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DOE OSTI2020

Remote Fiber Based Velocimeter For Interface Strength Measurements

The understanding of the performance of nuclear reactor plate fuels is essential when developing new fuel systems. The condition of pre and post irradiated plate fuels must be characterized effectively and efficiently prior to irradiation and then post irradiation within a high rad environment. The post characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization methods provide the capability to be remote and robust within a hot-cell environment. Laser based characterization can provide high spatial resolution appropriate for scanning and imaging large areas. The INL is currently developing a laser shock system for Post Irradiation Examination (PIE) station for use in the Hot Fuel Examination Facility (HFEF) at the INL. The laser shock system is designed to characterize fuel to cladding and cladding to cladding interface strength. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad interface. The shock wave propagates as a compressional wave through the fuel plate to the free (unconfined) backside surface and is reflected back through the test plate as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal delamination failure. The focus or this paper is on the accurate measurement of surface velocity to quantify the amount of energy traveling through the interfaces to the back surface of the fuel plate. The original fiber based velocimeter design was biased by parasitic back reflections. This paper will discuss the discovery of the back-reflection issue and the design changes to resolve the issue. Supporting data will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1973

Evaluation of a metal shear web selectively reinforced with filamentary composites for space shuttle application. Phase 3 Summary report: Shear web component testing and analysis

Three large scale advanced composite shear web components were tested and analyzed to evaluate application of the design concept to a space shuttle orbiter thrust structure. The shear web design concept consisted of a titanium-clad + or - 45 deg boron/epoxy web laminate stiffened with vertical boron/epoxy reinforced aluminum stiffeners. The design concept was evaluated to be efficient and practical for the application that was studied. Because of the effects of buckling deflections, a requirement is identified for shear buckling resistant design to maximize the efficiency of highly-loaded advanced composite shear webs. An approximate analysis of prebuckling deflections is presented and computer-aided design results, which consider prebuckling deformations, indicate that the design concept offers a theoretical weight saving of 31 percent relative to all metal construction. Recommendations are made for design concept options and analytical methods that are appropriate for production hardware.

Laakso, J. H.↗

NASA NTRS1972

Examination of UC-ZrC after long term irradiation at thermionic temperature

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

Yang, L.↗

DOE OSTI2017

Thermo-mechanical performance assessment of selected plates from MP-1 high power experiments

Monolithic plate-type fuel is a fuel form that is being developed for the conversion of high performance research and test reactors to low-enrichment uranium fuels. These fuel-plates are comprised of a high density, low enrichment, U-Mo alloy based fuel foil encapsulated in an aluminum cladding. To benchmark this new design, number of plates has been irradiated with satisfactory performance. As a part of continuing evaluation efforts, a set of plates covering range of operational parameters is scheduled to be tested during MP-1 irradiation experiments. It is necessary to evaluate the thermo-mechanical performance of plates during irradiation. For this, selected plates with distinct operational histories; covering low power, high power and high fission density were simulated. Fully coupled three-dimensional models of plates with a capability to evolve mechanical and thermal properties of constituent materials with irradiation time and burn-up were developed. The models input used projected parameters, including plate geometry, irradiation history and coolant conditions as input. The model output included temperature, displacement and stresses in the fuel, cladding and diffusion barrier. The fuel behavioral model considered inelastic behavior including volumetric swelling due to solid and gaseous products, irradiation induced creep, thermal expansion, conductivity degradation and plasticity. A visco-plastic behavioral model was used for the cladding that included thermal creep, irradiation hardening, growth due to fast neutrons and Mises plasticity. The plates were then simulated by using projected irradiation parameters. The resulting temperature, displacement and stress-strains were comparatively evaluated on the selected paths. The results were then compared with those of plates from previous RERTR experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Thermo-mechanical performance assessment of selected plates from MP-1 low power experiments

42 ENGINEERING↗

DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration

Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding) with solute additions of Mo, Ti, and Nb has been developed to provide the desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850°C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 10 23 m -3 ) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. In conclusion, the ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications.

