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DOE OSTI2025

Stochastic Ensemble Generation for Improved Characterization of Representing Geologic Variability in a Reservoir: IBDP Case Study for SMART Initiative

This document is a poster covering the findings from activities on training data generation, specifically geologic ensemble generation. The generated geologic realizations captured the range of possible permeability distributions of the subsurface at the Illinois Basin - Decatur Project (IBDP) site, based on available well log variabilities. The percentages of reservoirs and baffles in the injection zone and a truncation of baffle permeability led to more variance in the simulations. This will be used to build forward modeling, history matching, and optimization workflows. The geologic realizations were also ranked according to dynamic measures of hydraulic diffusivity, and simulations confirm a greater contrast between the reservoir and the baffles during injection.

stochastic ensemble generation↗

DOE OSTI2020

Thermal protection barrier for delaying access

A thermal protection barrier including a base layer having a plurality of baffles with perforated walls and a high melting temperature granular media disposed within the plurality of baffles. When the base layer is breached by a thermal threat, the granular media is operable to flow through the perforated walls of the plurality of baffles to provide a controlled flow of granular media to an area proximate the breach for thwarting the thermal threat.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Nuclear reactor support and seismic restraint with core retention cooling features

A nuclear island includes a nuclear reactor, a lateral seismic restraint, and a reactor core retention cooling system. The lateral seismic restraint includes a vertically oriented pin attached to one of the bottom of the lower vessel head and the floor underneath the nuclear reactor, and a mating pin socket is attached to the other of the bottom of the lower vessel head and the floor. The reactor core retention cooling system includes one or more baffles, optionally thermally insulating material, disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head. A plenum is defined between the one or more baffles and the exterior surface of a lower portion of the reactor pressure vessel. The one or more baffles may define a lower plenum inlet surrounding the lateral seismic restraint.

Edwards, Tyler A.↗

DOE OSTI2026

SOLPS-ITER modeling of a dedicated divertor for negative triangularity operation on DIII-D

The design of a new dedicated divertor for negative triangularity (NT) operation on DIII-D with neutral baffles and pumping is informed by SOLPS-ITER transport modeling. This dedicated NT divertor is the latest step in a progression of NT shapes with various divertor characteristics explored on DIII-D, including NT shapes at reduced triangularity and a campaign with stronger shaping that included new armored components on the outboard side. SOLPS simulations played a key role in these divertor designs. Interpretive simulations, using cross-field diffusivities constrained by experimental data in the NT Shelf shape were used to inform the design of the 2023 armor campaign components. A similar procedure used armor campaign data to predict conditions for the dedicated NT divertor. The predictive simulations were used to assess the divertor fluxes, detachment threshold, pumped flux, and neutral leakage. For the dedicated NT divertor, SOLPS simulations and two-point-modeling were used to show the relative impact of magnetic topology (mainly longer connection length) and divertor closure on the divertor conditions relative to the armor campaign. It is predicted that the dedicated NT divertor reaches detachment (measured by target ion flux rollover) at a lower upstream density (≈(1.75−−2.4)×1⁢019m−3) as compared to the armor campaign shape. For the preliminary design geometry, divertor closure reduces the neutral leakage by ≈10%. Parametric optimization indicating further ≈20%–60% improvement in the leakage flux and recycled flux crossing the pump entrance is possible for relatively minor changes to the divertor and baffle layout.

