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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

PHISICS/RELAP5-3D RESULTS FOR EXERCISES II-1 AND II-2 OF THE OECD/NEA MHTGR-350 BENCHMARK

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) High-Temperature Gas-Cooled Reactor (HTGR) Methods group currently leads the Modular High-Temperature Gas-Cooled Reactor (MHTGR) 350 benchmark. The benchmark consists of a set of lattice-depletion, steady-state, and transient problems that can be used by HTGR simulation groups to assess the performance of their code suites. The paper summarizes the results obtained for the first two transient exercises defined for Phase II of the benchmark. The Parallel and Highly Innovative Simulation for INL Code System (PHISICS), coupled with the INL system code RELAP5-3D, was used to generate the results for the Depressurized Conduction Cooldown (DCC) (exercise II-1a) and Pressurized Conduction Cooldown (PCC) (exercise II-2) transients. These exercises require the time-dependent simulation of coupled neutronics and thermal-hydraulics phenomena, and utilize the steady-state solution previously obtained for exercise I-3 of Phase I. This paper also includes a comparison of the benchmark results obtained with a traditional system code “ring” model against a more detailed “block” model that include kinetics feedback on an individual block level and thermal feedbacks on a triangular sub-mesh. The higher spatial fidelity that can be obtained by the block model is illustrated with comparisons of the maximum fuel temperatures, especially in the case of natural convection conditions that dominate the DCC and PCC events. Differences up to 125 K (or 10%) were observed between the ring and block model predictions of the DCC transient, mostly due to the block model’s capability of tracking individual block decay powers and more detailed helium flow distributions. In general, the block model only required DCC and PCC calculation times twice as long as the ring models, and it therefore seems that the additional development and calculation time required for the block model could be worth the gain that can be obtained in the spatial resolution

Advanced Reactor Technologies↗
DOE OSTI2016

Influence of hydride precipitation on the corrosion kinetics of Zircaloy-4: effect of the nano-structure and grain boundary properties of the zirconium oxide layer on oxygen diffusion flux - 2016-0073

Hydride precipitation at the oxide-metal interface is frequently proposed as causing the corrosion acceleration of Zircaloy-4 at high burnup in pressurized water reactors (PWRs). In order to identify the local mechanisms possibly involved, we studied the nano-structure of oxides formed on massive zirconium hydrides and reference Zircaloy-4 with an innovative grain mapping technique with the use of transmission electron microscopy (TEM). In autoclave PWR-conditions, the presence of a precipitated hydride phase, previously formed by a cathodic charging technique at the surface of Zircaloy-4, clearly increased the corrosion rate, and a higher oxygen diffusion flux along oxide grain boundaries is observed compared with the reference Zircaloy-4. The texture, grain size, and adjacent grain misorientation in the oxide layer were studied in detail along the direction of the oxide growth using an automated crystal orientation mapping technique associated with TEM for both pre-hydrided and reference Zircaloy-4 samples. The texture in the growth direction is similar on pre-hydrided and reference samples, but the grain-to-grain mis-orientations showed differences. Indeed, on the pre-hydrided sample, the misorientation of 90 deg. C with respect to the [001] monoclinic axis is less probable than in the reference oxide, and more mis-orientations of 50-70 deg. C and 120-150 deg. C, corresponding to larger mismatches between neighboring grain boundaries, are observed. A smaller average diameter of the columnar monoclinic grains is also clearly revealed for the oxide grown on the pre-hydrided sample that leads to a larger number of diffusion paths for oxidizing species. These results are discussed and used for simulating oxygen diffusion flux through the polycrystalline microstructure of the oxide layer as a function of the grain size. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data

Here, a computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [ 54 Fe(n,p) 54 Мn, 93 Nb(n,n') 93 mNb, and 58 Ni(n,р) 58 Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54 Fe, 56 Fe, 57 Fe, and 16 O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Molten Salt Reactor Experiment Simulation using Shift/Griffin

