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At least 1,351 records · Page 75
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

A mortar thermomechanical contact computational framework for nuclear fuel performance simulation

Nuclear fuel performance simulations involve the modeling of complex physical phenomena, ranging from fission gas release to fuel swelling and other temperature-induced effects. For light-water reactors (LWRs), swelling of the fuel and the pressure it imposes on the clad when they come into contact causes permanent clad deformation. Accurately characterizing the fuel-cladding interaction, which involves multiple physics, is essential to accurately simulate the fuel/cladding system. Thermomechanical modeling of this problem using a variationally consistent enforcement (e.g., a mortar approach) has been shown to improve the quality of results and facilitate convergence. Here, we present a general multiphysics computational framework for solving nuclear fuel problems using a mortar approach in BISON, a nuclear fuel performance code. In this study analyses show that using the mortar approach, which enables variationally consistent constraint enforcement, improves the quality of results as compared to the more commonly used node-on-face enforcement for representative LWR nuclear fuel simulations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Transient fuel performance analysis for the preliminary fuel concept of general atomics fast modular reactor

This study investigates the transient fuel performance of General Atomics Fast Modular Reactor (GA-FMR) during accident scenarios, focusing on the behavior of its innovative fuel system that combines high-assay low enriched uranium dioxide (HALEUO2) fuel with SiGA® ceramic matrix composite silicon carbide cladding. The preliminary fuel design’s response was analyzed during reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA) using BISON fuel performance analysis code, which included both the diffusion enhanced and BISON-FASTGRASS coupled UO 2 models. The RIA analysis demonstrated that effective reactivity control reduced fuel temperature, though with transient fission gas release resulting in additional tensile stress state on the cladding. LOCA simulations revealed differing predictions between the two models: the BISON UO 2 model showed more transient fission gas release but minimal pellet expansion, while the BISON-FASTGRASS UO 2 model predicted less pronounced fission gas release but more fuel swelling and thermal expansion, potentially leading to pellet-cladding mechanical interaction. Here, these findings highlight critical areas for fuel design optimization and identify knowledge gaps requiring further experimental and computational investigation to advance GA-FMR fuel development.

Lee, Soon K. [Argonne National Laboratory (ANL), A↗
DOE OSTI2024

Applying U.S. metal fuel experience to new fuel designs for fast reactors

With the increasing interest in small modular reactors or microreactors, developers are working to design and submit licensing approval requests of U–10Zr-fueled fast reactors. The developers and their proponents cite prior metal fuel experience (worldwide, but U.S. experience in particular for many developers) as the motivation and justification for their reactor concepts. The experience with metal fuel deployment in sodium-cooled fast reactors as well as the underlying irradiation testing database, provide a suitable basis for analytically justifying the use of metal fuel in new reactors. The evolution of metal fuel design and capability illustrates the importance of key fuel design parameters to consider in new applications of the prior experience: fuel smeared density, plenum-to-fuel volume ratio, the ratio of cladding radius to thickness, fuel composition, and cladding and duct materials. In-service operating and deployment conditions to be considered include fuel linear heat generation rate, fuel temperature, cladding temperature, peak burnup and peak fast fluence. Fuel designs and in-service conditions that are bounded by the database and experience are most easily addressed, but deviations from those previous parameters and conditions can be addressed by considering impacts on previously established behavior and applying other mitigating conservatisms, as appropriate. Here, the authors recommend any new deployment proceed with fuel surveillance and monitoring to mitigate risk, application of conservative measures to address uncertainties, and a fuel qualification program that addresses a range of in-service operating conditions with production fuel. The work reported should be of interest to students and regulators unfamiliar with metal fuel in fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Laser-based characterization of nuclear fuel plates

Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

The effect of Ta interface on the crystallization of amorphous phase change material thin films

The crystallization of amorphous GeTe and Ge{sub 2}Sb{sub 2}Te{sub 5} phase change material films, with thickness between 10 and 100 nm, sandwiched between either Ta or SiO{sub 2} layers, was investigated by optical reflectivity. Ta cladding layers were found to increase the crystallization temperature, even for films as thick as 100 nm. X-Ray diffraction investigations of crystallized GeTe films showed a very weak texture in Ta cladded films, in contrast with the strong texture observed for SiO{sub 2} cladding layers. This study shows that crystallization mechanism of phase change materials can be highly impacted by interface effects, even for relatively thick films.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2021