Lore, Jeremy [ORNL] (ORCID:000000029192465X)↗

DOE OSTI2026

Power and particle exhaust for the ARC fusion power plant

To successfully show that fusion is an attractive energy source, the ARCTM fusion power plant will need to operate with a robust, integrated power and particle exhaust solution. To maximise ARC’s fusion power output while avoiding excessive erosion of the plasma-facing components, we will need to radiatively dissipate most of the power crossing the last-closed flux surface, injecting radiating impurities such as argon or neon to access divertor detachment. Divertor detachment will need to be integrated with a high-performance core plasma, and with efficient impurity pumping to prevent the accumulation of helium ash in the core. To access and control detachment in high-performance plasmas, we have designed ARC with up–down-symmetric divertors, with secondary X-points in long, tightly baffled outer legs. Using a core-edge modelling workflow, we predict that with this divertor design, ARC will access detachment with modest argon seeding in the divertor (c Ar,div ∼0.9%), which should have minimal impact on the core ( Z eff,core <0.5) for reasonable argon enrichment (c Ar,div /c Ar,core =6). Due to the high predicted divertor neutral pressure (p div ∼20 Pa), we predict that ARC will sufficiently pump helium to limit ash accumulation in the core (c He,core <2%) for a helium enrichment of c He,div /c He,core =0.4. ARC’s divertor design is expected to increase the stability of a detachment front in the outer divertor leg, helping to prevent divertor reattachment during smaller heat-flux transients such as scrape-off-layer filaments associated with the quasi-continuous exhaust regime. However, this buffering will not be sufficient to prevent divertor reattachment during large type-I edge-localised modes (ELMs), and as such these will need to be avoided on ARC. Experiments on SPARC will be used to select an integrated scenario which avoids or mitigates type-I-ELMs while maintaining access to detachment, good core fusion performance and sufficient impurity exhaust. SPARC experiments will also be used to finalise ARC’s divertor design, by studying the impact of magnetic and first-wall geometry on detachment stability, impurity enrichment and neutral baffling under conditions similar to those expected for ARC. In conclusion, our analysis finds that ARC will have a viable power and particle exhaust solution which is compatible with high-power operations, and this solution will be validated in experiments on SPARC.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Evidence on the effects of main-chamber neutrals on density shoulder broadening

Evidence that density shoulder broadening is dependent on high main-chamber neutral density is presented. Shoulder broadening does not occur when the sources for main-chamber neutrals are minimized using divertor baffles and wide gaps to the first wall (~3× the density decay length). Removing the baffles or reducing the gap to the inner wall both act to increase the density shoulder amplitude in otherwise identical TCV discharges. Radial turbulent transport is correlated with shoulder amplitude.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2022

Scattered-light uniformity imager for diagnosing laser absorption asymmetries on OMEGA

Light scattered from a target is the most-direct measurement for diagnosing laser absorption in a direct-drive implosion. Observations from OMEGA implosions show much larger scattered-light asymmetries than predictions. A new instrument has been developed to absolutely measure the scattered-light intensity and nonuniformity for the purpose of diagnosing the asymmetry. The scattered-light uniformity imager (SLUI) diagnostic records the variation in scattered-light intensity over a transmission diffuser using a charge-coupled device (CCD)/lens assembly. At the standard operating position, an 11.3° (f/2.5) cone of light is collected. A stray light baffle, debris shield, and antireflection absorbing filter are also incorporated into the diagnostic payload inserted into the target chamber. The imaging parts of the diagnostic (light baffle, vacuum window, filters, lens, and CCD camera) are located outside the target chamber. Five SLUIs have been built and deployed in OMEGA’s ten-inch manipulator diagnostic ports, covering almost 5% of the emission surface, enabling an absolute scattered-light measurement should be within a few percent of the global average. Each SLUI system is calibrated offline, providing absolute scattered-light intensity measurements. Based on the measured point spread function, each diffuser plate image contains over 20 000 independent scattered-light absolute-intensity measurements of the variation over the collection cone. SLUI provides a platform to study scattered light and absorption asymmetries, and their possible sources.

