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LLC Submittal of NuScale Containment Leakage Integrity Assurance, TR-1116-51962, Revision 1
	ML19149A298
Person / Time
Site:	NuScale
Issue date:	05/28/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of New Reactors
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ML19149A297	List: ML19149A298;
References
	AF-0419-65334, LO-0419-65333 TR-1116-51962-NP, Rev. 1
	Download: ML19149A298 (91)
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LO-0419-65333 May 28, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of NuScale Containment Leakage Integrity Assurance, TR-1116-51962, Revision 1

REFERENCES:

1. Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Technical Reports Supporting the NuScale Design Certification Application (NRC Project No. 0769), dated December 2016 (ML17005A112)

2. NuScale Technical Report, NuScale Containment Leakage Integrity Assurance, TR-0116-51962, Revision 0, dated December 2016 (ML17005A134)

NuScale Power, LLC (NuScale) hereby submits Revision 1 of the NuScale Containment Leakage Integrity Assurance (TR-1116-51962) technical report. contains the proprietary version of the report entitled NuScale Containment Leakage Integrity Assurance. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 contains the nonproprietary version of the report entitled NuScale Containment Leakage Integrity Assurance.

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

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

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Omid Tabatabai, NRC, OWFN-8H12 : NuScale Containment Leakage Integrity Assurance, TR-1116-51962-P, Revision 1, proprietary version : NuScale Containment Leakage Integrity Assurance, TR-1116-51962-NP, Revision 1, nonproprietary version : Affidavit of Zackary W. Rad, AF-0419-65334 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0419-65333 :

NuScale Containment Leakage Integrity Assurance, TR-1116-51962-P, Revision 1, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0419-65333 :

NuScale Containment Leakage Integrity Assurance, TR-1116-51962-NP, Revision 1, nonproprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 NuScale Containment Leakage Integrity Assurance May 2019 Revision 1 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 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.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 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 otherwise 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.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 CONTENTS Abstract ....................................................................................................................................... 1 Executive Summary .................................................................................................................... 2 1.0 Introduction ..................................................................................................................... 4 1.1 Purpose ................................................................................................................. 4 1.2 Scope .................................................................................................................... 4 1.3 Background ........................................................................................................... 5 1.4 Containment Leakage Integrity Assurance ............................................................ 5 1.5 Regulatory Requirements ...................................................................................... 6 1.6 Abbreviations and Definitions ................................................................................ 8 2.0 Containment Leakage Integrity Assurance Overview ............................................... 10 2.1 NuScale Containment Vessel Structure .............................................................. 11 2.2 NuScale Containment Vessel Penetrations ......................................................... 12 2.3 Type B Testing ..................................................................................................... 15 2.4 Type C Testing ..................................................................................................... 15 2.5 Containment Overall Leakage Limits................................................................... 15 3.0 NuScale Containment System Design ........................................................................ 17 3.1 Containment Vessel Design ................................................................................ 18 3.2 Containment Penetrations ................................................................................... 19 3.2.1 Electrical Penetration Assemblies ....................................................................... 24 3.2.2 Emergency Core Cooling System Trip and Reset Body-to-Bonnet Seals ........... 27 3.2.3 Containment Vessel Flange ................................................................................ 28 3.3 Containment Isolation Valves .............................................................................. 29 3.3.1 Primary System Containment Isolation Valves .................................................... 30 3.3.2 Feedwater Isolation Valve ................................................................................... 31 3.3.3 Main Steam Isolation Valve and Bypass Valve .................................................... 32 4.0 Preservice Inspection and Testing .............................................................................. 33 4.1 Manufacturing Facility Testing and Inspection ..................................................... 33 4.2 Post-installation Testing and Inspection .............................................................. 33 4.3 Shipping and Receiving Controls ........................................................................ 34 5.0 Inservice Inspection and Inservice Testing ................................................................ 35 5.1 Inservice Inspection of the Containment System ................................................ 35

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.1.1 Inspection Elements ............................................................................................ 35 5.1.2 Weld Inspection ................................................................................................... 37 5.1.3 Bolted Flange Pressure Testing........................................................................... 37 5.1.4 Visual Inspections................................................................................................ 38 5.1.5 Steam Generator Inspections and Controls ........................................................ 38 5.1.6 Type B Testing ..................................................................................................... 38 5.2 Inservice Testing of the Containment System ..................................................... 40 5.3 Type B Local Leak Rate Testing .......................................................................... 40 5.3.1 Type B Test Method ............................................................................................. 40 5.3.2 Electrical Penetration Assemblies ....................................................................... 41 5.3.3 Ports and Manways ............................................................................................. 41 5.3.4 Emergency Core Cooling System Pilot Valve Bodies .......................................... 41 5.3.5 Containment Vessel Flange ................................................................................ 41 5.3.6 Bolting ................................................................................................................. 42 5.4 Type C Local Leak Rate Testing .......................................................................... 42 5.4.1 Type C Test Method............................................................................................. 42 5.4.2 Test Considerations ............................................................................................. 43 5.4.3 Primary System Containment Isolation Valves .................................................... 43 5.4.4 Secondary System Containment Isolation Valves ............................................... 44 5.5 Containment Leakage Rate Test Program .......................................................... 44 5.5.1 Containment Leakage Limits ............................................................................... 45 5.5.2 Test Frequency .................................................................................................... 45 5.5.3 Test Results and Reporting Requirements .......................................................... 45 5.5.4 Special Testing Requirements ............................................................................. 46 5.5.5 Multiple NuScale Power Module Testing ............................................................. 46 5.6 Type A Testing Challenges .................................................................................. 47 5.6.1 Temperature ........................................................................................................ 47 5.6.2 Temperature Change ........................................................................................... 48 5.6.3 Instrumentation .................................................................................................... 49 5.6.4 Allowable Leak Rate ............................................................................................ 50 5.6.5 Alternate Testing Arrangement Considered ......................................................... 50 6.0 Material Selection and Aging Degradation Leakage Rate Test Program ................. 52 6.1 Material Selection and Operating Conditions ...................................................... 52

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 6.1.1 Pool Water Chemistry .......................................................................................... 52 6.1.2 Reactor Coolant System Coolant Chemistry ....................................................... 53 6.2 Aging Degradation Assessment .......................................................................... 55 6.2.1 Fatigue ................................................................................................................ 55 6.2.2 Boric Acid Corrosion ............................................................................................ 56 6.2.3 Primary Water Stress Corrosion-Cracking .......................................................... 56 6.2.4 Stress Corrosion-Cracking of Austenitic Stainless Steels ................................... 57 6.2.5 Stress Corrosion-Cracking of Pressure-Retaining Bolting Materials ................... 58 6.2.6 Irradiation Embrittlement of Lower Containment Vessel...................................... 59 7.0 References ..................................................................................................................... 62 Appendix A. Containment Isolation Summary Figures ....................................................... 65 TABLES Table 1-1 Containment leakage integrity program elements ................................................. 4 Table 1-2 Abbreviations ......................................................................................................... 8 Table 1-3 Definitions .............................................................................................................. 9 Table 2-1 NuScale containment leak rate test comparison ................................................. 14 Table 2-2 Maximum allowable containment leakage rate limits .......................................... 16 Table 3-1 CNV bolted flange calculation applied preloads .................................................. 23 Table 5-1 ASME Section XI inspection comparison ............................................................ 36 Table 5-2 Summary of test and inspection elements........................................................... 39 Table 6-1 Containment vessel pressure-retaining materials ............................................... 54 Table 6-2 Target limits and monitoring frequency for pool water ......................................... 55 Table 6-3 Reactor coolant system coolant chemistry .......................................................... 55 Table A-1 Table of figuressimplified figures based on the following piping and instrumentation diagrams .................................................................................... 65 FIGURES Figure 3-1 Containment vessel head .................................................................................... 17 Figure 3-2 Containment vessel ............................................................................................. 18 Figure 3-3 Upper containment assembly bolted flange openings ......................................... 20 Figure 3-4 Containment vessel head/port flange (typical) .................................................... 21 Figure 3-5 Electrical penetration assembly modules (typical)............................................... 25 Figure 3-6 Electrical penetration assembly test ports (typical) ............................................. 26 Figure 3-7 Emergency core cooling system trip and reset valve assembly .......................... 28 Figure 3-8 Containment vessel flange seal and leak port ..................................................... 29 Figure 3-9 Primary system containment isolation valve dual-valve, single-body design ...... 31 Figure 5-1 Primary system containment isolation valves Type C testing .............................. 44 Figure 5-2 Reactor pressure vessel, containment vessel, ultimate heat sink temperature gradients .............................................................................................................. 48 Figure 6-1 SA-965 Grade FXM-19 forgings in lower containment vessel ............................. 61

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure A-1 Containment system ............................................................................................ 66 Figure A-2 Steam generator system ..................................................................................... 67 Figure A-3 Decay heat removal system ................................................................................ 68

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Abstract This technical report describes the NuScale Power, LLC (NuScale) Containment Leakage Integrity Program (CLIP). This program provides assurance that leakage integrity of containment is maintained and that containment leakage does not exceed allowable leakage rate values. The CLIP is a consolidation of programs described in the NuScale Design Certification Application (DCA). All CLIP elements are implemented under other programs as described in this report and the NuScale DCA. The requirements of 10 CFR 50, Appendix A, General Design Criterion 52 (GDC 52) state that containments shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure. The requirements of 10 CFR 50, Appendix J, Type A tests, include test specifications directly related to GDC 52 design requirements. The CLIP integrates:

containment vessel flange design that remains sealed at design pressure

preservice leak test at design pressure performed for all containment vessels

initial (first-of-a-kind) containment vessel preservice leak test at design pressure performed with the vessel fully assembled with all flanges in place

preservice 10 CFR 50, Appendix J, Type B testing

preservice 10 CFR 50, Appendix J, Type C testing

post-installation and repair inspection and testing

inservice inspection and examination

periodic 10 CFR 50, Appendix J, Type B testing

periodic 10 CFR 50, Appendix J, Type C testing This report provides relevant details of the NuScale containment vessel and containment systems designs, which support the CLIP in assuring containment leakage integrity. The NuScale CLIP provides leakage integrity assurance equivalent to the containment leakage testing requirements of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. This report provides supplemental information designed to inform the NRCs evaluation of NuScale Final Safety Analysis Report Section 6.2.6 and DCA Part 7, Section 7, GDC 52 exemption request.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Executive Summary This technical report (TR) describes NuScales Containment Leakage Integrity Program (CLIP).

