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At least 289 records · Page 16

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DOE OSTI2018

Results of Ring Compression Tests: Spent Fuel and Waste Disposition

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ductility↗

DOE OSTI2016

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

36 MATERIALS SCIENCE↗

DOE OSTI2020

Stability of zirconium hydride

This reports complete Milestone M2.2.1 Stability of zirconium hydride: Experimental investigation of the hydrogen concentration of encapsulated zirconium hydride before and after joining. In this milestone, we first report the successful fabrication of crack-free bulk zirconium hydride by using the ORNL static hydriding system. The fabricated hydride was characterized by using X-ray diffraction (XRD) and X-ray computed tomography (XCT). XRD results showed that the δ-phase zirconium hydride, α-zirconium, and ε-phase hydride were present. XCT confirmed the absence of the internal cracking. The hydrogen concentration was determined by the weight change before and after hydrogen loading, given the fact that the oxide phase was not found during XRD analysis. Two metallic cladding materials, i.e., molybdenum and FeCrAl, were selected to encapsulate the fabricated zirconium hydride pellets. Demonstration capsules were designed and machined. Mo crucible was successfully sealed by using a combination of electron beam welding and gas tungsten arc welding. Due to the undesired neutronics property of Mo, further investigation of this cladding material was not pursued. Electron beam welding and laser beam welding were employed to seal the FeCrAl capsules. XRD were performed on the zirconium hydride before and after the sealing. The results showed that the welding process has negligible impact on the hydrogen content in the FeCrAl-canned zirconium hydride. The presented work fulfilled the proposed work under this milestone and paves a pathway to fabricate single-phase zirconium hydride and to develop robust cladding materials to can bulk zirconium hydride.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Design and analysis of oxidation tests to inform FeCrAl ATF severe accident models

36 MATERIALS SCIENCE↗

DOE OSTI2017

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2018

Simulation of QUENCH-15 and Preliminary Pre-Test Predictions for QUENCH-19

Iron Chromium Aluminum alloys (FeCrAl) have been developed at Oak Ridge National Laboratory as an advanced fuel cladding for light water reactors (LWR). In contrast to zirconium-based alloys employed today, FeCrAl has slower oxidation kinetics and higher strength. A test (QUENCH-19) at Karlsruhe Institute of Technology’s QUENCH facility is planned to test and demonstrate the performance of FeCrAl cladding under postulated accident conditions. This test will replicate the previous QUENCH-15 test but will use FeCrAl instead of zirconium-based materials. During these tests, a small bundle of fuel rod simulators is electrically heated in a steam and argon environment. The temperature history, oxidation thickness, and in-situ hydrogen generation are the primary measurements of interest.The CORA/BWR code was previously developed and utilized in the 1980’s and 1990’s to simulate BWR bundle degradation during the CORA test series. A modified version of the CORA/BWR code, named CORA/BWR-QUENCH, has been developed to model the QUENCH test facility. The CORA/BWR-QUENCH code was used to model the QUENCH-15 test which used zirconium-based materials. The code predictions and test data were found to be in good agreement. The code was then used to model the planned QUENCH-19 test utilizing FeCrAl cladding, shroud, and spacer grids. The simulation predicts the peak cladding temperature to be 1620 K (~1347°C) for QUENCH-19 which is significantly lower than that during QUENCH-15 with the zirconium-based alloy. The predicted hydrogen generation is also significantly lower for the FeCrAl clad QUENCH-19 as compared to QUENCH-15: ~1.7 g versus ~48.8 g. These results align with expectations given the much slower oxidation kinetics of FeCrAl alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Draft of M2 Report on Integration of the Hybrid Hydride Model into INL's MBM Framework for Review

