ML24346A286

LLC - Response to SDAA Audit Question Number A-15.4.3-5

ML24346A286
Person / Time
Site:	05200050
Issue date:	12/11/2024
From:	NuScale
To:	Office of Nuclear Reactor Regulation
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Download: ML24346A286 (1)

Text

Response to SDAA Audit Question Question Number: A-15.4.3-5 Receipt Date: 06/17/2024 Question:

FSAR Figures 15.4-11 through 15.4-15 show NPM-20 system behavior supporting the limiting single CRA withdrawal evaluation in FSAR Section 15.4.3. However, the figures do not show whether a safe and stable condition is reached because they terminate roughly 90 seconds after event initiation. EC-0000-8579, which supports the single CRA withdrawal evaluation, appears to contain results supporting the conclusion that a safe and stable condition is reached for this event. Provide FSAR markup demonstrating that a safe and stable condition is reached for the limiting case of this event.

Response

Final Safety Analysis Report (FSAR) Figures 15.4-12 through 15.4-15 are replaced with figures that show an extended time sufficient to demonstrate that a safe and stable condition is reached for this event. FSAR Figure 15.4-11 is not replaced because it is not necessary to show the various reactivity terms for an extended time; the reactor power trend in the revised Figure 15.4-12 demonstrates reactivity is stable.

Markups of the affected changes, as described in the response, are provided below:

NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-65 Draft Revision 2 Audit Question A-15.4.3-5 Figure 15.4-12: Reactor Power (15.4.3 Control Rod Misoperation, Single Control Rod Assembly Withdrawal) 0 10 20 30 40 50 60 70 80 90 0

200 400 600 800 1000 1200 1400 1600 1800 Reactor Power (%)

Time (sec)

NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-66 Draft Revision 2 Audit Question A-15.4.3-5 Figure 15.4-13: Reactor Coolant System Pressure for Limiting Minimum Critical Heat Flux Ratio Case (15.4.3 Control Rod Misoperation, Single Control Rod Assembly Withdrawal) 1650 1700 1750 1800 1850 1900 1950 2000 2050 2100 2150 0

200 400 600 800 1000 1200 1400 1600 1800 RPV Lower Plenum Pressure (psia)

Time (sec)

NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-67 Draft Revision 2 Audit Question A-15.4.3-5 Figure 15.4-14: Reactor Coolant System Average Temperatures (15.4.3 Control Rod Misoperation, Single Control Rod Assembly Withdrawal) 515 520 525 530 535 540 545 550 555 560 565 0

200 400 600 800 1000 1200 1400 1600 1800 RCS Average Temperature (°F)

Time (sec)

NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-68 Draft Revision 2 Audit Question A-15.4.3-5 Figure 15.4-15: Reactor Coolant System Flow (15.4.3 Control Rod Misoperation, Single Control Rod Assembly Withdrawal) 0 200 400 600 800 1000 1200 1400 0

200 400 600 800 1000 1200 1400 1600 1800 RCS Flow (lbm/s)

Time (sec)