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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Harvesting Reactor Pressure Vessel Beltline Material from the Decommissioned Zion Nuclear Power Plant Unit 1

The decommissioning of the Zion Nuclear Power Plant (NPP) provided a unique opportunity to harvest and study service-aged reactor pressure vessel (RPV) beltline materials. This work, conducted through the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Program, aims to improve the understanding of radiation-induced embrittlement to support extended nuclear plant operations. Material segments containing the Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B Heat B7835-1 base metal, obtained from the intermediate shell region with a peak fluence of 0.7 × 10 19 n/cm 2 (E > 1.0 MeV), were extracted, cut into blocks, and machined into test specimens for mechanical and microstructural characterization. The segmentation process involved oxy-propane torch-cutting, followed by precision machining using wire saws and electrical discharge machining (EDM). A chemical composition analysis confirmed the expected variations in alloying elements, with copper levels being notably higher in the weld metal. The harvested specimens enable a detailed evaluation of through-wall embrittlement gradients, a comparison with the existing surveillance data, and the validation of predictive embrittlement models. This study provides critical data for assessing long-term reactor vessel integrity, informing aging-management strategies, and supporting regulatory decisions to extend the life of nuclear plants. This article is a revised and expanded version of a paper entitled, “Current Status of the Characterization of RPV Materials Harvested from the Decommissioned Zion Unit 1 Nuclear Power Plant”, PVP2017-65090, which was accepted and presented at the ASME 2017 Pressure Vessels and Piping Conference, Waikoloa, HI, USA, 16–20 July 2017.

harvesting beltline material↗
DOE OSTI2024

The effect of neutron irradiation on conventional and nanocrystalline nickel: NSUF Work Package UA-22PN080501 Milestone M3UA-22PN0805012 (Final Report)

Even though nanocrystalline materials (20-100 nm) present an unprecedented potential, scientific knowledge related to the effect of neutron irradiation on the mechanical properties and microstructure is still scarce. Most of the past studies were conducted using ion irradiation which may not have the same effect as neutron irradiation because of the smaller irradiation volume and the higher dose rate. To reach a firm conclusion on the potential of nanocrystalline materials for nuclear reactor applications, extensive study of model metals with different stacking fault energy (SFE) is required to elucidate their behavior in radiation environments. Nanocrystalline copper and nickel are typically chosen because they are commonly used as model FCC metals in studies of radiation effects. Nickel is an FCC metal with a high stacking-fault energy (~125 mJ/m 2 ) compared to copper (~45 mJ/m 2 ). Hence, microcrystalline and nanocrystalline nickel samples were irradiated in the INL’s Advanced Test Reactor (ATR), as a part of FY08 North Carolina State University NSUF Irradiation Experiment# 96 to evaluate the irradiation behavior of these materials. The objective of this FY20 NSUF project# 19122 is to perform PIE (at PNNL NSUF facility) on previously ATR-neutron irradiated (1.2 and 2.6 dpa; 80-89°C) nanocrystalline and microcrystalline nickel samples to investigate the changes in mechanical properties and microstructures and evaluate whether nanocrystalline nickel is relatively more radiation resistant compared to conventional microcrystalline nickel. To perform PIE at PNNL, sixteen neutron irradiated specimens (microcrystalline and nanocrystalline) were transferred from the NSUF Nuclear Fuels and Materials Library at INL. Experimental techniques such as SEM/EBSD, XRD, TEM, Vickers microhardness and tensile testing were employed to characterize the effect of neutron irradiation on the microstructure and mechanical properties of nanocrystalline nickel and compared them with corresponding characteristics of microcrystalline nickel.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Remaining Useful Strength (RUS) Prediction of SiCf-SiCm Composite Materials Using Deep Learning and Acoustic Emission

