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At least 811 records · Page 45

A study on pseudo interface wave technique for CRDM weld defects in nuclear power plants

The nuclear power plant inspection is very important for the safety issue. However due to some radiation and geometric problems, the detection of CRDM(Control Rod Drive Mechanism) can be very difficult by using conventional Ultrasonic Testing method. Also the shrink fit boundary condition can also be an obstacle for the inspection in this paper, instead of conventional Ultrasonic Testing, guided wave was used for the detection of some complicated structures. The CRDM nozzle was installed in reactor head with perfect shrink fit condition by using stainless steel. The wave amplitude distribution on the circumferential direction was calculated with various boundary conditions and the experimental result shows a possibility of the defect detection on J-groove weld.

36 MATERIALS SCIENCE↗

Simulated Aging of Metal Tritides by Ball Milling Induced Gas Incorporation

The study of tritium aging effects on materials requires a significant time commitment as a consequence of its 12.3-year half-life, making developmental studies prohibitively difficult and expensive. However, detailed knowledge of long-term aging effects is critical to the development of structural and storage materials for future fusion reactor technologies. As a result, multiple approaches to simulated aging effects have been investigated. Here, we report a method of simulated tritium aging achieved though the incorporation of trapped gases via high-energy ball milling of LaNi 4.25 Al 0.75 alloy storage material. Experimental results verify the presence of trapped gases by a combination of temperature programmed desorption and LECO chemical analysis. Following gas incorporation, we find that many of the degraded hydrogen sorption properties found in aged storage materials are reproduced by the ball milled powders.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (alloy 617) and Haynes 230 (alloy 230). Both alloys are considered to he the primary candidate structural materials for very high-temperature reactors (VITITRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900 degrees C for the effective stress range of 15-35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes was observed in all the studied conditions. Tertiary creep was found to he dominant over the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries were found to be the main reasons for the limited secondary regime and were also found to be the major causes of creep fracture. The creep curves computed using the adjusted creep equation of the form epsilon= cosh 1(1 rt) + P-sigma ntm agree well with the experimental results for both alloys at die temperatures of 850-950 degrees C.

36 MATERIALS SCIENCE↗

Designing Radiation Resistance in Materials for Fusion Energy

Proposed fusion and advanced (Generation IV) fission energy systems require high performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts of transmutant hydrogen and helium isotopes. After a brief overview of fusion reactor concepts and radiation effects phenomena in structural and functional (non-structural) materials, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials where vacancies are immobile at the design operating temperatures, or construct high densities of point defect recombination sinks. Environmental and safety considerations impose several additional restrictions on potential materials systems, but reduced activation ferritic/martensitic steels (including thermomechanically treated and oxide dispersion strengthened options) and silicon carbide ceramic composites emerge as robust structural materials options. Materials modeling (including computational thermodynamics) and advanced manufacturing methods are poised to exert a major impact in the next ten years.

Zinkle, Steven J↗

Stability of Y–Ti–O precipitates in friction stir welded nanostructured ferritic alloys

Nanostructured ferritic alloys, which have complex microstructures which consist of ultrafine ferritic grains with a dispersion of stable oxide particles and nanoclusters, are promising materials for fuel cladding and structural applications in the next generation nuclear reactor. This paper evaluates microstructure of friction stir welded nanostructured ferritic alloys using electron microscopy and atom probe tomography techniques. Atom probe tomography results revealed that nanoclusters are coarsened and inhomogeneously distributed in the stir zone and thermomechanically affected zone. Three hypotheses on coarsening of nanoclusters are presented. Finally, the hardness difference in different regions of friction stir weld has been explained.

