Text

Response to NRC Request for Additional Information Erai No. 9482
	ML18304A128
Person / Time
Site:	NuScale
Issue date:	10/26/2018
From:	; NuScale
To:	; Office of New Reactors
References
	RAIO-1018-62335
	Download: ML18304A128 (208)
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RAI0-1018-62335 Enclosure 2:

NuScale 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.: 06.02.0f01.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 Sectior:i 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(aJ,(cJ has not been updated for numerous geometry changes reported in (( }}2(a),(c), so it is not clear that NuScale Nonproprietary

a (( }}2(a),(cJ 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(aJ,(cl used (( }}2(a),(cJ percent conservatism in CNV free volume but that was reduced to (( }}2(a).(cl percent in Rev. 2. Please explain why a (( }}2caJ,(cJ 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

                                                                                                           -i and resubmit the updated NRELAP5 models and their results for the limiting DBEs for CRAM,                     i 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: The containment response analysis has been updated to incorporate the following changes: NPM Design Changes

Reduced the initial pool temperature from 140 deg-f to 110 deg-F;
Increased the initial pool level from 55 feet to 65 feet;
Revised MPS ECCS actuation high CNV level setpoint to a range of 264 inches to 300 inches;
Removal of ECCS actuation on low RPV level.

Analytical Model Changes

Implementation of NRELAP5 Version 1.4;
Resolution of errors in the analytical model:

o Correction of steadily increasing CNV vapor temperature preceding inadvertent opening of an RRV; o Increasing the CNV inside surface temperature to 240 deg-F above the pool level to correspond to the temperature at the CNV upper head, previously the temperature was set to a lower default value (approximately 11 O deg-F);

Use of a minimum containment free volume value of 6000 cu-ft, conservatively reduced below the minimum containment free volume of 6022 cu-ft, which accounts for RCS thermal expansion and includes an allowance for piping, valves, cabling and miscellaneous components such as platforms and ladders; NuScale Nonproprietary
Revised containment vessel (CNV) initial internal pressure to a conservative value of 3.0 psia;
CNV response analysis model updates to conform the CNV response analysis model to incorporate current relevant design data, consistent with the changes to TR-0516-49084, Sections 3.2.4.1 and 3.2.4.2, shown in the markup included with this eRAI response;
CNV response analysis model updates to incorporate the effects of the maximum mass of noncondensables, (approximately 66 lbm) that could exist within the CNV during operation and be released into the CNV during an event (approximately 65 lbm), for LOCA and valve opening events (see the response to eRAI 8776, Question 15.06.05-6 transmitted by NuScale letter RAI0-1017-56660, dated October 28, 2017);
Incorporation of the results of updated sensitivity analyses, including new sensitivity studies addressing the effects of noncondensable gas release rates and the impact of finer CNV axial volume, CNV heat structure radial and reactor pool nodalization models (see the response to eRAI 9380, Question 06.02.01.01.A-5 for more detail);
Updated model geometry inputs based on the current NPM design;
Conservative fuel inputs (e.g. core power shape and peaking, reactivity feedback, delayed neutron kinetic parameters, energy released per fission, scram worth, fuel thermal conductivity, fuel gap conductance) that are consistent with those used by the LOCA evaluation model;
Updated value of RCS fluid thermal expansion (i.e. 1.32%) based on the updated NPM geometry inputs in order to maximize RCS inventory release to the CNV.

The result of incorporation of the above containment analysis model changes results in an increase of the overall CNV peak internal pressure to 986 psia, which is below the CNV design pressure. Associated updates to the Containment Response Analysis Report, TR-0516-49084, are included with the response to this eRAI, Technical Specification 3.5.3, Ultimate Heat Sink is revised to modify the maximum bulk average pool temperature from 140 deg-F to 110 deg-F, and the allowance for reduced pool level raised from 55 ft. to 65 ft. The new temperature and pool level requirement is consistent with the initial conditions used in the peak containment pressure analysis. These changes are reflected in the markups included with this eRAI response. NuScale has completed analyses that support an increase of the containment design pressure from 1000 psia to 1075 psia by utilizing the stress margin reflected by TR-0516-49084, Table 5-

10. This results an approximate 8 percent margin between the CNV peak pressure determined by the updated CNV analysis and the new CNV design pressure.

NuScale Nonproprietary

The containment free volume was re-evaluated incorporating containment geometry changes and an allowance for RCS thermal expansion, piping, valves, cabling and miscellaneous components such as platforms and ladders. Incorporation of all of the above changes results in a net containment free volume of approximately 6022 cu-ft. Conservatively, 6000 cu-ft is established as a minimum containment free volume utilized as the assumed CNV free volume value in the containment response analysis. The FSAR has been updated to reflect this change. In addition, COL Item 6.2-3 is added requiring a COL applicant to confirm that the peak CNV pressure and temperature are bounded by the CNV design pressure and temperature values in consideration of the as-built CNV free volume with allowances for uncertainties and manufacturing tolerances. COL Item 6.2-3: A COL applicant that references the NuScale Power Plant design certification will

       -perform analysis that, in consideration of the as-built containment internal free volume, demonstrates that containment design pressure and temperature bounds containment peak accident pressure and temperature. The evaluation value for containment internal free volume must include margin to address the complex shapes of internal structures and components and manufacturing processes.

Details associated with the CNV design pressure increase, along with the ECCS actuation logic and setpoint changes, and the associated DCA impact, will be reflected in a supplemental response to this eRAI question. FSAR changes associated with the pool level and temperature change, will also be included in a supplemental response to this eRAI question. A sensitivity study evaluated the impact of a potential loss of pool level as a result of failure of non-seismically qualified piping connections and the non-seismically qualified dry dock gate attached between pool levels of 55' to 65'. For the sensitivity, the pool level was initialized at a 55' level, while the CNV wall temperatures were initialized at the same values from the base limiting case (i.e., crediting a 65' pool level for the initial CNV wall temperature). The impact on the peak pressure and temperature for the limiting cases is insignificant (< 1psia and 1 deg-F). Therefore, reactor pool volume above the 55' level is not credited at all for mitigating the containment response to limiting events. NuScale Nonproprietary

A sensitivity study evaluated the impact of higher initial CNV wall temperatures distributed at different elevations of the containment vessel in order to account for a potential lack of pool mixing or any uncertainty in heat transfer calculated between the RPV and CNV and to ensure that the pool temperature and initial CNV wall temperature distribution for the containment peak pressure analysis are conservative. For the sensitivity, the pool temperature was initialized at 115 deg-F instead of 110 deg-F, raising the initial CNV inner and outer wall temperatures below the pool level accordingly. This 5 deg-F increase is comparable to the largest code calculated temperature rise across the upper CNV wall heat structure of the containment vessel based on radiation and convective heat transfer from the RPV to the CNV. Assuming a uniform 5 deg-F increase among all initial CNV inner and outer wall temperatures yields a more conservative initial CNV wall temperature distribution at all elevations. Based on the sensitivity results, the impact of these higher initial CNV wall temperatures on the peak pressure and temperature for the limiting case is insignificant(< 1 psia and< 1 deg-F). This demonstrates that assuming a conservatively higher temperature distribution at the CNV wall does not significantly change the peak CNV pressure and temperature. It also demonstrates that the CNV peak pressure would remain within its safety limits when assuming a more conservative initial temperature distribution at the CNV inner and outer surfaces. The containment initial internal pressure is conservatively assumed to be 3.0 psia, corresponding to the maximum CNV internal pressure that ensures acceptable in-containment leakage detection performance at a pool temperature of 140 deg-F per FSAR Figure 5.2-3. At the revised pool high temperature limit of 110 deg-F, the maximum internal pressure is approximately 1.2 psia. Based on the results of the internal CNV pressure sensitivity study documented by the supplemental response to eRAI 8793, transmitted by NuScale letter RAI0-0917-56135, dated September 21, 2017, the peak CNV pressure determined assuming a initial internal pressure of 3.0 psia will be approximately 2 psi higher than the peak pressure corresponding the lower allowable internal pressure associated with a pool temperature of 11 O deg-F. This assumption provides additional conservatism in the updated CNV response analysis. In addition, DSRS Section 6.2.1.1.A, indicates that the technical rationale for application of this criterion for GDCs 16 and 50 is prevention of radioactive release to the environment. Containment leakage to the environment is governed by the differential pressure between the containment inner and outer surfaces. Therefore, it is appropriate to express the containment peak pressure margin to the design pressure in terms of the differential pressure between the peak calculated internal pressure and the external pressure acting against the outer surface of containment. NuScale Nonproprietary

NuScale analyses do not couple the internal and external containment pressures. Internal pressures are conservatively evaluated with an assumed external pressure of O psia. The atmospheric pressure, acting against the exterior CNV surface, is neglected. Atmospheric pressure, acting against the CNV exterior surface, provides approximately 15 psi additional margin, that is not credited by the CNV response analysis methodology. 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 peak pressure, is 986 psia, which is approximately 8 percent below the design pressure of 1075 psia, This demonstrates additional margin in the CNV design that is not considered by the containment response analysis. Impact on DCA: TR-0516-49084, Technical Specification 3.5.3, FSAR Section 6.2 and FSAR Figure 6.5-1 have been revised as described in the response above and as shown in the markup provided with the response to question 06.02.01.01.A-19 in this response. NuScale Nonproprietary

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-19 Even though the Final Safety Analysis Report (FSAR) Table 6.2-1 on containment design and operating parameters does not include the containment free volume, Table-6.5-1 on containment vessel key attributes does report containment free volume to be 1.1x107 in3 or 6365. 7 ft3, which is (( }}2(a),(c), as confirmed by the staff through an audit. The applicant is requested to update the FSAR to have consistent containment free volume across the DCA. NuScale Response: The containment free volume was re-evaluated incorporating containment geometry changes and an allowance for RCS thermal expansion, piping, valves, cabling and miscellaneous components such as platforms and ladders. Incorporation of all of the above changes results in a net containment free volume of approximately 6022 cu-ft. Accordingly, 6000 cu-ft is established as a bounding minimum containment free volume utilized as the assumed CNV free volume value in the containment response analysis. The FSAR has been updated to reflect this change. In addition, COL Item 6.2-3 is added requiring a COL applicant to confirm that the peak CNV pressure and temperature are bounded by the CNV design pressure and temperature values in consideration of the as-built CNV free volume with allowances for uncertainties and manufacturing tolerances. NuScale Nonproprietary

COL Item 6.2-3: A COL applicant that references the NuScale Power Plant design certification will perform an analysis that, in consideration of the as-built containment internal free volume, demonstrates that containment design pressure and temperature bounds containment peak accident pressure and temperature. The evaluation value for containment internal free volume must include margin to address the complex shapes of internal structures and components and manufacturing processes. Impact on DCA: FSAR Tables 1.8-2 and 6.5-1 have been revised as described in the response above and as shown in the markup provided with this response. NuScale Nonproprietary

NuScale Final Safety Analysis Report Interfaces with Certified Design RAI 01-61, RAI 02.04.13-1, RAI 03.04.01-4, RAI 03.04.02-1, RAI 03.04.02-2, RAI 03.04.02-3, RAI 03.05.01.03-1, RAI 03.05.01.04-1, RAI 03.05.02-2, RAI 03.06.02-6, RAI 03.06.02-15, RAI 03.06.03-11, RAI 03.07.01-2, RAI 03.07.01-3, RAI 03.07.02-651, RAI 03.07.02-652, RAI 03.07.02-8, RAI 03.07.02-12, RAI 03.07.02-1555, RAI 03.08.04-352, RAI 03.08.04-2351, RAI 03.08.04-2352, RAI 03.08.05-1451, RAI 03.09.02-15, RAI 03.09.02-48, RAI 03.09.02-67, RAI 03.09.02-69, RAI 03.09.03-12, RAI 03.09.06-5, RAI 03.09.06-6, RAI 03.09.06-16, RAI 03.09.06-1651, RAI 03.09.06-27, RAI 03.11-8, RAI 03.11-14, RAI 03.11-1451, RAI 03.11-18, RAI 03.13-3, RAI 04.02-152, RAI 05.02.03-19, RAI 05.02.05-8, RAI 05.04.02.01-13, RAI 05.04.02.01-14, RAI 05.04.02.01-19, RAI 06.02.01.01.A-18, RAI 06.02.01.01.A-19, RAI 06.02.06-22, RAI 06.02.06-23, RAI 06.04-1, RAI 09.01.01-20, RAI 09.01.02-4, RAI 09.01.05-3, RAI 09.01.05-6, RAI 09.03.02-3, RAI 09.03.02-4, RAI 09.03.02-5, RAI 09.03.02-6, RAI 09.03.02-8, RAI 10.02-1, RAI 10.02-2, RAI 10.02-3, RAI 10.02.03-1, RAI 10.02.03-2, RAI 10.03.06-1, RAI 10.03.06-5, RAI 10.04.06-1, RAI 10.04.06-2, RAI 10.04.06-3, RAI 10.04.10-2, RAI 11.01-2, RAI 12.03-5551, RAI 13.01.01-1, RAI 13.01.01-151, RAI 13.02.02-1, RAI 13.03-4, RAI 13.05.02.01-2, RAI 13.05.02.01-251, RAI 13.05.02.01-3, RAI 13.05.02.01-351, RAI 13.05.02.01-4, RAI 13.05.02.01-451, RAI 14.02-7, RAI 19-31, RAI 19-3151, RAI 19-38, RAI 20.01-13 Table 1.8-2: Combined License Information Items Item No. Description of COL Information Item Section COLltem 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.1 site-specific plant location. COL Item 1.1-2: A COL applicant that references the NuScale Power Plant design certification will provide the 1.1 schedules for completion of construction and commercial operation of each power module. COL Item 1.4-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.4 prime agents or contractors for the construction and operation of the nuclear power plant. COL Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site- 1.7 specific diagrams and legends, as applicable. COL Item 1.7-2: A COL applicant that references the NuScale Power Plant design certification will list additional 1.7 site-specific piping and instrumentation diagrams and legends as applicable. COL Item 1.8-1: A COL applicant that references the NuScale Power Plant design certification will provide a list of 1.8 departures from the certified design. COL Item 1.9-1: A COL applicant that references the NuScale Power Plant design certification will review and 1.9 address the conformance with regulatory criteria in effect six months before the docket date of the COL application for the site-specific portions and operational aspects of the facility design. COL Item 1.10-1: A COL applicant that references the NuScale Power Plant design certification will evaluate the 1.10 potential hazards resulting from construction activities of the new NuScale facility to the safety-related and risk significant structures, systems, and components of existing operating unit(s) and newly constructed operating unit(s) at the co-located site per 10 CFR 52.79(a)(31 ). The evaluation will include identification of management and administrative controls necessary to eliminate .or mitigate the consequences of potential hazards and demonstration that the limiting conditions for operation of an operating unit would not be exceeded. This COL item is not applicable for construction activities (build-out of the facility) at an individual Nu Scale Power Plant with operating NuScale Power Modules. COL Item 2.0-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate 2.0 that site-specific characteristics are bounded by the design parameters specified in Table 2.0-1. If site-specific values are not bounded by the values in Table 2.0-1, the COL applicant will demonstrate the acceptability of the site-specific values in the appropriate sections of its combined license application. COL Item 2.1-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.1 site geographic and demographic characteristics. COL Item 2.2-1: A COL applicant that references the NuScale Power Plant design certification will describe 2.2 nearby industrial, transportation, and military facilities. The COL applicant will demonstrate that the design is acceptable for each potential accident, or provide site-specific design alternatives. COL Item 2.3-1 : A COL applicant that references the NuScale Power Plant design certification will describe the 2.3 site-specific meteorological characteristics for Section 2.3.1 through Section 2.3.5, as applicable. COL Item 2.4-1: A COL applicant that references the NuScale Power Plant design certification will investigate 2.4 and describe the site-specific hydrologic characteristics for Section 2.4.1 through Section 2.4.14, except Section 2.4.8 and Section 2.4.1 0. Tier2 1.8-3 Draft Revision 3

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 5.4-1 : A COL applicant that references the NuScale Power Plant design certification will develop and 5.4 implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines," and applicable Electric Power Research Institute steam generator guidelines at the time of the COL application. The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity aAs aEEessieility assessment, steam plant rnrrosion prosuEt sepositioA assessmeAt, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting. COL Item 6.2-1: A COL applicant that references the NuScale Power Plant design certification will develop a 6.2 containment leakage rate testing program that will identify which option is to be implemented under 10 CFR 50, Appendix J. Option A defines a prescriptive-based testing approach whereas Option B defines a performance-based testing program. COL Item 6.2-2: A COL applicant that references the NuScale Power Plant design certification will verify that the 6.2 final design of the containment vessel meets the design basis requirement to maintain flange contact pressure at accident temperature, concurrent with peak accident pressure. CQL Item 6 2-3: A COL ai;ii;ilicant that references the NuS!;ale PQwer Plant design certifi!;a!iQn will i;ierform an 6.2 anal~sis that, in consideration Qf the as-built containment internal frgg volume, demonstrates that !::Qntainmeot design i;iremire aod timi12erat1.1re QQunds !;ontainment i;ieak a!;cident pcemne and tempecatuce Ibe ellaluatiQn llalue fuc S:Qntainment intemal fcee llQlume must include ma[gin tQ add[ess tbe S:Qrnple11 sbapes Qf internal strns:tu[es and S:QtDPQnents and rnanufactu[ing p[Qs:esses. COL Item 6.3-1: A COL applicant that references the NuScale Power Plant design certification will describe a 6.3 containment cleanliness program that limits debris within containment. The program should contain the following elements:

Foreign material exclusion controls to limit the introduction of foreign material and debris sources into containment.

                . Maintenance activity controls, including temporary changes, that confirm the emergency core cooling system function is not reduced by changes to analytical inputs or assumptions or other activities that could introduce debris or potential debris sources into containment.
                . Controls that limit the introduction of coating materials into containment .
                . An inspection program to confirm containment vessel cleanliness prior to closing for normal power operation.

COL Item 6.4-1: A COL applicant that references the NuScale Power Plant design certification will comply with 6.4 Regulatory Guide 1.78 Revision 1, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release." COL Item 6.4-2: Not used. 6.4 COL Item 6.4-3: Not used. 6.4 COL Item 6.4-4: Not used. 6.4 COL Item 6.4-5 : A COL applicant that references the Nu Scale Power Plant design certification will specify testing 6.4 and inspection requirements for the control room habitability system and control room envelope integrity testing as specified in Table 6.4-4. COL Item 6.6-1: A COL applicant that references the NuScale Power Plant design certification will implement an 6.6 inservice testing program in accordance with 10 CFR 50.55a(f). Tier 2 1.8-11 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems 6.2 Containment Systems 6.2.1 Containment Functional Design The containment is an integral part of the NuScale Power Module (NPM) and provides primary containment for the reactor coolant system (RCS). The NuScale containment system (CNTS) includes the containment vessel (CNV), CNV supports, containment isolation valves (CIVs), passive containment isolation barriers, and containment instruments. (See Figure 6.2- 1) 6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases The CNV is an evacuated pressure vessel fabricated from a combination of low alloy steel and austenitic stainless steel that houses, supports, and protects the reactor pressure vessel (RPV) from external hazards and provides a barrier to the release of fission products. The CNV is maintained partially immersed in a below grade, borated-water filled, stainless steel lined, reinforced concrete pool to facilitate heat removal. The CNV is an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Class MC (st eel) containment that is designed, ana lyzed, fabricated, inspected, tested and stamped as an ASME Code Class 1 pressure vessel. RAI 06.02 .01.0 1.A-l 8, RAI 06.02.01 .01 .A-l 9, RAI 08.01 -1 The CNTS, including the CNV, CIVs, and passive isolation barriers (refer to Section 6.2.4), provide a barrier that can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA (General Design Crite rion (GDC) SO). As a minimum, pressure retaining components that comprise the CNTS have a design pressure of at least 107500 psia and 550 degrees F, which bound the calculated pressure and temperatu re cond itions for any design basis event (DBE). In concert with the containment isolation valves (CIVs) and passive containment isolation barriers (discussed in Section 6.2.4), the CNV serves as a final barrier to the release of radioactivity and radiologica l contaminants to the environment (GDC 16). The CNV design specifications also take into consideration the pressures and temperatures associated with combustible gas deflagration. The CNV design includes no internal sub-compartments which elim inates the potential for collection of combustible gases and differential pressures resulting from postulated high-energy pipe breaks within containment. The CNV is designed to withstand the full spectrum of primary and secondary system mass and energy releases (loss-of-coolant accident (LOCA), valve opening events and non-LOCA) whi le consideri ng the worst case single active failure and loss of power conditions. Calculated peak containment pressures and temperatures are shown by analysis to rema in less than the CNV internal design pressure and temperature for analyzed events. Tier 2 6.2-1 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems RAI 06.02.01 .01 .A-18, RAI 06.02.01 .01.A-19 The limiting primary system pipe break (LOCi\) event peak pressure is 921 psia, resulting from a reactor coolant system injection line break. The LOCA peak pressure provides apprmcimately 8 percent margin to the CNV design pressure of 1000 psia. The peal< ON i,*rall temperature for this event is 523 degrees F. RAI 06.02.01.01 .A-18, RAI 06.02.01.01 .A-19 The overall ON peak pressure is 951 psi a resulting from inad 1,ertent opening of an emergency core cooling system i,*alve. The overall peak CNV temperature is 523 degrees r as discussed above. The peak pressure and ON wall temperature results for secondary system line break events are bounded by the LOCA results. These results demonstrate that the ON design provides margin to the ON design pressure of 1000 psia and ON design teA1perature of 550 degrees F. RAI 06.02.01.01.A-l 8, 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 986 psia, which is approximately 8 percent below the CNV design pressure of 1075 psi a. The LOCA event peak CNV pressure is 959 psia. RAI 06.02.01.01 .A-18, 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 1075 psia and CNV design temperature of 550 degrees F. The supporting analyses results are presented in Chapter 5 of the conta inment 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 requ ired 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. Tier 2 6.2-2 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems safety valve (RSV) opening and inadvertent ECCS valve (reactor vent valve (RW) and reactor recirculation valve (RRV)) opening. In primary system release events, the CNV design assures that sufficient inventory is retained within the CNV in a configuration that, when given a return path to the RPV (e.g., the ECCS RRVs), maintains the reactor core covered. Any primary system release event or secondary system release event transfers mass and energy into the CNV and has the potential to increase containment temperature and pressure. Peak calculated pressures and temperatures associated with these postulated events are bounded by the containment design pressure and temperature to ensure that the containment's functional capabilities are maintained under worse case conditions. Analyses of containment response to RCS 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. 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-l 8, RAI 06.02.01 .01.A-19 For releases from the RPV, the reactor coolant is condensed and collected until coolantthe condensate level within the -RPCNV has loi.'vered increased to the ECCS actuation setpoint. Actuation of the safety system opens the RWs 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 RWs 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 Tier2 6.2-7 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems

RCS discharge line break (Case 1)
RCS injection line break (Case 2)
RPV high point vent degasification supply line break (Case 3)
inadvertent opening of a RW (Case 4)
inadvertent opening of a RRV (Case 5)
steam line break
FWLB The above spectrum of postu lated 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 maxim izing the calculated peak containment pressure and temperatu re. The resu lts 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 resu lts. Each mass and energy release event analyzed also includes the consideration of the worst case single active failu re 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-l 9 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 setpoints-, fine CNV axial volume and radial CNV heat structure nodalization. BPY noncondensable release. and the single failure of one RRV to open . The peak calculated pressure is ~ 959 psia, providing a margin of -79116 psi a to the CNV Tier 2 6.2-10 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems design pressure of 107500 psia. The peak calculated temperature is 52§3- degrees F, providing a margin of 217 degrees F to the CNV design temperature of 550 degrees F. RAI 06.02.01 .01.A-18, RAI 06.02.01.01 .A- 19 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 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, aRG-minimum primary system flow, and single failure of one RRV to open. The results of single failure sensitivity studies demonstrated no adverse ON pressure impact for postulated single failures. The peak calculated pressure is 9865+ psia, providing a 4989 psia margin to the CNV design pressure of 107500 psia. Reference 6.2-1, Section 5.4 discusses the analytica l and design margin incorporated into the CNV design. RAI 06.02 .01 .01 .A-18, 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 assumes a loss of normal AC power and DC power that occurs simultaneously with a turbine trip,fine CNV axial volume and radial CNV heat structure nodalization, fine reactor pool nodalization, an IAB release pressure of 1200 psid, with DHRS available,low RCS flow and a failure of the associated FWIV to close. The peak calculated pressure is 44~ 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, with decay heat removal system (DHRS) a1,ailable, and a failure of the associated feedwater isolation valve (FWIV) to close. The peak calculated temperature is ~ 33 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 O psia and an external pressure resulting from 100 feet of pool water static pressure. Tier2 6.2-11 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems 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 (RW and RRV) are used to determine ECCS valve flows. Smaller values would result in less mass and energy release to containment; therefore, the maximum values are utilized. The containment response analysis methodology uses the heat transfer correlation package in the NRELAPS computer code. The LOCA evaluation model report demonstrates these correlations are applicable to the NPM design (Reference 6.2-2). The local fluid conditions and the local heat structure surface temperatures determine the heat transfer mode. Forced convection, natural convection, condensation, conduction, and nucleate boiling are included in the code and are selected if the local conditions are appropriate. Initial and boundary conditions are selected to maximize containment pressure and temperature response. Further details are provided in Reference 6.2-1. The containment response analysis methodology provides conservative modeling of the heat transfer to and from the CNV inside diameter, and from the CNV outside diameter to the reactor pool, to ensure a bounding peak CNV pressure and temperature response following a LOCA. The methodology includes the following elements:

radiative heating of CNV to maximize the initial inside diameter temperature and thereby minimize the initial condensation rate
high initial CNV pressure to maximize the non-condensable gas concentration
heat transfer from RPV outside diameter including convection and boiling heat transfer to the fluid in the CNV
condensation on CNV inside diameter including the effects of non-condensable gas
conservative low reactor pool level
conservative high reactor pool temperature RAI 06.02.01.01 .A-18, RAI 06.02.01 .01 .A-19
conservative low CNV free volume assumption. which accounts for RCS thermal expansion and includes an allowance for piping, valves. cabling and miscellaneous components such as platforms and ladders RAI 06.02.01.01.A-l 8, RAI 06.02.01.01.A-19 COL Item 6.2-3: A COL applicant that references the NuScale Power Plant design certification will perform an analysis that. in consideration of the as-built containment internal free volume, demonstrates that containment design pressure and temperature bounds containment peak accident pressure and temperature. The evaluation value for containment internal free volume must include margin to address the complex shapes of internal structures and components and manufacturing processes.