The CLIP, supported by the NuScale containment vessel (CNV) and containment system (CNTS) design, provides leakage integrity assurance for the NuScale containment. As discussed in the NuScale Design Certification Application (DCA), Part 7, Exemption Requests, NuScale is requesting an exemption from the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 52 and 10 CFR 50, Appendix J, which specify the design for and performance of preoperational and periodic integrated leak rate testing at containment design pressure.

The CLIP, supported by the design and analysis of the NuScale CNV and CNTS, provides leakage integrity assurance for the NuScale containment. The CLIP is a consolidation of programs described in the NuScale DCA. All CLIP elements are implemented under other programs as described in this report and the NuScale DCA. Each element of NuScales CLIP is consistent with a corresponding element of an approved program for reactor pressure vessels (RPVs) or large light water reactor (LLWR) containments. The primary CLIP elements that provide leakage integrity assurance include:

CNV flanges are designed to remain sealed at design pressure

factory inspection and testing, including preservice leak testing at design pressure with zero visible leakage, to ensure initial containment leakage integrity in accordance with an ITAAC

preservice and periodic Type B and C testing to ensure that overall containment leakage does not exceed allowable leakage rate values (i.e., quantifies overall containment leak rates)

ASME Section III, Class 1, design, construction, inspection, examination, and testing, and ASME Section OM and XI [(inservice testing (IST) and inservice inspection (ISI), repair and replacement, scheduled examinations, non-destructive examination methods, and flaw size characterization, including post-maintenance inspection, examination, and testing for CNV repairs or modifications)] to ensure continued leakage integrity (i.e., ensures that no unknown leak pathways develop over time)

Type B and C testing, inspections, and administrative controls (e.g., configuration management and procedural requirements for system restoration) to ensure leakage integrity associated with activity-based failure mechanisms [i.e., ensures that CNV flanges and containment isolation valves (CIVs) remain within allowable leakage rate values after system and component modifications or maintenance]

While the CLIP described in this report does not conform to GDC 52 and Type A testing requirements, the advanced NuScale design and CLIP provide more complete leakage integrity assurance than was considered when the subject regulations were adopted. This report provides a detailed overview of the key aspects of the testing, inspection, and design that ensures NuScales containment leakage integrity is maintained, including:

the overall containment leakage rate testing program, including the scope of the Type B and C testing to ensure adequate margin against design-basis leak rates

Type B testing adequacy is assured by:

CNV flanges are designed to remain sealed at design pressure

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 preservice design pressure leakage test of the CNV with CNV bolted flanges in place utilizing as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals installed to demonstrate no leakage at design pressure. The test is performed at design pressure and a minimum temperature of 70 degrees F and a maximum temperature of 140 degrees F the upper and lower halves of the CNV are assembled for the first module of the initial NuScale plant after successful testing, the upper and lower halves of all other CNVs may be tested separately covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design flange assembly utilizes positive verification to ensure proper flange loading from each stud The test configuration may utilize blanked off pipe ends in place of the containment isolation valves The acceptance criterion is no observed leakage from seals at examination pressure The ECCS trip valve and reset valve body-to-bonnet joint seals are not considered to be a flanged connection and are not included in the containment flange bolting calculation or preservice design pressure leakage test

the CNTS design as it applies to the containment function

the ISI program as it applies to the CNV

the IST program as it applies to CIVs

materials selection and aging degradation assessment As described in this report, the NuScale containment design and CLIP ensure that leakage integrity of containment is maintained and that containment leakage does not exceed allowable leakage rate values. This report provides supplemental information designed to inform the NRCs evaluation of NuScale Final Safety Analysis Report (FSAR) Section 6.2.6 and DCA Part 7, Section 7, GDC 52 exemption request.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 1.0 Introduction 1.1 Purpose The purpose of this technical report is to describe NuScales CLIP as well as the CNV and CNTS design elements that ensure leakage integrity. This report evaluates the NuScale plant design and CLIP against the requirements in 10 CFR 50, Appendix J (Reference 7.1.3) as incorporated in Design-Specific Review Standard (DSRS), Section 6.2.6 (Reference 7.1.6). This evaluation includes an assessment of the capability of the NuScale containment design to meet specific testing requirements in 10 CFR 50, Appendix J. This report identifies Type A requirements that will not be applied because the fundamental functionality is achieved differently. This report describes NuScales approach to Type B and Type C testing through an evaluation of the containment design.

This report provides supplemental information designed to inform the NRCs evaluation of NuScale FSAR Section 6.2.6 and DCA Part 7, Section 7, GDC 52 exemption request.

As shown in the table below, each element of NuScales CLIP is consistent with a corresponding element of an approved program for reactor pressure vessels (RPVs) or LLWR containments, which have been incorporated within the NuScale DCA. This report provides a consolidated description of inspection, testing, and examination elements from several programs described in the NuScale DCA related to containment leakage integrity. This report does not describe any elements that are not described in the NuScale DCA.

Table 1-1 Containment leakage integrity program elements CLIP Element NuScale DCA Requirement CNV flange design FSAR COL Item 6.2-2 Preservice inspection (TR Section 4) ASME III (FSAR 6.2)

Fabrication structural integrity testing (TR Section 4) ASME III (FSAR 6.2)

Preservice leakage testing FSAR 6.2.6, DCA Tier 1, (ITAAC)

Preservice Type B and C local leakage rate test (LLRT) Technical Specifications (TS) (DCA Part (TR Section 4) 4, Section 5.5.9)

Preservice Type B and C LLRT (TR Section 4) Initial Test Program (FSAR Table 14.2-43)

Post-installation/repair inspection & testing (TR Section 5) ASME III / XI (FSAR 6.2)

Post-installation/repair inspection & testing (TR Section 5) TS (DCA Part 4, Section 5.5.9)

Inservice inspection and examination (TR Section 5) ASME XI (FSAR 6.2)

Periodic Type B and C LLRT (TR Section 5) TS (DCA Part 4, Section 5.5.9) 1.2 Scope This report describes the CLIP for the NuScale design and evaluates the NuScale CLIP against 10 CFR 50, Appendix J. This report describes

the overall containment leakage rate testing (CLRT) program, including the scope and frequency of Type B and C testing of CNV penetrations.

the CNTS design as it applies to CNV design and the containment function.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1

materials selection and aging degradation as it applies to the containment pressure boundary

the ISI program as it applies to the CNV

the IST program as it applies to containment isolation valves (CIVs)

Type A integrated leak rate testing impediments 1.3 Background Pursuant to 10 CFR 52.7, NuScale is requesting an exemption from GDC 52. 10 CFR 50, Appendix J specifies Type A testing directly related to GDC 52. While Appendix J is not applicable to a design certification applicant, NuScale is also planning to request that the approval of the GDC 52 exemption within the NuScale Power Plant design certification include exemption from the requirements of 10 CFR 50, Appendix J Type A testing for plants referencing the NuScale design certification.

This technical report describes the NuScale containment testing, inspection, and design criteria that ensure leakage integrity of containment is maintained and that containment leakage does not exceed allowable leakage rate values.

1.4 Containment Leakage Integrity Assurance The NuScale CLIP provides containment leakage integrity by

demonstrating that the NuScale containment design can use LLRT to adequately ensure containment leakage integrity CNV flanges are designed to remain sealed at design pressure preservice design pressure leakage test of the CNV with CNV bolted flanges in place utilizing as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals installed to demonstrate no leakage at design pressure the upper and lower halves of the CNV are assembled for the first module of the initial NuScale plant after successful testing, the upper and lower halves of all other CNVs may be tested separately covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design flange assembly utilizes positive verification to ensure proper flange loading from each stud.

ensuring no unknown leakage pathways exist.

quantifying overall containment leak rates by LLRTs that provide accurate results for every potential leak path.

ensuring no unknown leak paths develop over time due to degradation.

ensuring no unknown leak paths develop due to activity-based failure mechanisms.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 1.5 Regulatory Requirements 10 CFR 52.47(a) states in part:

The application must contain a final safety analysis report (FSAR) that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and must include the following information:

(3) The design of the facility including:

(i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units; (ii) The design bases and the relation of the design bases to the principal design criteria The introduction to 10 CFR 50, Appendix A, states in part:

Also, there may be water-cooled nuclear power units for which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General Design Criteria must be identified and justified.

