Advanced Reactor Technologies

Continued development of system analysis codes has resulted in the recovery of conservatisms originally imposed on nuclear power reactors, allowing for an increase in the capacity of commercial nuclear reactors. These codes also play an instrumental role in the design and certification of new reactor systems. With the increased demand for passive natural circulation and gravity driven cooling options, these codes are met with the new challenge of simulating low pressure, low flow conditions. The objective of this work is to demonstrate the effectiveness of the widely used RELAP5/MOD3.3 code to simulate boiling, condensing and flashing flows under such conditions. Two-phase flow data in an internally heated vertical annulus with inner diameter of 19.1 mm and outer diameter of 38.1 mm is utilized for validation of the RELAP5/MOD3.3. The code calculation of pressure, temperature, void fraction, interfacial area concentration, and void weighted gas velocity along the 4.5 m test section is compared with data at five axial locations. In the 2.8 m heated section of the channel the code predictions compare favorably in general, although the error does increase at low system pressure. Beyond the heated length, code predictions of condensation and flashing show more noticeable disagreement along the 1.7 m unheated section. Condensation is consistently under-predicted. Flashing varies from relatively good agreement to complete failure, depending on the conditions at the exit of the heated section. User options related to boiling and condensation are also assessed, and shown to have marginal improvements in some conditions. In general the code consistently predicts the point of net vapor generation too soon along the heated length at low mass flux, over predicts the void fraction at the end of the heated length, and has large scatter in void fraction agreement at the end of the channel.

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The work presents validation of the TRAC/RELAP Advanced Computational Engine (TRACE) code for natural circulation two-phase flow in a vertical annulus. Natural circulation experiments were recently conducted for a vertical internally heated annulus at the Multiphase Thermo-Fluid Dynamics Laboratory at the University of Illinois. The experimental matrix consists of 107 experiments with system pressure in the range of 145 to 950 kPa and heat flux up to 275 kW/m2. Void fraction, gas velocity, and interfacial area concentration were measured in five axial locations along the test section with six measurements of bulk liquid temperature and pressure. To validate the capability of the TRACE code under natural circulation flow conditions, a complete model of the experimental facility was created and validated using forced convection and single-phase natural circulation data.

Sensitivity and uncertainty quantification were performed. The sensitivity to important simulation parameters was studied using Sobol’s variance decomposition and the Morris screening method. The sensitivity of boundary conditions on void fraction measurement was investigated. The sensitivity study has shown significant differences in model sensitivity between different experimental conditions. With heat flux being the most influential parameter for high-pressure cases without flashing and pressure, temperature and heat flux have a combined strong effect in the case of low-pressure experiments when flashing occurs. Additionally, higher uncertainty in void fraction prediction was observed for experimental conditions at low pressure with flashing.

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The validation of system analysis codes for nuclear reactor systems is required for the development and application of these computational tools. Designed as a comprehensive system analysis code for advanced nuclear reactors, SAS4A/SASSYS-1 requires validation of its physics model for capturing single-phase natural circulation behavior. To support the validation of SAS4A/SASSYS-1, high-precision experiments are performed capturing steady-state single-phase natural circulation on a scaled facility with comprehensive instrumentation. Dedicated tests are performed quantifying the critical modeling parameters, and a single-phase natural circulation benchmark dataset is obtained with well-documented uncertainty and comprehensive facility description. The validation is then performed against the dataset examining the capability of SAS4A/SASSYS-1 in simulating steady-state single-phase natural circulation. The experimental facility is modeled in the candidate code. Solution verification is performed using Richardson-extrapolation-based estimators which quantify and restrict numerical errors from discretization. Input uncertainty provided by the benchmark dataset is forward propagated through the candidate code, quantifying the output uncertainty in a Monte Carlo approach. The composition of the output uncertainty is also quantified through a variance-based sensitivity analysis. With the uncertainty quantified for each individual condition, a detailed comparison between the simulation results and experimental data is performed covering the whole dataset. The results show consistent agreement for all primary parameters. The current validation activity provides a valuable benchmark dataset for the validation of system analysis codes in capturing single-phase natural circulation and demonstrates satisfactory prediction capability of SAS4A/SASSYS-1 for steady-state single-phase natural circulation.