This report documents the development, demonstration and validation of a mesoscale, microstructural evolution model for simulation of zirconium hydride {delta}-ZrH{sub 1.5} precipitation in the cladding of used nuclear fuels that may occur during long-term dry storage. While the Zr-based claddings are manufactured free of any hydrogen, they absorb hydrogen during service, in the reactor by a process commonly termed ‘hydrogen pick-up’. The precipitation and growth of zirconium hydrides during dry storage is one of the most likely fuel rod integrity failure mechanisms either by embrittlement or delayed hydride cracking of the cladding. While the phenomenon is well documented and identified as a potential key failure mechanism during long-term dry storage (NUREG/CR-7116), the ability to actually predict the formation of hydrides is poor. The model being documented in this work is a computational capability for the prediction of hydride formation in different claddings of used nuclear fuels. This work supports the Used Fuel Disposition Research and Development Campaign in assessing the structural engineering performance of the cladding during and after long-term dry storage. This document demonstrates a basic hydride precipitation model that is built on a recently developed hybrid Potts-phase field model that combines elements of Potts-Monte Carlo and the phase-field models. The model capabilities are demonstrated along with the incorporation of the starting microstructure, thermodynamics of the Zr-H system and the hydride formation mechanism.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Assessment of Cladding-Cladding Delamination Failures in U-Mo Fuel Plates

Monolithic fuel is a candidate fuel form for the conversion of high-performance research reactors. This plate-type fuel consists of a high-density, U-Mo alloy-based fuel in a monolithic form that is sandwiched between zirconium diffusion barriers and encapsulated in an aluminum cladding. To qualify this fuel system, the program must demonstrate that the fuel meets the safety standards, has a predictable behavior, and performs well in a reactor. Although many plates have shown satisfactory thermo-mechanical performance in reactor, a small number of experimental plates exhibited failures, rendering performance of these plates unacceptable. Post-irradiation examinations of failed plates have revealed that various failures modes are present, including pillowing, blistering, buckling, warping, delamination, fracture, rupture, oxide spallation, etc. Previous studies have implied that such various operational and loading conditions could generate a complex stress-strain state, facilitating distinct failure modes that could impact the structural integrity. To understand such various failure modes, a comprehensive set of studies focusing on failure modes is being performed. As a part of these studies, this work studies the cladding-cladding separation failures that were observed in some of previous irradiation tests. To understand the root cause of cladding separation failures, selected plates from previous irradiation experiments were simulated under comparable operational conditions. Stress-strain state of the plates at the interface were examined. The results were comparatively evaluated to understand possible effects of stress state of the bond-line on delamination failures. This paper presents these findings, discusses such failure modes, and examines the influence of various factors on previously observed cladding delamination failures.

42 ENGINEERING↗

DOE OSTI2017

Evaluation of a Zirconium Recycle Scrubber System

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Electrorefining of Zircaloy-4 in LiCl-KCl to Decontaminate Irradiated Cladding - 16238

Electrorefining on unirradiated Zircaloy-4 specimen in LiCl-KCl was performed by managing the anode potential to identify whether high purity zirconium can be recovered in metallic form by electrorefining. Four different concentrations of ZrCl{sub 4} (0.1, 0.5, 1.0 and 2.0 wt%) were tested. Results indicate that high purity metallic zirconium was recovered at low concentration of ZrCl{sub 4} but ZrCl could be recovered at high concentration. The electrorefining experiments were simulated by a three-dimensional computational electrorefining model. Several physical properties were tested, and a set of physical properties was determined in the condition where chemical forms of cathode deposits estimated by the computational model accorded with the experimental results. Using those properties, co-deposition of zirconium and uranium was simulated in the geometry of a preliminary designed electro-refiner for irradiated Zircaloy claddings. Operating conditions that can lead to the formation of zirconium metal without ZrCl and uranium were calculated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Mechanistic model for stresses in the oxide layer formed on zirconium alloys

Zirconium alloys are widely used as the cladding material in pressurized-water-reactors. The oxide formed is subjected to compressive stresses, which relax over time. This may affect the protectiveness of the oxide layer by allowing crack formation. We present a mechanistic model to predict these stresses as a function of temperature and oxidation kinetics. Material parameters for elastic deformation, creep, and thermal expansion are taken from appropriate experimental studies and the resulting predictions for the evolution of the stress distributions are compared with other experimental data. Lastly, dislocation glide in the oxide is found to be the dominant mechanism of stress relaxation for temperatures below 900 K.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

Accident Tolerant Fuels↗

DOE OSTI2019

First-principles calculations of solute transport in zirconium: Vacancy-mediated diffusion with metastable states and interstitial diffusion

