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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A sensitivity analysis to predict the neutronics behavior of samples irradiated in the VTR rabbit system

We report a low-order neutronics model is developed to carry out hundreds of simulations efficiently and investigate the neutronics behavior of samples being irradiated in a test reactor setting under different geometrical constraints. The low-order model allowed for simulations that yield the expected neutronics behavior of any irradiated sample in any environment and allows for the calculation of highly accurate spatially averaged statistics and idealized spatial distributions in the neutron flux. Several benchmarks are performed to evaluate the performance and limitations of the low-order model revealing many important findings. The low-order model predicted the LHGR in the EBR-II driver fuel to within 2.34% by only simulating the fuel rod by itself, which served as a validation for the model. Sensitivity studies investigated 3% enriched UO 2 and U-10Zr being irradiated in the Versatile Test Reactor rabbit system. The analyses investigated a range of combinations of 15 radii and 5 heights for each sample in the rabbit system. Similar data sets are also provided for irradiations in the Advanced Test Reactor’s B-10 irradiation position, which is a thermal neutron spectrum environment. Generalized fits and fit coefficients are obtained for sample heating, reaction rate densities, and local multiplication rate characteristics, allowing the predictions of the neutronics behavior of the samples based on their geometrical constraints. The analyses and fits laid the groundwork for developing a user-end Multiphysics analysis framework to assist and accelerate irradiation experiment design and optimization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

The Effect of U-234 Content on the Neutronic Behavior of Uranium Systems

When analyzing uranium systems, the usual rule of thumb is to ignore the U-234 by assuming that it behaves neutronically like U-238. Thus for uranium systems, the uranium is evaluated as U-235 with everything else being U-238. The absorption cross section of U-234 is indeed qualitatively very similar to that of U-238. However, thermal absorption cross section of U-234 is about 100 times that of U-238. At low U-235 enrichments, the amount of U-234 is quite small so the impact of assuming it is U-238 is minimal. However, at high enrichments, the relative ratio of U-234 to U-238 is quite large (maybe as much as 1 to 5). Thus, one would expect that some effect of using the rule of thumb might be seen in higher enriched systems. Analyses were performed on three uranium systems from the set of Benchmarks. Although the benchmarks are adequately characterized as to the U-234 content, often, materials used in processing are not as well characterized. This issue may become more important with the advent of laser enrichment processes, which have little or no effect on the U-234 content. Analytical results based on the relationship of U-234 activity to that of U-235 have shown good predictive capability but with large variability in the uncertainties. Rucker and Johnson noted that the actual isotopics vary with enrichment, design of the enrichment cascade, composition of the feed material, and on blending of enrichments so there is considerable uncertainty in the use of models to determine isotopics. Thus, it is important for criticality personnel to understand the effects of variation of U-234 content in fissile systems and the impact of different modeling assumptions in handling the U-234. Analyses were done on LEU, IEU and HEU benchmarks from the International Handbook. These indicate that the effect of ignoring U-234 in HEU metal systems is non-conservative while it seems to be conservative for HEU solution systems. The magnitude of change in k-effective was as high as 0.4%, which has implications on selection of administrative margins and the determination of the upper subcriticality limit.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Depletion Analysis of Additively Manufactured Control Elements in HFIR

