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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Accelerate Nuclear Research and Development by Reducing Time and Cost Spend in the Pre-conceptual Design Phase of Advanced Reactor Experiments

The design process of every new concept, such as advanced nuclear reactors or associated experiments, starts with the pre-conceptual design phase. In this phase, the viability of a wide range of design options needs to be assessed quickly, to understand the operating envelope and its feasibility. A variety of physics models (thermal-hydraulics, neutronics, mechanical design, etc.) has to be considered at this very first design stage and optimum component sizes and materials (e.g. heat exchangers, piping, turbomachinery, coolant type, etc.) have to be chosen for a given set of boundary conditions (e.g. heat source, heat sink, flow rate, etc.). Detailed solutions such as provided by high fidelity methods like computational fluid dynamics (CFD), Monte Carlo methods, etc. and even lower fidelity tools such as system or subchannel codes, etc. are usually not used during the pre-conceptual design due to the relatively long time needed to create input models, the computational time to obtain a solution and the lack of flexibility to quickly investigate different combinations of components, individual component sizes and material properties. High fidelity tools are usually only employed in the conceptual design and later phases once a base concept has been identified during the pre-conceptual design stage. The current practice during the pre-conceptual design stage is that analysts collect the needed equations, material properties, closure laws, etc. and create ad-hoc solutions form scratch for every new problem. There clearly is a lack of a flexible scoping tool that can be used during pre-conceptional design before higher fidelity tools (as described above) come into play. To reduce user errors in ad-hoc solutions and increase fidelity and efficiency, this project aims to investigate and develop a user-friendly scoping tool to address the thermal-hydraulic designing needs during preconceptual experiment design, i.e. Thermal-hydraulic Research Universal Scoping Tool (TRUST). The success of TRUST will provide the nuclear engineers with an easy-to-use and affordable calculator for early reactor system design and optimization.

42 ENGINEERING↗
DOE OSTI2022

Uncertainty Quantification in High-Low Dynamic System Coupling using RAVEN and TRANSFORM

This work demonstrates new functionality and applications stemming from the development of high-fidelity to low-fidelity (high-low) coupling for system simulations and to further explore the capabilities of the Risk Analysis Virtual Environment (RAVEN) in the performance of uncertainty quantification in this kind of high-low coupled system models. The work builds from previous work on high-low coupling that utilized COBRA-TF (CTF), the high-fidelity subchannel analysis code, with a low fidelity model built in ORNL’s TRANSFORM, the system analysis code, utilizing the Functional Mock-Up Interface (FMI). Steady-state and transient analysis examples using the high/low coupled models generated from CTF and TRANSFORM/FMI are investigated. The workflows for both steady-state and transient coupled simulations are described. A steady-state parameter sweep and uncertainty analysis of the primary flow rates and reactor power are demonstrated. Likewise, a transient pump trip and power ramp sensitivity studies are also demonstrated. This work elucidates some of the potential benefits and future needs of using RAVEN for high/low system coupling analysis of energy systems. It also shows some of the difficulties that can be encountered in coupling system simulations.

Williams, Wesley↗
DOE OSTI2020

EVALUATION OF PRESSURE DROP CORRELATIONS FOR THE WIRE-WRAPPED ROD BUNDLES

The accurate prediction of pressure drop in the fuel rod bundle is of importance for the design and thermal-hydraulic analysis of nuclear reactor system. Due to very complex geometry of wire-wrapped fuel rod bundle geometry for the Sodium Fast Reactor (SFR), a simple method using hydraulic (equivalent) diameter is not sufficient to predict the pressure drop of rod bundle accurately. For several decades, many experimental works have been performed to develop an accurate correlation for wire-wrapped fuel bundle friction factor. In these works, the friction factor correlation of wire-wrapped fuel bundle has been developed as a function of Reynolds number of the flow and the design parameters such as fuel pin diameter, wire diameter, rod pitch, and wire lead length. The Cheng and Todreas correlations are well-known and used in many subchannel analysis code, such as CADET, TRIO-U, MATRA-LMW, SE2-ANL, etc. Novendstern model, Rehme model, Baxi and Dalle Donne model are also well-known and widely referred. Computational Fluid Dynamics is another typical approach to predict the pressure drop of wire-wrapped fuel rod bundle, but it requires the model validation because the accuracy of pressure drop prediction by the CFD simulation depends on what turbulence model is employed. The present work aims at evaluating the existing pressure drop correlations for the wire-wrapped fuel rod bundle and evaluating the pressure drop of 217-pin rod bundle by the CFD analysis.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

