ML19129A278

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LLC Submittal of Changes to Final Safety Analysis Report Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, Related to the Decay Heat Removal System and Emergency ...
ML19129A278
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Issue date: 05/09/2019
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LO-0519-65510 May 9, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Changes to Final Safety Analysis Report Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, Related to the Decay Heat Removal System and Emergency Core Cooling System Actuation Logic

REFERENCES:

Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Revision 1 to Resolution Plans for Module Protection System Actuation Logic Changes Related to Initiation of Decay Heat Removal System and Emergency Core Cooling System, dated May 6, 2019 (ML19126A322)

During a public teleconference with members of the NRC staff on February 19, 2019, NuScale Power, LLC (NuScale) discussed updates to Final Safety Analysis Report (FSAR) related to the decay heat removal system and emergency core cooling system actuation logic. The changes affect FSAR Tier 2 Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology. The Enclosures to this letter provides a mark-up of the FSAR pages incorporating revisions, in redline/strikeout format. NuScale will include this change as part of a future revision to the NuScale Design Certification Application. is the proprietary version of NuScale FSAR Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale FSAR Section 6.2, Containment Systems, and Technical Report TR-0516-49084 Containment Response Analysis Methodology.

This letter makes no regulatory commitments or revisions to any existing regulatory commitments.

If you have any questions, please feel free to contact Rebecca Norris at 541-602-1260 or at rnorris@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Omid Tabatabai, NRC, OWFN-8H12 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0519-65510 Page 2 of 2 05/08/2019 Enclosure 1: Changes to NuScale Final Safety Analysis Report Tier 2 Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, proprietary version Enclosure 2: Changes to NuScale Final Safety Analysis Report Tier 2 Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, nonproprietary version Enclosure 3: Affidavit of Zackary W. Rad, AF-0519-65613 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0519-65510 :

Changes to NuScale Final Safety Analysis Report Tier 2 Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0519-65510 :

Changes to NuScale Final Safety Analysis Report Tier 2 Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology, nonproprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

Potential single failures are considered in the containment response analysis methodology. Due to the simplicity of the NPM design, there are few candidate single failures for the secondary system mass and energy release scenarios. Failure of ECCS valves to open would obviously reduce the mass and energy release and are not analyzed. Failures of main steam isolation valves (MSIVs) or FWIVs to close are analyzed as sensitivity studies.

1.4.3 Initial and Boundary Conditions - Secondary System Initial conditions for secondary system line break containment response analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the initial conditions is consistent with applicable DSRS guidance. The selection process ensures that energy sources are maximized and energy sinks are minimized. Initial conditions associated with primary side parameters for MSLB analyses are similar to those described for the primary mass and energy release events, with exceptions noted by Reference 6.2-1. In addition to the primary system initial conditions, secondary system initial conditions for MSLB analyses are listed in Reference 6.2-1. Initial conditions associated with primary side parameters for FWLB analyses are similar to those described for the primary mass and energy release events, with one exception noted by Reference 6.2-1. The FWLB analyses use the same secondary system initial conditions as the steam line break (SLB) analyses. Boundary conditions for secondary system line break containment response analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the boundary conditions is consistent with applicable DSRS guidance. The selection process ensures that energy sources are maximized, and energy sinks are minimized. Boundary conditions assumed by MSLB analyses are the same as those used in primary release event analyses except for those listed by Reference 6.2-1. Boundary condition assumptions for FWLB analyses are the same as those used by MSLB analyses, with one exception discussed by Reference 6.2-1.

1.4.4 Description of Blowdown Model - Secondary System 8, RAI 06.02.01.01.A-19 The MSLB is modeled as a double-ended break of a main steam line inside the CNV that depressurizes the secondary system and pressurizes the CNV. Cross connected main steam lines downstream of the main steam isolation results in both SGs discharging to containment until the steam lines isolate. A low steam line pressurehigh containment pressure signal results in closure of the main steam and feedwater isolation valves, and reactor trip. and actuation of Subsequently, DHRS actuates after feedwater isolation; however, DHRS operation is not credited for this scenario. Actuation of DHRS establishes long-term decay heat removal using the unaffected SG and the DHRS. A single failure of the FWIV to close on the affected SG is mitigated by closure of the feedwater regulating valve. After the initiation of the break, there are two potential limiting events depending on the evolution of the scenario with continued AC power, or following a loss of normal AC and DC power.

2 6.2-19 Draft Revision 3

discussed by Reference 6.2-1.

8, RAI 06.02.01.01.A-19 The FWLB is modeled as a double-ended break of the largest feedwater pipe inside the containment that results in a depressurization of the affected SG and pressurization of the CNV. A high containment pressure signal results in closure of the main steam and feedwater isolation valves and, reactor trip and actuation of the DHRS. DHRS actuation occurs subsequently after feedwater isolation. Actuation of DHRS establishes long-term decay heat removal using the unaffected SG and the DHRS. A single failure of the FWIVMSIV to close on the affected SG allows more high energy steam to be discharged out of the FWLB prior to secondary side isolation than would occur if the associated FWIV failed to close.is mitigated by closure of the feedwater regulating valve. The limiting case, described by Reference 6.2-1, also assumes a loss of normal AC and DC power at time of turbine tripevent initiation, and that results in ECCS actuation and a loss of AC power to the pressurizer heaters. With the DHRS actuation, the primary system begins a gradual cooldown and depressurization. The maximum FWLB pressure and temperature occurs after the ECCS valves open.

