Engineering topics

Hu, Rui

Publications and source records attributed to Hu, Rui.

At least 109 records · Page 6

DOE OSTI2019

Uncertainty Quantification on SAM Simulations of EBR-II Loss-of-Flow Tests

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for SFR, LFR, and MSR/FHR. In prior work, SAM has performed well in modeling the EBR-II benchmark based on the SHRT-17 and SHRT-45R protected and unprotected loss-of-flow experiments. However, underlying uncertainties in the modeling parameters need to be quantified to gain confidence in any best-estimate simulations. In this work, a Python coupling framework between Dakota and SAM has been developed to investigate these uncertainties, alongside code enhancements in SAM for user-flexibility in perturbing physical closure models and fluid equations of state. 25 significant input parameters covering boundary conditions, geometry, physical closure models, and equations of state were identified with their uncertainties defined from literature. Two SAM transient models for SHRT-17 and SHRT-45R were perturbed on these input parameters using Latin Hypercube Sampling and the uncertainties in predicted maximum fuel, cladding, and coolant temperatures were quantified as the output responses. Spearman rank correlation of input parameters to the predicted temperatures indicated that the most sensitive parameters include the initial core power, pipe wall drag, fuel & coolant conductivity, and peak channel flow area. Sensitivity of input parameters was found to differ between the two transients and core channel modeling strategies. The overall perturbations did not result in large uncertainties in the predicted temperatures, despite the efforts made to obtain accurate yet conservative input uncertainties.

Mui, Travis↗

DOE OSTI2019

CIAE-ANL CRADA for SFR SAFETY ANALYSIS: Final CRADA Report

The purpose of this project is joint verification and validation of system analysis code capabilities for Sodium-cooled Fast Reactor (SFR) design and licensing. As the researchers and code developers at the Argonne National Laboratory (ANL) and China Institute of Atomic Energy (CIAE) independently build their own SFR system analysis capabilities, this collaboration facilitated exchange of information on, and evaluation of alternative approaches for, modeling the key phenomena (such as decay heat removal and natural circulation) and the safety-related components (such as pumps, heat exchangers, and outlet plenum) common in SFR systems. The analysis at ANL and CIAE also applied their respective computational capabilities and methods (in SAS4A/SASSYS-1 computer code at ANL and FR-Sdaso computer code at CIAE) on jointly identified benchmark problems for verification, and publicly available integral tests for validation, of their results. ANL team also shared the expertise on the established software quality assurance practices and the techniques for component reliability estimates to support probabilistic risk assessments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

SAM User's Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2019

Initial Industry Collaborations of the Center of Excellence

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In particular the NEAMS program is supporting the development of novel thermal-hydraulic codes. The Center of Excellence for Thermal Fluids Application in Nuclear Energy, launched in 2018, has as its key goals to serve as a front door to industry. The Center of Excellence for Thermal Fluids Applications in Nuclear Energy has recently launched a program to start collaborative efforts between the laboratories and industry with the objective of stimulating cooperation and increasing adoption of T/H tools developed under NEAMS by the industry-at-large. In particular, two industry partners agreed to participate in pilot short-term collaborations aimed at demonstrating the value of NEAMS tools to their designs. These are BWXT and Kairos who proposed projects related to HTGR and FHR designs respectively. In this report we present the results of these collaborations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Point Kinetics and Reactivity Feedback Modeling in SAM

Over the past few years, the rapidly rising interest in advanced nuclear reactor technology has led to an increased need for the developments and applications of advanced computational tools for modeling and simulation. System Analysis Module (SAM) is being developed at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses [1]. It utilizes the object-oriented application framework MOOSE [2] to leverage the modern software environment and advanced numerical methods [3]. To meet this need, the capabilities of SAM are being extended to enable the transient modeling, analysis, and design of advanced nuclear reactor systems. The objective of this summary is to present the development of Point-Kinetics modeling and reactivity feedback mechanisms in SAM. This summary also presents the new capability for modeling the thermal expansion reactivity feedback by coupling SAM and an external thermomechanics module (Tensor Mechanics) from MOOSE.

Hu, Guojun↗

DOE OSTI2019

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

MODELING AND SIMULATIONS OF DETERIORATED TURBULENT HEAT TRANSFER IN WALL HEATED CYLINDRICAL TUBE

The paper discusses the modeling and simulations of Deteriorated Turbulent Heat Transfer (DTHT) for a wall-heated fluid flows, which can be observed in gas-cooled nuclear power reactors during Pressurized Conduction Cooldown (PCC) event due to loss of force circulation flow. The DTHT regime is defined as the deterioration of normal turbulent heat transport due to increase of acceleration and buoyancy forces. The Computational Fluid Dynamics (CFD) tools such as Nek5000 and STAR-CCM+ can help to analyze the DTHT phenomena in reactors for efficient thermal-fluid designs. 3D CFD non-isothermal modeling and simulations were performed in a wall-heated circular tube. The simulation results were verified with two different CFD tools, Nek5000 and STAR-CCM+, and validated with an experimental data. The predicted bulk temperatures were identical in both CFD tools, as expected. Good agreement between simulated results and measured data were obtained for wall temperatures along the tube axis using Nek5000. In STAR-CCM+, the under-predicted wall temperatures were mainly due to higher turbulence in the wall region. In STAR-CCM+, the predicted DTHT was over 48% at outlet when compared to inlet heat transfer values.