The Department of Energy (DOE)’s NEAMS focuses its efforts on the development of advanced modeling and simulation (M&S) tools for light-water reactors (LWRs) and non–LWRs (i.e., molten salt reactors, high-temperature gas reactors, microreactors, and fast reactors). In the previous fiscal year, the Multiphysics Applications Driver Technical Area funded molten salt reactor (MSR) M&S at Oak Ridge National Laboratory (ORNL) to generate multigroup macroscopic cross sections with Shift for a MSRE 2D lattice model in Griffin. In addition, Shift’s capability to calculate gamma dose rates from activated components in the primary exchangers in a molten salt breeder reactor was also demonstrated. In fiscal year 2023, ORNL generated multigroup macroscopic cross sections using Shift for a 3D MSRE core model. MSRE depletion calculations using Griffin were also demonstrated in this fiscal year. For the depletion calculation, one-group microscopic cross sections for the 3D MSRE core were generated using Shift, and the decay transmutation library from ORIGEN was converted to an ISOXML file, which is required as input in Griffin. Several Monte Carlo codes, such as OpenMC and Serpent, were also used to benchmark and supplement multigroup cross sections generated by Shift. Multigroup libraries were generated with 8 and 20 group structures, and the study found the 8-group structure to be more accurate when comparing Griffin results to continuous energy (CE) Monte Carlo results. The average flux from CE Shift calculations is up to ~6% higher than the CE Serpent calculations because of different values applied for the energy released per fission (κ values). The average flux in the fuel salt calculated by Griffin using cross sections generated with Shift agrees well with the reference CE Shift solution; the same is valid for the corresponding Serpent results. The maximum relative error is ~6% and ~2% compared to the CE Shift and Serpent reference solutions, respectively. Meanwhile, the average flux calculated by Griffin in the graphite moderator shows a higher difference in the thermal range when compared to both reference Monte Carlo solutions; this result suggests a need for improvement in cross section generation for the graphite moderator in the thermal range in both Monte Carlo codes. Griffin depletion calculations using cross sections from Shift and Serpent were performed and compared against ORIGEN calculations, and the nuclide densities calculated by Griffin were found to be generally in agreement with those of ORIGEN. Because a different approach was taken to calculate the energy released per fission ( κ values) in Shift and Serpent, a difference in nuclide densities from differences in the average flux was observed between Griffin using Serpent and Shift cross sections. Griffin calculations with Shift cross sections produced higher average flux in the salt than with Serpent cross sections, leading to higher consumption of 235 U and higher production of 135 Xe. For time-dependent depletion calculations, cross sections were generated with Serpent, and Griffin’s results using these cross sections were compared to CE Serpent depletion results, demonstrating good agreement. The average difference in keff between Serpent and Griffin as a function of burnup is about 155 pcm. Similarly, good agreement with small differences up to ~0.5% was also noticed in the nuclide density of 235 U and 135 Xe. More details regarding the methodologies invoked to generate the cross section to make code-to-code comparisons are discussed further in this report. User feedback on Griffin and Shift capabilities that will enhance these calculations is provided in this report for future consideration. The work performed this fiscal year can be extended further for multiphysics coupling of Griffin-Pronghorn/SAM with Mole to study precursor flow and salt chemistry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Cell Dancoff-based embedded self-shielding capability for doubly heterogeneous particulate fuels in SCALE/Polaris

Polaris is a two-dimensional (2D) transport lattice physics capability in the SCALE code system for light-water reactor (LWR) analysis. SCALE/Polaris is being extended to treat doubly heterogeneous particulate fuels in LWRs with accident-tolerant fuels and prismatic high-temperature gas-cooled reactor (HTGR) fuel. Recently, a Dancoff-based Wigner-Seitz approximation capability based on the embedded self-shielding method (DWA-ESSM) was implemented into Polaris for efficient computation. A new double heterogeneity treatment capability based on DWA-ESSM in conjunction with the Hebert's collision probability method for double heterogeneity was developed and implemented for Polaris. The new capability was verified by performing benchmark calculations for the prismatic HTGR fuel compact problems with various design parameters through a code-to-code comparison between the Polaris and continuous-energy Monte Carlo results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Autonomous Radiation Survey Inside Contaminated Buildings at Fukushima Daiichi with the Riser UAV - 18299

