Search

Tritium (3H) has been increasingly researched when assessing the environmental impact of nuclear reactors and other nuclear facilities because it is widely present in nuclear systems and can easily...

Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or berylli...

Recent developments in high-magnetic-field fusion systems have created large incentives to develop flibe (Li2BeF4) salt fusion blankets that have four functions: (1) convert the high energy of fusi...

While assessing the environmental impact of nuclear power plants, researchers have focused their attention on radiocarbon (14C) owing to its high mobility in the environment and important radiological impact on human beings. The 10 MW high-temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China that adopted helium as primary coolant and graphite spheres containing tristructural-isotropic (TRISO) coated particles as fuel elements. A series of experiments on the 14C source terms in HTR-10 was conducted: (1) measurement of the specific activity and distribution of typical nuclides in the irradiated graphite spheres from the core, (2) measurement of the activity concentration of 14C in the primary coolant, and (3) measurement of the amount of 14C discharged in the effluent from the stack. All experimental data on 14C available for HTR-10 were summarized and analyzed using theoretical calculations. A sensitivity study on the total porosity, open porosity, and percentage of closed pores that became open after irradiating the matrix graphite was performed to illustrate their effects on the activity concentration of 14C in the primary coolant and activity amount of 14C in various deduction routes.

Aside from the historical Molten Salt Reactor Experiment, a few in-reactor loops, and one electron irradiation/corrosion facility, dedicated facilities to test the combined effects of molten salt corrosion and irradiation on materials do not currently exist. A major gap therefore exists in rapid, reactor-relevant materials testing capabilities which, if remedied, would greatly hasten molten salt reactor development. We present a new accelerator-based facility for rapid, simultaneous testing of molten salt-facing materials utilizing a proton beam as the radiation source. Introducing proton irradiation to a molten salt corrosion system poses specific engineering concerns in sample and corrosion cell design, operational stability, integration with the accelerator beamline, and radiation safety. This paper describes how these requirements were fulfilled with confirmatory tests and results.

Hastelloy N®, a nickel-based alloy, and 316 stainless steel are among the candidate structural materials being considered for the construction of the molten salt reactor (MSR). Most of the proposed MSR concepts use molten fluoride salts as coolant which can be quite corrosive to structural alloys. The results of studies on the corrosion behavior of the two alloys in molten Li2BeF4 (FLiBe) salt at 700°C are discussed. This salt is being considered as the primary coolant for MSR designs featuring solid fuel particles, but the reported results also provide insights into the corrosion in MSR designs where the uranium fuel is dissolved in the molten fluoride salt. Corrosion was observed to occur predominantly by de-alloying of Cr from the alloy surface and into the molten salt, with more pronounced attack occurring along the grain boundaries than in the bulk grains. The magnitude and the mechanisms of corrosion were different for corrosion tests performed in graphite and metallic capsules, a result warranting recognition given the coexistence of structural alloys and graphite in the molten salt medium in the MSR.

The MIT Nuclear Reactor Laboratory (NRL) has irradiated lithium-beryllium fluoride (flibe) salt as part of an on-going U.S. Department of Energy-funded Integrated Research Project to develop a Fluoride Salt High-Temperature Reactor (FHR). As part of this project, the NRL has carried out two irradiations of FHR materials in static flibe at 700°C in the MIT Research Reactor. These irradiations marked the start of a program evaluating the tritium production and release from the fluoride salt system at high temperature; in particular, there is interest in the evolution of tritium from the salt into solid materials and cover gasses. This paper describes the experience gained from the irradiation of flibe with respect to the detection of tritium. It covers the development of techniques for monitoring the evolution of tritium from the salt during irradiation and the factors particular to the FHR system that influence this process, including the radiolytic production and release of volatile fluorine and f...

Further understanding of tritium transport mechanisms in the combined molten fluoride salt and graphite environment is necessary for the design and licensing of a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). The three in-core fluoride salt irradiations completed at the Massachusetts Institute of Technology Reactor (MITR) are a useful parallel for studying transport phenomena expected in a FHR environment. During the irradiations, evolution of tritium from the flibe salt was monitored and compared to the calculated total generation rate. A difference of 22 ± 10% between the integrated calculated tritium generation rate and the total release was measured for the third MITR irradiation (FS-3). The fraction of tritium which was not released from the salt could be explained by tritium retention in graphite. For post irradiation examination, a thermal desorption furnace was used to heat nuclear graphite samples in order to release and measure retained tritium. The desorption analysis in this work utilized seven subsections of graphite from the second salt irradiation (FS-2); three from a disc of IG-110U and four from ARB matrix graphite. Observed desorption versus temperature as well as total tritium content in the samples after irradiation indicate that the graphites were not volumetrically saturated with tritium, but rather tritium retention was likely limited to the near-surface region. Measurements of the samples resulted in 2.90 ± 0.29 μCi/mm2 of tritium retained by IG-110U and 1.83 ± 0.31 μCi/mm2 for ARB during the 300 h FS-2 in-core irradiation. Based on the desorption measurements, the estimated total tritium retention in graphite from the FS-2 samples is consistent with the tritium release measurements from the FS-3 experiment.

In liquid–fueled molten salt reactor designs, materials will be exposed to a molten salt containing a multitude of fission products and other corrosive species. Little work has been done to understand the unique corrosion characteristics of materials in fission product-laden liquid–fuel systems. In this study, we conducted corrosion experiments up to 150 h in duration which exposed four commercial alloys (Hastelloy N, Incoloy 800H, 316L stainless steel, and Ni–201) to three molten salt compositions in order to better understand corrosion in liquid–fuel systems and inform reactor design. It was found that the presence of simulant fission product species, at predicted concentrations, in a highly corrosive FLiNaK + EuF3 molten salt does not lead to any detectable increase in corrosion at reactor–relevant conditions. No penetration of simulant fission product species into the samples was detected. The unique corrosion morphology of each of the alloys tested in this work is discussed. In particular, Ni–201 was found to be an ideal salt–facing material in molten fluoride systems, and is essentially immune to corrosion.

This study experimentally investigates the reduction efficiency of tritium permeation through 316 stainless steel tubing coated with alumina as a tritium permeation barrier (TPB) in support of the development of molten salt nuclear reactors, particularly for fluoride salt-cooled high-temperature nuclear reactors (FHRs). The TPB coatings composed ofan intermediate bond layer of NiCr, a transition layer of NiCr + alumina, and a pure alumina layer were successively added onto the outer surface of commercial 316 stainless steel tubing via plasma thermal spray. In order to generate a continuous gaseous tritium source, 35 g of purified natural-lithium FLiBe salt was irradiated by thermal neutron flux at 620 °C in the Massachusetts Institute of Technology Research Reactor (MITR). The preliminary results suggest that the TPB coatings on tube surfaces significantly reduced the tritium permeation rate at 700 °C. To get a better understanding of the TPB, the microstructure of the coated tubes was characterized with various techniques.