CTF Improved Drag Model and Flow Regime Transition Criteria

The demand for accurate prediction of two-phase flow behavior in a boiling water reactor (BWR) requires a comprehensive understanding of flow regime, void fraction, heat transfer, and pressure drop. The CTF subchannel code, which is used for the Thermal/Hydraulic (T/H) solution in the Consortium for Advanced Simulation of Light Water Reactors (CASL)-developed Virtual Environment for Reactor Application (VERA) core simulator, is being further developed for BWR applications. In support of this goal, the present work highlights some of the two-phase closure model developments towards improving the CTF void fraction prediction, especially for subcooled boiling. The drift-flux approach has been well-developed for upward dispersed two-phase flows and proven to be accurate in predicting void fraction in bubbly and slug flow regimes. In this work, these kinematic constitutive relations for the drift-flux velocity have been implemented into CTF to describe the interfacial drag of bubbly flow as an alternative to the existing model for better void fraction prediction. The success of these constitutive relations also relies on a good flow regime map that accounts for flow conditions and channel geometry. A more reliable flow regime transition criteria that account for the flow condition has also been implemented in this study for modeling the flow regime transition criteria. The newly implemented models are shown to give improved void fraction predictions in comparison to experimental data.

Hizoum, Belgacem↗
DOE OSTI2016

DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Coupling CTF to MPACT for Improved Thermal-Hydraulic Feedback for Transients

As computational power and techniques continue to advance, reactor analysis codes are afforded the opportunity to employ a more detailed physics model. One area where this has manifested is with coupled neutron kinetics and thermal-hydraulics codes for transient analysis, in which the feedback is critical for accurate modeling. The Michigan Parallel Characteristics Transport (MPACT) code is a 3-D method of characteristics neutron transport code capable of generating sub-pin level power distributions being developed by the University of Michigan through the Consortium for Advanced Simulation of LWRs (CASL) program. Currently, MPACT handles thermal-hydraulic feedback with an internal 1-D radial heat transfer model assuming constant moderator temperature. While this relatively simple treatment of thermal-hydraulic feedback has been shown to accurately model transients where Doppler feedback is the dominant feedback mechanism, coupling MPACT to a detailed thermal-hydraulic code will greatly expand the range of transient problems that can be accurately modeled. COBRA-TF (CTF) is a 3-D two-phase flow, subchannel thermal-hydraulics code maintained by Penn State University through CASL. CTF is based upon the Semi-Implicit Method for Pressure-Linked Equations, and it is especially well suited to model heat transfer and fluid flow phenomena associated with a nuclear reactor through a wide range of conditions. We present an initial effort coupling MPACT to CTF for transients, which provides a wider range of transient problems that can be modeled with increased accuracy. We have successfully coupled the thermal-hydraulics code CTF to the neutronics code MPACT for transient calculations by implicit and Crank-Nicolson methods, which have been demonstrated with single pin and 3x3 configurations. This capability extends the ability of MPACT to accurately model transients through a wider range of conditions than the existing thermal-hydraulic feedback, which is severely limited by the assumption of constant moderator temperature and density. By employing CTF for the thermal- hydraulic feedback, the transient physics can be described more completely at minimal computational expense. (authors)

42 ENGINEERING↗
DOE OSTI2016

Spacer Grid and Mixing Vanes Effects Study on Two-phase Flow Using Bubble Tracking Simulations

