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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Toward a Fully Integrated Multiphysics Simulation Framework for Fusion Blanket Design

Fusion is an attractive clean-energy solution, thanks to its various advantages, such as reduced radioactivity, little high-level nuclear waste, ample fuel supplies, and increased safety. However, the harsh operating environment introduced by a complex fusion plasma system makes design and integration of fusion blankets incredibly challenging and time-consuming. This work focuses on developing a fully integrated multiphysics simulation framework based on an advanced open-source platform—the Multiphysics Object-Oriented Simulation Environment (MOOSE)—to alleviate the difficulties in fusion blanket design and integration. MOOSE is a massively parallel finite element/volume multiphysics simulation platform that has been widely adopted within the nuclear fission community. Even though fission and fusion are fundamentally different, they involve similar multiphysics phenomena. A fully integrated open-source multiphysics simulation framework tailored for the fusion blanket design will be implemented by leveraging the well-established multiphysics capabilities in MOOSE. Once successfully developed, this fully integrated framework will rapidly evaluate a blanket design concept and offer insights for subsequent iterations. As the first step, we will mainly aim to integrate neutronics analysis, system thermal hydraulics simulation, and full 3-D heat transfer calculations. The efficacy of the integrated framework will be verified using an innovative solid ceramic blanket design. While the project’s final goal is to enable a fully integrated multiphysics simulation platform for various fusion blanket concepts, here this work, as a preliminary step, will mainly focus on a solid ceramic breeder helium-cooled blanket.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Modeling a Hypothetical Fusion Blanket Design Using MELCOR-TMAP

MELCOR (not an acronym) is a nuclear safety code developed by Sandia National Laboratories for the Nuclear Regulatory Commission. Idaho National Laboratory collaborates with Sandia to maintain and develop a version of MELCOR specifically for fusion device applications. Most recently, the tritium migration analysis program (TMAP) has been incorporated into MELCOR for fusion, in a version called MELCOR-TMAP. We show the results applying MELCOR-TMAP to model steady-state inventory and tritium retention in a liquid immersion blanket design for a hypothetical fusion device under steady-state operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Qualification and Commissioning of Helium Flow Loop Experiment for Blanket Design Measurements

Sufficient cooling of plasma-facing materials remains an outstanding challenge in the design of fusion reactor blankets in commercial power demonstration plants. Due to its chemical inertness and low neutron interaction cross section, pressurized helium is a candidate coolant fluid for such systems; however, helium has a small thermal mass compared to liquid coolants, potentially reducing heat removal performance. To address this need, a number of heat transfer enhancements have been proposed to improve the cooling efficiency of such components, thereby decreasing pumping power needs and improving overall plant efficiency. Toward this end, a helium flow loop experiment (HFLE) has been designed and commissioned to test advanced passive heat transfer enhancements in unit-cell test sections, providing necessary data for model validation and subsequent system design. The HFLE is designed to provide flow of pressurized (up to 4 MPa) helium at flow rates up to 80 g/s, enabling heat transfer and pressure drop measurements in test pieces at Reynolds numbers in excess of 180 000. To explore the effects of novel and complex heat transfer enhancements, test sections are produced via additive manufacturing, providing geometries not typically obtainable by conventional machining. Here in this work, we present results from HFLE commissioning and the initial thermal-hydraulic tests of an additively manufactured rifled-rib test section. Results are compared to smooth pipe correlations, and plans are described for future HFLE measurements. These preliminary experiments indicate the utility of the HFLE for heat transfer enhancement testing and simulation validation activities.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Development of integrated neutronics and thermal analysis capabilities to support design and optimization of fusion engineering demonstration facility systems and blanket design (Final Scientific/Technical Report)

A fusion engineering demonstration facility would be the first step in the commercial fusion plant developmental pathway that aims to be an engineering demonstration of tritium self- sufficiency. One of the primary requirements for tritium self-sufficiency, as well as to ensure low plant tritium release to the external environment, is to minimize the tritium loss through the internal components, which requires accurate predictions of the tritium behavior for a wide range of materials and system conditions. This grant addressed the lack of a comprehensive model which accounts for particular conditions, such as the significance of temperature fields and neutronics information, and the systematic uncertainty quantification of associated material properties that impact tritium generation, utilization, and loss mechanisms in the blanket.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Development of integrated neutronics and thermal analysis capabilities to support design and optimization of fusion engineering demonstration facility systems and blanket design (Final Scientific/Technical Report)

