Mechanisms of Retention and Transport of Fission Products in Virgin and Irradiated Nuclear Graphite
Lead Research Organisation:
University of Manchester
Department Name: Mechanical Aerospace and Civil Eng
Abstract
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Organisations
Publications
Bakenne A
(2021)
Microstructural Deformation and Modulus Changes of Nuclear Graphite due to Hydrostatic Pressure Loading
in Journal of Nuclear Materials
Bodey A
(2017)
Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging
in Journal of Physics: Conference Series
Grebennikova T
(2021)
Electrochemical decontamination of irradiated nuclear graphite from corrosion and fission products using molten salt
in Energy & Environmental Science
Krishna R
(2017)
An understanding of lattice strain, defects and disorder in nuclear graphite
in Carbon
Liu F
(2021)
The pores evolution in 3D view of nuclear graphite under ion beam irradiation
in Journal of Nuclear Materials
Marsden B
(2017)
Structural Materials for Generation IV Nuclear Reactors
Theodosiou A
(2018)
The complete oxidation of nuclear graphite waste via thermal treatment: An alternative to geological disposal
in Journal of Nuclear Materials
Theodosiou A
(2020)
An XPS/UPS study of the surface/near-surface bonding in nuclear grade graphites: A comparison of monatomic and cluster depth-profiling techniques
in Applied Surface Science
Theodosiou A
(2017)
Thermal oxidation of nuclear graphite: A large scale waste treatment option
in PLOS ONE
| Description | In order to fulfil the design and licensing requirements for advanced reactors, the U.S. Department of Energy (DOE) is interested in targeted research to quantify and understand the mechanisms of fission product (FP) transport and retention in nuclear graphite. This knowledge is important for many reasons, both during normal operation and accident conditions. From a safety viewpoint, it is essential that the movement of FPs can be predicted and understood in order to improve reactor safety and to minimise any effect on reactor operation. To achieve this improved understanding, a US/UK collaboration is underway, that will use an integrated experimental-computational approach to measure the diffusivities of seven highlighted fission products in a range of nuclear grade graphites. The pertinent transport/retention mechanisms will then be elucidated using a range of experimental analysis techniques and multiscale simulations. The graphite grades of interest are those that are considered potential candidates for use in the next-generation very/high temperature reactors (V/HTRs), fluoride salt-cooled high temperature reactors (FHRs) and molten salt reactors (MSRs). These grades include POCO (ZXF-5Q/AXF-5Q), NBG-18, PCEA, and IG-110 [1,2]. Other grades, used in the UK reactor fleet, PGA and Gilsocarbon will also be studied for comparison, along with HOPG as a reference. The FPs that are of interest are Iodine (I), Cesium (Cs), Krypton (Kr), Strontium (Sr), Ruthenium (Ru), Silver (Ag), and Europium (Eu). The purpose of the ion irradiation is to successfully implant FP ions into the graphite to allow for further experimental studies on the behaviour of the ions within the graphite matrix. Of particular interest are XPS and Raman analysis [3,4] which will be carried out before and after the irradiation experiments. Other effects of the ion bombardment, such as irradiation damage, will also be studied in detail. Subsequent work would then look to build on the initial phase of ion irradiation experiments and would look at other parameters such as increasing the fluence, changing the ion, and increasing the irradiation temperature. The aim would be to build up a detailed picture of FP behaviour within a range of different nuclear grade graphites, of various microstructure, in order to be able to predict and understand the physio-chemical mechanisms involved and hence ultimately improve the safety of the Gen IV reactor designs. |
| Exploitation Route | through journal publication and conference outputs |
| Sectors | Education,Energy |
| Description | Used in publications and in dissemination activities eg conferences and workshops |
| First Year Of Impact | 2019 |
| Sector | Energy |
| Impact Types | Societal,Economic |
| Title | GRAPHITE DECONTAMINATION |
| Description | The present invention relates to methods of decontaminating irradiated nuclear graphite. The method comprises immersing the irradiated nuclear graphite in a molten salt electrolyte, and subjecting the irradiated nuclear graphite to an electrochemical treatment. |
| IP Reference | WO2021250413 |
| Protection | Patent granted |
| Year Protection Granted | 2021 |
| Licensed | No |
| Impact | - |