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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Evaluation of effective-stress-function algorithm for nuclear fuel simulation

In a pressurized water reactor (PWR), the mechanical integrity of nuclear fuel is the most critical issue as it is an important barrier for fission products released into the environment. The integrity of zirconium cladding that surrounds uranium oxide can be threatened during off-normal operation owing to a pellet-cladding mechanical interaction (PCMI). To analyze the fuel and cladding behavior during off-operation, the fuel performance code should calculate an inelastic analysis in two - or three-dimensional calculations. In this paper, the effective stress function (ESF) algorithm based on a two-dimensional FE module has been implemented to simulate the inelastic behavior of the cladding with stability and accuracy. The ESF algorithm solves the governing equations of the inelastic constitutive behavior by calculating the zero of the appropriate effective-stress-function. To verify the accuracy of the ESF algorithm for an inelastic analysis, a code-to-code benchmark was performed using the commercial FE code, ANSYS 13.0. To demonstrate the stability and convergence of the implemented algorithm, the number of iterations in the ESF algorithm was compared with that in a sequential algorithm in the case of an inelastic problem. Consequently, the evaluation results demonstrate that the implemented ESF algorithm improves the efficiency of the computation without a loss of accuracy for an inelastic analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Evaluating transient fission gas release in high burnup light water reactor fuel during loss of coolant accident conditions via new capabilities

In this work, the role of transient fission gas release (tFGR) in the cladding burst behavior of high burnup fuel during a loss-of-coolant accident (LOCA) in commercial light water reactors was further investigated via the use of a new apparatus. During the LOCA-related temperature ramps of high burnup fuel, the release of fission gases exceeds the steady-state release observed under normal operating conditions. An enhancement was made to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) to probe the various factors influencing tFGR. Experiments were performed on commercially irradiated, zirconium-clad uranium dioxide fuel, and this paper details the design of the experimental setup, the initial test results, and the subsequent post-test analyses. Notably, the first test on high burnup fuel demonstrated a LOCA-relevant tFGR of 5.3% from an unpressurized fuel segment. The ultimate tFGR was 10.7% for beyond LOCA conditions. A follow-up test on similar fuel revealed a tFGR of 12.6% under comparable conditions. Microstructural analysis and an analysis of the released gas provide some insight regarding the source of tFGR in the fuel. Finally, a grain boundary bubble model may aid in the interpretation of the results and offer a guide for future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Physical and mechanical metallurgy of zirconium alloys for nuclear applications: a multi-scale computational study

In the post-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry is going to continue using advanced zirconium cladding materials in the foreseeable future, it become critical to gain fundamental understanding of the several interconnected problems. First, what are the thermodynamic and kinetic factors affecting the oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings (if any) could be used in order to gain extremely valuable time at off-normal conditions, e.g., when temperature exceeds the critical value of 2200°F? Thirdly, the kinetics of oxidation of such protective coating or braiding needs to be quantified. Lastly, even if some degree of success is achieved along this path, it is absolutely critical to have automated inspection algorithms allowing identifying defects of cladding as soon as possible. This work strives to explore these interconnected factors from the most advanced computational perspective, utilizing such modern techniques as first-principles atomistic simulations, computational thermodynamics of materials, diffusion modeling, and the morphological algorithms of image processing for defect identification. Consequently, it consists of the four parts dealing with these four problem areas preceded by the introduction and formulation of the studied problems. In the 1st part an effort was made to employ computational thermodynamics and ab initio calculations to shed light upon the different stages of oxidation of ziraloys (2 and 4), the role of microstructure optimization in increasing their thermal stability, and the process of hydrogen pick-up, both in normal working conditions and in long-term storage. The 2nd part deals with the need to understand the influence and respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T’s). For that goal, a description of the advanced plasticity model is outlined featuring the non-associated flow rule in hcp materials including Zr. The 3rd part describes the kinetic theory of oxidation of the several materials considered to be perspective coating materials for Zr alloys: SiC and ZrSiO 4 . In the 4th part novel and advanced projectional algorithms for defect identification in zircaloy coatings are described. In so doing, the author capitalized on some 12 years of his applied industrial research in this area. Our conclusions and recommendations are presented in the 5th part of this work, along with the list of used literature and the scripts for atomistic, thermodynamic, kinetic, and morphological computations.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229 Th, 230 Th, 232 U, 233 U, and 234 U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Program Management Plan EnCore ® Accident Tolerant Fuel (ATF) Program

