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At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Hydriding, Oxidation, and Ductility Evaluation of Cr-Coated Zircaloy-4 Tubing

Accident-tolerant fuel concepts have been developed recently in diverse research programs. Recent research has shown clear advantages of Cr-coated Zr cladding over bare cladding tubes regarding oxidation behavior under the design basis loss-of-coolant accident condition. However, limited data are available about the hydriding behavior of the Cr coating. For that purpose, Cr-coated Zricaloy-4 tubes were tested to investigate the effects of hydriding, oxidation, and postquench ductility behavior on coated Zr cladding. A high-power impulse magnetron sputtering (HiPIMS) process was used to produce a high-density coating on the Zircaloy-4 tube surface. Coated and uncoated Zircaloy-4 tube specimens underwent one-sided hydriding in a tube furnace filled with pure hydrogen gas at 425 °C. The tubing specimen ends were sealed with Swagelok plugs before the hydriding runs. For uncoated specimens, H analysis of the hydrided specimens indicated that the H content increased as the test time and initial pressure increased. However, almost no change was observed for the coated specimens that were hydrided under the same test conditions. After one-sided hydriding, the hydrided coated and uncoated specimens were exposed to steam at high temperatures for two-sided oxidation studies to simulate accident conditions. The coated specimens showed a slower oxidation: oxygen pickup was 50% lower than the uncoated specimens tested under the same conditions. Ring compression testing was performed to evaluate the embrittlement behavior of the Cr-coated specimens after hydriding and oxidation. The results indicated that the HiPIMS coating provides excellent protection from hydriding and oxidation at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closer of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Overview of a Versatile Loading System for Anisotropic Material Property Characterization

Additive manufacturing, cold rolling, and other thermomechanical treatments on metals, especially HCP, can cause texture and oriented grain structures resulting in anisotropic mechanical properties. This can lead to macroscopic response that significantly deviates from isotropic assumptions which motivates studying mechanical properties in multiple directions. In addition, multiaxial loading and complex stress states overlapping the above mentioned material anisotropy can lead to unexpected outcomes. Idaho National Laboratory developed an advanced mechanical testing system to study a range of uniaxial to multiaxial stress states while measuring the anisotropic response bringing insight into the overlap between material properties and stress states. High-temperature capability and stress- or strain-controlled loading is available to enable a variety of experiment types and conditions to measure elastic, plastic, and viscoplastic properties. This poster presents the design and capabilities of this system with preliminary results highlighting the benefits of multiaxial loading and anisotropic analysis in an integrated system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closure of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Pellet cladding mechanical interaction as a potential failure mechanism during a control rod drop accident in a boiling water reactor

Boiling water reactors (BWRs) represent approximately one-third of the operating fleet in the United States, contributing significantly towards the global effort in reducing carbon emissions. Even though most of the operating fleet has been in operation for quite some time, continued advancements in new nuclear fuel (such as accident tolerant fuel) or operating regimes (such as power up-rates and higher burnup operation) necessitates similar advancements in modeling and simulation capabilities. Bison, a component of the Virtual Environment for Reactor Applications (VERA), is a high-fidelity fuel performance code able to explore the fuel performance of a wide variety of fuel types in one-, two-, and three-dimensions. Until recently, the code had not been used for analyses of BWRs. Modeling capabilities have been added for Gd-bearing UO{sub 2} and pure zirconium liners. New models have been added based upon the U.S. Nuclear Regulatory Commission (NRC) guide-lines for hydrogen pickup in Zircaloy-2 claddings and pellet-clad mechanical interaction (PCMI) failure during a reactivity insertion accident (RIA), known as a rod drop accident (CRDA) in BWRs. Implementing and/or improving Bison modeling capabilities extended its analytical reach to areas beyond its original intended purpose. This paper demonstrates one of these capabilities as a proof of concept. Recently developed models enable Bison to provide an alternate approach for cladding integrity determination in CRDA evaluations, which currently use bounding conservative estimates. As part of this demonstration, NRC guidance on hydrogen-pickup and PCMI failure models were utilized in this research. Even though more research is needed in establishing right inputs and process in this area, this paper demonstrates Bison's ability to determine cladding integrity in a CRDA evaluation. This first of a kind demonstration is a proof of concept in this area, which could potentially be extended to a number of other areas where a more accurate cladding integrity determination would be needed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Specifications of FIPD Fission Gas Release Data