Nuclear decommissioning is constrained and defined by radiation and radioactive contamination. Where radiation is high and its distribution is unknown simply gathering the data needed for project planning can become a challenge. Fukushima Daiichi epitomises this challenge. The complex sequence of events at the plant following the March 2011 earthquake and tsunami has resulted in extensive physical damage and an essentially unpredictable distribution of contamination. TEPCO have applied a range of manual and robotics techniques to map the status of the plant, but many significant areas have remained no-go zones due to a combination of high radiation, physical hazards and obstacles. Further, even in areas where manual survey is feasible, there is strong desire to minimise human exposure to radiation by deploying robots where possible. Riser is a fully autonomous UAV equipped with an infrastructure-free 3D navigation and mapping system, inspection cameras and a CZT spectrometric measurement capability. The system has been custom designed specifically to meet the needs of the nuclear inspection. Riser is capable of accessing areas with dose rates of hundreds of milli-sievert per hour and producing 3D models and radiation surveys with accurately located data points. The data from Riser can subsequently be fused using Createc's N-Visage engine to produce three-dimensional contamination maps which can be used for predictive modelling of future D and D activity. Riser has been deployed at Fukushima Daiichi since March 2017. We present examples of data and analysis outputs, as well as general learning about deploying autonomous robots at a nuclear site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Design and screening of nanoprecipitates-strengthened advanced ferritic alloys

Advanced nuclear reactors as well as the life extension of light water reactors require advanced alloys capable of satisfactory operation up to neutron damage levels approaching 200 displacements per atom (dpa). Extensive studies, including fundamental theories, have demonstrated the superior resistance to radiation-induced swelling in ferritic steels, primarily inherited from their body-centered cubic (bcc) structure. This study aims at developing nanoprecipitates strengthened advanced ferritic alloys for advanced nuclear reactor applications. To be more specific, this study aims at enhancing the amorphization ability of some precipitates, such as Laves phase and other types of intermetallic phases, through smart alloying strategy, and thereby promote the crystalline®amorphous transformation of these precipitates under irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2018

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. Thus,to ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ~60μm. Discs 3mm in diameter were then oxidized at temperatures between 575°C and 625°C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575°C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development of random geometry capability in RMC code for explicit analysis of polytype particle/pebble - 18493

Monte Carlo method plays an important role in accurate simulation of High Temperature Gas-Cooled Reactors (HTR) and molten salt reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the innovative design of pebble bed, prism HTRs and molten salt reactors, there may be more than one kind of particles or pebbles in the system, such as the burnable poison particles, graphite pebbles or fuel pebbles with different enrichments. Different treatments have been used to deal with the double-heterogeneous geometry of these reactors. In some researches, the particle distributions in stochastic media are assumed the regular lattice, while in the others, the explicit packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) are used, by generating the random coordinates of the particles/pebbles. However, traditional geometry with repeated structure has difficulties in explicit modeling polytype particle/pebble, due to the arbitrary packing fractions of different kinds of particles or pebbles. Although, packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) could solve the polytype particle/pebble problems to some extent, but these methods are troublesome in geometry modeling and time consuming. In this paper, a new Random Universe Geometry method was developed and implemented in Monte Carlo code RMC, for explicit analysis of polytype particle/pebble in double heterogeneous geometry systems, such as HTRs and molten salt reactors. This method was applied to constructed mixed particles/pebbles problems and also simplified HTR-10 full core problem to test the feasibility, and the results show that the random geometry modeling can provide explicit packing fraction of different kinds of particles or pebbles. The results are also compared with that calculated by RSA method and regular structure, and agree quite well. This work would provide an effective tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Helium-implantation-induced lattice strains and defects in tungsten probed by X-ray micro-diffraction

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. Bombardment with energetic fusion neutrons causes collision cascade damage and defect formation. Interaction of defects with helium, produced by transmutation and injected from the plasma, modifies defect retention and behaviour. Here we investigate the residual lattice strains caused by different doses of helium-ion-implantation into tungsten and tungsten-rhenium alloys. Energy and depth-resolved synchrotron X-ray micro-diffraction uniquely permits the measurement of lattice strain with sub-micron 3D spatial resolution and ~10 -4 strain sensitivity. Increase of helium dose from 300 appm to 3000 appm increases volumetric strain by only ~2.4 times, indicating that defect retention per injected helium is ~3 times higher at low helium doses. This suggests defect retention is not a simple function of implanted helium dose, but strongly depends on material composition and presence of impurities. Conversely, analysis of W-1 wt% Re alloy samples and of different crystal orientations shows that both the presence of rhenium, and crystal orientation, have a comparatively small effect on defect retention. These insights are key for the design of armour components in future reactors where it will be essential to account for irradiation-induced dimensional change when predicting component lifetime and performance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterising Ion-Irradiated FeCr: Hardness, Thermal Diffusivity and Lattice Strain

