skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: AGC 2 Irradiated Material Properties Analysis

Abstract

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

Authors:
 [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1369362
Report Number(s):
INL/EXT-17-41165
TRN: US1701969
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; GRAPHITE; REACTOR DESIGN; HTR REACTOR; THERMODYNAMIC PROPERTIES; HTGR TYPE REACTORS; IRRADIATION; REACTOR TECHNOLOGY; Advanced Graphite Creep; high temperature reactor; post irradiation examination

Citation Formats

Rohrbaugh, David Thomas. AGC 2 Irradiated Material Properties Analysis. United States: N. p., 2017. Web. doi:10.2172/1369362.
Rohrbaugh, David Thomas. AGC 2 Irradiated Material Properties Analysis. United States. doi:10.2172/1369362.
Rohrbaugh, David Thomas. 2017. "AGC 2 Irradiated Material Properties Analysis". United States. doi:10.2172/1369362. https://www.osti.gov/servlets/purl/1369362.
@article{osti_1369362,
title = {AGC 2 Irradiated Material Properties Analysis},
author = {Rohrbaugh, David Thomas},
abstractNote = {The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.},
doi = {10.2172/1369362},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2017,
month = 5
}

Technical Report:

Save / Share:
  • This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results frommore » thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.« less
  • In 1983, three spherical buried high-explosive charges were fired to obtain dynamic material properties for in-situ dry alluvium. This report contains a concise summary describing the test-series objectives, rationale for gaging and placement, overall analysis, and recommendations for future tests. The key question concerning the design was what type of test should be fielded to give the most information about material properties used in calculating nuclear surface bursts. Particular requirements for measurement validation and redundancy of data were also key factors in the design. Several alternate approaches to data analysis by the various participants yielded results that pointed out themore » ability to determine uniaxially loading parameters of the material; however, stress difference and unloading estimates are still elusive. Important results, described in some detail, are the estimation of the dynamic loading uniaxial strain versus stress relationship, approach to development of a material properties test, analysis of stress gage performance, and recommendations for future material properties testing.« less
  • The Yucca Mountain Site Characterization Project is studying Yucca Mountain in southwestern Nevada as a potential site for a high-level nuclear waste repository. Site characterization includes surface- based and underground testing. Analyses have been performed to support the design of an Exploratory Studies Facility (ESF) and the design of the tests performed as part of the characterization process, in order to ascertain that they have minimal impact on the natural ability of the site to isolate waste. The information in this report pertains to sensitivity studies evaluating previous hydrological performance assessment analyses to variation in the material properties, conceptual models,more » and ventilation models, and the implications of this sensitivity on previous recommendations supporting ESF design. This document contains information that has been used in preparing recommendations for Appendix I of the Exploratory Studies Facility Design Requirements document.« less
  • Correlation studies have shown that hardening models currently available in the ABAQUS finite element code (isotropic, kinematic) do not accurately capture the inelastic strain reversals that occur due to structural rebounding from a rapidly applied transient dynamic load. The purpose of the Cyclic Material properties Test program was to obtain response data for the first several cycles of inelastic strain reversal from a cyclic properties test. This data is needed to develop elastic-plastic analysis methods that can accurately predict strains and permanent sets in structures due to rapidly applied transient dynamic loading. Test specimens were cycled at inelastic strain levelsmore » typical of rapidly applied transient dynamic analyses (0.5% to 4.0%). In addition to the inelastic response data, cyclic material properties for high yield strength (80 ksi) steel were determined including a cyclic stress-strain curve for a stabilized specimen. Two test methods, the Incremental Step method and the Companion specimen Method, were sued to determine cyclic properties. The incrementally decreasing strain amplitudes in the first loading block of the Incremental Step method test is representative of the response of structures subjected to rapidly applied transient dynamic loads. The inelastic strain history data generated by this test program will be used to support development of a material model that can accurately predict inelastic material behavior including inelastic strain reversals. Additionally, this data can be used to verify material model enhancements to elastic-plastic finite element analysis codes.« less