ML25013A221
| ML25013A221 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 01/13/2025 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25013A204 | List: |
| References | |
| LO-177590, LO-178078 | |
| Download: ML25013A221 (1) | |
Text
Question Number: DWO-SC-25 Receipt Date: 09/27/2024 Question:
Supporting Analyses (1) NuScale should describe in the FSAR the temperature approach method that will be used to monitor the time potentially in DWO conditions.
(2) NuScale should make available for staff review the basis for the approach temperature method, a representative demonstration of benchmarking the approach to appropriate data, and an example "stability map" or other operator aid.
Response
The attached markup to Standard Design Approval Application (SDAA) Final Safety Analysis Report (FSAR) Chapter 5.4.1.3 includes information about the approach temperature method that is used to monitor time potentially in density wave oscillation (DWO) conditions, including Figure 5.4-16, which is the approach temperature figure for the SDAA FSAR.
Engineering calculation EC-174500, Revision 1, DWO Approach Temperature Limit, is in the electronic reading room (eRR) and contains the basis, benchmarking, and an approach temperature map. Figure 5-8 of EC-174500 shows the approach temperature at DWO onset as a function of feedwater (FW) flow per steam generator (SG). The conclusions of EC-174500, Revision 1, pertinent to the SDAA FSAR are:
NuScale Nonproprietary NuScale Nonproprietary
((2(a),(c), ECI
The analysis demonstrates that the approach temperature limit line defines where DWO is precluded if RCS average temperature is above 520 degrees F and if main steam is superheated.
The NPM-20 reactor safety inputs to the module protection system for reactor coolant system hot temperature and main steam temperature are used to calculate approach temperature.
(( }}2(a),(c), ECI
(( }}2(a),(c), ECI This is defined as regions in the SDAA FSAR Figure 5.4-16, as shown in the attached markups. NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c),ECI The attached markups to SDAA FSAR Section 5.4.1.3 provide more information outlining how approach temperature relates to DWO in the steam generator tubes and include Figure 5.4-16. Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Material Referenced NuScale US460 SDAA 1.6-2 Draft Revision 2 Audit Item A-1.6-1, Audit Question DWO-SC-25, Audit Question DWO-SC-27, Audit Question DWO-SC-28 Table 1.6-1: NuScale Referenced Topical Reports Topical Report Number Topical Report Title Report Sections Incorporated by Reference Where Referenced (FSAR Section) MN-122626, Revision 21 NuScale Power, LLC Quality Assurance Program Description All 17.5 TR-1015-18653-P-A, Revision 2 Design of the Highly Integrated Protection System Platform All 7.0, 7.1, 7.2, 15.8, SDAA Part 7, Exemption 3 Chapter 7, Chapter 15 TR-131981-P, Revision 1 Methodology for the Determination of the Onset of Density Wave Oscillations 3.9, 5.4 TR-0516-49422-P, Revision 3 Loss-of-Coolant Accident Evaluation Model All 4.4, 5.2, 6.2, 6.3, 15.0, 15.6 TR-124587-P, Revision 0 Extended Passive Cooling and Reactivity Control Methodology All 5.4, 6.2, 15.0, 15.6, SDAA Part 4 TR-0920-71621-P-A, Revision 1 Building Design and Analysis Methodology for Safety-Related Structures Section B, Sections 3.0 - 7.0 3.5, 3.7, 3.8, 3B TR-0118-58005-P-A, Revision 2 Improvements in Frequency Domain Soil-Structure-Fluid Interaction Analysis Section B, Sections 3.0 - 8.0 3.7, 3A TR-0716-50351-P-A, Revision 1 NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces Section B, Sections 3.0 - 5.0 4.2 TR-0616-48793-P-A, Revision 1 Nuclear Analysis Codes and Methods Qualification Section B, Sections 2.0 - 8.0 4.3, SDAA Part 4 TR-0116-20825-P-A, Revision 1 Applicability of AREVA Fuel Methodology for the NuScale Design All 4.2, 4.3, 4.4, SDAA Part 7, Exemption 14 TR-108553, Revision 0 Framatome Fuel and Structural Response Methodologies Applicability to NuScale (Supplement 1 to TR-0116-20825-P-A, Revision 1; Supplement 1 to TR-0716-50351-P-A, Revision 1) All 4.2, 4.3, 4.4, SDAA Part 7, Exemption 14 TR-0116-21012-P-A, Revision 1 NuScale Power Critical Heat Flux Correlations All 4.4, 15.0, 15.6 TR-107522-A, Revision 1 Applicability Range Extension of NSP4 Critical Heat Flux Correlation (Supplement 1 to TR-0116-21012-P-A, Revision 1) All 4.4, 15.0, SDAA Part 4 TR-0915-17564-P-A, Revision 2 Subchannel Analysis Methodology All 4.3, 4.4, 15.0, 15.4, 15.6, SDAA Part 4