Ion-irradiated FeCr alloys are useful for understanding and predicting neutron damage in the structural steels of future nuclear reactors. Previous studies have largely focused on the structure of irradiation induced defects, probed by transmission electron microscopy (TEM), as well as changes in mechanical properties. Across these studies, a wide range of irradiation conditions has been employed on samples with different processing histories, which complicates the analysis of the relationship between defect structures and material properties. Furthermore, key properties, such as irradiation-induced changes in thermal transport and lattice strain, are little explored. Here we present a systematic study of Fe3Cr, Fe5Cr and Fe10Cr binary alloys implanted with 20 MeV Fe3+ ions to nominal doses of 0.01 dpa and 0.1 dpa at room temperature. Nanoindentation, transient grating spectroscopy (TGS) and X-ray micro-beam Laue diffraction were used to study the changes in hardness, thermal diffusivity and strain in the material as a function of damage and Cr content. Our results suggest that Cr leads to an increased retention of irradiation-induced defects, causing substantial changes in hardness and lattice strain. However, thermal diffusivity varies little with increasing damage and instead degrades significantly with increasing Cr content in the material. We find significant lattice strains even in samples exposed to a nominal displacement damage of 0.01 dpa. The defect density predicted from the lattice strain measurements is significantly higher than that observed in previous TEM studies, suggesting that TEM may not fully capture the irradiation-induced defect population. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

FeCr alloys↗
DOE OSTI2014

Diffusional Interaction between U-10 wt.% Zr and Fe at 903K, 923K and 953K (630°C, 650°C, and 680°C)

U-Zr metallic fuels cladded in Fe-alloys are being considered for application in an advanced Sodium-Cooled Fast Reactor (SFR) that can recycle the U-Zr fuels and minimize the long-lived actinide waste. To understand the complex fuel-cladding chemical interaction between the U-Zr metallic fuel with Fe-alloys, a systematic multicomponent diffusion study was carried out using solid-to-solid diffusion couples. The U-10 wt.% Zr vs. pure Fe diffusion couples were assembled and annealed at temperatures, 903, 923 and 953K for 96 hours. Development of microstructure, phase constituents, and compositions developed during the thermal anneals were examined by scanning electron microscopy, transmission electron microscopy and X-ray energy dispersive spectroscopy. Complex microstructure consisting of several layers that include phases such as U 6 Fe, UFe 2 , ZrFe 2 , α-U, β-U, Zr-precipitates, Χ, ε and λ were observed. Multi-phase layers were grouped based on phase constituents and microstructure, and the layer thicknesses were measured to calculate the growth constant and activation energy. The local average compositions through the interaction layer were systematically determined, and employed to construct semi-quantitative diffusion paths on isothermal U-Zr-Fe ternary phase diagrams at respective temperatures. The diffusion paths were examined to qualitatively estimate the diffusional behavior of individual components and their interactions. Furthermore, selected area diffraction analyses were carried out to determine, for the first time, the exact crystal structure and composition of Χ, ε and λ-phases. The Χ, ε and λ-phases were identified as Pnma(62) Fe(Zr,U), I4/mcm(140) Fe(Zr,U) 2 , and I4/mcm(140) U 3 (Zr,Fe), respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Mechanism of Methane Chemical Looping Combustion with Hematite Promoted with CeO 2