A fusion engineering demonstration facility would be the first step in the commercial fusion plant developmental pathway that aims to be an engineering demonstration of tritium self-sufficiency. One of the primary requirements for tritium self-sufficiency, as well as to ensure low plant tritium release to the external environment, is to minimize the tritium loss through the internal components, which requires accurate predictions of the tritium behavior for a wide range of materials and system conditions. This grant addressed the lack of a comprehensive model which accounts for particular conditions, such as the significance of temperature fields and neutronics information, and the systematic uncertainty quantification of associated material properties that impact tritium generation, utilization, and loss mechanisms in the blanket.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

3D high-fidelity automated neutronics guided optimization of fusion blanket designs

The compact Fusion Pilot Plant (FPP) is defined in the recent National Academies of Sciences, Engineering, and Medicine report as the next step of fusion energy demonstration with a $50$ MWe peak net electricity production, $Q_e$ greater than $1$, and at least $3$ hours of continuous operation. This fusion pilot plant will be a test bed enabling materials, designs, and fuel management assessment, and it will represent an engineering challenge because of its high-fusion power and compact design targets. Previous reactor data is limited to experiments operating in different design space ranges. Therefore, design iterations and assessments should rely on high-fidelity first-principle theoretical and computational models. The high-fidelity integrated modeling of the plasma is a fundamental part of fusion energy research. However, the whole device modeling is often neglected, utilizing low-fidelity, system-level analysis. Recently, the need for high-fidelity multi-physics modeling was recognized, resulting in a selection of integrated tools. Further, autonomous design optimization requires a streamlined framework that perturbs the design point, reruns the analysis, and examines the outputs. However, high-fidelity analysis requires complex geometry specification that is difficult to perturb. This work presents the parametric CAD generation tool TRACER and a new neutronic workflow. TRACER allows the perturbation of the geometry representation, creating geometry files ready for further analysis. The streamlined neutronic workflow allows efficient and accurate calculations. The two new tools coupled together were used to perform a 3D high-fidelity multi-objective, multi-input optimization of an "ARC Class" compact tokamak design. The workflow was driven by an optimization driver for full automation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

MHD flow in liquid metal blankets: Major design issues, MHD guidelines and numerical analysis

The design of breeding blankets represents the major challenge for fusion reactor engineering because of performance requirements and severe operating conditions in terms of heat load and neutron flux. Liquid metal alloys such as lead-lithium, due to their lithium content, can be used to breed tritium, one of the plasma fuel components, and owing to their high thermal conductivity, they may serve as coolants. On the other hand, there are technical issues related to the fact that the liquid metals are electrically conducting and interact with the plasma-confining magnetic field. Induced electric currents and generated electromagnetic forces affect velocity and pressure distribution in the blankets. Magnetohydrodynamic (MHD) flows for fusion applications have been often investigated in simplified geometries, such as pipes, ducts, bends, with focus on their fundamental features. These analyses are essential, since results remain valid as background for the development of blanket designs, even when a concept is dismissed. However, the conceptual study of fusion blankets requires to take into account the global multiple effects, that arise when the full system is considered. Progress made in fusion-related MHD research results from combined numerical and experimental activities. In this paper we review and summarize features of 2D and 3D MHD flows that are typical in liquid metal blankets, together with available correlations for MHD pressure losses. This knowledge can provide simple design MHD guidelines that support a preliminary estimate of MHD effects in a blanket concept, in terms of pressure drop and flow distribution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Assessment of thermal and radiation induced creep in the dual cooled lead lithium blanket

The creep in recently designed blanket for the fusion nuclear science facility is described by decomposing it into thermal, irradiation and cavity swelling creep. Initial estimates of thermal creep using elastic high temperature rules showed that the recent heat transfer improvements at the first wall (FW) is feasible, and the FW can withstand heat flux of 0.35 MW/m 2 at 3000 hrs without rupture. A viscoplastic model that is based on Norton's power law and relies on Multiphysics coupling of solid mechanics and heat transfer modules is used to capture the inelastic deformation in the blanket. Results showed that the design peak heat flux of 0.25 MW/m 2 produced maximum thermal creep strain of 0.45 % and the relaxation of the thermal stress at the FW. Irradiation creep is prescribed to be proportional to the displacement damage dose and applied stress. The displacements from irradiation creep radially decrease from the FW to the back wall, but the maximum deformation is found at the back wall that is connected to the Helium manifold due to the high stresses at the region. In conclusion, cavity swelling creep is dominant at steady state and combined with radiation creep to produce displacement of about 8 mm at the FW.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Evaluating Lithium-Lead Mixtures for Increased Tritium Breeding in Fusion Energy Blanket Systems: Exploring the Impact of Composition and Temperature

