ML25058A253
| ML25058A253 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 02/27/2025 |
| From: | Shaver M NuScale |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| LO-180013 | |
| Download: ML25058A253 (1) | |
Text
LO-180013 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com February 27, 2025 Docket No. 052-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Supplement to Chapter 15 US460 SDA Revision 1 to Revision 2 Snapshot
REFERENCE:
Letter from NuScale to NRC, NuScale Power, LLC Submittal of Chapter 15 US460 SDA Revision 1 to Revision 2 Snapshot, dated February 26, 2025 (ML25057A294)
The purpose of this letter is to provide additional updates for NuScale Power, LLC (NuScale)
Chapter 15 US460 SDA Revision 1 to Revision 2 Snapshot that were not included in the reference above. The attachment to this letter contains additional NuScales updates to Chapter 15 of the Standard Design Approval Application (Revision 2).
The attachment is nonproprietary.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Amanda Bode at 541-452-7971 or at abode@nuscalepower.com.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 27, 2025.
Sincerely, Mark W. Shaver Senior Director, Regulatory Affairs NuScale Power, LLC Distribution:
Mahmoud Jardaneh, Chief, New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Stacy Joseph, Senior Project Manager, NRC
LO-180013 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com
- Supplement to Chapter 15 US460 SDA Revision 1 to Revision 2 Snapshot
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-69 Draft Revision 2 Table 15.0-18: Margin to Critical Boron Concentration for Emergency Core Cooling System Cooldown Events Event (Time in Cycle)
Minimum Margin to Critical Boron Concentration (ppm)(1)
Core Concentration at Time of Minimum Margin (ppm)
Time of Minimum Margin to Critical (seconds)
Injection line break in CNV, RVV fails (BOC) 514 1014 31 Injection line break in CNV, RVV fails (MOC) 538 589 31 Injection line break in CNV, RVV fails (EOC) 134203 5
14972 Reactor component cooling water pipe break (slow bias ESB)High point vent break outside CNV (EOC) 30102 26 151,92037,070 Reactor component cooling water pipe break (fast bias ESBEOC) 2825 128 105,960118,600 (1) For cases with the margin to critical boron concentration greater than the core boron concentration, criticality can not be achieved by boron dilution without other positive reactivity addition.
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-71 Draft Revision 2 Table 15.0-20: Limiting Collapsed Liquid Level above Top of Active Fuel during Emergency Core Cooling System Cooldown Event Minimum Collapsed Liquid Level above Top of Fuel (ft)
Time of Minimum Level (seconds)
Discharge line break inside CNV 3.03.13 84,52049,295 Discharge line break outside CNV 2.11.57 64,99059,930 Injection line break inside CNV 2.93.12 85,95146,385 Injection line break outside CNV 2.21.81 66,46971,975 High point vent break line inside CNV 3.13.18 86,25745,395 High point vent break line outside CNV 3.23.40 86,15532,165 SGTF 1.80.23 69,07069,740 Inadvertent RRV 3.23.17 79,21947,300 Inadvertent RVV 3.03.17 84,98242,905 Inadvertent RVV with loss of AC and EDAS power - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after event 3.14.00 N/A Inadvertent RVV with loss of AC and EDAS power - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after event 3.64.44 N/A Inadvertent RVV with loss of AC and EDAS power - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after event 4.85.24 N/A
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-73 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Table 15.0-22: Input Parameters for Emergency Core Cooling System Extended Passive Cooling Analysis - Limiting Boron Cases Parameter Boron Precipitation Boron Transport Single failure one RVV fails to open one RVV fails to open Power availablity AC and DC power available AC and DC power available Decay heat 8095% of ORIGEN 8095% of ORIGEN(1)
Core power 100% RTP 100% RTP Reactor pool level 54 ft 54 ft Reactor pool temp 65°F 65°F Noncondensable gas zero zeronegligible(1)
ECCS valve flow capacity biased low biased low Initial boron concentration 1900 ppm 5 ppm ESB dissolution biasing fast fast ESB mass per dissolver 30 kg boron oxide 25 kg boron oxide ESB boron oxide pellet molar mass 69.6 g/mol 69.6 g/mol CNV wall condensate collection area -
ESB dissolver 35 sq ft 35 sq ft CNV wall condensate collection area -
ESB mixing tube N/A 140 sq ft Notes:
(1) A range of decay heat is evaluated as a result of reduced-power operation before an event. The decay heat used in the limiting boron transport case is conservatively representative of a downpower to 0 percent power for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> followed by a 5-hour ascent to 50 percent power.A negligible amount of noncondensable gas is included for calculational stability.
