ML24215A224
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Response to SDAA Audit Question Question Number: A-19.1-56 Receipt Date: 02/19/2024 Question:
In the SRM to SECY-93-087, the Commission approved the staffs position to require that analysis of multiple steam generator tube ruptures involving two to five steam generator tubes be included in the application for design certification for the passive PWRs. Further, the Commission stated that realistic or best-estimate analytical assumptions may be used to assess plant responses since the steam generator multi-tube rupture event is beyond the design basis requirements for PWRs.
The staff has reviewed NuScales steam generator tube failure (SGTF) probability as provided in FSAR Table 19.1-7, the SGTF Frequency Report (ER-102086), and the SGTF Probabilistic Risk Assessment Report (ER-P010-3782) to determine whether additional information is needed to support NuScales conclusion that multiple tube failures are not judged to be credible because of design characteristics. Based on the staffs review, it does not appear that scenarios involving multiple steam generator tube failures have been evaluated in the PRA for the SDA other than to treat the random failure of multiple steam generator tubes as not credible.
- 1. Describe how the occurrence of multiple SGTFs was evaluated in accordance with the Commission policy in SRM-SECY-93-0087 and its treatment in the PRA or justify the exclusion of such occurrences from the PRA.
- 2. Describe the plant response to the failure of two to five steam generator tubes. As part of the response, identify similarities and differences in the plants response compared to the failure of a single steam generator tube and include a justification for multiple steam generator tube failures not posing a challenge to the plants passive safety systems.
- 3. Identify any sensitivity studies that were performed to disposition the above concerns and discuss their adequacy for addressing these concerns.
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Response
- 1. As described in FSAR Section 19.1.4.1.1.4, in contrast to currently operating plants, the NuScale steam generator (SG) tubes are in compression (versus tension); secondary coolant is on the inside of the tubes and primary coolant is on the outside. With the tubes in a constant state of compression, NuScale ((2(a),(c), ECI and the simultaneous failure of multiple tubes is judged to be not credible in the NuScale helical coil SG design. Also, as described in SECY-93-087, the NRC has not received any report of multiple steam generator tube failures (SGTFs) in any U.S. or foreign plant.
The staffs concern in Part R of SECY-93-087 under Multiple Steam Generator Tube Ruptures for Passive PWRs was that an AP600 plant could respond in substantially different ways to the two accidents [single and multiple steam generator tube ruptures], and that an multiple SGTR event could pose substantial challenges to the plants passive safety systems. Because of the possible difference in the plant response to multiple steam generator tube ruptures (SGTRs), the staff was considering a multiple SGTR for the design basis of the AP600, as opposed to a single SGTR, which is the case in conventional plants. The SECY-93-087 also implies that considering only a single SGTR in the design basis for conventional plants is acceptable because plant conditions would probably be similar during the transient to those in a single SGTR. The staff summarizes their concern by stating, The staff recognizes, however, that the multiple SGTR frequency could be in the range of 10-3 to 10-4 per RY and that passive plant response for a multiple SGTR could significantly differ from that of a single tube rupture. This recognition has led the staff to conclude that rupture of more than a single tube should be considered within the design basis of the plant. NuScales estimated mean frequency of an SGTF is 4.6E-05 per module critical year for a single tube failure, and as discussed below, ((
}}2(a),(c), ECI Also discussed below, a multiple SGTF event does not challenge the plants safety systems.
- 2. The expected plant response to an SGTF includes a reactor trip on low pressurizer level, followed by secondary system isolation, which closes the steam and feedwater isolation valves on both SGs and actuates the decay heat removal system (DHRS). With secondary system isolation, the affected SG is isolated. Because the reactor pressure vessel and containment NuScale Nonproprietary NuScale Nonproprietary
vessel are partially immersed in the reactor pool, the single train of DHRS on the intact SG will provide passive heat transfer from the core to the ultimate heat sink. ((
}}2(a),(c), ECI For SGTF sequences involving failure of secondary system isolation, the loss of coolant inventory from the primary system to the secondary system is terminated when primary system pressure reaches secondary system pressure, which occurs after emergency core cooling system actuation. (( }}2(a),(c), ECI One of two trains of DHRS provides passive heat removal in the case of an isolated SGTF and one of two trains of the emergency core cooling system provides sufficient passive heat removal in the case of an unisolated SGTF.
- 3. In a bounding PRA sensitivity, both trains of DHRS are assumed failed. This sensitivity supports the identification of candidate risk-significant structures, systems, and components.
The results do not identify DHRS as a candidate for risk significance; the conditional core damage frequency and conditional large release frequency remain below the quantitative criteria for candidate risk-significant systems. ((
}}2(a),(c), ECI No changes to the SDAA are necessary.
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