ML24346A280
| ML24346A280 | |
| 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: ML24346A280 (1) | |
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Response to SDAA Audit Question Question Number: A-15.4.1-1S Receipt Date: 05/16/2024 Question:
Audit question A-15.4.1-1 was received by NuScale on April 22, 2024. NuScale provided a response on April 25, 2024. On May 16, 2024, NuScale received the following written feedback from the NRC:
The response to question A-15.4.1-1 states that (( 2(a),(c) To support its findings in this section, the FSAR must be supported by calculational results consistent with the approved methods. As such, any revised calculations must be completed and summarized in the FSAR. Further, NRC staff needs to audit the revised calculations and/or supporting justifications with respect to the analytical limits used in order to reach a finding on the applicability of the limit used in evaluating fuel centerline melt in UCBWS transients. Therefore, provide the FSAR markups (( }} 2(a),(c) for NRC staff audit.
Response
This response is a supplement to the April 25, 2024 response to address the feedback from May 16, 2024; the April 25, 2024 response remains unchanged in the electronic reading room (eRR). The subchannel analysis (EC-121739, Subchannel Analysis of Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power State) associated with the uncontrolled control rod assembly withdrawal from a subcritical or low power state analysis is revised to use NuScale Nonproprietary NuScale Nonproprietary
maximum fuel centerline temperature as an acceptance criterion rather than linear heat generation rate. Final Safety Analysis Report (FSAR) Section 15.4.1 is revised as indicated in the attached markups to incorporate the revised subchannel analysis. The revised subchannel analysis (EC-121739, Revision 3) is also provided in the eRR. 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 15.4-1 Draft Revision 2 15.4 Reactivity and Power Distribution Anomalies 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.4.1.1 Identification of Causes and Accident Description An uncontrolled control rod assembly (CRA) withdrawal from a subcritical or low power startup condition event could result in a rapid insertion of reactivity into the reactor core. There is an increase in reactor power due to the unexpected addition of reactivity as the CRA bank is withdrawn from the core. The core power increases at a faster rate than heat can be removed, resulting in an increase in reactor coolant system (RCS) temperature and a decrease in minimum critical heat flux ratio (MCHFR). An uncontrolled CRA withdrawal from a subcritical or low power startup condition is classified as an anticipated operational occurrence (AOO) as indicated in Table 15.0-1. 15.4.1.2 Sequence of Events and Systems Operation Audit Question A-15.4.1-1 The sequence of events for an uncontrolled CRA withdrawal from a subcritical or low power startup condition is provided in Table 15.4-1 for the limiting MCHFR and fuel centerline temperaturelinear heat generation rate (LHGR) case. The RCS pressure and secondary pressure are not acceptance criteria for this event. The RCS and secondary pressure are bounded by other AOO events. Unless specified below, the analysis of an uncontrolled CRA withdrawal from a subcritical or low power startup condition event assumes the control systems and engineered safety features perform as designed, with allowances for instrument inaccuracy. No operator action is credited to mitigate the effects of the CRA withdrawal. The rod control function of the control rod drive system (CRDS) provides reactivity control to compensate for rapid, short-term variations in the reactivity of the core. The rod control function is also used to maintain the measured RCS temperature at or near the programmed average coolant temperature. The CRDS rod control operational modes include manual mode, automatic mode, and insertion-only automatic mode. The CRDS rod control function could be in manual mode during startup conditions. An operator error or malfunction in the CRDS would have to occur to initiate a uncontrolled CRA withdrawal from a subcritical or low power startup condition. The expected normal travel rate of the CRAs is 6 in/min. However, the maximum allowed withdrawal rate of a CRA is 15 in/min, with a step size no greater than three-eighths inch. A spectrum of constant reactivity insertion rates that bounds this maximum rate and also bounds possible boron dilution scenarios is included in the analysis.