ML19343C756
| ML19343C756 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 01/08/2020 |
| From: | Vera Amadiz M NRC/NRR/DNRL/NRLB |
| To: | Michael Dudek NRC/NRR/DNRL/NRLB |
| Vera Amadiz M, 415-5861 | |
| Shared Package | |
| ML19343C747 | List: |
| References | |
| Download: ML19343C756 (13) | |
Text
January 08, 2020 MEMORANDUM TO:
Michael I. Dudek, Chief New Reactor Licensing Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation FROM:
Marieliz Vera, Project Manager /RA/
New Reactor Licensing Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation
SUBJECT:
U.S. NUCLEAR REGULATORY COMMISSION REPORT OF THE REGULATORY FOLLOW-UP AUDIT PERFORMED BETWEEN JUNE 24, 2019, THROUGH SEPTEMBER 30, 2019, REGARDING HELICAL COIL STEAM GENERATOR TUBE TESTING AT SIET LABORATORY NUSCALE POWER, LLC, STANDARD PLANT DESIGN CERTIFICATION On January 6, 2017, NuScale Power, LLC (NuScale) submitted a design certification (DC) application, for a Small Modular Reactor, to the U.S. Nuclear Regulatory Commission (NRC)
(Agencywide Documents Access and Management System (ADAMS) Accession Number ML17013A229). The NRC staff started its detailed technical review of NuScales DC application on March 15, 2017.
The NRC staff conducted an audit to examine NuScales helical steam generator tube modal testing related to the reactor internals comprehensive vibration assessment review. The onsite audit will take place at General Vessels in Cremosano and SIET S.p.A in Piacenza, Italy. The audit was initiated on June 24, 2019, and concluded on September 30, 2019, in accordance with the audit plan in ADAMS (ML19158A444).
The purpose of the audit was to: (1) gain a better understanding of the NuScale design; (2) verify information; (3) identify information that may require docketing to support the basis of the licensing or regulatory decision; and (4) review related documentation and non-docketed information to evaluate conformance with regulatory guidance and compliance with NRC regulations.
CONTACT: Marieliz Vera, NRR/DNRL 301-415-5861
The NRC staff conducted the audit via access to NuScales electronic reading room. The audit was conducted in accordance with the NRC Office of New Reactors (NRO) Office Instruction NRO-REG-108, Regulatory Audits.
The publicly available version of the audit report is enclosed.
Docket No.52-048
Enclosure:
- 1. Audit Summary - (Non-Proprietary)
- 2. Audit Summary - (Proprietary) cc: NuScale DC ListServ
Package: ML19343C747 Audit Summary Report - Public: ML19343C756 Audit Summary Report - Non-Public: ML19343C755
CSmith*
TLupold/SBailey*
MVera*
DATE 12/16/2019 12/18/2019 1/08/2020 1/08/2020
U.S. NUCLEAR REGULATORY COMMISSION
SUMMARY
AUDIT REPORT OF NUSCALE POWER, LLC, HELICAL COIL STEAM GENERATOR TUBE TESTING AT SIET LABORATORY June 24, 2019 - September 30, 2019 List of Participants NRC:
Yuken Wong, Senior Mechanical Engineer, Audit Lead Stephen Hambric, (NRC Consultant)
Marieliz Vera Amadiz, Project Manager NuScale:
Rich Danforth Olivia Hand Tamas Liszkai Hannah Rooks Maggie Wang Craig Langley Elisa Fairbanks Nadja Joergensen SIET:
Roberta Ferri Stefano Botti Andrea Achilli Carlo Randaccio Stefano Gandolfi Angelo Pozzi (Wintek)
Emanuelle Bianchessi (General Vessels)
1.0 BACKGROUND
Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Section 47, Contents of applications; technical information, states that:
The application must contain a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that construction conforms to the design and to reach a final conclusion on all safety questions associated with the design before the certification is granted. The information submitted for a design certification must include performance requirements and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the [U. S. Nuclear Regulatory Commission] NRC, and procurement specifications and construction and installation specifications by an applicant. The Commission will require, before design certification, that information normally contained in certain procurement specifications and construction and installation specifications be completed and available for audit if the information is necessary for the Commission to make its safety determination.