10 CFR 50, Appendix A, GDC 52 states:

Criterion 52 - Capability for containment leakage rate testing. The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

10 CFR 50.54(o) states in part:

Primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in appendix J to this part.

Appendix J to 10 CFR 50, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, states, in part:

One of the conditions of all operating licenses under this part and combined licenses under part 52 of this chapter for water-cooled power reactors as specified in § 50.54(o) is that primary reactor containments shall meet the containment leakage test requirements set forth in this appendix. These test requirements provide for preoperational and periodic verification by tests of the leak-tight integrity of the primary reactor containment, and systems and

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 components which penetrate containment of water-cooled power reactors, and establish the acceptance criteria for these tests. The purposes of the tests are to assure that (a) leakage through the primary reactor containment and systems and components penetrating primary containment shall not exceed allowable leakage rate values as specified in the technical specifications or associated bases; and (b) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment. These test requirements may also be used for guidance in establishing appropriate containment leakage test requirements in technical specifications or associated bases for other types of nuclear power reactors.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 1.6 Abbreviations and Definitions Table 1-2 Abbreviations Term Definition ASME American Society of Mechanical Engineers CES containment evacuation system CIV containment isolation valve CLIP Containment Leakage Integrity Program CLRT containment leakage rate testing CNTS containment system CNV containment vessel COL combined license CRDM control rod drive mechanism DCA design certification application DHRS decay heat removal system DSRS Design-Specific Review Standard ECCS emergency core cooling system EFPY effective full-power year EPA electrical penetration assembly FSAR Final Safety Analysis Report FWIV feedwater isolation valve GDC general design criteria I&C instrumentation and controls ILRT integrated leak rate test ISI inservice inspection IST inservice testing LLRT local leakage rate test LLWR large light water reactor MSIV main steam isolation valve NPM NuScale Power Module NPS nominal pipe size (ASME B36.10M)

PSCIV primary system containment isolation valve PWR pressurized water reactor PWSCC primary water stress corrosion-cracking RCPB reactor coolant pressure boundary RCS reactor coolant system RPV reactor pressure vessel RVV reactor vent valve SCC stress corrosion-cracking SCFH standard cubic feet per hour

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Term Definition SG steam generator SGS steam generator system SSCIV secondary system containment isolation valve TGSCC transgranular stress corrosion-cracking TR technical report TS technical specification UHS ultimate heat sink Table 1-3 Definitions Term Definition GDC 55 penetration A piping system (line) that is part of the reactor coolant pressure boundary (RCPB) and penetrates reactor containment. This type of penetration requires two, NRC Quality Group A, ASME Code Class 1, CIVs at each penetration. (References 7.1.1 and 7.1.2)

GDC 56 penetration A piping system (line) that connects directly to the containment atmosphere and penetrates the reactor containment. This type of penetration requires two NRC Quality Group B, ASME Code Class 2, CIVs at each penetration.

(References 7.1.2 and 7.1.9)

GDC 57 penetration A piping system (line) that penetrates reactor containment and is neither part of the RCPB nor connected directly to the containment atmosphere. Also known as a closed system. This type of penetration requires one NRC Quality Group B, ASME Code Class 2, CIV at each penetration.

(References 7.1.2 and 7.1.9)

La Maximum allowable containment leakage rate at pressure Pa Pa Peak CNV accident pressure

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 2.0 Containment Leakage Integrity Assurance Overview NuScale CLIP testing, inspection, and examination, supported by the design and analysis of the NuScale CNV and CNTS, ensure leakage integrity is maintained for the NuScale containment. The CLRT, in combination with other CLIP elements, verifies the leakage integrity of the reactor containment by testing that the actual containment leakage rates do not exceed the values assumed in the applicable safety analysis calculations for design basis events. The preoperational and periodic CLRT requirements and acceptance criteria that demonstrate leakage integrity of the CNTS and associated components are performed in accordance with 10 CFR 50, Appendix J and implemented through the licensees CLRT program described in Section 5.5.9 of the technical specifications, Part 4 of the NuScale DCA. The maximum allowable containment leakage rate is referred to as La, and this leakage rate is measured at peak containment accident pressure (Pa) (these terms are defined in 10 CFR 50, Appendix J).

The containment penetrations and containment isolation barriers are designed to permit the periodic leakage testing described in GDC 53 and 54 to verify leakage through the containment penetrations does not exceed the allowable leakage rate.

The design of the containment penetrations support performance of Type B and Type C testing in accordance with the guidance provided in Regulatory Guide 1.163 (Reference 7.1.4), ANSI/ANS 56.8 (Reference 7.1.10) and NEI 94-01 (Reference 7.1.12). The NuScale CNTS design accommodates both test method frequencies permitted by 10 CFR 50, Appendix J; Option A, Prescriptive Requirements and Option B, Performance-Based Requirements. Only Option A will be available to initial NuScale licensed plants, as there will not be sufficient performance history to use Option B. Initial COL applicants that reference the NuScale Power Plant design certification will develop a CLRT program which will identify Option A to be implemented under 10 CFR 50, Appendix J.

The NuScale containment is designed for all flanged joints to remain sealed at design pressure. The containment is initially inspected and tested at the factory, including ASME hydrostatic testing with an acceptance criterion of zero leakage, to verify that no unknown leak pathways exist. Additionally, a CNV preservice design pressure leakage test is performed that loads CNV bolted flange connections to containment design pressure and confirms no observed leakage under these conditions. Because all potential leakage pathways are known and testable, preservice and periodic Type B and C testing quantify the overall containment leakage rate to verify that maximum allowable leakage is not exceeded (i.e., the design and configuration of all potential leak pathways, including CNV flanges and CIVs, provide for LLRT results to meet containment integrated leakage rate acceptance criteria). Periodic inspection and testing verifies that no unknown leakage pathways develop over time (i.e., any potential through-wall degradation will be precluded as a credible mechanism for containment leakage).

Post-maintenance inspection and testing, including Type B and C testing and administrative controls, verify that no unknown leakage pathways develop due to activity-based failure mechanisms during maintenance or modifications.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 2.1 NuScale Containment Vessel Structure The NuScale CNV design ensures leakage integrity through design, inspection and testing other than as required by GDC 52 and Appendix J. NEI 94-01 (Reference 7.1.12) describes the purpose of 10 CFR 50, Appendix J for traditional large containment structures:

The purpose of Type A testing is to verify the leakage integrity of the containment structure. The primary performance objective of the Type A test is not to quantify an overall containment system leakage rate. The Type A testing methodology as described in ANSI/ANS-56.8-2002 serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

Continued leakage integrity of the NuScale CNV structure is ensured by precluding through-wall degradation as a credible leakage mechanism. The NuScale CNV is a welded metal vessel design, in contrast to existing pressurized water reactors (PWRs) that incorporate large containment building structures. The containment is designed for all flanged joints to remain sealed at design pressure. Manufacturing acceptance tests and inspections are similar to RPV tests and inspections, and are performed in a factory environment. Comprehensive ISI applying ASME Code Class 1 criteria also ensures no new leakage paths develop over the life of the plant due to degradation. All surface areas and welds are accessible for inspection. Additionally, a separate preservice design pressure leakage test is required for all containment vessels with CNV bolted flange connections in place to demonstrate no observed leakage utilizing as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals installed. This leakage test is required by an ITAAC. The first CNV of the initial plant shall be tested with the upper and lower halves of the containment vessel assembled. Penetration pathways are tested to Type B or C criteria at peak containment accident pressure. These features ensure that continued leakage integrity of the CNTS is maintained without the need for Type A testing.

The NuScale CNV design is different from traditional containments in several fundamental aspects. These design differences impact conformance with GDC 52 and Appendix J, and provide alternative means of assuring the leakage integrity of the NuScale containment. The major containment functional differences are:

The CNV is a high-pressure vessel with no internal subcompartments, an ASME Code Class MC component, constructed to ASME Code Class 1 vessel rules, constructed of all stainless steel clad or stainless materials.

All penetrations are either ASME Code Class 1 flanged joints capable of Type B testing or ASME Code Class 1 welded nozzles with isolation valves capable of Type C testing, or form part of a closed system inside containment.

All flanged joints are designed to remain in contact at accident temperature, concurrent with peak accident pressure.

During refueling, the reactor module, including the CNTS is physically moved by a crane to the refueling area. The upper and lower CNV shells are separated during outages for refueling, maintenance, and inspection. The CNV is designed to

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 accommodate comprehensive inspections of welds, including volumetric and surface inspections. All welds are accessible, and there are no areas that cannot be inspected. The CNV design allows for visual inspection of the entire inner and outer surfaces. Through-wall degradation can be identified prior to development of potential leak paths precluding this as a credible leakage mechanism.

During reassembly, positive verification is utilized to verify proper stud elongation to ensure proper loading on each flange seal.

During normal operation, the CNV is under a vacuum and is partially submerged in borated water. Automatic engineered safety feature actuation systems initiate on high containment pressure with the CNV still at partial vacuum conditions.

