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DOE OSTI2021

Material jet printing of transparent ceramic Yb:YAG planar waveguides

A new, to the best of our knowledge, 3D additive manufacturing technique utilizing particle-loaded ink jet printing to fabricate transparent ceramic Yb:YAG planar waveguides for laser gain media was demonstrated. Rheological optimization of YAG particle-loaded inks resulted in successful droplet formation and printing resolution. Planar waveguides composed of a Yb:YAG guide encased in undoped YAG cladding were printed with guide thicknesses ranging between 25 and 350 µm and consolidated to high optical quality via solid-state sintering. Sufficiently low optical (1–3%/cm) and intermodal scatter allowed single-mode propagation with a core/clad index difference of Δ n ∼ 5.0 × 10 − 4 (corresponding to 3 at.% Yb). The waveguides were cladding-pumped longitudinally with a 940 nm diode array resulting in 23.6% slope efficiency in 2 ms pulsed operation.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Ultra-high extinction ratio polarization beam splitter with extreme skin-depth waveguide

In this Letter, we present a high extinction ratio and compact on-chip polarization beam splitter (PBS), based on an extreme skin-depth (eskid) waveguide. Subwavelength-scale gratings form an effectively anisotropic metamaterial cladding and introduce a large birefringence. The anisotropic dielectric perturbation of the metamaterial cladding suppresses the TE polarization extinction via exceptional coupling, while the large birefringence efficiently cross-couples the TM mode, thus reducing the coupling length. We demonstrated the eskid-PBS on a silicon-on-insulator platform and achieved an ultra-high extinction ratio PBS ( ≈ <#comment/> 60 d B for TE and ≈ <#comment/> 48 d B for TM) with a compact coupling length ( ≈ <#comment/> 14.5 µ <#comment/> m ). The insertion loss is also negligible ( < <#comment/> 0.6 d B ). The bandwidth is > <#comment/> 80 (30) nm for the TE (TM) extinction ratio > <#comment/> 20 d B . Our ultra-high extinction ratio PBS is crucial in implementing efficient polarization diversity circuits, especially where a high degree of polarization distinguishability is necessary, such as photonic quantum information processing.

Ahmed, Syed Z.↗

DOE OSTI2026

All-ceramic channel waveguides fabricated via 3D printing

All-ceramic channel waveguides (CWGs) in Yb:YAG transparent ceramics have been fabricated for the first time, to the best of our knowledge, via direct ink write (DIW) and their laser performance has been demonstrated. Single filaments of Yb:YAG nanoparticle-loaded ink were extruded into undoped YAG; the Yb:YAG filaments formed the CWGs, surrounded by undoped cladding. Elemental mapping confirmed the Yb doping profile and waveguide integrity. Optical characterization showed low cladding scatter losses (<1.3%/cm at 1.3 µm), and laser testing with a 940 nm Ti:sapphire pump demonstrated efficient lasing at 1030 nm. The best-performing waveguide, with an elliptical cross-section (100 µm × 60 µm and a length of 1.4 cm), achieved a slope efficiency of 61% and a roundtrip loss of 12.4%. These results identify DIW as a promising approach for fabricating high-performance channel waveguides in transparent ceramics.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Fabrication of engineered dopant profiles in Er/Lu:YAG transparent laser ceramics via additive manufacturing

Transparent ceramic Er:YAG laser rods were fabricated via the direct ink write (DIW) method with engineered doping profiles featuring an Er-doped core with endcaps and core-clad structures. Laser rods up to 11 cm in length were produced which required development of a scalable process. To achieve this, multiple improvements were implemented, including printing the rods horizontally on a substrate, rather than vertically, eliminating the need for an external support structure and using a sacrificial drying layer to mitigate warping and defects. Highly transparent rods were achieved with optical scatter levels as low as 0.5%/cm (at 543 nm). A small refractive index difference of 5.7 ppm was measured at the interface between the Er-doped core and the Lu-doped endcaps and cladding. These results demonstrate DIW as a straightforward method for making good optical quality laser rods with engineered doping profiles to improve laser performance.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Effect of alloying elements, cold work, and hydrogen on the irradiation-induced growth behavior of zirconium alloy variants - 2016-0040

In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components-especially at high exposures-cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multi-year (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2X10{sup 26} m{sup -2} (E>1 MeV)-equivalent to maximum of approximately 37 dpa exposure-followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320±10 deg. C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM). All irradiation samples (typically flat rectangular coupons or curvilinear cutouts of cladding tubes sized 35 by 6.5 by 0.8 mm) were pre-filmed to avoid the uptake of impurity hydrogen from sodium-cooled BOR-60. A wide variety of samples representing standard LWR cladding alloys with and without pre-hydriding (approximately 116 to approximately 718 ppm) as well as special compositions with iron contents (100 to 4,000 ppm) were irradiated. The irradiation in BOR-60 was done in five different stages (eight micro-cycles) and lasted approximately 18 months with interim and final growth measurements made using a high precision-length measurement device. Results of the extensive investigation include: significant effects of Fe, Nb, and hydrogen additions; quantification of growth rates from low to very high fluences (dpas); measurement of volume changes; and correlation of growth with -component dislocation densities. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2022

Analyzing the effect of pressure on the properties of point defects in $\gamma\mathrm{U–Mo}$ through atomistic simulations