All fission gas release data stored in the Fuels Irradiation & Physics Database (FIPD) was originally measured using the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF). It is therefore called GASR data in FIPD. During the measurement of a sample, such as an irradiated EBR-II fuel element/capsule, a pinhole-sized region near the top of the element plenum was melted by a laser. Plenum gas then expanded into a calibrated volume (note: in this document, “sample” and “capsule/element” are used interchangeably consistent with GASR documents in FIPD). The pressure rise in the volume was recorded. Helium backfilling and expansion was then performed to determine the sample (e.g., fuel element plenum) volume using Boyle’s Law and assuming ideal gas behavior at constant temperature. With the plenum volume and the recorded pressure rise, the sample (e.g., fuel element plenum) pressure was derived with assumption of ideal gas law behavior. The plenum volume and pressure as well as the cladding temperature during the measurement were collected (GASR data in FIPD). Other records associated with the fission gas release data include: raw GASR data records including volumes and post-puncture pressures of seal head/sealing head and manifold, calibration data, backfilling gas pressure data, and the data analysis records. A sample(s) of the fission gas released from the plenum was collected by the GASR system into sample bottles. The chemical and isotopic composition of the gas sample was analyzed separately from GASR data, and will be discussed in a separate specification. The plenum volume, pressure, and cladding temperature during the measurement are typically utilized to determine the number of moles of gas in the plenum. This quantity is often compared to the number of moles of gas generated by fission events. However, calculating these values and their associated uncertainty is beyond the scope of this document, as it necessitates additional assumptions. The most important document to understand the FIPD fission gas data is the GASR operational manual (title: Gas Assay, Sample and Recharge System (GASR) operation and maintenance manual, HFEF/N OMM 4381, DOC. NO. W0018-0032-ES-00). This manual provides: (1) description of the GASR and the functions of each component (laser drilling, welding, seal head/sealing head, manifold, vacuum system, sample system, purge and gas tag system, etc.); (2) step-by-step guidance on calibrations, operations, and measurements; and (3) maintenance procedures and other details relating to the structure and operation of the GASR. Note that the original GASR operated until 2020. A new GASR with the same design and measurement methodology was installed in 2021. The specifications of the GASR presented on the HFEF website at this time are consistent with the ones given in the operational manual. The methods to calculate the plenum volume and pressures were not included in the operational manual, but were recorded in the legacy data analysis files. Details of the methods are given in Chapter 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

FY21 burst activities with coated Zircaloy-4 under accident conditions

This report summarizes the results of Advanced Fuels Campaign (AFC) accident-tolerant fuel (ATF) burst activities. Nuclear service grade Zry-4 was procured and coated with a 7-micron thick Cr coating. The coating quality was investigated, and there were several defects at the Cr/Zry-4 interface due to the surface roughness of the as received Zry-4 tubing. To provide insight into the effect of coating defects on the cladding performance under accident scenarios, the unirradiated uncoated and coated material was tested under loss of coolant accident (LOCA) and pellet cladding mechanical interaction (PCMI) reactivity insertion accident (RIA) conditions. The defected coating appeared to have no impact on the cladding performance under these scenarios when compared to the as-received cladding material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Tensile properties of the neutron irradiated HT-9 ACO-3 duct

HT-9, a ferritic-Martensitic steel, has attracted significant interest as a structural and cladding material for a variety of high dose nuclear component applications. This study examines the tensile properties of a HT-9 fuel assembly duct from the fast flux test facility, FFTF. The duct, from the Advanced Core Oxide-3 (ACO-3) irradiation test, was subjected to fast neutron irradiation exposure ranging from 3 dpa above the core top to nearly 155 dpa near the core mid-plane, spanning a temperature range from 382 °C at the lower end to 504 °C at the top. Due to the combination of thermal and neutron flux gradients along the core height. This yields a variety of irradiation damage processes which depend on core location. Tensile properties of specimens extracted at different positions along the height of the irradiated duct were measured at 25 °C, 200 °C and near their in-core irradiation temperature. The results show temperature has a greater impact than dose on post-irradiation strength and ductility. Two irradiation temperature regimes were determined, below irradiation temperatures of 420 °C irradiation hardening increases as irradiation temperature decreases, and above that temperature relatively minor changes are observed in tensile properties. These results are discussed in terms of expected microstructural changes, indicating that dislocation content is a strong contributor in these findings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Long-Length SiC/SiC Tube Irradiation Experiment at LWR Temperatures Inserted into HFIR

