ML24346A295
| ML24346A295 | |
| 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: ML24346A295 (1) | |
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Response to SDAA Audit Question Question Number: A-15.4.8-6 Receipt Date: 06/17/2024 Question:
In Table 5-3 of TR-0716-50350-P provides a list of sensitivity studies that are needed if default values are not used for rod ejection accident (REA) analyses. However, neither FSAR 15.4.8, nor FSAR 15.0.2.3 discusses whether the default or different values are used for the subchannel analyses. The staff notes that the sensitivity analyses are provided in EC-102431, Rev. 3, however the FSAR should reflect what was done for the analysis presented. Please clarify if default values are used for the subchannel analyses for those parameters in Table 5-3.
Please provide proposed FSAR markups to include a summary of information on what values are used in the subchannel analyses performed to produce the results in FSAR Section 15.4.8.
Response
Table 5-3 of TR-0716-50350-P, Revision 3, Rod Ejection Accident Methodology, identifies that sensitivity studies for axial nodalization, radial nodalization, convergence parameters, and convergence option deviations are only necessary if the subchannel default parameters are not used. Table 5-3 identifies that sensitivity studies for fuel heat transfer inputs, time-step size, and two-phase flow correlations are mandatory. ((2(a),(c) Therefore, the following clarification is provided: the default inputs from the NRC-approved subchannel methodology are used for axial nodalization, radial nodalization, convergence parameters, and convergence option deviations. Final Safety Analysis Report (FSAR) Section 15.4.8 is updated as shown in the attached markups to state that the default NuScale Nonproprietary NuScale Nonproprietary
NRC-approved subchannel methodology inputs are used for these items for the rod ejection analysis. 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 15.4-28 Draft Revision 2 The NRELAP5 thermal hydraulic analysis uses the power response calculated by S3K to simulate the power pulse associated with an REA. The NRELAP5 analysis provides boundary conditions to the downstream subchannel analysis to evaluate CHF, fuel temperature, and fuel enthalpy. The NRELAP5 cases provided for the downstream subchannel analysis are initialized with the following inputs: Initial power - The highest S3K power increase at each initial power level is selected. BOC and EOC conditions - The EOC conditions result in the highest S3K power increase at each initial power level, as shown in Table 15.4-20. However, BOC conditions at hot full power (HFP) are also included as they may be limiting for the downstream subchannel analysis. Average RCS temperature biased low and high - Although higher temperatures increase the vaporization and pressurization rate, lower temperatures are also included as they may be limiting for the downstream subchannel analysis. The RCS pressure and pressurizer level biased high - This combination maximizes the RCS pressurization rate. Allowances for instrument inaccuracy are accounted for in the analytical limits of mitigating systems in accordance with RG 1.105. Audit Question A-15.4.8-5 Key inputs and assumptions used in the subchannel analysis are provided in Reference 15.4-1 and Reference 15.4-2. The results of the subchannel analysis determine if there is any potential fuel damage resulting from an REA. The REA event-specific methodology is provided in Reference 15.4-4. NRELAP5 is also used to evaluate the maximum pressure response to an REA. A bounding assessment demonstrates the pressure acceptance criteria are satisfied for the NuScale Power Plant US460 standard design based on the MPS analytical limits of power, power rate, and pressurizer pressure identified in Table 15.0-7.The methodology in Reference 15.4-4 identifies that a generic assessment demonstrates the pressure acceptance criteria are generally satisfied for the MPS analytical limits of power, power rate, and pressure. The generic assessment using NRELAP5 is applicable to an REA for the NuScale Power Plant US460 standard design based on the MPS analytical limits in Section 7.1 and the REA peak power calculated in S3K. Audit Question A-15.4.8-5, Audit Question A-15.4.8-6 VIPRE-01 Model Key inputs and assumptions used in the subchannel analysis are provided in Reference 15.4-1 and Reference 15.4-2. Defaults from the Reference 15.4-1 and Reference 15.4-2 methodologies are used for axial nodalization, radial nodalization, convergence parameters, and convergence options in the subchannel analysis. The subchannel analysis includes performance of
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-29 Draft Revision 2 sensitivity studies to select inputs that ensure a reliable and converged solution and to ensure limiting values for acceptance criteria are determined in accordance with the methodology in Reference 15.4-4. The results of the subchannel analysis determine if there is any potential fuel failure resulting from an REA. 15.4.8.3.3 Regulatory Criteria Reference 15.4-4 discusses the various REA regulatory acceptance criteria and how they apply. A summary of these acceptance criteria are provided in this section. Fuel Cladding Failure The high temperature cladding failure threshold is expressed in cladding differential pressure. The peak radial average fuel enthalpy must be below 100 cal/g. For the NuScale Power Plant US460 standard design, the 100 cal/g limit is applied at all peak rod differential pressures. Fuel cladding failure is presumed if local heat flux exceeds the CHF analysis limit. The pellet-cladding mechanical interaction failure limit is a change in radial average fuel enthalpy of 33 cal/g or greater, bounding the excess hydrogen-dependent limit depicted in Figure 5-3 of Reference 15.4-4. If fuel temperature anywhere in the pellet exceeds incipient fuel melting conditions, then fuel cladding failure is presumed. Core Coolability Peak radial average fuel enthalpy shall remain below 230 cal/g. No fuel melt shall occur. Local heat flux shall not exceed the CHF analysis limit. Reactor Coolant System Pressure The maximum RCS pressure must remain below 120 percent of design pressure. Therefore, the peak pressure during an REA is limited below 2640 psia. 15.4.8.3.4 Fuel and Cladding Integrity Results SIMULATE-3K provides the power and reactivity response to an REA for each statepoint discussed in Section 15.4.8.3.2. The S3K analysis assumes the maximum reactivity insertion from ejecting the highest worth CRA for each of these statepoints. Table 15.4-20 provides the maximum power in percent HFP as well as the inserted reactivity of the ejected rod for a spectrum of initial power levels and times in cycle. Figure 15.4-18 provides the maximum power pulse, which occurs for EOC hot zero power conditions. The figure shows a rapid reactivity excursion when the rod is ejected, but the power pulse is}}