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Title: Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

Abstract

The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have beenmore » completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.« less

Authors:
 [1];  [2];  [1];  [2];  [2]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1335356
Report Number(s):
ORNL/TM-2016/612
NT0512000; NENT026; TRN: US1700827
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 97 MATHEMATICS AND COMPUTING; COMPUTERIZED SIMULATION; STEAM GENERATORS; FINITE ELEMENT METHOD; BOILING; PRIMARY COOLANT CIRCUITS; TUBES; FLUID MECHANICS; EVALUATION; SUPPORTS; PRESSURE DROP; MECHANICAL VIBRATIONS; BENCHMARKS; CORRELATIONS; DESIGN; MITIGATION; PERFORMANCE; HELICAL CONFIGURATION; TURBULENT FLOW; TWO-PHASE FLOW

Citation Formats

Pointer, William David, Shaver, Dillon, Liu, Yang, Vegendla, Prasad, and Tentner, Adrian. Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator. United States: N. p., 2016. Web. doi:10.2172/1335356.
Pointer, William David, Shaver, Dillon, Liu, Yang, Vegendla, Prasad, & Tentner, Adrian. Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator. United States. doi:10.2172/1335356.
Pointer, William David, Shaver, Dillon, Liu, Yang, Vegendla, Prasad, and Tentner, Adrian. 2016. "Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator". United States. doi:10.2172/1335356. https://www.osti.gov/servlets/purl/1335356.
@article{osti_1335356,
title = {Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator},
author = {Pointer, William David and Shaver, Dillon and Liu, Yang and Vegendla, Prasad and Tentner, Adrian},
abstractNote = {The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.},
doi = {10.2172/1335356},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • A test transient performed at a helical coil sodium-to-water steam generator test facility was simulated using the MINET code. It was determined that correct calculation of the sodium outlet temperature requires representation of heat capacitance of the structure.
  • The two helical coils representing the steam generator in the CNSG-IV were tested in the Steam Generator Test Facility of the Alliance Research Center (ARC). This facility combines the capabilities of the Hot Water Test Facility and the Once-Through Steam Generator (OTSG) Test Facility to test steam generators at full system pressures and temperatures for both the primary and secondary sides using water as the test fluid. The report provides the arrangement schematic of the Steam Generator Test Facility used during the functional performance tests of the helical coil steam generator (HCSG). It emphasizes only that equipment used during themore » helical coil test. An OTSG ('A' generator) was also used to provide an additional heat sink so the primary side furnace could be fired at a higher rate, yet be within its controllable range. (GRA)« less
  • The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes andmore » the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics. (GRA)« less
  • The U.S. Department of Energy (DOE) is currently funding research and development of a new high temperature gas cooled reactor (HTGR) that is capable of providing high temperature process heat for industry. The steam generator of the HTGR will consist of an evaporator economizer section in the lower portion and a finishing superheater section in the upper portion. Alloy 800H is expected to be used for the superheater section, and 2.25Cr 1Mo steel is expected to be used for the evaporator economizer section. Dissimilar metal welds (DMW) will be needed to join these two materials. It is well known thatmore » failure of DMWs can occur well below the expected creep life of either base metal and well below the design life of the plant. The failure time depends on a wide range of factors related to service conditions, welding parameters, and alloys involved in the DMW. The overall objective of this report is to review factors associated with premature failure of DMWs operating at elevated temperatures and identify methods for extending the life of the 2.25Cr 1Mo steel to alloy 800H welds required in the new HTGR. Information is provided on a variety of topics pertinent to DMW failures, including microstructural evolution, failure mechanisms, creep rupture properties, aging behavior, remaining life estimation techniques, effect of environment on creep rupture properties, best practices, and research in progress to improve DMW performance. The microstructure of DMWs in the as welded condition consists of a sharp chemical concentration gradient across the fusion line that separates the ferritic and austenitic alloys. Upon cooling from the weld thermal cycle, a band of martensite forms within this concentration gradient due to high hardenability and the relatively rapid cooling rates associated with welding. Upon aging, during post weld heat treatment (PWHT), and/or during high temperature service, C diffuses down the chemical potential gradient from the ferritic 2.25Cr 1Mo steel toward the austenitic alloy. This can lead to formation of a soft C denuded zone near the interface on the ferritic steel, and nucleation and growth of carbides on the austenitic side that are associated with very high hardness. These large differences in microstructure and hardness occur over very short distances across the fusion line (~ 50 100 ?m). A band of carbides also forms along the fusion line in the ferritic side of the joint. The difference in hardness across the fusion line increases with increasing aging time due to nucleation and growth of the interfacial carbides. Premature failure of DMWs is generally attributed to several primary factors, including: the sharp change in microstructure and mechanical properties across the fusion line, the large difference in coefficient of thermal expansion (CTE) between the ferritic and austenitic alloys, formation of interfacial carbides that lead to creep cavity formation, and preferential oxidation of the ferritic steel near the fusion line. In general, the large gradient in mechanical properties and CTE serve to significantly concentrate the stress along the fusion where a creep susceptible microstructure has evolved during aging. Presence of an oxide notch can concentrate the stress even further. Details of the failure mechanism and the relative importance of each factor varies.« less
  • Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary andmore » secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.« less