Evaluation of Creep-Fatigue Damage for Hot Gas Duct Structure of the NHDD Plant-Reference-Cited by-同舟云学术

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

Published:2010-05-04 Issue:3 Volume:132 Page:
ISSN:0094-9930
Container-title:Journal of Pressure Vessel Technology
language:en
Short-container-title:

Author:

Lee Hyeong-Yeon¹,Song Kee-Nam¹,Kim Yong-Wan¹

Affiliation:

1. Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yusong-gu, Daejeon 305-353, Korea

Abstract

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.

Publisher

ASME International

Subject

Mechanical Engineering,Mechanics of Materials,Safety, Risk, Reliability and Quality

Link

http://asmedigitalcollection.asme.org/pressurevesseltech/article-pdf/doi/10.1115/1.4000728/5875035/031101_1.pdf

Reference16 articles.

1. A Study of a Nuclear Hydrogen Production Demonstration Plant;Chang;Nucl. Eng. Technol.

2. Corum, J. M., and Blass, J. J., 1988, “Rules for Design of Alloy 617 Nuclear Components to Very High Temperatures,” Attachment 2, Draft Code Case for Alloy 617.

3. NUREG/CR-6816, 2003, “Review and Assessment of Codes and Procedures for HTGR Components,” Paper No. ANL-02/36.

4. ASME, 2007, “Rules for Construction of Nuclear Power Plant Components,” ASME Boiler and Pressure Vessel Code, Section III, Div. 1, Subsection NH, Class 1 Components in Elevated Temperature Service.

5. Creep-Fatigue Damage for a Structure With Dissimilar Metal Welds of Mod 9Cr-1Mo and 316L Stainless Steel;Lee;Int. J. Fatigue

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