Reference 6.2-1 , Table 3-2 shows the heat transfer correlations and models for all of the processes that could impact the CNV peak pressure and temperature response. Tier 2 6.2-13 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems The CNV modeling in t he NRELAPS LOCA containment response analysis model is similar to the LOCA Evaluation Model (Reference 6.2-2), with certain changes made to maximize the mass and energy release and consequential containment pressure and temperature, as detailed by Reference 6.2-1. The CNV and react or pool models for secondary system pipe break containment respon se analysis methodology are the same as the modeling for primary release events. Simplified diagrams of the nodalizatio n used in the containment response analysis methodo logy are provided by Reference 6.2-1. The primary system mass and energy release events analyzed include the following:

pipe breaks RCS discharge line RCS injection line RPV high point vent degasification line
inadvertent RW opening
inadvertent RRV opening The NuScale LOCA evaluation model divided the NPM LOCA scenarios into two phases for ph eno mena identification:
LOCA blowdown phase (Phase 1a)
ECCS recirculation (Phase 1b)

For primary system mass release events, the blowdown phase begins at break initiation or valve opening. Reactor coolant released into the containment volume pressurizes the containment volume and depressurizes the RPV. Pressurization of the containment and the decreased inventory within the RPV results in reactor trip and closure of the CIVs. The blowdown phase ends when the ECCS actuates the RWs and the RRVs. The ECCS actuation occurs as a result of any of the following conditions: RAI 06.02.01 .01 .A-18, RAI 06.02.01 .01 .A-19

low RPV le*,el
high CNV level
loss of normal AC power and the EDSS The RWs and RRVs open under the following conditions:
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 at that time.

Tier 2 6.2-14 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems

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.

6.2.1.3.3 Description of th e Blowdown Model - Primary System Release Events RAI 06.02.0 1.01.A-18, RAI 06.02.01 .01.A-19 During normal power operation (normal AC and DC power available), the primary system release scenarios start with the blowdown of the primary inventory through the pipe break or valve opening into the CNV. The reactor trips on high CNV pressure, and that causes a turbine trip along with main steam isolation and feedwater isolation. The primary system depressurizes as the CNV pressurizes, and the coolant inventory accumulates in the CNV. Steam released into the CNV condenses on the CNV inner surface that is cooled by conduction and convection to the reactor pool. When the primary system in*,rentory reaches the low level setpoint, or the CNV level reaches the high level setpoint, the ECCS actuates. The ECCS valves subsequently open as described in Section 6.2.1.3. The NRELAPS primary release event model is developed from engineering information, drawings and associated reference documents to develop a thermal-hydraulic simulation model that calculates the mass and energy released from the RCS during blowdown. The containment response analysis methodology assumes an initial power level of 1.02 times the licensed power level. The initial RCS volume and mass are consistent with that power level. The mass and energy release determined by t he containment response analysis methodology is based on the NRELAPS computer code, and the modeling approach is very similar to the Nu Scale LOCA Evaluation Model that complies with the applicable portions of 10 CFR 50 Appendix K. Specific changes to the LOCA Evaluation Model required to model primary system mass release events are described by Reference 6.2-1. A discharge coefficient of 1.0 is applied to the appl icable critical flow correlation. Reference 6.2-2 demonstrates the adequacy of the LOCA Evaluation Model two-phase and single phase choked and unchoked flow models for predictions of mass and energy release based on assessments of comparisons of NRELAPS mass flow predictions to experimental data. The containment response analysis methodology uses the heat transfer correlation package in the NRELAPS computer code. The LOCA Evaluation Model report demonstrates these correlations are applicable to the NPM design (Reference 6.2-2). Tier 2 6.2-16 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems 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. 6.2.1.4.2 Single-Failure Analysis - Secondary System 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. Fai lures of main steam isolation valves (MS IVs) or FWIVs to close are analyzed as sensitivity studies. 6.2.1.4.3 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. 6.2.1.4.4 Description of Slowdown Model - Secondary System RAI 06.02.01 .01 .A-18, RAI 06.02.01 .01 .A-19 Tier 2 6.2-19 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems 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 high containment pressure signal results in closure of the main steam and feedwater isolation valves, reactor trip and actuation of DHRS. 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 FWIY to dose 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. Analysis of the two above scenarios has determined that the case with continued AC power results in the peak CNV pressure and peak CNV temperature results, as discussed by Reference 6.2-1. RAI 06.02.01 .01 .A- 18, RAI 06.02.01.01 .A-1 9 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, reactor trip and actuation of the DHRS. Actuation of DHRS establishes long-term decay heat removal using the unaffected SG and the DHRS. A single failure of the FWWMSIV 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 FWIVfailed to close.+5-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. 6.2.1.4.S 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 ofthe add itional conservatisms applied in the initial and boundary condition assumptions applied to the secondary system components, as previously discussed. 6.2.1.4.6 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). 6.2.1.S 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 Tier 2 6.2-20 Draft Revision 3

NuScale Final Safety Analysis Report Containment Systems RAI 06.02.01.01.A-l 8, RAI 06.02.01 .01.A- 19 Table 6 .2-2: Containment Response Analysis Results Event Description Case Description CNV Pressure CNV Wall Temperature (psia) (OF) RCS Discharge Break Base Case 884ZQ5. 4994 _ 22 RCS Discharge Break Limiting Sensitivity Case Results WeW ~ lli RCS Injection Line Break Base Case 004.821 ~ ill RCS Injection Line Break Limiting Sensitivity Case Results ~ 959 ~ 5-2&2 PFessl:lFir:eF S13Fay Sl:l1313lyBEY'. t:ligb Base Case 55486 47e4Zl Point Vent Degasification Line Break PFessl:lFir:eF S13Fay Sl:l1313lyBEY'. t:ligb Limiting Sensitivity Case Results 8e4.2Ql 484482 Point Vent Degasification Line Break Inadvertent RW Actuation Base Case 8m.85..Q 484483. Inadvertent RW Actuation Limiting Sensitivity Case Results 8&Glli ~ Inadvertent RRV Actuation Base Case ~w ~ Inadvertent RRV Actuation Limiting Sensitivity Case Results %-1-2.8.Q, 5Geill Main Steam Line Break Limiting Results 4+9449 ,Q-7433 Feedwater Line Break Limiting Results 4-Q.416 4+4408 1 Limiting NPM primary/ secondary release event peak pressure. 2 Limiting NPM primary/ secondary release event peak temperature. Tier 2 6.2-63 Draft Revision 3

NuScale Design Control Document Fission Product Removal and Control Systems RAI 06.02.01 .01 .A-18, RAI 06.02.01.01 .A-19 Table 6.5- 1: Containment Vessel Key Attributes Parameter Value Design-basis containment leak rate 0.2 wt% per day Containment mioimum free volume 6QQQ cu-ft-l--.--1-*+-G7-ifl~ Tier 2 6.5-3 Draft Revision 3

Ultimate Heat Sink 3.5.3 3.5 PASSIVE CORE COOLING SYSTEMS (PCCS) 3.5.3 Ultimate Heat Sink LCO 3.5.3 Ultimate Heat Sink shall be maintained within the limits specified below:

a. Level ::::: 68 ft,

b. Bulk average temperature::;; 440110 °F, and

c. Bulk average boron concentration shall be maintained within the limit specified in the COLR.

APPLICABILITY : At all times . ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0 .3 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. Ultimate Heat Sink Level A.1 Suspend module Immediately

      < 68 ft and > 55 ft                            movements.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies in the refueling area . AND A.3 Restore Ultimate Heat Sink 30 Days Level to within limits . NuScale 3.5.3-1 Draft Revision 3.0

Ultimate Heat Sink 3.5.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Ultimate Heat Sink Level B.1 Initiate action to restore Immediately

   ~ ~ 65 ft.                    Ultimate Heat Sink Level to
                                 > ~ 5 ft.

AND B.2 Restore Ultimate Heat Sink 24 Hours Level to > ~ 65 ft. C. Ultimate Heat Sink bulk C.1 Suspend module Immediately average temperature not movements. within limits. AND C.2 Initiate action to restore Immediately Ultimate Heat Sink bulk average temperature to within limits. AND C.3 Restore Ultimate Heat Sink 14 Days bulk average temperature to within limits. D. Required Action and D.1 Be in MODE 2. 6 hours associated Completion Time of Condition A, B or AND C not met. D.2 Be in MODE 3. 36 hours NuScale 3.5.3-2 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 B 3.5 PASSIVE CORE COOLING SYSTEMS (PCCS) B 3.5.3 Ultimate Heat Sink BASES BACKGROUND The ultimate heat sink (UHS) consists of three areas identified as the reactor pool (RP), refueling pool (RFP) , and spent fuel pool (SFP). The pool areas are open to each other with a weir wall partially separating the SFP from the RP and RFP. The UHS water level indicates the depth of water in the UHS from the reactor pool floor (25 ft building elevation). The UHS supports or provides multiple safety and important functions including:

a. Acts as ultimate heat sink during postu lated design basis events,

b. Provides cooling and shielding of irradiated fuel in the spent fuel storage racks,

c. Limits releases from postulated fuel handling accidents,

d. Provides a reserve of borated water for fi lling the containment vessel in MODE 3,

e. Limits the temperature of the containment vessel and module during operations,

f. Provides shielding of radiation emitted from the core of an operating module, and

g. Provides buoyancy during module movement in MODE 4.

The UHS function is performed by providing a sufficient heat sink to receive decay heat from a module via the decay heat removal system (DHRS) heat exchangers and conduction through the containment vessel walls (Ref. 1) after a postu lated Emergency Core Cooling System (ECCS) actuation and after transition to long-term shutdown cool ing (Ref. 4i ). Irradiated fuel is stored in the SFP portion of the UHS that is separated from the balance of the pool by a submerged wal l. The submerged wall includes a weir that permits movement of new and irradiated fuel from the storage areas to a reactor during refueling , and also provides a means of inventory communication between the pool areas. The SFP provides cooling and shielding of the irradiated fuel in the storage racks, and provides sufficient water level to retain iodine fission product activity in the event of a fuel handling accident. Sufficient iodine activity NuScale B 3.5.3-1 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 BASES BACKGROUND (continued) will be retained to limit offsite doses from the accident to within the values reported in FSAR Chapter 15 (Ref. 4i ). During transients and shutdowns which are not associated with design basis events in which DHRS or ECCS is actuated, water from the RP is added to the containment vessel by the Containment Flood and Drain System (CFDS). After reaching an appropriate level in the containment, the reactor vent valves (RVVs) and reactor recirculation valves (RRVs) are opened to permit improved heat transfer from the reactor coolant system (RCS) to the containment vessel walls. During normal operations, the RP limits temperatures of the module by maintaining the containment vessel partially submerged in water. The water also provides shielding above and around the region of the core during reactor operations, limiting exposure to personnel and equipment in the area. In MODE 4, the module is transported from the operating position to the RFP area of the UHS. The UHS provides buoyancy as the module displaces pool water during the movement, thereby reducing the load on the reactor building crane. APPLICABLE During all MODES of operation and storage of irradiated fuel , the UHS SAFETY supports multiple safety functions . ANALYSIS The UHS level is assumed and credited in a number of transient analyses. The 68 ft level provides buoyancy assumed in the reactor building crane analysis and design to ensure its single-failure proof capacity during module movement in MODE 4. A UHS level of ee55 ft provides margin above the minimum level required to support DHRS and ECCS operation in response to LOCA and non-LOCA design basis events. The 65 ft level also assures the containment vessel wall temperature initial condition assumed in the peak containment pressure analysis. The UHS bulk average temperature is assumed and credited , directly or indirectly in design basis accidents including those that require DHRS and ECCS operation such as LOCA and non-LOCA design basis events. The bulk average temperature is also assumed as an initial condition of the peak conta inment pressure analysis. Note that the UHS sensible heat needed to heat the pool to boiling is not credited in the NuScale B 3.5.3-2 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 BASES APPLICABLE SAFETY ANALYSIS (continued) UHS safety analyses for pool inventory. Additionally, the UHS bulk average temperature is assumed in the buoyancy calculation of the reactor building crane load during movement of the module. The UHS bulk average boron concentration lower limit is established to ensure adequate shutdown margin during unit shut downs that are not associated with events resulting in DHRS or ECCS actuation, when the module is filled with RP inventory using the CFDS and the RRVs are opened . It also ensures adequate shutdown margin when the module is configured with the UHS inventory in contact with the reactor core, specifically in MODE 4 when the containment vessel is disassembled for removal , and in MODE 5. The upper limit on boron concentration is establ ished to limit the effect of moderator temperature coefficient (MTC) during localized or UHS bulk average temperature changes while the module and core are in contact with UHS water. The upper limit also provides assurance for criticality and boron dilution analyses. The ultimate heat sink level, temperature, and boron concentration parameters satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). LCO The UHS must provide an adequate heat sink to perform its UHS function . This is accomplished by providing a-sufficient submersion of the module and the mass of water that can be heated , and vaporized to steam if necessary, to remove decay heat via the decay heat removal system or conduction through the containment vessel walls and heat from irradiated fuel in the pool. The UHS level limits ensure that this level of module submersion and mass of water is available. The UHS bulk average temperature is an initial assumption of safety analyses. The limit on temperature preserves the analysis assumptions and permits crediting the pool to mitigate these events. It also provides margin for performance of the UHS function in that the pool must be heated before vaporization of the contents will begin. Determination of the UHS bulk average temperature is in accordance with approved procedures. The boron concentration must be within limits when the UHS contents are in communication with the RCS to preserve core reactivity assumptions and analyses. Determination of the bulk average boron concentration is in accordance with approved plant procedures. NuScale B 3.5.3-3 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 BASES APPLICABILITY The limits on UHS level, bulk average temperature and bulk average boron concentration are applicable at all times. The supported safety functions are applicable in all MODES and when irradiated fuel is being handled. The applicability is conservative and recognizes the passive nature and resistance to changes that are inherent in the pool design and operation ACTIONS A.1, A.2, and A.3 With the UHS level< 68 ft but> ~ 65 ft the UHS safety function is preserved , however the margin in the safety analyses of events related to handling of spent fuel is reduced . Also, the assumed buoyancy provided by the water volume displaced by the module is reduced . Required Actions A.1 and A.2 immediately suspend module movement and the movement of irradiated fuel assemblies. This reduces the likelihood of an event that would be adversely affected by the reduced water level. Suspension of movement does not preclude movement of a module or fuel assembly to a safe position. Additionally, Required Action A.3 , the UHS level must be restored to within limits within 30 days to restore the margin and assumptions of the safety analyses related to long-term cooling of the module and irradiated fuel. The 30 days is appropriate because the UHS safety function continues to be met even if a leak results in sudden draining of the pool to refill the dry dock. The level of> ~ 65 ft ensures adequate submersion of the containment vessel walls and more than 3 days of decay heat removal without further action. B.1 and B.2 If the UHS level is :S ~ 65 ft,--tRe an initial condition assumptions of the safety analysis regarding decay heat removal peak containment pressure may not be met if a leak results in sudden drainage to the dry dock or significant pool liner leakage . Action must be immediately initiated and continued to restore the UHS level to > a.§.5 ft. NuScale B 3.5.3-4 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 BASES ACTIONS (continued) C.1, C.2, and C.3 If the UHS bulk average temperature is> 440110 °F, actions must be taken to restore the UHS bulk average temperature to within the limit. 440110 °Fis the initial temperature assumed in the UHS boiling analysispeak containment pressure analysis calculations, and_§ consistent withconservative with respect to the RB Crane lifting capacity calculation . Additionally, tihe SFPC system in conjunction with the RFP cooling system is designed to maintain a UHS bulk average temperature of~ 440110 °F. D.1 and D.2 If the UHS level or bulk average temperature cannot be returned to within limits within the associated Completion Time, the unit must be brought to a condition where the decay heat of the unit with the potential to be rejected to the UHS is minimized. To achieve this status, the unit must be brought to MODE 2 within 6 hours and MODE 3 within 36 hours. The allowed Completion Times are reasonable, based on operating requirements, to reach the required unit conditions from full power conditions in an orderly manner. E.1, E.2, E.3, E.4, and E.5 If the UHS bulk average boron concentration is not within limits, actions must be initiated and continued to restore the concentration immediately. Additionally, activities that could place pool inventory in communication with the reactor core must be suspended. Therefore, CFDS flow into the containment must be immediately terminated, and disassembly of the containment vessel that would open the RCS to communication with the UHS also suspended . Additionally, module movement must be suspended and the movement of irradiated fuel suspended . The suspension of module and/or fue l movement shall not preclude completion of movement to safe position. NuScale B 3.5.3-5 Draft Revision 3.0

Ultimate Heat Sink B 3.5.3 BASES SURVEILLANCE SR 3.5.3.1 REQUIREMENTS Verification that the UHS level is above the required minimum level will ensure that the assumed heat capacity of the pool is available and the pool will provide the credited mitigation if an irradiated fuel handling accident occurs. Indication of UHS level including alarms when not within limits are available in the main control room . The Frequency is controlled under the Surveillance Frequency Control Program. SR 3.5.3.2 Verification that the UHS bulk average temperature is within limits ensures that the safety analyses assumptions and margins provided by the UHS remain valid. Key UHS temperatures are monitored and alarmed in the control room . The Frequency is controlled under the Surveillance Frequency Control Program . SR 3.5.3.3 Verification that the UHS bulk average boron concentration is within limits ensures that the assumed safety analyses assumptions and margins provided by the UHS boron concentration remain available. Plant operations with potential to significantly affect the UHS boron concentration are controlled and indicated in the control room. The Frequency is controlled under the Surveillance Frequency Control Program . REFERENCES 1. FSAR Chapter 6, "Engineered Safety Features." L FSAR Chapter 15, "Transient and Accident Analysis." NuScale B 3.5.3-6 Draft Revision 3.0

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn Containment Response Analysis Methodology Technical Report January, 2017 Draft Revision 1G Docket: PROJ0769 NuScale Power, LLC 1100 NE Circle Blvd ., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com © Copyright 2018 by NuScale Power, LLC © Copyright 2018 by NuScale Power, LLC

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC, and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC. The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary. © Copyright 2018 by NuScale Power, LLC ii

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1Q Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008820. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States-Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or othen,yise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. © Copyright 2018 by NuScale Power, LLC iii

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 CONTENTS Abstract .......................................................................................................................................1 Executive Summary ....................................................................................................................2 1.0 Introduction .....................................................*................................................................ 4 1.1 Purpose ......................... :....................................................................................... 4 1.2 Scope .................................................................:.................................................. 4 1.3 Abbreviations ..............................................................................,.......................... 5 2.0 Background .....................................................................................................................6 2.1 Regulatory Requirements ...................................................................................... 6 2.1.1 10 CFR 50 Appendix A - General Design Criteria for Nuclear Power Plants ....................................................................................................................6 2.1.2 Regulatory Guide 1.203 ........................................................................................7 2.1.3 Design Specific Review Standard for NuScale Small Modular Reactor Design ...................................................................................................................8 3.0 Analysis .........................................................................................................................22 3.1 Modeling Software ............................................................................................... 22 3.2 NRELAP5 Base Simulation Model Development ................................................ 22 3.2.1 RELAP5-3D© .......................................................................................................22 3.2.2 RELAP5-3D© Quality Assurance .........................................................................22 3.2.3 NRELAP5 Simulation Models ..............................................................................24 3.2.4 Containment ReponseAnalysis Base Model Development ................................ 31 3.3 Containment Response Analysis Methodology for Primary System Release Events ............................................................................ :...................... 45 3.3.1 Primary System Mass and Energy Release Methodology ................................. .45 3.4 Secondary System Containment Response Analysis Methodology ................... .49 3.4.1 Steam line Break Mass and Energy Release Methodology ................................ .49 3.4.2 Feedwater Line Break Mass and Energy Methodology ....................................... 51 3.5 Initial and Boundary Conditions ........................................................................... 53 3.5.1 Primary System Release Event Initial Conditions ............................................... 53 3.5.2 Primary System Release Event Boundary Conditions ........................................ 55 3.5.3 Main Steam Line Break Initial Conditions ............................................................ 59 3.5.4 Main Steam Line Break Boundary Conditions ..................................................... 60 3.5.5 Feedwater Line Break Initial Conditions .............................................................. 63 3.5.6 Feedwater Line Break Boundary Conditions ....................................................... 63

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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 4.0 Qualification and Assessment .................................................................................... 64 4.1 Assessment of Methodology and Data ................................................................ 64 4.1.1 Primary System Release Effects Code and Model Qualification ......................... 64 4.1.2 Secondary System Pipe Break Effects Code and Model Qualification ................ 65 4.2 Testing Results .................................................................................................... 66 4.2.1 NuScale Integral System Test Facility Testing ..................................................... 66 5.0 Results ...........................................................................................................................67 5.1 Primary System Release Scenario Containment Response Analysis ................. 67 5.1.1 Analysis Approach ............................................................................................... 67 5.1.2 Reference Analysis and Sensitivity Results ......................................................... 68 5.1.3 Primary Release Scenario Pressure and Temperature Results .......................... 68 5.2 Main Steamline Break Pressure and Temperature Results ............................... 113 5.3 Feedwater Line Break Pressure and Temperature Results ................................ 129 5.4 Margin Assessment ........................................................................................... 147 5.4.1 Atmospheric Pressure ....................................................................................... 147 5.4.2 Decay Heat Removal System Availability .......................................................... 150 5.4.3 Conclusion ......................................................................................................... 150 6.0 Summary and Conclusions ........................................................................................ 151 7 .0 References ................................................................................................................... 153 7 .1 Source Documents ............................................................................................ 153 7.2 Reference Documents ....................................................................................... 153 8.0 Appendicies .................................................................................................................154 8.1 Mass and Energy Input ..................................................................................... 154 8.2 Heat Sink Tables ................................................................................................ 154 8.2.1 Listing of Passive Heat Sinks ............................................................................ 167 8.2.2 Modeling of Passive Heat Sinks ........................................................................ 167 8.2.3 Thickness Groups ............................................................................................. 167 8.2.4 Properties of Passive Heat Sink Materials ........................................................ 168 TABLES Table 1-1 Abbreviations ......................................................................................................... 5 Table 2-1 Compliance with Design Specific Review Standard Section 6.2.1 ........................ 8 Table 2-2 Compliance with Design Specific Review Standard Section 6.2.1.1.A. ............... 10 Table 2-3 Compliance with Design Specific Review Standard Section 6.2.1.3 ................... 13 © Copyright 2018 by NuScale Power, LLC V

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 2-4 Compliance with Design Specific Review Standard Section 6.2.1.4 ................... 19 Table 3-1 New NRELAP5 models ....................................................................................... 23 Table 3-2 Containment vessel and reactor pool heat transfer modeling ............................ .40 Table 3-3 Primary system mass and energy release scenarios ......................................... .48 Table 3-4 Primary system initial conditions ......................................................................... 53 Table 3-5 Containment vessel and reactor pool initial conditions ........................................ 54 Table 3-6 Primary system boundary conditions ................................................................... 56 Table 3-7 Secondary system initial conditions .................................................................... 60 Table 3-8 Boundary conditions for the main steam line break containment response analysis methodology .......................................................................................... 61 Table 5-1 Initial conditions for primary system release event analyses ............................... 69 Table 5-2 Case 1 sequence of events - reactor coolant system discharge line break loss-of-coolant accident .................................................................................................. 70 Table 5-3 Case 2 sequence of events for limiting containment vessel temperature event - reactor coolant system injection line break loss-of-coolant accident ....... 75 Table 5-4 Case 3 sequence of events - RPV high point degasification line break loss-of-coolant accident. ...................................................................................... 91 Table 5-5 Case 4 sequence of events - inadvertent reactor vent valve opening event ...... 94 Table 5-6 Case 5 sequence of events - inadvertent reactor recirculation valve opening event. ...................................................................................................................99 Table 5-7 Main steam line break sequence of events ....................................................... 115 Table 5-8 Feedwater line break sequence of events ......................................................... 132 Table 8-1 Limiting Peak Pressure Case - Mass and Energy Release .............................. 154 Table 8-2 Limiting Peak Wall Temperature Case - Mass and Energy Release ................ 159 Table 8-3 Limiting Secondary Break Peak Pressure Mass and Energy Release .............. 165 Table 8-4 Passive heat sinks ............................................................................................. 167 Table 8-5 Thickness groups .............................................................................................. 167 Table 8-6 Physical properties of passive heat sink materials ............................................ 168 FIGURES Figure 3-1 NRELAP5 NuScale Power Module noding diagram ............................................ 27 Figure 3-2 NRELAP5 nodalization for non-loss-of-coolant accident evaluation model. ........ 29 Figure 3-3 NuScale module during power operation ............................................................ 32 Figure 3-4 NuScale module during emergency core cooling system operation .................... 32 Figure 3-5 NRELAP5 nodalization for reactor coolant system discharge line break loss-of-coolant accident. ...................................................................................... 34 Figure 3-6 NRELAP5 nodalization for reactor coolant system injection line break loss-of-coolant accident. ...................................................................................... 35 Figure 3-7 NRELAP5 nodalization for pressurizer spray supply line break and RPV high point vent degasification line loss-of-coolant accident.. ....................................... 36 Figure 3-8 NRELAPS reactor pool model ............................................................................. 39 Figure 3-9 Main steam llne break model ..............................................................................43 Figure 3-10 Feedwater line break model ................................................................................44 Figure 5-1 Case 1 containment vessel pressure - reactor coolant system discharge line break loss-of-coolant accident. ............................................................................ 71 Figure 5-2 Case 1 containment vessel wall temperature - reactor coolant system discharge line break loss-of-coolant accident ...................................................................... 72 © Copyright 2018 by NuScale Power, LLC vi

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 5-3 Case 2 primary system pressure - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) .................................................... 76 Figure 5-4 Case 2 pressurizer level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) .................................................... 77 Figure 5-5 Case 2 riser level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ............................................................................ 78 Figure 5-6 Case 2 primary temperatures - reactor coolant system injection line break loss-of-coolant accident. ...................................................................................... 79 Figure 5-7 Case 2 break and emergency core cooling system mass flowrate - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ........................................................................................... 80 Figure 5-8 Case 2 integrated loss-of-coolant accident and emergency core cooling system mass release - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ........................................................................... 81 Figure 5-9 Case 2 integrated loss-of-coolant accident and emergency core cooling system energy relea~e - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ........................................................................... 82 Figure 5-10 Case 2 containment vessel pressure - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ......................................... 83 Figure 5-11 Case 2 containment vessel level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ................................................... 84 Figure 5-12 Case 2 containment vessel vapor temperature - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) .................... 85 Figure 5-13 Case 2 containment vessel wall temperature profile -reactor coolant system injection line break loss-of-coolant accident (peak pressure case) .................... 86 Figure 5-14 Case 2 reactor pool temperatures - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) ................................................... 87 Figure 5-15 Case 2 energy balance - reactor coolant system injection llne break loss-of-coolant accident (peak pressure case) ................................................... 88 Figure 5-16 Case 2 containment vessel pressure - reactor coolant system injection line break loss-of-coolant accident (peak CNV wall temperature case) .................... 89 Figure 5-17 Case 2 containment vessel peak wall temperature - reactor coolant system injection line break loss-of-coolant accident (peak CNV wall temperature case) ............................................................................................... 90 Figure 5-18 Case 3 containment vessel pressure - high point vent line break loss-of-coolant accident ...............................................................................................................92 Figure 5-19 Case 3 containment vessel wall temperature - high point vent line break loss-of-coolant accident. ...... :............................................................................... 93 Figure 5-20 Case 4 containment vessel pressure - inadvertent reactor vent valve opening event .................................................................................................................... 95 Figure 5-21 Case 4 containment vessel wall temperature - inadvertent reactor vent valve opening event ...................................................................................................... 96 Figure 5-22 Case 5 primary pressure - inadvertent reactor recirculation valve opening event. ................................................................................................................. 100 Figure 5-23 Case 5 pressurizer level - inadvertent reactor recirculation valve opening event