On March 15, 2017, the U.S. Nuclear Regulatory Commission (NRC) accepted and docketed a standard design certification application (DCA) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17074A087) submitted by NuScale Power, LLC (NuScale), to certify its small module reactor design.
NuScale has submitted Technical Report (TR) TR-0716-50439, NuScale Comprehensive Vibration Assessment Program Analysis Technical Report, Revision 2, issued July 2019 (ADAMS Accession Nos. ML19212A777 (proprietary) and ML19212A776 (non-proprietary)),
which is referenced in DCA Part 2 Tier 2, Section 3.9.2.3, Dynamic Response Analysis of Reactor Internals under Operational Flow Transients and Steady-State Conditions, and Section 3.9.2.4, Flow-Induced Vibration Testing of Reactor Internals before Unit Operation.
NuScale has also submitted TR-0918-60894, NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, Revision 1, issued August 2019 (ADAMS Accession Nos. ML19214A250, ML19214A251, and ML19214A253 (proprietary) and ML19214A248 (non-proprietary)).
Previous versions of the comprehensive vibration assessment program (CVAP) analysis (TR-0716-50439) and measurement and inspection (MIP) (TR-0918-60894) technical reports did not contain sufficient information for the NRC staff to make a safety finding regarding the potential effects of flow-induced vibration (FIV) on the NuScale Power Module (NPM) RVI and Helical Coil Steam Generator (HCSG). Therefore, two earlier audits were conducted, in which the NRC staff examined NuScales internal documents, drawings, and test data. The audits were performed from May 16, 2017, through November 2, 2017, and September 5, 2018, through October 4, 2018. Detailed audit reports are available in ADAMS under Accession Nos. ML18023A091 (non-proprietary; ML18022A377 (proprietary) and ML18333A221 (non-proprietary); ML18333A222 (proprietary). This audit focused on the NuScale HCSG benchmarking and validation test facilities at SIET Laboratory in Piacenza, Italy. The results of this audit assisted in the NRC staffs decision regarding if NuScale HCSG tube flow testing data are needed to support the design certification and the subsequent safety finding of the adequacy of the HCSG tube design against FIV.
The NRC staff provided NuScale with an audit plan to facilitate the audit (ADAMS Accession No. ML19158A444). The NRC staff followed the NRO Office Instruction NRO-REG-108 (Revision 0), Regulatory Audits, in performing the audit of the NuScale design specifications.
At the SIET facility and using the electronic reading room from June 24, 2019, to September 30, 2019, NRC staff members from the Mechanical Engineering Branch of the Division of Engineering and Infrastructure in the NRC Office of New Reactors and NRC consultants conducted a regulatory audit of the NuScale helical coil steam generator tube flow-induced vibration test and analysis. The NRC staffs observations and findings are documented in the Audit Results section.
2.0 DOCUMENTS REVIEWED EC-A014-7439, Fluid Forces on the Steam Generator Tubes, Revision 0, dated September 9, 2019.
Presentation LO-0619-66059 TF-3 Modal Demonstration Testing and SIET Tour, dated June 27, 2019 (ML19182A046 (non-proprietary) and ML19182A047 (proprietary)).
3.0 AUDIT RESULTS Overview The main focus of the SIET audit was to establish reasonable assurance that vortex shedding (VS) lock-in and fluid-elastic instability (FEI) will not occur in the NPM HCSG tubing. The SIET TF-1 and TF-2 testing has been evaluated previously (ML18333A221), but the facilities were examined while on-site and new data analysis and accompanying calculations were presented by NuScale at the audit. The remainder of the audit focused primarily on the status of the TF-3 test facility, and structural dynamic modal testing in particular. Planned future flow testing in TF-3 was also discussed with NuScale. Finally, NuScale assessments of the effects of the thermal expansion regarding the expected tightness of the fit of the HCSG tubing during plant operation, were discussed.