Prognosis techniques for prediction of remaining useful life (RUL) are of crucial importance to the management of complex systems for they can lead to appropriate maintenance interventions and improvements in reliability. While various data-driven methods have been introduced to predict the remaining useful life (RUL) of machinery systems or batteries, no research has been reported on the remaining useful strength (RUS) prediction of silicon carbide fiber reinforced silicon carbide matrix (SiCf-SiCm) materials with pivotal role in its potential usage as a structural material in nuclear reactors and turbine engines. Knowledge of its degradation process is of the utmost importance to the manufacturers. For this purpose, two approaches based on the machine-learning techniques of random-forest (RF) and convolutional neural network (CNN) are proposed to predict the RUS of SiCf-SiCm using only acoustic emission (AE) signals generated during the material’s stress applying process. Experimental results show that the CNN models achieved better predictive performance than the RF models but the latter with expert-engineered features achieves better prediction for AE signals in the early stage of degradation. Additionally, our results demonstrate that both models can correctly predict the SiCf-SiCm RUS as evaluated by our robust testing method from which the best average root mean square error (RMSE) and Pearson correlation coefficient of 3.55 ksi units and 0.85 were obtained.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Enabling Load Following Capability in the Transatomic Power MSR

This project is dedicated towards designing a fuel processing system that enables liquid-fueled molten salt reactors (MSR) to load follow by removing the dissolved xenon in the fuel salt. As one of the Gen-IV nuclear reactor concepts, the molten salt reactor receives increasing development interests in the recent years. One distinguishing feature of the liquid-fueled molten salt reactor is its improved ability to operate in a load-following mode by including the unique online fission product removal system. Load-following means that the reactor changes its power output based on the demand on the grid. Most of the current operating nuclear reactors have limited load-following ability and operate as the base load on the grid. Due to the rapid increase of solar energy, the requirement on load-following capacity is significantly increased because of the varying power output of the solar panels, yet the traditional load-following capacity is expected to decrease as the decarbonization of the grid continues. Therefore, the ability to perform load-following operation for the nuclear reactors will greatly enhance the resilience of the grid and make nuclear energy more economically competitive. This load-following feature is included in many commercial molten salt reactor designs, such as the designs by Transatomic Power, Terrestrial Energy, and Flibe Energy. Unfortunately, detailed analysis of the fuel processing system for commercial scale MSRs is still lacking, as well as how the fuel processing quantitively impacts the load-following operation. Moreover, experimental data for many of the underlying physics of fuel processing is limited. This project aims to pave the way for the fuel processing technology to advance to the commercial stage by performing combined experimental and simulation research. During the project period, four interconnected aspects of the development of the fuel processing system in liquid-fueled molten salt reactors are investigated. These aspects are the simulation and analysis of the fission product removal system, the fuel cycle simulation, the coupled reactor neutronics and thermal hydraulics transient simulation, and the gaseous fission product removal experiment. Multiphase CFD simulations are performed for components of the processing systems, and simplified air-water experiments are carried out to provide validation data. It is concluded that the CFD simulation can satisfactorily predict the system level performance of the components, and engineering models are constructed based on this success. Fuel cycle analysis is performed for two representative MSR design, the MSBR and the Transatomic Power MSR. Open-source code SaltProc is developed to incorporate the unique fuel processing system of the MSRs. It is concluded that the removal of xenon is essential for load-following operation in thermal spectrum MSR and Molten Salt Breeder Reactor. For the Transatomic Power MSR, the xenon poisoning effect is negligible due to its relatively fast neutron spectrum, though the overall fuel cycle economics still benefits from the removal of xenon. Coupled reactor neutronics and thermal hydraulics transient simulation is performed specifically for the Transatomic Power MSR. It is concluded that the reactor core design could perform power ramping fast enough to satisfy load-following operation. Combining the findings from each aspect, it is concluded that the load-following operation of a thermal neutron MSR is dependent upon the removal of xenon, which could be achieved for a commercial sized reactor using continuous inert gas sparging in a separate system with reasonable dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Magnetostrictive Ultrasonic Waveguide Transducer for In-Pile Thermometry