                  .......................................................................................................................... 101 Figure 5-24       Case 5 riser level - inadvertent reactor recirculation valve opening event ........ 102 Figure 5-25       Case 5 primary temperature - inadvertent reactor recirculation valve opening event .................................................................................................................. 103

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1G Figure 5-26 Case 5 loss-of-coolant accident and emergency core cooling system flowrate - inadvertent reactor recirculation valve opening event ....................... 104 Figure 5-27 Case 5 integrated loss-of-coolant accident and emergency core cooling system mass flow rate - inadvertent reactor recirculation valve opening event ........................ :........................................................................... 105 Figure 5-28 Case 5 integrated loss-of-coolant accident and emergency core cooling system energy release - inadvertent reactor recirculation valve opening event .................................................................................................... 106 Figure 5-29 Case 5 containment vessel pressure - inadvertent reactor recirculation valve opening event (overall limiting pressure case) .................................................. 107 Figure 5-30 Case 5 containment vessel level - inadvertent reactor recirculation valve opening event .................................................................................................... 108 Figure 5-31 Case 5 containment vessel vapor temperature - inadvertent reactor recirculation valve opening event ...................................................................... 109 Figure 5-32 Case 5 containment vessel wall temperature - inadvertent reactor recirculation valve opening event ...................................................................... 11 O Figure 5-33 Case 5 containment vessel wall temperature profile - inadvertent reactor recirculation valve opening event ...................................................................... 111 Figure 5-34 Case 5 reactor pool temperature - inadvertent reactor recirculation valve opening event .................................................................................................... 112 Figure 5-35 Case 5 energy balance - inadvertent reactor recirculation valve opening event .................................................................................................... 113 Figure 5-36 Main steam line break steam generator pressure ............................................. 116 Figure 5-37 Main steam line break primary temperature ...................................................... 117 Figure 5-38 Main steam line break primary system pressure ............................................... 118 Figure 5-39 Main steam line break pressurizer level ............................................................ 119 Figure 5-40 Main steam line break and emergency core cooling system flowrate ............... 120 Figure 5-41 Main steam line break and emergency core cooling system integrated mass release ............................................................................................................... 121 Figure 5-42 Main steam line break integrated energy release .............................................. 122 Figure 5-43 Main steam line break containment vessel pressure ......................................... 123 Figure 5-44 Main steam line break containment vessel vapor temperature ......................... 124 Figure 5-45 Main steam line break containment vessel wall temperature ............................ 125 Figure 5-46 Main steam line break containment vessel level ............................................... 126 Figure 5-47 Main steam line break containment vessel wall temperature profile ................. 127 Figure 5-48 Main steam line break reactor pool temperature .. ,............................................ 128 Figure 5-49 Main steam line break energy balance .............................................................. 129 Figure 5-50 Feedwater line break steam generator pressure ............................................... 133 Figure 5-51 Feedwater line break primary temperature ........................................................ 134 Figure 5-52 Feedwater line break pressurizer level. ............................................................. 135 Figure 5-53 Feedwater line break riser level ........................................................................ 136 Figure 5-54 Feedwater line break primary system pressure ................................................. 137 Figure 5-55 Feedwater line break and emergency core cooling system flowrate ................. 138 Figure 5-56 Feedwater line break and ECCS integrated mass release ................................ 139 Figure 5-57 Feedwater line break and ECCS integrated energy release ............................. 140 Figure 5-58 Feedwater line break containment vessel pressure .......................................... 141 Figure 5-59 Feedwater line break containment vessel vapor temperature ........................... 142 Figure 5-60 Feedwater line break containment vessel wall temperature .............................. 143 Figure 5-61 Feedwater line break containment vessel level. ................................................ 144 © Copyright 2018 by NuScale Power, LLC viii

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 5-62 Feedwater line break containment vessel wall temperature profile ................... 145 Figure 5-63 Feedwater line break reactor pool temperature ................................................. 146 Figure 5-64 Feedwater line break energy balance ............................................................... 147 © Copyright 2018 by NuScale Power, LLC ix

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Abstract This report presents the NuScale Power, LLC, 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 Nu Scale Power Module (NPM) limiting peak pressure and temperature results determined using the methodology are presented. This report demonstrates

that the NuScale Power Module containment vessel design accommodates the limiting loss-of-coolant and non-loss-of-coolant events, with respect to peak accident pressure and temperature, including sufficient margin. This report also demonstrates conformance to 10 CFR 50 Appendix A, General Design Criteria (GDC) 16 and 50, and Principal Design Criterion (PDC) 38 along with compliance with relevant Acceptance Criteria given by the Design Specific Review Standard for NuScale Small Modular Reactor Design, Section 6.2.1 (Reference 7.1.4).

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 cool ing system valve-opening event analyses is similar w+tA-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 th is 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 ~ 526 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 ~ 959 psia , which affiG-results from a reactor coolant system injection line break case with a loss of normal AC and DC power. The LOCA event peak CNV pressure is below the CNV design pressure of 40001075 psia. The LOCA peak CNV pressure and wall temperature bound the main steamline break (MSLB) and feedwater line break (FWLB) results. © Copyright 2018 by NuScale Power, LLC 2

Containment Response Analysis Methodology Technical Report TR-05 16-49084-NP Draft Rev. 10 The overall limiting peak CNV accident pressure is approximately ~ 986 psia, wh ich is approximately 8 percent below the containment design pressure of 1075 psia. It results from an inadvertent reactor recirculation valve opening anticipated operational occurrence with a loss of normal AC and direct current (Dq power. The peak pressure of the limiting anticipated operational occurrence is also less than the CNV design pressure of 1000 psia. The CNV pressure for this limiting case is reduced to below 50 percent of the peak value in less than 2 hours, demonstrating adequate NPM containment heat removal. Section 5.4 discusses margin in the NPM design that is not included in the CNV design pressure rating or modeled in the containment response analyses. Design factors conservatively not credited include atmospheric pressure acting against the CNV shell stress margins, CNV cladding materialexterior surface and the availability of the decay heat removal system (DHRS). The containment response analysis methodology demonstrates that the NPM design has adequate margin to design limits and that it satisfies the requirements of General Design Criteria (GDC) 16, 50, and Principal Design Criterion (PDC) 38. © Copyright 2018 by NuScale Power, LLC 3

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 1.0 Introduction 1.1 Purpose The purpose of this report is to present the NuScale Power, LLC, methodology used to analyze the mass and energy (M&E) 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, and to present the NuScale Power Module (NPM)- limiting peak pressure and temperature results that are determined using the methodology. 1.2 Scope The scope of the Containment Response Analysis Technical Report comprises the M&E release from the spectrum of primary system and secondary system design basis transients and accidents and the resulting CNV pressure and temperature response. The duration of the analyses is sufficient to establish the CNV peak pressure and peak temperature for all events, and to demonstrate the decrease in pressure to one-half of the peak value within 24 hours. The NRELAP5 code , described in Reference 7.2.1, is used in this methodology. The simulation models used in the containment response analysis methodology are similar to the models used in the NuScale LOCA and non-LOCA methdoologies (Reference 7.2.1 and Reference 7.2.2). This report documents the differences compared to those methodologies and provides bounding analysis results for the limiting accident scenarios. Operation at rated power is the bounding initial condition for the limiting CNV pressure and temperature event scenarios for the NPM. Operation at rated power is the bounding initial condition because it has the maximum stored energy and decay heat. For the NPM , reduced power levels and shutdown conditions are non-limiting and do not need to be analyzed specifically. Chapter 2.0 describes the regulatory guidance that is applicable to the scope of the containment response analysis methodology and summarizes how the methodology meets the guidance. Chapter 3.0 describes the NRELAP5 computer code along with the qualification of the code for the scope of the containment response analysis methodology. Chapter 3.0 also describes the NRELAP5 model of the NPM used in the conta inment response analysis methodology. Chapter 4.0 describes validation and verification of the containment response analysis methodology as well as primary and secondary release event models, including the code and model qualification and conservativisms. Chapter 5.0 presents the containment response analysis methodology for primary system release events, the limiting scenarios and associated analysis results, and nominal case results demonstrating conservatism in assumed certain initial conditions. Chapter 5.0 also presents the containment response analysis methodology for secondary-system pipe breaks along with associated analysis results . Chapter 6.0 presents the report summary and conclusions. The methodology for simulation of the longer-term M&E release and CNV and NPM response that is used for establishing the equipment qualification (EQ) pressure and © Copyright 2018 by NuScale Power, LLC 4

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 temperature envelopes, and to demonstrate the long-term cooling capabilities of the NPM , are not included within the scope of this report. 1.3 Abbreviations Table 1-1 Abbreviations Term Definition AC _ _ __ _ alternating current ASME American Society of Mechanical Engineers ANS American Nuclear Society BPVC Boiler and Pressure Vessel Code CFR Code of Federal Regulations CNV containment vessel eves chemical and volume control system DC direct current DHRS decay heat removal system DSRS Design Specific Review Standard ECCS emergency core cooling system FSAR Final Safety Analysis Report FWIV feedwater isolation valve FWLB feedwater line break FWRV feedwater regulating valve GDC General Design Criteria IAB inadvertent actuation block ID inside diameter LOCA loss-of-coolant accident M&E mass and energy MSIV main steam isolation valve MSLB main steam line break NIST-1 NuScale Integral System Test Facility NPM NuScale Power Module NRC U. S. Nuclear Regulatory_ Commission OD outer diameter PDC Principal Design Criterion PIRT phenomena identification and ranking table PWR pressurized water reactor RCS reactor coolant system RPV reactor pressure vessel RRV reactor recirculation valve RSV reactor safety va~ RW reactor vent valve SG steam generator SMR small modular reactor SRP Standard Review Plan © Copyright 2018 by NuScale Power, LLC 5

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 2.0 Background The CNV is a compact, steel pressure vessel that consists of an upright cylinder with top and bottom head closures. The CNV is partially immersed in a below-grade reactor pool that provides a passive heat sink and is absent of internal sumps or subcompartments that could entrap water or gases. The CNV and the reactor pool are housed within a Seismic Category 1 Reactor Building. The unique nature of the NPM design necessitates development of a specific containment response analysis methodology. This technical report describes the thermal-hydraulic accident analysis methodology for primary and secondary system M&E releases into the CNV of the NPM, and the resulting pressure and temperature response of the CNV. This report presents the bases for the analysis methodology and results in support of Chapter 6 of the NuScale Final Safety Analysis Report (FSAR). The containment response analysis methodology and CNV peak pressure and temperature results are compared to applicable regu latory guidance, including the Design Specific Review Standard for NuScale Small Modular Reactor (SMR) Design , Section 6.2 .1 (Ref: 7.1.4). A spectrum of M&E release events is analyzed that bounds all of the LOCAs and valve-opening transients in the primary system and all secondary-system pipe-break accidents. The containment response analysis methodology uses conservative initial conditions and boundary conditions to ensure overall conservative results . The limiting results are shown to be less than the design pressure (40001075 psia) and the design temperature (550 degrees F) of the CNV. The qualification of the LOCA and non-LOCA methodologies presented in References 7.2.1 and 7.2.2, in particular the comparisons to separate effects tests and integral effects tests , are applicable for the containment response analysis methodology presented in th is report. The differences in the NRELAP5 simulation models used in the containment response analysis methodology as compared to the LOCA and non-LOCA models, along with the rationale for the selection of conservative initial and boundary conditions , are the subject of this report. Analysis results are presented for the limiting cases , along with nominal condition case results, demonstrating conservatism in certain initial conditions . 2.1 Regulatory Requirements The Nuclear Regulatory Commission (NRC) regulations and regulatory guidance applicable to the containment response analysis methodology are described in this section . The elements of the containment response analysis methodology that address each of these regulations and requirements are discussed . 2.1.1 10 CFR 50 Appendix A-General Design Criteria for Nuclear Power Plants The General Design Criteria (GDC) for Nuclear Power Plants, Appendix A to 10 CFR 50 (Ref: 7.1.2), include the NRC regulations applicable to the containment response methodology. Compl iance with GDC 16 and 50 and PDC 38 is as follows : General Design Criterion 16 - The analyses performed per the containment response analysis methodology are used to establish the limiting CNV pressure and temperature conditions resulting from the spectrum of design-basis primary system and secondary system M&E releases resulting from pipe breaks and valve actuations. The CNV is © Copyright 2018 by NuScale Power, LLC 6

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 designed to ensure that the design pressure and temperature limit are not exceeded as demonstrated by the analysis results. Principal Design Criterion 38 - The analyses performed per the containment response analysis methodology establish the performance of NPM containment heat removal and demonstrate that the containment peak pressure and temperature are rapidly reduced. The methodology addresses LOCAs, valve-opening events and secondary pipe breaks. Following containment isolation and opening of the ECCS valves, the containment heat removal function is passive and does not require electric power. The requirement to rapidly reduce the containment pressure and temperature is demonstrated by the peak pressure decreasing to less than 50 percent of the peak value consistent with Design Specific Review Standard (DSRS) Section 6.2.1.1.A (Ref: 7 .1.5). Potential single failures have been considered in the methodology, and the results of the analyses show that the safety functions can be performed including the limiting single failure. General Design Criterion 50 - The analyses performed per the containment response analysis methodology demonstrate that sufficient margin to the CNV design pressure and temperature is maintained. The methodology explicitly models all energy sources including energy in the steam generators (SGs). However, the energy from the post-LOCA oxidation of the cladding that is typical of light water reactors is not applicable to the NuScale design and is not included. Calculated cladding temperatures for design basis LOCAs are below the level where cladding oxidation occurs on a time scale of a LOCA event for the NPM. Therefore, this requirement is satisfied by the design that precludes fuel temperature reaching critical heat flux and any significant fuel cladding heatup. For the NPM loss-of-coolant accident evaluation model core coverage and a minimur;n critical heat flux ratio are significantly greater than the safety limit, which precludes the occurrence of cladding oxidation (see Reference 7.2.1, Section 2.2). The NRELAP5 code and model have been assessed to experimental data to demonstrate the capability to reliably simulate the scenarios of interest. Conservative values for initial conditions and boundary conditions ensure an overall conservative analysis result. 2.1.2 Regulatory Guide 1.203 Regulatory Guide 1.203, "Transient and Accident Analysis Methods" (Ref: 7 .1.3), describes a process that the NRC staff considers acceptable for industry use to develop and assess evaluation models used to analyze transient and accident behavior that is within the design basis of a nuclear power plant. An evaluation model is the calculational framework for evaluating the behavior of the reactor system during a postulated transient or design basis accident. The containment response analysis methodology is an extension of the NuScale LOCA and non-LOCA methodologies developed following the guidance of Regulatory Guide 1.203. This report references the LOCA and non-LOCA methodologies and identifies and justifies the differences in the containment response methodology when compared to those methodologies. © Copyright 2018 by NuScale Power, LLC 7

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 2.1.3 Design Specific Review Standard for NuScale Small Modular Reactor Design The NRC has issued "Design-Specific Review Standard for NuScale SMR Design" to guide the NRC staff review of the NuScale FSAR. This document replaces NUREG-0800, "Standard Review Plan." The NRC staff has specified the DSRS as an acceptable method for evaluating whether an application complies with NRC regulations for NuScale small modular reactor (SMR) applications, provided that the application does not deviate significantly from the design and siting assumptions made by the NRC staff while preparing the DSRS. The DSRS is used by NuScale as a guide to ensure that the containment response analysis methodology addresses all of the elements that NRC has included. Sections 2.2.3.1 through 2.2.3.4 describe how the containment response analysis methodology is consistent with the applicable DSRS guidelines, justify differences, or indicate non-applicability. 2.1.3.1 Design Specific Review Standard 6.2.1 Containment Functional Design The DSRS Section 6.2.1, "Containment Functional Design" (Ref: 7 .1.4 ), includes a high-level summary of an acceptable approach and content for a containment response analysis methodology, and references the lower-tier subsections with additional detail about the approach and contents. The comparison of the containment response analysis methodology to applicable content in DSRS Section 6.2.1 is provided in Table 2-1: Table 2-1 Compliance with Design Specific Review Standard Section 6.2.1 DSRS Section 6.2.1, p. 1 Containment Response Analysis Methodoloav The containment structure must be The containment response analysis capable of withstanding, without loss of methodology addresses LOCAs function, the pressure and temperature resulting from postulated limiting conditions resulting from postulated loss- breaks, valve-opening events, main of-coolant (LOCA), steam line, or steam line break (MSLB) accidents, feedwater line break accidents. and feedwater line break (FWLB) accidents. A conservative approach to modeling the full spectrum of break and valve sizes and locations is included. The limiting results are less than the CNV design pressure and temperature. The containment design basis includes The containment response analysis the effects of stored energy in the reactor methodology includes all primary coolant system, decay energy, and energy system and secondary energy sources from other sources such as the secondary that contribute to the M&E release. system, and metal-water reactions The energy from the post-LOCA including the recombination of hydrogen oxidation of the cladding that is typical and oxygen. of light water reactors is not applicable to the NuScale design and is not included as discussed by Section 2.1.1. © Copyright 2018 by NuScale Power, LLC 8

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 The subsequent thermodynamic effects in The containment response analysis the containment resulting from the release methodology uses the NRELAP5 of the coolant mciss and energy are system thermal-hydraulic analysis determined from a solution of the code. NRELAP5 solves the time-incremental space and time-dependent dependent conservation equations for energy, mass, and momentum mass, momentum, and energy. conservation equations. DSRS Section 6.2.1, p. 2 Containment Response Analysis Methodoloav GDC 50, among other things, requires that The containment response analysis consideration be given to the potential methodology models engineered safety consequences of degraded engineered features including NPM containment safety features, such as the containment heat removal and the ECCS with heat removal system and the ECCS, the conservative assumptions. Postulated limitations in defining accident single failures are considered. Initial phenomena, and the conservatism of and boundary conditions are selected calculation models and input parameters to maximize containment pressure and in assessing containment design margins. temperature response. Margin is maintained between the analysis results and the CNV design pressure and temperature limits (See Section 5.2.2). The regulation in 10 CFR 50Appendix The containment response analysis K. I.A provides the sources of energy that methodology includes all of the sources are required and acceptable to be of energy required in Appendix K.I.A included in determining the mass and with the following exceptions to Items 4 energy release from loss-of-coolant and 5: 4) Fission Product Decay: The accidents and secondary systems pipe American Nuclear Society (ANS)-5.1-ruptures. 1979 decay heat standard with a two-sigma uncertainty is used rather than 120 percent of the 1971 American Nuclear Society (ANS) standard. Consistent with DSRS 6.2.1.3, Section 11, Acceptance Criterion 1.C. v, the ANS-5.1-1979 standard is equal to the decay heat model given in Standard Review Plan (SRP) Section 9.2.5.

5) Metal-Water Reaction: The energy from the post-LOCA oxidation of the cladding that is typical of light water reactors is not applicable to the NuScale design and is not included as discussed in Section 2.1.1.

DSRS Section 6.2.1, p. 4 Containment Response Analysis Methodoloav The temperature and pressure profiles Methodology for simulation of the M&E provided in the applicant's technical release and CNV response that is used submittal for the spectrum of LOCA and for establishing the equipment main steam line break accidents are qualification pressure and temperature acceptable for use in equipment envelopes, and to demonstrate the qualification (i.e., there is reasonable long-term cooling capabilities of the assurance that the actual temperatures NPM, are outside of the scope of this and pressures for the postulated accidents report. © Copyright 2018 by NuScale Power, LLC 9

Containment Response Analysis Methodology Technical Report TR-0516--49084-NP Draft Rev. 10 will not exceed these profiles anywhere within the specified environmental zones, except in the break zone). 2.1.3.2 Design Specific Review Standard 6.2.1.1.A Containment The DSRS Section 6.2.1.1.A, "Containment" (Ref: 7.1.5), includes content related to containment design, including some elements that are associated with the capability to withstand M&E releases. The comparison of the conta inment response analysis methodology to applicable content in DSRS Section 6.2.1.1.A is provided in Table 2-2: Table 2-2 Compliance with Design Specific Review Standard Section 6.2.1.1.A DSRS Section 6.2.1.1.A, p. 1 Containment Response Analysis Methodoloav The temperature and pressure The containment response analysis conditions in the containment due to a methodology includes the spectrum of spectrum (including break size and primary release events resulting from location) of postulated loss-of-coolant postulated limiting breaks (LOCAs) and accidents (LOCAs) (i.e., reactor coolant valve openings, MSLB accidents, and system pipe breaks) and secondary FWLB accidents. The limiting results are system steam and feedwater line breaks less than the CNV design pressure and temperature. The effectiveness of static (passive) and The containment response analysis active heat removal mechanisms . methodology includes conservative modeling of passive heat removal systems (there are no active heat removal systems in the NuScale design). Specifically, conservatisms are employed in conservative assumed initial and boundary conditions, including the reactor pool to ensure a bounding peak CNV peak pressure and temperature following events involving release of mass and energy into the CNV. The performance of these systems is shown to be effective in limiting the CNV pressure and temperature response to within acceptable design limits. Conservatism in initial and boundary conditions is discussed in Section 3.5 DSRS Section 6.2.1.1 .A , p. 4 Containment Response Analysis Methodology To satisfy the requ irements of GDC 16 For the NuScale FSAR submittal , the and 50 regarding sufficient design results of the containment response margin, for plants in the design stage analysis methodology for the limiting (i.e. , at the construction permit (CP) or event scenarios are less than the CNV design certification (DC) stage) of review, design pressure and temperature. +Re the containment design pressure should oontainmont design pressure provides provide at least a 10% margin above the approximately 8 poroont margin to tho accepted peak calculated containment limiting LOG.A. peak CMV pressure. Tho I Ar""A ---,t, .... ---,..., ,.-- pressure following a LOCA, or a steam L..- -...J- .LL--.&.--,....,.. © Copyright 2018 by NuScale Power, LLC 10

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn or feedwater line break. Design margins or f"VVLB peak pressures. Additionally, of less than 10% may be sufficient, the overall peakThe overall limiting peak provided appropriate justification is ~ accident pressure resulting from an provided . For plants at the operating inadvertent reactor recirculation valve license (OL) or COL stage of review, the (RRV) opening event is approximately peak calculated containment pressure 986 psia, which is approximately 5§. following a LOCA, or a steam or percent below the GNVcontainment feedwater line break, should be less than design pressure:- of 1075 psia . It results the containment design pressure. from an inadvertent reactor recirculation valve opening anticipated operational occurrence with a loss of normal AC and DC power. Additional margin is provided by the NPM design to satisfy the requirements of GDC 16 and 50 as discussed in Section 5.4. To satisfy the requirements of GDC 38 to The containment response analysis rapidly reduce the containment pressure, methodology is applicable to the initial the containment pressure should be CNV response and demonstrates that reduced to less than 50% of the peak the peak pressure and temperature are calculated pressure for the design basis within the CNV design limits. The LOCA within 24 hours after the methodology also demonstrates that the postulated accident. If analysis shows CNV pressure decreases to less than 50 that the calculated containment pressure percent of the peak pressure within 24 may not be reduced to 50% of the peak hours to satisfy the requirements of calculated pressure within 24 hours, the Principal Design Criterion 38 for rap id organization responsible for DSRS reduction of containment pressure. Section 15.0.3 should be notified. Figure 5-29 demonstrates that the CNV pressure for the limiting case is reduced to less than 50 percent of its peak value in less than two hours. This demonstrates the CNV heat removal capabil ity. DSRS Section 6.2 .1.1.A, p. 5 Containment Response Analysis Methodology To satisfy the requirements of GDC 38 The containment response analysis and 50 with respect to the containment methodology models engineered safety heat removal capability and design features involving the containment heat margin , the LOCA analysis should be removal function and the ECCS . based on the assumption of loss of Conservative assumptions regarding offsite power and the most severe single safety feature performance, in failure in the emergency power system conjunction with conservative initial and (e.g ., a diesel generator failure), the boundary conditions, ensure that the containment heat removal systems (e.g. , CNV peak pressure and temperature a fan , pump , or valve failure), or the core analysis results following a primary cooling systems (e.g., a pump or valve system release are bounding (See failure) . The selection made should result Section 5.4 ). A limiting single failure is in the highest calculated containment considered (See Section 5.1.1). pressure Sensitivity cases considering the availability of power are performed to ensure that assumptions associated with availability of these systems ensure © Copyright 2018 by NuScale Power, LLC 11

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 limiting peak pressure and temperature results (see Section 3.5.2). There are no emergency diesel generators associated with the NPM design . Margin is maintained between the analysis results and the CNV design pressure and temperature limits for the limitinQ cases.