Previous discussions with NuScale led to the NRC staffs evaluation of the likelihood of an occurrence of HCSG VS and FEI in Section 3.9.2 of the Phase 2 Safety Evaluation Report (SER), to be based largely on the expected tightness of the fit between the tubing and tubing supports, which use arrays of tabs connected to support structures to hold the tubing. Tightly fitting tubes supported by more than one tab would effectively have clamped boundaries, significantly stiffening the tube modes (thereby increasing resonance frequencies) and increasing margin against VS and FEI1. NuScale had previously cited thermal expansion as a mechanism which would tighten the tube to support the fit, ensuring the clamped boundaries.
NuScale had also developed a preloading system for the TF-3 test structure which would press the supports and tubes together, emulating the expected thermal expansion effects.
Unfortunately, the assumption of clamped supports over most tubes is not substantiated by the audit results. An alternative means of establishing reasonable assurance of no HCSG FEI is therefore discussed at the end of this report and in the Phase 4 SER, which requires a DCA Part 2 Tier 1 commitment (see the Conclusions section).
TF-1 and TF-2 Test Facilities and Data Analysis Update The NRC staff toured both facilities at SIET, a small company of about 20 employees that specializes in safety testing, evaluation, and accreditation of the thermal hydraulic behavior of nuclear power components and systems. SIET is housed in a retired power plant in Piacenza, Italy. SIET is accredited by the Italian government and other agencies. The NRC audited SIET on quality assurance procedures associated with the TF-1 and TF-2 facilities in 2013 (ML14023A613). SIET designs and constructs its own test facilities and articles and calibrates its own instrumentation.
One of SIETs main facilities is its GEST Vessel (30 m high), which provides up to 8 kg/s of 1 The expected vortex shedding and fluid elastic instability frequencies are expected to be lower than the resonance frequencies of the tube modes. Higher tube mode resonance frequencies therefore, increase the ratio of resonance to excitation frequency (the safety margin).
superheated steam (maximum pressure and temperature of 10 MPa and 500 C), 40 kg/s of saturated steam (maximum pressure of 7 MPa), or 200 kg/s of water (maximum pressure and temperature of 17 MPa and 330 C). SIET also uses its IETI and SPES facilities (maximum power and pressure of 7 MW and 20 MPa) for testing and accrediting passive devices (like valves and heat exchangers), pumps, and special devices and instrumentation. SPES was used to test and evaluate the passive safety system response of the Westinghouse AP600 two loop pressurized water reactor due to postulated loss-of-coolant accidents in the early 1990s.
SIET has also worked for other nuclear companies, including Mitsubishi Heavy Industries and Toshiba. Finally, SIET is currently working on testing an anti-freezing system for liquid metal cooled reactors.
TF-1 is installed in the IETI facility. Three HCSG tubes are insulated and electrically heated to emulate the thermal hydraulics of the secondary coolant as it progresses upward. The inlet resistance was controlled by a needle valve and was adjusted to control density wave oscillation (DWO) behavior. All DWO frequencies are very low (much less than 1 Hz) and will not excite any structural or acoustic modes in the HCSG or other reactor vessel internals.
TF-2 is installed in the GEST facility. TF-2 models five columns of HCSG tubes, and like TF-1, was primarily tested to evaluate thermal hydraulic behavior of both primary and secondary coolant. However, the TF-2 tube inlets, outlets, and supports differ from the current design, which limits the facilitys usefulness for evaluating VS and FEI in the actual design. Long inlet and discharge manifolds feed and extract secondary coolant (steam) from the tube columns, so that discharge piping shapes are non-prototypic, and therefore not suitable for VS evaluations.