In Light Water Reactor (LWR) cores, the nuclear fuel rods are arranged together as rod bundles in a triangular or square pattern. The fuel rod bundles are supported by the spacer grids with mixing vanes. An understanding of spacer grid and mixing vanes effect is important since it has a direct impact on the turbulence structure in the flow and the heat removal efficiency from the fuel rod surface. Although a considerable amount of literature has been published on the experimental investigation of spacer grid and mixing vanes effect, the measurements were mainly focused on single-phase flow. The influence of spacer grid and mixing vanes on two-phase flow is yet to be fully understood. Some preliminary research efforts have been invested in nuclear community. For example, Wheeler et al. experimentally studied the effects of spacer grid in air-water two-phase flow. Although mixing vanes are not considered, they presented interesting results of time-averaged two-phase flow quantities at upstream and downstream locations of the spacer grid. Over the past decades, the rapid development of High-Performance Computing (HPC) is allowing for the increasingly larger-scale high-fidelity simulations. Thanks to this favorable historical trend, Direct Numerical Simulation (DNS), integrated with Interface Tracking Methods (ITM), has been emerging as a valuable tool to complement and expand our ability to understand the two-phase flow phenomenon. Meanwhile, in order to make best use of the large-scale two-phase flow simulations, advanced data analysis techniques are being developed, as they enable the extraction of detailed parameters about individual bubble behavior. The level set ITM is utilized in our DNS flow solver, which makes use of a signed distance field. The gas-liquid interface is modeled by the zero level set. The level set method can be readily used to distinguish different phases based on the sign of corresponding level set value (e.g. the sign is positive in the liquid phase while negative in the gas phase.). However, the traditional level set approach is not able to collect calculated values and associate with specific bubbles when multiple (even thousands of) bubbles present in the simulations of two-phase flow. This drawback hinders the collection of useful bubble information, which can give us valuable insights about bubbly flows. For example, how the different local fluid conditions could affect bubble interfacial forces, bubble deformation level, and eventually the bubble distribution throughout the whole domain. A bubble tracking methodology has been developed to address the drawbacks and improve the data collection efficiency of level set based two-phase flow simulations. In the present work, a single Pressurized Water Reactor (PWR) subchannel with reduced size spacer grid and mixing vanes is chosen as the computational domain. The springs and dimples of the spacer grid are not represented for simplicity in the present investigation. The bubble tracking methodology is applied to the simulations of two-phase bubbly flow through spacer grid and mixing vanes region. In the intended simulation, the liquid turbulence will be fully resolved by DNS while the two-phase behaviors are captured by level set ITM. Bubble tracking data are to be processed to investigate the spacer grid and mixing vanes effect on two-phase bubbly flow. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Improvement of Wall Drag Modeling Capabilities of CTF

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. It solves discretized forms of nine coupled equations for the mass, momentum, and energy equations for the liquid and vapor fields; the mass and momentum equation for the droplet field; and the energy equation for solid structures. This work focuses on improving the wall drag modeling capabilities of CTF by adding new friction models that account for surface roughness. Prior to this work, CTF used simple correlations that assumed that surfaces were smooth. In reality, different materials in the reactor have varying degrees of roughness that act to increase drag and frictional pressure drop. The frictional pressure drop is important, as it partly defines the driving head necessary for achieving certain flow conditions. This is particularly important in passively cooled systems that rely on natural circulation. Therefore, the effects of surface roughness cannot be neglected, as it affects flow rate and distribution, heat transfer and turbulence. (authors)

42 ENGINEERING↗
DOE OSTI2016

Uncertainty Analysis of CTF for the Uncertainty Analysis in Modeling Benchmark

This work describes the results of a quantitative sensitivity and uncertainty analysis of the thermal hydraulic subchannel code, Coolant Boiling in Rod Arrays - Three Field (COBRA-TF). Four steady state cases from Phase II, Exercise 3 of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Uncertainty Analysis in Modeling (OECD/NEA UAM) Benchmark are analyzed using the statistical analysis tool, Design Analysis Kit for Optimization and Tera-scale Applications (Dakota). This study supports the larger goal of the UAM Benchmark, which is to assess the uncertainty inherent in nuclear codes in a straightforward framework. The benchmark first focuses on assessing the uncertainty in stand-alone codes and then uses these results to propagate those uncertainties to coupled calculations. The work presented here adds to the existing results for this benchmark. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Two-phase Flow Regime Transition Study using Level-Set Method