1.4.5 Energy Inventories - Secondary System The energy inventories in the secondary system are the same as evaluated for the primary system mass and energy releases with the exception of the additional conservatisms applied in the initial and boundary condition assumptions applied to the secondary system components, as previously discussed.

1.4.6 Additional Information Required for Confirmatory Analyses - Secondary System Information supporting confirmatory analysis is contained in Table 6.2-1 and in the containment response analysis methodology report (Reference 6.2-1).

1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of the Emergency Core Cooling System For conventional pressurized water reactor designs, the ECCS system supplies water to the reactor vessel to reflood and cool the reactor core. The core reflooding rate for these plants depends directly on containment pressure (i.e., the core flooding rate increases with increasing containment pressure). Accordingly, a minimum containment pressure analysis for ECCS performance capability is applicable for these plants.

For the NuScale facility design, ECCS operation directly connects the RPV and CNV volumes and relies on the equalization of pressures within the two volumes. The ECCS flow consists of the RCS coolant that has condensed and collected within the CNV volume being returned to the RPV. The driving force for the flow back to the RPV is provided by the static head of coolant in the CNV that collects above the return to the 2 6.2-20 Draft Revision 3

Periodic inservice inspection of the containment heat removal surfaces is performed to ensure compliance with GDC 39 to assess for surface fouling or degradation that could potentially impede heat transfer from the CNV.

The CNV inspection elements are provided in Table 6.2-3. An inspection element is a combination of a component and the inspection requirements.

A description of the ISI requirements for Class 2 and 3 components is provided in Section 6.6.

1.7 Instrumentation Requirements Instrumentation is provided to monitor the conditions inside the containment and to actuate the appropriate engineered safety features, should those conditions exceed predetermined levels. Instruments are provided to measure containment pressure, temperature, and water level. Instrumentation to monitor RCS leakage into containment and compliance with RG 1.45 is described in subsection 5.2.5.

Containment pressure instrumentation is provided for continuous control room indication to monitor containment pressure boundary integrity, RCS pressure boundary integrity, and ECCS performance, and to support the actuation of critical safety functions (reactor trip, decay heat removal actuation, CVCS isolation, and containment isolation functions).

Containment pressure is measured and monitored by four narrow range, safety-related, instruments and two wide range nonsafety-related instruments. The narrow range sensors (transducer/transmitter type) are located inside the CNV wall enclosure near the top of containment. There are four independent channels of narrow range CNV pressure instrumentation. The wide range sensors (transducer/transmitter type) are located inside the CNV wall enclosure near the top of containment. There are two independent channels of wide range CNV pressure instrumentation.

Containment water level instrumentation is provided for continuous control room indication to monitor containment pressure boundary integrity, RCS pressure boundary integrity and ECCS performance and to support the actuation of critical safety functions.

Containment water level is measured and monitored by four safety-related instruments. The sensors (digital type) are located at the reactor pressure boundary interface. There are four independent channels of CNV water level instrumentation.

Containment air temperature instrumentation is provided for continuous control room indication to monitor the environment in containment.

2 6.2-22 Draft Revision 3

CNV near the top of the CNV head. There are two independent channels of CNV air temperature instrumentation.

Additional containment instrumentation design detail addressing power supplies, actuation logic and initiation signals for the engineering safety feature functions is provided in Section 7.

2 Containment Heat Removal Containment heat removal for accident conditions is based on CNV material selection and the physical configuration of the NuScale Power Plant design. The steel CNV is partially immersed in the reactor pool and heat is transferred to the water from the outer surfaces of the CNV in contact with the water. The continuous presence of cooling water on the outside of the CNV ensures an immediate, effective, and passive means for containment heat removal. The large inventory of water in the UHS ensures a supply sufficient for long-term containment heat removal.

Under normal operating conditions, the interior of the CNV is maintained in a dry condition under vacuum (<1 psia). The primary method of heat transfer from the outer surfaces of the RPV through the containment volume to the inner containment wall is by radiation. The radiated heat energy is conducted through the CNV wall to the exterior surface where it is released via convection into the reactor pool water. For normal operations, the water level in the pool submerges the CNV in 69 ft of water with the pool water level just below the upper head of the CNV. Most of the CNV is in contact with the pool water and used for containment heat removal during operations. Section 9.1.3 describes the removal of the heat from the pool water by the active reactor pool cooling system.

Containment vacuum is maintained during normal operation by the containment evacuation system (CES) described in Section 9.3.6. Maintaining the containment under a vacuum during normal operation serves to minimize heat transfer from the RPV to the CNV and the associated loss of efficiency.

During postulated primary and secondary release events into containment, the released inventory is collected and accumulates within the CNV. Actuation of the ECCS (opening of the RVVs and RRVs) and containment isolation provides for a natural circulation coolant pathway that circulates reactor coolant inventory through the containment volume back to the RPV and through the reactor core.

