Text

LLC Supplemental Response to NRC Request for Additional Information No. 466 (Erai No. 9482) on the NuScale Design Certification Application
	ML19304B471
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Site:	NuScale
Issue date:	10/31/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML19304B470	List: ML19304B471;
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	AF-0919-67237, RAIO-0919-67236
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RAIO-0919-67236 October 31, 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 Supplemental Response to NRC Request for Additional Information No. 466 (eRAI No. 9482) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 466 (eRAI No. 9482)," dated May 04, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 466 (eRAI No.9482)," dated October 26, 2018

3. NuScale Power, LLC Supplemental Response to NRC "Request for Additional Information No. 466 (eRAI No. 9482)," dated January 28, 2019

4. NuScale Power, LLC Supplemental Response to NRC "Request for Additional Information No. 466 (eRAI No. 9482)," dated May 22, 2019 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosures to this letter contain NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9482:

06.02.01.01.A-18 is the proprietary version of the NuScale Supplemental Response to NRC RAI No.

466 (eRAI No. 9482). 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 response.

This letter and the enclosed responses make no new regulatory commitments and no revisions to any existing regulatory commitments.

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0919-67236 If you have any questions on this response, please contact Rebecca Norris at 541-602-1260 or at rnorris@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Omid Tabatabai, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9482, proprietary : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9482, nonproprietary : Affidavit of Zackary W. Rad, AF-0919-67237 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0919-67236 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9482, proprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0919-67236 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9482, nonproprietary NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9482 Date of RAI Issue: 05/04/2018 NRC Question No.: 06.02.01.01.A-18 Conservatism in the NPM Initial Conditions for the CNV Safety Analyses To meet the General Design Criteria (GDCs) 16, 38, and 50 of Appendix A to 10 CFR Part 50 relevant to the containment design basis and guided by the Design-Specific Review Standard (DSRS) for NuScale Small Modular Reactor (SMR) Design Section 6.2.1, the staff is reviewing the applicant's analytical models and assumptions used in the containment response analysis methodology (CRAM) to determine if the licensing-basis safety analyses are acceptably conservative. Specifically, the staff needs to assess the conservatism of the licensing-basis models, constitutive/closure relations, model input parameters, and initial/boundary conditions used for the applicant's NPM design basis event (DBE) containment response analyses, in order to conclude that the results are valid over the applicable range of DBE conditions.

A limiting DBE model is expected to use the most conservative NPM initial and boundary conditions for the CNV safety analyses, based on the most biased reactor operating conditions and the limiting technical specifications. These initial conditions and assumptions should be based on the range of normal operating conditions with consideration given to maximizing the calculated peak containment pressure and temperature. In this regard, the applicant is requested to address the following three questions and update the FSAR, accordingly. The regulatory bases identified above are applicable to all questions in this RAI.

In Table 5-1 of the Containment Response Analysis Methodology Technical Report (TR-0516-49084-P, Rev. 0), the nominal CNV free volume is adjusted by ((2(a),(c) percent as a conservative initial condition for the containment response analysis to account for uncertainty in design, blockage in containment by components, such as, piping, etc. However, the base NRELAP5 model described in (( }}2(a),(c) has not been updated for numerous geometry changes reported in (( }}2(a),(c), so it is not clear that NuScale Nonproprietary

