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Nuscale - ACRS Subcommittee Presentation, FSAR, Steam Generator Design
	ML20035C789
Person / Time
Site:	NuScale
Issue date:	02/04/2020
From:	; NuScale
To:	; Office of Nuclear Reactor Regulation
References
	LO-0120-68573 PM-0220-68568, Rev 0
	Download: ML20035C789 (16)
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Presenters Kent Welter Engineering Chief, Testing and Analysis Kevin Spencer Mechanical Engineer, NSSS Engineering Joe Remic Supervisor, NSSS Component Analysis Brian Wolf Supervisor, Code Development Marty Bryan Licensing Project Manager 2

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Steam Generator Design - Open Session Agenda

Steam Generator Design

Steam Generator Inspection Program

Density Wave Oscillation Overview

ITAAC Closure Path for DWO

Preliminary Scoping Study

Planned Closure Activities Through ITAAC 3

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Steam Generator Design

Integral Helical Coil SG Design features

- Shell side is primary side - Tube side is secondary side

- Alloy 690 TT (1380 tubes, 77 - 87ft long, 5/8" OD)

- Low flow in primary (-1ft/sec)

- Tube wall degradation allowance (0.01 O" > ASME min wall)

- Support 100°/o volumetric inspection

- Normal access to shell side of tubes from below during refueling

Incorporation of Operating Experience

- Follow guidance of NEI 97-06 & EPRI (COL Item 5.4-1: Develop and implement a SG Program)

Flow restrictor design at SG tube inlet ensures acceptable tube flow fluctuations during operation 4

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Steam Generator Design (Cont'd)

UPPER SGSU y

SGSU ORT 5

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Steam Generator Inspection Program

Monitors performance and condition of SGs as part of ISi

Implements applicable portions of B&PV code Section XI and 10 CFR 50.55a(g)

Appendix B to 10 CFR 50 applies to implementation

Follows NEI 97-06 and EPRI guidance 1013706 Rev 7 "Steam Generator Management Program, PWR SG Examination Guidelines"

Includes

- Degradation assessment (including wear due to fretting)

- Tube integrity assessment

- Shell side integrity assessment 6

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Density Wave Oscillation - Overview

Consider a case for inlet flow reduction

- sinusoidal flow fluctuation Tota l Channe l Pressure D-op

Causes increase in void

Voids move through the channel Loca I Pressure [)'"ops

Voids movement also known as Density Wave Motion /J \

Density wave produces pressure drop fluctuations

Pressure drop is delayed with respect to the inlet flow fluctuations Inlet

At certain frequency Density Wave Time ~lay

- Total pressure drop is completely out of phase with the flow fluctuations Inlet Flow

- Effectively produces the negative pressure drop TIME

- Causes flow surge (March-Leuba, J., "Density Wave Instabilities in BWR", NUREGICR-6003, Oct. 1992)

- Cycle repeats

DWO can be self sustaining 7

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Density Wave Oscillation - Overview (continued)

DCA Rev. 3 Section 3.9.1 discusses the possibility of secondary flow oscillations during power ascent and descent

Section 5.4.1.2 addresses Inlet Flow Restrictor (IFR) sizing to limit DWO to acceptable limits

The NuScale Stability Topical Report TR-0516-49417 concludes that secondary oscillations do not challenge fuel thermal limits

Additional review of the DCA language is ongoing and adjustments will be made in a final revision as necessary 8

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ITAAC Closure Path for DWO

ITAAC 02.01.01 requires that an inspection is performed of the NuScale Power Module "ASME Code Class 1, 2, 3, and CS as-built component Design Reports to verify that the requirements of ASME Code Section Ill are met"

Tier 1 Table 2.1-2 defines the NuScale Power Module (NPM) ASME Code Class 1, 2, 3, and CS components:

Equipment Name ASME Code Section Ill RCS Integral RPV/SG/Pressurizer 1

Therefore, ITAAC 02.01.01 requires the inspection of the certified ASME Design Reports for the RPV and the steam generator 9

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ITAAC Closure Path for DWO (continued)

Subsection NCA of the 2013 Edition of the ASME Code defines requirements of what is to be included in ASME Design Specifications and ASME Design Reports.

