ML24346A276
| ML24346A276 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 12/11/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24346A130 | List:
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| References | |
| LO-175762 | |
| Download: ML24346A276 (1) | |
Text
Response to SDAA Audit Question Question Number: A-15.1.6-2 Receipt Date: 04/22/2024 Question:
Specify the modeled volume of reactor component cooling water in the containment flooding event and how this volume is determined.
The limiting event in FSAR Section 15.1.6 is a reactor component cooling water fluid pipe break.
FSAR Section 15.1.6.2, Sequence of Events and Systems Operation, states that a conservative total reactor component cooling water system volume is assumed. FSAR Section 15.1.6.3.2, Input Parameters and Initial Conditions, states that the containment conditions are biased so as to empty the entire volume of reactor component cooling water into containment.
However, the FSAR does not specify the modeled volume of reactor component cooling water or how this volume is determined. Increased reactor component cooling water volume will increase the heat transfer from the reactor pressure vessel and the severity of the transient.
Provide FSAR markup.
Response
The modeled inventory of reactor component cooling water in the containment flooding event in final safety analysis report Section 15.1.6 is approximately 24,300 kg. The response to audit item A-15.0.5-3 describes how this assumed inventory is determined. The assumed reactor component cooling water mass is added to final safety analysis report Section 15.1.6.2 as shown in the attached markup.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-24 Draft Revision 2 15.1.6.2 Sequence of Events and Systems Operation The event sequence table for the limiting containment flooding case is provided in Table 15.1-18.
Unless specified below, the analysis of the containment flooding event assumes the plant control systems and ESFs perform as designed, with allowances for instrument inaccuracy. No operator action is credited to mitigate the effects of a containment flooding event.
Audit Question A-15.0.5-3, Audit Question A-15.1.6-2 The containment flooding is initiated by a break in a CRDS line that transports reactor component cooling water inside containment. A conservative total RCCWS volume is assumedmass of approximately 24,300 kg is assumed as a bounding volume of the entire RCCWS inventory. The containment flooding cases assume one or more RCCWS pumps continue operating after the break. The number of operating RCCWS pumps assumed in the analysis is varied to determine the most limiting case.
The containment evacuation pumps could malfunction to cause a loss of containment vacuum scenario, but are assumed to operate at nominal capacity for the limiting containment flooding scenario in order to delay reaching the high containment pressure MPS setpoint.
Operator action is not credited for regulating control rod movement or increasing boron concentration, which ensures the maximum reactivity insertion is reached as the control system attempts to maintain RCS temperature by pulling the regulating control rods from the core.
The MPS high containment pressure signal is credited to provide protection against loss of containment vacuum and containment flooding events. In cases that result in a reactor trip, the same high containment pressure signal actuates SSI and DHRS to maintain reactor cooling.
There are no single failures that could make a containment flooding event more severe with respect to the acceptance criteria.
Normal AC power is assumed to be available for this event. A loss of AC power, either at event initiation or at reactor trip, is not a conservative condition for a containment flooding event. The loss of AC power results in a reactor trip, DHRS actuation, and containment isolation on other MPS signals earlier than the high containment pressure signal.
15.1.6.3 Thermal Hydraulic and Subchannel Analyses 15.1.6.3.1 Evaluation Model The thermal hydraulic analysis of the plant response to containment flooding is performed using NRELAP5. A description of the NRELAP5 model is