36 MATERIALS SCIENCE↗

NASA NTRS1973

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Siegel, B. L.↗

DOE OSTI2021

Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment

New fuel design and development currently requires 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance experiments to help reduce the timeline associated with fuel design and development. This study uses the Dynamical System Scaling (DSS) method to analyze cladding temperature data from the recent SETH-C experiment in the TREAT facility and accompanying BISON simulations to assess dynamic distortions occurring throughout the fast power excursion transient. The DSS analysis revealed that on the cool down from peak cladding temperature that the fuel radial power profile is the most sensitive modeling parameter with a heterogenous radial peaking factor corresponding to the lowest distortion compared to a uniform energy deposition. For the heat up to peak cladding temperature the heterogeneous radial power profile corresponded to the shortest process action. Finally, for the heat up to peak cladding temperature, the gap conductance model sensitivity was quantified using process action and shows that the default Light Water Reactor gap conductance model corresponded to the longest process action.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300°C to 400°C

The creep rupture of high burnup used nuclear fuel (> 45 GWD/MTU) cladding is regarded as one of the failure mechanisms during long-term dry storage. A high amount of zirconium hydride in the cladding matrix would degrade the mechanical properties of the cladding especially leading to delayed hydride cracking. To better understand the influence of zirconium hydride on the biaxial thermal creep performance and the crystalline texture stability of Zircaloy-4 cladding, the pressurized tube technique is applied to test the durability of the as-hydrided material. Tests were performed on as-received Zircaloy-4 tubular specimens as well as-hydrided ones with targeted 300 wt parts per million (wppm) or 750 wppm hydrogen. The biaxial creep experiments were conducted at temperatures from 300 degrees C to 400 degrees C and at equivalent stresses at mid-wall from 55 MPa to 65 MPa. The hydridation process prior to creep tests induces the formation of FCC delta-hydride platelets along the circumferential direction of the tube. This alignment and phase structure of hydride show no significant change after biaxial creep tests. The creep strain-rate negatively depends on the hydrogen content. The synchrotron wide-angle X-ray diffraction (WAXD) technique and electron backscatter diffraction (EBSD) analysis were applied for the study of the crystallographic orientation relationship. zirconium-to-hydride grain orientation follows Shoji-Nishiyama crystallographic relationship. This relationship is stable before and after creep deformation. These results of creep performance and texture stability of Zircaloy-4 claddings can help support the design basis of interim and long-term dry storage facilities.

creep↗

DOE OSTI2017

Al x In 1-x P LEDs with II-VI Cladding Layers for Efficient Red and Amber Emission. Final Scientific & Technical Report

Realizing solid state lighting systems with high efficiency and enhanced functionality will require the adoption of color-mixing (cm-LEDs) architectures combining highly efficient red, green, blue and amber light emitting diodes. Already, the hybrid architectures (hy-LEDs) pair phosphor-converted wide bandgap nitride LEDs with (Al x Ga 1-x ) 0.51 In 0.49 P red LEDs to improve the emission spectrum and reduce losses in the infrared. However, the efficiencies of (Al x Ga 1-x ) 0.51 In 0.49 P red and amber LEDs are still too low to truly enable these approaches. The DOE Solid-State Lighting R&D Plan highlights red and amber LED improvement as a long-term research goal. Excessive electron leakage out of the cladding layers, in particular, leads to strong thermal droop and red LED efficiencies of ~25% at normal operating temperatures. The typical device architecture for red and amber LEDs is based on indirect bandgap (Al y Ga 1-y ) 0.51 In 0.49 P cladding layers used to confine injected electrons and holes within a lower, direct bandgap (Al x Ga 1-x ) 0.51 In 0.49 P light-emitting active layer (x < y). The challenge with this approach is that the electron confinement is limited and decreases as the emission wavelength is shortened. Hence, thermal droop and overall device efficiency are very poor for amber LEDs. Innovative solutions to this problem must be sought through the use of new semiconductor materials and structures that will increase electron confinement. We plan to develop Al x In 1-x P LEDs with a II-VI semiconductor electron cladding layer to greatly increase the carrier confinement. Such an approach has the potential to improve red and amber LED efficiency, temperature stability, and light extraction. Al x In 1-x P active layers provide high direct bandgap energies needed to reach amber wavelengths without detrimental internal losses. ZnSe-based alloys are attractive candidates for the electron cladding layer because they can be engineered to provide high conduction band energies relative to the Al x In 1-x P active layer while maintaining roughly the same lattice constant. The lower index of refraction of the II-VI cladding layer will also aid light extraction. While extended defects have largely precluded the use of II-VI semiconductors as the light emitting layer in LEDs and lasers, the effects of these defects on device performance can be mitigated by limiting the use of II-VI materials to the thin, heavily doped cladding layer. Proposed research will focus on optimizing the quality of the Al x In 1-x P active layer and the heteropolar growth of the III-V/II-VI interface, tuning the II-VI cladding layer composition, and device design to maximize electron confinement and LED efficiency. Significant improvement in the performance of red and amber LEDs through this new approach is expected to advance the capabilities of next generation solid-state lighting technologies.

