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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Shear Punch Testing of BOR-60 Irradiated TEM Specimens

As a part of the project “High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation” an Integrated Research Program (IRP) project from the U.S. Department of Energy, Nuclear Energy University Programs (NEUP), TEM geometry samples of ferritic cladding alloys, Ni based super alloys and model alloys were irradiated in the BOR-60 reactor to ~16 dpa at ~370°C and ~400°C. Samples were sent to Los Alamos National Laboratory and subjected to shear punch testing. This report presents the results from this testing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Initial assessment of metallurgical interaction of clad/base metal systems

Molten Salt Reactors (MSRs), both liquid fuel and sold fuel (e.g., the fluoride salt-cooled high temperature reactors, FHR’s) offer several advantages relative to conventional, water-cooled nuclear power systems. However, currently approved ASME structural materials have not been optimized for use in molten salt environments and demonstrating materials performance in molten salts is challenging. Toward the goal of enabling MSR designs, corrosion resistant claddings offer an attractive solution to help mitigate corrosion concerns and enable the use of approved ASME materials. This report assesses metals and alloys that may be effective as corrosion resistant cladding with select structural alloys. The structural alloys of interest are capable of use at high temperatures ≥600°C for times ≥100,000 hours. Potential cladding alloys were assessed based on their corrosion resistance in both chloride and fluoride based salts, physical properties, mechanical properties, nuclear performance, and resistance to environmental-assisted cracking. Additionally both thermodynamic and kinetic computational modeling were used to assess interactions between the base metal and the cladding. Results to date have identified three candidate alloy systems that help enable MSR designs; nickel, molybdenum, and tungsten based claddings. This report discusses pros and cons of each class of cladding and outlines future work to develop these base metal / clad systems into industrial products suitable for use in commercial nuclear power systems.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Burst characteristics of advanced accident-tolerant FeCrAl cladding under temperature transient testing

Here assessment of burst characteristics of accident-tolerant fuel (ATF) claddings is essential to evaluate the safety margins of nuclear reactors, and better understanding can enable accelerated licensing for reactor concepts that include the new materials. Therefore, this study investigated the burst behavior of an ATF candidate of C26M, which is an iron–chromium–aluminum alloy (FeCrAl) under transient testing like the simulated loss-of-coolant accident (LOCA) conditions, except the water-quenching phase, in light-water reactors (LWRs). The effect of LOCA specimen length was assessed in terms of post-burst tube parameters. No critical length effect was determined on the burst pressure, burst location, and burst size measurements. Burst temperature showed larger variation likely due to its measurement approach used in this study. The post-test diametral strain was identified as the critical parameter to reduce the specimen length as compared to tube burst length or width. Postmortem optical metrology and digital image correlation were employed to determine the local strain-state during the simulated LOCA, which showed the loading path was close to equibiaxial conditions in regions away from the burst location, rather than internally pressurized conditions (or “plane-strain” tension). Pst-test microstructural characterizations of the FeCrAl revealed that ductile damage was present at the edge of the outer radial surface while the rest of the material was ruptured via cleavage at LOCA burst temperatures. Furthermore, internal grain boundary cracks were observed at locations away from the tube burst region. This behavior was considered to occur during the cooling down period of the simulated LOCA test. Overall, this study aimed to provide essential data for fuel performance code development by considering micromechanics of the deformation and failure.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Advanced Fuels Campaign Cladding & Coatings Meeting Summary

The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Steam Oxidation Testing in the Severe Accident Test Station

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative would significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Evaluation of Zircaloy-4 Welding and Hydrogen Charging Effects for Use as the Structural Material for the Target Solution Vessel and Support Lines of SHINE

The subcritical hybrid intense neutron emitter (SHINE) is a proposed 99 Mo production facility to be constructed in the United States. It uses a deuterium (D) and tritium (T) fusion source to cause subcritical fission in an aqueous uranium solution, which is cycled to collect the 99 Mo products. The target solution vessel (TSV) is the container of the aqueous uranium solution, so it must remain structurally sound during operational exposure to water, fission products, neutrons, and hydrogen isotopes. The TSV will be large enough to require welding, so the exposure of joints to these same conditions must also be considered. The candidate material being investigated is Zircaloy-4. While Zircaloy-4 has a long history in the nuclear industry, it has primarily been used as fuel cladding at service temperatures of ~300°C. Zircaloy-4 must be evaluated for behavior under neutron irradiation at SHINE-relevant temperatures of <100°C. In this project, the effects of hydrogen, welding, and neutron irradiation at low temperatures are investigated.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Uranium Zirconium Alloy Dissolution

This Special Instructions provides a standard method to dissolve irradiated or unirradiated UZr alloys. The scope of this procedure was developed to facilitate total dissolution of UZr solid samples with and without cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of surface modified Zr cladding by coating technology for ATF

