Search Menu
DOE PAGES title logo U.S. Department of Energy
Office of Scientific and Technical Information
Search Scholarly Publications

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

Journal Article · · Nuclear Engineering and Design
 [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)
+ Show Author Affiliations

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.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE); USDOE National Nuclear Security Administration (NNSA)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
1557662
Alternate ID(s):
OSTI ID: 1582893
Report Number(s):
INL/JOU-18-51979-Rev000
Journal Information:
Nuclear Engineering and Design, Vol. 335, Issue C; ISSN 0029-5493
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English
Citation Metrics:
Cited by: 3 works
Citation information provided by
Web of Science

Figures / Tables (23)

Fig. 1(p. 14)figure Fig. 1
Fig. 2(p. 15)figure Fig. 2
Fig. 3(p. 17)figure Fig. 3
Fig. 4(p. 18)figure Fig. 4
Fig. 5(p. 21)figure Fig. 5
Fig. 6(p. 22)figure Fig. 6
Table 1(p. 23)table Table 1
Table 2(p. 23)table Table 2
Table 3(p. 24)table Table 3
Fig. 7(p. 26)figure Fig. 7
Table 4(p. 28)table Table 4
Fig. 8(p. 30)figure Fig. 8
Table 5(p. 32)table Table 5
Fig. 9(p. 34)figure Fig. 9
Table 6(p. 36)table Table 6
Fig. 10(p. 37)figure Fig. 10
Fig. 11(p. 38)figure Fig. 11
Fig. 12(p. 39)figure Fig. 12
Fig. 13(p. 40)figure Fig. 13
Fig. 14(p. 41)figure Fig. 14
Fig. 15(p. 41)figure Fig. 15
Fig. 16(p. 43)figure Fig. 16
Fig. 17(p. 45)figure Fig. 17

Similar Records

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
MP-1
NED_9725
ATR
USHPRR-FQ
25228