a (( }}2(a),(c) percent CNV free volume adjustment is adequate and would also cover the thermal expansion of the reactor pressure vessel (RPV) under operating conditions. The staff also noted that (( }}2(a),(c) used (( }}2(a),(c) percent conservatism in CNV free volume but that was reduced to (( }}2(a),(c) percent in Rev. 2. Please explain why a (( }}2(a),(c) percent reduction in containment volume is justified for a NRELAP5 base model that may not reflect the current design. If necessary, please update the NRELAP5 base model and resubmit the updated NRELAP5 models and their results for the limiting DBEs for CRAM, as submitted in response to RAI 8783; or justify how the peak CNV pressure and temperature results remain conservative with an outdated base model. When the licensing basis containment analyses are updated, the FSAR and the decks also need to reflect the rise of initial CNV pressure from 2 psia to 3 psia, as was concluded by RAI 8793, Question 29717 (06.02.01-2). NuScale Response: As described by NuScale letter, LO-0819-66702, dated August 26, 2019, the results of additional ECCS valve testing at reactor operating pressure and temperature necessitated an adjustment to the inadvertent actuation block (IAB) range. The letter indicated that conforming changes to the DCA would be transmitted to the NRC for review. This supplement transmits the conforming changes reflecting the revised IAB operating range. Attached are updates to FSAR Sections 6.2, 6.3, TR-0516-49084, "Containment Response Analysis Methodology," and DCA Part 4, "Technical Specifications." These changes will be incorporated into the NuScale Standard Plant DCA Revision 4. Impact on DCA: FSAR Sections 6.2 and 6.3, DCA Part 4, FSAR Table 6.2, and related Technical Report TR-0516-49084, Containment Response Analysis Methodology, have been revised as described in the response above and as shown in the markup provided with this response. NuScale Nonproprietary

Containment B 3.6.1 BASES BACKGROUND (continued)

b. De-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.2, Containment Isolation Valves; and

c. The sealing mechanism associated with each containment penetration (e.g. welds, flanges, or o-rings) is OPERABLE (i.e.,

OPERABLE such that the containment leakage limits are met). APPLICABLE The safety design basis for the containment is that the containment SAFETY must withstand the pressures and temperatures of the limiting Design ANALYSES Basis Accident (DBA) without exceeding the design leakage rates. The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA) (Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA. The DBA analyses assume that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment is designed with an allowable leakage rate of 0.20% of containment air weight after a DBA per day (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 1), as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure 99486 psia (Pa) resulting from the limiting DBA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on containment leakage rate testing. La is assumed to be 0.20% per day in the safety analysis. Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY. The containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The containment is designed to maintain leakage integrity < 1.0 La. Leakage integrity is assured by performing local leak rate testing (LLRT) and containment inservice inspection. Total LLRT leakage is maintained

               < 0.60 La in accordance with 10 CFR 50, Appendix J (Ref. 1). Satisfactory LLRT and ISI examination are required for containment OPERABILITY.

NuScale B 3.6.1-2 Draft Revision 4.0

NuScale Final Safety Analysis Report Containment Systems RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S1, RAI 06.02.01.01.A-19 The overall limiting peak CNV pressure results from an inadvertent reactor recirculation valve opening anticipated operational occurrence with a loss of normal AC and DC power. The overall limiting CNV pressure is 986994 psia, which is approximately 65 percent below the CNV design pressure of 1050 psia. The LOCA event peak CNV pressure is 959 psia. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S1, RAI 06.02.01.01.A-19 The overall peak CNV temperature is 526 degrees F, resulting from a reactor coolant system injection line break. The peak pressure and CNV wall temperature results for secondary system line break events are bounded by the LOCA results. These results demonstrate that the CNV design provides margin to the CNV design pressure of 1050 psia and CNV design temperature of 550 degrees F. The supporting analyses results are presented in Chapter 5 of the containment response analysis methodology report (Reference 6.2-1). The supporting analyses are discussed by Reference 6.2-1, as well as Section 6.2.1.3 and Section 6.2.1.4. The CNV is evaluated to demonstrate it can withstand deflagration, incident detonation and deflagration-to-detonation events for 72 hours after event initiation. Structural analysis demonstrates that the CNV is capable of withstanding the resultant combustion loads with margin to stress and strain limits as required by 10 CFR 50.44. Further details are provided in Section 6.2.5. The structural and pressure retaining components of the CNV consist of the closure flanges and bolting, vessel shells, vessel top and bottom heads, nozzles and penetrations for piping and instrumentation, access and inspection ports, CNV support skirt, CNV support lugs, bolting for the RPV upper support ledge and the NuScale Power Module top support structure mounting assemblies. Section 3.8.2 provides additional design detail that includes a physical description of the geometry of the CNV and supports, plan views, and design criteria relating to construction techniques, static loads, and seismic loads. Instrumentation is provided to monitor containment parameters for normal operation, anticipated operational occurrences and accidents to include temperature, pressure, isolation valve position and liquid level (GDC 13 and 64). Section 7.1 discusses the containment parameters monitored. The integrated design of the RPV and CNV ensures that RCS leakage is collected within the CNV. In the event of primary system releases (e.g., LOCAs or valve opening events) the CNV provides for the retention of adequate reactor coolant inventory to prevent core uncovery or loss of core cooling. The reactor coolant water that collects in the CNV is passively returned to the reactor vessel by natural circulation via the emergency core cooling system (ECCS) described in Section 6.3. Under these conditions, the CNV transfers the sensible and core decay heat through its walls to the ultimate heat sink (UHS) and provides effective passive, natural circulation emergency core cooling flow. The containment is designed so Tier 2 6.2-2 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems