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Natural circulation is employed in new designs of light water reactors to enhance passive safety by maintaining flow and heat removal without pumps. Under low-pressure and low-flow-rate conditions, natural circulation is susceptible to two-phase instabilities leading to undesirable flow oscillations and operational difficulties. Flashing instability is one of the most widely reported low-pressure natural circulation instabilities, related to saturated vaporization triggered by a hydrostatic pressure drop in an adiabatic riser above a heated section. While existing studies have reported flashing instability experiments, modeling, and simulations including successes in matching numerical results and experimental data, solid yet clear analytical explanations for many of its qualitative features are still rare. To enhance the physical understanding beyond stability boundary prediction, the current work develops, validates, and analyzes a linear stability model of flashing instability. This model adopts a one-dimensional Drift-Flux Model simplified by physical assumptions and approximations, and it includes optional component models to match an actual facility for validation. Stability tests are performed on a 5-m-tall natural circulation loop, providing comprehensive benchmark data covering stability boundaries, one-dimensional transient signals, and periodic mean waveforms from local measurements. Validation confirms acceptable predictions of steady states, stability boundaries, and oscillation periods. The tractable model formulation leads to a closed-form characteristic function facilitating analytical manipulations and physical interpretations, based on which dominant pressure drop responses to inlet flow rate are extracted. The major instability mechanism is identified as a strong response of the two-phase driving force to the inlet flow rate that is delayed by enthalpy transportation through a long single-phase distance and can become an overwhelmingly destabilizing positive feedback under low-frequency perturbations. Experimentally reported qualitative features, including stability changes, timescale relations, and oscillation patterns, are analytically predicted and physically explained with clarity. In general, this study enriches experimental resources of flashing instability with a comprehensive dataset and provides a simple yet realistic analytical basis for physically understanding flashing instability beyond predicting stability boundaries.

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Low-pressure two-phase flow instabilities can potentially challenge start-up transients of water-cooled nuclear reactors. Predicting oscillations caused by flow instability is therefore paramount to reactor safety. While many thermal hydraulics system codes are developed as general-purpose tools, their performance is not guaranteed outside their optimized application ranges. The current study therefore performs validations of two thermal hydraulics system analysis codes, the Adaptive SYStem Thermal-hydraulics Version 3 (ASYST VER3) and Reactor Excursion and Leak Analysis Program MOD3.3 (RELAP5/MOD3.3), in reproducing oscillatory two-phase flow induced by low-pressure flashing instability. Benchmark conditions are selected from a novel dataset resolving void fraction with radial resolution and quantifying periodic behavior with ensemble averaging. Focusing on the prediction of transient two-phase phenomena rather than the determination of stability, validations are conducted by simulating a single channel under prescribed periodic boundary conditions. ASYST exhibits a systematic underprediction of void fraction in an adiabatic chimney downstream of a heated section. The prominent causal discrepancy is identified as a lost travelling void wave due to overpredicted condensation. RELAP5 noticeably overpredicts subcooled boiling and underpredicts flashing, which is consistent with existing validations against steady-state separate-effect tests under low pressure. This degraded code performance also suggests that in low-pressure transients prone to flashing instability, the confidence in RELAP5 derived from existing validations against integral-effect tests shall be carefully limited to validated capabilities in integral systems. In general, the current study fills the previous gap of knowledge about the performance of ASYST and RELAP5 in predicting detailed transient two-phase phenomena in low-pressure flashing-induced oscillations. The identified code defects reveal future directions for code improvements eventually towards reliable application of ASYST and RELAP5 under low pressure beyond their original calibration.