Guiding light via slippery liquid-infused porous surfaces

Slippery liquid-infused porous surfaces (SLIPSs) have been explored for many applications, taking advantage of their highly non-wetting property. In this work, we explore the SLIPS as a cladding material for waveguiding. SLIPSs are prepared by infusing perfluoropolyether oil to hydrophobized nanoporous surfaces of silicon. Power loss and transmission efficiency of an HeNe laser (1.82 mW and 632.8 nm) with varying incident angles were measured through microchannels consisting of the SLIPSs as cladding layers (noil = 1.30) and water (nwater = 1.33) as a core, compared to other cladding types including a planar silicon surface and the nanoporous surfaces in hydrophilic (Wenzel state) and hydrophobic (Cassie–Baxter state) conditions with no oil infused. Agreeing with Snell's law, a total internal reflection occurs at the incident angle as high as 14° for the SLIPSs. The waveguide loss at 14° is only 1.8 dB/cm for the SLIPSs, while those for planar silicon, hydrophilic nanoporous, and hydrophobic nanoporous surfaces are 5.9, 7.4, and 4.9 dB/cm, respectively. The power transmission efficiency of the SLIPSs is independent of the porosity because the surfaces are fully covered with the oil layer, whereas those of hydrophilic and hydrophobic nanoporous surfaces, whose pores are filled with water and air, respectively, depend on the porosity. The significantly lower power loss and the insensitivity to the surface porosity are advantages of the SLIPSs over the other surfaces and can benefit in waveguiding applications such as optofluidics.

Asawa, Kaustubh (ORCID:0000000197142555)↗
DOE OSTI2019

Overview of the SAS4A Metallic Fuel Models and Extended Analysis of a Postulated Severe Accident in the Prototype Gen-IV Sodium-Cooled Fast Reactor

The SAS4A safety analysis code, originally developed for the analysis of postulated severe accidents in oxide fuel sodium-cooled fast reactors (SFRs), has been significantly extended to allow the mechanistic analysis of severe accidents in metallic fuel SFRs. The SAS4A metallic fuel models simulate the metallic fuel thermomechanical and chemical behavior and track the evolution and relocation of multiple fuel and cladding components during the pretransient irradiation and during the postulated accident, allowing an accurate description of the changes in the local fuel composition. The local fuel composition determines the fuel thermophysical properties, such as freezing and melting temperatures, which in turn affect the fuel relocation behavior and ultimately the core reactivity and power history during the postulated accidents. Models describing the fuel-cladding interaction and eutectic formation, the effects of the in-pin sodium on the in-pin fuel relocation, and the postfailure reentry of the molten fuel and fission gas from the pin plenum have also been added. This paper provides an overview of the SAS4A key metallic fuel models emphasizing the postfailure metallic fuel relocation models included in the LEVITATE-M module of SAS4A. The capabilities of the SAS4A metallic fuel models are illustrated through an extended SAS4A analysis of a postulated unprotected loss-of-flow and transient-overpower accident in the metallic fuel prototype Gen-IV sodium fast reactor. The results show that the maximum relative power reached during the postulated accident is 1.19 P 0 . The favorable characteristics of the metallic fuel cause a significant decrease in net reactivity and relative power due to prefailure in-pin fuel relocation. Furthermore, negative net reactivity values persist after cladding failure, and the postfailure fuel relocation events occur at low and decreasing power levels.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Dynamic PRA Methods to Evaluate the Impact on Accident Progression of Accident Tolerant Fuels

Accident tolerant fuels (ATFs) are new nuclear fuels developed in response to the accident at the Fukushima power station in March 2011. The goal of ATFs is to withstand accident scenarios through better performance compared to currently employed fuels (e.g., small-scale hydrogen generation). This paper targets a method for evaluating and comparing ATF performance from a probabilistic risk assessment (PRA) perspective by employing a newly developed combination of event trees and dynamic PRA methods. Compared to classical PRA methods based on event trees and fault trees, dynamic PRA can evaluate with higher resolution the safety impacts of physics dynamics and the timing/sequencing of events on the accident progression without the need to introduce overly conservative modeling assumptions and success criteria. In this paper, we analyze the impact on the accident progression of three different cladding configurations for two initiating events [a large break loss-of-coolant accident (LB-LOCA) and a station blackout (SBO)] by employing dynamic PRA methods. The goal is to compare the safety performance of ATFs (FeCrAl and Cr-coated cladding) and the currently employed Zr-based clad fuel. We employ two different strategies. The first focuses on the identification of success criteria discrepancies between the accident sequences generated by the classical PRA model and the set of simulation runs generated by dynamic PRA using ATF. The second one, on the other hand, directly uses dynamic PRA to evaluate the impact of timing of events (e.g., recovery actions) on accident progression. In conclusion, by applying these methods to the LB-LOCA and SBO initiating events, we show how dynamic PRA methods can provide analysts with detailed and quantitative information on the safety impact of ATFs.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Advanced Autonomous Welding for Refabrication and Follow-On Testing of Previously Irradiated Nuclear Fuel

