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Title: On the flow induced vibration of an externally excited nuclear reactor experiment

Abstract

An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. Here, the objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific testmore » element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.« less

Authors:
 [1]; ORCiD logo [2];  [1];  [3];  [3];  [3];  [1];  [1];  [3];  [1];  [3]
  1. Oregon State Univ., Corvallis, OR (United States)
  2. Oregon State Univ., Corvallis, OR (United States); Idaho National Lab. (INL), Idaho Falls, ID (United States)
  3. 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); USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1557662
Alternate Identifier(s):
OSTI ID: 1582893
Report Number(s):
INL/JOU-18-51979-Rev000
Journal ID: ISSN 0029-5493
Grant/Contract Number:  
AC07-05ID14517
Resource Type:
Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 335; Journal Issue: C; Journal ID: ISSN 0029-5493
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; MP-1, NED_9725, ATR, USHPRR-FQ; 25228

Citation Formats

Latimer, Griffen, Marcum, Wade R., Howard, Trevor K., Jones, Warren, Phillips, Ann Marie, Woolstenhulme, Nicolas, Liu, Suyang, Weiss, Aaron, Campbell, Jed, Moussaoui, Musa, and Jensen, Colby. On the flow induced vibration of an externally excited nuclear reactor experiment. United States: N. p., 2018. Web. doi:10.1016/j.nucengdes.2018.05.007.
Latimer, Griffen, Marcum, Wade R., Howard, Trevor K., Jones, Warren, Phillips, Ann Marie, Woolstenhulme, Nicolas, Liu, Suyang, Weiss, Aaron, Campbell, Jed, Moussaoui, Musa, & Jensen, Colby. On the flow induced vibration of an externally excited nuclear reactor experiment. United States. doi:10.1016/j.nucengdes.2018.05.007.
Latimer, Griffen, Marcum, Wade R., Howard, Trevor K., Jones, Warren, Phillips, Ann Marie, Woolstenhulme, Nicolas, Liu, Suyang, Weiss, Aaron, Campbell, Jed, Moussaoui, Musa, and Jensen, Colby. Sat . "On the flow induced vibration of an externally excited nuclear reactor experiment". United States. doi:10.1016/j.nucengdes.2018.05.007. https://www.osti.gov/servlets/purl/1557662.
@article{osti_1557662,
title = {On the flow induced vibration of an externally excited nuclear reactor experiment},
author = {Latimer, Griffen and Marcum, Wade R. and Howard, Trevor K. and Jones, Warren and Phillips, Ann Marie and Woolstenhulme, Nicolas and Liu, Suyang and Weiss, Aaron and Campbell, Jed and Moussaoui, Musa and Jensen, Colby},
abstractNote = {An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. Here, the objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific test element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.},
doi = {10.1016/j.nucengdes.2018.05.007},
journal = {Nuclear Engineering and Design},
number = C,
volume = 335,
place = {United States},
year = {2018},
month = {5}
}

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Figures / Tables:

Fig. 1 Fig. 1: First three circumferential (top) and axial (bottom) modes

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