36 MATERIALS SCIENCE↗

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT and onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”

99 GENERAL AND MISCELLANEOUS↗

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fabrication and processing of next-generation oxygen carrier materials for chemical looping combustion

Among numerous methods of controlling the global warming effect, Chemical Looping Combustion is known to be the most viable option currently. A key factor to a successful chemical looping process is the presence of highly effective oxygen carriers that enable fuel combustion by going through oxidation and reduction in the presence of air and fuel respectively. In this study, CaMnO 3-δ was used as the base material and doped on the A-site (Sr or La) and B-site (Fe, Ti, Zn and Al) by 10 mol % of dopants. Solid state reaction followed by mechanical extrusion (optimized paste formula) was used as the preparation method A series of novel doped perovskite-type oxygen carrier particles (Ca x La (Or Sa) 1-x Mn 1-y ByO 3-δ (B-site = Fe, Ti, Al, or Zr)) were synthesized by the proposed extrusion formula. The produced samples were characterized with XRD, SEM, BET and TGA techniques. According to the results obtained from TGA analysis, the oxygen capacity of the samples ranged between 1.2 for CLMZ and 1.75 for CSMF. Reactivity and oxygen uncoupling behaviors of the prepared samples were also evaluated using a fluidized bed chemical looping reactor using methane as the fuel at four different temperatures (800, 850, 900, 950 °C). All of the oxygen carriers showed oxygen uncoupling behavior and they were able to capture and release oxygen. Mass-based conversion of the perovskites was calculated and temperature increase proved to increase the mass-based conversion rate in all of the samples under study. Gas yield was calculated at 950 °C as well, and results showed that CLMZ, CM and CSMF showed 100% gas yields and CLMF and CSMZ showed approximately 85% yield in fluidized bed reactor, which is a high and acceptable quantity. Based on extended reactor tests the modified calcium manganese perovskite structures (CSMF) can be a good candidate for future pilot tests.

30 DIRECT ENERGY CONVERSION↗

Report on Understanding and Predicting Effects of Thermal Aging on Microstructure and Tensile Properties of Grade 91 Steel for Structural Components

This report provides an update on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY17 (M2AT-17AN1602017), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has been produced and hot-rolled into plates.

36 MATERIALS SCIENCE↗

Report on the Initial Loading of SS3 Specimens from the First Commercial Heat of Alloy 709 in Sodium Materials Test Loops for Long-term Exposure

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy A709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18(M2NT-18AN050503011), under the Work Package NT-18AN05050301, “A709 Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of a mechanistic thermal aging model for Grade 91

This report provides a summary on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18 (M2NT-18AN050502061), under the Work Package NT-18AN05050206, “G91 Code Extension Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors. The thermal aging study of G91 involves three types of thermally-aged specimens: (1) specimens machined from two heats of G91, G91-H1 and H30176 that have been tested for the Advanced Materials Development Programs in the past years. These specimens were aged at 550, 600, and 650°C for times up to ~64,000 h; (2) specimens fabricated from the heads of the archived crept specimens (13 in total) of six heats of G91, H5349, H30394, H30383, H10148, H30176. These specimens were aged at 427 - 538°C and for times up to 132,647 h; (3) specimens made from a tube removed from the Kingston coal-fired power plant after exposure for 155,000 h at 550°C. These three sets of specimens provide a comprehensive data set for understanding and predicting the effect of long-term thermal aging on microstructure and tensile properties of G91. Microstructural analysis of the thermally-aged G91 revealed that thermal aging causes significant microstructural changes: (1) the tempered martensite formed during tempering is unstable, and suffers significant recovery during aging, manifested by an increase in subgrain width, a decrease in subgrain length, and a decrease in subgrain aspect ratio (i.e. subgrains become more equiaxed) as the aging temperature and time increase. The subgrain recovery is facilitated by the reduction of dislocation density within subgrains; (2) M23C6 carbides in the as-received G91 are distributed along grain and subgrain boundaries and play an important role as pinning obstacles against subgrain coarsening. M23C6 carbides coarsen during aging, and the coarsening of M23C6 carbides reduces the pinning force on the boundaries; (3) MX carbonitrides in the as-received G91 are distributed uniformly within subgrains, providing the precipitation hardening effect. They coarsen with increasing aging temperature and time, but the coarsening rate is low; (3) a new phase, the Laves phase intermetallic forms during aging. The formation of the Laves phase removes the Mo solutes from the matrix, reducing the solid solution strengthening effect. The Laves phase particles grow rapidly once they are nucleated, and they precipitate primarily along boundaries. They may have a pinning effect on boundaries, but should provide little to the alloy strength. The evolution of each microstructural constituent during thermal aging can be well described by individual microstructural models. The estimated activation energy for MX coarsening was 175 kJ/mol, equivalent to the activation energy for boundary and core diffusion in a-Fe, while the estimated activation energy for subgrain coarsening was significantly lower, ~96 kJ/mol. Thermal aging of G91 results in the reduction in the yield stress and the ultimate tensile strength but has an insignificant effect on the uniform and total elongations. The reduction in tensile strength is more pronounced as the aging temperature and time increases. The in situ tensile tests with high-energy X-rays revealed that sub-boundary strengthening plays a dominant role, and the strengthening role of M23C6 particles may be accounted for by subgrian boundary hardening, and three strengthening mechanisms can be superimposed to describe the tensile strength of G91: subgrain boundary hardening, MX precipitation hardening, and Mo solid solution strengthening. A microstructure – strength model was establ