47 OTHER INSTRUMENTATION↗

DOE OSTI2023

Mechanisms behind impurity spreading in a channeled, long-leg divertor configuration

One suggested benefit of long-leg, tightly baffled divertor configurations, a common feature included in the designs of many next-generation fusion devices, is access to a delocalized “spread” of impurity radiation along the leg, allowing for a substantial radiation region while maintaining both sufficient distance from the core to prevent pollution of the plasma fuel and avoiding local overheating of the plasma-facing materials. However, it is unclear whether this increased connection length and divertor volume afforded by the long-leg configuration is compatible with impurity transport, and whether it will enable this desired impurity radiation spread at all. Here, the SOLPS 4.3 code is used to assess the distribution of nitrogen and neon impurity radiation in a long-leg, tightly baffled divertor geometry. Further, there is a strong correspondence between the distribution of impurity radiation in the long leg and impurity recycling on the divertor side walls with both the nitrogen simulation set and the neon simulation set, suggesting that impurity spreading along these long legs is due to the interplay between the cross-field transport and associated neutralization of the impurity on the material surfaces.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Neutral leakage, power dissipation and pedestal fueling in open vs closed divertors

The effect of divertor closure on neutral leakage, power dissipation and pedestal fueling is explored using the SOLPS code for different densities including attached and detached divertor conditions for DIII-D open and closed divertors. Here, a metric for divertor closure in terms of the percentage of neutrals escaping the divertor is presented. At the detachment onset the population of neutrals escaping the divertor is 12% for the closed divertor and 35% for the open divertor. This results in a lower contribution of the convective term to the total heat flux and increased power losses in the closed divertor. The comparison between the two geometries reveals that for the same upstream density, the upstream temperature is 23% higher in the closed divertor. The physical mechanisms for detachment at lower upstream density in the closed divertor are also identified with the reduction of the convective term directly resulting in 40% decrease and the increased power losses contributing with a factor of 2. The baffling also lowers the neutral flux refueling the pedestal resulting in further reduction of the core density for achieving detachment. The results introduced here show that the divertor recycling is the primary fueling mechanism in DIII-D. It is shown that fueling occurring in the region between the X-point and the outer midplane is at least four times lower in the closed divertor due the baffling structure which enhances target neutral trapping and consequently reduced upstream leakage. Due to the openness of the inner leg to the common flux and HFS regions both open and closed divertor experience almost identical fueling at the HFS highlighting the need to also consider the inboard divertor for shape optimization.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Comparison between MAST-U conventional and Super-X configurations through SOLPS-ITER modelling

Abstract MAST-U has recently started operating with a Super-X divertor, designed to increase total flux expansion and neutral trapping, both predicted through simple analytic models and SOLPS calculations to reduce the plasma and impurity density detachment thresholds. In this study, utilising the SOLPS-ITER code, we are quantifying the possible gain allowed by the MAST-U Super-X and neutral baffling geometry, in terms of access to detachment. We show that a significant reduction of the upstream density detachment threshold (up to a factor 1.6) could be achieved in MAST-U, for the Super-X, as opposed to conventional divertor geometry, mainly through an increased total flux expansion, neutral trapping being found very similar between the different configurations. We also show that variations of the strike-point angle are complex to interpret in such a tightly baffled geometry, and that a case in which the target normal points more towards the separatrix does not necessarily imply a lower detachment threshold. As in previous calculations for TCV, we quantify the role of neutral effects through developing and applying a quantitative definition of neutral trapping.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Simulation of Helium Flow Visualization Apparatus for Studies of Blanket Cooling in Fusion Reactors

Flow visualization is essential to understanding helium cooling performance. This article investigates helium flow visualization in an apparatus to support the design of the blanket first wall for a fusion reactor. Helium’s safety advantages make it an attractive coolant, but effective cooling and flow visualization remain challenging. Sophisticated simulations are conducted in different test sections to address three key areas in the design of a test apparatus and cooling channel enhancements: the influence of viewing glass windows on flow patterns, the impact of increased heating, and the effectiveness of cooling structures like baffles. Viewing glass windows introduce flow recirculation region bifurcation and flow asymmetry, affecting flow patterns and necessitating careful analysis in future experiments. Increased heating results in helium flow detachment from the heated surface, leading to hot spots. Cooling structures, particularly baffles, prove effective in maintaining consistent attachment to the heated surface, improving the heat transfer performance. Further, this study also examines a tunable parameter in the turbulence model, highlighting the importance of accurate model tuning for future fusion reactor cooling designs. With an imminent helium flow visualization facility, these simulations will be used to optimize cooling structures on the heated wall to improve flow attachment and heat transfer efficiency. This work serves as a first investigation of the helium flow visualization apparatus for blanket cooling enhancement in fusion reactor design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗

DOE OSTI2022

Helical flow filter

A filter includes a filter outer wall, and a filter inner wall spaced apart from the filter outer wall. The filter inner wall and the filter outer wall define a filter void therebetween, and the filter inner wall and the filter outer wall extend along a filter axis. A volume of filtration material is positioned in the filter void, and a plurality of baffles extend between the filter inner wall and the filter outer wall. The plurality of baffles are configured to induce a helical component to an airflow entering the filter void at a first void end relative to the filter axis and exiting the filter void at a second void end opposite the first void end.

Hanson, Thomas David↗

DOE OSTI2023

Heat Exchanger Design for Waste Heat Recovery using Additive Manufacturing

With anticipated increases in global energy use and their associated emissions, there are many efforts to improve the emissions and efficiency of many industrial processes through waste heat recovery (WHR). To maximize improvements, the heat exchangers in WHR power cycles must be designed and optimized for a given heat source. Meanwhile, additive manufacturing (AM) is a single part manufacturing technique that provides the opportunity to develop unique and complex heat exchanger designs. Consequently, this work centers on the modeling, design, optimization, and testing of AM shell and tube heat exchangers for WHR. Specifically, a modified one-dimensional (1+1-D) and three-dimensional (3-D) heat exchanger models were developed and validated, with the 1+1-D model serving as the basis for design optimization. After modifications to the design to allow for AM, heat exchanger geometry optimizations were completed, targeting increased effectiveness across a range of physical parameters. Of the designs considered, nine were additively manufactured using Ti5553, and then tested using an external heat source and working fluid. Experimentally, the nineteen-tube, zero-baffle heat exchanger displayed the best overall performance, with the effects of dimensional accuracy, surface roughness, and baffles in the other designs also indicated. Next, the experimental results were used as inputs for the 1+1-D model. Although the original model was able to predict fluid outlet temperatures within 5% differences, larger variations in the heat exchanger performance parameters, temperature differences, and pressure drop necessitated optimizing the pressure drop and heat transfer calculations within the model. Thus, the resulting improved model was able to predict the heat transfer rates, heat source temperature difference, and effectiveness within 15% differences, while showing improvements in the pressure drops and working fluid temperature differences. The improved model serves as the foundation for future work to improve the design and performance of AM heat exchangers.

Spickler, Bailey Ann↗

DOE OSTI2023

Irradiation-assisted Stress Corrosion Cracking of PWRirradiated Type 347 Stainless Steel (NSUF CINR 19-16567)

This document serves as the final report for NSUF Project 19-16567 (Reference 1) and its supporting contracts (References 2 and 3). It documents completion of testing of specimens fabricated as previously described in Reference 4. At a high level, the purpose of the testing is to generate data to support continued operation of light water reactors (LWR) and provide further understanding of irradiation-assisted stress corrosion crack (IASCC) initiation in baffle-former bolt materials. Baffle-former bolts that had been extracted from commercial pressurized water reactors (PWRs) were machined into miniature 4-point bend specimens and TEM samples at the Westinghouse Churchill Site hot cell facility. These samples were then shipped to the University of Michigan to undergo IASCC initiation testing in simulated PWR primary coolant containing either LiOH or KOH additions. Overall, 15 samples were tested, 8 in water with KOH additions and 7 in water with LiOH additions, until indications of cracking were observed. Crack initiation stress was similar in both water environments at 50-60% of the irradiated yield stress. Post-test examination of the cracking sites, including scanning electron microscopy and transmission electron microscopy, demonstrated that the crack initiation mechanism and crack initiation rates were similar between the two different water chemistries. A grain boundary oxidation and oxide fracture mechanism was observed to be responsible for crack initiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Predicting deviations in ex-core detector response in Krsko NPP during an earthquake with Monte-Carlo neutron transport methods

