Fissions Per Initial Metal Atom Research Articles

TRISO Burnup-Dependent Failure Analysis of a HTGR Design-Basis Accident Using BISON

This work assesses the failure behavior of the silicon carbide (SiC) layer in TRIstructural ISOtropic (TRISO) fuel particles with BISON during both steady-state and transient conditions for the MHTGR-350 design by General Atomics. A one-dimensional BISON model for a uranium oxycarbide–bearing TRISO is developed to simulate a power level of 50 mW/particles up to 12.4% fissions per initial metal atom (FIMA). Stress predictions for the materials of interest are presented as a function of burnup, along with the SiC failure probability computed using Weibull statistics under the assumption of realistic SiC quality. A design-basis accident (DBA) that involves control rod withdrawal is then simulated in BISON at various burnup levels. Boundary conditions during the DBA are obtained from RELAP5-3D. The predicted SiC failure probability is 3 × 10−14%. This value does not increase during the transient as a function of burnup since irradiation induces a compressive hoop stress state that reaches a maximum absolute value of around 300 MPa. This compressive stress compensates the tensile loading conditions introduced by the transient. The tensile components appear amplified at higher burnups since they increase from 100 to 140 MPa going from 6% FIMA to 12.4% FIMA. Nonetheless, burnup provides a negligible impact on the overall failure probability predictions for SiC, as transient conditions do not translate to a stress increase toward tensile values at any of the considered burnups. Additionally, a two-dimensional (2-D) exploratory BISON model is developed to determine the effects of inner pyrolytic carbon (IPyC) crack formation on SiC. IPyC failure is predicted using Weibull statistics at 105 days from the beginning of irradiation, which is set as the time for crack initiation using XFEM in the 2-D BISON model. Stress concentrations induced by the crack cause the SiC failure probability to increase with respect to the intact particle case by approximately 10 orders of magnitude. Maximum stress concentrations are found at the time of crack formation, after which stress relaxation is observed during remaining steady-state irradiations and subsequent transient simulations. Considerations are made on the results of the 2-D model being contingent on both mechanical and anisotropy properties for pyrolytic carbon. Recommendations are provided for future work involving higher dimensionality models, further investigations on uncertainties, material properties, and additional designs, such as the fluoride salt-cooled high-temperature reactor, that operate at higher burnups.

Three-dimensional microstructural characterization of FBR MOX fuel and the contribution of microstructural features to the thermal conductivity of the fuel

Combination of microstructural characterization, property measurements, and phase field modeling is used to investigate fast breeder reactor (FBR) mixed oxide (MOX) fuel irradiated to burnup of 13.7% fissions per initial metal atom (FIMA). The fuel was characterized at different radial locations, which revealed that grey phase can be present in the central region if it nucleates on five metal precipitates (FMPs). In addition, in the mid-radial region FMPs do not diffuse out of the region once formed and the size of Pd–Te precipitates is dictated by the porosity present in the region. Thermal conductivity measurements (TCM) were conducted as a function of radial location and the microstructure of the fuel was correlated with the observed trend. Reconstructions of the 3D solid and gaseous fission product structures in different regions of the fuel were used to simulate the effective thermal conductivity (ETC) of the respective regions and determine which microstructural feature has the strongest impact on thermal conductivity. Based on conducted assessment, FMPs and Pd-Te precipitates improve local conductivity of central and mid-radial regions even in the presence of grey phase, but defects are primary contributor to the degradation of thermal conductivity on the periphery of the fuel.

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.

Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products

This work presents a comparison between the microstructural evolution of three annular fast-reactor mixed-oxide (MOX) fuel pellets irradiated to varying burnups at the Fast Flux Test Facility (FFTF). Fuel pellets irradiated to 3.4%, 13.7%, and 21.3% fissions per initial metal atom (FIMA) were examined using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) techniques. The cross-section of the low burnup pellet displayed minor structural changes, but the central annulus of the pellets at 13.7% and 21.3% FIMA shrank from their starting size. The high burnup fuel pellet featured streaking and porosity migration associated with columnar grain growth. The radial fission product distribution in each of the pellets had a higher number density of metallic particles >5 μm in diameter near the fuel centerline. Solid fission products in the fuel-cladding gap were observed in the low and intermediate burnup pellets. The low burnup sample showed minor accumulation of Ba in the gap, while the volatile Cs was primarily observed at the pellet surface. The intermediate burnup pellet displayed a porous mixture of fission products, consistent with the joint-oxide gain (JOG) that has been previously observed in fast-reactor MOX fuel pellets.

An improved method to evaluate the “Joint Oxyde-Gaine” formation in (U,Pu)O2 irradiated fuels using the GERMINAL V2 code coupled to Calphad thermodynamic computations

In this work, two different thermodynamic softwares, ANGE using the TBASE database, and OPENCALPHAD using the TAF-ID (Thermodynamics of Advanced Fuels – International Database), have been integrated into the GERMINAL V2 fuel performance code (of the PLEIADES platform) in order to evaluate the chemical state of (U, Pu)O2 fuel and fission products in sodium cooled fast reactors. A model to calculate the composition and the thickness of the “Joint-Oxyde Gaine” (JOG) fission product layer in the fuel-clad gap has been developed. Five fuel pins with a final burnup ranging between 3.8 and 13.4% FIMA (Fissions per Initial Metal Atom) have been simulated, and the calculated width of the fission product layer have been compared with post irradiation examinations. The two different thermodynamic softwares have been compared in terms of computation time and predicted fuel-to-clad gap chemistry. The main elements and phases encountered in the fission productlayer have been identified, and the impact of the changing oxygen potential has been explored.

Characterization of solid fission products in 13.7% FIMA MOX fuel using electron microscopy techniques

This work utilizes electron microscopy-based techniques to examine the radial behavior of solid fission products in plutonium (Pu) bearing mixed oxide (MOX) fuel irradiated to a burnup of 13.7% fissions per initial metal atom (FIMA). Metallic precipitates primarily consist of five fission products: ruthenium (Ru), rhodium (Rh), technetium (Tc), molybdenum (Mo), and palladium (Pd). The five metal precipitates (FMPs) examined in this work have low concentrations of Pd and Mo, with no major compositional differences along the fuel radius. A secondary Pd–Te metallic phase forms in cooler regions of the pellet, likely due to the diffusion of gaseous species away from the central void. X-ray chemical maps indicate that the Pd–Te phase can nucleate on the surface of FMPs before precipitating into separate particles. These particles were also found to alloy with iron (Fe) near the surface of the fuel pellet due to interdiffusion with the stainless-steel cladding. The insoluble perovskite oxide phase was found to form near the central void and at intermediate radial positions, but not at the fuel edge. These findings suggest that solid fission product phases form at varying counts and compositions along the fuel pellet radius, and thus should be considered when describing the thermal behavior of the fuel.

Radial microstructural evolution in low burnup fast reactor MOX fuel

Examination of irradiated fast reactor mixed-oxide (MOX) fuel was conducted to analyze microstructural behavior and defect evolution as a function of radial position. The fuel analyzed in this work was irradiated to 3.4% fissions per initial metal atom (FIMA) in the Fast Flux Test Facility. Scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) were used to analyze microstructural features and fission product secondary phases. A dual-beam focused ion beam (FIB)/SEM was used to prepare several samples along the fuel radius for transmission electron microscopy (TEM) defect analysis. The microstructure near the center of the fuel pellet showed separated grain boundaries accompanied by fission gas bubbles. The fuel pellet did not reach the threshold linear heating rate (LHR) needed for restructuring, formation of the columnar or equiaxed regions was not observed. Metallic fission products aggregated in pores along grain boundaries near the pellet center, but no metallic precipitates were visible in the outer region. Moreover, no appreciable amount of insoluble perovskite phase was formed. The dislocation loop density is highest near the fuel central void but decreases past the mid-point of the fuel radius likely due to the self-irradiation effects. Dislocation lines follow an inverse trend, with the lower density observed near the hottest regions and increasing in the pellet outer radius. These results indicate that the thermal gradient in MOX fuels heavily influences the localized fission product morphology and defect behavior.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