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-79 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-5: Boron Transport Analysis Concentrations - Injection Line Break at End of Cycle Case
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-80 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-6: Boron Transport Analysis Masses - Injection Line Break at End of Cycle Case
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-81 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-7: Boron Transport Analysis Concentrations - Reactor Component Cooling Water Line Break Case (Slow Bias Supplemental Boron)High Point Vent Line Break Case
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-82 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-8: Boron Transport Analysis Masses - Reactor Component Cooling Water Line Break Case (Slow Bias Supplemental Boron)High Point Vent Line Break Case
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-83 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-9: Boron Transport Analysis Concentrations - Reactor Component Cooling Water Line Break Case (Fast Bias Supplemental Boron)
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-84 Draft Revision 2 Audit Item A-15.0.5-1, Audit Item A-15.0.5-2 Figure 15.0-10: Boron Transport Analysis Masses - Reactor Component Cooling Water Line Break Case (Fast Bias Supplemental Boron)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-22 Draft Revision 2 A thermal hydraulic analysis is performed to provide the limiting boundary conditions for the downstream subchannel analysis, which evaluates the final CHF value. The limiting MCHFR case results from a 1237-percent split break located outside containment between the CNV upper headMSIV and primarysecondary MSIV. The following initial conditions are assumed in the analysis of the limiting MCHFR case.
Initial power level is assumed to be 102 percent of rated thermal power to account for a 2-percent measurement uncertainty.
BeginningEnd of cycle reactivity feedback with a high fuel temperature bias.
Conservative reactor trip characteristics are used, including a maximum time delay, holding the most reactive rod out of the core, and utilizing a bounding control rod drop rate.
The beginning-of-life SG is assumed, with no SG tube plugging.
The single failure identified for this case is the failure of an MSFWIV to close.
Automatic control of the regulating control rods is disabled.
Automatic control of the turbine control valve is disabled.
The results from the thermal hydraulic evaluations are used as input to the subchannel analysis to determine the limiting MCHFR for this event. The subchannel evaluation model is discussed in Section 15.0.2.
15.1.5.3.3 Results Steam Line Break Case Resulting in the Limiting Reactor Coolant System Pressure The sequence of events for the limiting RCS pressure SLB case event is provided in Table 15.1-11. Figure 15.1-29 through Figure 15.1-35Figure 15.1-34 show the transient behavior of key parameters for this SLB case. The effects of the postulated SLB on other systems are considered in Section 3.6 consistent with Branch Technical Position (BTP) 3-3 and BTP 3-4. Most SLB events have a cooldown effect on the RCS, where there is little or no pressurization of the RCS. However, this SLB case represents a small break in the steam line that causes core power to increase until reactor trip on high power rate. Feedwater is terminated because of SSI actuation on low PZR pressuremain steam superheat, causing a subsequent RCS heatup and pressurization until the DHRS actuates on high PZR pressure. This scenario results in the highest RCS pressure for the steam piping failure events analyzed.
Because of reactivity feedback from the cooldown, core power rises until reactor trip. The reactor safety valve (RSV) lifts during the re-pressurization following SSI and maintains RCS pressure within limits as shown in Table 15.1-11 and Figure 15.1-30.
