Catalyzed Oxidation of IG-110 Nuclear Graphite by Simulated Fission Products Ag and Pd Nanoparticles

To evaluate the stability of nuclear materials in high temperature gas reactors under air ingress conditions, catalytic oxidation of IG-110 graphite by two simulated fission products, metallic Pd and Ag, was studied in oxidative atmosphere and at temperatures up to 1000 °C using an integrated furnace, mass spectroscopy and infrared spectroscopy system. Transmission electron microscopy and X-ray diffraction studies show that Pd and Ag nanoparticles were successfully introduced onto powdery IG-110 graphite through an impregnation and subsequent heat-treatment process. The combined mass spectroscopy and infrared spectroscopy methods allow simultaneous analysis of two gaseous products, CO and CO2, and separate measurements of activation energy for their formation reactions. It was found that the introduction of Pd or Ag to IG-110 graphite substantially catalyzed the oxidation of graphite, characteristic of decreased onset temperatures for the oxidation of graphite. Moreover, the catalytic effects by Pd and Ag are considerably different based on measured concentration ratios of CO2 to CO as a function of oxidation temperatures. Ag makes the graphite oxidation commence at approximately 400 °C with CO2 being the dominant product. In contrast, Pd significantly increases the concentration ratio of CO2 to CO at temperatures higher than approximately 690 °C, although it decreases the onset temperature for the oxidation reaction to around 525 °C. To understand the catalytic difference, the mechanism of the graphite oxidation is discussed based on the changes of surface oxygen species on Ag and Pd

The Geozoic Supereon

Geological time units are the lingua franca of earth sciences: they are a terminological convenience, a vernacular of any geological conversation, and a prerequisite of geo-scientific writing found throughout in earth science dictionaries and textbooks. Time units include terms formalized by stratigraphic committees as well as informal constructs erected ad hoc to communicate more efficiently. With these time terms we partition Earth’s history into utilitarian and intuitively understandable time segments that vary in length over seven orders of magnitude: from the 225-year-long Anthropocene (Crutzen and Stoermer, 2000) to the ,4-billion-year-long Precambrian (e.g., Hicks, 1885; Ball, 1906; formalized by De Villiers, 1969)

Behavior of triplex SiC fuel cladding designs tested under simulated pressurized water reactor conditions

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.Includes bibliographical references (p. 101-107).A silicon carbide (SiC) fuel cladding for LWRs may allow a number of advances, including: increased safety margins under transients and accident scenarios, such as loss of coolant accidents; improved resource utilization via a higher burnup beyond the present limit of 62 GWd/MTU; and improved waste management. The proposed design, referred to as Triplex, consists of three layers: an inner monolith, a central composite, and an outer environmental barrier coating (EBC). The inner monolith consists of dense SiC which provides strength and hermeticity to contain fission products. The composite layer is made of SiC fibers, woven around the monolith, and then infiltrated with a SiC matrix. The composite layer adds strength to the monolith and provides a pseudo-ductile failure mode. The EBC is a thin coating of SiC applied to the outside of the composite to protect it against corrosion. The ends of the tubes may be sealed via the bonding of SiC end caps to the SiC tube. Triplex tube samples, monolith-only samples, and SiC/SiC bonding samples (consisting of two blocks bonded together) were tested in three phases as part of an evaluation of the SiC cladding system. A number of samples were exposed to PWR coolant and neutronic conditions using an incore loop in the MIT research reactor (MITR-II). Other samples remained in their as-fabricated states for comparison. First, mechanical testing revealed significant strength reduction in the Triplex samples due to irradiation-induced point defects, corrosive pitting of the monolith, and possible differences in the behavior of the Triplex components. Some manufacturing abnormalities were also discovered which could have compromised strength. The Triplex samples tested here were not as strong as reported in a previous study. SEM analysis was able to follow the propagation of cracks from initiation, at the monolith inner surface, to termination, upon breaching the EBC. The composite layer was found to be key in dissipating the energy driving the crack formation. Second, three SiC/SiC bonding methods (six samples total) were tested in the MITR-II to 0.2 dpa, and five of the six samples failed. SEM analysis indicates radiation induced degradation of the bond material. Dimensional and volume measurements established the anisotropic swelling of the two SiC blocks in each bond sample, which would have caused shear stresses on the bonds, contributing to their failure. Finally, thermal diffusivity measurements of the Triplex samples show substantial decreases with irradiation (saturating at about 1 dpa) due to the accumulation of phonon-scattering defects and corrosion of SiC. By 1 dpa, the thermal diffusivity/conductivity of this SiC cladding design is diminished to a value lower than that of Zircaloy. In the as-fabricated state, a large difference exists between the monolith-only and Triplex samples due to the phonon scattering centers at the interfaces of the layers. With irradiation this difference decreases, suggesting that similar corrosion and radiation damage effects exist in both the monolith and Triplex samples.by John D. Stempien.S.M