To study the response of ex-core detectors during an earthquake, an approach using Monte Carlo code MCNP was developed. As a first step to predict the response of the ex-core detectors a detailed MCNP model of the Krsko reactor core was created. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krsko nuclear power plant. Due to the outside the core position of the neutron detectors, the hybrid code ADVANTG is used to generate weight windows to speed up neutron transport outside the reactor core. To be able to use ADVANTG, fixed neutron source had to be reconstructed from the criticality core calculation. It was found that the power range ex-core detector movement has a negligible effect on the value of the thermal neutron flux at the active part of the detector. However, the 5 cm radial movement of the intermediate-range detector leads to 7 % - 8 % change in thermal neutron flux within the active intermediate-range detector region. The analysis continued with the evaluation of the effect of core barrel movement on the ex-core detector response. It was determined that the 2 mm core barrel radial oscillation can lead to ∼ 4% change in thermal neutron flux within the active detector region. It was evaluated that reactor core movement inside the baffle can lead up to ∼ 1.9% and ∼ 1.3% deviation in power-range and intermediate-range detector signal respectively. Analysis showed that the mechanical movement of ex-core neutron detectors could not explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Electrochemical leaching of spent LIBs: Kinetics, novel reactor, and modeling

The use of electrons as main reagent for the recovery and recycling of critical metals from spent lithium-ion batteries (LIBs) is a process electrification strategy that can be used to close the life-cycle loop of LIBs through more sustainable methods. Electrochemical leaching, a process that uses a reductant that is constantly regenerated electrochemically for the leaching of lithium-ion battery black mass (LIBBM), has shown high extraction efficiencies and sustainable scores. However, slow kinetics, reactor design challenges and lack of deeper understanding of the underlying processes are barriers to the optimization, scale-up, and market adoption of this technology. In this paper, a kinetic study and mathematical model for dissolving LIBBM is presented to better understand the underlying mechanisms aiming to reduce the processing time and make predictions for future design and scale-up. The effect of acid and electrochemically mediated reductant concentrations, LIBBM loading, and cathode/reactor designs were explored. As a result, the leaching time was reduced from 7h to under 1h at a pulp density of 73 g/L, without external heating. A novel reactor with parallel baffle electrodes (PBE) was developed, which significantly reduced the leaching time by improving convection in a stirred slurry electrochemical reactor. Dimensionless numbers were deduced from an unsteady state model, which can be used in dimensional analysis for future process design and scale-up.

25 ENERGY STORAGE↗

DOE OSTI2021

Image processing methods for Rayleigh scattering measurements of diesel spray mixing at high repetition rate

This work describes the diagnostic implementation and image processing methods to quantitatively measure diesel spray mixing injected into a high-pressure, high-temperature environment. We used a high-repetition-rate pulse-burst laser developed in-house, a high-speed CMOS camera, and optimized the optical configuration to capture Rayleigh scattering images of the vaporized fuel jets inside a constant volume chamber. The experimental installation was modified to reduce reflections and flare levels to maximize the images’ signal-to-noise ratios by anti-reflection coatings on windows and surfaces, as well as series of optical baffles. Because of the specificities of the high-speed system, several image processing techniques had to be developed and implemented to provide quantitative fuel concentration measurements. We find that these methods involve various correction procedures such as camera linearity, laser intensity fluctuation, dynamic background flare, as well as beam-steering effects. Image inpainting was also applied to correct the Rayleigh scattering signal from large scatterers (e.g. particulates). The experiments demonstrate that applying planar laser Rayleigh scattering at high repetition rate to quantitatively resolve the mixing of fuel and ambient gases in diesel jets is challenging, but possible. The thorough analysis of the experimental uncertainty and comparisons to past data prove that such measurements can be accurate, whilst providing valuable information about the mixing processes of high-pressure diesel jets.

47 OTHER INSTRUMENTATION↗