scholarly journals Assessment of very high-temperature reactors in process applications. Appendix II. VHTR process heat application studies

1977 ◽  
Author(s):  
J.E. Jones ◽  
W.R. Gambill ◽  
R.H. Cooper ◽  
E.C. Fox ◽  
L.C. Fuller ◽  
...  
Keyword(s):  
High Temperature ◽  
Heat Application ◽  
Process Heat ◽  
High Temperature Reactors ◽  
Very High

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

2008 ◽  
Author(s):  
William J. O’Donnell ◽  
Amy B. Hull ◽  
Shah Malik
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Structural Integrity ◽  
Cyclic Creep ◽  
Creep Rupture ◽  
Creep Fatigue ◽  
Asme Code ◽  
High Temperature Reactors ◽  
Gen Iv ◽  
Very High

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

Tritium Permeation Characterization of Materials for Fusion and Generation IV Very High Temperature Reactors

2015 ◽  
Vol 67 (3) ◽  
pp. 475-478
Author(s):  
S. Thomson ◽  
K. Pilatzke ◽  
K. McCrimmon ◽  
I. Castillo ◽  
S. Suppiah
Keyword(s):  
High Temperature ◽  
High Temperature Reactors ◽  
Tritium Permeation ◽  
Characterization Of Materials ◽  
Very High

Chromium Activity Measurements in Nickel Based Alloys for Very High Temperature Reactors: Inconel 617, Haynes 230, and Model Alloys

2009 ◽  
Vol 131 (6) ◽  
Author(s):  
Stéphane Gossé ◽  
Thierry Alpettaz ◽  
Sylvie Chatain ◽  
Christine Guéneau
Keyword(s):  
High Temperature ◽  
Heat Exchangers ◽  
Mass Spectrometric ◽  
Operating Conditions ◽  
Pure Substance ◽  
Haynes 230 ◽  
Inconel 617 ◽  
High Temperature Reactor ◽  
High Temperature Reactors ◽  
Very High

The alloys Haynes 230 and Inconel 617 are potential candidates for the intermediate heat exchangers (IHXs) of (very) high temperature reactors ((V)-HTRs). The behavior under corrosion of these alloys by the (V)-HTR coolant (impure helium) is an important selection criterion because it defines the service life of these components. At high temperature, the Haynes 230 is likely to develop a chromium oxide on the surface. This layer protects from the exchanges with the surrounding medium and thus confers certain passivity on metal. At very high temperature, the initial microstructure made up of austenitic grains and coarse intra- and intergranular M6C carbide grains rich in W will evolve. The M6C carbides remain and some M23C6 richer in Cr appear. Then, carbon can reduce the protective oxide layer. The alloy loses its protective coating and can corrode quickly. Experimental investigations were performed on these nickel based alloys under an impure helium flow (Rouillard, F., 2007, “Mécanismes de formation et de destruction de la couche d’oxyde sur un alliage chrominoformeur en milieu HTR,” Ph.D. thesis, Ecole des Mines de Saint-Etienne, France). To predict the surface reactivity of chromium under impure helium, it is necessary to determine its chemical activity in a temperature range close to the operating conditions of the heat exchangers (T≈1273 K). For that, high temperature mass spectrometry measurements coupled to multiple effusion Knudsen cells are carried out on several samples: Haynes 230, Inconel 617, and model alloys 1178, 1181, and 1201. This coupling makes it possible for the thermodynamic equilibrium to be obtained between the vapor phase and the condensed phase of the sample. The measurement of the chromium ionic intensity (I) of the molecular beam resulting from a cell containing an alloy provides the values of partial pressure according to the temperature. This value is compared with that of the pure substance (Cr) at the same temperature. These calculations provide thermodynamic data characteristic of the chromium behavior in these alloys. These activity results call into question those previously measured by Hilpert and Ali-Khan (1978, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: I. Alloy IN-643,” J. Nucl. Mater., 78, pp. 265–271; 1979, “Mass Spectrometric Studies of Alloys Proposed for High-Temperature Reactor Systems: II. Inconel Alloy 617 and Nimomic Alloy PE 13,” J. Nucl. Mater., 80, pp. 126–131), largely used in the literature.

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.

Development and Implementation of a 3D Heat Conduction Model for (Very) High Temperature Reactors Into the Reactor Dynamics Code DYN3D

2012 ◽  
Author(s):  
Silvio Baier ◽  
Ulrich Rohde ◽  
Soeren Kliem ◽  
Emil Fridman
Keyword(s):  
Heat Conduction ◽  
High Temperature ◽  
Block Type ◽  
Conduction Model ◽  
Heat Conduction Model ◽  
Reactor Dynamics ◽  
3D Effects ◽  
High Temperature Reactors ◽  
Short Time ◽  
Very High

The reactor dynamics code DYN3D was extended to treat phenomena in Block-type High Temperature Reactors (HTR). Therefor, a new heat conduction model was implemented into the code to tackle 3D effects of heat conduction and heat transfer. The first part of the paper describes the details of the heat conduction model. In the second part results of coupled neutron-kinetics/thermal-hydraulics calculations of steady state and short-time transients in block-type HTRs are discussed.

High Temperature Helium Technology on Process Heat Application of HTGR

1981 ◽  
Vol 84 (757) ◽  
pp. 1296-1303
Author(s):  
Takenori NAKANISHI ◽  
Tetsuro NAKADA ◽  
Masami ITOH ◽  
Masaki KITAGAWA
Keyword(s):  
High Temperature ◽  
Heat Application ◽  
Process Heat
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