core power equal to rated thermal power plus an allowance for uncertainties
conservative decay heat model
maximized RCS and fuel stored energy
secondary system stored energy sources These sources of energy are also included in secondary break release analyses, with the addition of energy resulting from feedwater pump runout.

Reference 6.2-1, Section 5.1 presents the results of the NRELAP5 base case analyses of the spectrum of primary mass and energy release scenarios for the NPM, that are determined using the containment response analysis methodology. Section 5.1 also describes sensitivity analyses used to determine the limiting primary release case assumptions for CNV pressure and wall temperature and presents their results. Sections 5.2 and 5.3 of Reference 6.2-1 present the results of NRELAP5 limiting analyses of main steam line and feedwater line break (FWLB) scenarios, respectively. Table 6.2-2 presents the results of the base case and limiting CNV pressure and wall temperature analyses for primary release (LOCA and valve opening events), as well as, limiting secondary system break scenarios. The sources and amounts of energy released to the containment and the post-accident time-dependence of mass and energy releases of postulated primary system events are described in Section 6.2.1.3. The sources and amounts of energy released to the containment and the postH-accident time-dependence of mass and energy releases of postulated secondary system pipe ruptures inside containment (main steam and feedwater line breaks) are described in Section 6.2.1.4. The capability to remove energy from the CNV (depressurization rate) is determined by the heat transfer rate from the CNV to the reactor pool. In all postulated events, containment pressure is shown to be reduced to less than 50 percent of the peak calculated pressure in less than 24 hours after the postulated accident (principal design criterion 38 (Section 3.1.4)). Specifically, for the limiting peak pressure case, the CNV pressure is reduced to less than 50 percent of its peak value in less than two hours. The NuScale CNV does not include one or more dedicated containment penetrations, equivalent in size to a single 3-foot diameter opening, to accommodate future installation of systems to prevent containment failure. As discussed in this section, the calculated peak containment pressures for design basis events remain less than the CNV internal design pressure. As discussed in Section 19.2.3, peak containment pressures do not challenge vessel integrity for any analyzed severe accident progression. Therefore, 10 CFR 50.34(f)(3)(iv) is not technically relevant to the NuScale design. Tier 2 6.2-4 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems mass and energy releases are provided in Section 6.2.1.3, and secondary system mass and energy releases are provided in Section 6.2.1.4. Graphical results for the limiting CNV pressure, temperature, and mass and energy release rates are shown in Figure 6.2-9 through Figure 6.2-14. Table 6.2-7 and Table 6.2-8 provide the sequence of events for the CNV peak pressure and peak temperature cases, respectively. In the event of a mass and energy release into CNV, a process of condensation and retention within the CNV facilitates the transfer of the energy to the UHS. Reactor coolant released from the RPV or main steam or feedwater released from the secondary system condenses on the relatively cool inner surface of the CNV wall. The resulting condensate flows down the inner CNV wall and collects in the bottom of the CNV shell. The vapor condensation and heat removal from containment is accomplished passively by transferring the energy through the CNV wall to the reactor pool. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-19 For releases from the RPV, the reactor coolant is condensed and collected until the condensate level within the CNV has increased to the ECCS actuation setpoint. Actuation of the safety system opens the RVVs and RRVs, further depressurizing the RPV and increasing the discharge of RPV inventory to the CNV. When RPV and CNV pressures approach equilibrium and the accumulated level in the CNV shell reaches a level where sufficient driving head is available, coolant flow from the CNV is returned to the RPV through the ECCS recirculation valves for core cooling. Opening of the RVVs and RRVs establishes the CNV shell as the outer boundary of the coolant circulation flow path. This method of passive coolant circulation and heat removal is further described in Section 6.2.2. For a secondary system mass and energy release into containment, the released steam or feedwater is captured within the CNV by closure of the CIVs. The collected inventory is condensed and retained with the heat energy transferred to the reactor pool. The design of the CNV is consistent with the functional requirements of the ECCS and its associated acceptance criteria. Acceptable models for evaluating emergency core cooling during the postulated mass and energy releases are defined in 10 CFR 50 Appendix K. The CNTS design provides for the isolation of process systems that penetrate the CNV. The design allows for the normal or emergency passage of fluids, vapor or gasses through the containment boundary while preserving the ability of the boundary to prevent or limit the escape of fission products in the event of postulated events. The containment isolation valves are described in Section 6.2.4. The CNV components and appurtenances are designed to ensure pressure boundary integrity for the life of the plant when considering fatigue, corrosion and wear. The CNV components and penetrations (piping, electrical and instrumentation and controls (I&C)) are designed for and tested to harsh Tier 2 6.2-7 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems The above spectrum of postulated release events bounds the primary and secondary release events for the NPM. The selection process used to determine initial conditions and boundary condition assumptions, reflecting the unique NuScale design, that are used for evaluation of containment response to postulated primary system mass and energy releases into containment are described in Reference 6.2-1, Section 3.5. Secondary system pipe break analysis initial and boundary condition assumptions and their selection process are described in Reference 6.2-1, Section 3.5. These initial conditions and assumptions are based on the range of normal operating conditions with consideration given to maximizing the calculated peak containment pressure and temperature. The results of NRELAP5 primary system release event analyses are presented by Reference 6.2-1, Section 5.1. Additionally, Reference 6.2-1, Section 5.1 discusses the insights obtained from the sensitivity studies, used to determine limiting assumptions and single failures, that create a bounding set of assumptions. These assumptions result in the limiting CNV peak temperature and pressure for primary release event Cases 1 though 5. Similarly, Reference 6.2-1, Sections 5.2 and 5.3 present the limiting CNV pressure and temperature results for main steam line and feedwater events, respectively, along with the analysis assumptions that provide these limiting results. Each mass and energy release event analyzed also includes the consideration of the worst case single active failure as identified by sensitivity cases and a determination of how the availability of normal AC and DC power affects the results, as described in detail by Reference 6.2-1. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S1, RAI 06.02.01.01.A-18S3, RAI 06.02.01.01.A-19 The limiting LOCA peak calculated containment pressure and temperature, based on the mass and energy release spectrum analyses, is postulated to occur as the result of a double-ended break of the RCS injection line. (Case 2). Considering the results of sensitivity analyses, the analysis assumes a combined simultaneous loss of normal AC power that occurs at event initiation, an inadvertent actuation block (IAB) release pressure of 1000 psid, conservatively biased ECCS actuation setpoint, fine CNV axial volume and radial CNV heat structure nodalization, RPV noncondensable release, and the single failure of one RRV to open.Considering the results of sensitivity analyses, the analysis assumes a combined simultaneous loss of normal AC power that occurs at event initiation, an inadvertent actuation block (IAB) release pressure of 1000 psid, conservatively biased ECCS actuation setpoint, fine CNV axial volume and radial CNV heat structure nodalization, no RPV noncondensable release, and the single failure of one RRV to open. The peak calculated pressure and temperature are listed in Table 6.2-2, which provide sufficient margin to CNV design pressure and temperature.is 959 psia, providing a margin of 91 psia to the CNV design pressure of 1050 psia. The peak calculated temperature is 526 degrees F, providing a margin of 24 degrees F to the CNV design temperature of 550 degrees F. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S1, RAI 06.02.01.01.A-18S3, RAI 06.02.01.01.A-19 Tier 2 6.2-10 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems The overall limiting peak calculated containment pressure, based on the mass and energy release spectrum analyses, is postulated to occur as the result of the spurious opening of a RRV anticipated operational occurrence (Case 5). The analysis models an expansion of the RCS fluid into the CNV volume and includes all relevant energy input from RCS, secondary and fuel stored energy sources, along with conservatively modeled core power and decay heat. Additional assumptions accounting for the results of sensitivity analyses, include the loss of both the normal AC power and the highly reliable DC power system (EDSS) postulated to occur at event initiation, an IAB release pressure of 1000 psid for the second RRV, and 900 psid for the RVVs, fine CNV axial volume and fine radial CNV heat structure nodalization, fine reactor pool nodalization, fast RPV noncondensable release rate, and minimum primary system flow.include the loss of normal AC power and highly reliable DC power system (EDSS) postulated to occur at event initiation and an inadvertent actuation block (IAB) release pressure of 1000 psid, fine CNV axial volume and radial CNV heat structure nodalization, fine reactor pool nodalization, RPV noncondensable release, minimum primary system flow, and single failure of one RRV to open. The peak calculated pressure is 986994 psia, providing a 6456 psia margin to the CNV design pressure of 1050 psia. Reference 6.2-1, Section 5.4 discusses the analytical and design margin incorporated into the CNV design. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S3, RAI 06.02.01.01.A-19 The peak calculated containment pressure resulting from a secondary side mass and energy release is postulated as the result of a double-ended steam line break inside containment. The analysis assumptions are documented by Reference 6.2-1, Section 5.2.assumes fine CNV axial volume and radial CNV heat structure nodalization, fine reactor pool nodalization, an IAB release pressure of 1200 psid, low RCS flow and a failure of the associated FWIV to close. The peak calculated pressure is listed in Table 6.2-2.449 psia. RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-19 The peak calculated containment temperature resulting from a secondary side mass and energy release is postulated as the result of a double-ended steam line break inside containment. The analysis assumes normal AC and DC power available, and a failure of the associated feedwater isolation valve (FWIV) to close. The peak calculated temperature is listed in Table 6.2-2.433 degrees F. The secondary system mass and energy release event results are bounded by the primary system mass and energy release events. The CNV external design pressure is 60 psia which is based on an internal pressure of 0 psia and an external pressure resulting from 100 feet of pool water static pressure. The environmental qualification of mechanical and electrical equipment exposed to the containment environment following a primary or secondary system mass and energy release inside containment is discussed in Section 3.11. Tier 2 6.2-11 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems The CNV instrumentation provided to monitor and record the required containment parameters and the capability to operate in post-accident environments is discussed in Chapter 7 and Section 3.11. 6.2.1.2 Containment Subcompartments A sub-compartment design basis and supporting analysis for mass and energy release is not relevant to the NuScale CNV because the CNV has no interior subcompartments. 