Evaluation of Creep-Fatigue Damage for Hot Gas Duct Structure of the NHDD Plant

2010 ◽  
Vol 132 (3) ◽  
Author(s):  
Hyeong-Yeon Lee ◽  
Kee-Nam Song ◽  
Yong-Wan Kim
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Inlet Temperature ◽  
Temperature Structure ◽  
Alloy 800H ◽  
Alloy 617 ◽  
Draft Code ◽  
Hot Gas ◽  
Creep Fatigue ◽  
Technical Issues

Evaluation of creep-fatigue damage has been carried out for the hot gas duct (HGD) structure in the nuclear hydrogen development and demonstration (NHDD) plant. The core outlet and inlet temperature of the NHDD plant are 950°C and 490°C, respectively. Case studies on high temperature design codes of the draft code case for Alloy 617, ASME boiler and pressure vessel code section III subsection NH (ASME-NH), and RCC-MR were carried out for the inner tube of the HGD for the candidate materials of Alloy 617 and Alloy 800H. Technical issues in application of the draft code case to a high temperature structure are discussed for the Alloy 617 material. Code comparison between the ASME-NH and RCC-MR for Alloy 800H has been carried out. The candidate material of the outer pressure boundary (cross vessel) of the HGD is Mod.9Cr-1Mo steel. The damage evaluation, according to the ASME-NH and RCC-MR for the cross vessel of Mod.9Cr-1Mo steel, has been conducted and their results were compared.

Evaluation of Creep-Fatigue Damage for Hot Gas Duct Structure of the NHDD Plant

2009 ◽  
Author(s):  
Hyeong-Yeon Lee ◽  
Kee-Nam Song ◽  
Yong-Wan Kim
Keyword(s):  
High Temperature ◽  
Fatigue Damage ◽  
Inlet Temperature ◽  
Temperature Structure ◽  
Alloy 800H ◽  
Alloy 617 ◽  
Draft Code ◽  
Hot Gas ◽  
Creep Fatigue ◽  
Technical Issues

Evaluation of creep-fatigue damage has been carried out for the HGD (hot gas duct) structure in the NHDD (Nuclear Hydrogen Development and Demonstration) plant. The core outlet and inlet temperature of the NHDD plant are 950°C and 490°C, respectively. Case studies on high temperature design codes of the draft Code Case for Alloy 617, ASME-NH and RCC-MR were carried out for the inner tube of the HGD for the candidate materials of Alloy 617 and Alloy 800H. Technical issues in application of the draft Code Case to a high temperature structure are discussed for Alloy 617 material. Code comparison between the ASME-NH and RCC-MR for Alloy 800H has been carried out. The candidate material of the outer pressure boundary (cross vessel) of the HGD is Mod.9Cr-1Mo steel. The damage evaluation according to the ASME-NH and RCC-MR for the cross vessel of Mod.9Cr-1Mo steel has been conducted and their results were compared.

An Evaluation of Creep-Fatigue Damage for the Prototype Process Heat Exchanger of the NHDD Plant

2011 ◽  
Vol 133 (5) ◽  
Author(s):  
Hyeong-Yeon Lee ◽  
Kee-Nam Song ◽  
Yong-Wan Kim ◽  
Sung-Deok Hong ◽  
Hong-Yune Park
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
High Temperature ◽  
Fatigue Damage ◽  
Inlet Temperature ◽  
Alloy 617 ◽  
Element Analysis ◽  
Creep Fatigue ◽  
Process Heat ◽  
Very High

A process heat exchanger (PHE) transfers the heat generated from a nuclear reactor to a sulfur-iodine hydrogen production system in the Nuclear Hydrogen Development and Demonstration, and was subjected to very high temperature up to 950°C. An evaluation of creep-fatigue damage, for a prototype PHE, has been carried out from finite element analysis with the full three dimensional model of the PHE. The inlet temperature in the primary side of the PHE was 950°C with an internal pressure of 7 MPa, while the inlet temperature in the secondary side of the PHE is 500°C with internal pressure of 4 MPa. The candidate materials of the PHE were Alloy 617 and Hastelloy X. In this study, only the Alloy 617 was considered because the high temperature design code is available only for Alloy 617. Using the full 3D finite element analysis on the PHE model, creep-fatigue damage evaluation at very high temperature was carried out, according to the ASME Draft Code Case for Alloy 617, and technical issues in the Draft Code Case were raised.