In Boiling Water Reactors (BWRs) the coolant exiting the core often has void fractions as high as 50%, resulting in high gas volume fraction two-phase flow regimes, such as annular or churn-turbulent flow. As the coolant boils along the subchannels it will experience flow regime transitions from bubbly to slug, slug to churn, and, finally, churn to annular. While these transition points are known to occur as a higher void develops along the flow path, the driving physics of the transition phenomenon is not well understood. Currently, experiment based correlations are the most reliable method for predicting where in the core the flow regime transitions will occur. However, experimentally, it's ever difficult to observe the details of the two-phase flow at prototypical pressures/temperatures, or, as with high fidelity experiments, the conditions differ from those in a reactor. Being able to predict these flow regime transitions is critical to safe and efficient reactor operations as well as maintaining fuel rod integrity by avoiding dryout. The characterization of flow regimes has been extensively studied, producing a wide set of flow regime maps based on characteristic parameters such as vapor and liquid superficial velocity, time dependent pressure drop, interfacial area density, etc. While these studies are useful in predicting the flow regime for certain geometry and flow conditions, they do not provide a universal predictive capability for any geometry under any conditions. Additionally, it is difficult to achieve reactor realistic pressure and temperature conditions in experiments. Direct Numerical Simulation (DNS) coupled with Interface Tracking Methods (ITMs) becomes advantageous in this case because the simulations are not limited by the material properties or to the real world conditions. DNS allows for analysis of flows that would be impossible to achieve, or observe in detail, experimentally. The finite element based DNS code, PHASTA (parallel, hierarchic, higher-order accurate, adaptive, stabilized transient analysis flow solver), used in this work utilizes the level set ITM and has been previously shown to properly resolve bubbly, slug, and annular flows, but not transitions between the two-phase flow regimes. The present work uses PHASTA to analyze flow in a pipe for various void fractions and superficial velocities. Flow regime transitions were induced to demonstrate the transient behavior of this phenomenon. With the unique advanced analysis capabilities developed in PHASTA that allow for individual bubble tracking, it becomes possible to extract new localized information about the flow behavior that is currently not attainable from experiments. The potential of this capability is to uncover the underlying physics of the flow regime transition phenomenon. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity Analysis of Passive Heat Removal Capability of FHR Based on Efficient Surrogate Model and GenFlow

Code Scaling, Applicability and Uncertainty Evaluation Methodology (CSAU) is used for quantifying reactor safety margins and works as an alternative method for licensing nuclear power plant. In the framework of CSAU, sensitivity and uncertainty analysis (SA and UA) should be performed. However, for large-scale complicated thermal hydraulics analysis simulated by code (e.g. RELAP5, CFD and subchannel analysis, the sensitivity and uncertainty analysis is really time-consuming. Furthermore, sometimes it is not feasible to perform SA and UA. Specifically, the computation time for Monte Carlo sampling or advanced LHS sampling can be neglected, but the computation time for thermal hydraulics code makes up main part of time during SA and UA process. In order to accelerate the speed of SA and UA, thermal hydraulics code can be improved based on advanced mathematic algorithm or using parallel method. However, the development for new mathematic method needs long time and strict verification and validation process is needed for a new code, which usually takes a couple of years or even longer time to have an approved code that can be used in nuclear industry. An alternative way to improve the speed of SA and UA is using surrogate model. The surrogate model is derived from the input as well as output data of the original code, assuming the code as a black box. The most common method is using response surface method (RSM) as a surrogate for the code, the input parameters are then sampled from their assumed distribution and the corresponding value of the response surface function is calculated. However, a couple of input parameters need hundreds or even thousands of original code calculations. The large number of runs is an obvious disadvantage of this method. Besides, the second disadvantage is an appropriate response surface function must be selected (e.g. quadratic, cubic, quartic, with or without cross product terms, or using a non-polynomial form) to avoid the overfit and underfit problem. This process is time consuming and expensive. Polynomial chaos expansion method is a reasonably fast and attractive alternative, stochastic model. The chaos expansion provides an efficient high-order method of including non-linear effects in stochastic analysis. Usually, the information about the model parameters and their probability distribution is limited and the dependence between them is not known. However, if arbitrary polynomial chaos is used, instead of assuming probability distribution for input parameters which may introduces additional subjectivity into the analysis, the available information can directly and most purely be used in stochastic analysis based on arbitrary data distribution. Thus, it is not necessary to know the exact probability density functions to perform SA and UA based on arbitrary polynomial chaos (aPC). The aPC has been applied for uncertainty quantification in many fields (e.g. in solid mechanics, fluid flows, thermal sciences), and become widely used in nuclear industry recently. Mk1 PB-FHR is a small modular fluoride-salt-cooled, high-temperature reactors (FHRs) that is a concept design by University of California, Berkeley (UCB). It has inherently safe features like natural-circulation-driven systems (e.g. direct reactor auxiliary cooling system (DRACS)) to passively remove decay heat from the reactor core during design basis and beyond design. It is necessary to perform SA and UA, since the licensing process of this new kind of reactor is under way and there are many design parameters that have not been fixed. Current nuclear commercial codes, such as RELAP5, have mainly been developed for light water reactors, which require the use of comprehensive two-phase flow models. This capability is not required to simulate single-phase flow phenomena in FHRs, whose baseline coolant, flibe, remains liquid up to 1430 deg. C at atmospheric pressure. The FHR advanced natural circulation analysis (FANCY) code has been developed specifically for FHR technology modeling and validation work has been performed by using CIET data in UCB. The code has been renamed General Flow (GenFlow) because it has been further developed to have the capability to model complicated pipework system based on object-oriented concept for user friendly purpose. Besides, it can be used to simulate both the force circulation as well as natural circulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Molten Salt Reactor Simulations using MPACT-CTF

