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Title: Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

Abstract

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements suchmore » as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Authors:
; ; ; ;
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
1214760
Report Number(s):
INL/JOU-14-32328
DOE Contract Number:
DE-AC07-05ID14517
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Technology; Journal Volume: 191
Country of Publication:
United States
Language:
English
Subject:
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; In-Pile Deformation and Measurement Instrumentatio

Citation Formats

K. L. Davis, D. L. Knudson, J. L. Rempe, J. C. Crepeau, and S. Solstad. Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor. United States: N. p., 2015. Web.
K. L. Davis, D. L. Knudson, J. L. Rempe, J. C. Crepeau, & S. Solstad. Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor. United States.
K. L. Davis, D. L. Knudson, J. L. Rempe, J. C. Crepeau, and S. Solstad. 2015. "Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor". United States. doi:.
@article{osti_1214760,
title = {Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor},
author = {K. L. Davis and D. L. Knudson and J. L. Rempe and J. C. Crepeau and S. Solstad},
abstractNote = {New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.},
doi = {},
journal = {Nuclear Technology},
number = ,
volume = 191,
place = {United States},
year = 2015,
month = 7
}
  • New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumentedmore » creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).« less
  • New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumentedmore » creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.« less
  • The influence of design parameters and burning on pellet/cladding interaction (PCI) of current boiling water reactor fuel rods was studied through in-core diameter measurement. Thinner cladding and a smaller diametral gap enhanced the PCI during startup. At constant power, fuel with SiO/sub 2/ added greatly reduced PCI due to relaxation. The fuel with a small grain size greatly reduced PCI due to densification. Preirradiation of rods up to 23 MWd/kgU caused a large PCI not only in a small gap but also in a large gap rod. Relaxation and permanent deformation was small. In the power increase experiment, one rodmore » experienced PCI failure. The spurt times of coolant radioactivity coincided well with the sudden drop of cladding axial strain and marked crack opening at the rod surface. The estimated hoop stress predicted by FEMAXI-III was 350 MPa at the failure.« less
  • The US Department of Energy (DOE) is embarking on a series of tests of tristructural isotropic (TRISO) coated-particle reactor fuel for the advanced gas reactor (AGR). As one part of this fuel development program, a series of eight fuel irradiation tests are planned for the Idaho National Laboratory's (INL's) advanced test reactor (ATR). The first test in this series (AGR-1) will incorporate six separate capsules irradiated simultaneously, each containing about 51,000 TRISO-coated fuel particles supported in a graphite matrix and continuously swept with inert gas during irradiation. The effluent gas from each of the six capsules must be independently monitoredmore » in near real time and the activity of various fission gas nuclides determined and reported. A set of seven heavily-shielded, high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based total radiation detectors have been designed and are being configured and tested for use during the AGR-1 experiment. The AGR-1 test specification requires that the fission product monitoring system (FPMS) have sufficient sensitivity to detect the failure of a single coated fuel particle and sufficient range to allow it to "count" multiple (up to 250) successive particle failures. This paper describes the design and expected performance of the AGR-1 FPMS.« less
  • Metallographic examination of heat exchanger tubes and fins which have been subjected to comparatively short time exposure in the Dorr--Oliver reactor environment indicates the following: (a) there has been no significant deterioration of fins or tubes fabricated from Incology 800H material; (b) the cobalt base braze alloy used to join the fins to the tubes showed no apparent attack. Micro-probe analysis revealed shallow internal attack on the fins, primarily oxidation, to a depth of approximately .002'' to .004''. A slight sulfur penetration was noted to a depth of about .002''.