4. To satisfy the requirements of GDC 38 The containment response analysis and 50 with respect to the containment methodology models engineered safety heat removal capability and design features including NPM containment margin , the containment response heat removal and the ECCS with analysis for postulated secondary conservative assumptions that maximize system pipe ruptures should be based containment pressure and temperature on the most severe single failure of the following a secondary system pipe secondary system isolation provisions rupture. For postulated secondary (e .g., main steam isolation valve fa ilure system pipe ruptures , a limiting single or feedwater line isolation valve failure). failure is considered , including main The analysis should also be based on a steam isolation valve or feedwater spectrum of pipe break sizes and reactor isolation valve (FWIV) failure. For the power levels. The accident cond itions NuScale design , full power and the selected should result in the highest maximum break size at each break ca lculated containment pressure or location are the limiting conditions. Initial temperature depending on the purpose and boundary conditions are selected to of the analysis. Acceptable methods for maximize containment pressure and the calculation of the containment temperature response (See Section 3.4 ).

environmental response to main steam Margin is maintained between the line break accidents are found in analysis results and the CNV design NUREG-0588, "Interim Staff Position on pressure and temperature limits. The Environmental Qualification of Safety- longer-term response for equipment Related Electrical Equipment. " qualification is not in the scope of th is reoort. 2.1.3.3 Design Specific Review Standard 6.2.1.3 Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents The DSRS Section 6.2.1 .3, "Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs)" (Ref: 7.1.6), includes the details of an acceptable approach and content for an M&E methodology for LOCAs. As noted , a comparison of NPM design reveals that some of the DSRS content is based on pressurized water reactor (PWR) large-break LOCA phenomena that are not applicable to the NuScale design. The comparison of the M&E methodology to applicable content in DSRS Section 6.2.1.3 is provided in Table 2-3: © Copyright 2018 by NuScale Power, LLC 12

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 2-3 Compliance with Design Specific Review Standard Section 6.2.1.3 DSRS Section 6.2.1.3, p. 3 Containment Response Analysis Methodoloav A. Sources of Energy. The sources of stored and generated The containment response analysis energy that should be considered in methodology includes reactor power; analyses of LOCAs include: reactor decay heat; stored energy in the core; power; decay heat; stored energy in the stored energy in the reactor coolant core ; stored energy in the reactor coolant system (RCS) metal , including the system (RCS) metal, including the reactor vessel and reactor vessel reactor vessel and reactor vessel internals; and stored energy in the internals; metal-water reaction energy; secondary system, including the SG and stored energy in the secondary tubing and secondary water. Metal-water system, including the steam generator reaction energy is not included in the tubing and secondary water. containment response analysis Calculations of the energy available for methodology as discussed in Section release from the above sources should 2.1.1 . be done in general accordance with the requ irements of paragraph I.A. in The containment response analysis Appendix K to 10 CFR Part 50, "Sources methodology models available energy of Heat during the LOCA." However, sources in accordance with the additional conservatism should be requ irements of 10 CFR Part 50, included to maximize the energy release Appendix K, paragraph I.A, with the to the containment during the blowdown exception of 1) metal-water reaction and subsequent phases of a LOCA. An energy is not included , and 2) the ANS-example of this would be accomplished 5.1-1979 decay heat standard with a by maximizing the sensible heat stored in two-sigma uncertainty is used rather the RCS and steam generator metal and than a factor of 1.2 with the 1971 ANS increasing the RCS and steam generator standard . Consistent with DSRS 6.2.1.3, secondary mass to account for Section II , Acceptance Criterion 1.C.v, uncertainties and thermal expansion. the ANS-5.1-1979 standard is equal to the decay heat model given in SRP The requirements of paragraph I.B in Section 9.2.5. Appendix K to 10 CFR Part 50, "Swelling and Rupture of the Cladding and Fuel The containment response analysis Rod Thermal Parameters," concerning methodology model of initial stored the prediction of fuel clad swelling and energy in the fuel is consistent with rupture should not be considered. This Paragraph I.A. 1 of Appendix K to 10 will maximize the energy available for CFR Part 50 . Fuel rods are initialized at release from the core. the maximum initial stored energy condition as determined by the fuel performance analysis. The fuel heat capacity values are conservatively increased to 115 percent of their nominal values to maximize fuel stored energy. The fuel thermal conductivity values are conservatively decreased to 85 percent of their nominal values to maximize fuel stored energy. The conta inment response analysis methodoloov includes conservative © Copyright 2018 by NuScale Power, LLC 13

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 elements that maximize the energy release including sensible heat stored in primary and secondary metal structures, and increasing the RCS mass to account for uncertainties and thermal expansion . The secondary mass is not a significant contributor and a nominal value is used . The containment response analysis methodology does not consider the fuel cladding swelling and rupture prediction requirements of paragraph I.B in Appendix K to 10 CFR Part 50. Calculated cladding temperatures for design basis LOCAs are below the threshold for cladding swelling and rupture . DSRS Section 6.2.1.3, p. 4 Containment Response Analysis Methodoloav B. Break Size and Location The containment response analysis

i. The staffs review of the applicant's methodology includes consideration of a choice of break locations and types is spectrum of break types discussed by discussed in SRP Section 3.6.2. Section 3.2.4 .1. Break locations are ii. Of several breaks postulated , the chosen such that M&E releases to break selected as the reference case containment are maximized .

should yield the highest mass and (( energy release rates, consistent with the criteria for establishing the break location and area . iii. Containment design basis calculations should be performed for a spectrum of possible pipe break sizes and locations to assure that the worst case has been identified.

                                                                        }}2(a ).(c)

C. Calculations In general, calculations of the mass and The containment response analysis energy release rates for a LOCA should methodology focuses on determining the be performed in a manner that maximum post-accident containment conservatively establishes the pressure and temperature. The containment internal design pressure methodology employs conservative (i.e., maximizes the post-accident elements to ensure an overall conta inment pressure response) . The conservative result. criteria qiven below for each phase of the © Copyright 2018 by NuScale Power, LLC 14

Containment Response Analysis Methodology Technical Report TR-0516--49084-NP Draft Rev. 10 accident indicate the conservatism that should exist.

i. Containment Analysis The analytical approach used to compute The M&E release determined by the the mass and energy release profile will containment response analysis be accepted if both the computer methodology is based on the NRELAP5 program and volume noding of the computer code, and the modeling reactor, piping and containment systems approach is similar to the NuScale LOCA are similar to those of an approved evaluation model, Reference 7.2.1 that ECCS analysis. The computer programs complies with the applicable portions of that are currently acceptable include 10 CFR 50 Appendix K. Specific CRAFT-2, and RELAPS, when a flow changes to the LOCA evaluation model multiplier of 1.0 is used with the required to convert it to a conservative applicable choked flow correlation . An methodology to model primary system alternate approach, which is also mass release events are described in acceptable, is to assume a constant Section 3.2.4 .1. The Moody critical flow blowdown profile using the initial model with a discharge coefficient of 1.0 conditions with an acceptable choked is used for saturated two-phase critical flow correlation . flow.

ii. Initial Slowdown Phase Containment Design Basis The initial mass of water in the reactor The containment response analysis coolant system should be based on the methodology assumes an initial power RCS volume calculated for the level of 1.02 times the rated power level. temperature and pressure conditions Initial RCS volume and mass are assuming that the reactor has been consistent with that power level . The operating continuously at a power level initial RCS volume conservatively at least 1.02 times the licensed power includes an allowance for RCS thermal level (to allow for instrumentation error). expansion . An assumed power level lower than the level specified (but not less than the licensed power level) may be used The containment response analysis provided the proposed alternative value methodology uses the conservative has been demonstrated to account for Moody critical flow model for two-phase uncertainties due to power level saturated fluid conditions consistent with instrumentation error. Appendix K. For subcooled fluid conditions the (( Mass release rates should be calculated }}2(a).(c) using a model that has been Reference 7.2.1, Sections 8.2.2 and demonstrated to be conservative by 8.2.3 demonstrates the adequacy of the comparison to experimental data. LOCA evaluation model two-phase and single-phase choked and un-choked flow Calculations of heat transfer from models for predictions of M&E release surfaces exposed to the primary coolant based on assessments of comparisons should be based on nucleate boiling heat of NRELAP5 mass flow predictions to transfer. For surfaces exposed to steam , experimental data. heat transfer calculations should be based on forced convection. The containment response analysis methodology uses the heat transfer Calculations of heat transfer from the correlation package in the NRELAPS secondary coolant to the steam computer code. The LOCA evaluation qenerator tubes should be based on model report demonstrates these © Copyright 2018 by NuScale Power, LLC 15

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 natural convection heat transfer for tube correlations are applicable to the NPM surfaces immersed in water and design (Ref: 7 .2.1 ). The local fluid condensing heat transfer for the tube conditions and the local heat structure surfaces exposed to steam. surface temperatures determine the heat transfer mode. Nucleate boiling and Calculations of heat transfer to the forced convection are included in the containment wall from released reactor code and are selected if the local steam should be such that the heat conditions are appropriate. removal from containment is -- conservatively underestimated so that The containment response analysis the containment pressure is maximized. methodology uses the heat transfer In regions where steam jetting occurs, correlation package in the NRELAPS heat transfer correlations that are based computer code. The LOCA evaluation on jetting of coolant (e.g. based on model report demonstrates these forced convection) may be used as correlations are applicable to the NPM appropriate. Correlations should be design (Ref: 7 .2.1 ). The local fluid appropriately conservative in regions conditions and the local heat structure away from jetting phenomena (e.g. surface temperatures determine the heat based on natural convection, as transfer mode. Forced convection, appropriate). All heat transfer correlations natural convection, condensation, used should be justified. conduction, and nucleate boiling are included in the code and are selected if Calculations of heat transferred from the local conditions are appropriate. condensed reactor water in the Initial and boundary conditions are containment sump into the containment selected to maximize containment wall and from the reactor vessel wall into pressure and temperature response the pooled sump water should be based (See Section 3.5). Steam jetting effects on appropriate heat transfer regimes for are not modeled. the conditions present in containment. Heat transfer through the containment vessel wall into the Reactor Building pool should be demonstrated to conservatively underestimate heat transfer to the pool. DSRS Section 6.2.1.3, p. 5 Containment Response Analysis Methodology iii. Postblowdown Recirculation Phase The containment response analysis (Cold Leg RRV Penetration Breaks Only) methodology uses the NRELAPS code After initial blowdown through a failed that has been determined to be capable RRV, which includes the period from the of modeling all of the phases of the accident initiation (when the reactor is in primary system release events for the a steady-state full power operation NPM design as discussed by Section condition) to the time that the RCS 3.2. NRELAPS predicts the evolution of equalizes to the containment pressure, the primary system release event the water remaining in the reactor vessel scenario, which includes the time of should be assumed to be saturated. pressure equalization and the time at Justification should be provided for the which flow of condensed water through duration of the recirculation period, which the RRVs into the reactor vessel occurs. is the time from the end of the blowdown As discussed in Section 3.1 .3, the to the time when flow from the containment response analysis condensed water in the containment methodology models applicable vessel sump comes back through the phenomena that contribute to RRVs into the reactor vessel. maximizing the M&E release and the © Copyright 2018 by NuScale Power, LLC 16

Containment Response Analysis Methodology Technical Report TR-0516-49084-N P Draft Rev. 1Q resulting containment pressure and temperature. Calculations of the refill rate should be based on the ECCS operating condition following the blowdown phase, where The "refill rate" is only applicable to large energy is released to the RCS primary PWRs. As discussed by the LOCA system by the RCS metal, core decay evaluation model report, the NPM design heat, and the steam generators. The precludes core uncovery (See calculated ECCS conditions should Reference 7.2.1 ). conservatively maximize the containment As discussed by Section 3.2.4.1, the pressure. containment response analysis methodology models applicable phenomena that contribute to Calculations of liquid entrainment, (i.e., maximizing the M&E release and the the carryout rate fraction), which is the resulting containment pressure and mass ratio of liquid exiting the core to the temperature. liquid entering the core, should be based on the NuScale full length emergency cooling heat transfer experiments or The concept of carryout rate fraction that conservatively scaled-up test results from is applicable to large PWRs is not subscale test. applicable to the NuScale design. As discussed by the LOCA evaluation model report, the NPM design precludes The assumption of steam quenching core uncovery, so there is no reflooding should be justified by comparison with phase (See Reference 7.2.1 ). applicable experimental data. Liquid As discussed by Section 3.2.4.1, the entrainment calculations should consider containment response analysis the effect on the carryout rate fraction of methodology models applicable the increased core inlet water phenomena that contribute to temperature caused by steam quenching maximizing the M&E release and the assumed to occur from mixing with the resulting containment pressure and ECCSwater. temperature. Steam leaving the steam generators The concept of steam quenching (that should be assumed to be superheated to occurs from mixing with ECCS water) the temperature of the secondary that is applicable to large PWRs is not coolant. applicable to the NuScale design because ECCS water is not injected into the core. As discussed by Section 3.2.4. 1, the containment response analysis methodology models applicable phenomena that contribute to maximizing the M&E release and the resulting containment pressure and temperature. The superheating effect described is a pressurized water reactor LOCA phenomenon that has minimal © Copyright 2018 by NuScale Power, LLC 17

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1Q. applicability to the NuScale design. For the NPM design, flow of primary steam over the SG tubes results in heat transfer based on the NRELAP5 heat transfer correlation package. This allows for superheating of the steam as determined bv the local conditions. DSRS Section 6.2.1.3, p. 6 Containment Response Analysis Methodoloav iv. Post-Recirculation Phase All remaining stored energy in the The stored energy is distributed as primary and secondary systems should predicted by the NRELAP5 modeling of be removed during the post-recirculation heat transfer to and from the primary phase. and secondary systems. The duration of Steam quenching on the containment the analysis is consistent with the LOCA vessel walls, due to pressure evaluation model and the applicable equalization between the reactor vessel figures-of-merit (See Reference 7 .2.1 ). and the containment vessel, should be The containment response analysis justified by comparison with applicable methodology considers steam

         .experimental data.                           condensation on the CNV walls, as The results of post-recirculation            discussed by Section 3.2.4.1. The analytical models should be compared to      NRELAP5 code and model have been applicable experimental data.                justified by comparison to applicable experimental data.

v. Decay Heat Phase The dissipation of core decay heat The containment response analysis should be considered during this phase methodology models the fission product of the accident. The fission product decay energy using the ANS-5.1-1979 decay energy model is acceptable if it is standard plus two-sigma uncertainty.

equal to or more conservative than the SRP Section 9.2.5 references the same decay energy model given in SRP ANS-5.1-1979 standard. Section 9.2.5. The described steam and water mixing Steam from decay heat boiling in the process does not occur in the NPM core should be assumed to flow to the design. Water flowing through the RRVs containment by the path which produces to the core inlet is below the water the minimum amount of mixing with the mixture level in the downcomer and condensed water flowing from the does not contact the steam produced by containment sump into the reactor vessel decay heat boiling. through the RRVs. 2.1.3.4 Design Specific Review Standard 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary Pipe Ruptures The DSRS Section 6.2.1.4, "Mass and Energy Release Analysis for Postulated Secondary Pipe Ruptures" (Ref: 7.2.2), includes the details of an acceptable approach and content for a M&E methodology for MSLBs and FWLBs. The comparison of the M&E methodology to applicable content in DSRS Section 6.2.1.4 is provided in Table 2-4: © Copyright 2018 by NuScale Power, LLC 18

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 2-4

Compliance with Design Specific Review Standard Section 6.2.1.4 DSRS Section 6.2.1.4, P. 4 Containment Response Analvsis Methodology

1. Sources of Energy.

The sources of energy that should be As discussed in Section 3.3, the containment considered in the analyses of steam and response analysis methodology includes all of the feedwater line break accidents include the sources of energy stored in the fluid and stored energy in the affected helical coil structures that contribute to the secondary line SG's metal, including the vessel tubing, break scenarios. This includes energy stored in feedwater line, and steam line; stored fluid contained in piping systems connected to the energy in the water contained within the break flowpath into the CNV. affected helical coil SG; stored energy in the feedwater transferred to the affected The containment response analysis methodology helical coil SG before closure of the considers a spectrum of pipe break sizes and isolation valves in the feedwater line; various plant conditions. However, the limiting stored energy in the steam from the initial conditions are at 102 percent rated power unaffected helical coil SG before the as the effect of SG liquid mass inventory and closure of the isolation valves in the feedwater flows is greatest at full power. (( helical coil SG crossover lines; and energy transferred from the primary coolant to the water in the affected helical coil SG during blowdown to include energy transferred to the draining DHRS heat exchanger water. The steam line break accident should be analyzed for a spectrum of pipe break sizes and various plant conditions from

                                                                                         }}2(a),(c) hot standby to 102 percent of full power.

The applicant need only analyze the 102-percent power condition if it can demonstrate that the feedwater flows and fluid inventory are greatest at full power.

2. Mass and Energy Release Rate In general, calculations of the mass and The containment response analysis methodology energy release rates during a steam or maximizes the CNV peak pressure and feedwater line break accident should be temperature. The Moody critical flow model with a performed in a conservative manner from discharge coefficient of 1.0 is used for saturated a containment response standpoint (i.e., two-phase fluid conditions. For subcooled and the postaccident containment pressure superheated fluid conditions the ((

and temperature are maximized). The }}2 <aJ.(c) A discharge following criteria indicate the degree of coefficient of 1.0 is used. conservatism that is desired: A. Mass release rates should be calculated using the Moody model (Ref. 6) for saturated conditions or a model that is demonstrated to be equally conservative. B. Calculations of heat transfer to the The containment response analysis methodology water in the affected helical coil SG uses the heat transfer correlation package in the should be based on nucleate boiling heat NRELAPS computer code. The non-LOCA transfer. evaluation model report demonstrates these correlations are applicable to the NPM design (Ref: 7.2.2). The local fluid conditions and the local heat structure surface temperatures © Copyright 2018 by NuScale Power, LLC 19

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1G 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 (See Section 3.5). C. Calculations of mass release should The containment response analysis methodology consider the water in the affected helical includes the water inventory stored in piping coil SG and feedwater line, feedwater systems connected to the break flowpath into the transferred to the affected helical coil SG CNV. The closure of isolation valves, with before the closure of the isolation valves consideration of a single failure, determines which in the feedwater lines and upon flooding sources of water contribute to the M&E release to with the DHRS heat exchanger inventory ensure limiting CNV peak pressure and in the affected loop, and steam in the temperature results. helical coil SG. D. If liquid entrainment is assumed in the The containment response analysis methodology steam line breaks, experimental data uses the two-phase flow and heat transfer models should support the predictions of the liquid in the NRELAPS code. The depressurization of entrainment model. A spectrum of steam the SG secondary will cause flashing in addition line breaks should be analyzed, beginning to the increase in primary-to-secondary heat with the double-ended break (DEB) and transfer. The initial liquid inventory in the SG decreasing in area until no entrainment is secondary will boil and flash, and additional calculated to occur. This will allow inventory will result from continued feedwater flow selection of the maximum release case. and from liquid in connecting pipes. The net effect If no liquid entrainment is assumed, a may include some liquid entrainment in the break spectrum of the steam line breaks should flow that is time dependent. An interfacial drag be analyzed beginning with the DEB and multiplier is available as a junction component decreasing in area until it has been option in NRELAPS to minimize liquid demonstrated that the maximum release entrainment. rate has been considered E. Feedwater flow to the affected helical The containment response analysis methodology coil SG should be calculated considering includes the water inventory stored in in piping the diversion of flow from the other helical systems connected to the break flowpath into the coil SG between the two feedwater pipes CNV. The increase in feedwater flow due to the to the common header with inlets to the depressurization of the helical coil SG is helical coil SG on opposite sides of the considered. The closure of isolation valves with reactor vessel, feedwater flashing, and consideration of a single failure determines which increased feedwater pump flow caused by sources of water contribute to the M&E release. the reduction in helical coil SG pressure. The net feedwater addition is calculated using An acceptable method for computing conservative modeling assumptions. feedwater flow is to assume all feedwater travels to the helical coil SG at the pump run-out rate before isolation. After isolation, the unisolated feedwater mass should be added to the available inventory in the helical coil SG. DSRS Section 6.2.1.4, p. 5 Containment Response Analysis Methodology © Copyright 2018 by NuScale Power, LLC 20

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 iii. Single-Failure Analyses Steam and feedwater line break analyses The containment response analysis methodology should assume a single active failure in considers single failures that affect the isolation of the steam or feedwater line isolation the main steam lines and feedwater lines. Non-provisions to maximize the containment safety valves are credited for isolation as a peak pressure and temperature. For the backup. assumed failure of a safety-related steam or feedwater line isolation valve, operation of nonsafety-related equipment may be relied upon as a backup to the safety-related equipment. © Copyright 2018 by NuScale Power, LLC 21

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 3.0 Analysis 3.1 Modeling Software The containment response analysis methodology uses the NRELAP5 system thermal-hydraulic code , which is a NuScale-modified version of the RELAP5-3D v 4.1 .3 code. NRELAP5 is used for all LOCA and non-LOCA transient and accident analyses, including the response of the CNV. The NRELAP5 simulation model used for the containment response analysis methodology is also similar to the NRELAP-5 simulation models used for the LOCA and non-LOCA methodologies, which are presented in References 7.2.1 and 7.2.2. The phenomena identification and ranking tables (PIRT) developed for the LOCA and non-LOCA methodologies are applicable to the containment response analysis methodology. The qualification of the LOCA and non-LOCA methodologies, in particular the comparisons to separate effects tests and integral effects tests, applicable to the containment response analysis methodology are presented in Section 4.1. The differences in the NRELAP5 simulation models used in the containment response analysis methodology as compared to the LOCA and non-LOCA models, along with the rationale for selection of conservative initial and boundary conditions , are the subject of this report. 3.2 NRELAPS Base Simulation Model Development 3.2.1 RELAPS-30 RELAP5-3D, version 4.1 .3 was used as the baseline development platform for the NRELAP5 code. RELAP5-3Dwas procured by NuScale and subsequently features were added to address unique aspects of the NuScale design and licensing methodology. The following is a brief description of the RELAP5-3D© code. The RELAP5-3D code has been developed for best-estimate transient simu lation of light water RCSs during postulated accidents. The code models the coupled behavior of the RCS and the core for LOCAs and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines , condensers, and secondary feedwater systems. The RELAP5-3D© code is based on a non-homogeneous and non-equilibrium model for the two-phase system that is solved by a fast, partially implicit numerical scheme to permit economical calculation of system transients. The code includes many generic component models from which general systems can be simulated. The component models include pumps, valves , pipes, heat releasing or absorbing structures, reactor kinetics, electric heaters, and control system components. In addition, special process models are included for effects such as form loss, flow at an abrupt area change, branching , choked flow, boron tracking , and noncondensable gas transport. 3.2.2 RELAPS-30 Quality Assurance NuScale Power procured RELAP5-3D©v.4 .1.3 from the Idaho National Laboratory through a commercial-grade dedication process that complies with NQA-1-2008 and NQA-1 a- © Copyright 2018 by NuScale Power, LLC 22

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 2009 requirements . The commercial-grade dedication evaluation determined that verification of certain of the critical characteristics required testing. Eleven test cases were identified for verification , figures of merit, and acceptance criteria. Included were models of the NPM along with Nu Scale proprietary test programs, legacy tests , and special feature tests. These cases constitute the matrix for commercial-grade dedication acceptance testing as discussed by the LOCA evaluation model report (Reference 7.2.1), Section 6.1.2. RELAP5-3D© v.4.1 .3 was then placed under the NuScale quality assurance program as NRELAP5 Version 0.0. Subsequent NRELAP5 versions were developed and placed under the NuScale Quality Assurance Program including the technical code revisions listed in Table 3-3 along with code corrections and administrative code revisions. 3.2.2.1 NRELAPS NRELAP5 is NuScale's proprietary system thermal-hydraulic computer code for use in engineering design and analysis. NRELAP5 was developed at NuScale, using RELAP5-3D©v.4.1.3 as the initial baseline. Chapter 6 of the LOCA Evaluation Model (Ref: 7.2.1) is a summary of the RELAP5-3D© code and the revisions incorporated by NuScale to produce the NRELAP5 code used in both the LOCA Evaluation Model and the Non-LOCA Evaluation Model (Ref: 7.2.2). The new models in NRELAP5 are listed in Table 3-3 along with the application in the containment response analysis methodology. Table 3-1 New NRELAP5 models New Model Application in Containment Response Analysis Methodology Condensation heat transfer Used for condensation heat transfer on

              * ((                                       the CNV inside diameter and inside the decay heat removal system (DHRS) heat exchanger tubes
                                       }}2(a ).(c)

Critical flow Used for two-phase saturated critical

Moody critical flow model for two- flow phase flow conditions Helical coil SG component Used for modeling the helical coil SGs
Heat transfer correlation
Friction correlation Pool heat transfer Churchill-Chu is used for modeling the
Churchill-Chu natural convection CNV outside diameter (OD), the reactor correlation correction to use bulk pressure vessel (RPV) outside fluid properties diameter, and outside the DHRS heat
Cooper pool boiling correlation exchanger tubes (vertical surfaces
Rohsenow pool boiling only).

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 New Model Application in Containment Response Analysis Methodology lnterfacial drag multiplier Used in containment response analysis

Input multiplier added to allow methodology to evaluate effect of liquid minimizing liquid entrainment in entrainment on break and valve flow break and valve flow Void drift velocity Used for two-phase flow
Kataoka-Ishii alternative formulation set to default Critical Heat Flux The 2006 Groenveld tables are used in
Reyes correlation the containment response analysis
Electric Power Research Institute methodology. CHF does not occur for all correlation with counter current LOCA and non-LOCA scenarios in the flow limitation or Groenveld as containment response analysis interpolation point for zero flow methodology.
Chang correlation
2006 Groenveld tables
Extended Hench-Levy correlation Dynamic gap conductance Not used in the containment response
Dynamic gap conductance model analysis methodology with optional pellet axis offset capability Boric acid solubiity Not used in the containment response
Compare boric acid analysis methodology concentration to solubility limit Decay heat Not used in the containment response
1971 ANS Standard including analysis methodology actinides 3.2.3 NRELAP5 Simulation Models This section presents the NRELAPS simulation models of the NPM that are used for the containment response analysis methodology. The NRELAPS models developed for the LOCA and non-LOCA evaluation models are used to develop the primary system (LOCA and valve opening events) and secondary system (MSLB and FWLB events) M&E release and containment response models, respectively. Substantive changes to the NRELAPS LOCA model are limited to those necessary for containment response analysis applications. Other changes are made to the LOCA model that do not significantly impact primary system release event containment analysis results (see Section 3.2.4 .1 ).

Changes to the NRELAPS non-LOCA model are limited to those necessary for containment response analysis applications. 3.2.3.1 NRELAP5 LOCA Evaluation Model The NRELAPS loss-of-coolant accident model input file is developed from engineering drawings, calculations, and reference documents. These sources of information provide the numerical information necessary to develop a complete thermal-hydraul ic simulation model of the NuScale SMR per the input fi le specification. The types of required information fall into the following NRELAPS input categories: © Copyright 2018 by NuScale Power, LLC 24

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

Thermal-hydraulic fluid volumes and connecting heat structures

             -   reactor vessel primary loop lower plenum core riser pressurizer SG primary side down comer
             -   reactor kinetics
             -   reactor vessel secondary system SG secondary main steam piping feedwater piping
             -   CNV
             -   reactor pool
             -   DHRS
             -   ECCS
             -   Chemical and volume comtrol system (CVCS) piping for RCS injection , discharge and pressurizer spray supply

Material properties
Control systems

             -   normal control systems pressurizer pressure pressurizer level Tavg steam pressure turbine load
            -    reactor protection system
            -    engineered safety feature controls The NRELAP5 NuScale Power Module model from which LOCA runs are initiated is described in the LOCA Evaluation Model in detail (Reference 7.2.1, Section 5.3) and is summarized in this report. The objectives of the NRELAP5 loss-of-coolant accident model are to analyze the LOCA break spectrum for the NPM and to demonstrate compliance with 10 CFR 50 Appendix K.

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Figure 3-1 is a simplified diagram of the nodalization selected to enable modeling of the phenomena that were determined to be important for the spectrum of LOCA scenarios. The LOCA primary system release scenarios start with the blowdown of the primary inventory through the pipe break into the CNV. The reactor trips on high CNV pressure, which causes a turbine trip along with main steam isolation and feedwater isolation . The primary system depressurizes as the CNV pressurizes, and the coolant inventory accumulates in the CNV. Steam released into the CNV condenses on the CNV inner surface that is cooled by conduction and convection to the reactor pool. When the primary system inventory reaches the low level setpoint, or the CNV level reaches the high level setpoint, and the pressure drop across the ECCS valves is less than the inadvertent actuation block (IAB) release pressure, the ECCS valves open. Opening of the reactor vent valves (RVVs) increases the primary depressurization rate and completes equalization of primary and secondary pressures. Opening of the RRVs establishes a flowpath for the inventory in the CNV to flow by gravity into the RPV for core cooling . The flowpaths through the break plus the RVV, and the flowpath through the RRV provide abundant core cooling that is sufficient to keep the core covered by a two-phase mixture that prevents any heatup of the fuel rod cladding . The NRELAP5 loss-of-coolant accident model includes the following additions to obtain a conservative LOCA analysis that meets the Appendix K requ irements:

conservative initial conditions at 102 percent of rated power level
with or without loss of normal alternating current (AC) power
high core power peaking factors
break junction modeling for the various break locations
Moody critical flow option
ANS 1973 decay heat standard with 1.2 factor and actinides
limiting single fa ilure assumption
ECCS actuation with conservative performance
conservative CNV modeling
conservative reactor pool modeling
conservative setpoints and actuation delays The LOCA evaluation model nodalization and each of these conservative LOCA modeling elements are evaluated in Section 3.2.4.1 for use in the primary system release event containment response analysis methodology. The adequacy of the NRELAP5 code and the LOCA model for modeling the primary system M&E scenarios is addressed in Sections 4.1 and 4.2.