36 MATERIALS SCIENCE↗

NASA NTRS1968

Single-mode-guide laser components.

Clad glass optical waveguides for design of lasers, modulators, couplers, detectors and filters

Schineller, E. R.↗

NASA NTRS1978

Temperature stabilization of microwave ferrite devices

Thin-film heating element for strip-line circulator is sandwiched between insulation and copper laminations. Disks conform to shape of circulator ferrite disks and are installed between copper-clad epoxy ground planes. Heater design eliminates external cartridges and reduces weight by approximately one-third.

Kaminsky, R.↗

DOE OSTI2016

Calibration and Validation of NRC Fuel Performance Models

Pacific Northwest National Laboratory maintains steady-state and transient fuel performance codes for the U.S Nuclear Regulatory Commission (NRC). The NRC uses these codes in performing confirmatory calculations of licensee fuel performance codes and safety analysis methods during the review and approval of these codes and methods. The NRC also uses these codes to perform calculations to direct regulatory research and to assist in the development of new regulations and regulatory guides. Because of the safety-related uses for these codes, it is critically important that the codes are capable of accurately modeling modern fuel and operations and that the models and code predictions are well validated using data collected from conditions relevant to in-reactor conditions. FRAPCON and FRAPTRAN have an extensive database of separate effects data used to validate individual material property and fuel performance models and an integral effects database used to assess the overall code prediction within the full range of fuel operations. The key code predictions that are of regulatory interest as they relate to various specified acceptable design limits are; fuel and cladding temperature, rod internal pressure, cladding hoop strain, cladding corrosion and hydriding, and fission gas release. To assess the ability of the code to accurately predict these values, the integral assessment database consists of measurements of fuel centerline temperature, fission gas release, void volume, cladding corrosion, and cladding hoop strain. This paper will discuss the activities that are performed to ensure that FRAPCON and FRAPTRAN represent the state of the industry and will describe the separate effect and integral effect databases that are used to validate the models and are continually being expanded to ensure validation over the full range of fuel operations.

Geelhood, Kenneth J.↗

DOE OSTI2017

Removal of Tritium from the Gaseous Effluents Produced by the Recovery of Zirconium from Used Fuel Cladding

U.S. regulations will require the removal of tritium ({sup 3}H) from gaseous effluents associated with nuclear fuel reprocessing and recycling. One process that may generate {sup 3}H-containing gaseous effluents is the recovery of zirconium from used fuel cladding. A chlorination process developed by Collins et al. is designed to convert the zirconium present in Zircaloy{sup R} cladding to recoverable ZrCl{sub 4}. The greatest economic benefit from including zirconium recycle as part of a larger used nuclear fuel (UNF) recycle effort would be a decrease in the disposal cost of used fuel cladding. Used fuel cladding is considered high-level waste (HLW). As zirconium is the second largest element by mass (25%) contained in most UNF assemblies (with uranium being the largest), recovery of the zirconium could result in a substantial decrease in the amount of HLW intended for geologic disposition. The reclaimed ZrCl{sub 4} product, if sufficiently free from radioactive impurities, could then be disposed of as low-level waste or reused in nuclear applications. The chlorination process uses chlorine gas (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium present in cladding to ZrCl{sub 4}. ZrCl{sub 4} is more volatile than other components in the cladding, causing evaporation of a relatively pure gaseous ZrCl{sub 4} stream. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. A fraction of the {sup 3}H (0-96%) produced in nuclear fuel during irradiation is found in zirconium-based cladding, possibly as zirconium hydride (the wide range is a remnant of various researchers' analyses of cladding of various irradiations and cooling times). Heating the cladding to high temperatures, such as would be done during the chlorination process, will release the {sup 3}H. At the temperatures of the zirconium recycle process, the {sup 3}H is likely to react with excess Cl{sub 2} to produce tritiated HCl ({sup 3}HCl), although this could be avoided if the cladding and fuel were treated to remove {sup 3}H prior to chlorination. A previous analysis showed that the concentration of {sup 3}HCl in the off-gas of the zirconium recycle process will be in the range of parts per billion to a few parts per million. Should {sup 3}H mitigation be required for the zirconium recycle process, it will be completed through the abatement of {sup 3}HCl from this off-gas stream. The intent of the testing described in this summary was to conduct proof-of-principle demonstrations for two {sup 3}HCl removal methods. The zirconium recycle off-gas stream differs from many industrial HCl-containing streams in that the concentration of HCl requiring removal is extremely low. Demonstrations of HCl removal at ppm concentrations were not found in the literature, making the proof-of-principle approach attractive as a way to identify the most promising method for continued development. An important consideration throughout these experiments was that current research efforts on the zirconium recycle process are conducted in a hot cell and any {sup 3}HCl removal methods should be easily deployed in a hot-cell environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

