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Geringer, Josina W.

Publications and source records attributed to Geringer, Josina W..

15 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Microstructure, electrical resistivity, and tensile properties of neutron-irradiated Cu–Cr–Nb–Zr

High strength, high conductivity copper alloys that can resist creep at high temperatures are one of the primary candidates for efficient heat exchangers in fusion reactors. Cu–Cr–Nb–Zr (CCNZ) alloys, which were designed to improve the strength and creep life of ITER Cu–Cr–Zr (CCZ) reference alloys, have been found to have comparable electrical conductivity and tensile properties to CCZ alloys. The measured creep rupture times for these improved alloys is about ten times higher than the ITER reference alloys at 90–125 MPa at 500 °C. However, the effects of neutron irradiation on these alloys, and the ensuing material properties, have not been studied; thus, their utility in a fusion reactor environment is not well understood. This study characterizes the room temperature mechanical and electrical properties of a neutron-irradiated CCNZ alloy and compares them to a neutron-irradiated ITER reference heat sink CCZ alloy. Tensile specimens were neutron irradiated in the High Flux Isotope Reactor (HFIR) to 5 dpa between 250 °C and 325 °C. Post-irradiation characterization included electrical resistivity measurements, hardness, and tensile tests. Microstructural evaluation used scanning electron microscopy, energy dispersive x-ray spectroscopy, and atom probe tomography to characterize the irradiation-produced changes in the microstructure and investigate the mechanistic processes leading to post-irradiation properties. Transmutation calculations were validated with composition measurements from atom probe data and used to calculate contributions to the increased electrical resistivity measured after irradiation. Comparisons with CCZ alloys in the same irradiation heat found that the post-irradiated CCNZ and CCZ alloys had comparable electrical resistivity. Although CCNZ alloys suffered more irradiation hardening than CCZ, the overall tensile behavior deviated very little from non-irradiated values in the temperature range studied.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Analysis of the ASME Code Rules for Subsection III-5-HHB (Composite Materials) for Current HTR Design Requirements

This document includes the critical analysis review of the American Society of Mechanical Engineers (ASME) Section III Division 5 Subsection HH Subpart B (HHB), including Mandatory Appendices, that was published in 2023. In the context of this document, reference to “the code” is specific to this subsection unless otherwise specified. A specific composites task group within the ASME Nonmetallic Design and Materials Working Group, with the support of external experts, was established to perform a gap analysis review. The significant findings are summarized here. The committee response with suggested action items are detailed in the body of the report.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Review of Recent Progress in Plasma-Facing Material Joints and Composites in the FRONTIER U.S.-Japan Collaboration

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, in this work, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material’s fabrication and benefits are described, and some discussion of remaining questions is given.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Codes and standards for ceramic composite core materials for High Temperature Reactor applications

Fiber-reinforced ceramic matrix composites are attractive for high-temperature nuclear applications due to excellent thermal and mechanical properties as well as reasonable-to-outstanding radiation resistance. Over the past 20 years, the use of ceramic matrix composite applications expanded to many commercial non-nuclear industries as fabrication and application of the technologies mature. The ASME Boiler Pressure Vessel Code, under Section III Division 5, provides the design and construction rules for High Temperature Reactor components. It published the first rules for ceramic matrix composites to be used for reactor core components. The rules lay out the quality requirements together with the design and materials criteria for the use and application of silicon carbide- and carbon-based matrix material technologies. As with the established graphite rules, the ceramic composite material rules are structured in Subsection HH (from Section III), that addresses the criteria for class SN nonmetallic core components. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in ASTM Committee C28 on Advanced Ceramics, with a current focus on ceramic composite tubes. This article describes the detail of the composites code, the design methodology and similarities to the graphite code, the guidance for the development of specifications for ceramic composites (for nuclear applications) including recent standard developments, and it mentions the next steps to support licensing aspects by validating the code with benchmarking data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Preparation & shipping of ten neutron irradiated Eurofer97 steel variants to HFIR for SANS experiments

At ORNL, ten variants of Eurofer97 steel were irradiated in the high flux isotope reactor (HFIR) to ~2.94 – 3.24 dpa at 300 ± 30 °C, as part of the EUROfusion Lot-IV collaboration. The irradiations were performed in non-instrumented rabbit capsules in the flux trap region, which included SS-J3 type tensile samples and M4-CVN multi-notch bend bars. Their full set of microstructure and mechanical properties were previously reported in multiple previous publications. In this project, half-broken irradiated and nonirradiated SS-J3 tensile samples from ten alloys, code-named H, I, J, K, L, M, N, O, P and E, were prepared and shipped for facilitating small angle neutron scattering (SANS) experiments at the HFIR general purpose (GP)-SANS beam line. This report summarizes the completed tasks which included canister moves at the Irradiated Materials Examination and Testing (IMET) hot-cell facility to retrieve the samples from long-term storage, loading of the ten irradiated samples at IMET inside lead (Pb) piglets that were specifically provided by ORNL for SANS experiments and radiological shipment from IMET facility to the HFIR hot cells for performing SANS experiments. In addition to the irradiated samples, ten nonirradiated half-broken pieces from the Eurofer97 variants were also provided to HFIR for SANS.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Neutron irradiation of tungsten in hydrogen environment at HFIR