Evaluation of Creep-Fatigue Damage for Prototype Process Heat Exchanger of the NHDD Plant

2010 ◽  
Author(s):  
Hyeong-Yeon Lee ◽  
Kee-Nam Song ◽  
Yong-Wan Kim ◽  
Sung-Deok Hong ◽  
Hong-Yune Park
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
High Temperature ◽  
Fatigue Damage ◽  
Inlet Temperature ◽  
Alloy 617 ◽  
Element Analysis ◽  
Creep Fatigue ◽  
Process Heat ◽  
Very High

A process heat exchanger (PHE) transfers the heat generated from a nuclear reactor to a sulfur-iodine hydrogen production system in the NHDD (Nuclear Hydrogen Development and Demonstration), and was subjected to very high temperature up to 950°C. An evaluation of creep-fatigue damage, for a prototype PHE, has been carried out from finite element analysis with the full three dimensional model of the PHE. The inlet temperature in the primary side of the PHE was 950°C with an internal pressure of 7MPa while the inlet temperature in the secondary side of the PHE is 500°C with internal pressure of 4MPa. The candidate materials of the PHE were Alloy 617 and Hastelloy X. In this study, only the Alloy 617 was considered because the high temperature design code is available only for Alloy 617. Using the full 3D finite element analysis on the PHE model, creep-fatigue damage evaluation at very high temperature was carried out, according to the ASME Draft Code Case for Alloy 617, and technical issues in the draft Code Case were raised.

A Viscoplastic Model for Alloy 617 for Use With the ASME Section III, Division 5 Design by Inelastic Analysis Rules

2021 ◽  
Author(s):  
M. C. Messner ◽  
T.-L. Sham
Keyword(s):  
High Temperature ◽  
Alloy 617 ◽  
Element Analysis ◽  
Material Models ◽  
Experimental Database ◽  
Viscoplastic Model ◽  
Analysis And Design ◽  
Inelastic Analysis ◽  
Wide Range ◽  
Creep Fatigue

Abstract The rules for the design of high temperature reactor components in Section III, Division 5, Subsection HB, Subpart B (HBB) of the ASME Boiler and Pressure Vessel Code contain two options for evaluating the deformation-controlled design limits on strain accumulation and creep-fatigue: design by elastic analysis and design by inelastic analysis. Of these options design by inelastic analysis tends to be less overconservative and produce more efficient designs. However, the HBB currently does not provide approved material models for use with the inelastic analysis rules, limiting their widespread use. A nonmandatory appendix has been developed to provide general guidance on appropriate material models and provide reference material models suitable for use with the design by inelastic analysis approach. This paper describes a viscoplastic model for Alloy 617 suitable for use with the HBB rules proposed for incorporation into the new appendix. The model represents the high temperature creep, creep-fatigue, and tensile response of Alloy 617 and accurately accounts for rate sensitivity across a wide range of temperatures. The focus in developing the model was on capturing key features of material deformation required for accurately executing the HBB rules and on developing a relatively simple model form that can be implemented in commercial finite element analysis software. The paper validates the model against an extensive experimental database collected as part of the Alloy 617 Code qualification effort as well as against specialized experimental tests examining the effect of elastic follow up on stress relaxation and creep deformation in the material.

