ML24346A247
| ML24346A247 | |
| 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: ML24346A247 (1) | |
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Response to SDAA Audit Question Question Number: A-15.0.3.7-2 Receipt Date: 05/22/2023 Question:
This item is a follow up to Audit item A-15.0.3.7-1. Please make the engineering calculation documents for the response to Audit item A-15.0.3.7-1 and also the engineering calculation documents for SDAA section 15.0.3.7.1 and SDAA section 15.0.3.7.2 available in the eRR.
Submit the response to Audit item A-15.0.3.7-1 on the docket. Provide an SDAA markup in eRR that includes the basis for the isolation time adequacy.
Response
The response to audit question A-15.0.3.7-1 discusses the dose analysis calculations for standard design approval application (SDAA) final safety analysis report (FSAR) Sections 15.0.3.7.1 and 15.0.3.7.2 and compares their assumptions to the event-specific transient results from the calculations associated with FSAR Sections 15.6.2 and 15.6.3, respectively. The calculations are provided in the electronic Reading Room (eRR) as indicated in the table below.
Topic Document Number in eRR Dose analysis calculation for Section 15.0.3.7.1 EC-112223, Revision 1 Dose analysis calculation for Section 15.0.3.7.2 EC-109242, Revision 2 Transient analysis calculation for Section 15.6.2 EC-122599, Revision 0 Transient analysis calculation for Section 15.6.3 EC-123941, Revision 1 The response to audit question A-15.0.3.7-1 will be submitted on the docket with the other audit questions that are identified as needing to be submitted. Markups to the SDAA FSAR are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-27 Draft Revision 1 15.0.3.7 Consequence Analyses of Design-Basis Source Terms 15.0.3.7.1 Failure of Small Lines Carrying Primary Coolant Outside Containment Failure of small lines carrying primary coolant outside containment is described in Section 3.2.5 of Reference 15.0-6. Section 15.6.2 describes the sequence of events and thermal-hydraulic response to a failure of small lines carrying primary coolant outside containment. The analysis in Section 15.6.2 shows the reactor core remains covered and no fuel failures occur.
Audit Question A-15.0.3.7-2 A bounding release of primary coolant from the RPV of 23,000 lbm is assumedis defined for the dose consequence analysis rather than using results from the transient analysis described in Section 15.6.2. The assumed primary coolant released from the RPV is 23,000 lbm. This value is calculated to bound the maximum possible decrease in pressurizer level before the containment isolation setpoint is reached, including uncertainty, plus allowance for continued mass release during signal processing and valve closure times. The assumed value of 23,000 lbm is confirmed to be bounding of the event-specific release from the transient analyses in Section 15.6.2.
The total mass released from the event is the sum of the mass released from the RPV and the primary coolant from CVCS equipment and piping. Primary coolant in the CVCS equipment (e.g., heat exchangers and filters) and piping within the RXB contribute less than 15,000 lbm additional primary coolant to the potential release.
Audit Question A-15.0.3.7-2 Containment isolation is assumed to occur at 30 minutes. The assumed value of 30 minutes is confirmed to be bounding of the event-specific isolation and iodine spiking time from the transient analyses in Section 15.6.2. The primary coolant masses described above are released over this duration. The release of activity is conservatively modeled as a direct release to the environment.
After containment isolation, primary coolant leaks through one containment isolation valve (the redundant in-series valve is assumed to fail open) at the maximum leak rate allowed by design-basis limits. The activity from this leak path is assumed to flow directly to the environment with no mitigation or reduction by intervening structures. After 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, the reactor is assumed to be shut down and depressurized, and releases through the containment isolation valve stop.
Iodine spike assumptions for this event are listed in Reference 15.0-6.
The primary coolant in the reactor vessel and CVCS equipment and piping in the RXB is assumed to initially contain the allowable concentration of dose equivalent (DE) I-131 of 5.8E-02 Ci/gm and DE Xe-133 of 16 Ci/gm.
There are no single failures for this event that could result in more severe radiological consequences.
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-28 Draft Revision 1 Doses are determined at the EAB, the LPZ, and for personnel in the MCR and the TSC. The MCR model is described in Section 15.0.3.6.1. The potential radiological consequences of the failure of small lines carrying primary coolant outside containment event are presented in Table 15.0-10.
15.0.3.7.2 Steam Generator Tube Failure Radiological consequences of the SGTF are calculated based on the guidance provided in Appendix F of RG 1.183 as described in Section 3.2.4 of Reference 15.0-6. Section 15.6.3 describes the sequence of events and thermal-hydraulic response to an SGTF. The analysis in Section 15.6.3 shows the reactor core remains covered and no fuel failures occur.
Audit Question A-15.0.3.7-2 For the faulted SG, a bounding release of primary coolant of 23,000 lbm is assumedis defined for the dose consequence analysis rather than using results from the transient analysis described in Section 15.6.3. The assumed primary coolant released from the reactor is 23,000 lbm. This value is calculated to bound the maximum possible decrease in pressurizer level before the secondary system isolation setpoint is reached, including uncertainty, plus allowance for continued mass release during signal processing and valve closure times. The assumed value of 23,000 lbm is confirmed to be bounding of the event-specific release from the transient analyses in Section 15.6.3. Secondary system isolation, including closure of MSIVs and MSIBVs, is assumed to occur at 30 minutes. The assumed value of 30 minutes is confirmed to be bounding of the event-specific isolation and iodine spiking time from the transient analyses in Section 15.6.3. The primary coolant release associated with the faulted SG is conservatively modeled as a direct release to the environment over this duration. The secondary system is not modeled and no credit is taken for holdup or dilution in the secondary system.