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

                                                                                            }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 3.2.3.2 NRELAPS Non-Loss-of-Coolant Accident Evaluation Models The NRELAP5 non-LOCA models are summarized in this section . The objectives of the NRELAP5 non-LOCA models are to analyze the spectrum of non-LOCA transients and accidents for the NuScale SMR, and to demonstrate compliance with the regulatory acceptance criteria . 3.2.3.2.1 Inadvertent Operation of Emergency Core Cooling System The inadvertent operation of ECCS events include the inadvertent opening of an RVV or an RRV. Both events involve an initial primary system M&E release through the inadvertently opened valve into the CNV, and a subsequent actuation of the remaining ECCS valves that results in a second M&E release into the CNV. The FSAR Section 15.6.6 describes the methodology for analyzing these events and is the starting point for developing the valve opening event models in the primary system containment response analysis methodology. 3.2.3.2.2 Secondary System Pipe Breaks The NRELAP5 non-LOCA model is the starting point for developing the MSLB and FWLB models in the containment response analysis methodology. Figure 3-2 shows the non-LOCA NRELAP5 nodalization diagram. © Copyright 2018 by NuScale Power, LLC 28

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

                                                                                                 }}2(a ),(c)

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 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 break 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 pressure low 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 reponse analysis methodology that focus on a conservative analysis of the CNV peak pressure and temperature response are detailed in Section 3.2.4.2.

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)

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 flash ing
DHRS actuation
with or without loss of normal AC and direct current (DC) electrical power
limiting single fa ilure The differences in the NRELAP5 feedwater line break modeling for the containment reponse 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 eva luation 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 NRELAPS 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 RWs 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 : © Copyright 2018 by NuScale Power, LLC 31

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

                                              /    Main steam isolation valve.s (MSIVs)
                                             ~  !/        Ma in feed isolation valves (MF IVs)

Reactor buildine pool Containment vessel Pressurizer Steam header Upper plenum DHRHX Hot leg riser Reactor recirculation vatves (RRVs) Core lower plenum Figure 3-3 NuScale module during power operation Figure 3-4 NuScale module during emergency core cooling system operation © Copyright 2018 by NuScale Power, LLC 32

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 LOCA Pipe Break and Valve Opening Modeling ((

                                                                                  }}2(a),(c)

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

                                                                               }}2(a),(c)

Figure 3-5 NRELAP5 nodalization for reactor coolant system discharge line break loss-of-coolant accident

 © Copyright 2018 by NuScale Power, LLC 34

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

                                                                                                 }}2(a),(c)

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

                                                                                                    }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 lI Conservative modeling of the LOCA pipe break spectrum and the valve opening events to ensure a bounding M&E release includes the following elements:

all break locations are considered
maximum credible break size at each location
critical flow with discharge coefficient of 1.0
saturated liquid - Moody critical flow
subcooled liquid - (( }}2(a),(c)
modified pressure volume work term
maximum RRV and RVV flow areas
liquid entrainment evaluated by use of interfacial drag multiplier in upper riser, riser upper plenum, pressurizer baffle, pressurizer, and downcomer Containment Vessel and Reactor Pool Models The CNV nodalization in the NRELAP5 loss-of-coolant accident and valve opening event containment response analysis model (Figure 3-6, Component 500) is modified compared tGconsistent with the LOCA evaluation model. The following substantive changes were made to the LOGA evaluation model to maximize M&E release and consequential containment pressure and temperature.

In addition to the five substantive modifications described above , the following non subtantive modifications are to the LOGA evaluation model:

       .     ((

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 A sensitivity analysis demonstrates that the above modifications to the LOCA evaluation model have a negligible effect on CNV analysis results . The CNV is maintained at a partial vacuum with an assumed high initial pressure (e.g.

       ~~ -0 psia), SGand with the maximum total mass of noncondensables that could exist within the CNV during operation (approximately 131 lbm), in order to capture the effects of CNV non-condensable gases. Also, during a LOCA or valve opening event, the maximum tota l mass of noncondensables that could exist within the RPV during operation are released to the CNV model, in order to capture the effects of RPV non-condensable gases are small but are modeled . Heat transfer to the Cfi.N vessel inside diameter (ID) is initially by radiative heating from the RPV, and that modeling sets the initial CNV wall temperature distribution . The LOCA or valve opening event M&E release into the CNV results in a rapid heating and pressurization of the CNV. The steam is condensed on the CNV inside diameter and the condensate film flows downward and forms a pool in the bottom of the CNV. As the CNV pool level rises boiling occurs on the RPV surface.

Heat transfer from the CNV outside diameter to the reactor pool initially maintains the vessel at a low temperature except for the upper section of the vessel that is above the pool surface elevation. Following the LOCA or valve opening event, the condensing of steam and convection from the CNV pool increases the vessel temperature , and heat transfer from the CNV outside diameter to the reactor pool increases. Heat transfer on the CNV outside diameter is by pool convection and pool nucleate boiling , except for the upper section that is not submerged in the reactor pool. The initial CNV wall temperatures above the pool level are maintained at 240 degrees F, which bounds the maximum CNV wall temperature that can exist during norma l operation . In the upper section heat transfer is by convection neglected due to amhe presence of insulation on the CNV upper head (See Figure 3-8). © Copyright 2018 by NuScale Power, LLC 38

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

                                                                                             }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-05 16-4 9084-N P Draft Rev. 10 Conservative modeling of the heat transfer to and from the CNV inside diameter, and from the CNV outside diameter to the reactor pool , to ensure a bounding peak CNV pressure and temperature response following a LOCA or valve opening event, includes the following elements:

        .    ((
        .   .[_
                                                                    }}2(a ),(c )

Table 3-2 shows the heat transfer correlations and models for all of the processes that could impact the CNV peak pressure and temperature response . These correlations and models, along with their applications , are described in greater detail in the LOCA Evaluation Model Report (Reference 7.2 .1). Table 3-2 Containment vessel and reactor pool heat transfer modeling Heat Transfer Process Correlation/Model Radiant heating from RPV outside Radiation enclosure model considered in analysis. diameter to CNV inside diameter This was not included in the model since inclusion of a radiation enclosure model has a negligible im12act on CNV 12eak 12ressure and tem12erature results . Convection from RPV outside diameter Vertical Sufaces to CNV pool ((

                                                                                           }}2(a },(c)

Non-Vertical Surfaces ((

                                                                                           }}2(a },(c}

(( }}2{a },{c} Condensation on CNV inside diameter lnterphase heat transfer Default model based on flow regimes Convection from CNV outside diameter Vertical Sufaces to reactor pool ((

                                                                                           }}2{a },{c}

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Heat Transfer Process Correlation/Model (( }}2(a),(C) Reactor pool mixing No mixing is modeled Reactor pool cooling to ambient Assumed adiabatic Reactor pool mixing with other modules No mixing with other modules is modeled 3.2.4.2 NRELAP 5 Secondary System Break Analysis Model Overview The NRELAP5 model used for secondary system pipe break ana lysis in the containment response analysis methodology is similar to the NRELAP5 model used in the non-LOCA accident FSAR Chapter 15 methodology (Section 3.2.3.2). The 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. The following changes 'Nere made to the non LOCA evaluation model to maximize M&E release and consequential containment pressure and temperature . © Copyright 2018 by NuScale Power, LLC 41

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 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 2018 by NuScale Power, LLC 42

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

                                                                                              }}2(a),(c)

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

                                                                                           }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Conservative modeling of the secondary pipe breaks to ensure a bounding M&E release includes the following elements:

        .    ((
                            }}2(a),(c)

Containment Vessel and Reactor Pool Models The CNV and reactor pool models for the MSLB and FWLB containment response analysis methodology are the same as the modeling for LOCA. Refer to Section 3.2.4.1. 3.3 Containment Response Analysis Methodology for Primary System Release Events Section 3.3 presents the details of the containment response analysis methodology for primary system releases resulting from primary system breaks and valve opening events. The NRELAP5 computer code described in Section 3.2.2.1 and the LOCA containment response analysis model described in Section 3.2.4.1 are applied using the methodology in this section to meet the NRC regulations and regulatory guidance in Section 2.0. 3.3.1 Primary System Mass and Energy Release Methodology 3.3.1.1 Loss-of-Coolant Accident Scenario Phenomena Identification and Ranking Table Results NuScale has performed and documented a PIRT for the LOCA scenarios resulting from primary system breaks and ECCS valve opening events. Loss-of-Coolant Accident Evaluation Model Report (Reference 7 .2.1 }, Chapter 4.0, summarizes the LOCA phenomena identification and ranking table. The results of the LOCA phenomena identification and ranking table were used in the development of the NRELAP5 code, the NRELAP5 LOCA model, and the LOCA evaluation model. The results of the LOCA scenario PIRT are directly applicable to the primary system M&E release and resultant CNV pressure and temperature response that are the focus of the 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 © Copyright 2018 by NuScale Power, LLC 45

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

loi.v RPV level
high CNV level
loss of normal AC power and the highly reliable DC power system The following design criteria govern RWs 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 RWs 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 through the RRVs into the reactor vessel starts to provide long-term core cooling via recirculation . This terminates the reactor vessel level decrease prior to core uncovery.

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Heat transfer to the CNV wall and to the reactor pool eventually exceeds the energy addition from the break flow and the RVV flow. When this occurs the period of peak containment pressure and temperature have been completed , and a gradual depressurization and cooling phase begins. 3.3.1.3 Event Scenarios and Break Spectrum The postulated primary system M&E release events include the following pipe break accidents and valve actuations. For the valve opening events, the specific FSAR events that result in actuation of that valve are listed.

Pipe breaks (LOCAs)

             -   FSAR 15.6.5 - RCS discharge line break LOCA ((                                       }}2(a),(c)
             -   FSAR 15.6.5 - RCS injection line break LOCA ((
                                                 }}2(a),(c)
             -   FSAR 15.6.5 - Pressurizer spray supply line break LOCA ((                        }}2(a ),(c)
             -   FSAR 15.6.5 - RPV high point degasification line LOCA ((                        }}2(a ),(c)

RSV actuation (( }}2(a),(c)

             -   FSAR 15.6.1 - Inadvertent RSV opening

RVV actuation ((( }}2(a ),(c) )

             -   FSAR 15.6.6 - Inadvertent RW open ing

RRV actuation (( }}2(a),(c)

             -   FSAR 15.6.6 - Inadvertent RRV opening The RPV high point degasification line, the pressurizer spray supply line, and the RSVs are all located near the top of the RPV. A LOCA in the RPV high point degasification line is the largest break size in this location and is analyzed in the containment response analysis methodology. The other two are non-limiting and are not analyzed.

One RW or one RRV can open as an initiating event due to an assumed mechanica l failure. The RVVs and RRVs all open following ECCS signal actuation and when the IAB design criteria discussed in Section 3.3.1.2 are met . The RPV high point degasification line break LOCA differs in that the break flow will be steam. The RCS break locations differ in that the discharge line connects to the downcomer, and the injection line connects to the riser. These three break locations plus the valve opening event locations fulfill the adequacy of the break spectrum with regard to location. The adequacy of the break spectrum with regard to break size is important in the timing of the ECCS valve opening , as the second M&E release resulting from the opening of the three RVVs is the dominant event for CNV pressure and temperature response . First, the maximum break size at each location is analyzed to ensure the maximum initial M&E release rate into the CNV during the first phase of CNV pressurization . Then , the sensitivity © Copyright 2018 by NuScale Power, LLC 47

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn of the opening time of the three RWs is addressed by analysis of a range of IAB release pressures for each break location. In this manner a lower IAB release pressure results in a delay in the RVV opening time . This is similar to a break size sensitivity because a range of break sizes would result in a range of depressurization rates and RVV opening times. However, by using th e maximum break size for all cases the maximum initial M&E release rate is used for all cases. This approach fulfills the adequacy of the break spectrum with regard to break size. ((

                                                                            }}2(a),(c)

In summary, the limiting postulated primary system M&E release scenarios consist of an initiating anticipated operational occurrence or accident, which may include a pipe break or RW or RRV valve opening, with a resultant an ECCS actuation signal causing all RWs and RRVs to MJy_open after the IAB design criteria discussed in Section 3.3.1.2 are met. Table 3-3 shows the primary system M&E release scenarios that are used to determine the limiting cases. Table 3-3 Primary system mass and energy release scenarios lniating Event Subsequent RW and RRV Analysis Actuations on ECCS Case LOCAin RCS Three RVVs and two RRVs actuate 1 discharge line from down comer LOCAin RCS Three RWs and two RRVs actuate 2 injection line from riser LOCA in RPV High Three RWs and two RRVs actuate 3 Point Degasification Line near top of vessel © Copyright 2018 by NuScale Power, LLC 48

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 lniating Event Subsequent RW and RRV Analysis Actuations on ECCS Case RVV opens due to a Two RWs and two RRVs actuate 4 mechanical failure RRV opens due to a Three RWs and one RRV actuate 5 mechanical fai lure 3.3.1.4 Identification of Bounding Events The bounding events for peak CNV pressure and for peak CNV temperature are identified by analyzing the spectrum of scenarios in Table 3-3 with conservative initial conditions and boundary conditions. Sensitivity stud ies are used to determine the bounding conditions and assumptions for the limiting cases. This is further discussed in Section 5.1.1. 3.4 Secondary System Containment Response Analysis Methodology Section 3.4 presents the details of the containment response analysis methodology for the secondary system pipe break accidents. The NRELAP5 computer code described in Section 3.2.2.1 and the secondary system containment response analysis model described in Section 3.2.4.2 are applied using the methodology in this section to meet the NRC regulations and regulatory guidance discussed in Section 2.0. The methodology for the main MSLB and the FWLB accident analyses is presented. 3.4.1 Steamline Break Mass and Energy Release Methodology 3.4.1.1 Non-Loss-of-Coolant Accident Event Phenomena Identification and Ranking Table Results NuScale has performed and documented a PIRT for the non-LOCA events. The results of the non-LOCA phenomena identification and ranking table are summarized in the non-LOCA evaluation model (Ref: 7.2.2). The results of the non-LOCA phenomena identification and ranking table are directly applicable to the secondary system M&E release and CNV pressure and temperature response that are the focus of the containment response analysis methodology. The basis for th is statement is that ((

                         }} 2 (a),(c) Therefore the non-LOCA phenomena identification and ranking table is also considered to be the secondary system conta inment response analysis PIRT.

3.4.1.2 Module Response The NPM initially responds to a MSLB inside the CNV with a simultaneous depressurization of the secondary system and a pressurization of the CNV. Feedwater flow out the break increases due to the decrease in backpressure and due to flash ing of the feedwater pipe inventory. The resulting high containment pressure signal causes an immediate actuation of the following safety features:

containment isolation including

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 adversechanqed by assuming a loss of normal AC and DC power (concurrent with the break), which results in an ECCS signal. 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 limiting events depending on the evolution of the scenario with continued normal AC power, or following a loss of normal AC and DC power. © Copyright 2018 by NuScale Power, LLC 50

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 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 term inated. 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 RWs 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 po,,;er results in the peak C~JV pressure and peak C~JV 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 Closure of the MS IVs 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

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 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 re lease 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 li nes (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 ft 2 , corresponding to one 5" Schedule 120 line between the FWIV and the FW tee . However, since these two geometries are located at the same reg ion (i.e. the FW tee), assuming a larger FW break size {0.1433 ft 2 ) is acceptable since it conservatively maximizes mass release to CNV. The limiting FWLB event is a double-ended rupture of the largest feedwater pipe ~ Schedule 120 I 4.563 in . ID), 1Nhich iswith a break area of 0.1433~ ft 2 . 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 fa ilure of the FWPJMSIV to close on the affected SG is mitigated by closure of the F'.0/RVbackup MSIV. The li miting 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 of power to the pressurizer heaters. W ith DHRS actuation 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. © Copyright 2018 by NuScale Power, LLC 52

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 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 EGGS actuation results in the peak C~JV pressure and peak C~JV 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 particu lar ((

                                            }}2(a ),(cl The SG initial conditions result from the NRELAP5

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn initialization process and are consistent with the conservative primary system initial conditions. The initial conditions for the CNV and the reactor pool are shown in Table 3-5. These initial conditions ensure that the CNV heat sink is minimized so that the peak containment pressure and temperature are modeled conservatively. Table 3-5 Containment vessel and reactor pool initial conditions Parameter Initial Condition Assumption Rationale ((

                                                                                          } }2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Parameter Initial Condition Assumption Rationale ((

                                                                             }}2(a),(c) 3.5.2    Primary System Release Event Boundary Conditions Boundary conditions for the spectrum of primary system M&E release analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the boundary cond itions 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. Due to the simplicity of the NPM design there are few postulated single failures for the primary system M&E release scenarios. Failure of ECCS valves to open would obviously reduce the M&E releaseare analyzed as sensitivity studies, and are not limiting. Failuresfailure of MSIVs or FWIVs to close are analyzed as sensitivity studies. considered, but they have minimal effect on the CNV pressure and temperature response as the secondary system response is not importantis immediately isolated for the primary side events . Table 3-6 presents the boundary conditions for the LOCA containment response analyses.

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 3-6 Primary system boundary conditions Parameter Boundary Condition Rationale Assumption (( i

                                                                                            }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Parameter Boundary Condition Rationale Assumption ((

                                                                   }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Parameter Boundary Condition Rationale Assumption ((

                                                                            }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 3.5.3 Main Steam Line Break Initial Conditions Initial conditions for the MSLB 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 6.2.1.4. The selection process ensures that energy sources are maximized and energy sinks are minimized. Table 3-4 presents the primary system initial conditions used for primary system release containment response analyses. ((

                                  }}2 (a),(c) Table 3-5 presents the CNV and reactor pool initial conditions used by the LOCA containment response analyses that are also used by the MSLB containment response analyses. Table 3-7 presents the secondary system initial conditions used by the MSLB containment response analyses.

Containment Response Analysis Methodology Technical Report TR-0516-49084-N P Draft Rev. 10 Table 3-7 Secondary system .initial conditions Parameter Initial Condition Rationale Assumption ((

                                                                                              }}2(a),(c) 3.5.4    Main Steam Line Break Boundary Conditions Boundary conditions for the MSLB mass and energy release analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the boundary conditions is consistent with the guidance in DSRS Section 6.2, and specifically DSRS Section 6.2.1.4. The selection process ensures that energy sources are maximized and energy sinks are minimized.

The largest break size is assumed to maximize the secondary system M&E release rate into the CNV and thereby maximize the resulting CNV pressurization and temperature increase. However, a subsequent primary system M&E release following ECCS actuation and delayed opening of the three RWs may result in the peak CNV pressure and temperature response for some scenarios. Also, opening of the RWs depends on the IAB © Copyright 2018 by NuScale Power, LLC 60

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 design criteria in Section 3.3.1 .2 being satisfied , and that may not occur until the DHRS has been operating for some period of time. As the DHRS cools the primary system, a delayed M&E release through the RWs will be smaller, and the second CNV pressurization wi ll be lower. Furthermore, the steam line break CNV pressure and temperature response remains bounded by the LOCA. Therefore, the maximum MSLB size is bounding and a break spectrum analysis is not necessary. Due to the simplicity of the NPM design , there are few postulated single failures for the secondary system M&E release scenarios. Failure of ECCS valves to open WGl:UG obviously reduceis considered for the M&E release and are not analyzedscenarios in which ECCS actuation occurs. Failures of MSIVs or FWIVs to close are analyzed as sensitivity studies. Table 3-6 presented the boundary conditions for the primary system containment response analysis methodology, and they are the same for the MSLB containment response analysis methodology except for those presented in Table 3-8. Table 3-8 Boundary conditions for the main steam line break containment response analysis methodology Parameter Boundary Condition Rationale Assumption ((

                                                                                              }}2(a ),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Parameter Boundary Condition Rationale Assumption ((

                                                                                          }}2(a),(c)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 3.5.5 Feedwater Line Break Initial Conditions Initial conditions for the FWLB mass and energy release analyses are selected to ensure a conservative CNV peak pressure and peak temperature resu lt. The process of selecting the initial conditions is consistent with the guidance in DSRS Section 6.2, and DSRS Section 6.2.1.4 specifical ly. The selection process ensures that energy sources are maximized and energy sinks are minimized. Table 3-4 presents the primary system initial conditions used by the LOCA containment response analyses. ((

                    }}2<a>.<c> Table 3-5 presents the CNV and reactor pool initial conditions used by the LOCA containment response analyses, and these initial conditions are also used by the FWLB containment response analyses. Table 3-7 presents the secondary system initial conditions used by the MSLB containment response analyses, and these initial conditions are also used by the FWLB containment response analyses.

3.5.6 Feedwater Line Break Boundary Conditions Boundary conditions for the FWLB mass and energy release analyses are selected to ensure a conservative CNV peak pressure and peak temperature result. The process of selecting the boundary conditions is consistent with the guidance in DSRS Section 6.2, and specifically DSRS Section 6.2.1.4. The selection process ensures that energy sources are maximized and energy sinks are minimized. Section 3.4.4 and Table 3-8 presented the boundary conditions used by the MSLB containment response analyses, these boundary conditions are also used by the FWLB containment response analyses, with the exception of the single failure eva luation that is discussed below. The largest break size is assumed to maximize the initial M&E release into the CNV. However, it is the subsequent second M&E release following ECCS actuation and open ing of the three RWs that results in the peak CNV pressure and temperature response . Also, opening of the RWs depends on the pressure differential decreasing to below the IAB release pressure, and that may not occur until DHRS has been operating for some period of time . Therefore , the initial break size is unimportant as the secondary M&E release is similar, and the sequence of events leading to the opening of the RWs is similar. Furthermore, the feedwater line break CNV pressure and temperature response is bounded by the LOCA. Therefore, a break spectrum analysis is not necessary. Due to the simplicity of the NPM design , there are few postu lated single failures for the secondary system M&E release scenarios. Failure of ECCS valves to open WGYW obviously reduceis considered for the M&E release and are not analyzed. scenarios in which ECCS actuation occurs. Failures of a MSIV or a FWIV to close are analyzed as sensitivity studies, with the FVVIV failure identified as limiting . © Copyright 2018 by NuScale Power, LLC 63

I Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 4.0 Qualification and Assessment 4.1 Assessment of Methodology and Data 4.1.1 Primary System Release Effects Code and Model Qualification The NRELAPS code has been qualified or assessed to the separate effects and integral effects tests as described by LOCA Evaluation Model Report (Reference 7 .2.1 ), Chapter 7.0 to demonstrate the capability to simulate LOCAs in the NPM . The results of the NRELAPS comparisons to data establish the capability of the code to model the NPM design for the LOCA analysis. The most important assessment activities were those comparing to integral LOCA tests conducted in the NIST-1 facility. The following two key known scaling distortions are relevant to the scope of the containment response analysis methodology:

          .    ((
                              }}2(a),(c)

Neither of the above phenomena have an impact on the peak CNV pressure. The first distortion is addressed by the containment response analysis methodology by closure of the MSIVs. The second distortion is addressed by the overall conservative modeling of CNV heat transfer in the containment response analysis methodology, which includes use of conservative initial conditions and boundary conditions that are discussed in Section 3.4 . The LOCA Evaluation Model Report (Reference 7 .2.1, Section 8.2) also presents the evaluation of the adequacy of the NRELAP5 code and LOCA Evaluation Model for modeling LOCAs in the NPM. The following action was identified as needed to address adequacy issues relative to the containment response analysis methodology: ((

                  }}2(a ),(c)

No additional qualification activities were performed for the LOCA containment response analysis methodology as the LOCA evaluation model qualification activities addressed in LOCA Evaluation Model Report (Ref: 7.2.1) are adequate.

 © Copyright 2018 by NuScale Power, LLC 64

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 4.1.2 Secondary System Pipe Break Effects Code and Model Qualification The LOCA-event code and model qualification as described in Section 4.1.1, which credits the LOCA Evaluation Model qualification activities, is generally applicable to the secondary system M&E release events. Additional NRELAP5 code and model qualification activities were included in the Non-LOCA Evaluation Model (Reference 7.2.2) with the focus being DHRS and SG heat transfer as they are of greater importance during non-LOCA events. The following NIST-1 facility and testing distortions are applicable to secondary M&E release containment analysis methodology:

        .    ((
                                                                                                  }}2(a),(c)

The secondary system M&E releases consist of the MSLB and the FWLB events, and both of these events involve asymmetric responses in the two SGs. Following break initiation the affected SG blows down into the CNV until the feedwater supply has been isolated. The unaffected SG is isolated from the affected SG following closure of the MS IVs, and then provides decay heat removal via the DHRS. A second M&E release for these events occurs for cases that include ECCS actuation on loss of normal AC and DC power coincident with the pipe break. The primary pressure gradually decreases during the DHRS cooldown phase, and the ECCS valves open when the differential pressure decreases to below the IAB release pressure. The NRELAP5 code and the containment response analysis model for the NPM are fully capable of modeling the secondary system M&E releases without directly applicable NIST-1 test data. The large body of NIST-1 separate effects and LOCA integral tests have demonstrated the capability of NRELAP5 to adequately model the NPM design. There are no additional phenomena associated with secondary M&E releases, and no additional qualification activities were performed for the secondary containment response analysis methodology. Additional justification for the above position is that the secondary system M&E release analyses for the NPM demonstrate that they are non-limiting compared to the primary system containment response analyses. This justification is further supported by the overall conservatism in the containment response analysis methodology. © Copyright 2018 by NuScale Power, LLC 65

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 4.2 Testing Results 4.2.1 NuScale Integral System Test Facility Testing A scaled facility of the NPM was constructed at Oregon State University, referred to as the NuScale Integral System Test Facility-1, or NIST-1, facility, to assist in validation of the NRELAP5 system thermal-hydraulic code. The facility is designed to perform various tests, including LOCA tests. A detailed description of NIST-1, the NRELAP5 model of the facility, and the NRELAPS validation testing is provided in Reference 7.2.1, Section 7.5. The NRELAP5 predictions of CNV pressure, level and temperature documented in Reference 7.2.1 show good fidelity to NIST-1 experimental measurements as follows. The CNV level and pressure response is predicted with reasonable to excellent agreement to RCS discharge line break experimental measurements as discussed by Reference 7.2.1, Section 7.5.6. ((

                 }}2(a),(c)

The CNV pressure response is predicted with reasonable to excellent agreement to spurious RW opening experimental data as discussed by Reference 7.2.1, Section 7.5.8. A separate high pressure condensation test described by Reference 7.2.1, Section 7.5.4 demonstrates that NRELAP5 has the capability to predict condensation rates for various pressures with reasonable to excellent agreement to experimental data. © Copyright 2018 by NuScale Power, LLC 66

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 5.0 Results 5.1 Primary System Release Scenario Containment Response Analysis This section presents the results of the NRELAP5 limiting analyses of the spectrum of primary system M&E release scenarios for the NPM, listed in Table 3-3, and secondary system break scenarios that are determined using the containment response analysis methodology presented earlier in this report. The case labels from Table 3-3 are used in the following discussion. 5.1.1 Analysis Approach The approach to determine the limiting peak CNV pressure event from the the spectrum of primary mass and energy release scenarios for the NPM, listed in Table 3-3, and the limiting peak CNV temperature for each primary release event was as follows: ((

                    }}2(a),(c)

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

                 }}2(a ),(c) 5.1.2    Reference Analysis and Sensitivity Results The following insights were obtained from the results of the NRELAP5 analyses of the five primary system M&E release cases and associated sensitivity studies.

