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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 1,639 records · Page 91
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Effects of elevated temperature exposure on the residual stress state and microstructure of PVD Cr coatings on SiC investigated via in situ X-ray diffraction and transmission electron microscopy

Pure Cr coatings are being considered as corrosion-resistant layers for SiC-SiC f composite fuel cladding in light water reactors. The residual stress state and interdiffusion behavior of Cr coatings deposited on SiC substrates using bipolar high-power impulse magnetron sputtering (B-HiPIMS) and direct current impulse magnetron sputtering (DCMS) were studied using in situ X-ray diffraction experiments at 300 °C, 700 °C, and 1200 °C. After elevated temperature exposure, evolution of the coating microstructure and coating/substrate interdiffusion was characterized by transmission electron microscopy. The compressive residual stress of the dense B-HiPIMS coating at room temperature was gradually relieved at 300 °C, while the stress transitioned to a tensile state at 700 °C. Grain growth of the Cr coating and a Cr–SiC interdiffusion layer consisting of newly formed Cr–C and Cr–Si phases were observed in the B-HiPIMS coating after exposure at 700 °C. The phases in the interdiffusion layer were identified to be Cr 7 C 3 , Cr 3 Si, and Cr 5 Si 3 C x experimentally and by thermodynamic predictions. For the DCMS coating, however, changes in residual stress and microstructure at the elevated temperatures were limited by the inherent porosity and inhomogeneous structure of the as-deposited coating. At 1200 °C, both coating types interacted rapidly with the underlying SiC substrate in the early stages of the experiment. The results from this study suggest that the characteristics of the Cr coatings do not significantly change at the reactor operating temperature of 300 °C, while changes in microstructure and stress state are expected in off-normal conditions of 700 °C and 1200 °C. Finally the results from this study suggest the B-HiPIMS coating is best-suited as a protective coating for SiC fuel cladding.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Measuring the elastic tensor of a monolithic SiC hollow cylinder with resonant ultrasound spectroscopy

Resonant Ultrasound Spectroscopy (RUS) has become a standard tool in material science to measure nondestructively the elastic properties of materials. The widespread adoption of this tool for applications related to nondestructive testing and structural health monitoring in the industry and other places has been restricted thus far by some requirements imposed by RUS on the sample to be tested, in particular the fact that the sample geometry should be relatively simple (e.g., cuboid, cylinder, or sphere). Here, we use RUS to determine the baseline material properties of a hollow cylinder made of monolithic SiC, a surrogate material and geometry representing nuclear fuel cladding for this study. Typically, elastic properties of the nuclear fuel cladding samples are inspected destructively (e.g., indentation) or with acoustic techniquesrequiring multiple and cumbersome measurements. Here, a relatively new RUS approach specifically developed to measure the elastic tensor of samples with arbitrary geometries is used. The results obtained from this RUS approach are compared to values reported in the literature. As a second verification step, vibrational mode shapes computed numerically with the elastic constants are compared to those measured on the sample using a 3D laser Doppler vibrometer, showing an excellent agreement.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes for ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special handling, machining, welding, and inspection of materials, if known, should also be communicated to the experiment fabrication and inspection team.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in the relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

RISMC industry application #1 (ECCS/LOCA). Core characterization automation: Lattice codes interface for PHISICS/RELAP5-3D

The United States Nuclear Regulatory Commission (U.S. NRC) is currently proposing rule-making “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated anymore as separated from other disciplines, but needs to be evaluated in a coupled manner with other physics. The Risk-Informed Safety Margin Characterization (RISMC) Pathway initiated several programs at Idaho National Laboratory (INL) to support the industry in the transition to the new rule. One of these programs is the Industry Application #1 (IA1). IA1’s goal is to develop an Integrated Evaluation Model (IEM) called LOCA Toolkit for the US (LOTUS). This tool connects five major disciplines involved in LOCA analysis, among them is (automated) core design (CD-A). Here, this paper focuses solely on the design of LOTUS' CD-A module as well as a demonstration application of the CD-A module to a generic Pressurized Water Reactor (PWR). The demonstration of the other LOTUS disciplines together with a demonstration example of the final risk-informed safety margin analysis is published in a separate paper. LOTUS is meant to provide the methodology framework into which the user can plug his established tools and codes. With the CD-A, the plant owner/operator will characterize his core design with the tools he has. He will then integrate his tools/methods into the LOTUS framework for the downstream safety analysis. For the current CD-A demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D, while cross section generation is done using the Studsvik lattice code HELIOS-2.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