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Two thermal hydraulics system codes, ASYST VER3 and RELAP5 MOD3.3, are assessed for their prediction of stability and limit cycles of low-pressure natural circulation with potential flashing instability. The benchmark conditions are from a nearly five-meter-tall water loop whose experiments have generated a published dataset covering both stability and limit-cycle oscillations. These conditions are simulated by a model of the entire system, and the target asymptotic behaviors are approached through sufficiently long transients under prescribed operational settings. ASYST is found not conservative for stability prediction, in the sense that the unstable operational range is underpredicted with discrepancy mainly in the high-subcooling stability boundary. Qualitative flow changes across the island of instability are however simulated satisfactorily, and for conditions correctly predicted as unstable, reasonable prediction is achieved for the oscillation period, time-averaged flow rate, and peak flow rate. Further comparison against experimental void fraction confirms that ASYST simulates physical two-phase behaviors in the limit cycles. RELAP5 predicts a much more stable system whose island of instability is significantly smaller than that of reality. Its simulated flow oscillations also noticeably deviate from the typical patterns of flashing instability, resembling sinusoidal waves following periods different from the experimental measurement. Underprediction of flashing and overproduction of subcooled boiling are identified as potential major defects in RELAP5, which calls for future model calibration and further investigation. In general, the current assessment contributes to the awareness of potential uncertainties when adopting ASYST and RELAP5 for predicting flashing instability and its induced transients.

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In the quest for a sustainable and climate-resilient future, significant interest is found in advanced reactor technologies in recent years including the liquid-fueled molten salt reactors (MSRs). Liquid-fueled MSRs stand out due to their unique characteristics, especially the potential online removal of fission products such as xenon-135. The removal of xenon can enhance fuel utilization and make the reactor more adaptable to load-following operations. However, the development of xenon removal system for MSRs requires improved understanding of xenon behavior in liquid-fueled MSRs. In this study, a system level Simulink model for liquid-fueled MSRs is developed and then adapted to study the xenon behavior in the Molten Salt Reactor Experiment (MSRE). The steady-state and transient xenon behavior in the MSRE is simulated and compared with available data and existing models in the literature. Good agreement is found between the simulation and the experiment under the same set of model parameters. The importance of xenon transfer between the core graphite, circulating bubbles, and fuel salt is highlighted. The Simulink model can be easily extended for future development of a xenon removal system in commercial scale liquid-fueled MSRs.

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The axial-flow centrifugal bubble separator designed for the gaseous fission product removal system in liquid-fueled molten salt reactors is simulated using the Eulerian two-fluid model coupled with the Adaptive Multiple Size Group method to account for the significant coalescence and breakup in the bubble separator. The behavior of the gas core in the bubble separator is mimicked by the symmetric interfacial area concentration model. The separator efficiency, local velocity, and pressure profiles at various conditions are compared with experimental data. Good agreement is found between the experiment and the simulation for the separator efficiency. With the coalescence and breakup being accounted for, the effect of the inlet void fraction on the separator efficiency is correctly captured. For the local pressure and velocity profiles, the agreement is only quantitative due to the simplifications on the geometry and potential limitations of the current computational fluid dynamics models. As good agreement is found for the separator efficiency, the sensitivity study is performed for various operational and design parameters with further simplified two-dimensional axisymmetric simulation.

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As one of the Gen-IV advanced reactors, liquid-fueled MSRs possess several distinctive features, one of which is the inclusion of a xenon removal system. The xenon removal system could enhance fuel utilization and remove a major barrier for load-following operation. In this study, a system level Simulink model of the Molten Salt Demonstration Reactor is developed. The model is used to perform a design analysis of the xenon removal system. Preliminary but quantitative requirements from fuel utilization and load following operation perspectives are applied. A reference system design that could satisfy the proposed requirements is obtained from the simulations. Moreover, a preliminary cost-benefit analysis is conducted using the reference design. It is found that the xenon removal system mostly benefits from load-following operation, and the major cost lies in the processing and storage of the fission-product containing gas in the off-gas system.

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Liquid-fueled molten-salt reactors have dynamic features that distinguish them from solid-fueled reactors, such that conventional system-analysis codes are not directly applicable. In this study, a coupled dynamic model of the Molten-Salt Reactor Experiment (MSRE) is developed. The coupled model includes the neutronics and single-phase thermal-hydraulics modeling of the reactor and validated xenon-transport modeling from previous studies. The coupled dynamic model is validated against the frequency-response and transient-response data from the MSRE. The validated model is then applied to study the effects of xenon and void transport on the dynamic behaviors of the reactor. Plant responses during the unique initiating events such as off-gas system blockages and loss of circulating voids are investigated.

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