In the event of a postulated MSLB or FWLB inside containment, the mass and energy released into the containment consists of the inventory present in the train with the break, including the content of the attached DHRS. The total inventory released considers the automatic isolation of the main steam and feedwater lines and single failures. The inventory released into containment flashes, condenses, and accumulates within the CNV.

The steel wall of the NuScale CNV provides for the direct (passive) transfer of containment heat (normal, transient, or accident conditions) to the UHS. There is no reliance on active components or electrical power. The design configuration provides the ability to reliably remove containment heat immediately in an accident and for at least 30 days as described 2 6.2-23 Draft Revision 3

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Executive Summary This report presents the NuScale Power, LLC, (NuScale) methodology used to analyze the mass and energy release into the containment vessel (CNV) for the spectrum of design basis transients and accidents, and the resulting pressure and temperature response of the CNV. The NuScale Power Module (NPM) limiting peak pressure and temperature results determined using the methodology are presented.

The containment response analysis methodology uses the NRELAP5 thermal-hydraulic code, which is a NuScale-modified version of the RELAP5-3D© v 4.1.3 code used for loss-of-coolant accident (LOCA) and non-LOCA transient and accident analyses, including the response of the CNV.

The NRELAP5 model used to model NPM performance for primary system LOCA and emergency core cooling system valve-opening event analyses is similar with to the model used in the LOCA evaluation model, described by Reference 7.2.1. The NRELAP5 model used for secondary system pipe-break analysis in the containment response analysis methodology is consistent with similar to the non-LOCA model described by the Non-LOCA Evaluation Model Report (Ref: 7.2.2).

Changes made to these models that maximize containment pressure and temperature response to primary and secondary system release events are described in this report. These changes conservatively maximize the mass and energy release and minimize the performance of the containment heat removal system and are consistent with acceptance criteria given by Design Specific Review Standard Section 6.2.1.3 (Ref: 7.1.6) and Design Specific Review Standard Section 6.2.1.4 (Ref: 7.1.7).

Other differences exist between the NRELAP5 model used to model NPM performance for primary system LOCA and emergency core cooling system valve-opening event analyses and the containment analysis model. These modeling differences, identified in Section 3.2.4.1, have a negligible impact on the CNV analysis results.

Initial and boundary conditions for the spectrum of primary system release containment response analyses and secondary system pipe break analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. These initial and boundary conditions are described in this report, along with the rationale for their selection.

The results of the NRELAP5 limiting analyses using the containment response analysis methodology are presented in this report. These analyses cover the spectrum of primary system mass and energy release scenarios for the NPM, and secondary system pipe break scenarios.

The limiting LOCA peak pressure and CNV wall temperature are a result of the reactor coolant system (RCS) injection line break. The LOCA limiting peak CNV wall temperature is approximately 523526 degrees F and it results from a reactor coolant system injection line break case, with a loss of normal alternating current (AC) power. The LOCA limiting peak internal pressure is approximately 921959 psia, which also results from a reactor coolant system injection line break case with a loss of normal AC and direct current (DC) power. The LOCA event peak CNV pressure is below the CNV design pressure of 10001050 psia. The LOCA peak CNV pressure and wall temperature bound the main steamline break (MSLB) and feedwater line break (FWLB) results.

© Copyright 20198 by NuScale Power, LLC 2

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 The FSAR Chapter 15 MSLB and FWLB scenarios start with the blowdown of the secondary inventory through the pipe break and into the CNV. The reactor trips on high CNV pressure or low steam line pressure, and that causes a turbine trip along with main steam isolation and feedwater isolation. One SG depressurizes as the CNV pressurizes, and an equilibrium is approached. The DHRS actuates, subsequent to feedwater isolation, and transfers decay heat to the reactor pool. Steam released into the CNV condenses on the CNV inner surface that is cooled by conduction and convection to the reactor pool.

The safety concern for the FSAR Chapter 15 main steam line breakMSLB scenario is the module response to the resulting overcooling, and the key boundary condition for the main steam line large-break scenario is the feedwater supplied to the affected SG. A single failure of the FWIV on the affected SG results in a continuation of feedwater flow until a delayed isolation occurs on feedwater regulating valve (FWRV) closure. The MSLB inside containment analysis includes the following modeling considerations:

  • break modeling with ((2(a),(c)
  • reactor trip on high CNV pressurelow steam line pressure
  • main steam isolation valves (MSIVs) actuation
  • feedwater isolation and regulating valves actuation
  • feedwater pump dynamic response
  • feedwater pipe inventory flashing
  • DHRS actuation
  • with or without loss of normal AC and direct current (DC) electrical power
  • limiting single failure The differences in the NRELAP5 MSLB modeling for the containment response analysis methodology that focus on a conservative analysis of the CNV peak pressure and temperature response are detailed in Section 3.4.13.2.4.2.