The TF-2 tube support assemblies are separated by [ ] degrees and are single point supports, whereas the current NuScale design uses eight support assembly locations, and up to three tabs could provide radial and vertical boundaries at each tube support.
Testing was performed under many conditions, with varying primary and secondary flow rates, pressures, and temperatures. The highest flow rate tested was [ ] that of the design, and no signs of VS or FEI were present in the vibration data.
NuScale continued to state that the lack of FEI behavior in TF-2, helps to support a safety finding for the actual design. However, the differences between the TF-2 and actual design discussed above, reduces confidence in NuScales reasoning. Therefore, the NRC staff cannot rely on TF-2 data and analysis to establish margin against FEI. However, NuScales best practices design approach and commitment to TF-3 flow testing prior to NPM initial startup should provide reasonable assurance of no FEI in the HCSG, provided sufficiently strong commitments are included in the design certification application (see further discussions below).
TF-1 and TF-2 Density Wave Oscillation Testing Steam generator inlet flow restrictors (SGIFRs) were not installed in the TF-1 and TF-2 test facilities. The TF-1 facility measured thermal hydraulic behavior explicitly, including DWO.
DWO behavior was observed but occurs at a very low frequency (much less than 1 Hz), and cannot excite any significant vibrational modes. DWO was also observed in TF2.
Based on concerns raised by the Advisory Committee on Reactor Safeguards (ACRS) at the June 19, 2019, NuScale design certification Full Committee meeting, NuScale was questioned regarding the TF-1 and TF-2 DWO behavior during the audit. NuScale stated that SGIFRs at the inlet to each actual SG tube design will provide the necessary secondary-side pressure drop to reduce DWO to acceptable limits. The SGIFR concept test data showed that the selected SGIFR design, provide the necessary secondary-side pressure drop that is expected to limit DWO. The minor design differences between the SGIFR final design and the tested design will have insignificant effect on the pressure drop and DWO.
TF-2 Benchmarking Update Three of the five TF-2 tubes were instrumented with strain gages, but only limited (five second) windows of data were acquired for each test condition. The TF-2 strain data have very high noise floors (due to the very short tests). Also, the data for boiling secondary flow conditions are corrupted by slowly oscillating naturally occurring temperature gradients (well below any structural resonance frequencies), which artificially elevate low frequency spectral levels.
NuScale attempted to remove these signals from its strain gage data to better compare measured spectra to those simulated using NuScales FIV analysis procedures.
NuScales TF-2 turbulent buffeting (TB) calculations were nearly complete at the time of the audit. NuScale applied both literature-based bounding forcing functions for internal and external flows and TF-1-based internal flow forces to structural finite element (FE) models of the TF-2 tubes. The simulated spectral peaks at low frequency tube resonances are significantly stronger than those measured, due to the conservatively high literature-based forcing functions.
The simulated tube vibrations based on TF-1-based internal forces are below the noise floor of the TF-2 measurements. It is therefore unlikely that the spectral peaks in the TF-1 internal pressure data (see ML18333A221 for further discussion of this topic) will cause significant FIV in the HCSG tubing. The updated calculations were later submitted as part of the revised CVAP MIP technical report (TR-0918-60894, Revision 1) in August 2019. The same conservative TB forcing function methodologies have been used to update design calculations, and to make pre-test predictions for TB-induced vibrations for TF-3.
Thermal Expansion Update NuScale presented a thermal expansion calculation of selected tubes and supports at the audit.
Two cases were analyzed; one with radial and vertical supports at the tabs, and the other with no supports, allowing free expansion. The tops of the tubes and support structures were fixed, since they are attached to large stiff structures. Thermal strains expand the tubes and support structures pushing them together. NuScale stated that the volumetric expansions are small with respect to the allowable tube to tab gaps, and that displacements due to thermal strains lead to larger relative displacements.