Neutron irradiation of tungsten with and without the presence of hydrogen is needed to understand the influence of hydrogen on microstructure development under fusion reactor conditions. However, there is a risk of ignition if air ingress occurs during seal welding of irradiation capsules in a pressurized hydrogen environment. Therefore, an irradiation capsule was designed that contains several disks of vanadium hydrides at a 30% hydrogen-to-metal atomic ratio. During irradiation, the hydrogen is released from the hydrides as the capsule temperature increases, so the irradiation capsule environment is mostly hydrogen when the capsule reaches its 400 °C design temperature. In this study, we describe the design and operating characteristics of this first-of-a-kind irradiation capsule.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

STEM imaging of irradiation induced defects in Eurofer97 steel variants irradiated in the EUROfusion collaboration

Ten exploratory variants of Eurofer97 reduced activation ferritic martensitic (RAFM) steel were irradiated in HFIR to ~2.94 – 32.4 dpa, 300±30 °C as a part of the EUROfusion collaboration. The irradiations were performed in rabbit capsules. ES21-22 capsule were used for irradiating SS-J3 tensile samples while M4CVN bend bar samples were irradiated in ES31-35 capsules. The mechanical properties and multi-length scale characterization of the microstructures of these ten steels, designated as H, I, P, J, K, L, M, N, O, P and reference E, in the nonirradiated form was reported in "Mechanical properties and microstructure characterization of unirradiated Eurofer-97 steel variants for the EUROfusion project, ORNL/SPR-2018/882," while their irradiated properties and some microstructure analysis was reported in "Post-irradiation examination of Eurofer97 steel variants irradiated to 2.5 dpa, ~300 °C in HFIR for the EUROfusion program, ORNL/SPR-2020/1440." Here, additional microstructures using TEM/STEM and STEM-EDX of the irradiated steels is presented. Characterization was performed at the Low Activation Materials Development and Analysis (LAMDA) laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation hardening and ductility loss of Eurofer97 steel variants after neutron irradiation to ITER-TBM relevant conditions

Ten Eurofer97 steel variants, produced by non-standard fabrication-processing routes and modified alloying chemistries, were studied by neutron irradiations in the high flux isotope reactor. The irradiations were performed to ITER-TBM relevant conditions of ~255–350 °C, 2.94–3.24 dpa. We quantified the irradiation-induced degradation of the steels using mechanical property tests. All the steels suffered from irradiation hardening, where a significant increase in Vickers microhardness and yield stress (σ YS ) occurred, accompanied with severe loss of tensile elongation. The extent of hardening was material dependent. For T irr = 300±30 °C, most steels showed σ YS increase in the range of ~30% to as high as ~66%, except for a low temperature tempered steel with σYS increase below 15%. Despite large losses in elongation, most failures were ductile. Significant post-necking ductility was retained with reduction in area (RA) between 65–75%, but <50% for low temperature tempered steels. The ultimate tensile stress to yield stress (σ UTS /σ YS ) ratios decreased significantly after irradiation, highlighting irradiation-induced strain hardening capacity reduction. No major effect of irradiation on the plastic instability stress (σ PIS ) and true fracture stress of the steels was observed. By comparing the tensile stresses in true stress units and with literature, the results suggest that RAFM steel designing should target materials with a large separation between σ PIS and σ YS , to ensure the materials can maintain large work hardening and uniform deformation capability after irradiation. The tensile data of the steels additionally revealed a compelling evidence of an inverse trend between the change in RA and increase in σ YS of the neutron irradiated Eurofer97 type steels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

HFIR SiC-SiC Composite Clad Tube Bowing Test: Pre-Irradiation Characterization

This report describes the pre-irradiation characterization of the SiC-SiC composite clad tube bowing experiment that will be irradiated in the High Flux Isotope Reactor (HFIR). There are concerns that SiCSiC fuel cladding in light water reactors could undergo bowing because of the non-uniform fast neutron flux profiles. This experiment is combined with the SiC-SiC channel box experiment, which has similar concerns regarding control blade movements in boiling water reactors. The HFIR experiment aims to validate the predicted deformation and stresses in the composite clad specimen after the exposure of fast neutron flux gradients. Significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were thoroughly characterized by using detailed 3D neutronic calculations. The 3D displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses to determine the predicted deformation that will affect the specimens. The tube specimens were characterized by using traditional dimensional inspection and surface profilometry to provide detailed information regarding the pre-irradiation condition as the first step to evaluate the radiation-induced deformation. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped by using a digital microscope and a multidimensional stage. This allows marker spacings to be accurately measured and compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment successfully completed two HFIR cycle irradiations, cycles 492 and 493, which finished on July 25, 2021.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Technological aspects in blanket design: Effects of micro-alloying and thermo-mechanical treatments of EUROFER97 type steels after neutron irradiation