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory is a flux-trap type research reactor which provides one of the highest steady-state neutron sources in the world. HFIR neutrons are used for a range of research applications, including scattering experiments, materials irradiation, and isotope production. The HFIR core operates at 85 MW and is light water cooled and moderated. The core consists of several concentric cylindrical regions, as depicted in Fig. 1. At the center of the core is the flux trap, where the thermal flux reaches a maximum on the order of 10{sup 15} neutrons/cm{sup 2}-s. Immediately surrounding the flux trap are two annular fuel elements, an inner fuel element and an outer fuel element, each loaded with hundreds of thin fuel plates curved into an involute shape. Surrounding the outer fuel element are two thin control elements (CEs), an inner CE (ICE) and an outer CE (OCE). The CEs are axially zoned into three regions according to absorption strength: the white regions contain only structural aluminum and do not significantly contribute to absorption; the gray regions contain tantalum (Ta) as the primary absorber with a concentration of 30% by volume; and the black regions contain europium in the form of Eu{sub 2}O{sub 3} with a concentration of 33% by volume. A thick beryllium reflector is the final core component surrounding the CEs. A particular challenge to efficient maintenance and continued operation of HFIR is fabrication of the CEs, the current methods for which are tedious, outdated, and expensive. In an effort to improve the operational economics of HFIR and to demonstrate the application of advanced manufacturing techniques to the production of nuclear reactor components, the ultrasonic additive manufacturing (UAM) process is being investigated for fabrication of the HFIR CEs. This application of UAM involves directly embedding small cylindrical compacts (∼5.6 mm diameter, ∼1.7 mm thick) containing 60% neutron poison by volume into an aluminum matrix, using energy from ultrasonic vibrations to bond the compacts to the aluminum. The CEs contain discrete regions of absorbing material, in contrast to the standard design with homogeneously distributed absorbers. This novel design necessitates an investigation into how the HFIR core neutronic behavior differs from the case with homogeneous CEs so as to ascertain the feasibility of using additively manufactured CEs in HFIR. From an operations perspective, it is desirable for the discrete CEs to match the performance of the homogeneous CEs as closely as possible so as to avoid reconsideration of established safety margins. Earlier analyses of HFIR under beginning of cycle (BOC) conditions have indicated that although absorption properties on a local scale differ significantly between CE designs, the macroscopic behavior (neutron multiplication and flux and power distributions) of the HFIR core is not significantly altered. This study now addresses CE performance throughout depletion of the HFIR core during a reactor cycle. These analyses help to construct a more complete picture of the feasibility of employing additively manufactured CEs in HFIR and will reveal any adjustments which must be made to the design in order to maintain acceptable performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Simultaneous {sup 233}U and{sup 235}U characterization through the assay of delayed neutron temporal behavior

Aqueous solutions containing dissolved uranium-233 and uranium-235 were irradiated for 60's in the SLOWPOKE-2 reactor at the Royal Military College of Canada. The temporal behavior of the delayed neutrons produced was recorded by the Facility's Delayed Neutron Counting (DNC) system. The percentage of uranium-233 as a function of total fissile mass present in each sample ranged from 0 to 100% and was predicted by the DNC system with average absolute errors of {+-} 4%. Future work will upgrade the system electronics and software to reduce both uncertainties in timings and electrical noise. Mixture analysis will also be expanded to include plutonium-239 and fissile materials contained in non-aqueous matrices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Conformational effect on small angle neutron scattering behavior of interacting polyelectrolyte solutions: a perspective of integral equation theory

We present small angle neutron scattering (SANS) measurements of deuterium oxide (D2O) solutions of linear and star sodium poly(styrene sulfonate) (NaPSS) as a function of polyelectrolyte concentration. Emphasis is on understanding the dependence of their SANS coherent scattering cross section I(Q) on the molecular architecture of single polyelectrolyte. The key finding is that for a given concentration, star polyelectrolytes exhibit more pronounced characteristic peaks in I(Q), and the position of the first peak occurs at a smaller Q compared to their linear counterparts. Based on a model of integral equation theory, we first compare the SANS experimental I(Q) of salt free polyelectrolyte solutions with that predicted theoretically. Having seen their satisfactory qualitative agreement, the dependence of counterion association behavior on polyelectrolyte geometry and concentration is further explored. Our predictions reveal that the ionic environment of polyelectrolyte exhibits a strong dependence on polyelectrolyte geometry at lower polyelectrolyte concentration. However, when both linear and star polyelectrolytes exceed their overlap concentrations, the spatial distribution of counterion is found to be essentially insensitive to polyelectrolyte geometry due to the steric effect.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

Deformation Behavior in Neutron Irradiated Generation II FeCrAl Alloys for Accident Tolerant Fuel Cladding

