ML24346A294
| ML24346A294 | |
| 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: ML24346A294 (1) | |
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Response to SDAA Audit Question Question Number: A-15.4.8-5 Receipt Date: 05/28/2024 Question:
Please provide proposed markup to the FSAR to include a discussion of the sensitivity studies that are required for using the methodology of TR-0716-50350-P-A, Rev. 3 for rod ejection accident analyses.
Response
Audit question A-15.4.8-1 requested making available for audit the calculations supporting Final Safety Analysis Report (FSAR) Section 15.4.8, including the documentation of the sensitivity studies described in Section 5.4.5 of TR-0716-50350-P, Rod Ejection Accident Methodology, Revision 3. In response to audit question A-15.4.8-1, NuScale provided the calculations and noted that the sensitivity studies were included in the calculations provided. Audit question A-15.4.8-1 was subsequently closed by the NRC.
Therefore, NuScale understands this question as a request to include discussion of the sensitivity studies into the FSAR. However, the detailed results of the sensitivity studies are beyond the level of detailed required for the FSAR. Although similar sensitivity studies were performed for the rod ejection event during the Design Certification Application (DCA) of the US600 design, the detailed results of the sensitivity studies were not required to be included in the FSAR for the DCA that was reviewed and approved by the NRC. Therefore, the detailed results of the sensitivity studies performed for the US460 are not added to the FSAR for the Standard Design Approval Application (SDAA). The specific types of sensitivity studies performed, their purpose, and their acceptance criteria are identified in TR-0716-50350-P, Revision 3, Section 5.4.5 and Table 5-3. Therefore, this level of information is also not added to the FSAR because the FSAR references TR-0716-50350-P, Revision 3 for the discussion of the methodology. Instead, NuScale revises FSAR Section 15.4.8 as shown in the attached markup NuScale Nonproprietary NuScale Nonproprietary
to clearly state that the sensitivity studies identified in TR-0716-50350-P, Revision 3 are performed to support the selection of inputs and identification of the limiting results that are then reported in the FSAR.
Note that the attached markup shows unrelated changes to the discussion of the bounding assessment for pressure acceptance criteria. Those changes were made previously for conformity with TR-0716-50350-P, Revision 3, but are not related to this audit question. The changes specific to this audit question are those directly below the Audit Question A-15.4.8-5 identifier in the left-hand margin.
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 15.4-27 Draft Revision 2 of moderate severity. Section 15.0.2 includes a discussion of the VIPRE-01 code and evaluation model.
Audit Question A-15.4.8-5 The REA event-specific methodology is provided in Reference 15.4-4.
15.4.8.3.2 Input Parameters and Initial Conditions SIMULATE-3K Model The power and reactivity data for an REA is calculated using S3K for the downstream thermal hydraulic and CHF analyses. The power response calculated by S3K is also used in the downstream analyses to determine the temperature and enthalpy responses of the fuel. The CHF analysis and the temperature and enthalpy responses of the fuel indicate if there is any fuel failure during an REA. The inputs, initial conditions, and conservatisms of the S3K rod ejection accident model are discussed in this section.
In order to maximize the possible reactivity insertion from an ejected rod, the non-shutdown CRAs are assumed to be at the PDIL with an uncertainty of 6 steps. The shutdown bank is positioned all rods out. Conservative reactor trip characteristics are applied to the REA model including:
highest worth CRA (other than the ejected rod) remains stuck out of the core reactor trip delay of 2 seconds maximum CRA drop time after reactor trip Conservative core characteristics are applied to the REA model to ensure the maximum reactivity insertion with minimum feedback. A top peaked power shape is applied to the REA model to maximize the effects of the ejected rod.
The uncertainty values of the DTC and moderator temperature coefficient (MTC) in the REA analysis are applied to the S3K values in the conservative (less negative) direction to minimize the fuel feedback effects that could mitigate the power response of an REA. A discussion of specific core parameter values and the associated biases used in the REA methodology is provided in the Rod Ejection Methodology Topical Report (Reference 15.4-4).
The S3K analysis provides REA power response calculations at the following power levels and times in cycle:
Power (%) - 0, 20, 50, 75, 100 Time in life - BOC, MOC, EOC NRELAP5 Model The NRELAP5 thermal hydraulic analysis uses the power response calculated by S3K to simulate the power pulse associated with an REA.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-28 Draft Revision 2 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 VIPRE-01 Model Key inputs and assumptions used in the subchannel analysis are provided in Reference 15.4-1 and Reference 15.4-2. The subchannel analysis includes performance of 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.