ML24346A297
| ML24346A297 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 12/11/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24346A130 | List:
|
| References | |
| LO-175762 | |
| Download: ML24346A297 (1) | |
Text
Response to SDAA Audit Question Question Number: A-15.6.2-2 Receipt Date: 06/17/2024 Question:
FSAR Section 15.6.2 states, Failure of lines carrying primary coolant outside containment is a non-mechanistic break in the CVCS injection line, CVCS discharge line, RPV high point vent line, or pressurizer spray line. The pressurizer spray line and high point vent line are the same size, are located at the top of the pressurizer, and exit through the containment head. A break in either of these lines is bounded by the CVCS injection and discharge lines. This text appears to have been held over from the DCA. Staff examined calculation EC-122599, Rev. 0, which states, (( 2(a),(c) Propose markups to Section 15.6.2 of the FSAR to clarify which cases are bounding for each of the figures of merit and the rationale, and clarify any text that could be easily misunderstood. Clearly indicate in the markups the presence of the venturis included in the calculation or reference to other FSAR sections containing this information. NuScale Nonproprietary NuScale Nonproprietary
Response
Final Safety Analysis Report (FSAR) Section 15.6.2 identifies the four chemical and volume control system (CVCS) lines that carry primary coolant outside containment and states that a spectrum of break sizes and locations is evaluated to determine the most severe consequences. The supporting calculation EC-122599, Revision 0, Failure of Small Lines Carrying Primary Coolant Outside Containment, previously provided in the electronic reading room, ((
}}2(a),(c) FSAR Section 15.6.2 is revised as shown in the attached markups to clarify that breaks in the four lines are considered rather than bounding one line by a different line. The revisions fulfill the NRC request to
[p]ropose markups to Section 15.6.2 of the FSAR to... clarify any text that could be easily misunderstood. In addition, FSAR Section 15.6.2 is revised as shown in the attached markups to mention the flow-restricting venturis and reference the FSAR section where more information about these design features is provided. The revision fulfills the NRC request to [c]learly indicate in the markups the presence of the venturis included in the calculation or reference to other FSAR sections containing this information. The results for this event are presented in FSAR Section 15.6.2.3.3 and are for the maximum reactor pressure vessel scenario. The limiting case is identified in FSAR Section 15.6.2.3.3 as a break in the CVCS makeup (i.e., injection) line. The description of the results in FSAR Section 15.6.2.3.3, Tables 15.6-1 through 15.6-3, and Figures 15.6-1 through 15.6-10 are identified as being associated with this maximum reactor pressure vessel scenario. Similarly, the discussion of inputs and assumptions in FSAR Section 15.6.2.3.2 is also identified as being for the maximum reactor pressure scenario. The audit question request to [p]ropose markups to Section 15.6.2 of the FSAR to clarify which cases are bounding for each of the figures of merit and the rationale is not applicable because the FSAR results already identify the case. FSAR Section 15.6.2.4 identifies that the results of the spectrum of break sizes and locations are bounded by the dose analysis assumptions in FSAR Section 15.0.3, but does not provide specific results for any case. Similarly, FSAR Section 15.6.2 does not provide minimum critical heat flux ratio (MCHFR) results. It is not necessary for the FSAR to identify the cases that are limiting for these figures of merit (i.e., dose and MCHFR) since their results are not described in the FSAR. Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-1 Draft Revision 2 15.6 Decrease in Reactor Coolant Inventory This section addresses design basis events associated with a potential unplanned decrease in reactor coolant system (RCS) inventory. 15.6.1 Inadvertent Opening of a Reactor Safety Valve Reactor safety valves (RSVs) provide over-pressure protection of the NuScale Power Module (NPM). The inadvertent opening of an RSV event can be caused by a mechanical valve failure and is classified as an anticipated operational occurrence (AOO) in Table 15.0-1. The inadvertent RSV actuation event is bounded by the inadvertent operation of the emergency core cooling system (ECCS), which opens both reactor vent valves (RVVs). Inadvertent opening of an RSV with a loss of the augmented direct current (DC) power system (EDAS) is evaluated, which also opens both RVVs. This event has similar results to inadvertent ECCS operation but is less limiting for MCHFR due to the loss of EDAS causing immediate insertion of the control rods. The inadvertent ECCS operation event, including an evaluation of the inadvertent opening of an RSV, is presented in Section 15.6.6. 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment Lines that carry primary coolant outside containment are the chemical and volume control system (CVCS) injection and discharge lines, pressurizer spray lines, and reactor pressure vessel (RPV) high point vent line. The CVCS lines extend from the RPV and exit the containment vessel (CNV) through double containment isolation valves (CIVs). Failure of lines carrying primary coolant outside containment is analyzed for thermal hydraulic effects and radiological consequences. This event is classified as an infrequent event, as shown in Table 15.0-1. 15.6.2.1 Identification of Causes and Accident Description Audit Question A-15.6.2-2 Failure of lines carrying primary coolant outside containment is a non-mechanistic break in the CVCS injection line, CVCS discharge line, RPV high point vent line, or pressurizer spray line. The pressurizer spray line and high point vent line are the same size, are located at the top of the pressurizer, and exit through the containment head. The CVCS injection and discharge lines contain flow-restricting venturis inside the CNV as described in Section 6.2.1.A break in either of these lines is bounded by the CVCS injection and discharge lines. To determine the most severe consequences of the failure of lines carrying primary coolant outside containment, a spectrum of break sizes and locations is evaluated. Primary coolant is released from the break into the Reactor Building (RXB) until CVCS CIVs close. The piping carrying primary coolant outside containment are not expected to fail during the life of the plant, so this event is classified as an infrequent event, as indicated in Table 15.0-1.