This report describes the pre-irradiation characterization and successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient at representative light-water reactor (LWR) temperatures of approximately 300 °C. The over-arching goal of this work is to validate thermomechanical models of SiC/SiC composite fuel assembly distortion used to predict the extent of lateral bowing and dose at which bowing reaches a maximum value. Ultimately, these models are used to ensure that fuel assembly distortion does not block coolant channels, interfere with control rod/blade movements, and/or impact local reactivity. The experiment contains six cladding tube specimens that are approximately 558.5 mm long with representative pressurized water reactor (PWR) outer and inner diameters of 9.5 mm and 7.5 mm, respectively. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along the outer surfaces of the specimen and mapped using analytic photography and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. A light curtain optical micrometer was also employed to measure the pre-irradiation bowing of each specimen. The experiment was successfully assembled and inserted into position VXF-19 for HFIR cycle 516, which began May 12, 2026. The experiment will be removed after cycle 519, which is currently scheduled to end in December 2026.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments

This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling & Simulation (M&S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. Further, the blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A LOCA Analysis Tool: Coupling RELAP5-3D to BISON

Experimental evidence illustrates that at burnups slightly above the current regulatory limit of a rod-averaged burnup of 62 MWd/kgU, the ceramic UO 2 inside light-water reactor fuel rods becomes susceptible to a phenomenon known as fuel fragmentation, relocation, and dispersal (FFRD) during a loss of coolant accident (LOCA) transient. The severity of FFRD is strongly influenced by the zirconium-based (Zircaloy) cladding behavior during the LOCA transient. A Technology Commercialization Fund (TCF) project was awarded to an Electric Power Research Institute (EPRI)/Idaho National Laboratory team to create a LOCA analysis tool that couples BISON to the systems/thermal-hydraulics code RELAP5-3D [1] for analysis of LOCA scenarios. In addition, further refinements to existing BISON models were identified as necessary to more accurately represent more recent experimental evidence from the Studsvik Cladding Integrity Project (SCIP) and other experimental programs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Decoding Early Candidacy of High Entropy Alloys for Nuclear Application using the Advanced Test Reactor through Predictive Methods and Combinatorial Testing

High Entropy Alloys (HEAs) are identified candidates for nuclear applications owing to their superb mechanical and thermal properties. In line with evaluating their candidacy, a challenge remains in their validity as structural replacements for extreme environments. Each compositional graded specimen consists up to four different compositions spanning the fueled zone for non-prototypical neutron irradiation testing in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). We have fabricated a set of four compositionally graded specimens using spark plasma sintering that include at least five of the following elements: Cr, C, Al, Zr, Mo, Nb, Ta, V, Ti, W, and Fe. In time for this meeting the role of temperature expected to play a role in-pile cladding chemical interactions (FCCI), mechanical interactions (FCMI), irradiation damage, creep, and resistance will be reported. The irradiated portion has a scheduled irradiation date in April 2020 and with post irradiation examination (PIE) to follow.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

MURR LEU structural and thermal hydraulics analyses: Part I – Preliminary irradiation thermo-mechanical behavior