Preliminary work has been completed to evaluate the deformation behavior of the Generation II FeCrAl alloy class. The Generation II alloy, C35M, was neutron irradiated within the High Flux Isotop e Reactor followed by post-irradiation examination to evaluate the mechanical properties, irradiated microstructure, deformation modes, and deformation microstructure. Complementary characterization techniques were used in this evaluation. It was found tha t aft er irradiation to1.8 dpa at 364 °C that the C35M specimen failed in a ductile manner. The observed deformation microstructure showed a high density of line dislocations, dislocation networks, and dislocation pile-ups. The observed performance and stru cture s are consistent with literature studies on high-Cr FeCr alloys that were thermally aged at 500°C. The observed results suggest that Generation II FeCrAl alloys retain ductility under low-dose, prototypical LWR irradiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Mechanical Behavior of Neutron Irradiated Refractory Multi-Principal Element Alloys Processed via Spark Plasma Sintering

The search for advanced materials capable of withstanding the extreme conditions of Generation IV reactors is a critical area in materials science research. These reactors operate under severe environments, including high temperatures, corrosion, stress, and irradiation damage. Consequently, there is a need for innovative alloy systems to ensure the reliability and longevity of proposed Generation IV reactor components. Refractory multi-principal-element alloys (RMPEA) have emerged as a promising candidate due to their exceptional properties. These alloys, characterized by their composition of multiple principal elements in near-equiatomic ratios, exhibit superior resistance to irradiation damage, reduced void swelling, enhanced microstructural stability, and minimal irradiation-induced hardening. While initial studies on RMPEAs have shown promising results, most research has been limited to thin films, nanocrystalline microstructures, and ion irradiation, which do not accurately represent the behavior of bulk materials. To address this gap, our research focused on the neutron irradiation of bulk RMPEAs. We aim to conduct comprehensive post-irradiation examinations (PIE) of RMPEAs irradiated at the Advanced Test Reactor at Idaho National Laboratory. The RMPEAs were synthesized using spark plasma sintering (SPS) with mechanically alloyed metallurgical powder. The RMPEA specimens are a MoNbTi alloy system with additions of -Zr, and -ZrV. Furthermore, PIE consisted of mechanical testing and advanced materials characterization. The mechanical testing consisted of sub-sized tensile testing, micro- and nano- indentation. Microstructural characterization included scanning electron microscopy and transmission electron microscopy. Mechanical testing coupled with advanced microscopy techniques provides insight into phase morphology and its effects on the mechanical properties of the RMPEA specimens. The results indicate that both pristine and irradiated RMPEA specimens exhibited brittle behavior during tensile testing, which can be attributed to their heterogeneous microstructure. The SPS manufacturing process did not include any post treatment, which resulted in a heterogeneous microstructure. Energy-dispersive X-ray spectroscopy revealed the presence of intermetallic such as laves phases within the microstructure. Specifically, Ti-rich precipitates were observed in the MoNbTi specimen, while Mo-rich precipitates were found in the MoNbTiZrV specimen. Nano-hardness testing of pristine samples showed that the laves phases exhibited higher hardness values compared to the matrix phase, suggesting that precipitate hardening is likely the dominant hardening mechanism in these specimens. The results from this work will be used to build a finite element model to predict mechanical behavior of future MPEA compositions. Thus, enabling for a streamlined approach to developing novel MPEAs for the nuclear industry.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Understanding TRISO Coated Particle Neutron Irradiation Behavior:Evolution of Advanced Micro Analysis and Electron Microscopy Approaches

Objectives and motivation• Provide data for fuel qualification in support of reactor licensing• Establish a domestic commercial TRISO fuel fabrication capability Approach • Focus is on developing and testing UCO TRISO fuel –Develop fuel fabrication and QC measurement methods, first at lab scale and then at industrial scale –Perform irradiation testingover a range of conditions (burnup, temperature, fast neutron fluence) –Perform post-irradiation examination and safety testingto demonstrate and understand performance during irradiation and during accident conditions –Develop fuel performance modelsto better predict fuel behavior –Perform fission product transport experiments to improve understanding and refine models of fission product transport