Containment vacuum pressure and leak rate into the CNV is constantly monitored during normal operation. The small containment volume and evacuated operating conditions allows wide-ranging detection capabilities for liquid or vapor in-leakage, providing an additional layer of leakage integrity assurance.

The NuScale CNV design is described in detail in Section 3.0.

2.2 NuScale Containment Vessel Penetrations The NuScale CNTS design supports leakage integrity assurance through inspection and testing other than as required by GDC 52 and Appendix J. When compared to traditional LLWR containments, the NuScale CNTS design is simple. The CNV has a low number of penetrations (40), all of which are either ASME Class 1 flanged joints capable of Type B testing, ASME Class 1 welded nozzles with isolation valves capable of Type C testing, or form part of a closed system inside containment [i.e. steam generator system (SGS) piping]. The CNV has no penetrations equipped with resilient seals. No instrument lines penetrate containment; therefore, there are no small diameter fluid lines without isolation capability that are not subject to Type B or C LLRT. There are no air locks, flexible sleeves, or nonmetallic boundaries. This simplicity of design provides for alternate means of assuring containment leakage integrity. This is primarily achieved by ensuring no unknown leak paths by ISI and accurate leakage rate measurements of all potential leak pathways by LLRT. Key features which ensure NuScale CNTS leakage integrity is maintained include:

CNV flanges are designed to remain in contact at accident temperature, concurrent with peak accident pressure.

As described in Section 2.1, the CNV is an ASME Code Class 1 pressure vessel with a relatively low volume and no internal subcompartments. This comparatively simple design (compared to existing LLWR designs) allows for identification of all potential leakage pathways.

The CNV pressure vessel preservice test and inspections are equivalent to RPV requirements, including hydrostatic testing requirements. This verifies that no unknown leakage pathways exist.

preservice design pressure leakage test of the CNV with CNV bolted flanges in place utilizing as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals installed to demonstrate no observed leakage at design pressure.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 the upper and lower halves of the CNV are assembled for the first module of the initial NuScale plant after successful testing, the upper and lower halves of all other CNVs may be tested separately covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design

The limited number of CNV penetrations have similar seal designs that are tested by Type B or Type C LLRT. This, and other aspects of the penetration design, allows accurate quantification of the overall leakage rate by LLRT.

The NuScale ISI program and planned CNV examinations will meet ASME Code Class 1 criteria. This ensures that no new unidentified leakage pathways develop over time.

Disassembly and reassembly procedures and controls of the CNV will be similar to the RPV. Positive verification is utilized to verify proper loading on each flange seal.

This ensures that these potential activity-based failure mechanisms do not degrade CNTS leakage integrity.

The CNV is an ASME Subsection NE, Class MC containment designed, fabricated, and stamped as an ASME Subsection NB, Class 1 pressure vessel, with overpressure protection provided in accordance with NE-7000. The CNV is made of corrosion-resistant materials, has a low number of penetrations, and no penetrations have resilient seals. The use of welded nozzles and testable flange seals at the containment penetrations ensure that Type B and C testing provide an accurate assessment of overall containment leakage rate.

The unique CNV and CNTS design allows testing and inspection options not suitable to current LLWR containment designs. Based on the containment vessel ASME pressure vessel design and its function, preferable methods of testing and inspection are available. Each element of NuScales CLIP is consistent with a corresponding element of an approved program for RPVs or LLWR containments.

The NuScale CNTS design is described in detail in Section 3.0. Table 2-1 compares elements of the NuScale CLIP with testing performed on the NuScale containment, RCPB, and traditional containments. The purpose of the table is to demonstrate that the testing is commensurate with the design and safety function of the NuScale containment.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Table 2-1 NuScale containment leak rate test comparison Reactor Coolant CLIP Program Pressure Boundary Element to Ensure Traditional NuScale Containment Testing for NuScale Essentially Leak- Containment and Other Licensed Tight Barrier Facilities Initial verification of Hydrostatic testing per Hydrostatic testing per Preservice ILRT structural integrity ASME Section III ASME Section III Factory - hydrostatic testing per ASME Section III Containment preservice Preservice ILRT Initial verification of leakage test (ITAAC) Hydrostatic testing per leakage integrity ASME Section III (leakage allowed below (no visible leakage prescribed limit) allowed)

On-site - preservice LLRT Administrative controls Prevention of Administrative controls Administrative controls such as configuration leakage from such as configuration such as configuration management and activity-based failure management and management and procedural mechanisms procedural requirements procedural requirements requirements for (degradation due to for system restoration for system restoration system restoration that system and/or that ensure that integrity that ensure that integrity ensure that integrity is component is not degraded by plant is not degraded by plant not degraded by plant modifications or modifications or modifications or modifications or maintenance) maintenance activities maintenance activities maintenance activities Detection of leakage RCS leak test -

from activity-based LLRT LLRT operational pressure failure mechanisms Design and construction Design and construction requirements, Prevention of requirements for CNV, inspections/

leakage from age- Design and construction inspections/ examinations based failure requirements for RCS, examinations performed performed in mechanisms inspections/

in accordance with accordance with ASME, (age-related examinations performed ASME, section XI, the section XI, the degradation) in accordance with maintenance rule and maintenance rule and ASME, section XI, the regulatory commitments regulatory maintenance rule and NuScale CNV design commitments regulatory commitments Detection of leakage allows for comprehensive RCS leakage detection from age-based ISI surface and weld failure mechanisms examination ILRT (age-related degradation)

Post-repair/ Hydrostatic testing per modification ASME Section XI Hydrostatic testing per ILRT/LLRT verification of ASME Section XI leakage integrity LLRT

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Reactor Coolant CLIP Program Pressure Boundary Element to Ensure Traditional NuScale Containment Testing for NuScale Essentially Leak- Containment and Other Licensed Tight Barrier Facilities Post-repair/

modification Hydrostatic testing per Hydrostatic testing per ILRT verification of ASME Section XI ASME Section XI structural integrity 2.3 Type B Testing Type B pneumatic tests (local penetration leak tests) detect and measure leakage across the pressure-retaining, leakage-limiting boundaries in the CNV. Preoperational and periodic Type B leakage rate testing is performed in accordance with 10 CFR 50, Appendix J, NEI 94-01, and ANSI-56.8 within the test intervals defined by the COL holder. The containment penetrations subject to Type B tests are identified in Appendix A.1. As described further in Section 3.0, the design of CNV penetrations allows accurate LLRT results to quantify overall containment penetration leak rates.

The design of NuScale CNV Type B penetrations is described in Section 3.2.

2.4 Type C Testing The CIVs are designed to support Type C pneumatic tests. Preoperational and periodic Type C leakage rate testing of CIVs is performed in accordance with the 10 CFR 50, Appendix J requirements, ANSI-56.8, and the COL holders technical specifications. The CIVs subject to Type C tests are identified in Section 3.3. As described further in Section 3.0, the design of CIVs allows accurate LLRT results to quantify overall CIV leak rates.

The design of NuScale CNV containment isolation valves is described in Section 3.3.

2.5 Containment Overall Leakage Limits Per 10 CFR 50, Appendix J, La is defined as the maximum allowable containment leakage rate in weight percent per day at peak containment accident pressure Pa. For NuScale, La has been selected to be 0.20 weight percent of the containment air mass per day at the peak containment accident pressure (Pa) provided in FSAR Tier 2, Section 6.2.1, over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . La is established as a safety analysis operational limit for the NuScale Power Plant design. The values are used in consequence calculations to confirm that accident radiological containment leakage to the environment is within acceptable limits.

A NuScale engineering evaluation of all containment penetrations and access flanges (all leakage pathways) determined that the design can reliably meet the 10 CFR 50, Appendix J leakage criteria using an La of 0.20 weight percent of containment air mass per day at design pressure. The engineering evaluation concluded that the combined maximum expected leakage from all local penetrations with conservative margin for degradation, is less than [0.60]La, which is the acceptance criterion for LLRT per 10 CFR

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 50 Appendix J. The 0.20 weight percent of containment air mass at 1000 psia per day was converted to 18.05 standard cubic feet per hour (SCFH) at 1000 psia. This value was rounded down to 17.5 SCFH at 1000 psia and used as a design parameter by consequence calculations. This is a conservatively bounding approach since 10 CFR 50 Appendix J requires La to be specified at Pa, which is less than 1000 psia.

Table 2-2 documents NuScale containment design basis leakage rate criteria. The CLRT leakage rate limits for LLRT will be developed from these design basis limits to meet 10 CFR 50, Appendix J leakage criteria.

Testing to meet [0.60]La at Pa ensures that the operational limit of 0.20 weight percent of containment air mass per day can be met. Reference 7.1.5, SRP Acceptance Criteria 2 states that to satisfy GDC 38 to rapidly reduce the containment pressure, the pressure should be reduced to less than 50 percent of the peak calculated pressure for the design basis loss-of-coolant accident within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the postulated accident. Following the peak containment pressure design basis accident, containment pressure drops from Pa to approximately 20 psia in four hours.