EVALUATION OF HOT CHANNEL FACTOR FOR SODIUM-COOLED FAST REACTORS WITH MULTI-PHYSICS TOOLKIT

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Hot channel factor (HCF)↗
DOE OSTI2021

Evaluation of ceria as a surrogate material for UO 2 in experiments on fuel cracking driven by resistive heating

A variety of normal operation and accident scenarios can generate thermal stresses large enough to cause cracking in light-water reactor (LWR) fuel pellets. Cracking of fuel pellets can lead to reduced heat removal, larger centerline temperatures, and localized stress in cladding all of which impact fuel performance. Furthermore, pellet cracking also contributes to a temperature reduction in the pellet since the pellet fragments tend to move towards the heat sink (cladding), and the heat flow remains predominantly radial despite the presence of cracks. It is important to understand the temperature profile on the pellet before and after cracking to improve cracking models in fuel performance codes However, in-reactor observation and measurement of cracking is very challenging owing to the harsh environment and design of fuel rods. Recently, an experimental pellet cracking test stand was developed for separate effects testing of normal operations and accident temperature conditions, using thermal imaging to capture the pellet surface temperature for evaluation of thermal stresses and optical imaging to capture the evolution of cracking in real time. Cracking experiments were initially performed using ceria (CeO 2 ) as a surrogate fuel material, which is useful for developing and demonstrating the experimental approaches but is also valuable in its own right for cracking model development and validation. A combination of induction and resistance heating was used for volumetric heat generation in the pellet creating a thermal gradient. The material properties of CeO 2 and UO 2 are reviewed and compared for use in model development. Simulations of the experiment were performed to evaluate the behavior of the surrogate (CeO 2 ) fuel in BISON. The measured temperature profiles from BISON models match reasonably well with the observed experiments for the ceria pellets before cracking. The findings from this work will help improve confidence in fracture models used for fuel pellets under similar in-reactor conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sensitivity of UO 2 fuel performance to microstructural evolutions driven by dilute additives

Use of dilute additives to nuclear fuel is being considered to increase the security of commercial fuel management through traceability of fabricated fuel elements. Taggants, as additives are denoted when included for traceability purposes, may also improve fuel performance, as demonstrated in Cr-containing uranium dioxide as described in the literature, and they may also improve fuel safety. In fact, studies have shown that some additives affect fuel material properties such as grain size and density after sintering. Given the possible range of elements that could be used as additives, the impact of such fuel property variations on the fuel’s thermomechanical behavior becomes relevant. These effects can be evaluated through a sensitivity study of standard fuel models to analyze changes in these properties using a fuel performance code. In this work, the BISON code is being used to investigate these effects through a 2D axisymmetric model of smeared UO 2 fuel pellets and ZIRLO® cladding under realistic pressurized water reactor core irradiation conditions. Here, randomly sampled densities and grain sizes within specified ranges are used as input parameters in the simulations, and several fuel model-related outputs are evaluated. The thermomechanical response of the cladding is also addressed in this study. The simultaneous variation of both input parameters offers a more comprehensive path to identify key sensitivities. Outputs explored include temperature, fission gas release, creep, and radial stress. Results show that although most of these outputs are sensitive to grain size to a certain extent, density mainly affects fuel temperature and elastic strain. Furthermore, sensitivities can vary depending on the radial position within the fuel pellet.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident (RIA) commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (TREAT), marking the first such tests in the U.S. in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The capsule initial conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO 2 . The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. Furthermore, the results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Improved correlations of the fuel/cladding liquid penetration rate with the out-of-pile transient database

Here, utilizing data from the Out-of-Pile Transient Database (OPTD), we developed advanced correlations for the cladding liquid penetration rate in metallic fuels used in sodium-cooled fast reactors (SFRs). Based on the available OPTD data, our research highlights the key role of Pu concentration in influencing the liquid penetration behavior, showing different threshold fuel-cladding interaction temperatures, with burnup playing a substantial role, particularly in fuels with a 19% Pu concentration. The penetration rate exhibits a parabolic, diffusion-limited behavior, thereby facilitating predictions of maximum liquid penetration depth over time. These correlations are significantly more accurate, under a variety of conditions, compared to existing single-parameter models, and enhance our understanding of SFR metallic fuel behavior, promoting safer and more efficient reactor operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Failure investigation and mitigation after experimental research reactor fuel plate deformation in an irradiation device