6.2.1.3 Mass and Energy Release Analyses for Primary System Release Events The NuScale containment receives the primary system mass and energy released following a postulated rupture of piping containing reactor coolant, inadvertent opening of an ECCS or RSV. The CNV response analysis methodology, described in technical report TR-0516-49084, - Containment Response Analysis Methodology Technical Report, Rev 0 (Reference 6.2-1), is an extension of the NuScale LOCA evaluation model that was developed in accordance with the guidance of Regulatory Guide (RG) 1.203, "Transient and Accident Analysis Methods" December 2005. Reference 6.2-1 identifies and justifies differences in the containment response methodology, in comparison to the LOCA evaluation model. The containment response analysis methodology is based on the NRELAP5 system thermal-hydraulic code with appropriate initial and boundary conditions. The NRELAP5 code is a NuScale modified version of the RELAP5-3D Version 4.1.3 code. The NRELAP5 code is used for LOCA transient analyses. The NRELAP5 code has been qualified or assessed to separate and integral effects tests as described by Reference 6.2-2 to demonstrate the capability of the code to model LOCAs in the NPM. The LOCA evaluation model qualification activities in Reference 6.2-2 are adequate to demonstrate the capability of the NRELAP code to model containment response to LOCA events. Modifications to the inputs to the code are incorporated in order to predict maximum containment peak pressures and temperatures for the various event scenarios by conservatively maximizing the mass and energy release while minimizing the performance of containment heat removal. The LOCA analysis break spectum modeled by the NuScale containment response analysis methodology considers the same break spectrum as considered by the LOCA evaluation model (Reference 6.2-2). The maximum break opening area is modeled. The inadvertent ECCS valve opening event analysis uses the largest valve area. Break locations are chosen such that mass and energy releases to containment are maximized. The dominant consideration is the timing of the second mass and energy release that occurs when the ECCS valves open. Sensitivity studies are performed on the timing of the ECCS valve opening, and this is effectively a break size sensitivity combined with the maximum initial mass and energy release for each break location. This approach assures that the limiting cases have been identified for the peak CNV pressure and peak CNV temperature. Critical flow is evaluated using the Henry-Fauske and Moody models for subcooled and two-phase flow conditions as discussed by Reference 6.2-1. The maximum valve area and Cv values (RVV and RRV) are used to determine ECCS valve flows. Smaller values Tier 2 6.2-12 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems containment pressure and temperature have been completed, and a gradual depressurization and cooling phase begins. Sensitivity cases are performed to determine the effect of loss of power (AC or DC) scenarios, as well as postulated single failures, on the primary system mass and energy release scenarios considered by the NuScale containment response analysis methodology. The insights obtained from the results of the sensitivity cases, discussed in Reference 6.2-1, are used to determine the limiting cases for CNV pressure and temperature. 6.2.1.3.1 Mass and Energy Release Data - Primary System Release Events The maximum containment peak pressure and peak temperature scenarios are determined by conservatively modeling the mass and energy release and minimizing the performance of the containment heat removal function of containment. Reference 6.2-1, Section 5.1 provides the results of the NRELAP5 limiting analyses of the spectrum of the five primary system mass and energy release scenarios for the NPM. The limiting primary system release event (Case 5) CNV pressure results are depicted by figures contained in Reference 6.2-1, Section 5.1. Graphical results for the limiting CNV pressure, temperature and mass and energy release rates are shown in Figure 6.2-9 through Figure 6.2-14. The limiting peak pressure and temperature results are below the CNV design pressure and temperature. 6.2.1.3.2 Energy Sources - Primary System Release Events The containment response analysis methodology models available energy sources identified by 10 CFR Part 50, Appendix K, paragraph I.A, with the exception of energy associated with fuel clad metal-water reaction, since calculated cladding temperatures for design basis LOCAs remain below the threshold for cladding oxidation. Energy sources addressed in the containment response analysis analyses include