The peak CNV pressure scenario is the RRV release (Case 5). The RRV mass and energy re lease causes an initial heatup and pressurization of the CNV, and then ECCS actuation resu lts in a second M&E re lease with all three RWs and second RRV open ing that pressurizes the CNV to the highest peak pressure.
The peak CNV wall temperature scenario is the CVCS injection line LOCA (Case 2) .

The break in this location combines a high temperature liquid initial M&E release followed by a high temperature M&E release through all three RVVs following an ECCS actuation signal.

The sensitivity parameters have only a small effect on the peak CNV pressure ~

psi / ""*1.9 percent) and temperature ("'+5 degrees F / °'*1 percent) results of the limiting cases. No single fai lures had a significant impact on the results for the limiting cases. The loss of power sensitivity that results in early ECCS actuation , and the IAB release pressure sensitivity that affects the timing of the opening of the ECCS valves ,

       . were the more important sensitivity parameters.

((

                                                          }}2(a ),(c) 5.1.3    Primary Release Scenario Pressure and Temperature Results The initial cond itions used by NRELAP5 analyses for each of the five cases in Table 3-4 are shown in Table 5-1 . The initial condition values in the second column of Table 5-1 are the nominal values plus the uncertainty or conservative allowance in parentheses. The assumed parameter values are consistent with the methodology as discussed by Section 3.5.1 and maximize heat sources while minimizing heat sinks. The decay heat conservatively used by these analyses is 120 percent of the 1979 ANS standard rather

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 than the methodology assumption (1979 ANS standard plus 2-sigma uncertainty). The 120 percent assumption bounds the required 2-sigma uncertainty required by the containment response analysis methodology (See Table 3-6). Table 5-1 Initial conditions for primary system release event analyses Parameter Conservative Containment Response Analysis Methodology Initial Condition H

                                                        }}2(a).(c)

The results of each of the five primary containment response analysis release analysis cases are summarized in the following sections, with more detailed results and discussion provided for the limiting CNV peak pressure scenario (Case 5 - RRV loss-of-coolant accident), and for the limiting CNV peak temperature scenario (Case 2 - RCS injection line break LOCA). 5.1.3.1 Case 1: Reactor Coolant System Discharge Line Break Loss-of-Coolant Accident The LOCA in the RCS discharge line initiates an M&E release from the downcomer into the CNV. The sequence of events is shown in Table 5-2. The CNV pressure response and temperature response are shown in Figures 5-1 and 5-2. The CNV peak GNV-pressure is

        ~ 705 psia for the reference case, and 900943 psia with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, 1200 psid IAB release pressuret,:. low-biased High CNV Level setpoint. fine CNV volume nodalization). The peak CNV wall temperature is 4QQ492 degrees F for the reference case , and ~ 510 degrees F with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, 9001200 psid IAB release pressure, ~             low-biased High CNV level signal ,

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 5-2 Case 1 sequence of events - reactor coolant system discharge line break loss-of-coolant accident Peak C NV Pressure Case Event Peak CNV Temperature Case Time (sec) Time (sec) LOCA in RCS discharge line For i;1eak i;1ressure case

                                        * -Loss of normal AC and DC power

ECCa actuation signal
FW/MS isolation WQ
Reactor trip Same
Containment isolation
MSIV closuFe
i;:w1v closure For peak pressure case Loss of normal AC power only for peak temperature case
Same High CNV pressure resulting in For 12eak 12ressure case
Containment isolation NIA1
MalV closure 51 5J Same
i;:vv1v closure -
ReactoF tFip

                                -For peak temperature case

Same ECCS actuation on IAB release 92 Same oressure ECCS valve open ing on pressure 4-&+95 difference beloi.v IAB release pressure NIA-Same

                                &- -   ---.1/ ----,-.,    ,r_ ----

ECCa actuation on highPeak CNV ~ for peak temperature reached : NIA109 ~ Same For 12eak 12ressure case : 510 °F For oeak temoerature case : Same EGGS val,-*e opening for Peak CNV 12ressure is reached: NIA112 4-WSame For i;1eak i;1ressure case : 943 i;1sia For oeak temperature case: Same 4W

                                .... ,I, ra..1\/          .,.. ,. . . . . ,n,...,... .......... :. . . ,

NIA

                                           -        ~               - --- r--*-

1c:.n'> ..., __ ____ C:\ NIA

                                ,.., __ 1, CNV pressure decreases to <50% of 1"'1\1\t * - - - - - - ** -- -
                                                                                                         *- e4.Q
-~     1900                                                                                                 NIASame peak pressure

Containment Response Analysis Methodology Technical Report TR-05 16-4 9084-NP Draft Rev. 10 1000 900 800 ro

*en
-a.

Q) 700

 !/)
 !/)

Q) 600 a.. cu 500 C

.....C Q) 400 z

0 300 X cu

200 100 0 0 200 400 600 800 1000 1200 1400 1600 1800 Time (sec) Figure 5-1 Case 1 containment vessel pressure - reactor coolant system discharge line break loss-of-coolant accident © Copyright 2018 by NuScale Power, LLC 71

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 500 CL 0 Q.) 450 ro Q.) Q. 400 E Q.) I-

~         350 z

0 X 300 ro

250 200 L...-~~-'--~~....__~~--'-~~-'--~~-'-~~-'-~~-'-~~--'--~~---'-----' 0 200 400 600 800 1000 1200 1400 1600 1800 Time (sec) Figure 5-2 Case 1 containment vessel wall temperature - reactor coolant system discharge line break loss-of-coolant accident 5.1.3.2 Case 2: Limiting Loss-of-Coolant Event - Reactor Coolant System Injection Line Break Loss-of-Coolant Accident The LOCA in the RCS injection line initiates an M&E release from the riser into the CNV. The results of the primary release event M&E release break spectrum analysis and sensitivity analyses have determined that Case 2 is the limiting LOCA peak pressure and overall limiting CNV wall temperature event. In addition, the analyses have shown that the Case 2 peak pressure results and CNV wall temperature results are -1 .I e and - 3.1 ~ percent higher, respectively, than the next highest result (Case 1); therefore, there is confidence that the overall limiting break location and scenario has been identified. The sequence of events is shown in Table 5-3 and detailed results for key parameters are shown in Figures 5-3 through 5-16. The peak CNV wall temperature is ~ 514 degrees F for the reference case, and ~ 526 degrees F with the combined effect of the adverse sensitivity parameters. The sensitivity parameters that contribute to the +-512 degrees F (-42.4 percent) increase are: f 1) the timing of ECCS valve opening as determined by the IAB release and high CNV level setpoints am:l-f2) the assumption of a loss of normal AC power. There was a small, adverse impact from the assumed (3) fine CNV volume & heat © Copyright 2018 by NuScale Power, LLC 72

Containment Response Analysis Methodology Technical Report TR-05 16-49084-NP Draft Rev. rn structure nodalization and (4) single failure of one RRV failing to open. The peak CNV pressure is 904894 psia for the reference case , and ~ 959 psia with the combined effect of the adverse sensitivity parameters. The sensitivity parameters that contribute to the

        +4-765 psi (-~7 .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...Q.l 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 FVVIV 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.lmv RPV level. In the containment response ana lysis 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 valves RWs . 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 ~ 952 seconds, and the open ing of the ECCS valves occurs at ~ 955 seconds (after the IAB release pressure is reacheda 3-second signal delay). The ECCS actuation and opening of the three RWs and one RRV causes the peak CNV wall temperature to occur at 4-+SG978 seconds. For the peak pressure case, the ECCS signa l actuates on IAB release pressurehigh C~N level at ~ 364 seconds, and the open ing of the ECCS valves occurs at 4-0M367 seconds. The ECCS actuation and opening of the three RWs and one RRV causes the peak CNV pressure to occur at 4400385 seconds. Then , as flow through the RWs dimishes, the primary and CNV © Copyright 2018 by NuScale Power, LLC 73

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 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. The primary system response for the RCS injection line LOCA CNV peak pressure case is shown in Figures 5-3 through 5-9. Figure 5-3 shows the primary pressure response. The initial depressurization phase due to the LOCA is followed by the rapid depressurization when the RVVs open. Figures 5-4 and 5-5 show the inventory in the pressurizer and in the riser. These figures show the expected trend of a decreasing level in the primary followed by a stabilization in inventory, with some liquid holdup in the pressurizer. A sensitivity study that decreased the interphase drag in the upper riser, riser upper plenum, pressurizer baffle, pressurizer, and the downcomer, with the intent of reducing liquid entrainment, showed that there was no adverse impact on the peak CNV pressure for this case. Figure 5-6 shows the primary coolant temperatures at six locations. Following ECCS actuation the temperatures converge and the cooldown proceeds. Figure 5-7 shows the LOCA and ECCS mass flowrates althoughincluding the spike in mass release when ECCS valves open is not sho 11,1n because of reduced plot frequency. Figures 5-8 and 5-9 show the integrated LOCA and ECCS mass flowrate and energy flowrate. Based on the integrated mass and energy flow rate plots, it is evident that the ECCS flow through the three RVVs into the CNV is significant. It is this M&E flow spike that causes the peak CNV pressure and wall temperatures to occur shortly thereafter as shown in Table 5-3 . The CNV and reactor pool responses for the RCS injection line LOCA peak pressure case are shown in Figures 5-10 to 5-15. Figure 5-10 shows the CNV pressure response and the limiting LOCA value of ~ 959 psia. This NRELAP5 analysis result is approximately g11% below the CNV design pressure of 40001075 psia. This is a key result of this limiting LOCA containment pressure response analysis case. Pressure increases rapidly to the peak value immediately following opening of the RVVs. Figure 5-11 shows the CNV liquid level increase as the unflashed break flow and condensed steam accumulates. Figure 5-12 shows the CNV vapor temperature . ((

                                                                             }}2(a ),(cl Figure 5-13 shows the temperature profile across the CNV wall at the 45 foot elevation . There is a large temperature gradient across the CNV wall. Figure 5-14 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 Figures 5-13 and 5-14 it is evident that the NPM design provides an effective heat sink for these short-term M&E analyses. Even with the conservative initial reactor pool level of ~ 65 ft above the pool floor and a temperature of 44011 O degrees F assumed in this analysis, the peak CNV wall temperature remains within the design limit.

Figure 5-15 shows the energy balance during the CVCS injection line LOCA and the trends of the heat sources and sinks. At approximately 44001000 seconds, the energy release © Copyright 2018 by NuScale Power, LLC 74

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 from the LOCA and the RVV valves decreases to below the energy transferred through the CNV wall. The CNV wall then continues to provide a strong heat sink for the sustained cooldown and depressurization of the module. As demonstrated by Table 5-3, the event progression for the RCS injection line LOCA peak pressure case and the peak CNV wall temperature case are similar. Accord ingly, only the CNV pressure and wall temperature figures will be presented for the peak CNV wall temperature case . Figure 5-16 shows the CNV pressure response for the RCS injection line LOCA peak wall temperature case. Figure 5-17 shows the CNV wall temperature response for the RCS injection line LOCA peak CNV wall temperature case and the overall limiting value of ~ 526 degrees F. This limiting NRELAP5 is less than the CNV design temperature of 550 degrees F. This is a key result of this limiting containment wall temperature response analysis case. Table 5-3 Case 2 sequence of events for limiting containment vessel temperature event - reactor coolant system injection line break loss-of-coolant accident Peak CNV Pressure Case Event Peak CNV Temperature Case Time (sec) Time (sec) LOCA in RCS injection li ne For Qeak Qressure case: _*_ Loss of normal AC power

FW/MS isolation WQ ~
Reactor triQ For Qeak temQerature case:
Loss of normal AC QOwer
FW/MS isolation High CNV pressure resulti ng in For Qeak Qressure case:
Containment isolation
MSIV closure 52 55~
F'WIV closure Same~

For Qeak temQerature case:

Containment isolation
T,Reactor *~:~

trip

                                    * -*-  -L * -
                                                  - .. *~

ECCS actuation on For Qeak Qressure case: 4-078364

IAB release Qressure ~ 952 For Qeak Qressure case:
hiah CNV level 4004-367 ECCS valve opening ~ 955 4400385 Peak CNV pressure ~ reached: NtA967

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Peak CNV Pressure Case Event Peak CNV Temperature Case Time (sec) Time (sec) For 12eak 12ressure case : 959 psia1 For peak temperature case: 939 psia Peak CNV temperature ~ deg reesreached: N./A384 ~ 978 For 12eak 12ressure case: 509 : F1 For Peak temperature case: 526 °F CNV pressure decreases to <50% of

 ~     2200                                                                 N./A-2500 peak pressure 2000 1800 1600
-ctl
.ii5 1400
-Q.

Q.) I...

J ti) ti) 1200 Q.)

I... a.. a.. 1000 0:: 800 600 400 0 500 1000 1500 2000 Time (sec) Figure 5-3 Case 2 primary system pressure - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 76

Containment Response Analysis Methodology Technical Report TR-051 6-49084-NP Draft Rev. 10 60 50

0 Q) Q) 40

.....J
.....C
  ~

Q) 30 a.. er N a.. 20 10 0 ,___ ____......_+-'-_ ___.__......_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ 0 500 1000 1500 2000 Time (sec) Figure 5-4 Case 2 pressurizer level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 77

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 50 g 0... 45 0::: 0 E 20 40 co E 0 35

a5 Q)

.....J "O

30

J O"

i

~      25 Cl) a.

cu 0 u 20 Q) Cl) Q: 15 ~~~~~~~_._~~~~~~-'-~~~~~~~'--~~~~~~_._~~-' 0 500 1000 1500 2000 Time (sec) Figure 5-5 Case 2 ri ser level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 78

Containment Response Analysis Methodology Technical Report TR-0516-4 9084-NP Draft Rev. 10 640 620 G:' 0 600 IJ)

  ~     580 ro
 '-     560 Q) a.

EQ) 540

~
"O      520
*s cr
~       500 U) 0 c:::    480 460 440 0                    500          1000                 1500                   2000 Time (sec)

SG Exit Liquid Temperature -+- Down comer Liquid Temperature ---M-Core Inlet Liquid Temperature -- Core Exit Liquid Temperature -e-Upper Riser Liquid Temperature ____._ PZR Liquid Temperature -e-Figure 5-6 Case 2 primary temperatures - reactor coolant system injection line break loss-of-coolant accident © Copyright 2018 by NuScale Power, LLC 79

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 3000 I I I I en E

.0 2500  -                                                                                              -
....ro Q) 2000  -                                                                                              -

0::: Q) en ro Q) Q) 0::: 1500 - - en en ro 2 1000 ... - en 0 0 w

-0      500  -                                                                                              -

C ro

ro

c::o 0 ... . . I I I

       -500                         I 0                     500              1000                  1500                 2000 Time (sec)

Figure 5-7 Case 2 break and emergency core cooling system mass flowrate - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 80

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

.0 2    60000 m

0::: Q) en m soooo Q) 0::: en j 40000 Cf) (.)

~    30000 "C

C: m

m 20000

Ill "C 2

 ~   10000 0) 2C:

O+-------......_______,.........._______.,________.,___ ___, 0 500 1000 1500 2000 Time (sec) Figure 5-8 Case 2 integrated loss-of-coolant accident and emergency core cooling system mass release - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 81

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 2

 ~    50000 Q)

II) m Q) o::: 40000 ei Q) C w Cf) 30000 (.) (.) w "O ffi 20000

ro

cc I2 10000 C 0 500 1000 1500 2000 Time (sec) Figure 5-9 Case 2 integrated loss-of-coolant accident and emergency core cooling system energy release - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 82

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1G 1000 900 800 ro

*w         700 0.
  ~

J f/)

600 f/)

0... 500 ro C Q) 400 C z 300 u 200 100 0 0 500 1000 1500 2000 Time (sec) Figure 5-10 Case 2 containment vessel pressure - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 83

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1G

20 z

-(.)

E 0

 ~       15 0

co E

- 0.....

Q) Q) 10 _J "O

 *5 0-

.:J

>         5 z

(.) o __________._________._________,._________.____, 0 500 1000 1500 2000 Time (sec) Figure 5-11 Case 2 containment vessel level - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 84

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 800 750 LL 0 700

J

+-'

650

Q) 0.. 600 E Q) I-Q) 550 E

J

~        500 z

(.) 450 X ctl

~        400 350 300 0                   500             1000                 1500                   2000 Time (sec)

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 500 450 LL a 4: 400 lO

 'Q" (U

(/) 350 Q) I...

J (U

I... Q) 300 a. E Q) I- 250

 ~      200 z

(.) 150 100 0 500 1000 1500 2000 Time (sec) CNV Wall ID Temperature at 45ft - + - CNV Wall Center Temperature at 45ft - M - - CNV Wall OD Temperature at 45ft ---+- Figure 5-13 Case 2 containment vessel wall temperature profile -reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 86

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 145 140

-u::-....

0 ( /) Q) 135

J cu Q) 130 a.

E Q) 125 f-0 0 Q.. 120 0 t5 cu Q) 115 0::: 110 105 0 500 1000 1500 2000 Time (sec) Reactor Pool Temperature at 17ft --+- Reactor Pool Temperature at 51ft -M-- Reactor Pool Temperature at 83ft ----- Figure 5-14 Case 2 reactor pool temperatures - reactor coolant system injection line break loss-of-coolant accident (peak pressure case) © Copyright 2018 by NuScale Power, LLC 87

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 14 12

~
~

Q) (.) C 10 co 8 co co

0) 6 Q)

C w 4 2 0 0 500 1000 1500 2000 Time (sec) Reactor Power CNV Heat Removal DHRS Heat Removal Steam Generator Power 8 Energy Transfer Through RPV Wall

8 Break and ECCS Energy Release Rate Figure 5-15 Case 2 energy balance - reactor coolant system injection llne break loss-of-coolant accident (peak pressure case)

Containment Response Analysis Methodology Technical Report TR-0516--49084-NP Draft Rev. 10 1000 900 800 ro

*u5
-Q.

Q) I...

J 700 en en Q) 600 I...

0... roC 500 I... Q) C 400 z u 300 X ro

200 100 0 0 500 1000 1500 2000 Time (sec) Figure 5-16 Case 2 containment vessel pressure - reactor coolant system injection line break loss-of-coolant accident (peak CNV wall temperature case) © Copyright 2018 by NuScale Power, LLC 89

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 500 u. 0 450

J

Q)

a. 400 E

Q) I-

~          350 z

() X 300 cu

250 200 '--~~~~~.__~~~~~.__~~~~~.__~~~~~......_~~~~~ 0 500 1000 1500 2000 Time (sec) Figure 5-17 Case 2 containment vessel peak wall temperature - reactor coolant system injection line break loss-of-coolant accident (peak CNV wall temperature case) 5.1.3.3 Case 3: Reactor Pressure Vessel High Point Degasification Vent Line Loss-of-Coolant Accident The LOCA in the RPV high point degasification line initiates an M&E release from the top of the pressurizer into the CNV. The sequence of events is shown in Table 5-4. The CNV pressure response and temperature response are shown in Figures 5-18 and 5-19. The CNV peak-GN-V pressure is a8e554 psia for the reference case, and ~ 901 psia with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, 1200 psid IAB release pressure}., high RCS flow, fine CNV volume nodalization). The peak CNV wall temperature is 471 479 degrees F for the reference case, and 484489 degrees F with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, goo1000 psid IAB release pressure, RRV single failure high RCS flow, fine CNV volume nodalization). Case 3 is non-limiting. © Copyright 2018 by NuScale Power, LLC 90

1- - -- - -- - - - - -- - - - -- - - -- - - - - - -- - - - ~ Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 5-4 Case 3 sequence of events - RPV high point degasification line break loss-of-coolant accident Peak CNV Pressure Case Event Peak CNV Temperature Case Time (sec) Time (sec) LOCA in RPV high point degasification line For geak gressure case only:

                                               * -Loss of normal AC and DC power WQ                                                                           SameQ

EGGS aGtl:1atieR 6ijRal
FW/MS isolation
Reactor trip
GeRtaiRFfleRt iselatieR
MSIV slesurs
FWIV slesurs High CNV gressure resulting in For geak gressure case:
Containment isolation 1 For geak temgerature case: 1
Reactor trig
FW/MS isolation
Loss of normal AC and DC gower assumed at turbine trig ECCS valves epsRactuation on bslew ~

9858 4e106 IAB release pressure 61 ECCS valve ogening 109 Peak CNV pressure t8e4reached : 44982 N./A128 For geak gressure case: 901 psia-1 For geak temgerature case: 894 gsia Peak CNV temperature ~ dsJrsssreached : N./A454 4-J-7478 For geak gressure case: 487 °F1 For geak temgerature case: 489 °F CNV pressure decreases to <50% of

          -~     1900                                                                  N./A-2000 peak pressure
         © Copyright 2018 by NuScale Power, LLC 91

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 1000 900 800 ro "in a. 700

 ~
 ~

en en Q) 600 0. roC 500 Q)

+-'

C 400 z> 0 300 X ro

200 100 0 0 200 400 600 800 1000 1200 1400 1600 1800 Time (sec) Figure 5-18 Case 3 containment vessel pressure - pressurizer spray supplyhiqh point vent line break loss-of-coolant accident © Copyright 2018 by NuScale Power, LLC 92

Containment Response Analysis Methodology Technical Report TR-05 16-4 9084-N P Draft Rev. rn 500 450 U::- 0 Q) I...

J 400 co I...

Q) a. E Q) r- 350

z () 300 X co

250 200 0 500 1000 1500 2000 Time (sec) Figure 5-19 Case 3 containment vessel wall temperature - pressurizer spray supplyhigh point vent line break loss-of-coolant accident 5.1.3.4 Case 4: Inadvertent Reactor Vent Valve Opening Anticipated Operational Occurrence The inadvertent RVVactuation anticipated operational occurrence (AOO) initiates an M&E release from the top of the pressurizer into the CNV. The sequence of events is shown in Table 5-5. The CNV pressure response and temperature response are shown in Figures 5-20 and 5-21 . The CNV peak--GN-V pressure is W-2856 psia for the reference case, and ggo911 psia with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, 1200 psid IAB release pressure , minimum primary systemlow RCS flowh fine CNV heat structure & reactor pool nodalization). The peak CNV temperature is 484483 degrees F for the reference case , and ~ 86 degrees F for the case with the combined effect of the adverse sensitivity parameters (loss of normal AC and DC power, 900 psid IAB release pressure). There was no sensitivity to any of the single failures. available, fine CNV volume nodalization). Case 4 is non-limiting. © Copyright 2018 by NuScale Power, LLC 93

Containment Response Analysis Methodology Technical Report TR-05 16-49084-N P Draft Rev. 10 Table 5-5 Case 4 sequence of events - inadvertent reactor vent valve opening event Peak CNV Pressure Event Peak CNV Temperature Case Case Time (sec) Time (sec) Inadvertent RVV actuation For Qeak Qressure case only: _*_ Loss of normal AC and DC power

eGG~ aGt1::1atieA SiQAal aOQ SameQ
FW/MS isolation
Reactor trip
GeAtaiAFAeAt iselatieA
M~IV Gl8Sl::IF8
FWIV Gles1::1r0 High CNV Qressure resulting in For Qeak Qressure case:
Containment isolation 0.2 0.2 For Qeak tem12erature case:
Containment isolation
Reactor triQ
FW/MS isolation ECCS 1,ial1,10s 8J:l8AiAQactuation on l::lelew ~

 &4-Z                          IAB release pressure eSn/a 10                           ECCS valves OQening                               n/a Peak CNV pressure fSWreached 7-927                                                                           N+A57 For Qeak Qressure case: 91 1 psiat For Qeak tem12erature case: 855 QSia Peak CNV temperature f reached N+A361                                                                          ~     37 For Qeak Qressure case: 482 lsQrses : Ft For Qeak tem12erature case: 486 °F CNV pressure decreases to <50% of peak
 -~     1700                                                                     N+A-2000 pressure

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 1000 900 800 ro "iii

-Q.

(1) L.. 700 Cl) Cl) (1) 600 L.. a.. cu 500 C L..

.....(1)C 400 z

u 300 X cu

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 500 I I I I 450 - - Ci:' 0 Q)

J 400 ro....

Q) a. E Q) I- 350 -

z 0 300 ~ co

250 - - I I I I 200 0 500 1000 1500 2000 Time (sec) Figure 5-21 Case 4 containment vessel wall temperature- inadvertent reactor vent valve opening event 5.1.3.5 Case 5: Limiting Overall Containment Vessel Pressure Event - Inadvertent Reactor Recirculation Valve Opening Anticipated Operational Occurrence The inadvertent RRV actuation initiates an M&E release from the downcomer into the CNV. The results of the primary release event M&E release break spectrum analysis and sensitivity analyses have determined that this AOO (Case 5) results in the limiting peak CNV pressure for all postulated events. The sequence of events is shown in Table 5-6 , and detailed results for key parameters are shown in Figures 5-22 through 5-35. The CNV peak GAA.Lpressure is ~ 941 psia for the reference case , and ~ 986 psia with the combined effect of the adverse sensitivity parameters. The sensitivity parameters that contribute to the +4-845 psi (-4-:94 .8 percent) increase are~ .(1) the timing of the ECCS valve opening as determined by the IAB release pressureand high CNV level setpoint; .(2) the assumption of a loss of normal AC and DC power; aR4-.(3) minimum primary system flow. There was no adverse impact from the single failure sensitivity studies. of an RRV; (4) fine CNV volume & heat structure and reactor pool nodalization: (5) fast RPV non-condensable release to CNV: and (6) low RCS flow. The peak CNV temperature is 49a492 degrees F for the reference case, and We512 degrees F with the combined effect of the adverse sensitivity parameters. The Case 5 peak CNV pressure case is the same case as the Case 5 peak CNV wall temperature case . (loss of normal AC power, low-biased high CNV level setpoint, single failure of an RRV, fine CNV volume & reactor pool nodalization). © Copyright 2018 by NuScale Power, LLC 96

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 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 RWs and the second RRV. The peak CNV wall temperature occurs following the RWs 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 RWs 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 RWs 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. Open ing of the RWsECCS valves occurs at

        ~ 171 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. Opening For the peak temperature case, opening of the three RWs 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 dimishes, 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 cool ing recirculation alignment.