Table 2-2 Maximum allowable containment leakage rate limits Leakage Rate Pressure Notes Containment Leakage Rate Evaluation Parameters 0.20 weight percent of 1,000 psia Leakage criteria used in consequence analysis, containment air mass per rounded down from 18.05 SCFH.

day 17.5 SCFH air converted leakage rate LLRT limits will be developed based on the values of Table 2-2, and will be based on a La at Pa and to meet < (0.60La).

The peak containment accident pressure (Pa) is identified in FSAR Tier 2, Section 6.2.1.

The NuScale CLRT is described further in section 5.3.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 3.0 NuScale Containment System Design The NuScale CNTS is designed as an ASME Code Class 1 CNV pressure vessel. The simplicity of the NuScale Power Module (NPM) design minimizes the number of containment penetrations required. There are a limited number of ports (7), manways (2), emergency core cooling system (ECCS) pilot valve penetrations (6), and electrical penetration assemblies (EPAs) (11) that all use similar bolted closure double O-ring seal designs. The CNV closure flange separating the upper and lower CNV assemblies uses the same seal design as the RPV and is similar to the port and manway seal design.

There are a limited number of fluid lines penetrating containment (14 total) (see Figure 3-1 and Figure 3-2). Eight fluid line penetrations are protected by dual CIVs, four are protected by a closed-loop SGS and a single secondary system containment isolation valve (SSCIV), and two are protected by a closed-loop inside and outside containment

[SGS and decay heat removal system (DHRS)].

Figure 3-1 Containment vessel head

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 3.1 Containment Vessel Design Approximately 90 percent of the CNV is submerged in the ultimate heat sink (UHS) that removes residual core heat during normal and accident conditions. The CNV has a design pressure and temperature referenced in FSAR Tier 2, Section 6.2.1. The CNV is a steel vessel with relatively low volume (approximately 6,144 ft3) compared to other PWR containments and has no internal subcompartments. The design prevents isolated pockets of concentrated gases. The upper portion of the CNV is constructed of low alloy carbon steel with stainless steel cladding on the inside and outside surfaces. The bottom portion of the CNV is constructed of stainless steel. The CNV will be factory fabricated, which facilitates enhanced fabrication quality and testing control.

Figure 3-2 Containment vessel

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 All CNV nozzles and penetrations are required to be either forged or welded connections; bellows sealed connections (which are common for LLWR containment penetrations) are not used. There are 14 CNV piping penetrations, eight of which are two-inch nominal pipe size (NPS 2) pipe penetrations that require Type C testing. All are isolated with CIVs of identical design and construction. The other six penetrations are main steam, feedwater and DHRS condensate penetrations that are connected to the steam generator (SG), which are not required to be Type C tested in accordance with 10 CFR 50, Appendix J, II.H. There are 11 EPAs on the CNV (Appendix A.1). There are nine ports and manways on the CNV, and there are six ECCS pilot valve penetrations (Appendix A.1). All Type B penetrations, including the CNV closure flange, have a similar seal design; the only difference is the size and model of the O-rings. These penetrations will undergo periodic LLRTs. All penetrations are either ASME Code Class 1 flanged joints capable of Type B testing, ASME Code Class 1 welded nozzles with isolation valves capable of Type C testing, or form part of a closed system inside containment.

There are 40 total CNV penetrations. The CNV closure flange also requires Type B testing. These penetrations are described in Appendix A.

No instrument lines penetrate containment; therefore, there are no small diameter fluid lines without isolation capability that are not subject to Type B or C LLRT. There are no air locks, flexible sleeves, or nonmetallic boundaries. There are no penetrations in the NPM design that would only be tested in an ILRT. Entry into the CNV is precluded during normal operation by personnel safety constraints and most openings will be submerged in the reactor pool. The integrity of the Type B pathways is not expected to be disturbed except when the NPM is in a refueling outage or disassembled for emergent maintenance activities. All Type B and Type C pathways will be tested to CNTS accident peak pressure (Pa). All Type C pathways are designed such that an individual valve can be tested in the same direction in which the valve would perform its safety function.

3.2 Containment Penetrations The CNV is designed to support Type B local penetration pneumatic leak tests to detect and measure leakage across the pressure-retaining, leakage-limiting boundaries that include flange openings (bolted connections) and EPAs. The CNTS penetration designs allow accurate LLRT results used to quantify the overall containment penetration leak rate. The following containment penetrations are subject to preoperational and periodic Type B leakage rate testing:

flanged access openings with bolted connections

EPAs

ECCS trip and reset valve body-to-bonnet seals

CNV closure flange All Type B penetrations are bolted closures that have dual metal O-ring seals with leak detection and testing ports between the seals. All Type B penetration assemblies are designed and constructed to ASME Code Class 1. The CNV closure flange has a similar double O-ring and test port arrangement. All CNV flanges are designed to remain in contact at accident temperature, concurrent with peak accident pressure. Figure 3-3 shows the location of the Type B bolted penetrations (CNV closure flange and ECCS pilot valves not shown).

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 3-3 Upper containment assembly bolted flange openings The nine access ports and manways are bolted closures that have dual metal O-ring seals with test ports between the seals. Figure 3-4 shows the CNV head and control rod drive mechanism (CRDM) access flange. This shows the double O-ring seal and test port arrangement. This seal and test port design is used for every Type B seal.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 3-4 Containment vessel head/port flange (typical)

Maintaining flange contact pressure is demonstrated by analysis showing that the flange surfaces of the bolted connection have no separation of the flanges from the inboard seal to the containment inner surface when peak accident pressure is applied to the inside of the containment vessel. This provides reasonable assurance that the seal would demonstrate similar flange gaps at peak accident pressure conditions as would be shown during the Type B test.

The following CNV bolted flanged openings are evaluated for maintaining contact pressure between the flanges from the inboard seal to the containment inner surface to show there is no separation of the flanges due to the inadvertent reactor recirculation valve (RRV) opening and chemical and volume control system (CVCS) injection line accident events:

Closure flange - 170 inch

Pressurizer (PZR) heater access ports - 44 inch (CNV31/32)

Steam generator (SG) inspection ports - 38 inch (CNV27/28/29/30)

Manway (shell) port - 38 inch (CNV26)

Control rod drive mechanism (CRDM) access opening - 67 inch (CNV25)

CRDM power penetration - 17 inch (CNV37)

Rod position indication (RPI) penetrations - 10 inch (CNV 38/39)

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1

Manway (head) - 18 inch (CNV24)

PZR power penetrations - 12 inch (CNV15/16)

Instrumentation & control (I&C) - channel A/B/C/D penetrations - 8 inch (CNV17/18/19/20)

I&C - division 1/2 penetrations - 3 inch (CNV8/9)

This evaluation ensures that the flanges maintain contact pressure and therefore do not pry apart while under peak accident pressure at a minimum. This provides reasonable assurance that the seal would demonstrate similar flange gaps at peak accident pressure as would be shown during the Type B test. Therefore, the leakage rate measured during the Type B test would be representative of leakage at peak accident pressure.

The seals for all of the CNV bolted flange connections are comprised of two concentric grooves with a seal positioned in each groove at each bolted flange connection. The seal material is a metal alloy and plated for improved sealing ability. The seal metal and plating are corrosion resistant based on exposure to the borated water chemistry. The seals are capable of a service temperature greater than the 550°F CNV design temperature. The preload applied to the bolted flange connections is sufficient to apply the seal seating load with sufficient preload remaining to maintain flange contact pressure at peak accident pressure conditions.

An analysis to demonstrate that the NuScale flange designs would maintain contact pressure, at peak bounding pressure, for assumed seal characteristics, was performed in the CNV bolted flange calculation. This calculation resulted in the stud preload values for each flange and these values are provided in Table 3-1 below. COL Item 6.2-2 ensures that the final design of the containment vessel, including the final seal design, meets the design basis requirement to maintain flange contact pressure at accident temperature concurrent with peak accident pressure.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 ((

                                                                 }}2(a),(c)

Table 3-1 CNV bolted flange calculation applied preloads ((

                                                                                                      }}2(a),(c)

