Microstructural and Micro-Chemical Evolutions in Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 tristructural-isotropic (TRISO) fuel particles were fabricated using slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. Extensive microstructural and analytical characterizations were conducted to correlate those differences with the fuel kernels’ responses to neutron irradiations in terms of irradiated fuel microstructure, fission products’ chemical and physical states, and fission gas bubble evolutions. The studies used state-of-the-art transmission electron microscopy (TEM) equipped with energy-dispersive x-ray spectroscopy (EDS) via four silicon solid-state detectors with super sensitivity and rapid speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen in order to create representative irradiation conditions with fuel burnup in the range of 10.8 to 18.6% fissions per initial metal atom (FIMA) and time-average volume-average temperatures varying from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% and 14.03 wt.% for the AGR-1 and AGR-2 fuel kernels, respectively. The TEM results showed significant microstructural reconstructions in the irradiated fuel kernels from both the AGR-1 and AGR-2 fuels. There are four major phases: fuel matrix of UO2 and UC, U2RuC2, and UMoC2—in the irradiated AGR-2 fuel kernel. Zr and Nd form a solid solution in the UC phase. The UMoC2 phase often features a detectable concentration of Tc. Pd was mainly found to be located in the buffer layer or associated with fission gas bubbles within the UMoC2 phase. EDS maps qualitatively show that rare-earth fission products (Nd et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR-1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there was a high number of rod-shaped precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The differences in irradiated fuel kernel microstructural and micro-chemical evolution when comparing AGR-1 and AGR-2 TRISO fuel particles may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role. Limited electron energy-loss spectroscopy (EELS) characterizations of the AGR-2 fuel kernel show almost no carbon in the UO2 phase, but a small fraction of oxygen was detected in the UC/UMoC2 phase.

Development of Security and Safety Fuel for Pu-Burner HTGR: Part 2 — Design Study of Fuel and Reactor Core

JAEA (Japan Atomic Energy Agency) has conducted feasibility studies of the fuel and of the reactor core for the plutonium-burner HTGR (High Temperature Gas-cooled Reactor). The increase of the internal pressure, which is caused by generations of CO gas and stable noble gases, is considered to be the one of the major causes of TRISO (TRI-structural ISO-tropic) fuel failure at high burn-up. The CO gas is generated by the chemical reaction of the graphite making up the buffer layer with the free-oxygen released from the fuel kernel by fission. The stable noble gases, which are fission products, are also released from the fuel kernel. Although it is considered very difficult to suppress the increase of the partial pressure of the stable noble gases because of its chemically inert nature, the increase of the CO gas partial pressure can be suppressed by reducing the free-oxygen mole concentration using a chemical reaction. ZrC acts an oxygen getter, which reduces the free-oxygen generated with fission reaction. An increase of the CO gas partial pressure with burn-up in a TRISO fuel is expected to be suppressed by coating ZrC on a fuel kernel. A PuO2-YSZ (Yttria Stabilized Zirconia) fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter under a HTGR temperature condition was examined based on a thermochemical calculation. A preliminary feasibility study on the 3S-TRISO fuel that enables to attain a high burn-up around 500 GWd/t was also conducted focusing on a fuel failure caused by an increase of the internal pressure. Additionally, a preliminary nuclear analysis was conducted for the plutonium-burner HTGR with a fuel shuffling in the radial direction. As a result, the thermochemical calculation result showed that all the amount of the free-oxygen is captured by a thin ZrC coating under 1600°C condition. The plutonium-burner HTGR will be designed to suppress fuel temperature to be lower than 1600°C under severe accident conditions, and hence it was confirmed that coating ZrC on the fuel kernel is very effective method to suppress the internal pressure. The internal pressure the 3S-TRISO fuel at 500 GWd/t is calculated to be lower than 60 MPa, which allows to prevent the fuel failure, and hence the feasibility of the 3S-TRISO fuel was also confirmed. Additionally, the results of the whole core burn-up calculations showed that the fuel shuffling in the radial direction allows to achieve the high burn-up around 500 GWd/t. It also showed that the temperature coefficient of reactivity is negative value during the rated power condition through the operation period.

Development of Security and Safety Fuel for Pu-Burner HTGR: Part 4 — Study on 3S-TRISO Fuel Fabrication

In order to develop 3S-TRISO fuel for Pu-burner High Temperature Gas Reactor (HTGR), we conducted lab. scale experiments such as preparation test of simulated fuel kernel; CeO2-YSZ particle, and coating pre-test with simulated kernel. In the preparation test, based on the actual achievement of manufacturing fuel for High Temperature Engineering Test Reactor (HTTR)[1], we tried to fabricate some CeO2-YSZ particles through external gelation process. As a result, we successfully obtained the manufacturing parameters that can prepare good particles. In addition, we carried out some parametric coating test with fluidized-bed equipment and ZrO2 particle as simulated ZrC coated fuel kernel, and obtained the prospect of the possibility to coat the layer having desired thickness.

Reactivity Hold-Down Technique for a Soluble Boron Free PWR Using TRISO Particle Fuel

TRISO coated particle, developed for HTGR initially, has advantages of nuclear proliferation-resistance and fuel integrity against the release of fission products. In this paper, a 350MWt small sized PWR core design utilizing TRISO fuel concept is presented. TRISO particles are dispersed in graphite matrix to form the fuel compact, and then the fuel compact is clad by Zircaloy-4 cladding to form a fuel rod. The graphite matrix increases thermal conductivity of fuel compact, so that the fuel average temperature would be well below conventional PWRs’. In order to simplify reactor design, operation and maintenance, soluble boron free concept while operation is introduced. The emphasis of the study is put on the reactivity hold-down technique for the 350MWt PWR core. Excess reactivity is suppressed through a combination of Pu-240 adding with Gd2O3 loading. Pu-240 is added into UO2 fuel kernel of some assemblies, and Gd2O3 rods are loaded in other assemblies. The non-fissile plutonium isotope Pu-240 has a considerably high thermal neutron capture cross section compared to U-238, so that the Pu-240 added fuel can greatly suppress excess reactivity over burnup. Besides, reactor core life would be extended by adding proper amount of Pu-240 for its converting into Pu-241 which is a fissile isotope. Combining Pu-240 adding with Gd2O3 loading, the designed core reaches an average core burnup of approximately 58GWD/t, as well as a core life of nearly 6EFPY.

Chemical Compatibility at High Temperature between the Carbide Fuel UC or (U,Pu)C and SiC Cladding for the Gas cooled Fast Reactor

The chemical compatibility at high temperature between the fuel kernel (U,Pu)C and SiC cladding, the reference materials for the GFR reactor, is studied. For that purpose, a thermodynamic database on the U-Pu-C-Si system was developed with the Calphad method to calculate the phase diagrams. Differential thermal analysis experiments were performed to measure phase transition temperatures in Si-U and C-Si-U systems. According to the calculated isopleth section between the hyperstoichiometric uranium carbide UC1.02 and SiC, the materials shall not react below 2056 K, the temperature at which a liquid phase shall form. These calculations are in good agreement with two chemical compatibility tests performed at 1873 K and 2073 K between the materials. Calculations were also performed to study the chemical interaction between the mixed carbide (U,Pu)C1.04 and SiC. The presence of plutonium in the fuel kernel lowers the liquid formation temperature of 167 K.