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 The primary system response for the Case 5 inadvertent RRV opening event (peak pressure case) is shown in Figures 5-22 through 5-28. Figure 5-22 shows the primary pressure response. The initial depressurization phase due to the RRV opening is continued by the rapid de pressurization when the RWs open. Figures 5-23 and 5-24 show the inventory in the pressurizer and in the riser. These figures show the expected trend of a decreasing level in the primary followed by a stabilization in inventory, with some liquid holdup in the pressurizer. A sensitivity study that decreased the interphase drag in the upper riser, riser upper plenum , pressurizer baffle, pressurizer, and the downcomer with the intent of reducing liquid entrainment, showed that there was no adverse impact on the peak CNV pressure for this case. Figure 5-25 shows the primary coolant temperatures at six locations. Following ECCS actuation the temperatures converge and the cooldown proceeds. Figure 5-26 shows the RRV opening and ECCS mass flowrates . It is evident that the ECCS flow immediately following ECCS actuation , mainly the flow through the three RWs into the CNV, is significant. It is this flow spike that causes the peak CNV pressure and wall temperatures to occur shortly thereafter as shown in Table 5-6. Figures 5-27 and 5-28 show the integrated LOCA and ECCS mass flowrate and energy flowrate . The CNV and reactor pool response for the Case 5 inadvertent RRV actuation LOCA is shown in Figures 5-29 to 5-34. Figure 5-29 shows the CNV pressure response and how pressure rapidly increases to the limiting peak value of ~ 986 psia. This limiting NRELAP5 result can be compared to the CNV design pressure of 40001075 psia. This is a key result of this limiting containment response analysis case. Figure 5-29 also demonstrates the long term cool ing capability of the UHS. CNV pressure is reduced to below 50 percent of the peak value within two hours of accident initiation. Figure 5-30 shows the CNV liquid level increase as the unflashed break flow and condensed steam accumulates. Figure 5-31 shows the CNV vapor temperature . Initially, flashing of the break flow at low CNV pressure results in a temperature decrease . ((

                 }} 2 (aJ,(cJ Figure 5-32 shows the peak CNV wall temperature and the limiting va lue of We492 degrees F. Figure 5-33 shows the temperature profile across the CNV wall at the 45 foot elevation. There is a large temperature gradient across the CNV wal l. Figure 5-34 shows the reactor pool temperatures for a range of elevations. The reactor pool temperature does not increase significantly for the short duration of these M&E release analyses. From Figures 5-31 through 5-34 it is evident that the CNV wall is the significant heat sink in the short-term. Even with the conservative initial reactor pool level of aa65 ft above the pool floor and a temperature of 44G110 degrees F assumed in these analyses, the CNV wall is capable of maintaining the peak CNV pressure within the design limit.

Figure 5-35 shows the energy balance during the RRV loss-of-coolant accident and the trends of the heat sources and sinks. At approximately +OG750 seconds the energy release from the LOCA and the RW valves decreases to below the energy transferred through the CNV wall. The CNV wall then continues to provide a strong heat sink for the sustained cooldown and depressurization of the module. © Copyright 2018 by NuScale Power, LLC 98

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 5-6 Case 5 sequence of events - inadvertent reactor recirculation valve opening event Peak CNV Pressure Event Peak CNV Tem12erature Case Case Time (sec) Time (sec) WQ Inadvertent RRV actuation :Actuation Q For peak temperature case

Loss of normal AC power
FW/MS isolation For peak pressure case

_*_Loss of normal AC and DC power

ECCa actuation signal
FW/MS isolation
Reactor trip
ContaiAFflent isolation
MalV closure
FWIV closure High CNV pressure resulting in:

For peak temperature case 0.4

Containment isolation 0.4
Reactor trip For peak pressure case
Containment isolation

~   74                     ECCS valve openingactuation on                168 differentiat For peak temperature case

high CNV level For peak pressure belewcase

_*_IAB release pressure nnn--;,-n 1-

77 ECCS valve ooenina 171

~ 91                       Peak CNV pressure reached :                   182 For peak pressure case: 986 psia For oeak temoerature case: 967 osia
~   596                    Peak CNV temperature reached :                180 For peak pressure case : 492 °F For oeak temoerature case: 512 °F
-~     1800                CNV pressure decreases to <50% of             -1800 peak pressure

Containment Response Analysis Methodology Technical Report TR-05 16-49084-NP Draft Rev. 10 2000 1800 1600

-ro
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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 60 50

0 Q) (]) _J 40 C (]) e (]) 30 0.... c::: N 0.... 20 10 0 0 200 400 600 800 1000 1200 1400 1600 Time (sec) Figure 5-23 Case 5 pressurizer level - inadvertent reactor recircu lation valve opening event © Copyright 2018 by NuScale Power, LLC 101

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480 460 0 200 400 600 800 1000 1200 1400 1600 Time (sec) SG Exit Liquid Temperature - - - Downcomer Liquid Temperature ____.._ Core Inlet Liquid Temperature - - Core Exit Liquid Temperature -a-- Upper Riser Liquid Temperature --tt-- PZR Liquid Temperature --+- Figure 5-25 Case 5 primary temperature - inadvertent reactor recirculation valve opening event © Copyright 2018 by NuScale Power, LLC 103

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 2500 I I I I I I ui E 2000 .... -

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0 200 400 600 800 1000 1200 1400 1600 Time (sec) Figure 5-27 Case 5 integrated loss-of-coolant accident and emergency core cooling system mass flow rate - inadvertent reactor recirculation valve opening event © Copyright 2018 by NuScale Power, LLC 105

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

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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 1Q. 1000 900 800 ro

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200 100 0 0 200 400 600 800 1000 1200 1400 1600 Time (sec) Figure 5-29 Case 5 containment vessel pressure - inadvertent reactor recirculation valve opening event (overall limiting pressure case) © Copyright 2018 by NuScale Power, LLC 107

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 z 20 (_) 0 E 0

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

........ 900                                                                                         -

LL 0 800 - - 700 ... - 600 ... - 500 ... - 400 ..._~~-'-'~~~~*~~~---~~-......_*~~--'-'~~~~*-----~~-..._*~~--'-'~~----' 0 200 400 600 800 1000 1200 1400 1600 Time (sec) Figure 5-31 Case 5 containment vessel vapor temperature - inadvertent reactor recirculation valve opening event © Copyright 2018 by NuScale Power, LLC 109

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u 150 100 0 200 400 600 800 1000 1200 1400 1600 Time (sec) CNV Wall ID Temperature at 45ft --+- CNV Wall Center Temperature at 45ft ~ CNV Wall OD Temperature at 45ft ~ Figure 5-33 Case 5 containment vessel wall temperature profi le - inadvertent reactor recircu lation valve open ing event © Copyright 2018 by NuScale Power, LLC 111

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 145 LJ.. 0 140 Cl) 135

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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 14 12

~
~

Q) u 10 C co 8 co al 0) 6

Q) C w 4 2 0 0 200 400 600 800 1000 1200 1400 1600 Time (sec) Reactor Power CNV Heat Removal M DHRS Heat Removal Steam Generator Power

                                                                                            *B Energy Transfer Through RPV Wall

8 Break and ECCS Energy Release Rate Figure 5-35 Case 5 energy balance - inadvertent reactor recircu lation 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 pressurization that reaches the 9.5 psia high pressure setpoint. The following automatic actions occur on high CNV pressure:
containment isolation including MS IV and FWIV closure
reactor trip
turbine trip
DHRS actuation (Note: no DHRS operation is credited for this event analysis)

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 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 © Copyright 2018 by NuScale Power, LLC 113

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 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 operation following 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 decrease 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 449449 psia at 4442 seconds. This limiting NRELAP5 result can be compared to the CNV design pressure of 4G001075 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 ~        28 degrees F at 4941 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 5-4 7 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 evident that the CNV wall is the significant heat sink for these containment response analyses. Even with the conservative initial reactor pool level of a565 ft above the pool © Copyright 20 18 by NuScale Power, LLC 11 4

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 floor and a temperature of 440110 degrees F assumed in these analyses, the CNV wall is capable of maintaining the peak CNV pressure and temperature within the design limit. Figure 5-49 shows the energy balance during the MSLB and the trends of the heat sources and sinks. At approximately a00300 seconds the energy release from the MSLB has dimished, and the energy transfer through the CNV wall and from the DHRS to the pool dominate. This energy balance is consistent with the cooldown of the primary system shown in Figure 5-37. Table 5-7 Main steam line break sequence of events Time (sec) Event 0 MSLB 0- ~ Low steam line pressure resulting in

Reactor trip
Turbine trip
MSIV closure
FWIV closure
DHRS aotuation High CNV pressure resulting in
Containment isolation

         ~ 4                          Closure of FWRV complete 4442                         Peak CNV pressure 484 1                        Peak CNV temperature
         -200                         CNV oressure decreases to <50% of peak pressure

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 1200 1000 -*en ro

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a::: 480 460 0 200 400 600 800 1000 Time (sec) SG Exit Liquid Temperature --+- Downcomer Liquid Temperature -*- Core Inlet Liquid Temperature -- Core Exit Liquid Temperature --a-Upper Riser Liquid Temperature -- PZR Liquid Temperature --e-Figure 5-37 Main steam line break pri mary temperature © Copyright 2018 by NuScale Power, LLC 117

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 1950 1900 1850 ro

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z 150 0 100 L-~~~~~_._~~~~~----~~~~~ .........~~~~~--'~~~~~--' 0 200 400 600 800 1000 Time (sec). CNV Wall ID Temperature at 45ft - - - t - CNV Wall Center Temperature at 45ft - * - CNV Wall OD Temperature at 45ft - - Figure 5-47 Main steam line break containment vessel wall temperature profile © Copyright 2018 by NuScale Power, LLC 127

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 124 122 u. 0 ( /) Q.) 120 ro Q.) 118 a. E Q) I-0 116 0 a.. 0 t3 114 ro Q.) a::: 112 110 0 200 400 600 800 1000 Time (sec) Reactor Pool Temperature at 14ft Reactor Pool Temperature at 28ft Reactor Pool Temperature at 44ft Reactor Pool Temperature at 57ft Figure 5-48 Main steam line break reactor pool temperature © Copyright 2018 by NuScale Power, LLC 128

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 14 12

e, 10 (I) (.) C 8 ro ci5 co Cl 6 (I) C LU 4 2 0 0 200 400 600 800 1000 Time (sec) Reactor Power CNV Heat Removal DHRS Heat Removal Steam Generator Power Energy Transfer Through RPV Wall Break and ECCS Energy Release Rate 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 pressure:

containment isolation including MSIV closure and FWIV closure
DHRS actuation
reactor trip
tl:.1rbine 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 of the emergency core cooling system RRVs and RVVs does not occur until the pressure differential decreases to below the IAB release pressure. In the containment response

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn analysis methodology the IAB release pressure is an important analysis input as it determines the second M&E release into the CNV via the RWs . A higher IAB release pressure results in an earlier opening of the ECCS valves when the RCS is hotter. The peak CNV pressure and peak CNV wall temperature occur following the RW actuation , after the CNV has been preheated by the initial M&E release. Sensitivity studies of single failures have determined that a failure of a ~ MSIV to close had an adverse impact on the CNV peak pressure and temperature res ults. Following the alignment of the module to mitigate the initial secondary blowdown phase, the secondary system pressure and inventory decrease due to the loss of inventory. The CNV pressurizes and the steam condenses on the cold ID of the CNV. The condensate flows down the CNV walls and accumu lates with unflashed secondary 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. After the end of the secondary blowdown phase decay heat removal is via the DHRS. Opening of the ECCS valves occurs at 12,012 11,566 seconds when the pressure differential decreases to below the 1200 psid IAB release pressure. Th is causes the CNV peak GN-V-pressure (442416 psia) and the peak CNV wall temperature (444407 degrees F) at - 11 ,60012,058 and - 12,20011 ,870 seconds, respectively. As flow through the RWs dimishes, the primary system 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. The module response for the FWLB analysis is shown in Figure 5-50 through Figure 5-64. Figure 5-50 shows the SG pressure response with the affected SG (SG2) depressurizing via blowdown out the break into the CNV and stabilizing at a low pressure of approximately 30 psia . The unaffected SG (SG1 ) pressure fluctuates in response to DHRS heat transfer. The affected SG repressurizes by reverse break flow on ECCS valve open ing. Then , both SGs depressurize as ECCS heat transfer dominates. Figure 5-51 shows the gradual primary system cooldown due to DHRS , and the increase in the cooldown rate with the opening of the ECCS valves . Figure 5-52 shows the relatively steady pressurizer level decrease during DHRS cooling and then a rap id level decrease when ECCS valves open . Figure 5-53 shows the riser level remain ing full until the ECCS valves open , and then level rapidly decreases before stabilizing. Primary system pressure (Figure 5-54) gradually decreases during the DHRS cooldown period due to loss of pressurizer heaters and then rapidly depressurizes on ECCS valves opening. Figure 5-55 through Figure 5-57 show the break and ECCS mass release , the integrated mass release , and the integrated energy release into the CNV, respective ly. The FWLB flow rate and integrated mass release is not significant due to the small SG inventory. Due to the insignificance of the secondary break flow, the effect of liquid entrainment is also insignificant. The primary system M&E release through the three RWs is the significant M&E release event for the FWLB accident. The CNV and reactor pool responses for the FWLB are shown in Figure 5-58 through Figure 5-63. Figure 5-58 shows the CNV pressure response. The initial M&E release results in the CNV pressu rizing to -60 psia before heat transfer to the CNV wall results in pressure stabilizing at-15 psia. Then pressure rapidly increases to the limiting peak value of 4424 16 psia following opening of the RWs. This limiting NRELAP5 resu lt can be © Copyright 2018 by NuScale Power, LLC 130

r Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 compared to the CNV design pressure of 4-0001075 psia and to the limiting MSLB and primary release event results. The FWLB peak CNV pressure result is higher than the MSLB result, but is bounded by the limiting LOCA results. This is a key result in th is FWLB containment response analysis. Figure 5-59 shows the CNV vapor temperature . ((

                                                                                             }} 2 (a).(cl Figure 5-60 shows the peak CNV wall temperature and the limiting value of 444407 degrees F.

Th is limiting NRELAP5 result can be compared to the CNV design temperature of 550 degrees F, and to the limiting MSLB and LOCA results . The FWLB is bounded by both the MSLB result and the limiting primary release event results. Th is is a key resu lt in this FWLB containment response analysis. Figure 5-61 shows the CNV level response with an initial level increase fo llowing the initial M&E release, and the second level increase following the delayed opening of the ECCS valves. Figure 5-62 shows the temperature profile across the CNV wall at the 45 foot elevation . A significant temperature gradient exists. Figure 5-63 shows the reactor pool temperature for a range of elevations. Clearly the reactor pool temperature does not increase significantly through the time of peak CNV pressure and temperature . Even with the conservative initial reactor pool level of ~ 5 ft above the pool floor and a temperature of 440110 degrees F assumed in these analyses, the CNV wall is capable of maintaining the peak CNV pressure and temperature within the design limit. Figure 5-64 shows the energy balance during the FWLB and the trends of the heat sources and sinks. The DHRS and CNV wall heat sinks combine to exceed the ECCS energy release and resu lts in a sustained cooldown of the primary system as shown in Figure 5-51 .

 © Copyright 2018 by NuScale Power, LLC 131

Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Table 5-8 Feedwater line break sequence of events Time (sec) Event 0 FWLB

Loss of normal AC and DC 12ower at event initiation resulting in ECCS actuation signal
Reactor triQ
Turbine trio High CNV pressure res1:.1lting infollowed by:

0 - 2.,6

Containment isolation

                                            -e.!..._MSIV     closure
                                            -e.!..._ FWIV    closure
                                            -e.!..._ DHRS actuation

Reactor trip

                                            * +1:Jrsine trip

boss of norFflal AG anEI l:;)G po,,.;er on t1:Jrl3ine trip res1:.1lting in eGG~ act1:Jation signal w -12 Peak CNV pressure from secondary M&E release

                   ~       11.566      ECCS valve opening on differential pressure below IAB release pressure (<1200psid)

Peak CNV pressure

                 ""12,Qag - 11.600
                 ""12,2QQ- 11.870     Peak CNV temperature
                      -13,000         CNV pressure decreases to <50% of peak pressure

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          © Copyright 2018 by NuScale Power, LLC 140

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                © Copyright 2018 by NuScale Power, LLC 141

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                   © Copyright 2018 by NuScale Power, LLC 142

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© Copyright 2018 by NuScale Power, LLC 144
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 400 Ci:' 350 0 ii= Li)
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z 0 150 100 0 5000 10000 15000 20000 Time (sec) CNV Wall ID Temperature at 45ft - - CNV Wall Center Temperature at 45ft -w-- CNV Wall OD Temperature at 45ft - - Figure 5-62 Feedwater line break containment vessel wall temperature profile © Copyright 2018 by NuScale Power, LLC 145
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 240 LL 0 220 ( /) Q.) 200
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Containment Response Analysis Methodology Technical Report TR-0516-4 9084-NP Draft Rev. 10 14 12
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C w 4 2 0 0 5000 10000 15000 20000 Time (sec) Reactor Power --+-- C NV Heat Removal ~ DH RS Heat Removal --- Steam Generator Power --a-- Energy Transfer Through RPV Wall ------- Break and ECCS Energy Release Rate ~ Figure 5-64 Feedwater line break energy balance 5.4 Margin Assessment The following subsections discuss the analytical and design margin incorporated into the NPM design. Section 5.4.1 describes margin inherent in the enhanced requirements imposed on the CNV as an American Society of Mechanical Eng ineers (ASME) Boiler and Pressure Vessel Code (BPVC) Class 1 vessel. Section 5.4.2 describes conservative modeling assumptions in the containment peak pressure and temperature analysis, along with the results of a nominal case quantifying the conservatism resulting from some of these assumptions. 5.4.1 Containment 'Jessel Design Atmospheric Pressure Margin The CNV consists of an upper section and a lmver section joined at tho closure flange about one third of the way up tho vessel. Tho CNV upper section is made of SA 508 Grado 3, Class 2 lov.i alloy steel with 0.125 in . stainless stool cladding on tho inside surfaces and 0.250 in . stainless stool cladding on the outside surface. The top head of tho upper section is a torispherical head and tho vessel shell has an OD of 176.75 in . and a base metal wall thickness of 3.00 in. Tho C~JV lm*,or section is made of SA Q65 P:XM 1Q austenitic stainless stool and does not require any cladded surfaces. The bottom head on tho lower section is a torispherical head and tho vessel shell consists of throe regions. The bottom region , in tho area of the fuel , has an OD of 134.50 in. and a wall thickness of 3.00 in ., tho © Copyright 2018 by NuScale Power, LLC 147
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 middle region is a transition between the diameters of the bottom and top region . The top region has an OD of 176.50 in . with a 1Nall thickness of 3.250 in. Containment vessel shell to shell , shell to head , and penetration to shell welds are high quality, full penetration welds performed in the vessel fabrication shop. Material properties of the structural portions of the CNV are summarized in Table 5 9. Table 5 9 Containment vessel material properties Material Prnperty (ksi} Material Sm Sy Sv 440-~ aa0-~ 4--~ aa0-~ 440-~ aa0-~
,6, 5GB Grago a, JG..

-0 JG..

-0 ~ 5-&:4 ~ ~
Glass~ GA 9e5 FXM ~g ~ ~ ~ ~ 99:-7 83-4 The CNV has a design pressure of 1,000 psia and design temperature of 550 degrees-~ The limiting LOCA peak containment pressure is 921 psia and the o,,erall peak containment pressure is 951 psia . The design pressure meets the requirements of ASME Boiler and Pressure Vessel Code (BPVC) , Section Ill , Paragraph NGA 2142.1(a) and NB 3112.1(a) by bounding the most severe Level A service level pressure and ASME BPVC , Section 111 , Paragraph NE 7120(b) by not exceeding service limits specified in the design specification. The CNV is designed , fabricated , inspected and tested as an ASME Code Class 1 component in accordance toASME BPVC , Section Ill , Subsection ~JB . Pressure boundary forgings and 'Neid filler materials are tested for mechanical and fracture toughness to the requirements of ASME BPVC , Section Ill , Article NB 2000 . The CNV is a high quality, shop fabricated vessel , made to the requirements ofASME BPVC , Section Ill , Article ~JB 4000 , with low alloy welds post weld heat treated . Many requirements between an NB and MC vessel are similar. However, one substantive difference between an NB and MC class vessel is in weld inspection. The main welds forming the pressure boundary shell are Category A, B and C full penetration , butt ,,velds. In an NB class vessel these welds are required to have a volumetric and either liqu id penetrant or magnetic particle inspection performed per ASME BPVC , Section Ill , Subarticle NB 5200 . The corresponding welds in an MC class vessel only require a fully radiographed inspection per ASME BPVC , Section Ill , Subarticle NE 5200 . By only performing a radiograph examination of the 'Neid , all potential flaws in the weld may not be detected, which reduces the quality and potentially the strength of the weld . After fabrication of the CNV is completed , a shop hydrostatic test of the vessel is performed perASME BPVC, Section 111 , Article NB 6000. Before hydrostatic testing , 100 percent of the pressure boundary welds are inspected. Inspection is performed to ASME BPVC , Section Ill , Subsubarticle NB 5280 and Subarticle IVVB 2200 using examination methods of ASME BPVC Section V except as modified by ASME BPVC , Section 111 , Paragraph NB 5111 . The hydrostatic test is done to a minimum pressure of 1,265 psia (125 percent) and a maximum pressure of 1,340 psia at a minimum temperature of 70 degrees-F and a maximum temperature of 140 degrees-F. The hydrostatic pressure and © Copyright 2018 by NuScale Power, LLC 148
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 temperature is held for a minimum of 10 minutes. The pressure is then decreased to design pressure and held for a minimum of fo1:1r homs and then inspected for leaks. After the test is completed , the press1:1re bo1:1ndary welds are inspected again to the same req1:1irements 1:1sed before the test. The NB 6000 hydrostatic test is performed to a greater press1:1re then req1:1ired by l'!>Hi: 6000. l'!>JE 6321 specifies a minim1:1m test pross1:1ro of only 110 percent and NE 6322 a maxim1:1m of 116 percent. The N1:1Scale CNV is tested to a press1:1re margin 15 percent greater than conventional steel containment struct1:1res. The ASME BPVC provides allowable stress limits to prevent gross failure of vessels. Analysis of primary stresses for the CNV shows that the most limiting cross sections occ1:1r in the general section of the shells. Table 5 1O s1:1mmarizes tho design condition (1 ,000 psia and 550 degrees-f) stress results for tho 1:1ppor and lower CNV. The table shows that the 1:1ppor and lower Cl'!>JV stresses in the shells have 3.8 percent and Q.5 percent margin to the ASME alloi.vable stresses , respective ly. The 1:1pper and lower CNV rated design press1:1re co1:1ld be increased by the margins to 1,038 psia and 1,095 psia , respectively. A comparison of the limiting LOCA peak press1:1re (Q21 psia) and the maximum overall peak containment pross1:1re (Q51 psia) to the increased press1:1re ratings of tho 1:1pper and lower shells shows is provided by Table 5 10, v,rhon tho margin between the maxim1:1m membrane stresses to tho ASME allowable stresses is considered. Table 5 1O Design cond ition stress s1:1mmary IAGFeased Maxim1:1m Allowable StFess to Increased Increased oveFall CN-V Membrane Membrane a110,...iable GNV rated bOG,~, peak peak Section Stfess Stress, Sm margin press1:1re press1:1re press1:1re {ksij {ksij fo/o} {psia) margin{%) Margin
~ ~ ~ ~ 4-;038 ~ 9:4 Gt>JV sl=!ell bGwef ~ ~ ~ 4-;Wa ~ ~
Gt>JV sl=!ell Additionally, the 1:1pper Cl'!>JV shell is low alloy steel with stainless steel cladding on both sides. The base metal of the 1:1pper section is 3.00 in. Cladding adds an additional 0.375 in . to the thickness . This is an increase of 12.5 percent in th ickness and therefore an increase in pressure capacity. The increase in thickness proportionally increases pressu re capacity of the CNV. While in tho ASME Code calculations cladding is neglected, tho additional material does provide additional margin. When this margin is considered the increased pressure rating of the upper CNV becomes 1.125*1,038 psia - 1,168 psi . The margin of maxim1:1m containment peak pressure (Q51 psia) to the increased rated press1:1re of the upper CNV is 22.8 . percent. NuScale analyses do not couple the internal and external containment pressures. Internal pressures are conservatively evaluated with an assumed external pressure of O psia. The atmospheric pressure, acting against the exterior CNV surface, is neglected . The overall limiting peak CNV peak pressure results from an inadvertent reactor recirculation valve opening anticipated operational occurrence with a loss of normal AC © Copyright 2018 by NuScale Power, LLC 149
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 and DC power. The overall limiting CNV peak pressure, is 986 psia, which is approximately 8 percent below the design pressure of 1075 psia, which occurs at a CNV elevation at the bottom of the CNV. Atmospheric pressure and reactor pool hydrostatic head, acting against the CNV exterior surface, provides approximately 22 psi additional margin, that is not credited by the CNV response analysis methodology. This demonstrates additional margin in the CNV design that is not considered by the containment design pressure ratingresponse analysis. 5.4.2 Decay Heat Removal System Availability The LOCA (Case 2) and AOO Gase 2 and {Case 5} are performed with and without DHRS available to estimate the impact of DHRS availability on the CNV peak pressure response. The DHRS is conservatively not credited in the design basis containment response analysis cases. The NRELAP5 code has not been validated to cover DHRS performance during LOCAs or valve opening events. However, the DHRS is a single-failure proof safety-related system that can be credited in the future, with additional NRELAP5 val idation, if the CNV pressure margin is reduced for any reason (design changes). The results of the DHRS available cases indicate that more than 25 psiaabout 37 psi additional margin could be gained by credit for DHRS availability. 5.4.3 Conclusion The NPM design provides sufficient margin to satisfy the requirements of GDC 16 and 50. The LOCA peak pressure provides approximately 11 % and the AOO peak pressure provides approximately 8% margin to the CNV design pressure of 1075 psia, to address the acceptance criteria given by DSRS Section 6.2.1.1 .A (See Table 2-2). The CNV response to the limiting LOCA event and AOO transient are conservatively calculated and demonstrate that the peak calculated pressures are below the CNV design pressure and decrease in pressure to one-half of the peak value within 24 hours. Further assurance of sufficient margin is provided through consideration of the robust designatmospheric pressure and hydrostatic head, acting against the CNV exterior surface and construction of the NPM and conservative assumptions related to the creditingavailability of the DHRS system in the containment response analysis. Consideration of external pressure acting against the CNV exterior surface reduces the differential pressure across the CNV wall by about 22 psi . The determination of NPM design pressure, in accordance with ASME Class 1 criteria, is conservative relative to Class MC and CC containments. This design pressure does not consider the additional margin provided by the internal and external cladding of the upper CNV shell. The effect of DHRS actuation in reducing peak containment pressure was not credited in the containment response analysis. Consideration of the effect of DHRS actuation, along with external pressure acting against the CNV exterior surface, reduces the differential across the CNV wall by approximately 59 psi. The containment response analysis methodology, analysis results and further conservatisms related to design and system operation provide assurance that the NPM design demonstrates sufficient margin to satisfy the requirements of GDC 16 and 50. © Copyright 2018 by NuScale Power, LLC 150
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 6.0 Summary and Conclusions This report presents the NuScale containment response analysis methodology for determining primary system and secondary system mass and energy releases and the resultant CNV pressure and temperature response for the NPM . A spectrum of LOCAs and ECCS valve opening events were analyzed along with the MSLB and FWLB accidents. The scope of the methodology is the short-term CNV response for comparison to the CNV pressure and temperature design limits. Equipment qualification and the long-term NPM response are not in the scope of this report. The containment response analysis methodology uses the NRELAP5 code , which originates from the RELAP5-3D© code . The NRELAP5 code includes new capabilities added by NuScale to enable modeling of the design features and transient response of the NPM. The NRELAP5 model of the NPM used in the containment response analysis methodology is based on the NuScale LOCA and non-LOCA evaluation models with limited revisions and additions necessary for application in the containment response analysis methodology. NuScale has completed LOCA and non-LOCA phenomena identification and ranking tables . The results of the PIRTs have been used in the development of the NRELAP5 code and model. The NRELAP5 LOCA and non-LOCA models have been assessed by comparison to generic separate effects tests and intergral effects test, as well as to the NuScale design-specific NIST-1 facility separate effects and integral LOCA tests . The containment response analysis methodology is shown to meet the intent of Section 6.2 of the NuScale DSRS. Based on the systematic application of conservative initial conditions and boundary conditions in the containment response analysis methodology, the margin in the containment response analysis methodology is judged to be sufficient. Conservative NRELAP5 demonstration analyses of the containment response analysis methodology have been performed for a spectrum of primary system LOCAs and ECCS valve opening events, and for the MSLB and FWLB accident secondary system events. Sensitivity stud ies have been used to identify the bounding scenarios and trends. The following insights were obtained:
The bounding scenarios for both peak CNV pressure and temperature were determined to be primary system release events. The secondary system break events may include ECCS actuation , which essentially combines an initial secondary system M&E release with a subsequent primary system M&E release, but they are non-limiting scenarios.