TEM Characterization of Nanostructured Grade 91 Neutron Irradiated to 6 DPA at 300 °C and 500 °C

Previously, samples of coarse grained (CG, grain diameter > 1 µm), ultra-fine grained (UFG, 1 µm> grain diameter> 100 nm), and nanocrystalline (NC, 100 nm> grain diameter) grade 91 steel were neutron irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Grade 91 is a ferritic/martensitic steel alloy with a nominal composition of Fe-9%Cr-1%Mo by weight, and it is being studied for applications in advanced nuclear reactor designs including as a fuel cladding material in gas cooled reactors. To assess the viability of this alloy for nuclear applications, the role of grain boundary size has not been sufficiently explored as a mechanism for developing radiation-tolerant materials. To achieve an UFG microstructure the material was processed using equal channel angular pressing (ECAP), and the NC microstructure was achieved through high pressure torsion (HPT). Such samples of various geometries for different types of characterization were irradiated to a dose of 6 displacements per atom (DPA) at separate temperature conditions of 300 °C and 500 °C. Thus, there were a total of 6 experimental conditions, three grain sizes (CG, UFG, and NC), each irradiated at two different temperature conditions (300 °C and 500 °C).

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Mechanical Property Assessment of Unirradiated Cladding After Exposure to Time at Temperature

Light water reactors in the United States are operated within inherent safety limits designed to account for anticipated operational occurrences (AOOs). Unlike more severe reactor transients, these events permit the potential reuse of reactor cladding, highlighting the nuanced role of thresholds in influencing reactor efficiency compared to design-basis or beyond-design-basis transient conditions. The threshold designated for AOO peak cladding and fuel temperatures are currently grounded around thermal hydraulic limits such as surpassing the critical heat flux (CHF), rather than in fuel-cladding material performance. This CHF criterion therefore represents a conservative approach to AOO transients during the reactor lifetime. Extending this threshold to higher temperatures than CHF can result in additional operation efficiency gains. Here, a material performance-based approach is assessed to better understand the physical material limits of short time scale exposures to elevated temperatures. This approach, termed time at temperature (t@T), references fuel cladding microstructural and mechanical response to rapid temperature transients in absence of radiation damage. The results of this work are expected to help identify margin for the United States reactor fleet, which could be utilized to increase operational efficiencies through decreased reactor outages.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

The Building Science Advisor: A Web-Based Tool to Assess the Durability of Building Envelope Components

The Department of Energy (DOE) and Oak Ridge National Laboratory (ORNL) have developed an innovative new design tool that will put building science expert advice in the hands of every building designer who cares to manage moisture risk in high-R envelope components. As modern buildings become increasingly more airtight, constructed with modern labor-saving materials, and equipped with air-conditioning, they have become less forgiving to moisture intrusion. The Building Science Advisor (BSA) provides building science knowledge and advice based on expert experience, field measurements, laboratory tests, and computer simulations.BSA users are prompted to enter relevant information about the building location, design, and material selection options like cladding, structural system, and insulation. Influential factors such as climate, building air tightness, material properties, and internal moisture loads are also considered to estimate and compare the moisture durability performance of several design options. If the BSA deems the assembly’s performance unsatisfactory, it will provide the reasons why, and suggest necessary changes to wall design to ensure more robust performance.Using the BSA tool enables building designers to confidently select assembly design characteristics that achieve their design goals with the least moisture durability risk. Links to design-specific guidance are also be provided to help users manage any remaining risk. This tool will further enable DOE’s Building Technologies Office (BTO) to meet its long-term energy goal of a 50% reduction in building energy consumption by reducing builder concern about using highly energy efficient wall systems.

Desjarlais, Andre Omer↗