PM-0614-7392-NP - NuScale Comprehensive Flow-Induced ...
Transcript of PM-0614-7392-NP - NuScale Comprehensive Flow-Induced ...
NuScale ComprehensiveFlow-Induced Vibration
Program
Dr. Tamas Liszkai, P.E.
Reactor Module Design Supervisor
July 23, 2014
NuScale Nonproprietary
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Agenda" Overview
* Abbreviations
* Scope
* CVAP plan
* Key elements of CVAP
* Flow velocity
* Test plan
* Summary
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Purpose" Address and outline steps for NuScale SSC and its FOAK
design in the area of
- Regulatory Guide 1.20 Rev. 3, Comprehensive VibrationAssessment Program
- scope of program
- FIV analysis and testing of SSC
" Provide a description of additional activities planned forthese programs to support DCA application and the COLapplication
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Abbreviations
ASME American Society of Mechanical Engineers
BPVC boiler and pressure vessel code
CFD computational fluid dynamics
CIV containment isolation valve
CNSG consolidated nuclear steam generator
CNV containment vessel
COL combined operating license
CRA control rod assembly
CRDS control rod drive system
CVAP comprehensive vibration assessment program
FEI fluid elastic instability
FIV flow-induced vibration
FOAK first-of-a-kind
HCSG helical coil steam generator
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AbbreviationsI Reco ese nenl
ICI in-core instrument
ISI inservice inspection
IST inservice testing
MARAD U.S. Dept of Transportation-Maritime Administration
MS main steam
NDE nondestructive examination
PSD power spectral density
PWR pressurized-water reactor
RCPB reactor coolant pressure boundary
RCS reactor coolant system
RMS root mean square
RQTP Reactor Qualification Test Plan
RVI reactor vessel internals
SG steam generator
SSC structure, system, and component
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Scope" NuScale design is a FOAK plant and meets the definition
of prototype design criteria as delineated in RG 1.20
" Due to the NuScale integral design, SSCs included in theCVAP are
- RVI perASME BPVC, Section III, Subsection NG
- HCSG RCPB tubes
- NuScale Power Module main steam and main feedwater piping upto, and including, main steam isolation valves
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Pre-DCA Submittal CVAP Plan
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Post-DCA CVAP Plan
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NuScale Power Module Components
ClVs
CNV •
RXM piping
RW -
RPV
integrated PZRbaffle plate andsteam plenum
SG annularspace
RVI lower riserassembly
RVI core supportassembly
CIV = containment isolation valve
CNV = containment vessel
PZR = pressurizer
RPV = reactor pressure vessel
RRV = reactor recirculation valve
RVI = reactor vessel internals
RVV = reactor vent valve
RXM = reactor module
SG = steam generator
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Flow-Induced Vibration Mechanisms
Phnmeo Scenn Criteria
Fluid elasticinstability
Vortex shedding
Turbulent buffeting
Acoustic resonance
1. Array of cylinders (minimum one row) (i.e., geometry)
2. Array pitch/diameter < 2.0; array must sufficiently confine fluid to allow
feedback between adjacent cylinders
1. Bluff body (or edge of a cavity in-line with flow) (i.e., geometry)
2. Subject to cross-flow
3. Absence of downstream structures to disrupt vortices
1. Subject to turbulent flow (axial, cross-flow or combination)
1. Suitable geometry to generate an acoustic resonance, typically a hollow or
cavity2. Single phase gas/vapor environment within hollow/cavity
Leakage flowinstability
1.2.3.
1.2.