The NRC staff asked NuScale how other forces might affect the tube to support contact, such as gravity, primary flow, Coriolis effects of the secondary coolant, and internal and external static pressures. NuScale later quantified these effects in internal document EC-A014-7439 and compared them to those of the thermal expansion. The net effects of thermal and operational loading should induce contact between most, if not all, tubing and adjacent supports in the vertical and/or radial directions. However, TF-3 flow testing includes only two of these effects; flow and gravity (partially, since there is no water inside the tubes), which is conservative (leading to potentially looser support conditions and lower resonance frequencies) and is therefore, acceptable.
TF-3 Status of construction The entire TF-3 test structure (housing, supports, tube arrays) is being constructed at General Vessels (GV) near Cremosano, Italy. GV specializes in pressure vessel and heat exchanger manufacturing. Currently, the two outermost tube columns (13 and 12) have been constructed and installed, with three more columns pending (11-9). For structural dynamic/modal testing the entire vessel is oriented horizontally and rests on rollers, so that it may be rotated to test different sections. A platform extends through the vessel for engineers to adjust instrumentation and perform impact hammer testing.
The TF-3 is being built using nearly the same construction and assembly methods planned for the prototype. This is the first attempt to build a NuScale HCSG (at least part of one), and as a result, some key aspects of tube/support interaction have now become apparent. First, the tube to tube support (via triads of tabs) tightness of the fit is highly variable, with a roughly equal amount of tubes considered loose and tight. The eight circumferentially oriented tube supports are not equally spaced, leading to four long-span regions and four short-span regions around a full 360-degree traverse of tubing. The short spans appeared generally tight, but many long spans could easily be moved back and forth between supports by hand. Second, NuScales preloading system, which combines internal rings with radial bolts pushing against the inner surfaces of the support structures and external pneumatic actuators pushing against the outer surfaces, does not tightly clamp all tubes into the supports, as was originally hoped, due to radial clearances between the tubes and support structures. The preloading system was intended to emulate thermal expansion of the tubes and improve tube to support tightness of the fit, but instead only tightens the fit between the support structures.
A 3D printed model of three tube columns and supports was provided to better understand the steam generator assembly and fitup. Each support structure has tab arrays to hold the tubes.
Each tube is inserted between a triad of tabs, one in the center on one side of a tube, and two outer tabs on the opposite side. The tabs are punched out of the support structure metal, and finish machined to remove any sharp edges. The center tab in each three-tab row is longer than the side tabs and fits into a vertical track machined into the back of the radially adjacent support structure. These tab/track combinations provide tangential (hoop) stiffness but allow vertical relative motion via sliding. The outer tabs press against the back walls of the next tube support to provide radial stiffness. Radial stiffness is increased with the preloading system, but only of the support to support connections. To date, no means of tightening the tube to tab tightness is available. Therefore, NuScale stated that it is planning flow testing, assuming the worst-case condition of as-installed loose tube to support connections. Since this is conservative, it is therefore acceptable.
Dynamic instrumentation and testing The inner tube column (12) was tested dynamically during the audit. Some column 12 tubes are instrumented with permanent internal accelerometers. Installation of columns 11-9 is pending, with tests planned for 11 and 9 (both accelerometers and strain gages will be included in these columns). The instrumented tubes are welded, with internal instrumentation wire fed out of the tubes to the Data Acquisition System (DAS). Plans are being made to ensure pre-tensioning of the internal instrumentation wire to minimize its impact on damping. The tubes are bent into curved and pitched sections to fit into the steam generator. After assembly, temporary accelerometers are also installed using clamped mounts with small metal blocks.
All testing during the audit was performed by SIET engineers under NuScales direction. The DAS uses 24-bit digital signal processing and is based on National Instruments (NI) architecture. Data are acquired continuously, from both hammer and shaker excitation, so that post-processing may be done afterwards, including removal of very low frequency contamination. On-site checks of damping are based on the half-power bandwidth (AKA 3 dB down) method. Post-processing using more accurate modal parameter estimation is done with a Labview library module.