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-32 Draft Revision 2 Table 15.1-3: Decrease in Feedwater Temperature (15.1.1) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 21216 psia Maximum SG pressure 2420 psia 145475 psia MCHFR 1.435 1.7159
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-34 Draft Revision 2 Table 15.1-5: Sequence of Events for Limiting Decay Heat Removal SystemSteam Generator Overfill Case (15.1.2 Increase in Feedwater Flow)
Event Time [s]
FW flow begins to increase 0
Reactor trip and SSI actuation on low main steam superheat 158 SSI valves fully closed (single failure of one FWIV) 2518 DHRS actuation on high main steam pressure 3225 Maximum SG level (SG with failed FWIV) 12088 DHRS heat removal exceeds core decay heat 2335
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-36 Draft Revision 2 Table 15.1-7: Increase in Feedwater Flow (15.1.2) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 218891 psia Maximum SG pressure 2420 psia 15071471 psia MCHFR 1.435 1.695
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-39 Draft Revision 2 Table 15.1-10: Increase in Steam Flow (15.1.3) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 212985 psia Maximum SG pressure 2420 psia 15071253 psia MCHFR 1.435 1.557
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-40 Draft Revision 2 Table 15.1-11: Sequence of Events (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case)
Event Time [s]
SLB occurs 0
Reactor trip actuated (high power rate) 5634 SSI actuated (low main steam superheatpressurizer pressure) 14278 SSI valves closed except assumed failed MSFWIV 88152 DHRS actuated (high pressurizer pressure) 901509 RSV lifts 1742838 Peak RCS pressure reached 18381742
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-41 Draft Revision 2 Table 15.1-12: Sequence of Events (15.1.5 Steam Piping Failure - Limiting Steam Generator Pressure Case)
Event Time [s]
SLB occurs 0
Low main steamHigh pressurizer pressure limit is reached 0.146 Reactor trip, SSI, DHRS actuated 248 SSI actuated 2
SSI valves closed (except failed FWIV) 1258 DHRS actuation 32 Peak SG pressure reached 82111
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-43 Draft Revision 2 Table 15.1-14: Steam Piping Failure - Inputs Parameter Nominal Biases RCS Pressure Limiting Case SG Pressure Limiting Case MCHFR Limiting Case Core power (MWt) 250
+2%
+2%
+2%
Pressurizer pressure (psia) 2000
+70
+70
+-70 Pressurizer level (%)
60
+8
+8
+-8 RCS flow rate Range in Table 15.0-6 Low Low Low RCS average temperature (°F) 540
+5
+5
+5 SG pressure (psia) 475
+350
+35
+350 FW temperature (°F) 250
+-20
-20
+-20
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-44 Draft Revision 2 Table 15.1-15: Steam Piping Failure (15.1.5) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 2284 psia Maximum SG pressure 2420 psia 1440421 psia MCHFR 1.435 1.6355
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-46 Draft Revision 2 Table 15.1-17: Loss of Containment Vacuum/Containment Flooding (15.1.6) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 209687 psia Maximum SG pressure 2420 psia 1304407 psia MCHFR 1.435 2.5009
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-77 Draft Revision 2 Figure 15.1-29: Break Flow Rate (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case) 0 20 40 60 80 100 120 140 0
5000 10000 15000 20000 25000 30000 Break Flow (lbm/s)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-78 Draft Revision 2 Figure 15.1-30: Reactor Pressure Vessel Lower Plenum Pressure (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case) 1600 1700 1800 1900 2000 2100 2200 2300 0
5000 10000 15000 20000 25000 30000 RPV Lower Plenum Pressure (psia)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-79 Draft Revision 2 Figure 15.1-31: Reactor Power (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case) 0 20 40 60 80 100 120 0
5000 10000 15000 20000 25000 30000 Reactor Power (%)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-80 Draft Revision 2 Figure 15.1-32: Reactor Coolant System Average Temperatures (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case) 460 480 500 520 540 560 580 600 620 0
5000 10000 15000 20000 25000 30000 RCS Temperature (°F)
Time (sec)
RCS average temperature RCS cold temperature RPV lower downcomer temperature RCS hot temperature vol-avg moderator temperature
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-81 Draft Revision 2 Figure 15.1-33: Reactor Coolant System Flow (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case) 0 200 400 600 800 1000 1200 1400 1600 0