Experimental research reactor fuel testing is conducted in the Belgian Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) in dedicated irradiation vehicles or rigs. One such vehicle allows flat full-size fuel plates to be irradiated by inserting them into slotted baskets that captures a narrow portion of the longitudinal edges of the plates. The motion of the fuel plates within the baskets is possible within the narrow slots and thus, the plate is considered to be unattached. The design intentionally omits fixing mechanisms of the fuel plates to the baskets to facilitate the inspection and repositioning of the plates between the irradiation cycles and the accommodation of thermal expansion of the plates in the lateral direction. However, loosely inserted fuel plates have weak structural boundary conditions allowing for larger out-of-plane deflections caused by hydrodynamic loads exerted by the flowing coolant, as compared to those of fixed plates. Unexpected large deformations of plates occurred in several irradiation cycles that further resulted in a loss of cladding integrity. These deformations could not be attributed to a single source. This triggered a series of thermal hydraulic, structural, and fluid-structure interaction analyses aiming at understanding the observed phenomenon. The analyses revealed that, for a certain combination of unfavorable manufacturing and assembly tolerances, fuel plate edges could escape out of the slots in the irradiation basket due to the hydrodynamic load. Subsequently, the plate could become wedged inside the basket coolant channel opening. This resulted in reduced coolant flow and accelerated temperature increase and thermal expansion of the plate while under irradiation. This unfavorable feedback loop could then lead to excessive plate surface temperatures, deformed plates and cladding failure, as was observed in the experiments. These analyses not only provided a probable cause of the fuel plate failures, but also resulted in a new and improved design of the irradiation basket to avoid these issues in the future. In conclusion, a series of recent successful irradiations confirm that the sources of failures were identified correctly, and the implemented mitigations were adequate.

BR2↗
DOE OSTI2018

Lanthanide migration and immobilization in metallic fuels

It is recognized that the lanthanide fission products can enhance the fuel-cladding chemical interaction (FCCI), which is a key concern of using metallic fuels such as U-Zr in a sodium-cooled fast reactor. The present work conducts a critical review on the analysis of lanthanide functions in FCCIs. Available in-pile and out-of-pile data are first collected and analyzed, then the theories for lanthanide migration and redistribution in the fuel in operation are analyzed. For mitigating FCCIs, one of the effective method is applying fuel additives for immobilizing lanthanides. The review investigates four candidates of fuel additives (Pd, Sn, Sb and In), and it is concluded that Sb can be the best candidate among the four elements based on current available thermodynamic data and microstructure characterizations. Considering that FCCIs lead to formation of metallic alloys/compounds between lanthanides and steel cladding constituents (Fe, Cr, and Ni), the review also analyzes the thermodynamic data such as enthalpy of formation of the alloys and/or metallic compounds and identify the solidus temperature of each alloy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

An approach to grid-to-rod fretting wear modeling using dynamic mortar contact

This article presents the development of a grid-to-rod fretting (GTRF) modeling methodology that utilizes dynamic mortar contact. Here, we leverage a recently developed computational framework for modeling thermomechanical contact in dynamic simulations from the nuclear fuel performance code Bison. Usage of mortar contact ensures a smooth distribution of normal and frictional contact forces, displacements, and velocities on the contact interface, thus facilitating an accurate computation of wear. The integration within an advanced nuclear fuel performance code enables analysis of complex interactions between the fuel, cladding, and spacer grid. Such interactions include swelling, creep, fuel fragmentation, burnup, plenum pressure, and fission gas release. Our methodology is demonstrated via the modeling of two spacer grid geometries and the influence of fuel-cladding mechanical contact on the generation of fretting wear.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Additively manufactured strain sensors for in-pile applications

Accurate, real-time monitoring of strain in fuel, cladding, and structural components of nuclear reactors is critical to better understand radiation induced phenomena during reactor tests and operations. The data provided is crucial to verify physics-based, multiscale modeling and simulation efforts, which aim to shorten the timeline for the development of new nuclear materials. Resistive strain gauges have limited performance during in-pile experiments due to the harsh operating conditions and limited physical space between fuel and cladding components. In this work, aerosol jet printing using silver nanoparticle inks was used to fabricate interdigitated electrode capacitive strain gauges on aluminum alloy 6061 tensile specimens. Here to simulate the temperatures of a traditional light water reactor, the capacitive strain gauges were tested with a mechanical test frame up to 300 °C and compared to commercially available bondable resistive strain gauges. The printed capacitive strain gauges exhibited a gauge factor of 1.0 and showed higher reproducibility and predictability of strain sensing performance than the resistive strain gauge. The results demonstrate the potential of aerosol jet printing to fabricate strain sensors with predictable performance and reduced invasiveness for high-temperature applications with confined spacing.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Effect of Al 2 O 3 coating on fretting wear performance of Zr alloy