Zirconium alloys are the most widely used nuclear fuel cladding materials for light water power reactors where irradiation damage causes solute redistribution, leading to degradation of alloy properties such as corrosion resistance. Designing radiation-tolerant zirconium alloys requires a thorough understanding of the atomic-scale transport behavior of the alloying elements in Zr. In this work, we perform density function theory calculations to investigate the diffusion of Sn, Cr, Fe, Be, Al, and Ni in the hexagonal close-packed (HCP) Zr matrix. We develop a methodology to accurately model the metastable vacancy states along the basal migration path, known to occur in group IV metals. We compute the vacancy-mediated solute diffusion coefficients and drag ratios using the kinetic Monte Carlo method and an analytic Green's function method—the agreement between the two validates our methodology. The computed diffusion coefficients of Sn and Al show good agreement with the experimental data and we expect these solutes to diffuse via the vacancy-mediated mechanism. We use a Green's function approach, parameterized with data from density functional theory calculations, to compute the interstitial diffusion coefficients of Cr, Fe, Be, and Ni in the HCP Zr lattice. The computed diffusion coefficients of Cr, Ni, and Be agree with the experimental measurements within one order of magnitude, while those of Fe are within two orders of magnitude of the experimental measurements. The drag ratios for Cr, Fe, Be, and Ni are positive up to 1235 K, which suggests that nonequilibrium vacancy fluxes could drag these solutes toward sinks such as dislocation loops and grain boundaries. We also compute the transport coefficients without including the metastable states, and using the eight- and thirteen-frequency model. Our results show significant differences in drag ratio for the eight- and thirteen-frequency model predictions compared with the Green's function methodology, but smaller errors in the solute diffusivity. Combining interstitial and vacancy-mediated diffusivities, we predict the unusual result that increased vacancy concentration slows down solute diffusivity, while a sufficiently high vacancy concentration can change the dominant mechanism to an accelerated vacancy-mediated diffusion.

36 MATERIALS SCIENCE↗

DOE OSTI2016

High Temperature Fuel Cladding Chemical Interactions Between Unirradiated TRIGA Fuels and 304 Stainless Steel

High temperature fuel cladding chemical interactions (FCCI) between Training-Research, Isotopes-General-Atomics (TRIGA) fuel elements and their 304 stainless steel (304SS) cladding are of interest to develop an understanding of the fuel-behavior during transient reactor scenarios. TRIGA fuels are composed of uranium particles dispersed in a zirconium hydride (Zr H) matrix encased in a 304SS cladding. At high temperatures, the fuel and cladding have the potential experience FCCI. A large number of interactions can take place in this multi-component fuel system. Emphasis is given to the development of low melting point phases. Diffusion couples of TRIGA fuel vs. 304SS were annealed at 1000°C for 5 and 24 hours. Characterization was then carried out via scanning electron microscopy and transmission electron microscopy. The as-received and annealed microstructures of the fuel and the interactions regions between the fuel and the 304SS will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

Here, we present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ~15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Development of Innovative Non-Concentrated Alloy (NCA) Fuel Cladding for Advanced Nuclear Reactors

As the nuclear energy sector advances toward next-generation reactors, the need for high-performance fuel cladding materials has become increasingly urgent. Traditional alloys like zirconium and stainless steel are reaching their performance limits under higher temperatures, more corrosive coolants, and extended irradiation. This report presents the development of a new class of fuel cladding materials based on Non-Concentrated Alloys (NCAs) ? multi-element systems designed to deliver enhanced mechanical strength, corrosion resistance, and radiation tolerance. Through a combination of computational modeling (Computer Coupling of Phase Diagrams and Thermochemistry: CALPHAD), simulation-guided alloy selection, and experimental fabrication via arc melting and spark plasma sintering (SPS), three strategic alloy design paths were explored: (1) FeCrAl-based NCAs, (2) refractory-lean neutron-efficient alloys, and (3) equimolar high-entropy compositions. Microstructural analysis confirmed the formation of stable body-centered cubic BCC_A2 phases, while mechanical testing demonstrated hardness values significantly exceeding those of conventional cladding materials. The results highlight the tunability of NCA systems and their potential for balancing strength and ductility ? a critical consideration for in-reactor performance. Looking forward, future work will focus on thermomechanical optimization, CALPHAD refinement, and benchmarking against industry standards to enable scalable deployment. This work not only advances the science of nuclear materials but also supports broader goals in nuclear safety, performance, and nuclear energy innovation.

36 - MATERIALS SCIENCE↗