Tritium transport, corrosion, and Fuel performance modeling in the Fluoride Salt-Cooled High-Temperature Reactor (FHR)

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.Cataloged from PDF version of thesis.Includes bibliographical references (pages 294-305).The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is a pebble bed nuclear reactor concept fueled by tristructural isotropic (TRISO) fuel particles embedded in graphite spheres and cooled by a liquid fluoride salt known as "flibe" (7LiF-BeF2). A system of models was developed which enabled analyses of the performance of a prototypical pebble bed FHR (PB-FHR) with respect to tritium production and transport, corrosion, TRISO fuel performance, and materials stability during both normal and beyond design-basis accident (BDBA) conditions. A model of TRITium Diffusion EvolutioN and Transport (TRIDENT) was developed and benchmarked with experimental data. TRIDENT integrates the effects of the chemical redox potential, tritium mass transfer, tritium diffusion through pipe walls, and selective Cr attack by tritium fluoride. Systems for capturing tritium from the coolant were proposed and simulated with TRIDENT. A large nickel permeation window reduced the tritium release rate from 2410 to 800 Ci/EFPD. A large gas stripping system reduced tritium release rates from 2410 to 439 Ci/EFPD. A packed bed of graphite located between the reactor core and the heat exchanger reduced peak tritium release rates from 2410 to 7.5 Ci/EFPD. Increasing the Li-7 enrichment in flibe from 99.995 to 99.999 wt% reduced both the tritium production rate and the necessary sizes of tritium capture systems by a factor of 4. An existing TRISO fuel performance model called TIMCOAT was modified for use with PBFHRs. Low failure rates are predicted for modern uranium oxycarbide (UCO) TRISO fuels in a PBFHR environment. Post-irradiation examinations of surrogate TRISO particles determined that the outer pyrolytic carbon layer is susceptible to cracking if flibe were to freeze around the particles. Chemical thermodynamics calculations demonstrated that common constituents of concrete will not be stable in the event they contact liquid flibe. The chemical stability of fission products in reference to the coolant redox potential was determined in the event the TRISO UCO kernel is exposed to flibe during a BDBA. Noble gases (Kr and Xe) will escape the coolant. Cesium, strontium, and iodine are retained in the salt. All other important radionuclides are retained in the kernel or within the coolant system.by John Dennis Stempien.Ph. D

Catalyzed oxidation of IG-110 nuclear graphite by simulated fission products Ag and Pd nanoparticles

To evaluate the stability of nuclear materials in high temperature gas reactors under air ingress conditions, catalytic oxidation of IG-110 graphite by two simulated fission products, metallic Pd and Ag, was studied in oxidative atmosphere and at temperatures up to 1000 °C using an integrated furnace, mass spectroscopy and infrared spectroscopy system. Transmission electron microscopy and X-ray diffraction studies show that Pd and Ag nanoparticles were successfully introduced onto powdery IG-110 graphite through an impregnation and subsequent heat-treatment process. The combined mass spectroscopy and infrared spectroscopy methods allow simultaneous analysis of two gaseous products, CO and CO2, and separate measurements of activation energy for their formation reactions. It was found that the introduction of Pd or Ag to IG-110 graphite substantially catalyzed the oxidation of graphite, characteristic of decreased onset temperatures for the oxidation of graphite. Moreover, the catalytic effects by Pd and Ag are considerably different based on measured concentration ratios of CO2 to CO as a function of oxidation temperatures. Ag makes the graphite oxidation commence at approximately 400 °C with CO2 being the dominant product. In contrast, Pd significantly increases the concentration ratio of CO2 to CO at temperatures higher than approximately 690 °C, although it decreases the onset temperature for the oxidation reaction to around 525 °C. To understand the catalytic difference, the mechanism of the graphite oxidation is discussed based on the changes of surface oxygen species on Ag and Pd