Real-time and reliable temperature measurement on nuclear fuels is crucial for the safe operation of existing pressurized-water reactors and future advanced nuclear reactors. Magnetostrictive materials deform when subjected to a magnetic field or exhibit magnetization variation when stressed. Here, based on these properties, this study prototyped an ultrasonic thermometer (UT) consisting of a magnetostrictive waveguide, a dc coil providing appropriate magnetic biasing, and an ac coil generating an acoustic impulse and detecting the resulting acoustic echoes. By tracking the time of flight between the excitation pulse and the echoes, the UT can potentially detect nuclear fuel cladding temperature from a long distance. In this study, magnetostrictive iron–gallium alloys, or Galfenol, were selected as the waveguide due to their large magnetostriction, superior temperature survivability, excellent radiation resilience, and high mechanical robustness. A multiphysics finite-element model considering electrical, magnetic, and mechanical dynamics in the magnetostrictive UT was then developed. The model exhibited an error of 0.24% in time-of-flight simulation and, therefore, enabled computer-aided design and guided signal processing. Between room temperature and 120 °C, the new Galfenol-based UT exhibits a linear sensitivity of 162.8×10 ₋6 °C ₋1 , which is 51.7% higher than a previous magnetostrictive UT based on iron–cobalt–vanadium alloys.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

In-situ and ex-situ characterization of ion-irradiated AM materials

Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

In situ irradiation of spent nuclear fuels

To improve the economics of commercial nuclear reactors, nuclear vendors and utilities are seeking approval for increased burnup usage of the existing nuclear fleet. This is especially critical for meeting the clean energy initiative by the United States Government, Department of Energy (DOE). However, one of the key challenges the nuclear industry faces in this regard is that the regions exposed to high burnup and low temperatures exhibit a fine-grained microstructure with large bubbles known as high-burnup structure (HBS) [1]. The formation of HBS has been correlated to the diminished performance of the reactor, as well as fuel fragmentation and pulverization during transient and accidental conditions [2]. Therefore, it is paramount to understand the mechanisms for HBS formation along with its impact on the properties and performance of nuclear fuels. While existing programs, such as Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Advanced Fuel Campaign (AFC) are focusing on evaluating the performance impact of HBS, the physical mechanisms contributing to HBS formation are still not fully understood. In addition, having predictive capabilities and sound understanding of the microstructural evolution of nuclear fuel is essential for accelerated development, qualification, and deployment of new nuclear materials and novel reactor designs for advanced nuclear reactors. There is a lack of consensus among researchers regarding the mechanisms leading to such restructuring observed in HBS. Grain subdivision due to polygonization versus recrystallization, continuous versus discrete recrystallization occurring in tandem or conjunction, etc., have been proposed and debated. In general, it is hypothesized that defect accumulation and dislocation interaction within the grains cause the realignment of dislocations into grain boundaries, leading to the new subgrain formation, which over time transforms into new grains. However, due to the lack of transient data, the importance of fission rate, irradiation, thermal, and stress history of the fuel on the restructuring could not be assessed. In situ microstructural evolution under various irradiation conditions is desired to bridge this gap. Alternatively, phase-field-based models have been developed to capture HBS formation via discrete recrystallization utilizing the classical nucleation approach [3–5]. However, in these models, the grain nucleation criteria are often defined based on empirical relations for burnup and fission gas density leading to dislocation density change. A mechanistic approach to capture the dislocation interaction with the microstructural features leading to grain subdivision is lacking.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Neutron spectroscopy of plutonium using a handheld detection system

