ML24346A273
| ML24346A273 | |
| 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: ML24346A273 (1) | |
Text
Response to SDAA Audit Question Question Number: A-15.1.3-2 Receipt Date: 05/13/2024 Question:
The last row of Table 15.1-8 states that SSI and DHRS actuation occur on the high main steam pressure signal. However, FSAR Section 15.1.3.2 does not list high main steam pressure as one of the signals credited for SSI or DHRS actuation. Please provide FSAR markups clarifying the signals credited for SSI and DHRS actuation as well as the event sequence.
Response
The description of the module protection system (MPS) signals credited in the increase in steam flow event in Final Safety Analysis Report (FSAR) Section 15.1.3.2 is inaccurate. The current description in FSAR Revision 1 identifies the high main steam superheat signal as providing protection and being credited for secondary system isolation (SSI). Review of the output associated with the supporting calculation (EC-120241, Revision 0, NPM-20 Increase in Steam Flow/Inadvertent Opening of Steam Generator Relief or Safety Valve Analysis - previously provided in the electronic reading room (eRR)) identifies that the high main steam pressure signal provides protection and is credited for SSI rather than the high main steam superheat signal. The high main steam pressure signal also results in actuation of the decay heat removal system (DHRS). Therefore, FSAR Section 15.1.3.2 is revised as shown in the attached markup to add discussion of the high main steam pressure signal and delete discussion of the high main steam superheat signal. This revision ensures consistency with FSAR Table 15.1-8; no changes to Table 15.1-8 are required.
NuScale entered a similar inaccuracy with the wording in EC-120241 into its corrective action program.
NuScale Nonproprietary NuScale Nonproprietary Markups of the affected changes, as described in the response, are provided below:
NuScale Final Safety Analysis Report Increase in Heat Removal by the Secondary System NuScale US460 SDAA 15.1-12 Draft Revision 2 Because of the cooling of the RCS during an increase in steam flow event, the coolant in the downcomer increases in density. This increase in density can affect the power level detection by the excore neutron detectors. In order to account for this effect, the high power trip is adjusted to account for downcomer coolant temperature.
Audit Question A-15.1.3-2 In increased steam flow events that result in a reactor trip, the subsequent actuation of SSI and the DHRS are credited with maintaining reactor cooling. The MPS signals credited for SSI actuation are high PZR pressure, low main steam pressure, high main steam pressuresuperheat, and low main steam superheat.
The MPS signals credited for DHRS actuation areis high PZR pressure and high main steam pressure.
There are no single failures that could result in a more severe outcome of the limiting increase in steam flow event with respect to the acceptance criteria. Two possible failures that could affect the transient are the failure of one of the MSIVs or one of the FWIVs. However, these single failures could only occur after secondary side isolation or DHRS actuation, which occurs coincident with or after the reactor trip, and the MCHFR has already occurred. Therefore, neither single failure is modeled in the case presented in this section.
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 an increase in steam flow event because FW is lost, which reduces the overcooling event.
15.1.3.3 Thermal Hydraulic and Subchannel Analyses 15.1.3.3.1 Evaluation Model The thermal hydraulic analysis of the plant response to an increase in steam flow is performed using NRELAP5. A description of the NRELAP5 model is provided in Section 15.0.2. The NRELAP5 model is based on the design features of an NPM. The non-LOCA transient modifications to the NRELAP5 model are discussed in Section 15.0.2. The relevant boundary conditions from the NRELAP5 analyses are provided to the downstream subchannel CHF analysis.
The subchannel core CHF analysis is performed using VIPRE-01. VIPRE-01 is a subchannel analysis tool designed for general-purpose thermal-hydraulic analysis under normal operating conditions, operational transients, and events of moderate severity. Section 15.0.2 contains a discussion of the VIPRE-01 code and evaluation model.
15.1.3.3.2 Input Parameters and Initial Conditions As discussed in Section 15.1.3.2, the limiting increase in steam flow is an increase of 25 percent normal steam flow. Steam flow is assumed to linearly