Structural Evaluation With Elastic and Inelastic Analysis Methods on a High Temperature Storage Tank Subjected to Static and Dynamic Loadings

2020 ◽  
Author(s):  
Wen Wang ◽  
Xiaochun Zhang ◽  
Xiaoyan Wang ◽  
Maoyuan Cai
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Molten Salt ◽  
Fatigue Damage ◽  
Storage Tank ◽  
Power Plants ◽  
Elastic Analysis ◽  
Inelastic Analysis ◽  
Analysis Methods ◽  
Creep Fatigue

Abstract The structural integrity of reactor components is very essential for the reliable operation of all types of power plants, especially for components operating at elevated temperature where creep effects are significant and where components are subjected to high-temperature alteration and seismic transient loading conditions. In this article, a molten salt storage tank in high temperature thorium molten salt reactor (TMSR) is evaluated according to ASME-III-5-HBB high temperature reactor code. The evaluation based on 3D finite element analyses includes the load-controlled stress, the effects of ratcheting, and the interaction of creep and fatigue. The thermal and structural analysis and the application procedures of ASME-HBB rules are described in detail. Some structural modifications have been made on this molten salt storage tank to enhance the strength and reduce thermal stress. The effects of ratcheting and creep-fatigue damage under elevated temperature are investigated using elastic analysis and inelastic analysis methods for a defined representative load cycle. In addition, the strain range and the stress relaxation history calculated by elastic and inelastic methods are compared and discussed. The numerical results indicate that the elastic analysis is conservative for design and a full inelastic analysis method for estimating input for creep-fatigue damage evaluation need to be developed.

scholarly journals Strain Rate Sensitivity of Alloys 800H and 617

2013 ◽  
Author(s):  
J. K. Wright ◽  
J. A. Simpson ◽  
R. N. Wright ◽  
L. J. Carroll ◽  
T. L. Sham
Keyword(s):  
High Temperature ◽  
Flow Stress ◽  
Strain Rate ◽  
Strain Rate Sensitivity ◽  
Rate Sensitivity ◽  
Temperature Limit ◽  
Applied Strain ◽  
Alloy 800H ◽  
Alloy 617 ◽  
Temperature Heat

The flow stress of many materials is a function of the applied strain rate at elevated temperature. The magnitude of this effect is captured by the strain rate sensitivity parameter “m”. The strain rate sensitivity of two face–center cubic solid solution alloys that are proposed for use in high temperature heat exchanger or steam generator applications, Alloys 800H and 617, has been determined as a function of temperature over that range of temperatures relevant for these applications. In addition to determining the strain rate sensitivity, it is important for nuclear design within Section III of the ASME Boiler and Pressure Vessel Code to determine temperature below which the flow stress is not affected by the strain rate. This temperature has been determined for both Alloy 800H and Alloy 617. At high temperature the strain rate sensitivity of the two alloys is significant and they have similar m values. For Alloy 617 the temperature limit below which little or no strain rate sensitivity is observed is approximately 700°C. For Alloy 800H this temperature is approximately 650°C.

High Temperature Design and Damage Evaluation of Mod.9Cr-1Mo Steel Heat Exchanger

2011 ◽  
Author(s):  
Hyeong-Yeon Lee ◽  
Jong-Bum Kim ◽  
Jae-Hyuk Eoh ◽  
Yong-Bum Lee ◽  
Hong-Yune Park
Keyword(s):  
High Temperature ◽  
Heat Exchanger ◽  
Fatigue Damage ◽  
Effective Length ◽  
Outer Diameter ◽  
Element Analysis ◽  
Test Loop ◽  
Creep Fatigue ◽  
Shell And Tube ◽  
Critical Locations

High temperature design and evaluation of creep-fatigue damage for sodium-sodium heat exchanger, DHX (Decay heat exchanger) in a sodium test loop have been conducted. The DHX is a shell- and tube-type heat exchanger with outer diameter of 21.7mm, thickness of 1.65mm and effective length of 1.73m. The DHX shell and tube materials were Mod.9Cr-1Mo steel. The temperatures of shell inlet and shell outlet in the DHX are 510°C and 308°C, respectively, while the temperatures of tube inlet and outlet are 254°C and 475°C, respectively. Three dimensional finite element analysis was conducted for the DHX and evaluation of creep-fatigue damage at several critical locations of the heat exchanger was carried out according to the elevated temperature design codes of the ASME Section III Subsection NH and RCC-MR. Evaluations on the integrity of the DHX and code comparisons were carried out for the critical locations of the DHX.

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