core power initialized at 102 percent of rated thermal power (163.2 MW).
decay heat modeled using the 1979 ANS standard decay heat model with a 1.2 multiplier.
RCS stored energy based on conservative initial conditions of pressure, average RCS temperature and pressurizer level that consider the normal operating range including instrumentation uncertainties and deadband.
stored energy in vessel internal structures.
RCS piping inside containment.
stored fuel energy.
stored secondary energy (steam generator (SG) tubes, main steam and feedwater piping inside containment) based on conservative initial conditions of steam pressure and feedwater temperature that consider the normal operating range including instrumentation uncertainties and deadband.

Tier 2 6.2-15 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems The containment response analysis methodology for secondary system events is based on the NRELAP5 system thermal-hydraulic code with appropriate initial and bounding conditions. The NRELAP5 code is a NuScale modified version of the RELAP5-3D Version 4.1.3 code. The NRELAP5 code is used for non-LOCA transient analyses. The NRELAP5 non-LOCA model described in the non-LOCA transient analysis methodology report (Reference 6.2-4) is used to develop the MSLB and FWLB analyses in the containment response analysis methodology. Modifications to the inputs to the code are incorporated in order to predict maximum containment peak pressures and temperatures for various event scenarios accomplished by conservatively maximizing the mass and energy release while minimizing the performance of containment heat removal. The limiting MSLB event and FWLB event are double-ended ruptures of the largest main steam line and feedwater line pipes. Secondary system mass and energy releases consist of the MSLB and FWLB events with the asymmetric responses in SGs included. The affected SG blows down into the CNV and the feedwater supply and main steam lines are isolated. Conservative modeling of secondary system mass and energy release scenarios ensures a bounding analysis. All breaks are considered with a maximum break size at each location. Critical flow is evaluated using the Moody and Henry-Fauske models for subcooled and two-phase flow conditions as discussed in Reference 6.2-1. The containment response analysis methodology uses the heat transfer correlation package in the NRELAP5 computer code for secondary system pipe break analysis. The Non-LOCA transient analysis methodology report demonstrates these correlations are applicable to the NPM design (Reference 6.2-4). The local fluid conditions and the local heat structure surface temperatures determine the heat transfer mode. Nucleate boiling heat transfer is included in the code and is selected if the local conditions are appropriate. For the helical coil SG, other heat transfer modes exist as the coolant enters as subcooled liquid and exits as superheated steam. Initial and boundary conditions are selected to maximize containment pressure and temperature response. A description of each postulated secondary system mass and energy release event is provided in Reference 6.2-1. Results of the limiting analyses are provided in Reference 6.2-1 and for FWLBs. The secondary system mass and energy analyses are fully bounded by the primary system limiting events. 6.2.1.4.1 Mass and Energy Release Data - Secondary System Similar to primary system mass and energy release scenarios, the maximum containment peak pressure and peak temperature scenarios for secondary system releases into containment are determined by conservatively modeling the mass and energy release and minimizing the performance of the heat removal function of containment. Tier 2 6.2-18 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems operating, maintenance, testing, and postulated accident conditions over its 60-year design life. Section 6.2.1 provides additional detail. For ferritic materials classified as pressure-retaining components of the RCPB, the requirements of ASME BPVC Section XI Appendix G apply. The NuScale ferritic containment pressure boundary conforms to ASME BPVC, Section II material specifications and meets the fracture toughness criteria and testing requirements identified in ASME BPVC Section III, Division 1, NB-2300 for containment pressure boundary components fabricated of ferritic materials. Fracture prevention of the containment pressure boundary is assured. RAI 06.01.01-8S1, RAI 06.01.01-9S1 The Alloy 718 heat treatment requirements discussed in Section 3.13.1.1 are applied to the Alloy 718 CNV main closure flange studs. Portions of the lower CNV have 60-year design fluence in excess of 1E+17 neutrons/cm2, E > 1 MeV, with the peak fluence in the lower CNV not exceeding 5.5E+18 neutrons/cm2, E > 1 MeV. To remove neutron embrittlement concerns in the CNV, the lower CNV is made of austenitic stainless steels which are more resistant to neutron embrittlement than ferritic materials. 