Chemical looping combustion (CLC) is a promising technology for fossil fuel combustion that produces sequestration-ready CO{sub 2} stream, reducing the energy penalty of CO{sub 2} separation from flue gases. An effective oxygen carrier for CLC will readily react with the fuel gas and will be reoxidized upon contact with oxygen. This study investigated the development of a CeO{sub 2}-promoted Fe{sub 2}O{sub 3}-hematite oxygen carrier suitable for the methane CLC process. Composition of CeO{sub 2} is between 5 and 25 wt % and is lower than what is generally used for supports in Fe{sub 2}O{sub 3} carrier preparations. The incorporation of CeO{sub 2} to the natural ore hematite strongly modifies the reduction behavior in comparison to that of CeO{sub 2} and hematite alone. Temperature-programmed reaction studies revealed that the addition of even 5 wt % CeO{sub 2} enhances the reaction capacity of the Fe{sub 2}O{sub 3} oxygen carrier by promoting the decomposition and partial oxidation of methane. Fixed-bed reactor data showed that the 5 wt % cerium oxides with 95 wt % iron oxide produce 2 times as much carbon dioxide in comparison to the sum of carbon dioxide produced when the oxides were tested separately. This effect is likely due to the reaction of CeO{sub 2} with methane forming intermediates, which are reactive for extracting oxygen from Fe{sub 2}O{sub 3} at a considerably faster rate than the rate of the direct reaction of Fe{sub 2}O{sub 3} with methane. These studies reveal that 5 wt % CeO{sub 2}/Fe{sub 2}O{sub 3} gives stable conversions over 15 reduction/oxidation cycles. Lab-scale reactor studies (pulsed mode) suggest the methane reacts initially with CeO{sub 2} lattice oxygen to form partial oxidation products (CO + H{sub 2}), which continue to react with oxygen from neighboring Fe{sub 2}O{sub 3}, leading to its complete oxidation to form CO{sub 2}. The reduced cerium oxide promotes the methane decomposition reaction to form C + H{sub 2}, which continue to react with Fe{sub 2}O{sub 3}/Fe{sub 3}O{sub 4} to form CO/CO{sub 2} and H{sub 2}O. This mechanism is supported by the characterization studies, which also suggest that the formation of carbonaceous intermediates may affect the reaction rate and selectivity of the oxygen carrier.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Effects of Minor Alloying Elements and Impurities on Long-Range Ordering in Ni-Cr-Based Alloys

There is growing concern about long-range ordered Ni2Cr phase forming during thermal aging of Ni-Cr-based alloys (e.g., 690, 625), which leads to lattice contraction, hardening, and embrittlement, possibly limiting the service life of light water reactor components. The timespan when this phase transformation occurs can vary in orders of magnitude between model and commercial alloys and is likely related to the presence of major (e.g., Fe, Mo) and minor alloying elements (e.g., Ti, Si, Nb). These minor alloying elements have been hypothesized to bind with vacancies and therefore limit diffusion. Contrarily, impurities (e.g., P, S) are believed to enhance vacancy diffusion and accelerate the Ni2Cr phase transformation because of a vacancy drag mechanism. This study implements experimental isothermal aging of model Ni-Cr binary and Ni-Cr-X ternary alloys, where X is Ti, Si, Nb, and P to help understand how each minor alloying element affects the Ni2Cr phase transformation timeline. Synchrotron-based X-ray diffraction and hardness measurements are used to quantify the evolution of the Ni2Cr phase transformation through a variety of short aging times to identify the onset of long-range order. These experimental results are compared to first principles simulations from literature to understand the mechanism by which each element impacts vacancy diffusion and the formation of the long-range ordered phase, Ni2Cr.

36 - MATERIALS SCIENCE↗
DOE OSTI2014

Mesh Copy/Move/Merge Tool for Reactor Simulation Applications

Reactor core simulations require the construction and mesh generation for core models consisting of lattices of fuel and other rods grouped into assemblies, and lattices of assemblies of several types grouped into a core model. A set of tools has been described for generating assembly and core lattice models. Both rectangular and hexagonal lattices are supported. The tools operate in three stages. First, assembly models of various types can be generated by the AssyGen tool, based on input describing the content of unit cells, the arrangement of unit cells in the lattice, and the extent of the lattice and any surrounding material. After generating the assembly model, the model is meshed with the CUBIT mesh generation toolkit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, they are arranged in a core model using the CoreGen tool. The input for CoreGen is similar to that of AssyGen, with assembly models substituted for unit cells. AssyGen and CoreGen also annotate the models with material and volume groupings necessary for specifying materials and boundary conditions required by the analysis. The AssyGen and CoreGen tools are packaged in the open-source MeshKit library for mesh generation; download and build instructions are included in this document.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