This project investigates hypo- and hyper-eutectic liquid lithium-lead (PbLi) mixtures as a coolant in fusion blanket systems, with a focus on testing the compatibility of compositions with better tritium breeding ratios (TBR). The 12-month study concentrates on static corrosion experiments, comparing silicon carbide (SiC) corrosion rates in Li-rich and eutectic Li-Pb mixtures, and examining interactions between Mo and PbLi. The research seeks to determine the maximum operating temperatures for this material combination, acknowledging the balance between Li-Pb melting temperature and achievable TBR to unlock new opportunities for blanket design. This work lays the groundwork for future collaboration between Kyoto Fusioneering and Oak Ridge National Laboratory (ORNL). Findings could enhance the TBR of fusion blanket designs, and impact fusion energy development by confirming material compatibility between different Li-Pb mixtures with SiC and Mo at higher temperatures, including at temperatures relevant for very high-temperature blankets (1,000 °C), which could enable higher electricity conversion efficiencies and commercial applications using process heat.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Modeling of Transport Processes in Liquid-Metal Fusion Blankets: Past, Present, and Future

The successful development of robust breeding blanket systems will strongly rely on computational tools for predicting the complex behavior of the electrically conducting liquid-metal (LM) breeder flowing in the complex-shaped blanket ducts in the presence of a strong plasma-confining magnetic field, volumetric heating, and tritium generation. Associated transport processes involve magnetohydrodynamic (MHD) flows, heat transfer, corrosion, and tritium transport. This paper is an overview of past and present efforts in the development, application, and verification and validation (V&V) of such computational tools. As a result of the ongoing campaign on V&V of computer codes for LM blankets, the international fusion community has identified several candidates that promise to become real blanket design and analysis tools in the near future. Among them are HIMAG, MHD-UCAS, COMSOL Multiphysics, ANSYS FLUENT, ANSYS CFX, and OpenFOAM. The progress, over the last decade, in the application of such codes in blanket studies is tremendous. This is illustrated with two examples for a dual-coolant lead-lithium (DCLL) blanket: (1) integrated computer modeling for the recently designed DCLL blanket in the United States and (2) application of the code MHD-UCAS to the analysis of PbLi flows and heat transfer in a generic DCLL blanket prototype at high Hartmann (Ha ~ 10 4 ) and Grashof numbers (Gr ~ 10 12 ). Here, this paper also presents an approach to the development of a new integrated computational tool called the virtual dual-coolant lead-lithium (VDCLL) blanket, which elaborates the existing U.S. MHD code HIMAG.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Demonstration of RELAP5-3D for transient analysis of a dual coolant lead lithium fusion blanket concept

Our research aims to build a multiphysics framework for transient analysis of the Dual- Coolant Lead-Lithium (DCLL) blanket design of the proposed Fusion Nuclear Science Facility (FNSF) for all Onticipated Operational Occurrences (AOOs) using Reduced Order Models (ROMs). This paper presents an assessment of using RELAP5-3D for transient thermal-hydraulic blanket analysis of the DCLL design. Preliminary validation studies for the RELAP5-3D properties were conducted using a representative vertical flow loop model and a comparison study using experimental data from the thermal convection corrosion loops at ORNL. These studies inspected the thermal hydraulic response of Lead-Lithium Eutectic (PbLi) systems within the RELAP5-3D code; mainly pressure drop and heat transfer, as compared to analytical calculations and experimental data. The Magnetohydrodynamic (MHD) pressure drop effect for liquid metal under the effects of a uniform magnetic field was implemented into our models using an equivalent forms loss coefficient based on the phasic momentum equations. We verified our implementation of MHD pressure drop within RELAP5-3D using a systematic approach consisting of two studies. Models were developed to perform comparison studies based on experimental data from the Argonne Liquid metal EXperiment (ALEX) facility and the Magnetohydrodynamic PbLi Experiment (MaPLE) facility. Using the developed validation basis, we built a simplified DCLL channel model within RELAP5-3D, inspected MHD pressure drop and performed transient temperature analysis. The DCLL channel model used radial heating data from the International Thermonuclear Experimental Reactor (ITER) which was normalized to the FNSF design. Here, a representative startup transient was performed using a relative power curve from the initial ramp up of a long-term power pulse of the Tore Supra reactor. This study demonstrated that RELAP5-3D ROMs are capable of accurate transient analysis of the DCLL blanket. We have since utilized MCNP to develop heating profiles based on the FNSF design and are working toward implementing them into our DCLL model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Optimal Chloride Salt Mixture for a Fusion Blanket