- Design Specifications

NCA-2142.2 requires that Design Specifications identify all loadings (e.g. pressure, temperature, mechanical loads, cycles, and/or transients) and the service limits a component will experience

>>, Loading combinations for the RPV (including SG tubes) defined in Table 3.9-3 of DCA

>> Transient (TH) loads are based on time history of design basis transients, described in DCA Section 3.9.1.

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ITAAC Closure Path for DWO {continued)

Subsection NCA of the 2013 Edition of the ASME Code defines requirements of what is to be included in ASME Design Specifications and ASME Design Reports.

- Design Reports

NCA-3260 requires that the Design Report evaluate the loads and load combinations as defined in the Design Specification to the applicable acceptance criteria are met.

NCA-5350(4) requires the Design Report be certified by a Registered Professional Engineer (RPE) competent in the applicable field. The RPE confirms all loads and load combinations specified in the Design Specification have been addressed and satisfied.

NCA-831 O(a) permits application of the ASME 'N' Certification Mark to the Reactor Pressure Vessel which provides certification by an Authorized Nuclear Inspector that all ASME examinations and testing have been completed.

Therefore, the resolution of DWO will be achieved through ITAAC 02.01.01 activities related to the RPV and the steam generator 11 PM-0220-68568 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Preliminary Scoping Study

Steam Generator Tubes

- The applicable primary side conditions were applied to the exterior of the SG tubes and DWO full flow reversal was considered on the interior (secondary side) for the

total length of the tube.

- The resultant alternating stress due to this oscillation is below the ASME Code endurance limit; therefore, preliminary analyses demonstrate that DWO full flow reversal will not result in any fatigue damage to the steam generator tubes.

Feedwater Plenum Tube-to-Tubesheet Weld

- Preliminary analyses using bounding DWO transient definitions, assuming complete tube dry out, and conservative tube to plenum interaction, resulted in alternating stresses above the ASME Code endurance limit.

The DWO transient definition in this vicinity is undergoing further evaluation.

The preliminary DWO alternating stress in relation to the ASME endurance limit, coupled with the analytical conservatisms applied in this evaluation, enable NuScale to confidently predict the final alternating stress due to DWO in this region will be below the ASME Code endurance limit.

ITAAC 02.01.01 requires the inspection of the Design Report in which this evaluation will be documented.

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Planned Closure Activities Through ITAAC

ITAAC 2.01.01 inspection is performed of the NuScale Power Module "ASME Code Class 1, 2, 3, and CS as-built component Design Reports to verify that the requirements of ASME Code Section Ill are met"

- NRELAPS will provide thermal-hydraulic time histories of design transients in support of the fatigue analysis, including fluid temperatures, pressures, nominal flow rates, and quality

- TF-1 and TF-2 test data will provide a basis for parameters used to calculate film coefficients and resulting stresses during DWO

- Thermal response of the tube is much faster than the DWO frequency, so a quasi-static analysis can be performed

- The stresses from all loadings and transients (including those from DWO) will be combined and compared to allowable stress limits

- The goal of the SG ASME Code calculations is to confirm that the alternating stresses on the SG tubes, the tube-to-tubesheet welds, and the tubesheets due to DWO stresses are below the ASME endurance limit; thereby enabling these components to withstand infinite cycles 13 PM-0220-68568

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. NuScale Conclusion

The successful completion of ITMC by the licensee constitutes the basis for the NRC determination to allow operation of a facility certified under 10 CFR 52 14 PM-0220-68568 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Acronyms ASME American Society of Mechanical PWR Pressurized Water Reactor Engineers RCS Reactor Coolant System B&PV Boiler and Pressure Vessel RPV Reactor Pressure Vessel BWR Boiling Water Reactor SG Steam Generator CFR Code of Federal Regulations TH Code Thermal Hydraulic Code DCA Design Certification Application DHRS Decay Heat Removal System DWO Density Wave Oscillation EPRI Electric Power Research Institute FSAR Final Safety Analysis Report FW Feedwater IFR Inlet Flow Restrictor ISi Inservice Inspection ITAAC Inspection, Test, Analysis and Acceptance Criteria NEI Nuclear Energy Institute NPM NuScale Power Module NSSS Nuclear Steam Supply System 15 PM-0220-68568

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