The performance of follow-on experiments using irradiated nuclear fuel at any point in its lifecycle is a critical step in understanding phenomena and behavior. Transient experiments with high-burnup fuel can deepen the understanding of fuel fragmentation, relocations, and dispersal under loss-of-coolant accidents. An advanced autonomous welding process to refabricate commercial fuel rods inside a hot cell was created and tested to enable flexible experiment approaches on fuels irradiated in commercial and test reactors. Irradiated light water reactor fuel test pins from experiments performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory were used to demonstrate the refabrication process. The welding process was found to be sensitive to welding parameters but flexible such that multiple passes could be performed on the same location until a hermetic weld was obtained. The refabrication of rodlets and successful welds was also found to be sensitive to the preparation of the irradiated cladding and endcaps. Thorough defueling of the fuel at the weld location and proper sizing of the endcaps and backing material mitigated these issues. The use of strategically located heat sinks in contact with the cladding and endcap materials also increased welding and refabrication success. For this work, the test pins were sectioned to remove the original endcaps and fuel was removed from both ends of each rodlet. The reassembly of the rodlets was then completed in four steps, which included the press fitting of new endcaps, the circumferential welding of rodlet endcaps to the cladding, rodlet pressurization in a pressure chamber, and seal welding the rodlet under pressure. The integrity of the refabricated rodlets was then verified via helium leak checking inside a vacuum chamber. The advanced welding system is capable of refabricating rodlets up to 380 mm in length, and repressurizing them up to 15 500 kPa. The refabricated lengths of the rodlets used in this work ranged from 149 to 165 mm and the refabricated fuel stack heights ranged from 70.4 to 79.8 mm. Here, the rodlets were pressurized with argon to an average pressure of 3617 kPa, and the average leak rate after refabrication was 6.7 • 10 -8 cm 3 • s -1 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

P2M Simulation Exercise on Past Fuel Melting Irradiation Experiments

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK·CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW·m -1 , average burnup 27 GWd·tU -1 ], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW·m -1 , average burnup 48 GWd·tU -1 ), carried out in 1987 in the BR2 reactor at SCK·CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Drying of Spent Nuclear Fuel: Considerations and Examples

A recently revised American Society for Testing and Materials consensus standard guide for drying of spent nuclear fuel (SNF) provides considerations and guidance for preparing SNF for its safe storage in a sealed dry storage system. The standard discusses (1) needs for drying, (2) techniques to dry, and (3) demonstration of adequate dryness. No specific approach is prescribed since the effective techniques and needs for drying depend on the specifics of the fuel and dry storage system. Here, this paper discusses these topics using examples for both Zr-alloy-clad commercial SNF and for aluminum-alloy-clad research reactor SNF. Residual water can include free water (liquid and/or vapor), physisorbed water bound to internal surfaces, and chemisorbed water incorporated into surface films, such as (oxy)hydroxides. The potential impacts of these residual waters are corrosion/oxidation, radiolytic breakdown into gaseous and/or reactive species, and canister pressurization. For commercial SNF, inadvertent free water, even up to large amounts (e.g., 10+ mol), is not predicted to cause adverse corrosion degradation, except possible fuel oxidation for breached SNF. For aluminum-clad SNF, the production of radiolytic hydrogen with contribution from the chemisorbed water in its hydrated oxides is a primary consideration. For both SNF systems, canister pressurization is predicted to be well within the canister design, and flammability would not pose a safety concern using an oxygen limit of 5 vol % criterion. (Flammability control can be achieved by limiting either hydrogen or oxygen, and an oxygen limit is expected to be easier to meet in the presence of radiolytic H 2 generation.) The two primary technologies for SNF drying, vacuum drying and forced-gas dehydration, are described herein, and drying tests and campaigns using these methods are cited. Dryness criteria and the methods used to detect and measure residual (free) water are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Thermal-Hydraulic-Mechanical Modeling and Simulation of Sodium-Potassium–Cooled MARVEL Microreactor Core