Long spans of two tubes were selected for dynamic testing demonstrations for the audit; one loosely fitting and the other tightly fitting. Short spans have much higher resonance frequencies with significant margin against FEI. Impact hammer testing of the loose and tight tubes allowed initial scoping of resonance frequencies and mode shapes. The hammer was instrumented so that transfer functions of acceleration and applied force could be computed. Peaks in the transfer functions show the tube resonances clearly.
The lowest frequency modes are dominated by vertical motion (along the axis of the overall steam generator and in the direction of flow), as expected (this is the lowest stiffness direction; modes in the radial direction are stiffened by the curvature of the tubing). An electrodynamic shaker applied more controlled input forces to evaluate the dependence of resonance frequencies and damping on vibration amplitude. However, audible rattling occurred during the high amplitude testing. This rattling led to nonlinearities in the structure resonant response, shifting the resonance frequencies themselves as the particular tabs in contact with the tubes changed over time. Resonance frequency shifts make damping estimation extremely difficult (half power bandwidth calculations are meaningless with a shifting resonance frequency). See below for more details on damping assessments.
NuScale exercised the ANSYS FE models of the tubes to further investigate the observed resonance frequencies for low excitation amplitudes (linear behavior, with the tube to tab contact remaining constant throughout the test). The loosely fitting tube span has lower resonance frequencies which correspond well to ANSYS calculations with a pivot boundary condition assumed. Pivots restrict translational motion (except in the tube axis direction), but not rotational motion. The tightly fitting span has higher resonance frequencies, which correspond better to clamped boundary conditions, which restrict rotational motion since more than one tab actively restricts tube motion. These clamped tubes have much higher resonance frequencies, providing improved margin against FEI. Both boundary conditions were evaluated in the updated analysis CVAP technical report (TR-0716-50439, Revision 2), with the most conservative critical velocities reported.
The dynamic measurements show clear groups of modes clustered together, each mode with a different resonance frequency. These resonance frequency differences are due to both mass loading of the instrumented tube span by the accelerometers and mounting blocks, but more importantly due to variability of the tube to tab support locations. The manufacturing and installation variabilities in the tube to tab geometries lead to many possible combinations of tube to tab support locations, and therefore many possible effective span lengths. NuScale adjustments of active tab locations in its FE model show similar variability in resonance frequencies.
The compressive preloading system was exercised, and the loose column was retested. The preloading shifted resonance frequencies only slightly upward, by less than 10 percent, and slightly increased apparent damping. The system does not, however, generate clamped boundaries for loosely fitting tube spans. It is uncertain whether NuScale will continue using the preloading system for future flow testing.
The inlet tubing (potentially susceptible to VS since downstream flow is not obstructed), is doubly curved and stiff, and measured resonance frequencies are higher than those of the fairly straight long-span regions, and also higher than the vortex shedding frequencies, increasing margin against lock-in. NuScales ANSYS model generally matches the measured frequencies, when adjusting which tab(s) actively support the tubing. NuScale has updated its calculations which show improved margin against HCSG VS lock-in. With this improved margin, NuScale later reduced assumed damping for VS assessments to 1 percent as recommended in Regulatory Guide 1.20.
NuScale plans to justify the use of 1.5 percent damping for its HCSG FEI assessments using the TF-3 modal testing data. It provided test results that show damping increases significantly with increasing vibration amplitude. As vibration amplitude increases, damping increases initially due to frictional losses between the tube and supports (tabs or radial wall contact). As vibration further increases however, rattling is audible, and energy is lost due to repeated contact. Also, as stated above, the resonance frequencies shift sporadically during testing, leading to wide apparent peaks in the processed vibration spectra, which is inappropriate for damping estimation using a half-power method. While these data definitely support damping higher than 1 percent, they are being acquired under loose tube/support fit conditions.