It is expected that the accident tolerance fuels (ATF) will provide an enhanced fuel performance of active cooling in the reactor core for a longer time period during design-basis events as well as maintain or improve the fuel performance during normal operations and operational transients. Thus, various ATF cladding concepts have being suggested and developed in many countries. Among them, the coated cladding concept is studied in this work after the considerations of various performances in both the accidents and normal operation of LWRs. In the surface coated cladding concepts, the combination of coating material and coating method is a very important factor to determine the ATF cladding performance. As a new coating material, Cr-based alloy was developed to increase the corrosion resistance under normal operation as well as the oxidation resistance during an accident. From the high temperature oxidation test at 1200 Celsius degrees, the Cr-based alloy showed a superior oxidation resistance to Zircaloy-4 cladding, and showed more improved corrosion resistance than Zircaloy-4 cladding under 360 Celsius degrees PWR simulated loop conditions. The 3D laser coating and a combination of arc ion plating and laser beam scanning were developed to obtain a good adhesion property of corrosion resistance alloy coated on a Zr substrate. The corrosion resistance alloy (Cr-based alloy and FeCrAl alloy) coated layer used by two types of coating methods was maintained without peeling off during the ring compression test at room temperature, although some cracks were formed on the coated layer. After a high-temperature oxidation test at 1200 Celsius degrees for 2000 s, a very thin oxide layer, which was a few microns, was formed on the corrosion resistance alloy layer. However, the ZrO{sub 2} phase, which was thicker than 100 μm was formed on the Zr alloy cladding surface. Thus, it is known that the high-temperature oxidation resistance of commercial Zr alloy cladding can be considerably improved by the coating of corrosion resistant alloys. In addition, the partial oxide dispersion strengthened (ODS) treatment using a laser beam scanning (LBS) method was applied to decrease the ballooning and rupture opening during accident conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

High temperature oxidation of PCD coated Zr alloy

Polycrystalline diamond (PCD) coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys and decrease corrosion rate of zirconium alloy during standard operation. Zirconium alloys are widely used as cladding and construction material in almost all types of nuclear reactors, where usually creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper zirconium alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations (Raman spectroscopy) were done for zirconium alloy covered with polycrystalline diamond layer before and after high temperature steam exposure. Weight increase and hydrogen release were measured during steam exposure. The results show that the oxidation kinetics of PCD coated samples is similar to kinetics of uncoated samples at almost all temperatures, but with lower hydrogen production and lower weight gains. Most important difference is in later beginning of breakaway oxidation at 1000 Celsius degrees, when breakaway oxidation starts on PCD coated sample 15 minutes later. Raman spectroscopy revealed graphitization of the PCD layer and the diamond sp{sup 3} bounds are lost and the Raman signal consists only from the two graphite-related D bounds. It implies conversion of PCD layer to the homogeneous non-crystalline carbon layer with a large content of graphitic phase. Overall hydrogen production and weight gain values are generally lower then for uncoated samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advanced Manufactured Strain Sensors for Extreme Environments

Experimentation at irradiation test facilities are essential for reducing the innovation time of developmental fuels, fuel cladding, and structural materials employed in next-generation nuclear reactors. However, due to the harsh conditions generated in such reactors and limited instrumentation space, the evaluation of a material’s mechanical properties is often limited to characterization after the materials have been removed from reactor conditions, or post-irradiation examination. These experiments are costly, time-consuming, and fail to capture the critical time-evolving phenomena that occur during the irradiation experiments. Advanced manufactured digital image correlation patterns and strain sensor devices serve as two promising technologies that can be deployed in the confined and challenging orientations of these irradiation experiments while also providing key insight on salient materials phenomena (i.e., mechanical properties). To guide the development of the printed strain sensors prior to their deployment in critical experiments, the adhesion strength between the substrate and printed film interface is measured via tensile testing and a non-contact laser-induced spallation technique. The establishment of these process control steps helped guide the successful fabrication and testing of direct-write strain sensing devices discussed in this work. The fabrication process controls are necessary for enabling the sustained operation of these strain sensing device through experimentation and minimize the potential for premature failure.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Advanced Manufactured Strain Sensors for Extreme Environments

Experimentation at irradiation test facilities are essential for reducing the innovation time of developmental fuels, fuel cladding, and structural materials employed in next-generation nuclear reactors. However, due to the harsh conditions generated in such reactors and limited instrumentation space, the evaluation of a material’s mechanical properties is often limited to characterization after the materials have been removed from reactor conditions, or post-irradiation examination. These experiments are costly, time-consuming, and fail to capture the critical time-evolving phenomena that occur during the irradiation experiments. Advanced manufactured digital image correlation patterns and strain sensor devices serve as two promising technologies that can be deployed in the confined and challenging orientations of these irradiation experiments while also providing key insight on salient materials phenomena (i.e., mechanical properties).To guide the development of the printed strain sensors prior to their deployment in critical experiments, the adhesion strength between the substrate and printed film interface is measured via tensile testing and a non-contact laser-induced spallation technique. The establishment of these process control steps helped guide the successful fabrication and testing of direct-write strain sensing devices discussed in this work. The fabrication process controls are necessary for enabling the sustained operation of these strain sensing device through experimentation and minimize the potential for premature failure.