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-2 Draft Revision 2 The module protection system (MPS) is credited to protect the NuScale Power Module (NPM) in the event of an uncontrolled CRA withdrawal from a subcritical or low power startup condition. The following MPS signals protect the NPM in the event of an uncontrolled CRA withdrawal from a subcritical or low power startup condition: high power (at 25 percent of full power for startup conditions) source range (SR) and intermediate range (IR) log power rate high SR count rate The RCS pressure control with heaters and spray is assumed to function normally to delay the trip on high pressurizer pressure. Pressurizer level control is disabled. No single failure could occur during an uncontrolled CRA withdrawal from a subcritical or low power startup condition event that would result in more severe conditions for the limiting case. 15.4.1.3 Thermal Hydraulic and Subchannel Analyses 15.4.1.3.1 Evaluation Models The thermal hydraulic analysis of the NPM response to an uncontrolled CRA withdrawal from a subcritical or low power startup condition is performed using NRELAP5. The NRELAP5 model is based on the design features of the NPM. The non-loss-of-coolant accident (non-LOCA) NRELAP5 model is discussed in Section 15.0.2. The relevant boundary conditions from the NRELAP5 analyses are provided to the downstream subchannel critical heat flux (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. Limiting axial and radial power shapes are used in the subchannel analysis to ensure a conservative MCHFR result, in accordance with the methodology described in the Subchannel Analysis Methodology Topical Report (Reference 15.4-1) and the Statistical Subchannel Analysis Methodology Topical Report (Reference 15.4-2). Section 15.0.2 includes a discussion of the VIPRE-01 code and evaluation model. 15.4.1.3.2 Input Parameters and Initial Conditions Audit Question A-15.4.1-1 A spectrum of initial conditions is analyzed to find the limiting reactivity insertion due to an uncontrolled CRA withdrawal from a subcritical or low power startup condition. The initial conditions of the transient evaluation result in a conservative calculation. Table 15.4-2 provides key inputs and associated biases for the limiting MCHFR and fuel centerline temperatureLHGR case. The following initial conditions and assumptions ensure the results have sufficient conservatism.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-3 Draft Revision 2 The minimum initial power assumed for this analysis is 1 watt. The transient analyses for this event evaluate cases with initial power ranging from this minimum power of 1 watt to 15 percent of full power (consistent with use of the low setting for the high power analytical limit). The Standard Review Plan (SRP) guidance states that minimizing initial power provides the most conservative conditions because it provides the maximum power peak. However, this is not the case for the NuScale Power Plant US460 standard design. The SR log power rate trip signal prevents a significant power increase at lower powers. The least negative/most positive reactivity feedback coefficients are used to minimize the reactivity feedback. No maximum bank worth is credited to limit the total reactivity insertion. The positive reactivity inserted by the CRA withdrawal is modeled as a constant reactivity addition beginning at the transient initiation. To bound the reactivity insertion at the maximum rod speed and the reactivity insertion from possible boron dilution scenarios, a range of reactivity insertion rates from 0.0001 $/sec to 0.0500 $/sec is analyzed. Conservative reactor trip characteristics are used, including a maximum time delay, holding the most reactive rod out of the core, and using a bounding control rod drop rate. Allowances for instrument inaccuracy are accounted for in the analytical limits of mitigating systems in accordance with Regulatory Guide (RG) 1.105. Audit Question A-15.4.1-1 The results from the thermal hydraulic evaluation are used as input to the subchannel analysis to determine the MCHFR and maximum fuel centerline temperatureLHGR for this event. The subchannel model is discussed in Section 15.0.2. 15.4.1.3.3 Results Audit Question A-15.4.1-1 The sequence of events for a representative uncontrolled CRA withdrawal from a low power or startup condition is provided in Table 15.4-1 for the limiting MCHFR and fuel centerline temperatureLHGR case. Figure 15.4-1 through Figure 15.4-4 show the transient behavior of key parameters for the case that is limiting with respect to MCHFR and fuel centerline temperatureLHGR. Audit Question A-15.4.1-1 The CRA bank begins to withdraw at the transient initiation, which begins to raise power, RCS temperature, and RCS pressure. The case that is limiting for MCHFR and fuel centerline temperatureLHGR has an initial power and reactivity insertion rate that initially avoids the high power rate trip and allows a heatup and pressurization before reactor trip. The reactor power rises until the
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-4 Draft Revision 2 high power (25 percent) limit is reached and MPS initiates a reactor trip after the delay. The peak power, maximum RCS pressure, and MCHFR occur near the time of reactor trip. Audit Question A-15.4.1-1 The limiting cases for an uncontrolled CRA withdrawal from a low power or startup condition demonstrate margin to the acceptance criteria. The limiting MCHFR for this event is above the CHF analysis limit and the limiting fuel centerline temperatureLHGR is below the design limit. The limiting values for these acceptance criteria are shown in Table 15.4-3. 15.4.1.4 Radiological Consequences The leakage-related radiological consequences of this event are bounded by the design-basis accident analyses presented in Section 15.0.3. 15.4.1.5 Conclusions The acceptance criteria for this AOO are met for the limiting cases. The thermal margin limits are met.