The US Department of Energy and industry have shown significant interest in advanced reactors, as evidenced by over 1.3 billion dollars of private investments in companies to develop advanced reactor concepts. More than seven of these advanced reactor concepts use molten salt reactor (MSR) technology. Based on this interest in MSR technology, ORNL has invested in the development of the MSR simulation capability, extending some capabilities in VERA to model this high-profile reactor concept. In this paper, the MPACT deterministic neutron transport capability and the COBRA Two-Fluid (CTF) subchannel capability are adapted to handle liquid fuel MSRs. Specifically, this work focuses on modifications to the depletion algorithms to account for isotopic mixing, continuous feed, and continuous removal. Models will be added to account for drift of precursors and decay heat emitters as the liquid fuel flows through the reactor coolant system. Modifications to CTF, including the addition of molten salt properties, are also addressed. Finally, the coupled MPACT-CTF capability is used to simulate the 30-year operation of an MSR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Validation of CTF Droplet Entrainment and Annular/Mist Closure Models using Riso Steam/Water Experiments

This document summarizes the work to validate the droplet entrainment and de-entrainment models, as well as two-phase closure models in the CTF (Coolant Boiling in Rod Arrays - Two Fluids) code by comparison with experimental data obtained at Risoe National Laboratory. CTF is the modernized version of the subchannel code, COBRA-TF, being jointly developed by Oak Ridge National Laboratory and North Carolina State University for applications in the Consortium for Advanced Simulation of Light Water Reactors (CASL). The Risoe data included a series of more than 250 steam/water experiments that were performed in both tube and annulus geometries over a range of various pressures and outlet qualities. Experimental conditions were set so that the majority of cases was in the annular/mist flow regime. Measurements included liquid film flow rate, droplet flow rate, film thickness, and two-phase pressure drop. CTF was used to model 180 of the tubular geometry cases, matching experimental geometry, outlet pressure, and outlet flow quality to experimental values. CTF results were compared to the experimental data at the outlet of the test section in terms of vapor flow fractions, and entrained liquid flow fractions, and pressure drop per unit length.

42 ENGINEERING↗
DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Influence of Axial Power Distributions on Rod-to-wall Gap

In a scaled-down (3 x 3, 4 x 4, 5 x 5 or 6 x 6) rod bundle thermal hydraulic experiment, channel walls were often applied to surround the experimental rod bundle and provide a hydraulic flow boundary for the experiments. However, in the actual PWR (Pressurized Water Reactor), there are no walls surrounding the rod bundle fuel assembly. These non-representative walls can lead to non-prototypical conditions in the rod bundle experiments. There are several researches about the wall effects in rod bundle experiments. Based on hydraulic consideration, Zang developed a wall effect factor by subchannel analysis method to investigate the wall effects of different scale rod bundles. He did not pay attention to the wall effects in thermal aspects. Li studied the wall effects of different scale rod bundles of tight arrangement spiral-fin fuel assembly by CFD. The existing researches mostly focus on the wall effects of different scale rod bundles. There are very few, if any at all, studies focus on how to avoid or minimize wall effects through proper experimental design by adjusting the rod-to-wall gap (the gap between the channel wall and outer rod wall) for rod bundles of various scales. In addition, the axial power distribution is different in different periods of the life cycle in the PWR core. For example, it usually is outlet peak power distribution in the axial direction at the beginning of life span. However, it may turn to the inlet peak power distribution when the control rods are pulled out from the high burn up of fuel in the reactor core. Both flow dynamics and heat transfer mechanism change depending on various axial power distributions. Subsequently, it is important to study the impacts of the wall effects on flow and heat transfer with different axial power distributions in rod bundle experiments. That is, in the evaluation of rod-to-wall gap in rod bundle experiments, the influence of axial power distributions should be examined. In this paper, the rod-to-wall gap was calculated in axially uniform and non-uniform power, including inlet peak, cosine, and exit peak heat flux distributions. The rod-to-wall gap calculation results were then compared with respect to two different criterion: the average fluid temperature and the maximum wall temperature criterion, to study the influence of axial heat flux profile on radial thermal distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Large Eddy Simulation of turbulent flow in rod bundle with spacer grids