The safety concern for the FSAR Chapter 15 FWLB scenario is the module response to the overheating caused by a loss of the SG heat sink and the resulting primary system and secondary system pressurization. The key boundary conditions are the DHRS performance, which limits the peak secondary pressure, and the reactor safety valve (RSV) capacity, which limits the peak primary pressure. A single failure of the MSIV on the intact SG results in a small decrease in secondary inventory during the transition to DHRS operation, and a conservative minimum secondary heat sink. The FSAR Chapter 15 FWLB inside containment analysis includes the following modeling considerations:

  • break modeling with (( }}2(a),(c)

© Copyright 20198 by NuScale Power, LLC 31

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10

  • reactor trip on high CNV pressure
  • MSIVs actuation
  • FWIVs actuation
  • feedwater pump dynamic response
  • feedwater pipe inventory flashing
  • DHRS actuation
  • with or without loss of normal AC and direct current (DC) electrical power
  • limiting single failure The differences in the NRELAP5 feedwater line break modeling for the containment response analysis methodology that focus on a conservative analysis of the CNV peak pressure and temperature response are detailed in Section 3.4.2. The adequacy of the NRELAP5 code and the non-LOCA models for evaluation of the secondary system release scenarios is addressed in Sections 4.1 and 4.2.

3.2.4 Containment Reponse Analysis Base Model Development 3.2.4.1 NRELAP5 Primary Release Event Analysis Model Overview The NRELAP5 model used to model NPM performance for primary system (LOCA and ECCS valve opening) release event analyses is similar to the model used in the LOCA evaluation model described in Section 3.2.3.1. The NPM geometry inputs and conservative fuel inputs in the containment response analysis model are consistent with those used by the LOCA Evaluation Model. The following substantive differences are related to the objective of determining the maximum containment peak pressure and peak temperature scenarios. This is accomplished by conservatively maximizing the M&E release and minimizing containment heat removal. Figure 3-3 is an illustration of the NPM during power operation that shows the main design features. Figure 3-4 illustrates the ECCS mode of operation and shows the RVVs and RRVs along with the CNV and reactor pool that provide containment heat removal and ultimate heat sink. The nodalization diagram in Figure 3-1 plus the changes described in this section constitute the NRELAP5 model used to simulate primary release scenarios resulting from bounding breaks and valve opening events. The following modification is included in the primary release event containment response analysis model: 

        *   ((
                        }}2(a),(c)

© Copyright 20198 by NuScale Power, LLC 32

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 3-3 is an illustration of the NuScale Power Module during power operation that shows the main design features including the DHRS that actuates, subsequent to feedwater isolation, for secondary line breaks. For some secondary line break scenarios actuation of the DHRS results in a slow cooldown of the primary system and an eventual opening of the ECCS valves and a second M&E release, when a loss of power to the ECCS valve actuator solenoid occurs. Additions and modifications to this model for the secondary system M&E release analysis are the feedwater system model, the pipe break model, the CNV and the reactor pool model. These modifications to the model are described below. Feedwater System Model The feedwater system is an important boundary condition for the secondary system M&E release analyses. The initial secondary inventory in the helical coil SG is small and does not by itself cause a significant CNV pressurization following a secondary line break. The main source of mass is the feedwater system due to an assumed single failure of the FWIV on the affected helical coil SG. Also, the feedwater pump is assumed to respond to the decrease in helical coil SG pressure by a corresponding increase in feedwater flow. Feedwater flow continues to supply the affected helical coil SG until the FWRV automatically closes to back up the FWIV. Secondary Pipe Break Model The secondary pipe break spectrum modeling in the containment response analysis methodology is the same as in the Non-LOCA Methodology, with the limiting break size being the double-ended break. Figure 3-9 shows the NRELAP5 model of the MSLB. The break is modeled by closing the normal flow path (Valve 910) and by opening two break junctions (Valves 911 and 912) that start the break flow to the CNV at the appropriate elevations. Figure 3-10 depicts the NRELAP5 model of the FWLB. The break is modeled by closing the normal flow path (Valve 913) and by opening two break junctions (Valves 914 and 915) that start the break flow to the CNV at the appropriate elevations. Main steam isolation valve closure isolates the unaffected SG from the affected SG. A single failure of one MSIV to close is addressed by automatic closure of the secondary MSIV on each steam line. © Copyright 20198 by NuScale Power, LLC 44

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 containment response methodology. The basis for this statement is that CNV pressure and temperature is a figure-of-merit in the LOCA phenomena identification and ranking table. Therefore, the LOCA scenario PIRT is also considered to be the LOCA containment response analysis methodology PIRT. 3.3.1.2 Module Response The typical response of the NPM to a primary system M&E release is characterized by a simultaneous depressurization of the primary system and pressurization of the CNV. The module response depends on the size of the break or valve opening, the location of the release as that determines if the release is steam or liquid or two-phase, and the timing of the M&E releases. The resulting high containment pressure signal causes an immediate actuation of the following safety features:

  • containment isolation, including closure of MSIVs closure of FWIVs closure of backup MSIVs (non-safety) closure of FWRVs (non-safety)
  • reactor trip
  • turbine trip
  • DHRS actuation Any steam that is released through the break or valve condenses on the cold inner surface of the CNV. Condensate and any unflashed break liquid accumulates into a pool on the bottom of the CNV. The primary system level decreases due to the break or valve flow.