The MCHFR for the limiting case is above the CHF analysis limit, as shown in Table 15.4-3. Fuel centerline temperatures do not exceed the melting point. Audit Question A-15.4.1-1
As discussed in Reference 15.4-1, a steady-state LHGR protection limit can be applied to an uncontrolled CRA withdrawal from a subcritical or low power startup condition event to ensure the fuel centerline temperatures do not exceed the melting point. The maximum fuel centerline temperatureLHGR for the limiting case is below the limit, as shown in Table 15.4-3. 15.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power 15.4.2.1 Identification of Causes and Accident Description A spurious CRA withdrawal that occurs when the reactor is at power leads to an addition of positive reactivity into the reactor, an increase in core power, and a corresponding increase in heat flux. Because of the time lag in the response of the secondary system, the heat removal from the steam generators follows the heat increase in the primary system. The result is an increase in RCS temperature and pressure. These conditions could challenge design pressures and the specified acceptable fuel design limits (SAFDLs). The power range neutron excore detectors provide high power and high power rate core protection. For cases where the reactivity insertion is sufficiently slow, the high pressurizer pressure, high RCS hot temperature, and high RCS average temperature limits provide protection. These MPS limits are analyzed for a spectrum of uncontrolled CRA withdrawal conditions to ensure protection functions are actuated to prevent the violation of the design safety limits.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-33 Draft Revision 2 Audit Question A-15.4.1-1 Table 15.4-1: Sequence of Events for Limiting Minimum Critical Heat Flux Ratio and Fuel Centerline TemperatureLinear Heat Generation Rate Case (15.4.1 Uncontrolled Control Rod Assembly Withdrawal from Subcritical or Low Power Startup Condition) Event Time [s] Rod withdrawal initiates 0 High power (25%) limit reached 201 Reactor trip actuated 203 Maximum power occurs 205 MCHFR occurs 205 Maximum RCS pressure occurs 205
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-34 Draft Revision 2 Audit Question A-15.4.1-1 Table 15.4-2: Key Inputs for Limiting Minimum Critical Heat Flux Ratio and Linear Fuel Centerline TemperatureHeat Generation Rate Case (15.4.1 Uncontrolled Control Rod Assembly Withdrawal from Subcritical or Low Power Startup Condition) Parameter Nominal Bias Initial power 500 kW N/A1 Initial RCS flow rate 80 kg/sec Low2 Pressurizer pressure 2000 psia Nominal Pressurizer level 50% Nominal RCS average temperature 351.9 °F Nominal MTC N/A3 BOC (most positive/least negative) FTC N/A3 BOC (least negative) Reactivity insertion rate 0.00094 $/sec N/A 1A spectrum of initial powers is analyzed, and this value provided the limiting results. 2The initial RCS flow rate varies as a function of the initial power. RCS flow is biased low for the initial power level. 3Nominal values are not specified; only most negative and most positive/least negative values are specified.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-35 Draft Revision 2 Audit Question A-15.4.1-1 Table 15.4-3: Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition (15.4.1) - Limiting Analysis Results Acceptance Criteria Limit Analysis Value MCHFR 1.435 10.0 Maximum fuel centerline temperatureLHGR 4791 °F14.2 kW/ft 1007 °F6.0 kW/ft}}