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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 Anticipated 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 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 using 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. A representative startup transient was performed using a relative power curve based on current PWR power ramp limits. 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. [ABSTRACT FROM AUTHOR]

• We assessed the hydrogen migration and redistribution model in the BISON code for component-level tritium transport. • Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising. • A calibration method resulted in improved BISON predictions. We assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution.. The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising. [ABSTRACT FROM AUTHOR]

• State-of-the-art literature review of multiphysics modeling current status and needs. • Separate- and integral-effects experiments for the validation of the multiphysics analysis are needed for particular components of interest for the DCLL blanket. • Robust sensitivity and uncertainty (S/U) analysis must be integrated with the facility design to identify the significance of influencing parameters involved in tritium transport, along with quantification of their uncertainties towards output parameters. The development and demonstration of fusion power plant facilities require advancement of multiphysics modeling capabilities. These capabilities provide an integrated platform for design and analysis of a fusion engineering demonstration power plant. We present a state-of-the-art literature review of the status and needs of multiphysics modeling, including neutronics analysis, and thermal hydraulic/mechanical simulation tools in support of fusion powerplant system design and integration. In this paper we use the Fusion Nuclear Science Facility (FNSF) concept as an example system to assess future needs for advancement of multiphysics modeling capabilities. Specifically, we discuss the latest progress on R&D missions in the literature which are specific to the neutronics and thermal hydraulics design of the breeding blanket concept chosen for the FNSF, the Dual Coolant Lead Lithium (DCLL) blanket. We identify future R&D needs to be addressed through the development of multiphysics modeling capabilities that tie fusion neutronics to tritium breeding ratio (TBR), neutron wall loading (NWL), material damage and transmutation, shutdown dose, and safety assessments. The capabilities will support the design and licensing of a fusion engineering demonstration facility as well as a future engineering demonstration plant. We conclude that more integrated and prototypic tritium experiments must be developed to establish the database needed for the verification and validation of tritium transport simulations. Robust sensitivity and uncertainty (S/U) analysis must be integrated with the facility design to identify the significance of influencing parameters involved in tritium transport, along with quantification of their uncertainties towards output parameters. This enables design optimization of the DCLL blanket during the R&D phase for a fusion engineering demonstration facility. Additional separate- and integral-effects experiments for the validation of the multiphysics analysis is needed for particular components of interest for the DCLL blanket, such as the flow channel inserts (FCIs), along with multi-fidelity simulations which incorporate the magnetohydrodynamic (MHD) effects, and coupled analysis techniques that can capture the interactions between the thermal hydraulic effects and the tritium transport behavior. Simultaneously, the systematic S/U analysis for the multiphysics modeling framework must prove the reliability of the model to help define the technical gaps that need to be addressed by subsequent R&D activities for a fusion engineering demonstration facility. Lastly, S/U analysis of the uncertainties in design parameters for the evaluation of steady and transient safety analysis would be the final requirement that will elucidate the importance of these design parameters, as well as the quantification of their uncertainties on several identified key Figures of Merit (FoM) applicable to the safety analysis to provide guidance for the final design optimization of the DCLL. [ABSTRACT FROM AUTHOR]