Correlation of analysis with high level vibration test results for primary coolant piping
- Brookhaven National Lab., Upton, NY (United States)
- Nuclear Regulatory Commission, Washington, DC (United States)
Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results.
- Research Organization:
- Brookhaven National Lab., Upton, NY (United States)
- Sponsoring Organization:
- USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 5424926
- Report Number(s):
- BNL-NUREG-47397; CONF-920631-25; ON: DE92012517
- Resource Relation:
- Conference: American Society of Mechanical Engineers pressure vessel and piping conference, New Orleans, LA (United States), 21-25 Jun 1992
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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SEISMIC EFFECTS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
COMPUTER CODES
DYNAMIC LOADS
MECHANICAL VIBRATIONS
REACTOR SAFETY
SCALE MODELS
TESTING
COOLING SYSTEMS
ENRICHED URANIUM REACTORS
POWER REACTORS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
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Nonbreeding
Light-Water Moderated
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