Fretting wear caused by flow-induced vibration (FIV) is one of the predominant tribological failure modes in nuclear power plants and industry. This paper presents an experimental study of the effect on fretting performance of alumina (Al 2 O 3 ) coatings produced by atomic layer deposition (ALD) on Zirlo (a Zr alloy), a common nuclear fuel cladding alloy. A coating with uniform thickness of about 750 nm was deposited onto a Zirlo flat. Because the typical thickness of fuel cladding is 1 mm, an ultra-thin coating is required. The hardness and elastic modulus of the ALD coating were measured by nano-indentation techniques. With a pin-on-flat contact configuration in low-amplitude reciprocating contact, fretting performance was measured for uncoated Zirlo, Al2O 3 -coated Zirlo, and Al 2 O 3 -coated Zirlo followed by heat treatment. Results showed that the as-deposited Al 2 O 3 coating provided slight improvement in fretting wear performance. However, the heat-treated coating showed two orders of magnitude reduction in fretting wear of the Zr alloy. Wear in the uncoated material involves some local plastic deformation and debris accumulation at the contact interface. In the Al 2 O 3 -coated material, wear occurred by transfer of transparent Al 2 O 3 oxide into the steel pin counterface. Localized removal of coating from the Zirlo alloy flat resulted in a “shark whale” pattern in a scanning electron microscopy image. There was minimal debris accumulation with the heat-treated sample.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In Situ High-Temperature Synchrotron Diffraction Studies of (Fe,Cr,Al) 3 O 4 Spinels

The modeling of a loss-of-coolant-accident scenario involving nuclear fuels with FeCrAl cladding materials in consideration to replace a Zircaloy requires knowledge of the thermodynamics of oxidized structures. At temperatures higher than 1500 degrees C, oxidation of FeCrAl alloys forms (Fe,Cr,Al) 3 O 4 spinels. In situ high-energy X-ray diffraction in a conical nozzle levitator installed at beamline 6-ID-D of the APS was used to study the structural evolution of the oxides as a function of the temperature. Single-phase (spinel) and multiphase (spinel-corundum-FeAlO 3 ) regions are mapped as a function of the temperature for three different compositions of FeCrAl oxidation products. The thermal expansion coefficients and cation distribution in the spinel structure have been refined. The temperature at which complete melting of the fuel cladding is expected has been determined by the liquidus temperatures of the oxidized products to be between 1657 and 1834 degrees C in a 20% O 2 /Ar atmosphere using the cooling trace method. The liquidus temperature increases with increasing Al and Cr content in the spinel phase.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Unique [Mn 6 Bi 5 ] - Nanowires in KMn 6 Bi 5 : A Quasi-One-Dimensional Antiferromagnetic Metal

In this paper, we report a new quasi-one-dimensional compound KMn 6 Bi 5 composed of parallel nanowires crystallizing in a monoclinic space group C2/m with a = 22.994(2) Å, b = 4.6128(3) Å, c = 13.3830(13) Å and β = 124.578(6)°. The nanowires are infinite [Mn 6 Bi 5 ] - columns each of which is composed of a nanotube of Bi atoms acting as the cladding with a nanorod of Mn atoms located in the central axis of the nanotubes. The nanorods of Mn atoms inside the Bi cladding are stabilized by Mn–Mn bonding and are defined by distorted Mn-centered cluster icosahedra of Mn 13 sharing their vertices along the b axis. The [Mn 6 Bi 5 ] - nanowires are linked with weak internanowire Bi–Bi bonds and charge balanced with K + ions. The [Mn 6 Bi 5 ] - nanowires were directly imaged by high-resolution transmission electron microscopy and scanning transmission electron microscopy. Magnetic susceptibility studies show one-dimensional characteristics with an antiferromagnetic transition at ~75 K and a small average effective magnetic moment (1.56 μ B /Mn for H ∥ b and 1.37 μ B /Mn for H ⊥ b) of Mn from Curie–Weiss fits above 150 K. Specific heat measurements reveal an electronic specific heat coefficient γ of 6.5(2) mJ K –2 (mol-Mn) -1 and a small magnetic entropy change ΔS mag ≈ 1.6 J K –1 (mol-Mn) -1 across the antiferromagnetic transition. Finally, in contrast to a metallic resistivity along the column, the resistivity perpendicular to the column shows a change from a semiconducting behavior at high temperatures to a metallic one at low temperatures, indicating an incoherent-to-coherent crossover of the intercolumn tunneling of electrons.

36 MATERIALS SCIENCE↗