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Title: Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

Abstract

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. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being moremore » cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

Authors:
 [1];  [1];  [1];  [1];  [1]
  1. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy - Nuclear Energy Advanced Modeling and Simulation (NEAMS)
OSTI Identifier:
1409207
Report Number(s):
ANL/NE-16/49
136860
DOE Contract Number:
AC02-06CH11357
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS

Citation Formats

Merzari, E., Yuan, Haomin, Kraus, A., Solberg, Jerome, and Ferencz, Robert M. Coupled Calculations in Helical Steam Generator: Validation on Legacy Data. United States: N. p., 2016. Web. doi:10.2172/1409207.
Merzari, E., Yuan, Haomin, Kraus, A., Solberg, Jerome, & Ferencz, Robert M. Coupled Calculations in Helical Steam Generator: Validation on Legacy Data. United States. doi:10.2172/1409207.
Merzari, E., Yuan, Haomin, Kraus, A., Solberg, Jerome, and Ferencz, Robert M. 2016. "Coupled Calculations in Helical Steam Generator: Validation on Legacy Data". United States. doi:10.2172/1409207. https://www.osti.gov/servlets/purl/1409207.
@article{osti_1409207,
title = {Coupled Calculations in Helical Steam Generator: Validation on Legacy Data},
author = {Merzari, E. and Yuan, Haomin and Kraus, A. and Solberg, Jerome and Ferencz, Robert M.},
abstractNote = {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. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.},
doi = {10.2172/1409207},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • Three steam generator transient test cases, that were simulated using the MINET computer code, are described, with computed results compared against experimental data. The MINET calculations closely agreed with the experiment for both the once-through and the U-tube steam generator test cases. The effort is part of an ongoing effort to validate the MINET computer code for thermal-hydraulic plant systems transient analysis, and strongly supports the validity of the MINET models.
  • 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