Deuterium-tritium fusion reactors cannot operate for a significant period without a closed tritium fuel cycle, according to a recent National Academies of Sciences (NAS) report on bringing fusion reactors to the US electrical grid. This fact places breeder blankets as one of the foundational systems for self-sustained fusion reactor operation. The typical functional requirements for a breeder blanket system include producing tritium, absorbing kinetic energy, transporting thermal energy, and being environmentally attractive. State-of-the-art research on liquid blankets has converged to primarily focus on (LiF) 2 and BeF 2 (FLiBe) molten salts and a metallic eutectic of lead and lithium (PbLi), but “virtually all of the technologies related to the tritium fuel cycle are at a low technological readiness level”. This work sought to explore optimum blanket configurations as it aligned with the Oak Ridge National Laboratory (ORNL) FY 2023 Laboratory Directed Research and Development Program’s research priority of developing and expanding the current understanding of fusion blanket science and technology. This purpose of this work was to address ORNL research priorities and NAS recommendations by investigating novel liquid blanket materials that could provide self-sustaining operation and draw on experience from research on molten salts used for advanced fission reactors, concentrated solar, and thermal energy storage. The hypothesis when proposing this research was that there could be chloride-based blanket designs that can exceed the tritium breeding ratios of (FLiBe) molten salt blankets while reducing the use of Be (FLiBe), avoiding the generation of HF (FLiBe), and minimizing magnetohydrodynamic (MHD)-perturbed flow fields (PbLi). The fastest and most cost-effective path to deploying liquid fusion breeder blankets could be from maximizing the synergistic technological overlap between fusion, fission, concentrated solar, and thermal energy storage industries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

Abstract The limited tritium resources available for the first fusion power plants (FPPs) make fuel self-sufficiency and tritium inventory minimization leading issues in FPP design. This work builds on the model proposed by Abdou et al (2020 Nucl. Fusion 61 013001), which analyzed the fuel cycle (FC) of a DEMOnstration nuclear FPP-class FPP with a time-dependent system-level model. Here, we use a modified version of their model to analyze the FC of an Affordable, Robust, Compact (ARC)-class tokamak and two versions of a Spherical Tokamak for Energy Production (STEP)-class tokamak. The ARC-class tokamak breeds tritium in a 2LiF + BeF 2 liquid immersion blanket, while the STEP-class tokamak breeds tritium utilizing either a liquid-lithium blanket design or an encapsulated breeding blanket. A time-dependent system-level model is developed in Matlab Simulink ® to simulate the evolution of tritium flows and tritium inventories in the FC. The main goals of this work are to assess tritium self-sufficiency of the ARC- and STEP-class designs and to determine quantitative design requirements that can be used to analyze the adequacy of a proposed FC system. These design requirements are aimed at achieving a low tritium inventory doubling time ( t d ) and a low start-up inventory ( I s t a r t u p ) while keeping the required tritium breeding ratio (TBR r ) as low as possible. We also consider how improvements in FC technology and POs affect TBR r and I s t a r t u p . The model results show that TBR r for ARC- and STEP-class FPPs should be achievable if the tritium burn efficiency (TBE) reaches 0.5%–1% (TBR r < 1.2). This assumes significant, but attainable, improvements over current abilities. However, the model results indicate that an FPP must achieve ambitious performance targets, including FPP availability > 70%, tritium processing time < 4 h, and the implementation of direct internal recycling (DIR). If future research yields major improvements to achievable TBE, it may be possible to achieve tritium self-sufficiency while operating at lower availability and without implementing DIR.