On the discrepancies between FIMA and specific burnup

Nuclear fuel consumption is most commonly described by using two indicators: Fissions per Initial Metal Atom (FIMA) and energy generated per unit mass of initial heavy metal (GWd/tHM). FIMA indicates physical consumption of nuclides, while burnup is practical for industrial applications and comparisons. Both quantities are used as indicators in fuel cycle research and development studies.In this paper we demonstrate that the two analyzed indicators are not exactly proportional to each other. The exact ratio between these factors is derived to be dependent on the content of nuclear fuel. More precisely, it is a function of the ratio of the average energy generated in a system per fission to the average atomic mass of heavy metal. The predictions are proved by computation tests using the SERPENT code applied to common PWR assembly geometry. For the fuels comprising fissile isotopes of uranium and plutonium, the observed FIMA discrepancies are as big as 1.8% at burnup equal to 750 GW d/tHM. Concerning realistic content of industrial fuels such as UOX, MOX and 233U/Th loadings, the differences of FIMA reach 0.10% at about 48 GW d/tHM. The trends have been well predicted by the approximated theoretical model. The conclusions are meaningful for any comparisons of fuel cycle scenarios and are valid for any burnup calculations.

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800°C during post-irradiation safety tests. Fourteen compacts (a total of ∼58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800°C, the data suggest that release of these elements through intact coatings may become significant after ∼100h. Cesium was very well retained by intact SiC layers, with a fractional release <5×10−6 after 300h at 1600°C or 100h at 1800°C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300h at 1600 or 1700°C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ∼210h at 1800°C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

“Trapped fraction” effect in microfuel with plutonium fuel

Results of the investigation into the thermodynamics of two types of microfuel (with oxygen getter and without it) with plutonium fuel for various degrees of burnup are presented. The behavior of the trapped fraction of Ag, Ce, Cd, Cs, La, Mo, Pu, Pd, Ru, Sr, Te, and Y is investigated. The fraction of any fission product bound into stable chemical compounds with other components of the system and excluded from the diffusion process is called the trapped fraction. An abrupt increase in the trapped fraction of cesium and, correspondingly, a decrease in free cesium during the burnup of ∼26% FIMA (fissions per initial metal atom) or more are found for the microfuel containing no oxygen getter. This leads to a substantial nonlinear burnup dependence of the trapped fraction and should be the cause of an abrupt decrease in the Cs output from the microfuel. It seems likely that the found effect is associated with the formation of carbonate Cs2CO3 in the plutonium fuel. In the case of microfuel containing the oxygen getter, no formation of cesium carbonate occurs and the trapped fraction of cesium is almost independent of the burnup.

Microstructural characterization of high burn-up mixed oxide fast reactor fuel

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

Distribution of fission products in irradiated graphite materials of HTGR fuel assemblies: Third and fourth OGL-1 fuels

Axial, circumferential and radial distributions of fission products in the graphite sleeve, inner-tube and block of irradiated high temperature gas-cooled reactor (HTGR) fuel assemblies were measured by gamma spectrometry, lathe sectioning and beta counting with ion-exchange separation. Some distinctive peaks of 144Ce and 125Sb in their axial profiles, together with the very high activity level of fission products are ascribed to the failure of coated fuel particles. The effective retention capability of the graphite sleeve was observed for 90Sr, 106Ru, 125Sb, 144Ce and 155Eu; whereas not for 134Cs and 137Cs. Silver-110m was detected in graphite materials of the fourth OGL-1 fuel assembly with an increased burnup of 1.96% fissions per initial metal atom (FIMA). Effective in-pile diffusion coefficients of 90Sr, 125Sb and 144Ce in the graphite sleeves have been estimated using the Fickian diffusion theory.