The U.S. Department of Energy's Microreactor Program, with Idaho National Laboratory's development of a nuclear microreactor applications test bed named MARVEL, aims to support R&D for the deployment of small, transportable reactors across civilian, industrial, and defense sectors. The MARVEL microreactor, an 85-kWth thermal fission reactor, incorporates TRIGA nuclear fuel and a sodium-potassium eutectic as its primary coolant, designed for safety and efficiency, with natural circulation eliminating the risk of critical heat flux conditions. The reliance on natural circulation for primary cooling means the reactor avoids using fuel spacers to minimize core pressure drop, which could disrupt the primary coolant's natural flow. However, the reactor core’s tight P/D ratio of 1.05, in the absence of fuel spacers, could pose a risk of fuel rod contact and increased peak cladding temperatures. To ensure the reactor safety, this study conducted computational modeling and simulations to investigate the reactor's thermal-hydraulic-mechanical characteristics, including the reactor core heat transfer coefficients, the potential for rod-to-rod contact, and assessed its impact on peak cladding temperature and overall reactor safety. The computational analyses of the MARVEL microreactor core revealed that the thermal deformation of fuel rods under worst-case scenario may lead to the fuel rod contact, but the peak cladding temperatures will remain significantly lower than the safety criteria, ensuring the safety operation reactor without fuel spacers under normal operating conditions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2015

The MaNGA integral field unit fiber feed system for the Sloan 2.5 m telescope

We describe the design, manufacture, and performance of bare-fiber integral field units (IFUs) for the SDSS-IV survey Mapping Nearby Galaxies at Apache Point Observatory (MaNGA) on the the Sloan 2.5 m telescope at Apache Point Observatory. MaNGA is a luminosity-selected integral-field spectroscopic survey of 10{sup 4} local galaxies covering 360–1030 nm at R∼2200. The IFUs have hexagonal dense packing of fibers with packing regularity of 3 μm (rms), and throughput of 96 ± 0.5% from 350 nm to 1 μm in the lab. Their sizes range from 19 to 127 fibers (3–7 hexagonal layers) using Polymicro FBP 120:132:150 μm core:clad:buffer fibers to reach a fill fraction of 56%. High throughput (and low focal-ratio degradation (FRD)) is achieved by maintaining the fiber cladding and buffer intact, ensuring excellent surface polish, and applying a multi-layer anti-reflection (AR) coating of the input and output surfaces. In operations on-sky, the IFUs show only an additional 2.3% FRD-related variability in throughput despite repeated mechanical stressing during plate plugging (however other losses are present). The IFUs achieve on-sky throughput 5% above the single-fiber feeds used in SDSS-III/BOSS, attributable to equivalent performance compared to single fibers and additional gains from the AR coating. The manufacturing process is geared toward mass-production of high-multiplex systems. The low-stress process involves a precision ferrule with a hexagonal inner shape designed to lead inserted fibers to settle in a dense hexagonal pattern. The ferrule ID is tapered at progressively shallower angles toward its tip and the final 2 mm are straight and only a few microns larger than necessary to hold the desired number of fibers. Our IFU manufacturing process scales easily to accommodate other fiber sizes and can produce IFUs with substantially larger fiber counts. To assure quality, automated testing in a simple and inexpensive system enables complete characterization of throughput and fiber metrology. Future applications include larger IFUs, higher fill factors with stripped buffer, de-cladding, and lenslet coupling.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2026

Atomistic modeling of lanthanide diffusion in refractory body-centered cubic molybdenum

Lanthanide fission products can strongly interact with candidate cladding alloys, but their transport properties in refractory metals remain poorly understood. Here, in this work, we investigate the atomic-scale diffusion behavior of La, Ce, Pr, and Nd in body-centered cubic (bcc) molybdenum, a potential candidate for advanced nuclear cladding. Self-consistent mean-field transport modeling is performed to evaluate the fission product transport and vacancy mobility, informed by first-principles and nudged elastic band calculations of vacancy formation energies, migration barriers, and solute–vacancy binding characteristics. Compared with bcc Fe, lanthanide solutes in bcc Mo exhibit slower tracer diffusion due to higher vacancy formation and migration energies. Furthermore, the calculations reveal that the influence of fission products on migration barriers in bcc Mo are not as extensive in range compared to bcc Fe. Among the studied lanthanides, La exhibits the strongest vacancy binding while also being the fastest diffuser in Mo. These findings highlight how refractory bcc alloys can reduce fission product infiltration, offering valuable insight into the development of durable cladding systems for advanced reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2015