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Report on the initiation of planned FY18 short and intermediate term creep rupture tests and creep-fatigue tests on the first commercial heat of Alloy 709

This report provides an update on the initiation of planned FY18 short and intermediate term creep rupture tests and creep-fatigue tests on the first commercial heat of Alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY18 (M3NT-18AN050502031), under the Work Package NT-18AN05050203, “A709 Code Case Testing” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

FY19 Report on the Status of Sodium Exposure Tests of the A709 First Commercial Heat

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy A709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The focus of the FY19 program was (a) to complete the sodium exposure of sub-sized specimens from A 709 lab heats and 316H (a reference material) and (b) to continue sodium exposure of sub-sized specimens from A 709 commercial heat 58776.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Reconstruction of 3D Concrete Microstructures Combining High-Resolution Characterization and Convolutional Neural Network for Image Segmentation

After water, concrete is the second most used material in the world. Concrete’s forming adaptability and low-cost constituents make it a predominant material used in the construction of civil infrastructures in nuclear power plants such as concrete biological shields, containment buildings, turbine buildings, fuel handling and storage buildings, underground piping for cooling, cooling towers, and so on. Depending on environmental and operating conditions, these passive structures are subject to time-dependent phenomena that can either enhance (e.g., continued hydration) or degrade concrete’s structural performance. Unlike components such as the reactor pressure vessel and the primary circuit, concrete composition varies regionally because it is manufactured using local aggregates and cement. Hence, concrete performance metrics over time cannot be derived confidently using empirical relations. Alternatively, the specific characteristics of the local concrete constituents and their assemblage in concrete must be considered.

36 MATERIALS SCIENCE↗

Mechanical creep testing and characterization of Cr coated Zircaloy

Zirconium based cladding has been developed for use in light water reactor fuel applications due to the material’s desirable low neutron scattering cross section. To help improve corrosion resistance in accident conditions, such as a loss of coolant accident (LOCA), a chromium coating can be applied to the outer diameter of the Zircaloy tube. Cladding strength at reactor relevant temperatures is important for the structural performance. Stress relaxation creep testing was performed on Cr coated Zircaloy-4 to measure strain rate as a function of stress for 350 and 650°C. These were performed at both 5 ± 1 % and 7.5 ± 1 % strain. The lower temperature provides results near normal operating temperature, while the higher temperature provides results relevant to off-normal conditions where recrystallization will take place. Characterization was also performed to analyze the structure of the Cr coating and Zircaloy-4 as well as the interface between the two.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

