ML24346A324

LLC - Response to SDAA Audit Question Number A-19.1-64

ML24346A324
Person / Time
Site:	05200050
Issue date:	12/11/2024
From:	NuScale
To:	Office of Nuclear Reactor Regulation
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Download: ML24346A324 (1)

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Response to SDAA Audit Question Question Number: A-19.1-64 Receipt Date: 04/29/2024 Question:

The staff reviewed the response to A-19.1-55, item 3d, and SDAA FSAR Section 19.1.4.1.1.5 on Thermal-Hydraulic Uncertainty, specifically the three-step approach. The SDAA FSAR is written as if the three-step approach was implemented in its entirety for the SDAA. In reality, NuScale performed a limited scope evaluation for the SDAA to confirm the conclusions and reliability probabilities of the full scope evaluation (i.e., the entire three-step approach) performed during DCA and did not reproduce or recalculate the passive safety system response surface or reliability probabilities. Therefore, for the FSAR to reflect what was done for the SDAA and for the staff to make its finding based on this information, the staff is requesting that NuScale update the SDAA FSAR to describe the analysis performed specifically for the SDA.

The staff proposes the following FSAR markup:

For the SDA, NuScale performed a limited scope evaluation to confirm the conclusions and reliability probabilities of the full scope evaluation performed for the US600 certified design.

Instead of developing passive safety system response surfaces specific to the US460 design, bounding scenarios and thermal hydraulic parameters were identified from the full scope evaluation performed for the US600 certified design. These were evaluated with a US460-specific NRELAP5 model assuming successful reactor trip and containment isolation, and no credit for operator actions. The objective of this analysis was to incorporate design differences and determine effect of these differences, if any, on the conclusions of full scope evaluation for the US600 certified design. Based on this evaluation, NuScale concluded that a full scope re-evaluation was not necessary for the US460 design.

Response

NuScale revised Section 19.1 of the US460 Standard Design Approval Application to describe the approach used for passive safety system reliability.

Markups of the affected changes, as described in the response, are provided below:

NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Probabilistic Risk Assessment NuScale US460 SDAA 19.1-26 Draft Revision 2 resampled with the intended distributions into the previously calculated system response for comparison with the failure metric.

Audit Question A-19.1-64 NuScale performed a full scope passive safety system reliability evaluation to support the US600 certified design (Docket No.52-048). To support the US460 Standard Design Approval Application, NuScale performed a limited scope evaluation that consisted of identifying the most challenging scenarios and thermal-hydraulic parameter biases for the full scope evaluation, incorporating changes reflecting the US460 design, and simulating the responses with a US460-specific NRELAP5 model. The results confirm that the passive reliability rates for the ECCS and DHRS from the full scope evaluation are applicable to the US460 design.

Table 19.1-9 provides the calculated probabilities for failures of passive heat transfer.

Human Error Probabilities An HRA is performed to identify potential human failure events (HFEs) and to systematically estimate the probability of those events using bounding methods in the absence of as-operated facility information. The methods used in other nuclear power plant PRAs, as found by surveying the literature, and the methods applied in the NuScale PRA produce comparable HFE values. Both pre-initiator and post-initiator human actions are considered in the HRA. The HRA primarily applied the approach provided in NUREG/CR-4772 (1987) to estimate pre-initiator operator actions using the Accident Sequence Evaluation Program Human Reliability Analysis Procedure methodology and primarily NUREG/CR-6883 (2005) to estimate the post-initiator operator actions using the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) methodology.

Pre-initiator or latent errors, also referred to as Type A HFEs, can occur as a result of maintenance, testing, or calibration activities (before an initiating event) resulting in unavailability of the associated equipment when demanded. During maintenance, testing or calibration, equipment may be disabled or placed in an abnormal alignment that may render the function of that equipment unavailable. Human errors can occur when restoring or realigning the equipment into the normal configuration. A failure during these activities that results in equipment not being restored or aligned to normal is considered a pre-initiator human error. Consistent with the Accident Sequence Evaluation Program Human Reliability Analysis Procedure methodology, the following summarizes the process used to evaluate pre-initiator HFEs:

Identify activities and practices that may adversely impact the availability of mitigating systems if performed incorrectly.