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

A Study of the Annealing Behavior of Neutron Irradiated Graphite

Recycle is being considered as a disposal option for irradiated graphite from gas cooled reactors. Thermal annealing was performed on irradiated graphite samples to establish what fraction of in-crystal (displacement damage) and ex-crystal (pore generation) damage could be recovered. The sample dimensions and electrical resistivity were measured after isochronal annealing at 500, 1000, 1500, and 2000 C. Sample dimensions were unaffected by annealing. Some fraction of the resistivity was unannealable at 2000 C, this behavior which was attributed to (a) structural changes resulting from pore generation (also causing dimensions to be unaffected by annealing), or, (b) defect structures in irradiated graphite that persist even at an annealing temperature of 2000 C. The resistivity annealing behavior was different in low irradiation temperature samples, Tirr<~400 C and high irradiation temperature samples Tirr > ~400 C, which indicated defect structures established during irradiation were different in these two temperature regimes. Moreover, the extent of recovery was much greater in the specimens with higher irradiation temperatures, and was small for lower dose low temperature specimens. Evidently, some in-crystal defects that scatter or trap electrons are still present in the crystal lattice after annealing and their tenacity depends upon the irradiation temperature and accumulated neutron dose.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Spin-orbit driven magnetic insulating state with J eff =1/2 character in a 4d oxide

The unusual magnetic and electronic ground states of 5d iridates has been shown to be driven by intrinsically enhanced spin-orbit coupling (SOC). The influence of appreciable but reduced SOC in creating the manifested magnetic insulating states in 4d oxides is less clear, with one hurdle being the existence of such compounds. Here we present experimental and theoretical results on Sr 4 RhO 6 that reveal SOC dominated behavior. Neutron measurements show the octahedra are both spatially separated and locally ideal, making the electronic ground state susceptible to alterations by SOC. Magnetic ordering is observed with a similar structure to an analogous J eff =1/2 Mott iridate. We consider the underlying role of SOC in this rhodate with density functional theory and x-ray absorption spectroscopy and find a magnetic insulating ground state with J eff =1/2 character.The unusual magnetic and electronic ground states of 5d iridates have been shown to be driven by intrinsically enhanced spin-orbit coupling (SOC). The influence of appreciable but reduced SOC in creating the manifested magnetic insulating states in 4d oxides is less clear, with one hurdle being the existence of such compounds. Here, we present experimental and theoretical results on Sr 4 RhO 6 that reveal SOC dominated behavior. Neutron measurements show the octahedra are both spatially separated and locally ideal, making the electronic ground state susceptible to alterations by SOC. Magnetic ordering is observed with a similar structure to an analogous J eff =1/2 Mott iridate. We consider the underlying role of SOC in this rhodate with density functional theory and x-ray absorption spectroscopy, and find a magnetic insulating ground state with J eff =12 character.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2014

Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Triply Periodic Minimal Surfaces Lattice Neutronics Scoping Studies

Triply Periodic Minimal Surfaces (TPMS) have seen extensive research in heat exchanger development due to their enhanced heat transfer performance. When applied to nuclear reactor cores, TPMS can more efficiently dissipate heat to the coolant, permitting increased power densities. The increased power density can be used to develop smaller reactors while retaining power output. Although TPMS core geometries promise enhanced heat transfer, the neutronic behavior of such geometries require further inquiry. In this study, the neutron multiplication and buckling are compared between variations of TPMS and traditional pincell core designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Beam-contamination-induced compositional alteration and its neutron-atypical consequences in ion simulation of neutron-induced void swelling

Although accelerator-based ion irradiation has been widely accepted to simulate neutron damage, neutron-atypical features need to be carefully investigated. In this study, we have shown that Coulomb force drag by ion beams can introduce significant amounts of carbon, nitrogen, and oxygen into target materials even under ultra-high vacuum conditions. The resulting compositional and microstructural changes dramatically suppress void swelling. By applying a beam-filtering technique, introduction of vacuum contaminants is greatly minimized and the true swelling resistance of the alloys is revealed and matches neutron behavior closely. These findings are a significant step toward developing standardized procedures for emulating neutron damage.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