The ability to distinguish multiple forms of plutonium from one another, such as oxide and metal, is paramount in areas of nuclear nonproliferation and international safeguards. In its metal form, plutonium can be readily used in a nuclear weapon, while oxide forms are associated with nuclear reactor fuel. Oxide-based plutonium forms emit neutrons with an energy spectrum that is significantly different from the fission neutrons that are emitted from plutonium metal. Organic scintillation detectors output pulses that are proportional to the neutron energy deposited, and therefore present a means of distinguishing these plutonium forms based on their energy spectra. In this work, metal and oxide forms of plutonium were measured using a handheld detection system based on an organic glass scintillator. Monte Carlo modeling of these experiments was performed to provide insight into the origin of the features in the observed light output spectra. Through analysis of multiple regions of these spectra, in a matter of minutes we were able to unambiguously discriminate oxide and metal plutonium forms from one another and from a plutonium-beryllium neutron source, which was considered for comparison because these sources are commonly used in industrial applications. The ability to discriminate weapons-usable material from nuclear reactor fuel has applications in nuclear treaty verification and safeguards.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Development of a MOOSE-based crystal plasticity model with irradiation defect evolution for irradiation creep in 316

Irradiation creep and irradiation swelling are two of the lifetime limiting factors for structural materials in nuclear reactors. These mechanical effects are driven by irradiation defect evolution and the interaction of those defects with dislocations in the microstructure. We present here a coupled cluster dynamics – crystal plasticity approach to model irradiation swelling and creep behavior in 316 SS. The time-dependent evolution of irradiation defects is calculated with a cluster dynamics approach and passed to the crystal plasticity model to compute the dislocation evolution. We show the impact of the irradiation defect evolution on the stress state in the material, which drives inelastic deformation through dislocation motion. The inelastic deformation in the 316 SS is dependent on the dose rate, where the inelastic deformation driven by the early-stage irradiation defect evolution determines the mechanical behavior of the 316 SS.

316 Stainless Steels↗
DOE OSTI2023

Nuclear Material Control & Accounting for Pebble Bed Reactors (FY 2023 Summary Report)

This report discusses the work done under the US Department of Energy NE-5 Advanced Reactor Safeguards and Security Program during FY 2023. It provides a summary of material control and accounting (MC&A) for pebble bed reactors (PBRs) and addresses some of the main challenges with current PBR MC&A approaches that will inform safeguards and security by design efforts. The efforts to date have focused on tristructural isotropic (TRISO) pebble fuel material accounting and control including working with partners in industry, loss and production of nuclear material as part of reactor operations, burnup modeling and measurements, uncertainty quantifications for such modeling and measurements, statistical approaches needed, and measurement methods. The unique fuel management and utilization in a PBR, where the fuel in spherical form is introduced and circulates through the reactor, poses special challenges for MC&A. This contrasts with traditional water-cooled reactors in which the fuel is contained in large assemblies and can be easily identified and counted. Even online fueled reactors, such as the CANDU reactors (none of which operate in the United States), are significantly different because the fuel is still contained in relatively large assemblies, is uniquely identified, and the number of assemblies that pass through the core on an annual basis is much fewer than the hundreds of thousands that circulate in a PBR, none of which are uniquely identified. Additionally, the nature of the TRISO fuel results in very low heavy metal loading with each pebble containing less than 10 g of uranium and on the order of less than 1 g of fissile material. This low fuel density and the robustness of the TRISO particles are major features of the TRISO fuel from a safety basis as each TRISO particle and pebble acts as a containment for the nuclear material and fission products during normal and accident conditions. This also results in very low plutonium loading per pebble during normal operations, which is on the order of 0.1 g at full burnup. A major feature of PBRs is that they will allow for significantly higher burnup, on the order of 160 GWd/THM compared to the burnup of traditional LWRs, which is on the order of 45 GWd/THM. This is achieved by monitoring the pebbles as they circulate through the reactor and allowing them to be reintroduced into the core until the desired burnup is achieved and they are removed from the reactor and enter the spent fuel storage areas.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Multiscale and Machine Learning Modeling for Additive Manufacturing