The ECCS actuates on the following conditions:

  • low RPV level
  • high CNV level
  • loss of normal AC power and the highly reliable DC power system The following design criteria govern RVVs and RRVs opening:
  • If the pressure differential across the valves is greater than the IAB threshold when the ECCS signal actuates, then the valves stay closed until the pressure differential decreases to below the IAB release pressure
  • If the pressure differential across the valves has decreased to below the IAB threshold pressure when the ECCS signal actuates, then the valves open and the IAB release pressure is not used Opening of the RVVs increases the depressurization rate, and the primary system and CNV pressures approach equalization. As the pressures equalize, the break/valve flow decreases. With pressure equalization and the increase in the CNV pool level, flow

© Copyright 20198 by NuScale Power, LLC 48

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 closure of primary main steam isolation valves closure of FWIVs closure of backup main steam isolation valves (non-safety) closure of FWRVs (non-safety)

  • reactor trip
  • DHRS actuation
  • turbine trip As the secondary system depressurizes, the feedwater pump flowrate increases in response to the decrease in SG pressure. Closure of the MSIVs separates the affected SG from the unaffected SG, thereby reducing the mass and energy release. Actuation of the DHRS on the unaffected SG establishes long-term decay heat removal. Closure of the FWIVs terminates the supply of secondary inventory and the affected SG boils dry. The initial primary system transient is a moderate overcooling event that does not result in ECCS actuation. Steam that is released through the break condenses on the cold inner surface of the CNV. Condensate accumulates into a pool on the bottom of the CNV. As the break flow decreases, the CNV pressure and temperature decrease and the time period of the peak values is completed. The peak pressure and temperature are significantly less than for a LOCA due to the smaller secondary inventory that is released prior to feedwater isolation.

The typical MSLB scenario is more severe when a single failure is considered. The limiting single failure is a failure of the FWIV to close on the affected SG. Closure of the FWRV is credited in this scenario, but the much longer stroke time results in a higher CNV peak pressure and temperature. Isolation of the feedwater ends the mass and energy release, and the CNV pressure and temperature then decrease due to heat transfer to the reactor pool through the CNV. When this occurs the period of peak containment pressure and temperature have been completed, and a gradual depressurization and cooling phase begins via the DHRS. The above MSLB scenario is made more adversechanged by assuming a loss of normal AC and DC power (concurrent with the break), which results in an ECCS signal and DHRS actuation following secondary system isolation. Subsequent primary system depressurization resulting from heat transfer via the DHRS along with a loss of power to the pressurizer heaters leads to ECCS actuation when the pressure differential decreases to below the IAB release pressure. Opening of the RVVs results in a second M&E release from the primary system, and the peak CNV pressure and temperature from this second release may be higher thanclose to the initial peak from the secondary system M&E release. 3.4.1.3 Limiting Event Description The limiting MSLB event is a double-ended rupture of the largest main steam line (12 in. Schedule 120 / 10.75 in. ID), which is a break area of 0.6303 ft2. Both SGs blow down into the CNV until the MSIVs close. After the initiation of the break there are two potential © Copyright 20198 by NuScale Power, LLC 52

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 limiting events depending on the evolution of the scenario with continued normal AC power, or following a loss of normal AC and DC power. For the scenario with continued normal AC power, the affected SG continues to blow down until feedwater is isolated including a single failure of the FWIV on the affected SG. This results in an extended period of feedwater delivery until the FWRV closes. The availability of power to the pressurizer heaters maintains primary system pressure and there is no ECCS actuation. The peak CNV pressure and temperature occurs as a result of the blowdown of the affected SG, and then the event is terminated. For the scenario with a loss of normal AC and DC power concurrent with the break, the feedwater pump stops and the delivery of feedwater to the affected SG is less than the case with continued normal AC and DC power. The loss of normal AC and DC power causes an ECCS actuation signal and a loss of power to the pressurizer heaters. With DHRS actuation the primary system begins a gradual cooldown and depressurization. The IAB prevents the ECCS valves from opening until the pressure differential decreases to below the IAB release pressure. Opening of the RVVs initiates a primary system M&E release with the CNV pre-heated and pressurized from the initial MSLB M&E release. This second M&E release has the potential to produce the peak CNV pressure and wall temperature results. Continued heat transfer through the CNV wall to the reactor pool results in a gradual cooldown and depressurization. Analysis of the two above scenarios has determined that the case with continued normal AC power results in the peak CNV pressure and peak CNV temperature results. 3.4.2 Feedwater Line Break Mass and Energy Methodology 3.4.2.1 Module Response The NPM initially responds to an FWLB inside the CNV with a reduction in the secondary heat sink due to the loss of feedwater flow, a depressurization of the affected SG as it blows down, and a pressurization of the CNV. Feedwater flow out the break increases due to the decrease in backpressure and due to flashing of the feedwater pipe inventory. The resulting high containment pressure signal causes an immediate actuation of the following safety features:

  • containment isolation including closure of primary main steam isolation valves closure of FWIVs closure of backup main steam isolation valves (non-safety) closure of FWRVs (non-safety)
  • reactor trip
  • DHRS actuation
  • turbine trip