This Program Management Plan (PMP) is developed to support Westinghouse’s vision of developing and commercializing the Accident Tolerant Fuel. The near term goal of this Program is to deliver 2 fuel assemblies that consists of LTRs for Exelon Byron Unit 2 in spring of 2019. The longer term goal of this program is deployment of LTAs in 2022 of a mixture of coated zirconium cladding and SiC cladding with ADOPT and U3Si2 fuels. This is a high complexity program and is one of the largest research and development programs in Westinghouse and as such, the project scope is expected to change continuously. Integrated change control will be invoked as necessary. The Program consists of multiple funding streams (internal through Westinghouse innovation funding), external (through the DOE). The DOE funding consists of two phases: 1. Phase 2 – funded by DOE award number DE-NE0008222. 2. Phase 2B – funded by DOE – Westinghouse proposal reference # LTR-AMER-MKG-18-826, Rev. 1, August 23, 2018 (10/18 to 2/21), Award No. DE-NE0008824. 3. Cost share contributions from Westinghouse and program partners.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Towards Using Eshelby Calculations to Enhance Kinetic Model for Zirconium Hydride Precipitation

A C++ library (called Eshelby ) was implemented in fiscal year 2015 based upon the formulas documented in this report. The library implements a generalized version of Eshelby's inclusion problem. The library was written as a set of functions which can be called from another program; the principle intended use cases are kinetic models of precipitate formation in zirconium claddings where use of the Eshelby library provides needed elastic energy density calculations, as well as calculations of stress and strain in and around precipitates; it is intended that the library will be made open source. For isotropic inclusions in the form of oblate and prolate ellipsoids, the Eshelby library can be used for nearly any relevant/appropriate shape parameters to calculate strains, stresses and energy density at interior and exterior points. The Eshelby library uses a combination of analytical formulas and numerical routines making it very extensible. For example, the library can can easily be extended to include inclusions such as spheres since analytical expressions exist for the required elliptic integrals; similarly, general ellipsoids do not have analytical results for the required elliptic integrals but those integrals can be numerically evaluated and thus fit naturally into the Eshelby library. This report documents all formulas implemented in the Eshelby library and presents some demonstration calculations relevant to the intended application.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1977

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Clement, J. D.↗
DOE OSTI2016

Computational and Experimental Examination of Simulated Core Damage and Relocation Dynamics of a BWR Fuel Assembly - 16013

The Japan Atomic Energy Agency (JAEA) plans a large-scale test to evaluate damage and relocation behavior of Boiling Water Reactor (BWR) core materials consisting of fuel rods (zirconia as fuel simulant), channel boxes, control blade assemblies as well as lower core-support structures. Its purpose is to contribute to the understanding of the core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing anticipated at JAEA in the future, JAEA desires preliminary investigations to examine melting test pieces, being composed of a mixture of zirconia and metals by a non-transferred thermal plasma heating system. The purpose of such tests is to verify the materials and test piece will be heated by plasma as a primary heat source to the target temperature of ca. 2900 K and to collect data about the relocation behavior of the high-temperature materials. This paper describes the first in a series of tests demonstrating the effective use of thermal plasma technology to recreate the dynamics of a typical fuel core bundle channel assembly section. Results from preliminary computational simulations of simplified fuel bundle assemblies are presented illustrating the effectiveness of a 150 kW, non transferred (NTR) configuration of a thermal plasma argon jet to heat the assembly within a highly-insulated containment vessel. An experimental test program using the computational analyses as a basis and a Phoenix Solutions Co (PSC) commercial model PT150, 150 kW plasma torch is described and the test results are presented illustrating the time versus temperature history of the fuel bundle assembly melting dynamics. Details on plasma operation are discussed including: power requirements, current flow, voltage levels achieved, argon gas flow rates, processing times and standoff distances of the torch discharge from the top of the fuel bundle assembly (achieved with remote control of the axial (vertical) translation of the torch within the containment processing chamber). Photos taken from the top of the processing chamber are presented illustrating the various phases of melting characteristics of the zircaloy-clad, zirconium oxide filled fuel tubes, all within a stainless steel tube bundle channel centered within an insulated zirconia crucible. Based on lessons learned from this initial test campaign, a brief discussion of the requirements and features of a follow-on Phase II test program is provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage

Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO 2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 °C were terminated by quenching with water; transient tests were performed up to 1500–1600 °C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. Finally, the behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Alloying effect of niobium and tin on the zirconium alloy fuel cladding behavior at high temperature oxidation in steam - 2016-0072

As the materials for fuel claddings for water-cooled reactors, binary and multicomponent zirconium alloys zirconium-niobium (Zr-Nb) (E110, M5), zirconium-tin-iron (Zr-Sn-Fe) (Zircaloy-2,-4), and zirconium-niobium-tin-iron (Zr-Nb-Sn-Fe) (E635, ZIRLO{sup R} ) most commonly are used. Improvement extends, in particular, by varying their composition by niobium, tin, and iron. At the same time, importance is given to ensuring the safe operation of the fuel rods not only in normal conditions but also in emergency situations such as a loss of coolant accident. In this paper, we present the research results of studies of the influence of alloying elements niobium, tin, and iron on corrosion resistance and embrittlement at high-temperature steam oxidation of fuel claddings based on the alloy types Zr-xNb (x=1.0/2.5) and Zr-xNb-ySn-zFe (x=0.6/2.4; y=0.24/1.1; z=0.18/0.34) made by using zirconium sponge. We conducted high-temperature steam oxidation tests at temperatures of 1,000 deg. C, 1,100 deg. C, and 1,200 deg. C with continuous measurement of the weight gain during the experiment and subsequent cooling in steam with rate ∼ 20 deg. C/s. We conducted studies of kinetics of high-temperature oxidation in steam, structural-phase state changes, alloying elements distribution, absorbed hydrogen content, and residual ductility after rapid cooling of the oxidized specimens. We revealed the difference in the oxidation kinetics of the materials studied, which decreased as the oxidation temperature increased. Likewise, an increase of the tin content in the alloy influenced a specimen's breakaway oxidation intensification at 1,000 deg. C. This shows that the mechanical properties of fuel claddings oxidized under the same conditions depend on their alloying composition, and the hydrogen fraction is absorbed by the specimen and formed by the oxidation of ZrO{sub 2}, α-Zr(O), and ex-β layer structures, which in turn varies depending on the alloy composition. The increase of niobium, tin, and iron content in the zirconium alloy leads to a decrease in the residual ductility of fuel claddings after high-temperature steam oxidation. According to obtained results, among the materials studied, Zr-1.0Nb alloy is the most resistant to embrittlement during high-temperature steam oxidation at temperatures of 1,000-1,200 deg. C. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Mechanistic understanding of zirconium alloy fuel cladding Performance - 2016-0095

A review is presented of work performed in our group over the years in the areas of radiation damage, corrosion, hydrogen pickup, hydriding, and the mechanical behavior of zirconium alloy nuclear fuel cladding with the goal of developing a greater mechanistic understanding of cladding performance in service. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Alloying effect of Nb and Sn on the Zirconium alloy fuel claddings behavior at high temperature oxidation in steam - 6.1 STP-2016-0072

As the material of fuel claddings for water cooled reactors most commonly used binary and multicomponent zirconium alloys Zr-Nb (E110,M5), Zr-Sn-Fe (Zircaloy-2,-4) and Zr-Nb-Sn-Fe (E635, Zirlo) are used, which improvement extends, in particular, by varying their composition by Nb, Sn and Fe. At the same time, the importance is given to ensuring the safe operation of the fuel rods not only in normal conditions but also in emergency situations such as LOCA. In this paper the researches results of the studies of influence of alloying elements Nb, Sn (and Fe) on corrosion resistance and embrittlement at high temperature steam oxidation of fuel claddings based on alloys types Zr-xNb (x=1.0/2.5) and Zr-xNb-ySn-zFe (x=0.6/2.4; y=0.24/1.1; z=0.18/0.34) made with using zirconium sponge are presented. High temperature steam oxidation tests were carried out at temperatures of 1000, 1100 and 1200 deg. C with continuous measurement of the weight gain during the experiment and subsequent cooling in steam with rate ∼ 20 deg. C/s. The studies of kinetics of high-temperature oxidation in steam, structural-phase state changes, alloying elements distribution, absorbed hydrogen content and residual ductility after rapid cooling of the oxidized specimens were made. The difference in the oxidation kinetics of the materials studied was revealed, it decreases with the oxidation temperature increase. At the same time, an increase of the Sn content in the alloy influences on specimens 'breakaway' oxidation intensification at 1000 deg. C. It is shown that the mechanical properties of fuel claddings oxidized under the same conditions depend on their alloying composition, the hydrogen fraction absorbed by the specimen and formed by the oxidation ZrO2, a-Zr(O) and 'ex-β' layers structure, which in turn varies depending on the alloy composition. The increase of Nb, Sn and Fe content in the zirconium alloy leads to decrease in residual ductility of fuel claddings after high temperature steam oxidation. According to obtained results the most resistant to high-temperature steam oxidation and embrittlement at temperatures of 1000-1200 deg. C among the materials studied is Zr-1Nb alloy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Engineering-Scale Development/Demonstration of Zirconium Recovery from Used Duel Cladding