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 The metal temperature of the CNV is taken from a heat transfer analysis of accident conditions at the time CNV peak pressure occurs. The accident event pressure transient is evaluated by analyzing a pressure equal to or greater than the peak pressure for that accident. The temperature transient inside of the CNV and the reactor pool outside of the CNV for the accident conditions are used as the bulk fluid temperature for inside and outside of the CNV respectively. The CNV metal surface temperatures are used to calculate the heat transfer coefficients on the CNV model. The differential temperature between the cooler pool side of the CNV and the inside surface of the CNV creates additional compression at the inside surface of a bolted flange due to thermal expansion of metal nearer the inside surface. The CNV is also evaluated at a uniform 140°F temperature with a pressure equal to or greater than the CNV design pressure. This evaluation simulates the expected conditions for the preservice design pressure leakage test. Evaluation at these pressure and temperature conditions provides assurance that leakage will not be detected during the preservice design pressure leakage test. Additionally, this evaluation demonstrates the preservice design pressure leakage test is evaluating the bolted flanges under a condition that bounds what would be seen at peak accident pressure conditions. The emergency core cooling system (ECCS) trip and reset pilot valve body-to-bonnet seals are tested and qualified per ASME QME-1 and designed per ASME B&PVC Section III, Subarticle NB-3500. These valves will be pressure tested and functionally qualified separately which will include a series of tests at reactor coolant system (RCS) operating pressure, which is above the containment design pressure. The trip and reset valves for the reactor vent valves (RVVs) will be tested under steam service at operating reactor pressure and temperatures which bounds accident conditions. The trip and reset valves for the RRVs are of a similar design. These qualification tests provide additional assurance of leakage integrity beyond what is done for the CNV bolted flanges as they are performed under steam service rather than water. Successful completion of the qualification tests will show that the body-to-bonnet joint does not develop a new leak pathway when pressurized with steam and justifies exclusion of these valves from a similar bolted flange analysis as performed for the other bolted connections. Additionally, this bolted connection is subject to 10 CFR Part 50, Appendix J, Type B testing and ASME Section XI in-service testing (IST) at RCS operating pressure before initial operation, as well as prior to going into operation after each outage. The ECCS trip and reset pilot valve body-to-bonnet bolted connection will be Type B tested to CNV design pressure and leakage measured. The pressure will be increased to RCS design pressure exceeding the RCS operating pressure requirement, per FSAR Section 5.4.2.1, and examined for leakage. This testing at each start-up makes demonstrating by analysis how the bolted connection performs under pressure not necessary. 3.2.1 Electrical Penetration Assemblies The NuScale EPAs use sheathed modules with a glass-to-metal sealing technology that is not vulnerable to thermal or radiation aging, does not require periodic maintenance, and can achieve a less-than-minimum detectable leak rate (Figure 3-5). See © Copyright 2019 by NuScale Power, LLC 24

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Section 5.3.2 for further discussion. The performance of the glass-to-metal EPA seal has been proven in currently operating nuclear plants. The EPA with installed modules is bolted to CNV flange penetrations similarly to the flanged access ports. Figure 3-5 depicts the pressurizer heater power supply EPA. This configuration is typical for all NuScale EPAs. The NuScale design includes the ability to test the double O-ring seals by pressurizing between the seals of the EPA similarly to the flanged access ports. EPA modules are provided with a test port for local leak rate testing (Figure 3-6) and are Type B tested periodically in accordance with the requirements of the owner's Appendix J testing program. Figure 3-5 Electrical penetration assembly modules (typical) © Copyright 2019 by NuScale Power, LLC 25

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 3.2.2 Emergency Core Cooling System Trip and Reset Body-to-Bonnet Seals There are five ECCS main valves supported by 11 trip and reset (pilot) valves for actuation. The 11 pilot valves are located on six penetrations on the upper CNV assembly. Each trip and reset valve body is located outside the CNV and is a component of the RCPB. The trip and reset valves are aligned to static RCS head during normal operation. The RCPB is the valve body welded to the CNV penetration safe end. The only joint is at the body-to-bonnet and is sealed with a double O-ring and test port arrangement similar to the rest of the Type B penetrations. The trip and reset valves have no containment or RCS isolation function; therefore, the test criteria for these penetrations is a Type B test of the double O-ring joint for each valve. Figure 3-7 shows the dual trip and reset valve design and the location of the body-to-bonnet joints. The test port is not shown. All trip and reset valve pressure boundaries are designed and constructed to ASME Code Class 1. © Copyright 2019 by NuScale Power, LLC 27

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 3-7 Emergency core cooling system trip and reset valve assembly 3.2.3 Containment Closure Flange The CNV closure flange allows disassembly of the CNV for refueling, maintenance, testing, and inspection of the NPM. It has a double metal O-ring seal with test port design similar to the RPV flange. © Copyright 2019 by NuScale Power, LLC 28

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 3-8 Containment vessel flange seal and leak port The test port is attached to a test line that is routed near the CNV head manway. 3.3 Containment Isolation Valves The CNTS containment isolation valve designs allow accurate Type C LLRT results used to quantify overall containment penetration leak rate. There are eight containment penetrations with primary system containment isolation valves (PSCIVs); all are NPS 2. Four of these piping penetrations are part of the RCPB (GDC 55). Each of these penetrations is protected by dual, ASME Code Class 1 PSCIVs of identical design. Four piping penetrations are open to containment atmosphere (GDC 56). Each of these ASME Code Class 2 penetrations is protected by dual ASME Code Class 1 PSCIVs of identical design. There are six secondary system piping penetrations in the CNV, none of which meet the criteria for Type C testing. These six penetrations are connected to a closed loop inside containment, the SGS (GDC 57). Four of these penetrations (main steam and feedwater lines) are protected by single ASME Code Class 2 SSCIVs and nonsafety backup valves. The other two penetrations (DHRS condensate return lines) are protected by an ASME Code Class 1/2 closed loop inside containment and an ASME Code Class 2 closed loop outside containment. © Copyright 2019 by NuScale Power, LLC 29

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 All CIVs, both primary and secondary systems, have the same actuator design and are quarter turn, ball valves. The size and ball design varies between primary and secondary valves, but the majority of design features are identical. 3.3.1 Primary System Containment Isolation Valves The PSCIVs are the only piping penetration isolations that are required to meet the Appendix J Type C test criteria. Both GDC 55 and GDC 56 penetrations are protected by PSCIVs of identical design, size, code class, and construction. The PSCIVs have a dual-actuator, single-body arrangement. Four PSCIVs protect GDC 55 penetrations and are designed to ASME Code Class 1. Four PSCIVs protect GDC 56 penetrations and are designed to ASME Code Class 2; all eight PSCIVs are constructed to ASME Code Class 1. The RCCW/CRDS piping penetrations are classified as GDC 56 penetrations; however, these lines do contain an ASME Code Class 2 closed loop inside containment. The NuScale design uses dual PSCIVs on the inlet and outlet of the RCCW/CRDS piping penetration for two reasons. First, two automatic isolations are required for single failure criteria to prevent a unique event in which a failure in RCCW/CRDS could initiate an ECCS actuation on high CNV water level. Secondly, the RCCW has several mechanical (non-welded) connections inside containment and would present challenges as a GDC 57 containment barrier. The ASME Code Class 2 CRDS piping provides a defense-in-depth barrier for containment isolation, but it is not credited as a GDC 57 closed loop. © Copyright 2019 by NuScale Power, LLC 30

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 3-9 Primary system containment isolation valve dual-valve, single-body design Each dual valve assembly is welded directly to the CNV head integral vessel nozzle safe end with a butt weld. The containment flooding and drain system, containment evacuation system (CES), and reactor component cooling water system valves are classified as Quality Group B, but are specified to be designed and constructed to ASME Code Class 1. 3.3.2 Feedwater Isolation Valve There are two feedwater containment penetrations. Both of these piping penetrations are NPS 5. Each penetration is protected by a dual ASME Code Class 2 SSCIV. The feedwater isolation valve (FWIV) has a dual-actuator, single-body arrangement. Each FWIV consists of an actuated isolation valve and a self-actuating check valve. The FWIV actuated isolation valve is a hydraulic-to-open, compressed gas-to-close, ball valve. It is the inboard valve of the dual-valve arrangement. The outboard valve is a safety-related nozzle check valve. The function of the feedwater isolation check valve is to close more rapidly than the FWIV in the event of a feedwater line break outside containment to preserve DHRS inventory. The feedwater isolation check valve has no containment isolation function and no other specific leakage criteria. Neither the FWIV nor the feedwater isolation check valve has an Appendix J Type C test requirement as defined © Copyright 2019 by NuScale Power, LLC 31

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 by Appendix J. This is typical for the FWIV of a PWR design; however, the FWIV is classified as IST Category A, having specific leakage criteria to maintain DHRS inventory. The dual-assembly is welded to a short section of piping welded directly to the CNV head integral vessel nozzle safe end with a butt weld arrangement. 3.3.3 Main Steam Isolation Valve and Bypass Valve There are two main steam containment penetrations. Both of these piping penetrations are NPS 12. Each penetration is protected by an ASME Code Class 2 SSCIV with an integral bypass valve. The main steam isolation bypass line is NPS 2. The main steam isolation valve (MSIV) has a single-actuator, single-body arrangement. The bypass valve is integral to the MSIV in a parallel arrangement. The MSIV and bypass valve are hydraulic-to-open, nitrogen-to-close, ball valves. The function of the MSIV and bypass valve is containment isolation, main steam isolation, and DHRS boundary preservation during decay heat removal actuation. Neither the MSIV nor the bypass valve has an Appendix J Type C test requirement as defined by Appendix J. However, both the MSIV and bypass valves are classified as IST Category A, having specific leakage criteria to maintain DHRS inventory. Each valve assembly is welded to a short vertical section of piping welded directly to the CNV head integral vessel nozzle safe end with a butt weld. © Copyright 2019 by NuScale Power, LLC 32

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 4.0 Preservice Inspection and Testing 4.1 Manufacturing Facility Testing and Inspection The CNV is hydrostatically tested in the factory in accordance with ASME Subsection NB-6000. The water-filled CNV is pressurized to a minimum of 25 percent over design pressure for at least 10 minutes. Pressure is then reduced to design pressure prior to examining for leaks. The acceptance criterion is no leakage indications at the examination pressure (design pressure). Nondestructive examination of the CNV in the factory includes:

All pressure-retaining and integrally-attached materials examination meets the requirements of NB-5000 and NF-5000 using examination methods of ASME Boiler and Pressure Vessel Code Section V.
All clad surfaces are magnetic particle or liquid penetrant examined in accordance with NB-2545 or NB-2546, respectively, of Reference 7.1.7 prior to cladding.
ASME Code Class 1 pressure boundary examinations are in accordance with NB-5280 and IWB-2200 using examination methods of ASME Boiler and Pressure Vessel Code Section V as modified by NB-5111. Preservice examinations shall include 100 percent of the pressure boundary welds.
ASME Code Class MC examinations are subsumed by NB exam requirements. The Class MC examination is in accordance with IWE-2200. In addition, due to the high pressure design of the CNV, the preservice examination requirements of IWB-2200 are applied (Reference 7.1.7).
Final preservice examinations are performed after hydrostatic testing, but prior to code stamping.

