larger space- and time-scales.
Changsong Liu, Xuebang Wu et al., ISSP-CAS, Hefei. Multiscale study of the interactions of hydrogen with irradiation-induced defects at entire temporal evolution stages. Focused on applied modelling (MD and KMC) of microstructure damage and fundamental (DFT) modelling of hydrogen in damaged tungsten.
Guangnan Luo, Chonghong Zhang et al., IPP-CAS, Hefei. Effect of high energy ion irradiation damages on deuterium permeation and retention in tungsten. The team will produce damaged tungsten and study it on the EAST tokamak and by laboratory exposure.
Marie-France Barthe, Charlotte Becquart et al., CNRS. Modeling and experimental validation of hydrogen behaviour in tungsten for tokamak. Multi-institute project for experiments and applied modelling of irradiation damage and hydrogen behaviour in damaged tungsten. Models range from MD to continuum in a multiscale approach.
Christian Grisolia, Bernard Rousseau et al., CEA. Quantification of tritium implantation in tungsten-based fusion materials. Multi-institute project focused on experiments, applied modelling and some fundamental modelling of hydrogen (tritium) in damaged tungsten.
Matej Mayer, Thomas Schwarz-Selinger et al., IPP Garching. Deuterium Retention in Tungsten Damaged by High-energetic Tungsten Ions. The group will produce damage using MeV tungsten (and other) beams. The damage will be characterized and effects on hydrogen retention will be studied using nuclear reaction analysis, thermal desorption spectroscopy, transmission electron microscopy and other. Experimental work is supported by modelling.
Bernard Unterberg, Jochen Linke et al., Forschungszentrum Jülich. Radiation induced degradation of tungsten grades under thermal and plasma exposure and development of advanced tungsten materials. Study of combined effects of irradiation and heat load on surface and microstructure of tungsten; studies of hydrogen interaction with damaged tungsten; studies of different tungsten grades. A hot cell may be available near the end of the project period.
Shishir Deshpande, P. M. Raole et al., IPR, Gandhinagar. Radiation damage and H/D retention studies on ion-irradiated tungsten and its alloys - experiments and modeling. Studies of irradiation induced damage and its effects on the retention of hydrogen isotopes in tungsten and in the W-La2O3 alloy.
Akira Hasegawa et al., Tohoku University. Effects of neutron irradiation on damage structure evolution of tungsten in fusion device. Studies that involve real neutrons and associated neutron damage; the team has a long history in the field. The focus of this project is on the detailed characterization of microstructure following irradiation.
Yuji Hatano, Yasuhisa Oya et al., Toyama University. Hydrogen isotope retention in neutron-irradiated tungsten and tungsten alloys. Study of tritium migration in damaged tungsten. The proposed research fits in a USA-Japan collaboration on tritium in fusion materials.
Mizuki Sakamoto, Hideo Watanabe et al., Plasma Research Center, University of Tsukuba. Microstructure and hydrogen isotope retention in tungsten with radiation induced defects. Multi-institute project on the creation and characterization of microstructure damage by MeV surrogate irradiation and the behaviour of hydrogen in damaged tungsten.
Takuji Oda, Hyung-Jin Shim et al., Seoul National University. Evaluation of tritium inventory in irradiated tungsten by atomic-scale modeling. Fundamental (DFT) and applied (MD, KMC) modelling of tritium behaviour in radiation-damaged tungsten.
Yury Gasparyan, Alexander Pisarev et al., National Research Nuclear University "MEPhI", Moscow. Deuterium retention in radiation damaged tungsten materials. Work connected to experiments at IPP Garching with modelling support at MEPhI.
Boris Khripunov, Alexander Ryazanov et al., Kurchatov Institute, Moscow. High flux plasma effect on tungsten damaged by high energy ions. Kurchatov has both the MeV surrogate irradiation facilities and the plasma exposure (LENTA) facilities.
Sabina Markelj et al., Microanalytical Center, Department for Low and Medium Energy Physics, Josef Stefan Institute, Ljubljana. Hydrogen retention in self-damaged and He irradiated tungsten and alloys for PFC. Use in-situ nuclear reaction analysis (NRA) during atomic beam exposure to determine dynamic deuterium retention in undamaged and pre-damaged tungsten. The influence on D retention of damage due to high energy He beam will be studied also by simultaneous irradiation by 4He and deuterium atomic beam.
Segei Dudarev et al., CCFE, Abingdon. Gas desorption from irradiated tungsten under radiation-induced swelling conditions. Fundamental study of the evolution of damage (vacancy migration, interaction of vacancies, helium bubbles, swelling) and effects of material temperature upon evolution of damage. The fundamental studies are used to develop applied large-scale models for microstructure evolution and gas transport.
Brian D. Wirth et al., University of Tennessee, Knoxville (UTK). Plasma surface interactions in tungsten involving He. Fundamental calculations to calibrate potentials for large-scale MD studies and the execution of such studies.
Jean-Paul Allain, David Ruzic et al., University of Illinois. Plasma-material interactions research on hydrogen permeation in irradiated tungsten. A variety of surrogate irradiations will be used and the materials will be exposed to plasma at Illinois.
Masashi Shimada, Brad Merrill et al., Idaho National Laboratory. Tritium behavior in neutron-irradiated tungsten using TPE divertor plasma simulator and TMAP mass transport code. INL is the lead USA institute in the USA-Japan collaboration on tritium in irradiated fusion materials. This project combines studies on the production and characterization of damage and studies on hydrogen behaviour in damaged tungsten.
CRP F44003 on Primary Radiation Damage Cross Sections (2013-2017; first RCM in Nov 2013) engages participants from the nuclear data and materials research communities to determine ways to characterize radiation damage from different sources including neutrons and charged particles with different energy spectra. The aim is to improve upon the so-called NRT-dpa (Norget, Torrens and Robinson displacements per atom) standard and describe radiation damage using as few parameters as possible, choosing parameters that provide the best possible correlation to relevant material properties.
CRP F13013 on Investigations on Materials under High Repetition and Intense Fusion Pulses (2011-2015) is devoted to experimental studies using accelerators of the response of candidate plasma facing materials (different tungsten grades and coatings, beryllium, CFC, SiC) of fusion devices to extreme energy and particle loads. The main issue addressed in CRP F13013 is macroscopic damage to wall material (melting, blistering, ablation).
CRP F43020 on Data for erosion and tritium retention in beryllium plasma-facing materials (2012-2017) is concerned with the other material that is of most interest for ITER, where beryllium will be the main wall material and tungsten will be used in the divertor.