Narrow annular flow path exists (i.e., geometry)
Flexible structure in annulus, bounded by fixed surface
Annular flow path is diverging (restriction at inlet to annulus) or parallel
Noncircular cross section (i.e., geometry)
Aspect ratio (length/width) in prevailing direction of flow is 4.0 or greater (for
rectangular structure)Galloping/flutter
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Component List for FIV
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Component FIV Applicability Screening
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SG FIV Screening
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RVI Riser Assembly FIV Screening
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RVI Core Support FIV Screening
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Vibration and Stress Analysis" Determination of critical characteristics for FIV
* Common input to FIV analysis
" Basis for testing development
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Key Elements for Structural Analysis
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Key Elements for Thermal Hydraulics
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Key Elements for Damping
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Key Elements for Turbulence
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Average and Maximum Flow Velocities'
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Flow Velocity ComparisonAvrg Veoct at10%Pwe ftUta Downcomer*Ioe Upe Snternals
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EPR1 N/A 24 16 30 5
API1000 2 N/A 19 16 40 5
US-APWR 3 N/A 23 14 30 5
SONGS 4 18 N/A N/A N/A
1. UK EPR DCD Chapters 3, 4, and 52. AP1000 DCD Chapters 4 and 53. US-APWR DCD Chapter 4 and 54. U.S. Nuclear Regulatory Commission, "San Onofre Nuclear Generating Station-NRC Augmented Inspection Team Report
05000361/2012007 and 05000362/20112007," July 20125. Cross flow velocity estimated as half hot leg velocity
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NuScale Reactor Qualification Test Plan
e The Reactor Qualification Test Plan (RQTP) includestesting required to support the DCA
e Includes SG FIV testing needed to support CVAP
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CRA/ICI guide tubes
- most complex geometry outside SG in RCS flow path
- need for flow testing is under evaluation to establish boundingforcing functions
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Pressures Due to Secondary Side Boiling
9 SIET TF-1 test
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TF-1 Test Conditions-Pressure Measurements
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TF-1-Power Spectral Density
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Fluid Elastic Instability (FEI)* Review of previous (public domain) testing, indicates no FEI tests of
full helical tube arrays, only helical segments
- Chen: testing at Argonne National Lab (ANL)l
- B&W: testing of CNSG design for MARAD 2
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" FEI design data (ASME Appendix N-1 300) is based on straight tubearrays
* Limited Chen and B&W testing of helical segments indicated straighttube array data should bound helical array FEI
* To provide additional confidence in FEI design margins for NuScaleHCSG, full helical bundle FEI test will be performed
il Chen, S.S., et al., "Tube Vibration in a Half-Scale Sector Model of a Helical Tube Steam Generator," Journal of Sound and Vibration, Vol.
91, pages 539-569 (1983).21 Glasser, R.P., "Experimental Evaluation of Helical Consolidated Nuclear Steam Generator (CNSG) Tubes and Supports, MA-RD-920-
76019, November 1975.
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NuScale FEI Test* Will utilize existing test section for fluid heated test (TF-2)
at SIET
- test section employs a NuScale tube support concept
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9 Testing will include secondary flow (including two-phase)
- investigate the potential for unknown effects due to two-phasesecondary flow forcing coupled with primary flow
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Turbulent Forcing Functions" NuScale has evaluated all geometries in the reactor
vessel internals (including SG and SG tube supports) toidentify turbulent forcing functions
" Turbulent forcing functions consist of three components
- correlation length (A)
- convective velocity (v)
- power spectral density (PSD)
" Based on geometry and/or correlations, specificcorrelation lengths and convective velocities for all RVIand SG components have been identified
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SG Power Spectral Densities* Limited open source PSDs exist for
- HCSG tube external flow 1'2
- annular turbulent flow 2
- parallel flow over 2D surfaces
- cross flow around 1 D structures
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ii American Society of Mechanical Engineers, Boilerand Pressure Vessel Code, Section III, "Rules for Construction of Nuclear FacilityComponents," Appendix-N, 2007 Edition with no Addenda.21 Au-Yang, M.K., "Flow-Induced Vibration of Power and Process Plant Components," ASME Press, New York, 20113i Snyder, M., et al., Progress in the Generation of Flow Turbulence Excitation Forces from CFD Analyses and Experimental Data, Paper ID000127, 6th International Conference on Nuclear Thermal Hydraulic, Operations and Safety (NUTHOS-6), Nara, Japan, October 4-8, 2004.
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Turbulent Annular Flow PSD1
E+O
E-I
E-2
(-,
0
0 1.0 2.0 3.0 4.0
F=JRJ/V
5.0
11 Au-Yang, M.K., "Flow-Induced Vibration of Power and Process Plant Components," ASME Press, New York, 2011
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HCSG Tube Random Vibration
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HCSG Tube Modal Analysis
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HCSG Tube RMS Response
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Summary" NuScale CVAP to satisfy RG 1.20 requirements
" Screening for all potential FIV mechanisms inside RPV
" Analyses or evaluations are performed for all susceptiblecomponents
" Generally very low susceptibility due to low flow rates
" Pre-DCA testing needs have been identified
" The key is to provide sufficient details in the DCA, suchthat the NuScale analysis, testing, and inspectionmethodology including uncertainties for FIV can beevaluated by the NRC
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