Uranium–molybdenum (U–Mo) alloys in monolithic fuel foil are the primary candidate for the conversion of high-performance research reactors in the USA. Monolithic fuel is utilized in a plate-type design with a zirconium diffusion barrier and aluminum cladding. These fuel types are unique in that they contain no plenum for the release of fission gases, which, in conjunction with the aluminum cladding, can lead to large stress states within the fuel. The nature of how fundamental processes of radiation damage, including the evolution of point defects, under such stresses occur is unknown. In this work, we present molecular dynamics simulations of the formation energy of point defects under applied stress. We report this work will allow for the implementation of stress-dependent microstructural evolution models of nuclear fuels, including those for both fission gas bubble growth and creep, which are critical to ensure the stable and predictable behavior of research reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

The Development of Reversible Bending Fatigue Tester and Its Application to High Burnup Spent Nuclear Fuel Vibration Integrity Investigation under Normal Transportation

For high burn-up spent nuclear fuel (SNF), it is expected that the spent fuel cladding will have a high population of microcracks and hydrides, including macro-hydrides and micro-hydrides. This will reduce the stress intensity required to advance the crack growth. The linking of these microcracks during vibration loading also reduces the fatigue threshold/incubation period significantly, accelerating fatigue failure. In addition to the damages in cladding, the microstructure of comprising fuel pellets and interfaces of fuel rod have changed dramatically after the high burn-up process. These modifications have a direct impact on the structural integrity and vibration response of SNF rods in transportation.

Fuel Transport↗

DOE OSTI2014

Ion irradiation testing and characterization of FeCrAl candidate alloys

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Fast Spectrum Experiment Fuel Pin Payload Specification for Package 9330

This specification supports work to be conducted by AREVA Federal Services under Idaho National Laboratory (INL) Subcontract No. 45559, Release 50, including Statement of Work (SOW) -11417 addressing package transportation licensing. This document provides transportation payload specification for experiment fuel pins produced in the performance of INL CRADA 13-CR-15. The payload description provided in this document will be used to obtain licensing necessary to enable transportation of unirradiated experiments in ATR Fresh Fuel Shipping Containers (FFSCs). The FFSC has a current U. S. Department of Energy (DOE) certificate, USA/9330/AF-96, revision 0, which is available on the DOE RAMPAC website or upon request from the INL or DOE. Its desired payload will consist of individual experiment fuel pins (a.k.a. rods) loaded in a FFSC’s internal fuel handling enclosure (FHE), a.k.a. internal canister. Each package and its FHE arrangements should be prepared for a payload that contains less than critical mass and is certified for ground and air transportation per NRC 10 CFR 71 and International Atomic Energy Agency (IAEA) criteria. The unirradiated experiment fuel pins are all fabricated with uranium-based fuel and ferritic-martensitic stainless steel, HT-9, cladding and end plugs. The payload may include pins with and without spacer-wires. Solid pins will have minimal amounts of sodium bonding between the fuel and inside cladding. Annular pins will contain He gas installed at atmospheric pressure (~12.7 psia). Uranium fuel will be of varying enrichments with a maximum bounding enrichment of 94% U-235. Blanket material consists of depleted uranium and zirconium of varying weight percent. Section 2 provides pin parameters for NRC 10CFR71 and IAEA evaluations. Supplemental criteria titled functional and operating requirements is intended to convey other requirements, including those that protect the payload, rather than to demonstrate package compliance with regulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

RERTR-12 Post-irradiation Examination Summary Report

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 on four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal property measurements conducted in FY 2014 and compares results to values obtained from literature and measurements performed in FY 2013, where applicable, along with appropriate discussion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Assessment of MARMOT: A Mesoscale Fuel Performance Code

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Training package 1 for slitting data analysis

This document and accompanying files are intended as a first training package on how to analyze slitting data. The end goal is to have Idaho National Laboratory (INL) personnel trained to analyze future slitting data taken in the INL Hot Cell on clad, Low-Enriched Uranium (LEU) fuel plates. This first data package will cover data analysis for a monolithic material (as compared to a layered material like the clad fuel plates). The additional issues for layered specimens will be covered in a future training package.

42 ENGINEERING↗

DOE OSTI2015

Formed HIP Can Processing

The intent of this report is to document a procedure used at LANL for HIP bonding aluminum cladding to U-10Mo fuel foils using a formed HIP can for the Domestic Reactor Conversion program in the NNSA Office of Material, Management and Minimization, and provide some details that may not have been published elsewhere. The HIP process is based on the procedures that have been used to develop the formed HIP can process, including the baseline process developed at Idaho National Laboratory (INL). The HIP bonding cladding process development is summarized in the listed references. Further iterations with Babcock & Wilcox (B&W) to refine the process to meet production and facility requirements is expected.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Measure Guideline: Deep Energy Enclosure Retrofit for Double-Stud Walls

This Measure Guideline describes a deep energy enclosure retrofit solution that provides insulation to the interior of the wall assembly with the use of a double-stud wall. The guide describes two approaches to retrofitting the existing walls—one that involves replacing the existing cladding and the other that leaves the cladding in place. This guideline also covers the design principles related to the use of various insulation types and provides strategies and procedures for implementing the double-stud wall retrofit. It also includes an evaluation of important moisture-related and indoor air quality measures that need to be implemented to achieve a durable high-performance wall.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2015

Status update of the BWR cask simulator

The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on heat load and the effect of simulated wind on a simplified below ground vent configuration.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