36 - MATERIALS SCIENCE↗
DOE OSTI2016

AREVA enhanced accident tolerant fuel program - Current results and future plans

The purpose of this paper is to provide a brief overview of the enhanced accident tolerant fuel (EATF) concepts investigated by the AREVA team in Phase 1 to support the U.S Department of Energy Enhanced Accident Tolerant Fuel program and concerning light water reactors. The main goal of the 'fuel pellets with additive' concept is to improve the economics, performance and operational flexibility of fuel in light water reactors while featuring improved safety performance under normal and off-normal conditions. Within this concept 2 nuclear fuels were explored: Cr{sub 2}O{sub 3} doped UO{sub 2} pellet and composite UO{sub 2} pellets. The second concept is the coating of zirconium cladding whose main goal is to reduce cladding oxidation and hydrogen generation during accident conditions. The coating materials were chosen to be compatible with the specific deposition process, to provide robustness in the nuclear environment and have a relatively small neutron cross section. Three coatings were explored: chromium coated claddings, electrophoretic deposition of coating, and MAX phase coatings. MAX phases are a family of the layered compounds with a general formula M{sub n+1}AX{sub n} with n = 1, 2, or 3, where M is a transition metal, A is an A-group element, and X is either C or N. Another improvement to the fuel cladding that was considered is to replace the zirconium cladding altogether with an alternative cladding material which can substantially improve reactor response during postulated accident conditions. Two concepts were investigated: molybdenum claddings and SiC/SiC composite cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Bulk and surface controlled diffusion of fission gas atoms

Fission gas retention and release impact nuclear fuel performance by, e.g., causing fuel swelling leading to mechanical interaction with the clad, increasing the plenum pressure and reducing the gap thermal conductivity. All of these processes are important to understand in order to optimize operating conditions of nuclear reactors and to simulate accident scenarios. Most fission gases have low solubility in the fuel matrix, which is especially pronounced for large fission gas atoms such as Xe and Kr, and as a result there is a significant driving force for segregation of gas atoms to extended defects such as grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. Several empirical or semi-empirical models have been developed for fission gas release in nuclear fuels, e.g. . One of the most commonly used models in fuel performance codes was published by Massih and Forsberg. This model is similar to the early Booth model in that it applies an equivalent sphere to separate bulk UO 2 from grain boundaries represented by the sphere circumference. Compared to the Booth model, it also captures trapping at grain boundaries, fission gas resolution and it describes release from the boundary by applying timedependent boundary conditions to the circumference. In this work we focus on the step where fission gas atoms diffuse from the grain interior to the grain boundaries. The original Massih-Forsberg model describes this process by applying an effective diffusivity divided into three temperature regimes. In this report we present results from density functional theory calculations (DFT) that are relevant for the high (D 3 ) and intermediate (D 2 ) temperature diffusivities of fission gases. The results are validated by making a quantitative comparison to Turnbull's and Matzke's data. For the intrinsic or high temperature regime we report activation energies for both Xe and Kr diffusion in UO 2±x , which compare favorably to available experiments. This is an extension of previous work. In particular, it applies improved chemistry models for the UO 2±x nonstoichiometry and its impact on the fission gas activation energies. The derivation of these models follows the approach that used in our recent study of uranium vacancy diffusion in UO 2 . Also, based on the calculated DFT data we analyze vacancy enhanced diffusion mechanisms in the intermediate temperature regime. In addition to vacancy enhanced diffusion we investigate species transport on the (111) UO 2 surface. This is motivated by the formation of small voids partially filled with fission gas atoms (bubbles) in UO 2 under irradiation, for which surface diffusion could be the rate-limiting transport step. Diffusion of such bubbles constitutes an alternative mechanism for mass transport in these materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Performance of Waterlogged LWR fuel during RIA