deteriorated turbulent heat transfer↗

DOE OSTI2019

Three-dimensional flow model development for thermal mixing and stratification modeling in reactor system transients analyses

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the mull-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Here, based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Thermal hydraulic model of the molten salt reactor experiment with the NEAMS system analysis module

System analysis codes have a long history of providing best-estimate and conservative safety analysis for both light water and advanced reactor technologies, including molten salt reactors. As interest continues to expand with advanced reactor concepts, system analysis codes will need revisions to accommodate the behavior of these technologies. Legacy system analysis codes will need to be updated to the latest numerical techniques to shorten execution time and increase the accuracy of results. One example of a modern system analysis code that already encompasses these characteristics is the System Analysis Module (SAM). One key objective of this paper was to review available information for system code modeling of the Molten Salt Reactor Experiment (MSRE) from sources in the open literature and collect the information from these open sources in one place for the first time. This supports the potential objective of developing an open specification for system code analysis for MSRE steady state and transients with and without reactor kinetics. Data from actual MSRE tests will serve as the basis for code-to-code comparison exercises, including the MSRE zero power physics tests, the fuel pump start-up and coast down tests, and the natural circulation transient. The objective is to produce a code-to-code benchmark with a standardized set of comparison problems, recognizing the limitations of the original data. To demonstrate an initial application of this objective and the usefulness of compiling this open data, two Molten Salt Reactor Experiment (MSRE)-related models were developed to evaluate SAM for liquid fueled molten salt reactors. One model was the SAM MSRE hydraulic mockup, which provided experimental data for pressure drop measurements. The second model was the complete MSRE primary loop. The MSRE primary loop model incorporated a fluoride salt fuel/coolant with heat transfer in both the core and heat exchanger. For both the hydraulic mockup and MSRE primary loop models, a holistic 1-D system description was built using open documentation, an open description that can be readily modified and applied for any system analysis code. SAM results for the pressure drop of the hydraulic mockup model were within 6% with measurements. Coolant temperatures for the primary loop model matched the expected axial change in temperature from historical calculations. Using alternative coolant properties obtained from the literature, corresponding to salts with different actinide contents, returned similar trends in core temperature profiles. A thermal hydraulic demonstration of a loss-of-flow transient showed the importance of coupling SAM thermal hydraulic analysis to neutronics. Here, this coupling is essential for simulating MSR transients with system analysis codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

lmprovements and Path Forward for Regulatory Acceptance of SAS4A/SASSYS-1

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The NRC’s approval of the license application is contingent on, among other things, satisfactory demonstration of the design basis and response to transient and accident scenarios using accepted codes and methods. This effort seeks to improve the regulatory acceptability of the SAS4A/SASSYS-1 advanced reactor design and safety analysis system software by identifying and addressing gaps in the SAS4A/SASSYS-1 documentation basis that support software qualification and dedication. The software’s unique capabilities to assess inherent fast spectrum feedback effects and passive safety features, which facilitate assessment of key safety metrics, position it as a key licensing tool in the field of sodium-cooled fast reactors (SFR).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Advanced Model Developments in SAM for Thermal Stratification Analysis during Reactor Transients

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Modeling and Simulations of Thermal-Fluid Phenomena Related to Pressurized Conduction Cooldown in High Temperature Gas Reactor

Under the support of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program, an effort is being pursued to support High Temperature Gas Reactors (HTGR) technology development and its modeling and simulation needs. There is a particular need for advanced modeling & simulation tools to predict thermal-fluid behavior in the nuclear reactor primary system—especially the core and lower and upper plena—during safety-related transients.To help accomplish this, we must provide confidence in the reliability and accuracy of the thermal fluid simulation tools being developed under the NEAMS program for HTGR applications. The benchmark simulations of the available experimental tests would help us in this effort.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Grain Refinement by Authigenic Inoculation Inherited from the Medium-Range Order Structure of Ni-Cr-W Superalloy

The combination of liquidus casting and thermal control solidification furnace was applied to obtain a fine-grained ingot. A rapid quenching method and x-ray diffraction measurement were used to investigate the effect of authigenic inoculation on grain refinement. The structure factor S(Q) of liquid Ni-Cr-W superalloy at 1400 °C (Liquidus temperature) and bright-field image of the microstructures quenched from 1400 °C have been measured by the high-temperature x-ray diffractometer and the transmission electron microscopy (TEM), respectively. The results show that a pre-peak exists on a S(Q) curve at the liquidus temperature. The clusters of atom in rapidly quenched microstructures obtained by isothermal heat treatment at 1400 °C were studied using TEM. Meanwhile, the effect of isothermal different temperatures on rapidly quenched microstructures was studied. The results also show that there are only the globular, equiaxed grains distributed in the solidification structure. These particles are inherited from the medium-range order structure, which is beneficial for grain refinement. The normalized work-hardening rate-strain curve indicates the work-hardening rate of fine grain is higher than that of conventional grain at the same temperature and the same deformation.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Preliminary SAM Assessment