BOILER Experiment Material Characterization and HFIR Irradiation Status

Pre-oxidized alumina-forming austenitic (AFA) steels have been previously identified as candidate alloys for structural components in lead-cooled fast reactors (LFRs). They offer compatibility with liquid Pb, high-temperature strength, formability, and cost advantages. However, variations in Ni content can affect the formation and stability of the Al 2 O 3 layer, influencing compatibility with liquid Pb. The effect of fast neutron irradiation on Al 2 O 3 stability in liquid Pb also requires evaluation. Therefore, understanding how Ni concentrations impacts pre-oxidized AFAs under combined extremes of irradiation and liquid metal corrosion is essential before safe deployment. The Behavior Of In-situ Lead Environments & Radiation (BOILER) experiment was developed under the Nuclear Science User Facilities (NSUF) program to integrate alloy development, irradiation experiment design, and irradiated materials characterization. In this effort, two pre-oxidized AFA steels with 20 wt% and 25 wt% Ni, hereinafter referred to as GA05-20Ni and GA05-25Ni, were produced. An irradiation experiment was then planned for the High Flux Isotope Reactor (HFIR), designed for passive heating of irradiation rabbit capsules from gamma heating in the HFIR flux trap (1 × 10 15 n/cm 2 ·s, >0.1 MeV). This heating melts Pb and exposes the pre-oxidized AFA steel specimens to nominal temperatures of 400 and 650°C. Detailed neutronics and thermal analyses were performed, though based on nominal design rather than as-built, as-irradiated conditions. This report documents further characterization of the pre-oxidized AFAs in the unirradiated condition. It also includes as-built thermal analysis using measured component dimensions, updated fill gas concentrations, and actual HFIR irradiation positions. Finally, the report summarizes capsule fabrication, current irradiation status, projected completion, estimated damage accumulation, and initial plans for post-irradiation examination plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Characterization of ceramic fuel powder packing fractions to support INFLUX

Triply periodic minimal surface (TPMS)-based structures show marked potential in novel nuclear reactor fuel designs, as their high surface area-to-volume ratio increases the efficiency of heat transfer out of the fuel, enabling safer, more innovative reactor designs. This milestone report addresses the role of dUO 2 powder processing route on the fill behavior of TPMS-based cladding shells to understand and advance the feasibility of manufacturing TPMS-based nuclear fuel forms. dUO 2 powder was processed through either a dry granulation route, varying consolidation pressure, or through milling, varying milling time, milling method and milled size distribution. The lowest tapped bulk densities (TBD), but best powder flowabilities, were obtained when testing unprocessed dUO 2 powder which was prone to self-agglomeration and formed low-density spheroids. The highest TBD and lowest flowabilities were obtained when using powder produced by hammer-milling dUO 2 powder to pass through a 200-mesh sieve, which led to particles with angular morphologies. Powder produced by dry granulation exhibited TBD that varied according to the consolidation pressure used to form the initial pellets and exhibited improved flowabilities when compared to hammer-milled material. Because of the large span of granule sizes formed as well as the irregular shape associated with the granules, a packing fraction of 0.69 was achieved, exceeding the analytical solution for random close packing of mono-sized spheres. TPMS polymer shells were loaded with unprocessed, granulated, and hammer-milled dUO 2 powders, and their qualitative packing behaviors were analyzed using x-ray computed tomography (xCT). TBDs calculated after loading TPMS polymer shells were 10-20% lower when compared to tapped bulk density measurements taken in a glass graduated cylinder, indicating a non-trivial impact on the tapped bulk density of either the TPMS channel size, TPMS channel surface material, powder cohesiveness, or a combination of the two parameters. A metallic zircaloy-4 TPMS shell will be loaded with hammer-milled dUO 2 powder upon receipt of the shell from Oak Ridge National Laboratory (ORNL) and shipped to Idaho National Labs (INL) for subsequent hot isostatic pressing (HIP) densification experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