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Task 1 In-vessel Tritium Source Terms

2015-5-7 11:29| View Publisher: root| View: 3394| Comments: 0

1 Scope

This task addresses the safety issues of tritium retention and permeation properties in materials (bulk, co-deposition layers, dust, flakes and breeder materials), development of measurement techniques for tritium quantification and accountability, tritium removal from plasma-facing-component materials, and the development of associated computational tools.

2 Progress ( 2014 )

(China) Hydrogen extraction from PbLi by gas-liquid contact method and tritium release experiments on Li4SiO4 pebbles were completed in China Academy of Engineering Physics. Tritium behaviour (solubility/permeability) measurement in CLAM steel was completed in Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Science (CAS). Tritium recovery from solid breeder (Li4SiO4) and FeAl/Al2O3 coatings were performed in China Institute of Atomic Energy (CIAE). Deuterium/helium retention study in VPS-W/Cu and related experiment research in HT-7/EAST were performed at Institute of Plasma Physics Chinese Academy of Science (IPP-CAS) and Beihang University.

(EU) Long-term fuel retention behaviour was investigated via gas balance and post mortem analysis from tiles in the ITER-like-wall (solid Be in the main chamber and a combination of bulk W and W coatings in the divertor) campaign in JET.  As laboratory experiments, deuterium retention studies in plasma facing components (RAFM, bulk W, alloys and W coatings) were carried out with high-energy self-W ion bombardment.

(Japan) Deuterium retention in advanced tungsten materials was conducted under high fluence conditions at Osaka University. Tritium behavior in tritiated wastes by imaging plate and deuterium retention study with low-energy high-flux deuterium plasma exposure onto RAFM were carried out under the collaboration with University of Toyama and Japan Atomic Energy Agency.

(USA) Deuterium retention study in W was carried out with high-energy self-W ion bombardment in PISCES linear plasma device at University of California-San Diego.  Deuterium retention study in neutron-irradiated molybdenum was conducted in Safety Tritium and Applied Research facility at Idaho National Laboratory.

3 Future Plane ( 2015 )

(China)  INEST/CAS will perform corrosion study in lead-lithium eutectic and tritium permeation test in tritium permeation barrier material fabricated by electrochemical deposition.  CIAE will study performance/flexibility of Al-Pack cementation coating fabrication method to complex geometries.  IPP-CAS will investigate plasma surface interaction of ultra-fine grained/nano-crystalline tungsten.

(EU)  JET will continue investigating long-term fuel retention behavior with post mortem analysis of the tile from the ITER-like-wall campaign.  IPP-Garching will continue studying deuterium retention in plasma facing components (RAFM, bulk W, alloys and W coatings) with high-energy ion bombardment (by W self ion bombardment).

(Japan)   Osaka University will continue deuterium retention study in advanced tungsten materials under high fluence conditions with for different experimental conditions. JAEA will continue deuterium retention study onto RAFM under Low-Energy, High-Flux D Plasma Exposure with different experimental conditions for RAFM to investigate the modification of surface state and D behavior in RAFM.

(USA)  ITER grade tungsten will be irradiated with neutrons up to 1.5 dpa at High Flux Isotope Reactor (HFIR) under the new US-Japan collaborative project: PHENIX (2013-2018). The irradiation will be carried out with thermal neutron shielding to better represent the fusion neutron energy spectrum, and will be transported to the Idaho National Laboratory (INL) for the Tritium Plasma Experiment (TPE) plasma exposure test in 2015-2018. INL will investigate tritium retention from new sets of single crystal HFIR-irradiated tungsten.

4 Experts

1 C In-vessel Tritium Source Terms








Volker Philipps (Retired)


Hirofumi Nakamura


Kihak IM


A. V. Markin


Masashi Shimada


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