Fuel pins with small defects could allow for the gradual ingress of water into the cladding during normal operation. If these 'waterlogged' fuel pins are subjected to a rapid power change, the internal pressurization could greatly exceed that of a normal fuel pin due to rapid conversion of the water to steam. This could induce either a premature or a more energetic failure during a design basis RIA than in non-defected fuel pins. Experimental studies on this type of fuel are very limited and do not appear to have been fully reported in the open literature. Results from a variety of tests conducted at the Transient Reactor Test Facility (TREAT), Special Power Excursion Reactor Test (SPERT) facility, Nuclear Safety Research Reactor (NSRR), and Power Burst Facility (PBF) are reviewed in this paper. These tests were intended to evaluate both the physical consequences of failure and the threshold and causes for waterlogged fuel failure during RIA. The available data suggests that although waterlogged fuel pins may burst at a lower energy deposition threshold, they do not pose a significant new hazard during design basis accidents. However, data is limited to tests conducted on fresh fuel samples only.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Advanced Characterization of Fuel Pellet and Cladding from the HB Robinson NPP

In order to predict the behavior and response of materials to extreme environments of a nuclear reactor, comprehensive models have been developed over the past several decades. These models are typically validated against carefully designed single or double variable experiments which may be systematically altered in order build up a model of behavior under the conditions being examined. In reality, this does not necessarily compare well with the real application situations whereby the behavior of the material under irradiation is determined by multiple variables changing simultaneously e.g. temperature, irradiation flux and fluence, and pressure. This is particularly so when considering the highly dynamic environment of a fuel assembly within a nuclear power plant (NPP) core. This work aims to provide near-atomic resolution microstructural and chemical data of a fuel/clad specimen that has been used in service within a commercial NPP. Atom probe tomography and S/TEM characterization of the cladding showed the presence of small scale clusters containing Fe and Cr, along with segregation of these minor solutes to dislocations. SEM examination of the fuel pellet showed relatively large fission gas bubbles that had burst due to the metallographic polishing process to prepare the sample for examination. SEM-EDX mapping revealed the presence of Xr bubbles located beneath the surface that still retained their Xe concentration. Metallic precipitates were also observed. S/TEM-EDX of the fuel material also revealed the presence of small metallic precipitates. In addition, small Xe bubbles were also found to be present in the material, lying along grain boundaries only. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Problem Specification for FY12 Modeling of UNF During Extended Storage

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the Advanced Modeling and Simulation Office (AMSO) of the US Department of Energy, Office of Nuclear Energy (DOE/NE) has invested in the initial extension and application of advanced nuclear simulation tools to address relevant needs in evaluating the performance of used nuclear fuel (UNF) during extended periods of dry storage. There are many significant challenges associated with the prediction of the behavior of used fuel during extended periods of dry storage and subsequent transportation. The initial activities are focused on integrating with the Used Fuel Disposition (UFD) Campaign of the DOE/NE and a demonstration that the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) for modeling the mechanical state of the cladding after decades of storage. This initial focus will model the long-term storage of the UNF and account for the effect, and generation, of radially and circumferentially oriented hydride precipitates within the cladding and predict the end of storage (EOS) mechanical state (stress, strain) of the cladding. Predicting the EOS state of the cladding is significant because it (1) provides an estimate of the margin to failure of the cladding during nominal storage operation and it (2) establishes the initial state of the fuel for post-storage transportation. Because there are significant uncertainties associated with the storage conditions, hydride precipitate formation, and the beginning of storage (BOS) condition of the UNF, this will also allow for the development of a rigorous capability to evaluate the relative sensitivities of the uncertainties and can help to guide the experimental and analysis efforts of the UFD Campaign. This document is focused on specifying the problem that will be solved with AMPFuel. An associated report, documents the specifics of the constitutive model that will be developed and implemented in AMPFuel to account for the presence and predict the generation of the hydride precipitates. This report satisfies the deliverable for the DOE Office of Nuclear Energy, Advanced Modeling and Simulation Office, milestone M3MS-12OR0605083, 'Definition of Problem Specification,' which defines the problem to be solved that will satisfy milestone M2MS-12OR0605081, 'Demonstration of the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code for modeling UFD.' This document should provide sufficient detail to model a high burnup pressurized water reactor (PWR) fuel rod to provide an estimate of the end of storage (EOS) mechanical state of the cladding. The fuel rod and irradiation history are based on seven cycles of irradiation in the CP and L H.B. Robinson nuclear reactor, which achieved a discharge burnup of 66.682 MWd/kgU. The fuel has been experimentally examined for storage conditions by Argonne National Laboratory for the NRC. In addition, we have compiled a list of key factors that have been shown to strongly influence the EOS state of the fuel and have identified baseline values and ranges of uncertainties that will be considered. The simulations that will be performed have been described in detail and include the modeling assumptions and boundary conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

High-performace cladding-pumped erbium-doped fibre laser and amplifier

We report cladding-pumped erbium-doped fibre laser and amplifier configurations. Through fibre design optimisation, we have achieved a record-high laser slope efficiency, 40 % with respect to absorbed pump power ({lambda} = 976 nm), and an output power of 7.5 W. The erbium-doped fibre amplifier efficiency reaches 32 %.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