4.2 Preservice Design Pressure Leakage Testing A separate preservice design pressure leakage test is performed on the CNV. This test is performed to ensure that the integrated leakage of the CNV meets design criteria. This test is performed on every NuScale CNV and shall contain the following elements:

This test is required under a separate ITAAC.
As-designed flange covers shall be installed with the design bolting materials, design bolting assembly preloads, and design seals installed.
CNV bolted flanges shall be in place. Covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design.
The upper and lower halves of the CNV are assembled for the first module of the initial NuScale plant. After the first CNV for the initial plant is tested successfully, the upper and lower halves of all other containment vessels may be tested separately.
The CNV is pressurized with water to design pressure, held for 30 minutes, and no observed leakage shall be visible from any joint.
A COL Item requires the applicant to verify that the CNV design meets the design basis requirement to maintain flange contact pressure at accident temperature.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1

The test configuration may utilize blanked off pipe ends in place of the containment isolation valves
The acceptance criterion is no observed leakage from seals at examination pressure
The ECCS trip valve and reset valve body-to-bonnet joint seals are not considered to be a flanged connection and are not included in the preservice design pressure leakage test or containment flange bolting calculation.

4.3 Post-installation Testing and Inspection Preservice inspections and local leak rate testing after installation at the plant site verify the leakage integrity of containment, including verification that no degradation of containment leakage integrity occurred during shipping and installation. 4.4 Shipping and Receiving Controls In addition to the post-installation testing, 10 CFR 50, Appendix B controls ensure that the leakage integrity assurance, provided by preservice tests and inspections performed in a factory environment, is maintained throughout shipping and receiving processes. Quality assurance controls in accordance with 10 CFR 50, Appendix B, Section VII, and Section XIII ensure the quality of the CNV and CNV components throughout shipping and receiving operations. Shipping and handling requirements ensure that these activities do not result in damage or deterioration of CNV components. Procurement controls ensure that material and equipment conform to the procurement requirements and design specifications, including verification upon receipt. These quality assurance processes have not yet been established; however, as required by Appendix B, the controls will be included in the quality assurance programs of the manufacturing facility and COL holder with NRC oversight. Typical controls, as described in NQA-1, include:

measures for packaging, shipping, receiving, storage, and handling of items, and for the inspection, testing, and documentation to verify conformance to specified requirements
purchased items shall be inspected to verify conformance to specified procurement and design requirements
handling, storage, and shipping, of items shall be controlled to prevent damage, in accordance with established procedures
for critical, sensitive, or high-value items, specific procedures for handling, storage, packaging, shipping, and preservation
for critical, sensitive, or high-value items, specific procedures of special receiving inspection instructions
receiving inspection shall verify by objective evidence such features as:

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.0 Inservice Inspection and Inservice Testing ISI and IST are required by 10 CFR 50.55a(g) and (f) (Reference 7.1.1), respectively, and ensure that periodic requisite inspection and testing is performed on the CNTS to ensure leakage integrity is maintained. Type B testing is specified in the COL holders ISI plan and Type C testing specified in the COL holders IST plan. Both the ISI and IST programs are an integral part of the CLIP. 5.1 Inservice Inspection of the Containment System The ISI provides an essential function for the CLIP by confirming CNTS integrity and ensuring no new leakage paths are present. Age-based failure mechanisms are prevented and detected through the compact and accessible design of the CNV, along with inspections and examinations performed in accordance with the ASME Code Section XI Division 1 (Reference 7.1.8, hereafter referred to as Section XI). The NuScale CNV is an ASME Code Class 1 vessel. The CNV components are constructed of stainless steel or are clad on interior and exterior surfaces with stainless steel and are fully inspectable. Periodic, comprehensive ISI ensures that a degradation mechanism is detected and addressed before CNV integrity is threatened. The requirements for inspection of passive components (structures, welds, supports, etc.) are provided in ASME Section XI. The ASME Code defines ISI requirements for ASME Class 1, Class 2, Class 3, and Class MC components. The CNV is classified as a Class MC containment. The CNV is designed, constructed, and inspected to ASME Code Class 1. The ISI program specifies Type B local penetration leak tests, which are pneumatic pressure leak rate tests of the containment penetrations, such as openings, flanges, and EPAs. 5.1.1 Inspection Elements The NuScale primary CNV design is different from traditional containments. The major differences are summarized as:

The CNV is a high-pressure vessel.
The CNV provides the containment heat removal function to transfer decay heat from the fuel to the UHS.
During normal operation, the CNV is under a vacuum and is mostly submerged in borated water.
During refueling, the CNV is physically moved by a crane to the refueling area while loaded with fuel.
The lower CNV is exposed to a higher neutron flux than typical containments.
Although the CNV is a Class MC component, it is being constructed to ASME Class 1 vessel rules.
The inside of the CNV is inaccessible by personnel during startup and normal operation.
The low-alloy portion of the CNV is clad on its inside and outside surfaces.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Table 5-1 compares inspection requirements of NuScales adoption of ASME Class 1, Section XI, IWB to those required by ASME Class MC, Section XI IWE (Reference 7.1.8). Table 5-1 ASME Section XI inspection comparison IWE (Class MC) Requirement IWB (Class 1) Requirement IWE-1231 (a)(3) The CNV is fully accessible from both sides for visual exam. Only 80 percent of the containment boundary is required to be accessible for visual a single sided visual exam. IWE-1232 The CNV has no inaccessible areas. Exemptions are given for inaccessible areas which are embedded in concrete or visually blocked by plant structures, equipment, or components. IWE-1241 No parts of the NuScale containment fall under IWE-1241 for augmented examination due to the Augmented examinations dual clad geometries. No painted or other coatings are used on the NuScale containment. IWE-2500 These requirements are met by following IWB-2500 B-N-1, which requires a VT-3 of all Two inspections that would apply from Table IWE- accessible areas. 2500-1 (E-A) are a general visual of the CNV (Item E1.10) and a VT-3 of wetted surfaces of submerged areas (E1.11). IWE-2500 Requirement met by adopting IWB-2500 B-G-1 and B-G-2. Surface examinations are performed VT-1 exam of CNV bolted connections is required at least once per interval when bolting is removed. by Table IWE-2500-1 (E-A). Other IWE IWB Visual examination only In addition to inspection requirements for a metal containment, includes multiple surface and volumetric exams. © Copyright 2019 by NuScale Power, LLC 36

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Based on the high pressure and the safety function of the CNV, NuScale determined that enhanced inspection requirements are needed for the CNV. Therefore, NuScale will inspect the CNV to ASME Class 1 requirements. The CNV inspection elements are listed in Table 5-2. Several of the CNV inspection elements are similar to those required for the RPV. 5.1.2 Weld Inspection All CIVs are located outside the CNV. The ASME Section XI reduced-ISI requirements for small primary system pipe welds between the CNV and the CIVs are not applied to these welds. Welds between the CNV and the CIVs are ASME Code Class 1 and are inspected with a volumetric and surface exam at each test interval. The CNV design allows comprehensive inspections of welds, including volumetric and surface inspections. All pressure boundary welds are accessible and there are no areas that cannot be inspected. The basis for a NuScale ISI program for the CNV is shown in Table 5-2, which describes weld inspection locations and requirements. The specified surface, volumetric (ultrasonic), and visual examinations ensure that no new leakage paths are created over the service life of the CNV. 5.1.3 Bolted Flange Pressure Testing All flanges on the CNV and RPV have dual O-rings with a test port between the O-rings to allow for leak testing. Nozzles with flanges are listed in Appendix A.1. All CNV flanges are tested in accordance with 10 CFR 50, Appendix J, Type B criteria. The seal design of the RPV flanges is identical to the CNV. Leak testing of the RPV flanges is performed each time they are removed to ensure proper sealing. With the exception of the main CNV and RPV flange bolting, all bolts are connected to threaded inserts. There are no inspection requirements for the attachment welds for the threaded inserts. The EPAs are bolted-flange arrangements. The flange seal is leak tested as described in Sections 5.3.1 and 5.3.2. The EPA sheath modules are design and tested to have a negligible leak rate and only require an LLRT for post-maintenance activities. All EPAs are pressure tested periodically in accordance with 10 CFR 50, Appendix J, Type B criteria. If necessary, an EPA can be removed to pressure test a glass module. Section 5.3.2 provides additional discussion on EPA design and leakage integrity. The only pressure-retaining bolting greater than two-inches is in the RPV and CNV main flanges. The RPV and CNV use the same stud and bolt design. The RPV and CNV use the same tools and controls to disassemble and reassemble each vessel. These bolts are inspected per Section XI (Reference 7.1.8) Category B-G-1. Surface examination is performed when bolting is removed. The CNV and RPV main flange bolting is required to be removed and inspected once each interval. All pressure-retaining bolting in the CNV and RPV two-inches or less in diameter are Section XI Category B-G-2. These bolting assemblies require a VT-1 each interval if removed. This includes all flange bolting in the CNV and RPV. © Copyright 2019 by NuScale Power, LLC 37