NuScale Final Safety Analysis Report Material Referenced NuScale US460 SDAA 1.6-3 Draft Revision 2 TR-108601-A, Revision 4 Statistical Subchannel Analysis Methodology (Supplement 1 to TR-0915-17564-P-A, Revision 2) All 4.3, 4.4, 15.0, 15.4, 15.6, SDAA Part 4 TR-0915-17565-P-A, Revision 4 Accident Source Term Methodology Section B, Sections 3.0 - 5.0 2.3, 3.11, 3C, 12.2, 15.0, SDAA Part 4 TR-0516-49416-P, Revision 4 Non-Loss-of-Coolant Accident Analysis Methodology All 5.2, 6.2, 7.1, 15.0, SDAA Part 4 TR-0716-50350-P, Revision 3 Rod Ejection Accident Methodology All 4.3, 15.0, 15.4 TR-0516-49417-P-A, Revision 1 Evaluation Methodology for Stability Analysis of the NuScale Power Module Section B, Sections 3.0 - 10.0 4.4, 15.4, 15.0, 15.9 TR-0420-69456-NP-A, Revision 1 NuScale Control Room Staffing Plan None 18.5, SDAA Part 4 TR-0515-13952-NP-A, Revision 0 Risk Significance Determination Section D, Section 3.0 17.4, 18.6, 19.1, 19.3 Table 1.6-1: NuScale Referenced Topical Reports (Continued) Topical Report Number Topical Report Title Report Sections Incorporated by Reference Where Referenced (FSAR Section)
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-8 Draft Revision 2 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 Audit Question DWO-SC-25, Audit Question DWO-SC-27, Audit Question DWO-SC-28 Flow instabilities, such as density wave oscillation (DWO), 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). 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. Audit Question DWO-SC-25, Audit Question DWO-SC-27, Audit Question DWO-SC-28 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, tA comparison between RCS hot temperature and main steam temperature (Section 7.2, System Features) is used to determine the approach temperature, which correlates with in-tube conditions indicating the potential for DWO onset. Time is counted in DWO when operating in Region 1 in Figure 5.4-16 or when the RCS average temperature is less than 520 degrees F or there is not
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-9 Draft Revision 2 main steam superheat; this time is accrued against the Section 3.9.1 cyclic limits for ASME design transients. Operation in Region 2 occurs when RCS average temperature is greater than or equal to 520 degrees F with main steam superheat and indicates that there is no DWO in the SG tubes; therefore, time is not accrued against the Section 3.9.1 cyclic limits. The boundary between Region 1 and Region 2 in Figure 5.4-16 does not indicate that there is DWO in the SG tubes; there is margin from the boundary between Region 1 and Region 2 to DWO onset in the SG tubes. The SG and inlet flow restrictor design assures that DWO transient conditions are acceptable to meet applicable ASME BPVC criteria. Audit Question DWO-SC-22 Procedures that address actions to perform during possible DWO conditions are identified in Section 13.5.2, Operating and Maintenance Procedures. 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 requiring inspection or repair is in an accessible position to minimize time and radiation exposure during refueling and maintenance outages. The SG tube inspections and testing meet requirements of the SG program. Performance of a preservice volumetric examination on the entire length of the SG tubing meets specifications in Table IWB-2500-1 (B-Q). A preservice eddy
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-37 Draft Revision 2 5.4-8 American Society of Mechanical Engineers, "Forged Fittings, Socket-Welding and Threaded," ASME B16.11-2011, New York, NY. 5.4-9 NuScale Power, LLC, "Evaluation Methodology for Stability Analysis of the NuScale Power Module," TR-0516-49417-P-A, Revision 1. 5.4-10 NuScale Power, LLC, "NuScale Comprehensive Vibration Assessment Program Analysis Technical Report," TR-121353, Revision 0. Audit Question DWO-SC-25, Audit Question DWO-SC-27, Audit Question DWO-SC-28 5.4-11 "Methodology for the Determination of the Onset of Density Wave Oscillations (DWO)," TR-131981-P, Revision 1.
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-68 Draft Revision 2 Audit Question DWO-SC-25, Audit Question DWO-SC-27, Audit Question DWO-SC-28 Figure 5.4-16: Approach Temperature for NPM-20 0 5 10 15 20 0 100 200 300 400 500 600 700 800 900 Approach Temperature (degrees F) Feedwater Flow (gpm/steam generator) Region 2 Region 1}}