© Copyright 20198 by NuScale Power, LLC 53

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Closure of the MSIVs separates the affected SG from the unaffected SG, thereby reducing the mass and energy release. Actuation of the DHRS on the unaffected SG establishes long-term decay heat removal. Closure of the FWIVs terminates the supply of feedwater to the break, and the affected SG dries out and ends the secondary mass and energy release. The primary system transient is initially a moderate overheating event that is stabilized by DHRS heat transfer, and does not result in ECCS actuation. Any steam that is released through the break condenses on the cold inner surface of the CNV. Condensate accumulates along with unflashed break liquid into a pool on the bottom of the CNV. As the break flow decreases, the CNV pressure and temperature decrease and the time period of the peak values is completed. The typical FWLB scenario is potentially more severe when a single failure is considered. The postulated single failures are a failure of the FWIV to close, or a failure of the MSIV to close, on the affected SG. Closure of the nonsafety-related FWRV, or closure of the non-safety secondary MSIV to close, is credited in this scenario, but the longer stroke times result in a higher CNV peak pressure and temperature. Isolation of the feedwater ends the secondary system mass and energy release, and the CNV pressure and temperature then decrease due to heat transfer to the reactor pool through the CNV and via the DHRS. When this occurs the period of peak containment pressure and temperature have been completed, and a gradual depressurization and cooling phase begins. The above FWLB scenario is made more adverse by assuming a loss of normal AC and DC power concurrent with turbine trip that results in an ECCS actuation signal. The loss of pressurizer heaters causes a gradual primary system depressurization during the DHRS cooldown, and subsequent opening of the RVVs when the pressure differential decreases to the IAB release pressure. Opening of the RVVs initiates a second M&E release. 3.4.2.2 Limiting Event Description For each feedwater train, the FW line geometry inside CNV changes from one 5 Schedule 120 line (between the FWIV and the FW tee) to two 4 Schedule 120 lines (between the FW tee and the FW plenum). The 0.1433 ft2 FW line break area used in the CNV analysis represents the total area of two 4 Schedule 120 lines between the FW tee and the FW plenum. The maximum break area of a single FW line inside CNV is actually 0.1136 ft2, corresponding to one 5 Schedule 120 line between the FWIV and the FW tee. However, since these two geometries are located at the same region (i.e. the FW tee), assuming a larger FW break size (0.1433 ft2) is acceptable since it conservatively maximizes mass release to the CNV. The limiting FWLB event is a double-ended rupture of the largest feedwater pipe (5 in. Schedule 120 / 4.563 in. ID), which iswith a break area of 0.14331136 ft2. The affected SG and its feedwater pipe blow down into the CNV. The unaffected SG responds to the depressurization of the affected SG until the MSIV closes. The feedwater piping on the affected SG then continues to blow down until feedwater is isolated by FWIV closure. A single failure of the FWIVMSIV to close on the affected SG is mitigated by closure of the FWRVbackup MSIV. The limiting case also assumes a loss of normal AC and DC power at time of turbine tripevent initiation, and that results in ECCS signal actuation and a loss © Copyright 20198 by NuScale Power, LLC 54

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 of power to the pressurizer heaters. With DHRS actuation, subsequent to feedwater isolation, the primary system begins a gradual cooldown and depressurization. The IAB prevents the ECCS valves from opening until the pressure differential eventually decreases to below the IAB release pressure. Opening of the RVVs combines a subsequent primary system M&E release with the initial feedwater line break M&E release and results in a significantly more severe CNV pressure and temperature response. Analysis of the above scenarios has determined that the case with loss of normal AC and DC power and ECCS actuation results in the peak CNV pressure and peak CNV temperature. A single failure of the FWIV on the affected steam generatior to close, and minimum initial primary system pressure, are included in the limiting case based on sensitivity analysis results. 3.5 Initial and Boundary Conditions 3.5.1 Primary System Release Event Initial Conditions Initial conditions for the spectrum of primary system release containment response analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the initial conditions is consistent with the guidance in DSRS Section 6.2.1.3. The selection process ensures that energy sources are maximized and energy sinks are minimized. Table 3-4 presents the primary system initial conditions for the primary system release containment response analyses. Table 3-4 Primary system initial conditions Parameter Conservative containment Rationale response analysis methodology Initial Condition (( 

                                                                             }}2(a),(c)

The initial conditions in the secondary system, in particular (( 

                                                                                                     }}2(a),(c)

© Copyright 20198 by NuScale Power, LLC 55

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 structure nodalization and (4) single failure of one RRV failing to open. The peak CNV pressure is 904894 psia for the reference case, and 921959 psia with the combined effect of the adverse sensitivity parameters. The sensitivity parameters that contribute to the 

        +1765 psi (~27.3 percent) increase are: (1) the timing of ECCS valve opening as determined by the ECCS actuation setpoint, (2) the assumption of a loss of normal AC and DC power, and 3) high primary system flow. There was a small, adverse impact from the assumed (3) single failure of one RRV failing to open (4) fine CNV volume & heat structure nodalization and (5) the RPV noncondensable release to CNV. The detailed discussion of the Case 2 results that follow are for the limiting peak CNV pressure and temperature cases.