The University of Missouri Research Reactor (MURR) is expected to be converted from highly enriched uranium (HEU, ≥ 20 wt% U-235) U-Al x dispersion fuel to low-enriched uranium (LEU, < 20 wt% U-235) with U-10Mo monolithic fuel. This work introduces high-fidelity irradiation thermo–mechanical (T-M) analysis of the MURR LEU focusing on changes in coolant channel gap thickness. Three-dimensional (3D) finite element (FE) models were developed to simulate the irradiation T-M behavior of the MURR LEU element with all 23 curved fuel plates, the two side plates, and the combs. It was shown that channel gap thickness changes were influenced not only by plate thickness variations due to fuel swelling and creep but also by the radial displacement of consecutive MURR LEU plates. Modeling the fuel element assembly captured side plate displacements, which were shown to reduce radial fuel plate displacements towards the convex side. The maximum local radial displacement in the element was predicted at the end of life (EOL) as 23.7 mil (602.0 µm) on the lateral centerline of plate 23 towards the convex side. The maximum stripe-averaged reduction in channel gap thickness, particularly relevant for thermal hydraulics (TH) safety analysis, was calculated as 15.9 mil (403.9 µm) in single-side heated channel 24 (the outermost channel). These results account for the thermal resistance from the oxide build-up on cladding surfaces which was shown to be up to 0.82 mil (20.8 µm) thick. It was demonstrated that accounting for oxide layer thermal resistance led to a 10 °C higher peak fuel temperature and a 4.4 mil (111.8 µm) greater maximum local radial displacement. The impact of the calculated channel gap thickness changes on the MURR LEU TH safety analysis is evaluated in Part II.

Irradiation thermo-mechanical analysis↗
DOE OSTI2025

Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

Here, a novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.

Bismuth↗
DOE OSTI2022

Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility

Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2015

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

This paper reviews test data for cermet fuel samples developed in the 1960's to better quantify Nuclear Thermal Propulsion (NTP) cermet engine performance, and to better understand contemporary fuel testing results. Over 200 cermet (W-UO2) samples were tested by thermally cycling to 2500 deg (2770 K) in hydrogen. The data indicates two issues at high temperatures: the vaporization rate of UO2 and the chemical stability of UO2. The data show that cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance, while other approaches yield smaller, incremental improvements. Data is very limited above 2770 K, and this complicates predictions of engine performance at high Isp. The paper considers how this material performance data translates into engine performance. In particular, the location of maximum temperature within the fuel element and the effect of heat deposition rate are examined.

Ceramic Nuclear Fuel↗
DOE OSTI2026

Full-core high-burnup BWR LOCA fuel performance analysis and FFRD susceptibility

The susceptibility of the boiling water reactor (BWR) Limerick Unit 1 to fuel fragmentation, relocation, and dispersal during a postulated large-break loss-of-coolant accident (LBLOCA) was calculated using a multiphysics framework. The simulations include full-core, rod-resolved neutronic, thermal hydraulic, and fuel performance models using the VERA, TRACE, and BISON codes. This work focused on the transient BISON simulations, which include both the normal operation and LBLOCA periods in the same simulations. Cladding integrity was assessed using two correlations that are included with BISON. make page break Several new BWR-specific features were recently added to BISON. This work represents the first time these features have been included in a core-scale set of simulations. This study hence evaluates the performance of these new models for an operating reactor with realistic operating conditions. Simulation results showed that cladding integrity was maintained (i.e., no rods burst). Finally, future work to improve BWR and PWR predictions using this framework is suggested.

BISON↗
NASA NTRS1980

Potentialities of TEC topping: A simplified view of parametric effects

An examination of the benefits of thermionic-energy-conversion (TEC)-topped power plants and methods of increasing conversion efficiency are discussed. Reductions in the cost of TEC modules yield direct decreases in the cost of electricity (COE) from TEC-topped central station power plants. Simplified COE, overall-efficiency charts presented illustrate this trend. Additional capital-cost diminution results from designing more compact furnaces with considerably increased heat transfer rates allowable and desirable for high temperature TEC and heat pipes. Such improvements can evolve of the protection from hot corrosion and slag as well as the thermal expansion compatibilities offered by silicon-carbide clads on TEC-heating surfaces. Greater efficiencies and far fewer modules are possible with high-temperature, high-power-density TEC: This decreases capital and fuel costs much more and substantially increases electric power outputs for fixed fuel inputs. In addition to more electricity, less pollution, and lower costs, TEC topping used directly in coal-combustion products contributes balance-of-payment gains.

Morris, J. F.↗