Additive manufacturing (AM) techniques provide the opportunity to simultaneously design new materials and components with complex structures in less time, enabling faster material developments. Even though compositionally similar, the texture of the materials produced by such techniques is significantly different from conventionally manufactured materials. Additively manufactured materials produces highly heterogeneous microstructure within a single build. Such variations in the microstructure make qualifying AM products challenging for extreme environment applications. Understanding the AM process and its influence on the materials’ microstructures/properties is paramount for evaluating the workability and performance of the manufactured materials. The performance of AM materials for advanced nuclear reactor applications is of interest to the Advanced Materials and Manufacturing Technologies (AMMT) program under the Department of Energy Office of Nuclear Energy. Hence, considering the microstructural variabilities in the AM products and their impact on the performance of the material, it is important to correlate the process conditions to the final product and establish a process-structure-property- performance (PSPP) correlation for AM materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Interaction of Polymethyl Methacrylate with Boehmite-Filmed Aluminum Cladding Under Gamma Irradiation

This paper presents an overview of ongoing work to qualify the Advanced Test Reactor (ATR) driver fuel elements that have been affected by irradiation-degraded polymethyl methacrylate (PMMA) flux wands. Irradiation testing was performed on PMMA material in contact with aluminum clad material. The cladding was prefilmed with a boehmite oxide layer, an important feature of the ATR driver fuel. The effects on the boehmite layer due to gamma irradiation of the PMMA-aluminum clad system were investigated. PMMA embrittlement, followed by softening and degradation, occurred at high radiation levels. Adhesion between the cladding and irradiated PMMA was observed. Flow testing at prototypic ATR flow rates demonstrated the effective removal of the adhered material. Measurements of the boehmite layer thickness were performed, and Raman spectroscopy was utilized to detect the presence of boehmite in the irradiated PMMA material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

SEMI-EMPIRICAL MODELING OF IRRADIATION-INDUCED DIMENSIONAL CHANGE IN H-451 NUCLEAR GRAPHITE

Nuclear graphite has been used as a moderator material in nuclear reactor designs dating back to the first reactor to reach criticality, Chicago Pile 1, in 1942. In addition, it is anticipated to be used in the conceptual Generation four (GenIV) Molten-salt reactors (MSRs) and the High-temperature gas-cooled reactors (HTRs). The macroscopic dimensional change observed in irradiated nuclear graphite is a property change of significant importance. Largely, volumetric change provides valuable insight into the in-service lifetime of graphite components used in nuclear reactors. The dimensional change behavior varies amongst each grade of nuclear graphite due to processing techniques and the resulting microstructure. In this work, historic data for nuclear graphite H-451 is revisited. A semi-empirical methodology is proposed to describe the dimensional change behavior as a function of temperature for nuclear graphite H-451. The turnaround dose, or when there is a reversal of the dimensional change from contraction to expansion, is proposed to be a thermally activated process and thus can be described by an Arrhenius model. On the atomic scale, H-451 is sp2-bonded carbon atoms with some degree of disorder regardless of orientation. Towards that end, the activation energy is assumed to be a constant irrespective of orientation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) and its alloys are used as fuel cladding in nuclear reactors. These materials have their low absorption cross-section for thermal neutrons and have good mechanical and thermal properties. In addition, they are used as tritium getters in tritium-producing burnable absorber rods (TPBARs), in the form of nickel-plated Zircaloy-4 tubes. To produce 3 H, the breading blankets such as lithium aluminate are irradiated in pressurized water. 3 H produced in this process is captured by the getter and 3 H reacts chemically with Zr to form metal hydride (ZrTx). The precipitation of metal hydride results in not only the volume increase and hydrogen embrittlement of zircaloy but also changes in getter’s chemical properties. This degrades the getter’s performance. Therefore, understanding the behavior of 3 H (such as diffusion and solubility) in zircaloy is important to understand getter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Annual Report on NCSP Technical Support task in BNL during FY24

This work reports on the investigations done during FY 24 in collaboration with two 2024 Summer undergraduate interns through the DOE SULI and BNL SURP programs. These efforts were partially funded by the Nuclear Criticality Safety Program through the Technical Support Succession Plan task. The first project, developed by Ian Snider, focused on testing the impact of uncertainties in thermal cross sections for multiple materials in nuclear reactor simulations. The second project aimed implement concurrent into BNL’s machine-learning code to correct spin mis-assignments in neutron resonances, the Bayesian Resonance Reclassifier (BRR).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Laser Powder Bed Fusion Additive Manufacture Nb1Zr Development