Hydroforming of elliptical cavities

Activities of the past several years in developing the technique of forming seamless (weldless) cavity cells by hydroforming are summarized. An overview of the technique developed at DESY for the fabrication of single cells and multicells of the TESLA cavity shape is given and the major rf results are presented. The forming is performed by expanding a seamless tube with internal water pressure while simultaneously swaging it axially. Prior to the expansion the tube is necked at the iris area and at the ends. Tube radii and axial displacements are computer controlled during the forming process in accordance with results of finite element method simulations for necking and expansion using the experimentally obtained strain-stress relationship of tube material. In cooperation with industry different methods of niobium seamless tube production have been explored. The most appropriate and successful method is a combination of spinning or deep drawing with flow forming. Several single-cell niobium cavities of the 1.3 GHz TESLA shape were produced by hydroforming. They reached accelerating gradients E acc up to 35 MV/m after buffered chemical polishing (BCP) and up to 42 MV/m after electropolishing (EP). More recent work concentrated on fabrication and testing of multicell and nine-cell cavities. Several seamless two- and three-cell units were explored. Accelerating gradients E acc of 30–35 MV/m were measured after BCP and E acc up to 40 MV/m were reached after EP. Nine-cell niobium cavities combining three three-cell units were completed at the company E. Zanon. These cavities reached accelerating gradients of E acc = 30–35 MV/m. One cavity is successfully integrated in an XFEL cryomodule and is used in the operation of the FLASH linear accelerator at DESY. Additionally the fabrication of bimetallic single-cell and multicell NbCu cavities by hydroforming was successfully developed. Several NbCu clad single-cell and double-cell cavities of the TESLA shape have been fabricated. The clad seamless tubes were produced using hot bonding or explosive bonding and subsequent flow forming. The thicknesses of Nb and Cu layers in the tube wall are about 1 and 3 mm respectively. The rf performance of the best NbCu clad cavities is similar to that of bulk Nb cavities. The highest accelerating gradient achieved was 40 MV/m. The advantages and disadvantages of hydroformed cavities are discussed in this paper.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Small-angle neutron scattering measurements of δ-phase deuteride (hydride) precipitates in Zircaloy 4

Small-angle neutron scattering (SANS) measurements have been performed under ambient conditions to characterize deuteride (hydride) particles in Zircaloy 4, a fuel cladding material used in pressurized light-water nuclear reactors. Hydrogen pickup by the cladding leads to a rim structure in which large circumferential hydride plate-like particles preferentially form on the cooler water-side region of the cladding. Deuterium substitution has been used to increase the coherent response and decrease the incoherent background of the SANS measurements. Four bulk deuterium concentrations were investigated, approximately 100, 400, 500 and 1000 parts per million by weight (w.p.p.m.) deuterium, as well as a zero-deuterium-concentration reference sample. The net SANS response from the deuteride phase was determined at all concentration values after subtraction of the reference SANS response, which effectively subtracted the strong scattering from second-phase particles in as-received Zircaloy. The net SANS response consisted of strong Porod scattering from deuteride particles over the entire measured Q range (0.005–0.4 Å -1 ). The net SANS response was anisotropic at concentrations greater than 100 w.p.p.m. and required elliptical averaging analysis. A significant sample orientation effect on the intensity of the SANS response was observed, due to preferential alignment of deuteride particles. The effect of ex situ applied stress at elevated temperature on deuteride phase dissolution and reprecipitation was investigated; a weak effect was observed with SANS that could not be confirmed by optical microscopy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Thermal Stability of the Dot-in-Well Gain Medium for Photonic Crystal Surface Emitting Lasers

Self-assembled quantum dots (QDs) embedded in InGaAs quantum wells (QWs) are used as active regions for photonic-crystal surface-emitting lasers (PCSELs). An epitaxial regrowth method is developed to fabricate the dot-in-well (DWELL) PCSELs. Here, the epitaxial regrowth starts with the growth of a partial laser structure containing bottom cladding, waveguide, active region, and the photonic crystal (PC) layer. The PC layer is patterned to realize the cavity. Subsequently a top cladding layer is regrown to complete the laser structure. During the regrowth of the top cladding layer, the partial laser structure is subjected to high growth temperatures in excess of 600 °C resulting in an unintentional annealing of the active region. This annealing of the active region can alter the QDs by changing their size resulting in a blue shift in photoluminescence (PL) and narrowing PL emission. This effect results in the misaligning of the gain peak and the cavity resonance, resulting in sub-optimal lasing performance. DWELL active regions are known to have better thermal stability compared to both QDs and QWs and could be an ideal candidate for regrown PCSELs. We successfully demonstrate an optically-pumped epitaxially-regrown DWELL PCSEL with an emission wavelength of 1230 nm operating at room temperature. Furthermore, the DWELL active region shows excellent emission wavelength stability and intensity despite the high temperature regrowth process.

77 NANOSCIENCE AND NANOTECHNOLOGY↗