ML24215A121
| ML24215A121 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 08/02/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24215A000 | List:
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| References | |
| LO-169995 | |
| Download: ML24215A121 (1) | |
Text
Response to SDAA Audit Question Question Number: A-5.4.1.3-1 Receipt Date: 09/15/2023 Question:
The NRC staff noted that Section 5.4.1.3 in the SDAA points to TR-121353 for the SG Comprehensive Vibration Assessment Program screening and performance analysis. Please provide a clear summary in the FSAR of the screening and performance analysis results performed for the SG Comprehensive Vibration Assessment comparable to the summary for the Stability Performance discussion in Section 5.4.1.3.
Response
The information provided in Section 5.4.1.3 of the Standard Design Approval Application (SDAA) about the Comprehensive Vibration Assessment Program (CVAP) technical report (TR-121353) is equivalent to the information provided on the stability analysis topical report (TR-0516-49417). Section 5.4.1.3 of the SDAA identifies both reports with their appropriate reference.
The information provided in Section 5.4.1.3 of SDAA regarding the stability analysis topical report (TR-0516-49417) is not a summary of the stability analysis topical report but rather contains additional information on how the stability topical report addresses density wave oscillations (DWO). The attached markup to SDAA moves that information to a new paragraph to avoid confusion.
Information on the CVAP technical report is available in TR-121353, which is part of SDAA.
Section 2.3.1 and Section 2.3.2 of TR-121353 contain information on screening for flow induced vibration (FIV) phenomena applicable to steam generator system components. Performance analysis results for the FIV phenomena specific to the steam generator tubes is in the following sections of TR-121353:
NuScale Nonproprietary NuScale Nonproprietary
Section 3.2.1 for fluid elastic instability
Section 3.2.2 for vortex shedding
Section 3.2.3 for turbulent buffeting Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-7 Draft Revision 2 describes the compatibility aspects of the secondary coolant chemistry that provide corrosion protection for stainless steels and nickel alloys, including the SG components exposed to the secondary system coolant, and Section 10.3.5, Water Chemistry, describes the secondary water quality control program. The SGs are flushed during NPM startup and shutdown to establish initial chemistry for power operations or refueling.
Section 11.1.2, Design Basis Secondary Coolant Activity, addresses estimated radioactivity design limits for the secondary side of the SGs during normal operation. The radiological effects associated with an SG tube failure are in Section 15.0.3, Design Basis Accident Radiological Consequences.
5.4.1.3 Performance Evaluation The RCS natural circulation flow loop is entirely within the RPV, thereby eliminating distinct RCS piping loops and the associated potential for a large pipe break (i.e., large break LOCA) event. This design, combined with the intertwined SGs tube bundle configuration, eliminates the potential for asymmetric core cooling and temperatures as a result of a loss of a single SG function. Isolation or other loss-of-heat transfer capability by either of the two intertwined SGs does not introduce asymmetrical cooling in the reactor coolant system because the tube configuration of the remaining functional SG continues to provide symmetrical heat removal from the reactor coolant flowing in the downcomer of the reactor vessel.
The primary coolant system operates at a higher pressure than the secondary system, resulting in the SG tubes being in compression. This configuration reduces the likelihood of a tube failure and eliminates the potential for pipe whip due to tube-side jetting.
Feedwater enters the SG tubes at their lowest point. As it rises through the tubes, it undergoes a phase change and heats above saturation temperature before exiting the SG tubes as superheated steam. The configuration keeps the steam-water interface fluid, and the superheated steam at the top of the tubes separated from the subcooled liquid at their bottoms. This configuration minimizes the hydraulic instabilities that could introduce potential sources of water hammer.
Stability Performance Flow instabilities, such as density wave oscillation, may arise in individual SG tubes because of fluid brought to boiling conditions as it travels up the tubes. Inlet flow restrictors at the FW inlet plenum interface provide the necessary pressure drop to preclude unacceptable secondary flow instabilities. Acceptable instabilities are tube mass flow fluctuations that do not cause reactor power oscillations that could exceed fuel design limits, and that result in applicable ASME BPVC criteria being met.
Stability analyses are documented in TR-0516-49417, Evaluation Methodology for Stability Analysis of the NuScale Power Module (Reference 5.4-9).
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-8 Draft Revision 2 Audit Question A-5.4.1.3-1 The stability analysis documented in Appendix A of Reference 5.4-9 shows that the main effect of density waves in the tubes of the helical coil SGs is a small reduction in the effective heat transfer coefficient between the two sides of the SG. The unstable flow oscillations impact on heat transfer in individual tubes does not affect the overall heat transfer to the primary side because the flow oscillations in the tubes are not in-phase and thus their individual effects cancel out.
Significant primary flow oscillations are not excited by the instabilities in the SG tubes.
Analyses regarding the susceptibility of the NPM to develop DWO conditions use the approach documented in Appendix B of TR-131981-P, Methodology for the Determination of the Onset of Density Wave Oscillations (DWO),
Reference 5.4-11. Results show that the combination of operating conditions and inlet flow restrictor design allow for margin to DWO onset at all nominal power levels from 20 percent to 100 percent power, which is the power generation range for turbine operation. While DWO may occur during limited operational times at low power levels, the SG and inlet flow restrictor design assures that DWO transient conditions are acceptable to meet applicable ASME BPVC criteria.
Comprehensive Vibration Assessment Program Performance The results of the Comprehensive Vibration Assessment Program screening and performance analysis for the SG is in technical report TR-121353, "NuScale Comprehensive Vibration Assessment Program Analysis Technical Report,"
(Reference 5.4-10).
Section 17.4, Reliability Assurance Program, describes the reliability assurance plan used for SG reliability evaluation; the guidance in Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, describes the determination of SG risk significance.
5.4.1.3.1 Allowable Tube Wall Thinning under Accident Conditions The SG tubes have a nominal wall thickness of 0.050 in. The design adds a lifetime degradation allowance of 0.010 in. to the calculated ASME BPVC minimum SG tube wall thickness per NB-3121 (Reference 5.4-3). This degradation allowance provides margin for potential in-service tube degradation mechanisms (e.g., general corrosion, erosion, wear). This degradation allowance also includes margin for SG tube wall thickness manufacturing tolerances, including wall thinning due to tube bending. The SG tubes construction meets the rules of ASME BPVC,Section III, Subsection NB.
5.4.1.4 Tests and Inspections The SGs testing and inspection ensures conformance with the design requirements described in Section 5.2.4, RCPB ISI and Testing. Equipment