Physics↗
DOE OSTI2025

New developments and verification of fusion blanket simulation capabilities in the MOOSE framework

Multiphysics modeling capabilities have a crucial role to play in the accelerated deployment of fusion energy. To that end, we developed new multiphysics fusion blanket simulation capabilities in the Multiphysics Object-Oriented Simulation Environment (MOOSE). Firstly, we expanded on the existing capabilities of the previously published work, by coupling 3D tritium transport modeling capabilities using the Tritium Migration Analysis Program, version 8 (TMAP8) to an existing tool including thermal hydraulics, fully three-dimensional (3D) heat transfer, and loosely coupled neutronics analysis. Secondly, we performed a thorough verification of the new capabilities and increased testing code coverage to meet MOOSE’s software quality standards. The MOOSE framework follows a strict software quality assurance plan to be Nuclear Quality Assurance, Level 1 compliant. The new multiphysics fusion blanket simulation capabilities are now held to the same standard. Thirdly, to demonstrate MOOSE’s new fusion blanket modeling capabilities, we performed a fully integrated, multiphysics simulation of a 3D solid ceramic breeder blanket design. This proof-of-concept simulation provides the temperature and tritium distribution across the blanket. In conclusion, the combined efforts towards software quality and the development of multiphysics coupling capabilities provide an effective and reliable framework for modeling solid ceramic fusion blankets using MOOSE.

modeling and simulation↗
DOE OSTI2025

Construction of 3D MHD pressure drop correlation and flow characterization in the contraction region of a fusion blanket manifold

Inlet and outlet manifolds are typical components of liquid metal (LM) blanket designs of a fusion power reactor to be used to distribute the LM flow into breeding channels and collect it at the exit of the blanket. High pressure loss in the magnetohydrodynamic (MHD) flows featuring abrupt geometrical changes is one of the main feasibility issues of such designs. Recently, optimization studies were conducted to construct 3D MHD pressure drop correlations for a LM flow in an electrically insulating manifold with gradual expansion. Here, the 3D computational approach developed in that study is applied to the outlet manifold featuring gradual contraction. A systematic analysis was performed with a total number of 135 flow cases computed with COMSOL Multiphysics for Hartmann numbers 1000 < Ha < 10,000, Reynolds numbers 100 < Re < 12,000, and contraction angles 45° < θ < 75° for a fixed contraction ratio of 4. The effects of Ha, Re and θ on the flow recirculation, development length and the total pressure drop were carefully examined. A linear regression analysis was used to determine the power rule of pressure drop coefficient k related to Ha and Re, demonstrating a good match with the Ludford layer theory. Eventually, a correlation for the 3D MHD pressure drop coefficient was constructed as a function of Ha, Re and θ. Further, the results were compared against the inlet manifold. It was found that the flow in the inlet manifold exhibits larger recirculation zones. In the investigated range of Ha, Re and θ, the pressure drop coefficient k of the LM MHD flow in the gradual contraction is only slightly lower (< 8 %) than that in the gradual expansion.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

New Developments and Verification of the Fusion Blanket Simulation Capabilities in the MOOSE Framework

This poster contains new developments and updates in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework related to fusion modeling and simulation, and forms the basis for an on-going effort in modeling whole fusion devices at Idaho National Laboratory using MOOSE. In this work, a slice of the midplane of a proposed ceramic breeder blanket design for the Fusion Nuclear Science Facility is simulated using a combination of neutronics, 1D system level thermal hydraulics for helium, 3D heat conduction, and tritium transport. Previous work focused on the engineering scale full assembly of a blanket sector, but the simulations here form the basis for software quality and verification efforts to increase confidence as the large scale simulations continue. This poster was presented at the Idaho National Laboratory Intern Poster Session 2023.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mitigated Dissimilar Mass Transfer with Aluminized RAFM Steels in Flowing PbLi at 650°C

To evaluate dissimilar material interactions in flowing eutectic Pb-16at. %Li between fusion-relevant materials and to establish a maximum operating temperature of future blanket designs, a series of thermal convection loop (TCL) experiments with flowing PbLi have been conducted. Following a 1000-h, 550°C to 650°C TCL experiment, a thin reaction product was observed on the surface of monolithic, high-purity SiC. To identify the source of the observed dissimilar material interaction and to understand the reaction kinetics, an identical 2000-h TCL exposure was conducted, but less reaction of the SiC specimen was observed. Characterization of the FeCrAlMo (alloy APMT) loop tubing and the PbLi-exposed austenitic stainless steel (SS) assembly parts suggested that the formation of the reaction layer on the surface of the PbLi-exposed SiC did not originate from the coated reduced activation ferritic martensitic (RAFM) steel, but from the dissolved loop materials (liquid metal exposed APMT tubing and type 316 SS fittings). Here, the results, after 2000 h, suggest that the dissimilar material reaction between the Al-coated RAFM steel and the SiC has very slow reaction kinetics within a temperature gradient of 550°C to 650°C.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