For the intact SG, the primary coolant associated with the maximum primary-to-secondary leak rate allowed by design-basis limits is assumed to be released directly to the environment. The secondary system is not modeled and no credit is taken for holdup or dilution in the secondary system. The direct release of primary coolant to the environment associated with the intact SG is assumed to be terminated after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> when the RCS is depressurized and primary and secondary system pressures equalize.
Iodine spike assumptions for this event are listed in Reference 15.0-6. The primary coolant contains an assumed concentration of 5.8E-02 Ci/gm DE I-131 for the coincident iodine spike scenario and 3.5 Ci/gm DE I-131 for the pre-incident iodine spike scenario. For both iodine spiking scenarios, the primary coolant is assumed to contain 16 Ci/gm DE Xe-133.
A single failure of an MSIV or MSIBV to close at the time of secondary system isolation is mitigated by closure of the secondary MSIV or secondary MSIBV and does not result in more severe radiological consequences.
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-4 Draft Revision 1 response of Figure 15.6-1 and Figure 15.6-2, and Figure 15.6-5 through Figure 15.6-8, a low pressurizer level signal initiates the CVCS containment isolation valves to close, isolating the break flow from the RCS. The system continues to cool using DHRS (Figure 15.6-4), RCS flow stabilizes (Figure 15.6-9), RCS temperature (Figure 15.6-5) and fuel temperature (Figure 15.6-8) stabilize and continue to decline, the core remains subcritical (Figure 15.6-10), and the water level is above the top of the active fuel (Figure 15.6-7). The system response shows that the event terminates and the NPM reaches a safe, stabilized condition.
15.6.2.4 Radiological Consequences Audit Question A-15.0.3.7-2 Section 15.0.3 provides the radiological consequences of a failure in small lines carrying primary coolant outside containment. Assumptions for primary coolant release (23,000 lbm) and containment isolation time (30 minutes) in Section 15.0.3 are bounding of the limiting results from the spectrum of break sizes and locations evaluated in this section.
15.6.2.5 Conclusions The acceptance criteria for an infrequent event are listed in Table 15.0-2. These acceptance criteria, followed by how the NuScale Power Plant US460 standard design meets them are listed below.
- 1) Potential core damage is evaluated on the basis that it is acceptable if the minimum departure from nucleate boiling ratio (DNBR) remains above the 95/95 DNBR limit. Minimum critical heat flux ratio (CHFR) is used instead of minimum DNBR, as described in Section 4.4.2.
The fuel integrity is not challenged by a break of a CVCS line outside containment. The fuel temperatures decrease upon the reactor trip and DHRS actuation, as shown in Figure 15.6-8, and the water level remains above the top of the active fuel, as shown in Figure 15.6-7. In addition, the event is bounded by the rapid depressurization predicted during the inadvertent ECCS actuation, which is analyzed for critical heat flux as presented in Section 15.6.6.
- 2) RCS pressure should be maintained below 120 percent of the design value.
Table 15.6-3 presents the results. The RCS pressure for the limiting RCS pressure scenario is below the acceptance criterion.
- 3) The main steam pressure should be maintained below 120 percent of the design value.
Table 15.6-3 presents the results. The main steam pressure, presented as steam generator pressure, is below the acceptance criterion.
- 4) The containment pressure should be maintained below the design pressure.
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-8 Draft Revision 1 Figure 15.6-20. The SG levels are presented in Figure 15.6-21. The RCS flow and average temperature response are shown in Figure 15.6-23 and Figure 15.6-24, respectively. DHRS flow is presented in Figure 15.6-25 and total reactivity for the event is shown in Figure 15.6-26.
The MPS is credited to protect the plant in the event of SGTF. The following MPS signals provide the plant with protection during an SGTF:
high pressurizer pressure, low pressurizer pressure, and low pressurizer level.
The MSIVs, MSIBVs, secondary MSIVs, and secondary MSIBVs are credited for isolating the faulted SG, depending on the scenario. All of these valves are qualified for the conditions analyzed and are designed to close under design basis conditions.
15.6.3.4 Radiological Consequences Audit Question A-15.0.3.7-2 Section 15.0.3 provides the radiological consequences of the SGTF. Assumptions for primary coolant release (23,000 lbm) and secondary system isolation time (30 minutes) in Section 15.0.3 are bounding of the limiting results from the spectrum of SGTF conditions evaluated in this section.
15.6.3.5 Conclusions The acceptance criteria for a postulated accident are listed in Table 15.0-2. These acceptance criteria, followed by how the NuScale Power Plant US460 standard design meets them, are listed below.
- 1) Potential core damage is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95/95 DNBR limit. MCHFR is used instead of minimum DNBR, as described in Section 4.4.2.
The fuel integrity is not challenged by a SGTF. The water level remains in the pressurizer level region, above the top of the active fuel, as shown in Figure 15.6-13 and Figure 15.6-22. In addition, the event is bounded by the rapid depressurization predicted during the inadvertent ECCS operation event, which is analyzed for critical heat flux and presented in Section 15.6.6.
- 2) RCS pressure should be maintained below 120 percent of the design value.
Table 15.6-7 presents the results of the limiting scenarios. The RCS pressure is below the acceptance criterion.
- 3) The main steam pressure should be maintained below 120 percent of the design value.