5000 10000 15000 20000 25000 30000 RCS Flow (lbm/s)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-82 Draft Revision 2 Figure 15.1-34: Decay Heat Removal System Flows Train 1 (15.1.5 Steam Piping Failure - Limiting Reactor Coolant System Pressure Case)
-10 0
10 20 30 40 50 60 0
5000 10000 15000 20000 25000 30000 DHRS Flow (lbm/s)
Time (sec)
DHRS 1 flow DHRS 2 flow
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-84 Draft Revision 2 Figure 15.1-36: Pressure Steam Generator 1 Secondary Pressures (15.1.5 Steam Piping Failure -
Limiting Steam Generator Pressure Case) 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500 0
200 400 600 800 1000 1200 1400 1600 1800 Pressure (psia)
Time (sec) steam pressure 1 steam pressure 2 FW plenum 1 pressure FW plenum 2 pressure
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-86 Draft Revision 2 Figure 15.1-38: Break Flow Rate (15.1.5 Steam Piping Failure - Limiting Minimum Critical Heat Flux Ratio Case) 0 50 100 150 200 250 0
500 1000 1500 2000 2500 3000 3500 4000 Break Flow (lbm/s)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-87 Draft Revision 2 Figure 15.1-39: Event Reactivity (15.1.5 Steam Piping Failure - Limiting Minimum Critical Heat Flux Ratio Case)
-35
-30
-25
-20
-15
-10
-5 0
5 10 15 0
500 1000 1500 2000 2500 3000 3500 4000 Reactivity ($)
Time (sec) total reactivity fuel reactivity feedback moderator reactivity feedback regulating bank reactivity scram reactivity
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-88 Draft Revision 2 Figure 15.1-40: Reactor Power (15.1.5 Steam Piping Failure - Limiting Minimum Critical Heat Flux Ratio Case) 0 20 40 60 80 100 120 140 0
500 1000 1500 2000 2500 3000 3500 4000 Reactor Power (%)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-89 Draft Revision 2 Figure 15.1-41: Volume Average Fuel Temperature (15.1.5 Steam Piping Failure - Limiting Minimum Critical Heat Flux Ratio Case) 400 500 600 700 800 900 1000 1100 1200 1300 0
500 1000 1500 2000 2500 3000 3500 4000 Vol-Avg Fuel Temperature (°F)
Time (sec)
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-90 Draft Revision 2 Figure 15.1-42: Secondary Pressures (15.1.5 Steam Piping Failure - Limiting Minimum Critical Heat Flux Ratio Case) 0 100 200 300 400 500 600 700 800 900 1000 0
500 1000 1500 2000 2500 3000 3500 4000 Pressure (psia)
Time (sec) steam pressure 1 steam pressure 2 FW plenum 1 pressure FW plenum 2 pressure
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-22 Draft Revision 2 The limiting FWLB for minimum critical heat flux ratio is a double ended guillotine break in an FW line outside of containment coincident with a loss of AC power.
The limiting MCHFR sequence of events is provided in Table 15.2-23.
The limiting FWLB for SG pressure occurs from a double ended guillotine4 percent break in an FW line inside containment coincident with a loss of AC power at the time of the break. The peak secondary pressurization is a function of the delay in DHRS actuation and establishing heat removal.
Table 15.2-22 provides the sequence of events for the limiting secondary pressure case.
The limiting DHRS cooling case involves a 4 percentdouble ended guillotine break inof an FW line inside containment. This break results in the complete loss of one train of DHRS. NoA loss of AC power is assumed at the time of the break. Due to the smaller size of the break and loss of AC power, the high PZR pressure MPS signal occurs before the high CNV pressure MPS signal. Reactor trip, SSI, and DHRS actuation occur on high PZR pressure and containment isolation occurs on high CNV pressure.Upon break initiation, pressure inside of the CNV rapidly increases, reaching the high CNV pressure analytical limit and actuating reactor trip, containment isolation, DHRS, and SSI. The remaining DHRS loop provides cooling to the module and is sufficient to remove 100 percent of decay heat and drive flow through the core. This event is not limiting for any of the acceptance criteria. The sequence of events for this case is provided in Table 15.2-24.
The MPS is credited to protect the NPM in the event of a FWLB. The following MPS signals provide the plant with protection during a FWLB:
high power rate low main steam pressure high PZR pressure high CNV pressure The actuation of a single RSV is credited for ensuring pressures in the RCS do not exceed the acceptance criteria.
The failure of the FWIV to close on the failed SG results in a limiting value for RCS pressure and MCHFR. A single failure of the FWIV is also assumed in the limiting DHRS case. The nonsafety-related FWRV is credited to close when single failure of the FWIV is assumed. Failure of the FWIV is not limiting for SG pressure or for the limiting DHRS case.