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-2 Draft Revision 2 15.6.2.2 Sequence of Events and Systems Operation Audit Question A-15.6.2-2 The analysis considers the rupture of the CVCS injection and discharge lines outside containment. Primary coolant is discharged from the line break into the RXB until a CVCS isolation signal occurs. The CVCS can be isolated from the RCS by two redundant safety-related CIVs located outside containment on the exit and entry containment penetrations. All four CVCS lines isolate on a containment isolation signal. In addition, the pressurizer spray line and the RPV high point vent line isolate on low pressurizer pressure. A spectrum of break sizes and break locations are analyzed to determine the most severe consequences. The analyses show that the reactor trips, and the decay heat removal system (DHRS) actuates to remove decay heat, but ECCS actuation setpoints are not reached. Since ECCS does not actuate, once the CVCS CIVs close, the reactor coolant remains in the RPV, as opposed to discharging into containment, and shutdown of the NPM proceeds using DHRS. Since these events are terminated by containment isolation and do not rely on ECCS for mitigation, the transients are treated as non-loss-of-coolant accident (non-LOCA) events. Dose considerations for breaks in CVCS lines outside containment are addressed in Section 15.0.3. Therefore, the analysis of CVCS line breaks in this section focuses on maximizing the RCS pressure for addressing acceptance criteria. Table 15.6-1 provides the sequence of events for the maximum RCS pressure scenario. 15.6.2.3 Thermal Hydraulic and Subchannel Analyses 15.6.2.3.1 Evaluation Model The thermal hydraulic analysis of the plant response to the failure of lines carrying primary coolant outside containment is performed using NRELAP5. The NRELAP5 model is based on the design features of an NPM. The non-LOCA NRELAP5 model is discussed in Section 15.0.2. 15.6.2.3.2 Input Parameters and Initial Conditions Audit Question A-15.6.2-2 This evaluation considers the rupture of the CVCS injection line or CVCS discharge lines located outside the containment boundary. The assumptions and initial conditions of the evaluations are selected to maximize the severity of the accident by maximizing the RCS pressure. Unless specified below, the analyses assume the control systems and engineered safety features perform as designed, with allowances for instrument inaccuracy. No operator action is credited to mitigate the effects of a CVCS line break outside containment.
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-3 Draft Revision 2 Table 15.6-2 provides inputs and assumptions. The maximum RCS pressure scenario is an equivalent five percent cross-sectional area CVCS injection line break. The following are key input parameters: Audit Question A-15.6.2-2 core power (102 percent) - The higher power increases the thermal energy in the RCS at event initiation.enthalpy in the riser, where the makeup line is located, and the density in the downcomer, where the letdown line is located, is maximized at the maximum power of 102 percent. pressurizer pressure - In order to delay the low pressurizer pressure signal, which initiates CVCS isolation and terminates the break flow from the NPM, the nominal steady state pressure of 2000 psia is increased by the pressure uncertainty of 70 psia. pressurizer level (68 percent) - The pressurizer level is increased by the level measurement uncertainty of eight percent in order to delay the low pressurizer level reactor trip and CNV isolation. A combination of core parameters is used to provide a limiting power response. Sensitivity cases show the beginning-of cycle (BOC) core parameters maximize RCS pressure. Loss of normal alternating current (AC) power conditions and no loss of normal AC power conditions are examined at the start of the event and concurrent with a reactor trip. On a loss of normal AC power, the turbine is tripped and feedwater is lost. The module protection system (MPS) remains powered so safety systems are not automatically actuated. The small line failure outside of containment is detected by the MPS on low pressurizer pressure or low pressurizer level. When the turbine is tripped, the turbine stop valves close, leading to a decreased capacity of the steam generators to remove heat from the RPV. The heatup causes the pressurizer pressure to increase, the water density to decrease, and the pressurizer level to increase, which delays the event termination and maximizes RCS pressure. Therefore, a loss of AC power at the start of the event is conservative, as confirmed in sensitivity studies. No single failure is assumed. 15.6.2.3.3 Results Figure 15.6-1 to Figure 15.6-10 show the system response to the failure of lines carrying primary coolant outside containment. Table 15.6-3 contains the results of the event. Audit Question A-15.6.2-2 The maximum RPV pressure scenario starts with an equivalent five percent cross-sectional area break of the CVCS injectionmakeup line with a coincident loss of normal AC power. The turbine stop valves close as a result of the loss of normal AC power, increasing the steam line pressure and RPV pressure (Figure 15.6-1 and Figure 15.6-2). A high pressurizer pressure signal occurs, initiating a reactor trip, SSI and DHRS actuation. The reactor trip is evident in the reactor power decrease depicted in Figure 15.6-3 and DHRS flow shown}}