Pressurized water nuclear reactor fuel assemblies are made of fuel rods held by spacer grids, which are the core components of the nuclear power plants. Hot fuel rods are cooled by the coolant water flow, and the spacer grids with mixing vanes distort the mainstream, producing the swirling flow which can enhance the heat transfer to improve CHF (Critical Heat Flux). Heat transfer between fuel rods and coolant is a crucial part of thermal-hydraulic analysis. For the purpose to improve the thermal hydraulic performance of the fuel assemblies, many experiments have been conducted to understand the flow structures and three-dimensional behaviors of the turbulence in the rod bundle sub-channels and then to optimize the design of the spacer grid. In the experimental work, which requires lots of efforts to get reliable data, LDA (Laser Doppler Anemometry) and PIV (Particle Image Velocimetry) are generally used to measure the velocity profile and turbulent intensity in the sub-channels. Computational Fluid Dynamics (CFD) is another useful way to study the characteristics of the detailed flow condition and heat transfer phenomena in the subchannels. The main issue of the single-phase CFD is to accurately predict the flow and turbulence characteristics. RANS method is generally used to simulate the induced lateral velocity and the heat transfer process. With the development of computer technology, some more precise but previously time-consuming turbulent models, such as LES (Large Eddy Simulation), are gradually used for the simulation of rod bundle flow with current advanced CPU resources. Ikeno and Kajishima explained the decay of swirling turbulent flow in the rod bundle using LES method. Benhamadouche acquired the turbulence induced forces on the central rod of the rod bundle through performing the LES of the flow downstream the spacer grid. Chang and Tavoularis conducted hybrid simulations with SAS SST and LES of a 5 x 5 rod bundle, and several vortices were identified in the vicinity of the vanes and the generation mechanisms, evolution progress and influence on the velocities of these vortices were clarified. In the paper we present CFD simulations of turbulent flow in rod bundle sub-channels with spacer grids of two different experiments. One is MATiS-H benchmark experiment with scaled spacer grids, which is conducted by KAERI, and is used as a CFD benchmark test, and the other is a 5 x 5 rod bundle with typical spacer grids using PIV technology, with the cooperation of Harbin Engineering University. The main objectives of the calculations are to validate LES methodology by the benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Study of effect of PWR cold leg temperature gradient on reactor core condition

Effects of temperature and flow gradients in Westinghouse designed three-loop Pressurized Water Reactor (PWR) cold legs, the piping between the main coolant pump and the reactor vessel, were evaluated using Computational Fluid Dynamics (CFD) code STAR CCM+ and coupled neutronic and thermal-hydraulic (T/H) code system VERA. In the parametric study, several symmetric and asymmetric temperature gradients that were significantly larger than those observed from plant measurements were applied to the cold leg inlets for comparison with the base case without any temperature gradient. A CFD model using the STAR-CCM+ code was developed for a portion of the RCS region between the Reactor Coolant Pump (RCP) and the core inlet based on previously validated modeling approach. The CFD simulation results were processed for the temperature and flow rate distributions at the core inlet as input to the VERA calculations. The VERA code system consists of COBRA-TF (CTF) for thermal-hydraulics, MPACT for reactor physics and neutron transport, and ORIGEN for isotopic depletion. The VERA model was for depletion calculations of a high-burnup loading pattern with the reactor core in pin-by-pin and subchannel resolution. The results of the study indicate that the postulated temperature gradients within the PWR cold legs do not result in any significant changes in the core inlet temperature distributions and the core power distributions during the reactor operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Comparisons of Nodal diffusion and whole-core transport methods for multiple cycles of a small Light water reacto

As part of a cooperative research and development agreement between the US department of Energy's Oak Ridge National Laboratory and Holtec International subsidiary SMR LLC, core simulator models of the SMR-160, a light water-cooled and -moderated, natural circulation, small modular reactor, were developed and compared. The model results indicate pin-by-pin power histories, critical boron concentration, and other key performance indicators useful for developing multi-cycle core loading patterns. One model uses industry standard nodal methods with a simple thermal-hydraulics solution, and the other uses an advanced deterministic neutronics solver with a coupled subchannel calculation. The goals of the comparison are to provide the reactor designer with a detailed and coupled core physics reference solution, inform the reactor designer of nodal model best practices based on the code-to-code comparison, and set up a workflow for performing subsequent detailed core simulator calculations. The results of this work will be used in the development of coupled neutronic, thermal hydraulic, and fuel performance models for advanced analysis of the SMR-160. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