15.2.8.3 Thermal Hydraulic and Subchannel Analyses 15.2.8.3.1 Evaluation Model The thermal hydraulic analysis of the NPM response to an FWLB is performed using NRELAP5. The NRELAP5 model is based on the design features of a NuScale module. The non-LOCA NRELAP5 model is discussed in
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-37 Draft Revision 2 Table 15.2-5: Loss of External Load - Turbine Trip - Loss of Condenser Vacuum - Steam Generator Maximum Pressure Case Sequence of Events Event Time (sec)
Event initiator - TT 0
Turbine stop valves fully closed (assumption) 0 Reactor trip, SSI and DHRS actuation, PZR heater trip (high PZR pressure) 89 Assumed loss of normal AC power 9
Secondary system isolation complete 189 DHRS actuation valves full open 389 Time of peak secondary pressure 6393
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-39 Draft Revision 2 Table 15.2-7: Loss of External Load - Turbine Trip - Loss of Condenser Vacuum - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS Pressure 2420 psia 23010 psia Maximum SG Pressure 2420 psia 145126 psia MCHFR 1.435 2.409
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-45 Draft Revision 2 Table 15.2-13: Loss of Non-Emergency Alternating Current Power - Reactor Coolant System Pressure and Steam Generator Pressure Limiting Cases Sequence of Events Event Time [s]
Loss of AC power 0
Control rod insertion begins (reactor trip on high PZR pressure) 89 Secondary system isolation occurs 89 DHRS actuation valves begin to open 89 RSV lifts 10 Maximum RCS pressure 11 Maximum SG pressure (RCS pressure case) 51 CNV isolation 60 Maximum SG pressure (SG pressure case) 61
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-48 Draft Revision 2 Table 15.2-16: Loss of Non-Emergency Alternating Current Power - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 230514 psia Maximum SG pressure 2420 psia 14591391 psia MCHFR 1.435 1.932.40
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-54 Draft Revision 2 Table 15.2-22: Feedwater Line Break Sequence of Events - Peak Steam Generator Pressure Case Event Time [s]
Failure that initiates event 0
Loss of AC power 0
Maximum power 3
Control rod insertion begins (reactor trip on high pressurizercontainment pressure) 54 Secondary system isolation valves fully closed 5
DHRS actuation valves fully open 3534 Maximum SG pressure 6090 Maximum SG level 35393689
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-56 Draft Revision 2 Table 15.2-24: Feedwater Line Break Sequence of Events - Limiting Decay Heat Removal System Case Event Time [s]
Failure that initiates event 0
Loss of AC power 0
Maximum power 3
Control rod insertion begins (reactor trip on high pressurizercontainment pressure) 54 Secondary system isolation valves fully closed 54 Maximum RCS pressure 9
DHRS actuation valves fully open 354 Maximum SG pressure 6088 Maximum RCS pressure 380 Maximum SG level 35393439
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-58 Draft Revision 2 Table 15.2-26: Feedwater Line Break - Limiting Analysis Results Acceptance Criteria Limit Analysis Value Maximum RCS pressure 2420 psia 231620 psia Maximum SG pressure 2420 psia 14461424 psia MCHFR 1.435 2.4001
NuScale Final Safety Analysis Report Decrease in Heat Removal by the Secondary System NuScale US460 SDAA 15.2-60 Draft Revision 2 Table 15.2-28: Sequence of Events for Inadvertent Operation of Decay Heat Removal System - Limiting Peak Steam Generator Pressure Case Event Time (s)
Transient initiation (inadvertent isolation of oneboth SGs) 0 High pressurizer pressure analytical limit reached 6
RTS actuation - control rod insertion begins 8
Secondary system isolation valves closed 9
DHRS actuation valves fully open 38 Maximum SG pressure reached 9738
NuScale Final Safety Analysis Report Increase in Reactor Coolant Inventory NuScale US460 SDAA 15.5-8 Draft Revision 2 Table 15.5-3: Summary of Results - Chemical and Volume Control System Malfunction Acceptance Criteria Limit Analysis Value Maximum RCS Pressure 2420 psia 22886 psia Maximum SG Pressure 2420 psia 1285357 psia MCHFR 1.4345 2.3210
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-31 Draft Revision 2 Table 15.6-3: Failure of Lines Carrying Primary Coolant Outside Containment - Results Maximum Reactor Pressure Vessel Pressure Scenario Parameter Acceptance Criteria Value Maximum RCS Pressure 2640 psia 23032275 psia Maximum SG Pressure 2640 psia 14671392 psia
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-33 Draft Revision 2 Table 15.6-5: Steam Generator Tube Failure - Sequence of Events - Limiting Steam Generator Pressure Event Time (s)
SGTF (7100 percent tube area split break) at top of SG 0
Loss of AC power 0
Loss of power to pressurizer heaters 0
Reactor trip actuation (lowhigh pressurizer pressurelevel) 654 DHRS actuation 654 Secondary system isolation 654 MSIV fully closed (SG isolated) 15 CNV isolation 6054 CVCS isolation 6054 PZR heater isolation 654 MSIV fully closed (SG isolated) 664 Maximum SG pressure reached 18455999
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-35 Draft Revision 2 Table 15.6-7: Steam Generator Tube Failure - Results Parameter Acceptance Criteria Value Limiting Case Maximum RCS Pressure 2640 psia 22962304 psia Limiting Reactor Pressure Maximum SG Pressure 2640 psia 20382276 psia Limiting Steam Generator Pressure