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory (Argonne) for advanced non-LWR safety analysis. To assist NRC to assess SAM capabilities for advanced reactor safety analysis and licensing at the NRC, a series of verification and other standard tests (OSTs) are modeled in SAM and code simulation results are compared with available analytical results. This report documents the preliminary SAM assessment using a matrix of six test problems. Each test problem is further examined with different boundary conditions, system configurations, or modeling options. Although relatively simple, these tests cover the basic equation models, basic component models, and basic system level processes and phenomena that must be modeled for advanced reactor safety analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Ambient and building condition effects modeling of air-cooled natural circulation systems

Throughout the testing program at Natural convection Shutdown heat removal Test Facility (NSTF), strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under identical test procedures. Significant analysis efforts were conducted to gain understanding of these influences, and quantify the system response of the test facility. In this paper, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions, including wind and temperatures. Coefficients in the model were determined based on best fitting of the experimental data set. The predictive capability of the empirical model was demonstrated by applying it to new sets of experimental data. Finally, after implementing the empirical model in both system-level thermal fluids and computational fluid dynamics modeling of the facility, simulations from both approaches resulted in accurate predictions in system natural circulation flow rates.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2018

MicroRNA-30e-5p suppresses non-small cell lung cancer tumorigenesis by regulating USP22-mediated Sirt1/JAK/STAT3 signaling

MicroRNA-30e-5p (miR-30e-5p) is a tumor suppressor that is known to be downregulated in non-small cell lung cancer (NSCLC). However, how miR-30e-5p inhibits NSCLC tumorigenesis is not known. Ubiquitin-specific peptidase 22 (USP22) is upregulated in NSCLC and promotes tumorigenesis via a Sirt1-JAK-STAT3 pathway. In this study, we investigated whether miR-30e-5p inhibits tumor growth by targeting USP22 in NSCLC. Our results reveal that miR-30e-5p expression was correlated negatively with USP22 in NSCLC tissues. Luciferase reporter assays showed that miR-30e-5p negatively regulated USP22 expression by binding to a specific sequence in the 3UTR. MiR-30e-5p overexpression and USP22 knockdown significantly inhibited tumor growth in vivo and induced cell cycle arrest and apoptosis in NSCLC cells in vitro. The effects of miR-30e-5p inhibition were prevented by USP22 knockdown. MiR-30e-5p inhibited SIRT1 expression and increased expression of p53 and the phosphorylated form of STAT3 (pSTAT3). Furthermore, miR-30e-5p prevented USP22-mediated regulation of SIRT1, pSTAT3, and p53 expression. Taken together, these findings suggest that miR-30e-5p suppresses NSCLC tumorigenesis by downregulatingUSP22-mediated Sirt1/JAK/STAT3 signaling. Our study has identified miR-30e-5p as a potential therapeutic target for the treatment of NSCLC.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2017

Evolution of Σ3n CSL boundaries in Ni-Cr-Mo alloy during aging treatment

Highlights: • Σ3{sup n} boundary distribution characteristics are changed during aging treatment at 600 °C. • The increase in fractions of Σ9 and Σ27 boundaries depends on interaction between Σ3{sup n} boundaries. • The consumption of Σ3{sub c} boundaries is modeled with migration of random HABs. • Generation mechanism of Σ3{sub ic} boundaries differs from traditional models. - Abstract: The evolution of Σ3{sup n} (n = 1,2,3) boundaries in a Ni-Cr-Mo alloy during an aging treatment at 600 °C was investigated based on electron backscatter diffraction (EBSD) measurements. Incoherent Σ3 (Σ3{sub ic}) and coherent Σ3 (Σ3{sub c}) boundaries were identified using the single-section trace analysis method. The results show that the Σ3{sup n} boundary distribution characteristics are changed and that a Pt{sub 2}Mo-type Ni{sub 2}(Cr,Mo) superlattice phase precipitates during the aging treatment. Through the analysis and comparison of the average grain diameters measured under the different conditions, the migration behaviors of both the Σ3{sub ic} boundaries and random high-angle boundaries (HABs) are found. The migration of the Σ3{sub ic} boundaries provides the probability of interaction between Σ3{sup n} boundaries. This interaction is responsible for some different types of triple junctions, the increase in Σ9 and Σ27 boundary fractions and interrupted connectivity of the network of random HABs. However, the Σ3{sub c} boundaries' consumption caused by the migration of random HABs and the difficulty of forming new Σ3{sub c} boundaries lead to the decreased fraction of Σ3{sub c} boundaries. The further discussion about the increase in the fraction of Σ3{sub ic} boundaries indicates that the generation mechanism of the Σ3{sub ic} boundaries during the aging treatment differs from the models applied to recrystallization and subsequent grain growth processes.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

Hu, Rui↗