6.2.8 References 6.2-1 NuScale Power, LLC, "Containment Response Analysis Methodology Technical Report," TR-0516-49084, Rev. 20. 6.2-2 NuScale Power, LLC, "Loss-of-Coolant Accident Evaluation Model," TR-0516-49422, Rev. 0, December 2016. 6.2-3 NuScale Power, LLC, "Combustible Gas Control," TR-0716-50424, Rev. 01. 6.2-4 NuScale Power, LLC, "Non-Loss-of-Coolant Transient Analysis Methodology Report," TR-0516-49416, Rev. 10, January 2017. 6.2-5 NuScale Power, LLC, "Long-Term Cooling Methodology," TR-0916-51299, Rev. 1, August 2019. 6.2-6 NuScale Power, LLC, "NuScale Containment Leakage Integrity Assurance," TR-1116-51962, Rev. 10, December 2016. 6.2-7 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition, Section XI Division 1, "Rules for Inservice Inspection of Nuclear Components," 2013 edition, Section XI Division I, New York, NY. Tier 2 6.2-61 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-18S3, RAI 06.02.01.01.A-19 Table 6.2-2: Containment Response Analysis Results3 Event Description Case Description CNV Pressure CNV Wall Temperature (psia) (°F) RCS Discharge Break Base Case 705 492 RCS Discharge Break Limiting Sensitivity Case Results 943 510 RCS Injection Line Break Base Case 894 514 RCS Injection Line Break Limiting Sensitivity Case Results 959 5262 RPV High Point Vent Degasification Base Case 554 471 Line Break RPV High Point Vent Degasification Limiting Sensitivity Case Results 901 489 Line Break Inadvertent RVV Actuation Base Case 856 483 Inadvertent RVV Actuation Limiting Sensitivity Case Results 911 486 Inadvertent RRV Actuation Base Case 941 492 Inadvertent RRV Actuation Limiting Sensitivity Case Results 9869941 512 Main Steam Line Break Limiting Results 449 433 Feedwater Line Break Limiting Results 416 408 1 Limiting NPM primary/secondary release event peak pressure, includes IAB operating range sensitivity. 2 Limiting NPM primary/secondary release event peak temperature. 3 Results reflected in this Table do not consider the impact of sensitivity studies performed to address a revised IAB operating range as discussed by Reference 6.2-1, Section 5.1.1, except as stated in Note 1. Tier 2 6.2-63 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems Table 6.2-7: Sequence of Events - Peak Containment Vessel Pressure Case Event Time (s) Transient initiated by an inadvertent RRV #1 opening into the CNV 0 High CNV pressure analytical limit is reached 0.4 Reactor trip 2 The upper end of the IAB release pressure range is reached (1000 psid) and 75 RRV #2 opens The lower end of the IAB release pressure range is reached (900 psid) and the 84 RVVs open Peak CNV internal pressure is reached (994 psia) 99 Peak CNV wall temperature is reached (for peak pressure case) 616 CNV drops to less than 50% of peak pressure value Approximately 1800 Tier 2 6.2-96 Draft Revision 4

NuScale Final Safety Analysis Report Containment Systems Table 6.2-8: Sequence of Events - Peak Containment Vessel Temperature Case Event Time (s) Transient initiated by a CVCS injection line break into the CNV 0 High CNV pressure analytical limit is reached 3 Reactor trip 5 High containment level actuation limit is reached 952 All ECCS valves open 955 Peak CNV internal pressure is reached for peak CNV temperature case 967 Peak CNV wall temperature is reached (526 °F) 978 CNV pressure drops to less than 50% of peak pressure Approximately 2500 Tier 2 6.2-97 Draft Revision 4

Tier 2 NuScale Final Safety Analysis Report Figure 6.2-9: Maximum Containment Internal Pressure - Peak Containment Vessel Pressure Case

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Tier 2 NuScale Final Safety Analysis Report Figure 6.2-10: Break and Emergency Core Cooling System Mass Release Rate - Peak Containment Vessel Pressure Case

                                                              
                                                              
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