The proposed experiments allow for further investigation of nuclear graphite samples tested during the 2016 molten flibe salt in-core irradiation at the Massachusetts Institute of Technology Reactor (MITR). Measurements in this proposal can also expand the current understanding of hydrogen trapping in neutron irradiated graphite and produce useful results relevant for advanced reactors. The graphite samples included will first be characterized to examine the microstructure changes caused by high-temperature molten salt and neutron irradiation. Irradiated and as-received graphite samples will then be monitored for hydrogen solubility at elevated temperatures using constant pressure gas charging. The motivation of this work is to directly measure hydrogen retention in graphites after neutron irradiation and understand the relevant mechanisms responsible for the changes in solubility.



A main graphite grade of interest is IG-110 because of its availability in previous literature as well as its use in previous MITR irradiations. High-purity IG-110U discs 8mm diameter and 2mm thick have been irradiated at MITR up to a fast fluence of 3.4E20 n/cm2 (E>0.1 MeV), which is expected to create a noticeable change in graphite microstructure and hydrogen solubility based on studies in literature. The graphite interplanar distance (d002) will be measured using X-ray diffraction, which can be correlated to the overall degree of graphitization. Defects in the graphite structure created by neutron damage are believed to decrease graphitization after irradiation as well as the increase the overall trapping of hydrogen. Samples will also be examined with a scanning electron microscope (SEM) to observe the grain structure and pore size distribution with and without neutron damage. Analysis with SEM can help determine whether other microstructural metrics can explain differences in graphite hydrogen retention. In addition to the IG-110U samples, one other grade of nuclear graphite will be selected for the baseline characterization.



Direct measurements of hydrogen retention in irradiated and as-received graphites will be made with a gas sorption analyzer equipped with a sample heater. The hydrogen retention measurements will be carried out over a representative range of expected fluoride-salt-cooled high-temperature reactor (FHR) temperatures and tritium partial pressures, namely 500-700ᵒC and 0.1-10kPa H2. The relationship between total retention and hydrogen pressure can be used to determine whether hydrogen is retained in a molecular form or whether dissociative retention occurs. An intermediate vacuuming and recharging step can be added to the procedure to differentiate between high-energy, persistent retention, and easily removable hydrogen. The weak/strong retention procedure is useful to examine the extent to which neutron irradiation increases concentration of low energy and high energy trapping sites. If successful, these measurements aim to be the most representative hydrogen solubility data for the FHR because of the relevant graphite grades, irradiation temperatures, hydrogen charging temperatures and hydrogen partial pressures.

The proposed research tasks are aligned with the Department of Energy Office of Nuclear Energy mission, primarily through materials research relevant to advanced reactor development. Fluoride-salt-cooled high-temperature reactors (FHRs) are featured in the Advanced Reactor Concepts program as a promising nuclear technology. Research into retention of tritium in graphite at elevated temperatures is also relevant for high temperature gas-cooled reactor designs. Since 2011, three NEUP three-year Integrated Research Projects (IRPs) have been awarded which together focus on establishing a path forward to the commercialization of FHRs.



For the FHR, mitigating the environmental release of tritium is a key technical issue. Several options have been proposed to partition tritium from the coolant for safe storage, such as a permeation window, gas sparger, carbon absorber bed, and permeation barrier coatings. Currently, it is difficult to fully assess which option is the optimal tritium management solution due to the lack of data relevant to FHR conditions in open literature. The IG-110U graphite samples proposed for investigation can serve as a useful data source because they were irradiated in molten flibe salt at a temperature and neutron flux that is representative for the FHR design. Additionally, the hydrogen measurements of these graphite samples will use a characteristic temperature and partial pressure for the gas retention studies. Therefore, the nuclear graphite hydrogen retention and microstructure characterization studies in this proposal are of high value to FHR designers.