Next generation fission and fusion nuclear reactors require materials that can withstand operating temperatures greater than 500 °C, neutron irradiation doses of up to 200 displacements per atom (dpa), and potentially corrosive coolants such as the alkali liquid metals sodium, lithium, and NaK (Na33K eutectic alloy). Refractory alloys, such as Nb1Zr (Nb-1wt%Zr) and Molybdenum alloy TZM (Mo-0.5wt%Ti-0.08wt%Zr) have been traditionally considered viable candidates for advanced fission and fusion reactor concepts. However, it is relatively difficult to generate complex geometries of interest from these alloys using traditional manufacturing methods. In addition, there needs to be a concentrated effort to address refractory metal challenges at elevated temperature operation. In order to generate complex geometries of interest, modern manufacturing techniques are considered to increase the technological readiness level (TRL), cost-effectiveness, and schedule savings. This work focused on the continued development of laser powder bed fusion (L-PBF) additive manufacturing (AM) to improve both design flexibility, evaluate microstructure and properties, and ultimately accelerate the TRL and qualification of these processes and alloys for components to potentially be put into service. Niobium alloy Nb1Zr was identified through a down-selection process outlined in previous reports as a candidate to develop in L-PBF AM. Historically, Nb1Zr had been explored for high temperature fast spectrum fission reactors for both terrestrial and space applications. Molybdenum alloy TZM has also been considered for these reactor concepts due to exceptional high-temperature strength, creep resistance, and stability under irradiation. L-PBF AM of TZM has previously been investigated at LANL under the Microreactor program, NASA, ORNL, and in academia. However, due to the crack prone nature of TZM, L-PBF AM of TZM resulted in significant microcracking and additional development is required to pursue viable maturation. Other AM methods have been found to be more successful in printing TZM, and those alternatives approaches are discussed in this effort. The efforts detailed in this report focused on continued development of Nb1Zr through L-PBF and development of TZM via L-PBF and electron powder bed fusion (E-PBF). The objective of this work was to further the development of these AM techniques for the chosen refractory alloys, elucidating and addressing associated challenges through characterization of several demonstration builds. At LANL, Nb1Zr builds were completed using an EOS M290 and M400 machines, and a refractory alloy-dedicated L-PBF system, the Xact Metal XM200G, was installed. The XM200G primary purpose was to do the Nb1Zr parameter development process; however, due to difficulties associated with the machine installation and qualification process, it was decided to pivot development to the larger M400 and M290 machines. Although the supply of Nb1Zr powder was limited, it was sufficient to generate sub-scale metallographic specimens for the purpose of parameter development. This was first accomplished on the EOS M400 then the M290 due to machine schedule availability. Further development of TZM has been initiated at the University of Texas El Paso (UTEP) under contract with LANL to use both a heated build envelope L-PBF machine and E-PBF machine that have been found in the literature to mitigate microcracking. UTEP was provided with TZM powder and build plates to support parallel TZM parameter development across both machines. As part of the contract, UTEP will also be conducting microstructural characterization once optimized process parameters have been identified. The optimized process parameters for each machine will be used to generate a series of metallographic, mechanical, and surface finish specimens for subsequent characterization and testing. In the next section, we provide a detailed discussion of the methodology used for investigating the feasibility of leveraging these alloys for use in advanced reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation

Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni–Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5–9 displacements per atom (dpa) and temperatures ranging from 430–683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni 3 Si stoichiometry in a Ni core–Si shell structure. This core–shell structure provides insight into a potentially unique nucleation and growth mechanism—nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A New NCSP Nuclear Criticality Safety Training and Pipeline Program

Nuclear criticality safety (NCS) training is vital and mandatory for NCS professionals to ensure the safe processing, handling, storage, and transportation of fissionable materials outside nuclear reactors. The requirement for this training is prescribed in several international and domestic guidance documents.

(NCS) Training↗