The limiting M&E release scenario is characterized by an initial heatup and pressurization of the CNV due to the LOCA or ECCS valve opening , and then the subsequent opening of the RVVs on following the pressure differential decreasing to below the IAB release pressure. It is the second M&E release that drives the CNV to the peak CNV pressure and peak CNV wall temperature results .

The heat capacity of the CNV wall , rather than heat transfer to the reactor pool, provides the short-term heat sink to limit the peak CNV pressure and temperature.
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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn
For the limiting cases the results of the sensitivity studies, including postulated single fai lures and the loss of normal AC and DC power, showed only a limited impact (""'.:::.1 percent) on the key figures-of-merit. The loss of normal AC and DC power and the timing of ECCS valve opening were the most important sensitivity parameters.
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 ~ 526 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 pressure is 92-4-approximately 959 psia_.. which ffiSG-results from a reactor coolant system injection line break case, with a loss of normal AC and DC power. The LOCA event peak CNV pressure is below the CNV design pressure of 4GOO 1075 psi a. The LOCA peak CNV pressure and wall temperature bound the main steamline break (MSLB) and feedwater lind break { FWLB} results. The overall limiting peak CNV accident pressure is approximately 951 psia and it 986 psia, which is approximately 8 percent below the containment design pressure of 1075 psia . It results from an inadvertent reactor reci rculation valve opening AGG, anticipated operational occurrence with a loss of normal AC and DC power. +R-isThe peak pressure of the limiting anticipated operational occurrence is also less than the CNV design pressure of 4G001075 psia. The CNV pressure for th is limiting case is reduced to below 50 percent of the peak value in less than 2 hours, demonstrating adequate NPM containment heat removal. Section 5.4 discussed margin in the NPM design that is not included in the CNV design pressure rating or modeled in the containment response analyses . Design factors conservatively not credited include CNV shell stress margins, CNV cladding material atmospheric pressure acting against the CNV exterior surface and the availability of the DHRS. The containment response analysis demonstrates that the NPM design has adequate margin to design limits and that it satisfies the requriements of GDC 16 and 50 and PDC 38. © Copyright 2018 by NuScale Power, LLC 152
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 7 .0 References 7.1 Source Documents 7.1 .1 U.S. Code of Federal Regulations , Title 10, Part 50. 7.1.2 U.S Code of Federal Regulations, "Appendix A to Part 50 - General Design Criteria for Nuclear Power Plants, " (1 OCFR50, Appendix A) . 7.1.3 U.S. Nuclear Regulatory Commission, "Transient and Accident Analysis Methods," Regulatory Guide 1.203, December 2005. 7.1.4 U.S. Nuclear Regulatory Commision, "Design Specific Review Standard for NuScale SMR Design," Section 6.2.1, June 2016. 7.1.5 U.S. Nuclear Regulatory Commision, "Design Specific Review Standard for NuScale SMR Design," Section 6.2.1.1.A, June 2016. 7.1 .6 U.S. Nuclear Regulatory Commision, "Design Specific Review Standard for NuScale SMR Design," Section 6.2.1.3, June 2016. 7 .1 .7 U.S. Nuclear Regulatory Commision, "Design Specific Review Standard for NuScale SMR Design," Section 6.2 .1 .4, June 2016. 7.2 Reference Documents 7 .2.1 Nu Scale Power, LLC , "LOCA Evaluation Model," TR-0516-49422, Revision 0. 7.2.2 NuScale Power, LLC , "Non-LOCA Transient Analysis Methodology Report," TR-0516-49426 , Revision 0. © Copyright 2018 by NuScale Power, LLC 153
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 8.0 Appendicies 8.1 Mass and Energy Input The purpose of this Appendix is to present the mass and energy release to the CNV during the limiting LOCA event (Case 2), the limiting overall peak CNV pressure event (Case 5) and the limiting secondary system release event (MSLB).,_.,. up to the time that the peak pressu re is reduced to one half its value. 8.2 Heat Sink Tables The purpose of this Appendix is to present the passive heat sink characteristics credited in the containment response analysis methodology. Table 8-1 Limiting Peak Pressure Case - Mass and Energy Release Time (s) <1l Mass Release (lbm/s) Enerav Release (Btu/s) Q G G
~ Q Q ~ Ja~ .7~~ HJ~4Q ~ a29.rn9 2iQJ97 ~ 49a .~as 247i7~ ~ 4 ~~ .279 2JeSQ2 rna .a2a J97.222 227J2J ~ JSQ.QQJ 2rnaJG ~ 27.eJa rn2e .9~ ~49Hrn ~ ~ 4 rn .ae ~J~HiSQ ~29.e4a ~~ S2.2J ~rngrng ~ 9i7.7~7 QQQ~~7 ~ J~ .eaa SeJ.7~2 SQJSQJ ~ rn2J.Q~ 77~94J ~ JJ.eea rn~e.4e 74J749 434:-8-7 9e7 .Ja eS722i ~ Ja.e7a SJ~ .2S4 eGrn24 ~ 7~i.QQJ a247JS ~ J7.eSa aeG.S92 4J2S94 ~ 4~9.77i J44SJi ~ J9.e9a 279.J42 2477S9 44(h.7 rns.447 HQS7~ ~4~ .7Ga ~a7.Ge9 ~229Ga ~ n.4~4a aaSaG.e J4.&.-d+ 4a.aGJ4 JJSae.7 ~ Je.9~9s 2SQQ~ .a
© Copyright 2018 by NuScale Power, LLC 154
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Time (s)<1 1 Mass Release (lbm/s) Enerav Release (Btu/s)
~ G.nHgQ 27Q.4J2 ~ i .ea22 gJg7.4 eaG.G7a i .JJ4J7 i aQJ.4 i ~ i .Q24Q7 i227 .47 87G.Hla ii .8632 ~aarn .7 97G.2Ja i a.a4e2 ~
44-W:4 Q.GeiQg7i 88 .d8i8 i JgQ _Qa Q.Hl4gQi 24g _Je4
~ Q_gg4gij7 §Qgg _77 4-794:-2 ~ eaG .sQs ~ G.eanai aG2J .J4 2Gi4 .7a G.G2aQJa2 42 .i4ia ~ G.JQGe2a aG7 .ng ~ iG7.Jgg iJG4Ga 22GG.Ga G.4a4i2e aa8 .a24 22go.2a G.iai9J2 207 .iaa 24GG.aa G.JGia9a JeQ .278 ~ o.2Jrngg 27a .ae2 244G.ea 4e_gg4g a7a47.2 ~ Q.Hl7Qi2 22s .7ea ~ Q.2eJGJi J4J .QQ4 284i .ea G.ieeJG4 2Hl .QQ7 JG42 .ia G.i4Jaig rng _§Jg J242 .ea G.i2922a i 9Q _g§4 JaOO o.iHl7aa J4gJ .2a 0.00 0.00 0.00 1.00 520.85 257627.45 2.00 523.1 7 258750 .86 3.00 521 .93 258164 .72 4.00 519.26 256962 .72 5.00 515.99 255607 .71 6.00 512.44 254280 .63 7.00 508.73 253062 .33 8 .00 504.94 251984 .33 9.00 500.89 250943 .23 10.00 497.07 250168 .15 11.00 492.84 249319.52 12.00 488.50 248522.89 13.00 483.98 247730.73 14.00 479.49 247027 .59
© Copyright 2018 by NuScale Power, LLC 155
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 15.00 475.20 246477 .05 16.00 470.79 245889 .66 17.00 466.32 245272 .77 18.00 461 .90 244650 .14 19.00 457.90 244207 .37 20 .00 454.78 244180.29 40.00 412.85 228110.54 60.00 378.12 216874 .65 61.00 377.02 215906 .03 62.00 375.81 215107.74 63.00 374.48 214449 .07 64.00 373.13 213733.48 65.00 371.90 212887.33 66.00 370.89 212000 .58 67.00 369.82 211170.35 68.00 368.63 210506 .14 69.00 367.35 209910 .68 70 .00 366.22 209185 .02 71 .00 365.31 208336 .35 72.00 364 .44 207525 .80 73.00 363.47 206812 .03 74.00 362.44 206204 .10 75.00 361.43 205589 .78 76.00 360.48 204939 .76 77.00 1105.21 1079703.71 78.00 1263.49 1270075.68 79.00 1169.96 1206666.55 80.00 976.53 1004884.36 81 .00 838.30 909863 .68 82 .00 735.11 697397 .33 83 .00 2156.32 1334963.85 84.00 1611.53 1038460.23 85.00 1230.48 794378 .32 86.00 502.46 523334 .05 87.00 1515.48 927023.91 88 .00 231 .91 254414 .90 89.00 155.94 177983.24 90.00 209.84 172368.89 91 .00 161 .79 130981 .20 92.00 144.75 115382.50 93.00 139.61 110392.72 94.00 136.23 106487.96 95.00 132.60 102460.47 © Copyright 2018 by NuScale Power, LLC 156
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 96.00 128.69 98461 .93 97.00 126.62 96438.44 98.00 123.56 93419.72 99.00 120.86 90594 .20 100.00 118.67 88154.41 101.00 116.65 86096.24 102.00 115.36 84829.11 103.00 114.12 83538.32 104.00 112.74 82081.88 105.00 111.91 81198.07 106.00 111.10 80460 .96 107.00 110.12 79604 .71 108.00 108.94 78503.95 109.00 107.37 77011.06 110.00 106.08 75843.21 111.00 105.1 4 75092.50 112.00 104.38 74545.22 113.00 104.15 74478.90 114.00 103.05 73560.53 115.00 102.05 72792.56 116.00 101 .1 9 72219.92 117.00 100.41 71743.60 118.00 99 .60 71236.55 119.00 98.66 70603.35 120.00 97.70 70001.60 140.00 79.37 58965.27 160.00 67 .06 51663.90 180.00 58.34 46108.22 200.00 51 .94 41845.43 220.00 47.36 38792 .00 240.00 43.26 360 13.35 260.00 37.48 32408.39 280.00 35 .65 30650 .56 300.00 33 .86 29094.61 320.00 32.43 27834.92 340.00 30.90 26514.72 360.00 29.81 25566.75 380.00 28.61 24579.84 400.00 27.29 23548.94 420.00 26 .1 5 22680.53 440.00 25.18 21945.98 460.00 24 .11 21085.17 480.00 23.28 20464.64 © Copyright 2018 by NuScale Power, LLC 157
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 500.00 22.39 19824.76 520.00 21.45 19148.74 540.00 20.52 18497.43 560.00 19.61 17896.37 580.00 18.64 17263.95 600.00 17.69 16652.23 620.00 16.71 16050.69 640.00 15.59 15391.62 660.00 14.30 14661.37 680 .00 12.98 13926.77 700.00 8.86 11910.40 720.00 7.42 11136.13 740.00 6.43 10578.58 760.00 5.72 10159.74 780.00 5.16 9842.23 800.00 4.63 9508.67 820.00 4.17 9219.58 840.00 3.77 8959.33 860.00 3.42 8737.91 880.00 3.09 8525.88 900.00 2.79 8327.20 920.00 2.56 8181 .55 940.00 2.29 8002.94 960.00 2.05 7836.06 980.00 1.77 7617.34 1000.00 1.63 7541 .69 1020.00 1.43 7411.43 1040.00 1.25 7284.12 1060.00 1.08 7165.44 1080.00 0.71 6883.31 1100.00 0.69 6923.14 1120.00 0.57 6829.92 1140.00 0.45 6740.14 1160.00 0.34 6655.13 1180.00 0.23 6569.41 1200.00 0.12 6484.89 1220.00 0.02 6404.77 1240.00 -0.07 6334.75 1260.00 -0.1 6 6270.01 1280.00 -0.23 6208.31 1300.00 -0.31 6149.76 1320.00 -0.37 6094.65 1340.00 -0.44 6042.12 © Copyright 2018 by NuScale Power, LLC 158
Containment Response Analysis Methodology Technical Report TR-051 6-49084-NP Draft Rev. 10 1360.00 -0.50 5993.93 1380.00 -0.55 5944.68 1400.00 -0.61 5895.22 1420.00 -0.66 5849.06 1440.00 -0.71 5805.89 1460.00 -0.75 5766.42 1480.00 -0.79 5728.05 1500.00 -0.83 5689.85 1520.00 -0.87 5652.53 1540.00 -0.91 5616.71 1560.00 -0.94 5581.06 1580.00 -0.97 5547.13 1600.00 -1 .01 5513.44 1620.00 -1 .04 5476.9 1 1640.00 -1 .07 5442.10 1660.00 -1 .10 5406.73 1680.00 -1.1 3 5375.81 1700.00 -1. 15 5347.79 1720.00 -1 .17 53 19.46 1740.00 -1.19 5292.06 1760.00 -1 .21 5265.43 1780.00 -1.23 5239.46 1787.10 -0.90 5371.96
1. RWRRV opens at WQ seconds.
Table 8-2 Limiting Secondary Break Peak PressureWall Temperature Case - Mass and Energy Release Time (s)U~> Mass Release (lbm/s} Energy Release (Btu/s) Q Q Q 4-:00a 479.rna a9~44~
~ 4rn.44g a~a40~ ~ ae7 .7Ga 4ae779 4-:@ d~4 .QQ~ 4ggggg &:OOa ~g7 _9Qij ae4na &-:00 ~4e.a97 a~arno ~ ~4g _13g~ ~QQ7ae ~ ~ae .a77 rngaaB ~ 7.oga rna .g~g ~oaQQa ~ ~44.~g~ ~gggg13 ~ ~~ g _gga ~aoao7 aa.Ha gg _rnag ~Qaijea
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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Time (s)<12 > Mass Release (lbm/s) Energy Release (Btu/s)
~ 5g_7g77 74755 .5 ~ a:u9gg 429Q1.4 47--0+ 1a.G77e rn9H .~ ~ e.9799~ 9G4e .27 e4:-0Qa e.2g9g4 g~~2 .n -7&:4-a 12.7J95 rn425 .a ~ 1J.2QH H11g .e ~ 21 _g1 ae 2122Q.1 rng _a15 e.a2eoa gJ94 _g7 ~ 4.47799 eQQJ _g9 130.425 11 .ee93 15241.7 444-,4g 1.94533 27rn .e9 ~ Q.421951 754 .1Q3 1e2.515 3.3e5ee 4553.49 ~ 3.9ee29 53rn .~2 rn4.e95 2.717Q3 31G7.e1 4-9&.+a ~ _35g~2 ~975.G4 2Q@ _gg5 1 .43lee 2075 .e ~ 3.43159 4e24 .47 221 .915 2.9792g 4Q39.31 239.97 2.20944 313Q.41 25G.G2 ~ .47a4a 2~ ~ a.27 201 .075 1.45141 2Q92 .53 272.13 1 .53G05 211~ .9 213 .rn5 o.lee597 133Q.24 ~ o.@91397 1rn3 _5g 3Ga.29a G.a3erne 9G2 .e7e ~ Q.419932 751 .eo1 327.4Ga Q.344~47 ea2.rng 331.4e Q.3115g9 @og _g24 341 .51 o.255509 534 .7Q7 359.5e5 0.227770 49e .~42 ~ Q.217111 5e9 .137 3g~ .e7a G.aa293 9G9 .Ga4 ~ Q.99W42 14e5.49 403 .715 o.1532G9 12go.2e 444.84 g_ggg42g 12Ge.54 425 .195 1.009~7 4a4-7 ~ g_gg7@2e ~ JGG.Ga
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Containment Response Analysis Methodology Technical Report TR-051 6-4 9084-NP Draft Rev. 10 Time (s)11a1 Mass Release (lbm/s) Enerav Release (Btu/s) 447 g_grne~ ~~Q5.Q5 45g _g55 G.~Grn7 4H.QQ~ 4-W:-44 Q.394~75 eea.475 4gg _rna G.~rn~~~ g.4g~g4
~ G.JQijeaa 55~ .Q4~
0.00 0.00 0.00 1.00 79.18 48243.13 2 .00 79.30 48314.98 3.00 79.57 48482.36 4 .00 79.92 48690.46 5.00 80.29 48914.73 6.00 80.64 49119.33 7.00 80.83 49230.89 8.00 80.91 49270.28 9.00 80.89 49252.24 10.00 80.81 49 187.48 11 .00 80.70 49080.36 12.00 80.59 48944.34 13.00 80.50 48782 .97 14.00 80.43 48601.19 15.00 80.38 48398.72 16.00 80.37 48197.13 17.00 80.45 48031 .03 18.00 80.63 47922 .32 19.00 80.88 47846.57 20.00 81 .14 47787.04 40.00 82.67 46764.30 60.00 81.49 45804.67 80.00 80.30 44989 .66 100.00 78.89 44200 .28 120.00 77.54 43524.76 140.00 76.25 42962.65 160.00 74.95 42446.83 180.00 73.72 41987.01 200.00 72 .58 41572.17 220.00 72.65 41690.47 240.00 71.49 41296.20 260.00 70.43 40984.25 280.00 69.50 40786.43 300.00 68.37 40548.94 320.00 66.60 39942.51 © Copyright 2018 by NuScale Power, LLC 161
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. rn 340.00 64.62 38980.23 360.00 64.08 38649.04 380.00 63.87 38472.63 400.00 63.87 38426.77 420.00 63.75 38342.67 440.00 63.54 38216.47 460.00 63.37 38113.27 480.00 63.22 38024.68 500.00 63.11 37955.82 520.00 63.02 37892.62 540.00 62.94 37835 .79 560.00 62.87 37788.37 580.00 62.82 37745.00 600.00 62.77 37705.34 620.00 62.70 37657.63 640.00 62.66 37616.46 660.00 62.61 37577.23 680.00 62.57 37538 .52 700.00 62.51 37491 .14 720.00 62.47 37449.59 740.00 62.43 37409.24 760 .00 62.38 37367.66 780.00 62.32 37313.38 800.00 62.23 37246.63 820.00 62.17 37187.16 840.00 62.12 37138 .68 860.00 62.10 37095.76 880.00 62.10 37068.80 900.00 62.08 37029 .61 920.00 62.03 36977.23 940.00 61 .95 36914 .79 950.00 61.85 36852 .69 951.00 6 1.84 36846.42 952.00 61.83 36840 .27 953.00 61 .82 36834 .32 954.00 61.81 36828.52 955.00 1352.36 1370453.47 956.00 1193.26 1214639.97 957.00 996.84 1039740.72 958.00 841 .22 878708.54 959.00 677.92 705524.91 960.00 553.42 574452 .22 961.00 488.95 500751 .61 © Copyright 2018 by NuScale Power, LLC 162
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 962.00 378.97 386018.43 963.00 286.76 292965.95 964.00 206.23 210777.85 965.00 146.88 149539.51 966.00 104.54 106021.53 967.00 77.48 80524.71 968.00 64.94 67469.99 969.00 56.93 58992 .81 970.00 51.97 53697 .29 971 .00 48.08 49749.34 972.00 45.23 46977.89 973.00 41 .99 43648.53 974.00 39.38 41252.21 975.00 38.24 40286.01 976.00 36.20 38304.67 977.00 34.52 36582.13 978.00 32.77 35013.81 979.00 32.66 35063.81 980.00 30.51 33275.33 1000.00 7.6 1 16951 .11 1020.00 4.87 14668.09 1040.00 3.50 13415.22 1060.00 2.36 12351 .95 1080.00 1.38 11456.21 1100.00 0.53 10619.46 1120.00 -0 .19 9953.30 1140.00 -0 .71 9465.05 1160.00 -1 .09 9093.51 1180.00 -1.49 8686.42 1200.00 -1.89 8306 .51 1220.00 -2 .31 7904 .29 1240.00 -2 .65 7592.29 1260.00 -2.97 7271.04 1280.00 -3. 19 7042.75 1300.00 -3.38 6834.40 1320.00 -3.48 6721 .77 1340.00 -3 .67 6547.47 1360.00 -3.78 6403.45 1380.00 -3 .90 6250.04 1400.00 -3 .96 6129.98 1420.00 -4 .00 6046.52 1440.00 -4 .11 5896.39 1460.00 -4.17 5796.09 © Copyright 2018 by NuScale Power, LLC 163
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 1480.00 -4.17 5752.75 1500.00 -4.25 5623.53 1520.00 -4.26 5552 .52 1540.00 -4.27 5462.45 1560.00 -4.30 5384 .27 1580.00 -4 .31 5317 .01 1600.00 -4 .31 5267.64 1620.00 -4 .34 5193.43 1640.00 -4 .24 5213.03 1660.00 -4.28 5114.53 1680.00 -4 .30 5051 .19 1700.00 -4 .27 5022.07 1720.00 -4.27 4961 .03 1740.00 -4.20 4963.80 1760.00 -4 .10 4979.97 1780.00 -3.99 4982.59 1800.00 -3.79 5020.53 1820.00 -3.51 5109.37 1840.00 -3.34 5139.58 1860.00 -3.25 5130.42 1880.00 -3.15 5149.35 1900.00 -3 .12 5103.51 1920 .00 -3 .11 5060.80 1940.00 -3 .10 5028.42 1960.00 -3 .08 4996.18 1980.00 -3.04 4966.92 2000.00 -3 .09 4867.16 2020.00 -3.38 4746.53 2040.00 -3.43 4697.04 2060.00 -3.29 4710 .26 2080.00 -3 .16 4724.90 2100.00 -3 .04 4760.70 2120.00 -2.93 4781 .73 2140.00 -2.86 4788.56 2160.00 -2 .82 4767.95 2180.00 -2 .78 4736.15 2200.00 -2.73 4727.76 2220.00 -2 .61 4763.77 2240.00 -2 .55 4748.07 2260.00 -2.51 4731 .84 2280.00 -2.46 4724.87 2300.00 -2.40 4733.99 2320.00 -2 .34 4744.78 © Copyright 2018 by NuScale Power, LLC 164
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 2340.00 -2.24 4783.38 2360.00 -2 .24 4743.72 2380.00 -2 .23 4713.57 2400.00 -2 .18 4719.62 2420.00 -2 .34 4626.02 2440.00 -2.26 4644.15 2460.00 -2 .05 4698.36 2472 .90 -2.26 4601 .93 ~1- Break initiated at O seconds. Table 8-3 Limiting Secondary Break Peak Pressure Mass and Energy Release Time {s}< 1> Mass Release {lbm/s} Energ)l Release {Btu/s} 0.00 0.00 0.00 1.00 429.13 529471 .38 2.00 333.15 413594.14 3.00 311 .07 393366 .92 4.00 290.94 371550.11 5.00 264.05 339779.42 6.00 238.67 309056.90 7.00 216.91 282123.38 8.00 192.24 250588.45 9.00 118.64 154655.48 10.00 118.04 154001 .57 15.00 173.00 218097.88 16.00 180.15 224434 .34 17.00 184.92 227523.77 18.00 187.88 229107.05 19.00 187.54 226782.47 20.00 185.84 223894.90 25.00 159.89 196109.64 30.00 131.22 164114.01 35.00 99.41 126511 .96 40.00 55.26 71015.79 45.00 34.88 44867.91 46.00 13.36 17175 .87 47.00 7.15 9186 .32 48.00 6.09 7825 .33 49.00 5.59 7178 .04 50.00 3.25 4170 .70 55.00 2.81 3593.22 60.00 2.54 3252 .16 © Copyright 2018 by NuScale Power, LLC 165
Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 Time (s} <1 > Mass Release (lbm/s} Energ~ Release (Btu/s} 80.00 0.73 943 .37 100.00 0.57 725.24 120.00 0.48 609.39 140.00 0.49 623 .51 160.00 0.35 444.30 180.00 2.14 2752 .23 196.55 4.19 5420 .36 Break initiated at O seconds. © Copyright 2018 by NuScale Power, LLC 166
Containment Response Analysis Methodology Technical Report TR-0516-4 9084-N P Draft Rev. 10 8.2.1 Listing of Passive Heat Sinks The containment vessel shell is the only passive heat sink cred ited in the containment response analysis methodology. 8.2.2 Modeling of Passive Heat Sinks Table 8-4 Passive heat sinks Exposed Shell Total Passive Heat Material Thickness, Group Surface Area Volume, Total Mass, Surface Sink (Vessel steel in by Thickness ft3 lbm Area, ft2 plate) Group, ft2 ((
}}2(a),(c) 8.2.3 Thickness Groups Table 8-5 Th ickness groups Material Group Thickness Range, Designation in SA-240 304L (Stainless Steel) 1 :{ }}2(a),(c)
SA-240 304L (Stainless Steel) ~ {
}}2(a ),(c)
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Containment Response Analysis Methodology Technical Report TR-0516-49084-NP Draft Rev. 10 A-965 FXM-19240 304L (Stainless Steel), { }}2(a),(c) A-508 Grade 3 (Carbon Steel) 8.2.4 Properties of Passive Heat Sink Materials Table 8-6 Physical properties of passive heat sink materials Material Density, lbm/ft3 Specific Heat, Thermal Conductivity, Btu/lbm-°F Btu/hr-ft-°F SA-240 304L (Sta inless 5o 1.12 0.1137 8.6 Steen
$A-508 Grade 3 (Carbon ~83.84 0.1067 23 .7 Steel)
SA-965 FXM-19 (Stainless 487 .296 ::l .1142 M
~
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ML18304A128

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