Additional damping measurements will be made of the full test assembly, and during FIV flow testing to justify the FEI 1.5 percent damping assumption.
TF-3 plans for flow testing NuScale considered modal characterization of the tube columns critical for establishing ranges of flow testing needed to identify onset of FEI. Modal testing will be performed without preloading, and will be considered bounding, since some tube long spans may have pivot supports (leading to low resonance frequencies). NuScale has not yet decided whether to use the preloading system during flow testing, since its effects are minimal and do not provide prototypic tubing boundary conditions.
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Given the NRC design certification review schedule, very little of the TF-3 data will be available to support a safety finding. Therefore, a clear definition of VS and FEI onset and necessary safety margins were later defined in DCA Part 2, Tier 1 and TR-0918-60894, Revision 1. These margins included estimated bias factors between the TF-3 test facility and the NPM, as well as analysis and testing uncertainties. The NRC staff evaluated these margins and the supporting documentation, and found them to be acceptable in Section 3.9.2 of the Phase 4 SER.
4.0 CONCLUSION
The NRC staff is satisfied with NuScales previous TF-1 and TF-2 testing, and with plans for TF-3 testing with regard to TB, FEI, and VS. Key conclusions from the audit are below.
SIET and NuScale are well qualified to perform the TF-3 testing. The manufacturing company (General Vessels) is also well qualified to construct selected components.
Both have significant prior experience working with heat exchangers and nuclear power plant components.
Thermal expansion, coupled with additional forces caused by gravity, primary coolant flow, static pressures, and Coriolis forces induced by helically swirling secondary flow, should improve tightness of tubing to support the fit. This will increase resonance frequencies, increasing the margin against FEI, and reduce the possibility of any unsupported locations. However, the improved tightness of the fit in the actual NPM will also reduce damping due to reduced frictional losses and reduced likelihood of rattling.
This effect will not be known prior to NPM initial startup testing. However, regions of the HCSG will be instrumented during initial startup to ensure no significant FIV mechanisms occur.
Modal testing of the current state of the TF-3 test structure reveals loose fits between tubing and supports. Long span tube regions exhibit either pivot or clamped conditions, which agree well with ANSYS FE model results. However, there are many loose long spans, leading to many lower frequency resonances which affects margin against FEI.
However, the actual margin will be established during TF-3 flow testing in the future.
VS/lock-in in the HCSG is unlikely due to the higher than expected resonance frequencies of the HCSG inlet tubes.
The TF-3 testing is vital in confirming that VS and FEI does not exist in the NuScale HCSG. NuScale is also motivated to establish the FEI onset conditions of the HCSG to support future power uprate applications. In the DCA Part 2 Tier 1, NuScale made the following commitment:
TF-3 is the test facility designed to study fluid elastic instability, vortex shedding, and turbulence due to primary side flow in helical steam generator tubes. Testing consists of modal testing in air and in water, and primary side flow testing with extensive instrumentation to detect vibration.
Prototypes of the SG assembly will undergo TF-3 testing and meet the acceptance criteria in accordance with the Initial Test Program Steam Generator Flow-Induced Vibration Test. The results of the testing will be reviewed and approved in accordance with the NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report prior to loading fuel in the first ever NPM. This one-time testing satisfies TF-3 testing requirements for subsequent NPMs built in accordance with the approved design.
NuScale has followed best design practices to minimize the likelihood of FEI. The TF-3 test facility will be a conservative representation of the actual HCSG conditions, since thermal expansion effects (which will improve tightness of the fit) are not included. Also, the TF-3 facility allows flow speeds which far exceed what will be possible (and allowable) in an actual NPM, allowing NuScale to establish margins against FEI, should it occur. This best practice approach, combined with a Tier 1 commitment to require delivery of TF-3 test data validating margin against FEI and VS, provides reasonable assurance that the HCSG will not experience significant FIV-induced damage during the life of a NPM.