Catalyzed Oxidation of Nuclear Graphite by Simulated Fission Products Sr, Eu, and I

The influence of three fission products Sr, Eu, and I on the oxidation of IG-110 nuclear graphite was studied in the temperature range of 400 to 1000 °C. Sr and Eu were introduced as chlorides, and I was introduced as NaI. The temperature dependence of both CO2 and CO production during the graphite oxidation measured with mass spectroscopy and infrared spectrometry shows that the introduction of these three compounds to graphite significantly decreases the onset temperature for the oxidation of graphite. Among the three compounds, NaI is the most active towards the oxidation reaction, characterized by a significant decrease of the onset temperature from approximately 650 to 400 °C before and after its introduction to graphite. Separate measurements of CO2 and CO concentration at varying temperatures enable the calculation of the activation energy for the formation of CO2 and CO. The activation energies for the oxidation of pure and fission product-impregnated graphite samples decrease in the following order: standard IG-110 graphite, EuCl3-impregnated IG-110, SrCl2-impregnated IG-110, and NaI-impregnated IG-110. This trend indicates that the three compounds catalyze the oxidation of graphite at temperatures relevant to the operation of high-temperature gas-cooled reactors. Furthermore, it is found that the three compounds can also affect the molar ratio of reaction products CO2 and CO, and the rates of the graphite oxidation. At temperatures higher than about 850 °C, the impregnated samples exhibit lower CO2: CO ratios than the pure graphite. Different from EuCl3 and NaI, the introduction of SrCl2 decreases the graphite oxidation rates at temperatures higher than about 770 °C. Their catalytic mechanism can be understood based on a redox cycle of the intermediate active species, promoting the dissociation of molecular oxygen and transfer to the carbon

EPMA-based mass balance method for quantitative fission product distribution comparison between TRISO particles

Two irradiated AGR-2 TRISO particles were chosen to demonstrate a recently developed mass balance technique in which EPMA-generated concentration data were used to determine fission product mass on a layer-by-layer basis in TRISO particles. EPMA-calculated fission product masses for most fission products in the two particles were within ± 20% of their ORIGEN-modelled masses. Results show that Sr, Ba, and Eu accumulate preferentially in the carbon-rich kernel periphery on the particles’ side that lacks a gap between the buffer and IPyC. In addition, the more mobile elements—Cs, Sr, Pd, and Ag, accumulate in greater quantity in the outer layers of particle AGR2-223-RS34 compared to particle AGR2-223-RS06, which has relatively more of those elements’ mass located in the kernel and kernel periphery, suggesting enhanced fission product transport in particle AGR2-223-RS34. This model can be used better understand and test fission product transport in TRISO particles. Graphical abstract: [Figure not available: see fulltext.]Accepted Author ManuscriptRST/Reactor Physics and Nuclear Material

Effects of Microstructure on the Oxidation Behavior of A3 Matrix-Grade Graphite

The oxidation behavior of matrix-grade graphite in air- or steam-ingress accident scenarios is of great interest for high-temperature gas reactors (HTGRs). In this study, the microstructures of two variants of matrix-grade graphite based on the German A3-3 and A3-27 formulations were characterized with scanning electron microscopy (SEM), transmission electron microscopy (TEM), and Raman spectroscopy, and correlated to oxidation behavior observed through thermogravimetric analysis (TGA) and differential scanning calorimetry (DSC). Through TEM imaging and selected area electron diffraction (SAED), a higher volume fraction of partially graphitized carbon was identified in the A3-3 type graphite than in the A3-27 type. This structure is believed to have contributed to the accelerated oxidation exhibited by A3-3 in the chemical reaction-controlled oxidation regime