Recognizing that the zirconium (Zr) in used nuclear fuel (UNF) cladding is the second largest mass in UNF assemblies, process development studies for the recovery and purification of Zr were initiated and have been in progress at Oak Ridge National Laboratory (ORNL) for the last several years as a major element of waste minimization for advanced nuclear fuel recycle. Together with uranium recovery and recycle, Zr recovery would divert >90% (wt.) of the UNF assemblies from emplacement in a geologic repository by converting the Zr to low-level waste or reuse/recycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.

cladding mechanical performance↗
DOE OSTI2021

Unveiling damage mechanisms of chromium-coated zirconium-based fuel claddings at LWR operating temperature by in-situ digital image correlation

Here, we investigate the coupled thermomechanical fracture mechanisms of coated nuclear fuel claddings at Light Water Reactor (LWR) operating temperatures. These coated claddings are a highly attractive, near-term solution, which addresses the demands for accident-tolerant fuel systems and provide greater oxidation resistance. However, the fracture mechanisms of these coatings, which may create channels for oxidation ingression, must be fully understood prior to implementation. Thus, high-temperature expanding plug experiments were conducted on coated cladding specimens at a temperature of 315 °C, consistent with the operating environment of LWRs. Furthermore, in-situ thermomechanical deformation was measured with stereo digital image correlation during heating and mechanical testing to separately resolve contributions of thermal and mechanical strain. Digital image correlation, supported by acoustic emissions (AE) detection, was also leveraged to track cracking activity during loading. Coating fracture was found to initiate at total hoop strains of 0.34%. The thermal deformation of the coated claddings was investigated via finite element simulations, revealing a bidirectional stress-state within the coating with axial and circumferential strains of 0.026 and 0.031%. This bidirectional stress-state was attributed with the generation of off-axis fracture pattern within the coating as identified via post-experiment scanning electron microscopy. Thus, this study unveiled critical, coupled thermomechanical mechanisms governing the coating fracture of coated claddings at LWR temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2026

High Temperature High Vacuum Mechanical Property Assessment of Zirconium Nuclear Fuel Cladding

This report presents the mechanical characterization of a specific Zry-4 cladding batch serving as the foundation for a diverse range of fuel performance research at Oak Ridge National Laboratory (ORNL). This effort supports research needs for the U.S. Department of Energy (DOE), particularly regarding evaluating accident tolerant fuel (ATF) cladding coating concepts, expanding understanding of cladding response to loss-of-coolant accidents (LOCA) transients, refining post-critical heat flux (CHF) limits (t@T), and upcoming irradiation campaigns. The central objective was to define the baseline performance of the substrate Zircaloy-4 (Zry-4) material leveraged across ORNL Advanced Fuel Campaign (AFC) efforts through controlled high-temperature vacuum tensile testing. This work begins to address gaps in existing models where implementation based on nominal heat-treatment labels, such as stress relief annealed (SRA), often fail to capture the interplay of recovery, recrystallization, and grain growth. To quantify this, data was benchmarked against the Pacific Northwest National Laboratory (PNNL) stress strain model to determine where this material falls in comparison to assumed values for materials in the same heat treatment regime. Analysis of the tensile data revealed that this specific SRA batch exhibits a transitional microstructural state best described by an effective cold-work (CW) parameter of 0.09, diverging from the previous estimation of 0.5 for SRA materials. Additionally, comparative testing of Cr coated specimens demonstrated no distinct difference in axial strength relative to the bare substrate. This suggests that the strengthening benefits of Cr coatings observed in burst scenarios are driven by residual stress mechanisms acting solely in the hoop direction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