The sequence of events (Table 5-3) show that in the first seconds following the occurrence of a LOCA in the RCS injection line many automatic responses occur to transition the module from full power operation to an alignment that mitigates the initial LOCA blowdown phase. The break flow into the CNV causes a rapid pressurization that reaches the 9.5 psia high pressure setpoint. The following automatic actions occur on high CNV pressure:

  • containment isolation resulting in MSIV and FWIV closure
  • reactor trip
  • turbine trip
  • DHRS actuation (Note: DHRS actuation is conservatively not credited in the primary system containment response analysis methodology)

As a conservative assumption, either a loss of normal AC power or a loss of normal AC and DC power is also assumed to occur at the time of the break and the ECCS signal is actuated on high CNV level or IAB release pressure.low RPV level. In the containment response analysis methodology the ECCS setpoints are important analysis input as they determine the time of the second primary system M&E release into the CNV via the ECCS valvesRVVs. The peak CNV pressure and peak CNV wall temperature occur following the ECCS valvethis RVV actuation, after the CNV has been preheated by the initial LOCA M&E release. Following the alignment of the module for the LOCA blowdown phase, the primary system pressure and inventory decrease due to the loss of inventory through the LOCA. The CNV pressurizes and the steam condenses on the cold ID of the CNV. The condensate flows down the CNV walls and accumulates in a pool in the CNV lower head. The cold CNV wall absorbs the energy of the condensed steam and starts to heat up by conduction. Eventually the energy is transferred through the CNV wall to the reactor pool, and the pool temperature slowly increases. For the peak CNV wall temperature case, the ECCS signal actuates on high CNV level at 872952 seconds, and the opening of the ECCS valves occurs at 1158955 seconds (after the IAB release pressure is reacheda 3-second signal delay). The ECCS actuation and opening of the three RVVs and one RRV causes the peak CNV wall temperature to occur at 1180978 seconds. For the peak pressure case, the ECCS signal actuates on IAB release pressurehigh CNV level at 1078364 seconds, and the opening of the ECCS valves occurs at 1081367 seconds. The ECCS actuation and opening of the three RVVs and one RRV causes the peak CNV pressure to occur at 1100385 seconds. Then, as flow through the RVVs dimishes, the primary and CNV © Copyright 20198 by NuScale Power, LLC 75

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 The sequence of events (Table 5-6) shows that in the first seconds following the occurrence of an inadvertent RRV event many automatic responses occur to transition the module from full-power operation to an alignment that mitigates the initial blowdown phase. The break flow into the CNV causes a rapid pressurization that reaches the 9.5 psia high pressure setpoint. The following automatic actions occur on high CNV pressure:

  • containment isolation resulting in MSIV and FWIV closure
  • reactor trip
  • turbine trip
  • DHRS actuation (Note: DHRS actuation is not credited in the primary system containment response analysis methodology)

As a conservative assumptionFor the peak temperature case, a loss of normal AC power, along with DC power is assumed to occur at the time of the break. RRVs and RVVs opening does not occur until the high CNV level setpoint is reached. In the containment response analysis methodology the high CNV level setpoint is an important analysis input as it determines the second primary system M&E release into the CNV through the RVVs and the second RRV. The peak CNV wall temperature occurs following the RVVs opening after the CNV has been preheated by the initial M&E release. For the peak pressure case, a loss of normal AC and DC power is also assumed to occur at the time of the break. This results in an ECCS signal. However, RRVs and RVVs opening does not occur until the differential pressure across the valve decreases to below the IAB release pressure. In the containment response analysis methodology the IAB release pressure is an important analysis input as it determines the second primary system M&E release into the CNV through the RVVs and the second RRV. The peak CNV pressure and peak CNV wall temperature occur following the RVVs opening after the CNV has been preheated by the initial M&E release. Following the alignment of the module for blowdown, the primary system pressure and inventory decrease due to the loss of inventory. The CNV pressurizes and the steam condenses on the cold interior wall of the CNV. The condensate flows down the CNV walls and accumulates along with unflashed break liquid in a pool in the CNV lower head. The cold CNV wall absorbs the energy of the condensed steam and starts to heat up by conduction. Eventually the energy is transferred through the CNV wall to the reactor pool, and the pool temperature slowly increases. Opening of the RVVsECCS valves occurs at 127171 seconds for the peak temperature case (when the high CNV level setpoint is reached) and at 77 seconds for the peak pressure case (when the RCS pressure decreases to below the 1000 psid IAB release pressure), as determined by the results of sensitivity analyses. OpeningFor the peak temperature case, opening of the three RVVs and the second RRV results in the peak CNV pressure and wall temperature at 143 and 189 seconds, respectively.182 and 180 seconds, respectively. For the peak pressure case, opening of the three RVVs and the second RRV results in the peak CNV pressure and wall temperature at 91 and 596 seconds, respectively. As flow through the RVVs diminishes, the primary and CNV pressures converge, and continued heat transfer to the CNV leads to a gradual cooldown and depressurization phase. Pressure equalization enables recirculation flow from the CNV pool through the RRVs to establish the long-term cooling recirculation alignment. © Copyright 20198 by NuScale Power, LLC 99