Presently available data on neutron irradiation damage raise doubts on the feasibility of using EUROFER97 steel for a water-cooled starter blanket in a DEMO reactor, since the ductile-to-brittle transition temperature (DBTT) increases significantly for irradiation temperatures below 350°C. The additional DBTT shift caused by H and He transmutation can only be estimated based on very few results with isotopically tailored EUROFER97 steel. Conservative calculations show that the DBTT of EUROFER97 steel could exceed the operating temperature in water-cooled starter blankets within a relatively short time period. This paper presents results from a EUROfusion funded irradiation campaign that was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The paper compares ten newly developed reduced activation ferritic-martensitic (RAFM) steels irradiated to a nominal dose of 2.5 dpa at 300°C. The post-irradiation experiments using Small Specimen Test Technology included hardness, tensile, and fracture mechanics tests combined with fractography and microstructure analysis are presented. Results show that micro-alloying EUROFER97-type steels influenced the mechanical properties but a dominating impact on irradiation damage resistance could not be identified. In contrast, specific thermo-mechanical treatments lead to better DBTT behavior. Discussion about irradiation response to heat treatment conditions is also given. Here, despite requiring data also at high dpa values, the results indicate that with these modified materials an increased lifetime and potentially also an increased operating temperature window can be achieved compared to EUROFER97.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

HFIR SiC Bowing Test Ready to Insert

This report describes the successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient. Excessive bowing of a SiC/SiC channel box in a boiling water reactor could potentially interfere with control blade movements. Similar concerns exist for SiC/SiC fuel cladding in light water reactors. The experiment described herein will provide experimental validation of the structural response of a miniature SiC/SiC channel box and tube specimens with pressurized water reactor diameters during irradiation. The significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were characterized using detailed three-dimensional neutronic calculations. The three-dimensional displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses that predicted the deformation and stresses in the channel box specimen. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped using a digital microscope and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment was successfully assembled and is scheduled for insertion during HFIR cycle 492, which is currently scheduled to run from May 25, 2021 to June 18, 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Thermal diffusivity of irradiated tungsten and tungsten-rhenium alloys

The Japan-US PHENIX project irradiated tungsten materials in the RB-19J capsule experiment in the High Flux Isotope Reactor (HFIR). A gadolinium (Gd) shielding was used to absorb the thermal neutrons and reduce rhenium and osmium generation in tungsten. Pure tungsten and K-doped W-3% Re samples were irradiated at 532 – 662 °C to dose of 0.21-0.46 dpa, with the grain orientation perpendicular or parallel to the disk surface. Thermal diffusivity measurements were performed from 100 °C to 500 °C. Furthermore, additional measurements followed after annealing up to 900 °C. Irradiated pure tungsten specimens showed similar thermal diffusivity results compared with an unirradiated W-1% Re specimen in another study. The transmutation amount of Re was calculated to be about 0.52% for those specimens that showed good agreement with this study. Specimens irradiated in this study to different doses presented almost the same thermal diffusivity. Annealing up to 800 °C resulted in no recovery of thermal diffusivity. These results show that the contribution of crystalline defects to degradation of thermal diffusivity is quite limited. In addition, the thermal diffusivity of the irradiated specimens was getting close to that of the unirradiated specimens at elevated temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tensile properties of powder-metallurgical-processed tungsten alloys after neutron irradiation near recrystallization temperatures

The tensile properties of powder-metallurgical-processed Pure W, K-doped W, W-3%Re, and K-doped W-3%Re were examined after neutron irradiation up to 0.7 dpa at 910–1020 °C with a thermal neutron shield in the High Flux Isotope Reactor (HFIR). After irradiation, recrystallized Pure W (R) exhibited a brittle fracture mode, while recrystallized K-doped W-3%Re (R) exhibited a ductile fracture mode at 500 °C. K-doped W-3%Re (R) has fine grains, and hence, contains a considerable number of grain boundaries that act as sinks for irradiation defects. Solid solute Re in the W matrix could improve not only the mechanical properties of W, but also its resistance to neutron irradiation. At 500 °C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at ~1000 °C did not induce hardening of stress-relieved (SR) W materials, but SR W materials tended to exhibit a decrease in the ultimate tensile strength (UTS) and an increase in total elongation (TE). Here, the softening due to the recovery and recrystallization of SR W materials and the hardening due to the formation of irradiation defect clusters were balanced during irradiation at ~1000 °C, and ductility was exhibited without an increase in strength.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of HFIR neutron irradiation on fracture toughness properties of standard and Ni-doped F82H

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. Here, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58 Ni, and F82H doped with 1.4% 60 Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness.

36 MATERIALS SCIENCE↗