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 5-35 Case 5 energy balance - inadvertent reactor recirculation valve opening event 5.2 Main Steamline Break Pressure and Temperature Results The sequence of events (Table 5-7) show that in the first seconds following the occurrence of a MSLB many automatic responses occur to transition the module from full power operation to an alignment that mitigates the secondary system blowdown. The break flow into the CNV causes a rapid SG depressurization that reaches the 9.5 psia highlow steam line pressure setpoint. The following automatic actions occur on high CNVlow steam line pressure:

  • containment isolation including MSIV and FWIV closure
  • reactor trip
  • turbine trip
  • DHRS actuation Immediately following the low steam line pressure signal, the high CNV pressure signal is reached, resulting in containment isolation. Following the alignment of the module to mitigate the secondary blowdown, the secondary system pressure and inventory decrease due to the loss of inventory through the break. With continued normal AC power the

© Copyright 20198 by NuScale Power, LLC 115

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 feedwater pump initially continues to operate and supply the SGs. Feedwater isolation then terminates the supply of feedwater to the affected SG and effectively mitigates the event. The CNV pressurizes and the steam condenses on the cold ID of the CNV. The condensate flows down the CNV walls and accumulates in a pool in the CNV lower head. The cold CNV wall absorbs the energy of the condensed steam and starts to heat up by conduction. Eventually the energy is transferred through the CNV wall to the reactor pool, and the pool temperature slowly increases. The module response for the MSLB is shown in Figures 5-36 through 5-51. Figure 5-36 shows the SG pressure response with the affected SG (SG2) depressurizing via blowdown out the break into the CNV. The unaffected SG (SG1) initially depressurizes until the MSIV closes, and then gradually pressurizes following DHRS actuationisolation. Figure 5-37 shows the primary system temperature response due to the initial secondary system blowdown and then following DHRS actuation secondary side isolation. Figure 5-38 shows the primary system pressure response with the initial depressurization following secondary system blowdown, and then the pressure increasing from operation of the pressurizer heaters during DHRS operationfollowing secondary side isolation. Figure 5-39 shows that the pressurizer level rapidly decreases during the initial overcooling, and then gradually decreasesincreases in response to the decreaseincrease in primary temperatures during DHRS operationfollowing secondary side isolation. Figures 5-40 through 5-42 show the secondary system mass release, the integrated mass release, and the integrated energy release into the CNV, respectively. The liquid entrainment in the break flow was negligible, and therefore the sensitivity study on interphase drag upstream of the break flow was not necessary. The CNV and reactor pool responses for the MSLB are shown in Figures 5-43 to 5-48. Figure 5-43 shows the CNV pressure response. The pressure rapidly increases to the limiting peak value of 419449 psia at 4142 seconds. This limiting NRELAP5 result can be compared to the CNV design pressure of 10001050 psia, and to the limiting primary release event result. The MSLB result is bounded by the limiting LOCA (Case 2) and overall limiting primary release event result (Case 5). This is a key result in this MSLB containment response analysis. Figure 5-44 shows the CNV vapor temperature. (( 

                      }}2(a),(c) Figure 5-45 shows the peak CNV wall temperature and the limiting value of 427428 degrees F at 4641 seconds. This limiting NRELAP5 result can be compared to the CNV design temperature of 550 degrees F, and to the limiting LOCA result. The MSLB result is bounded by the limiting primary release event result (Case 2).

This is a key result in this MSLB containment response analysis. Figure 5-46 shows the CNV level response. Figure 5-47 shows the temperature profile across the CNV wall. There is a large temperature gradient. Figure 5-48 shows the reactor pool temperatures for a range of elevations. The reactor pool temperature does not increase significantly for the short duration of these analyses. From these results it is © Copyright 20198 by NuScale Power, LLC 116

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 5-49 Main steam line break energy balance 5.3 Feedwater Line Break Pressure and Temperature Results The sequence of events (Table 5-8) show that in the first seconds following the occurrence of an FWLB many automatic responses occur to transition the module from full power operation to an alignment that mitigates the initial secondary system blowdown phase. The break flow into the CNV causes a rapid pressurization that reaches the 9.5 psia high pressure setpoint. The following automatic actions occur on high CNV pressurefollowing an assumed loss of normal AC and DC power at the time of event initiation in the limiting case:

  • containment isolation including MSIV closure and FWIV closure
  • DHRS actuation
  • reactor trip
  • turbine trip As a conservative assumption a loss of normal AC and DC power is also assumed to occur at the time of turbine tripevent initiation. This results in an ECCS signal. However, opening

© Copyright 20198 by NuScale Power, LLC 131

LO-0519-65510 : Affidavit of Zackary w. Rad, AF-0519-65513 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying response reveals distinguishing aspects about the method by which NuScale develops its containment response analysis methodology. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed response entitled Changes to NuScale Final Safety Analysis Report Section 6.2, Containment Systems, and Technical Report TR-0516-49084, Containment Response Analysis Methodology. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. AF-0519-65513 Page 1 of 2

(5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR § 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on May 09, 2019. Zackary W. Rad AF-0519-65513 Page 2 of 2}}

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