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 The RPV and CNV require a leakage test (VT-2), Section XI Category B-P, at normal operating pressure after each refueling outage. In the NuScale design, leakage is continuously monitored in the CNV. This leakage monitoring is used to meet VT-2 exam requirements according to Section XI IWA-5241(c). During normal operation, the CNV is in a vacuum, so leakage would be from the pool to the inside of the CNV. The CNV leak detection system is able to detect leakage from both the RPV and CNV during normal operation. 5.1.4 Visual Inspections The ASME Code Class MC, Section IWE, requires only visual examination for structures, systems, and components subject to normal degradation and aging. However, based on the high pressure and safety functions of the CNV, the NuScale ISI program requires the CNV to meet ASME Code Class 1 requirements similar to the RPV. The CNV design allows visual inspection of the entire inner and outer surfaces; therefore, developing an undetected leak through the metal pressure boundary is unlikely. 5.1.5 Steam Generator Inspections and Controls The SG forms part of a GDC 57 closed-loop containment barrier for PWRs; therefore, its integrity and its failure mechanisms contribute to the integrity of the containment boundary. The NuScale SG design is different from traditional SGs. Major differences include:

The SG is located inside the RPV; it is not a separate component attached by RCS piping.
The tubes are helically coiled in the annular space around an upper riser.
Steam is generated on the inside of the tubes; lower pressure is on the tube interior.

The SG is a GDC 57 closed-loop system isolated by single SSCIVs (MSIV and bypass valve, FWIV). The SG is an ASME Code Class 1 RCPB. Detailed inspection requirements for the SG tubing and tube-to-tube sheet welds are part of the ISI program. An SG program is established in NuScale DCA Part 4, Technical Specification 5.5.4 to ensure that SG tube integrity is maintained. 5.1.6 Type B Testing Type B testing is local pneumatic pressure leak rate testing of containment penetrations in accordance with 10 CFR 50 Appendix J, such as EPAs, ports, manways, ECCS pilot valve bodies, and the CNV closure flange. It is an ISI test that is specified in the COL holders ISI plan. The NuScale ISI program specifies Type B LLRTs (Table 5-2). Table 5-2 is a summary of test and inspection elements in the NuScale ISI program for the CNV. © Copyright 2019 by NuScale Power, LLC 38

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Table 5-2 Summary of test and inspection elements Description Exam Category Examination Method CNV shell and head welds B-A Volumetric CNV support welds B-A, B-K, F-A Surface, volumetric Volumetric or not required CNV nozzle-to-shell welds B-D per B-D, Note 1 Nozzle-to-safe end dissimilar metal welds Note: Safe end is a short length of Surface and volumetric, Class 1 pipe that is welded between the B-F surface forged CNV pipe penetration and the CIV body. Exempted by IWB-1220 due ECCS pilot valve body to safe end welds B-J to small size Surface and volumetric PSCIV (GDC 55) body to safe end welds B-J Exemption per IWB-1220 due to small size not applied Surface PSCIV (GDC 56) body to safe end welds C-F-1 Exemption per IWB-1220 due to small size not applied SSCIV body to safe end welds C-F-1 Surface and volumetric Decay heat removal inner and outer safe C-F-1 Surface and volumetric end-to-piping welds CNV ports, manways, EPAs and ECCS Appendix J Type Pneumatic leakage pilot valve body-to-bonnet seals B VT-2 CNTS leakage test B-P Required for all pressure retaining components VT-3 for wetted surfaces General visual for surfaces CNV exterior surface N/A that are normally dry. Based on the requirements from IWE-2500-1 (E-A). CNV interior surfaces B-N-1 VT-3 Pressure retaining bolting material, B-G-1 Volumetric greater than two inches Pressure retaining bolting, two inches or B-G-2 VT-1 less © Copyright 2019 by NuScale Power, LLC 39

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.2 Inservice Testing of the Containment System The NuScale standard IST Plan, as defined in FSAR Section 3.9.6, identifies all valves with specific leakage criteria. Valves with specific leakage criteria as a containment boundary are identified as LTJ, a valve with an Appendix J Type C leakage test requirement. The IST Plan also specifies test frequencies that are pursuant to the requirements of 10 CFR 50, Appendix J, Option A, (III)(D)(3). The NuScale IST program identifies all valves in the NuScale design that have a Type C test requirement. The NuScale IST plan specifies IST requirements for the NuScale Power Plant design. 5.3 Type B Local Leak Rate Testing Type B tests of the double O-ring seals on the containment bolted closures are performed by local pressurization at containment peak accident pressure, Pa. Pressurized gas, such as air or nitrogen, is applied to the leak test ports, which are provided between the two O-ring seals in each bolted closure, and the pressure-decay over time or the leak flow rate is measured. All Type B tests use either the pressure-decay or flow-makeup method of detection as described in Reference 7.1.10. For the pressure-decay method, a test volume is pressurized with air or nitrogen to at least Pa. The rate of decay of pressure in the known test volume is monitored over time to calculate a leakage rate. For the flow-makeup method, the required test pressure is maintained in the test volume by making up test fluid, such as air or nitrogen, through a calibrated flowmeter. The makeup-flow rate is the leakage rate from the test volume. The design combined leakage rate for all penetrations and valves subject to Type B and Type C tests is limited to less than (0.60) La. An overall leak rate of less than (0.60) La will be confirmed by LLRT prior to the startup of each NPM. In accordance with 10 CFR 50, Appendix J, Type B tests are performed during each reactor shutdown for refueling or other convenient intervals in accordance with the CLRT Program. 5.3.1 Type B Test Method Pa, the bounding peak containment accident pressure, given in FSAR Tier 2, Section 6.2.1, has been calculated to be less than 1,000 psia. All double O-ring (Type B) seals are tested with air or nitrogen at a pressure not less than Pa. All Type B penetrations are tested each refueling outage. An as-found test is required to be performed before any Type B penetration is opened or manipulated in any way that would affect the as-found test. See Section 5.4.2 for a discussion of test considerations, including preconditioning. Test equipment is installed on the test port that is located between the double O-ring seals. The seal is then tested with compressed air or nitrogen using either the pressure-decay or flow-makeup method to measure the leakage as specified in the CLIP. Once as-found testing is performed and documented, the penetration can be opened. Just inside the CNV head manway is a small tubing connection to the CNV flange test port. The Type B test rig is connected here and an as-found test of the CNV flange is performed. Once the refueling outage is completed and penetrations are closed for the final time, an as-left Type B test is performed on penetrations (Reference 7.1.3). If a penetration was © Copyright 2019 by NuScale Power, LLC 40

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 not opened and no bolts were manipulated, and if the as-found test was within CLIP acceptance criteria, then no further tests on that penetration are necessary. The CNV flange is tested twice after it is reassembled. The first time will be in the refueling area to ensure the new CNV flange O-rings are installed properly and are sealed. The as-left test occurs after the NPM is moved to the operating bay. The as-left test ensures that CNV movement had no adverse effect on the CNV flange seal. After the CNV closure flange seal is tested, then the CNV head manway cover can be reinstalled and tested. 5.3.2 Electrical Penetration Assemblies The EPA sheath modules are installed and tested at the factory. Glass-to-metal seals (penetrations), exclusive of the flange-to-nozzle seals, are designed for leakage rates not to exceed 1.0 x 10-3 standard cm3/s (1.27 x 10-4 SCFH) of dry nitrogen at design pressure and at ambient temperature, including after any design basis event (Reference 7.1.11). Glass-to-metal seals typically achieve leak rates in the undetectable range, 1.0 x 10-7 standard cm3/s of dry nitrogen at design pressure and at ambient temperature. The glass-to-metal module seal is an established sealing technology that is not vulnerable to thermal or radiation aging and does not require periodic maintenance or testing. The module-to-EPA seal is periodically tested at the Type B testing frequency discussed in Section 5.5.2. The EPA flange seal is the same double O-ring seal design of all Type B penetration seals. The required installation acceptance criterion for leakage rate of each EPA is 1.0 x 10-2 standard cm3/s (1.27 x 10-3 SCFH) per Reference 7.1.11. The leakage margin allotment for Type B testing is preliminarily selected to be 50 times the installation acceptance criterion. This leakage margin for EPA contribution to overall containment leakage supports maintaining overall containment leakage to less than (0.60) La. 5.3.3 Ports and Manways All CNV access port seals and manway seals are the identical double O-ring design. The leakage performance of these seals is expected to be similar to the EPAs based on an evaluation of leakage performance for off-the-shelf metal seals. 5.3.4 Emergency Core Cooling System Pilot Valve Bodies There are six NPS 3 containment penetrations for the ECCS trip and reset valve assemblies. A Type B test is required at the double seal between the valve bonnet and body (see Figure 3-7). The rest of these valve bodies are self-contained metal barriers that form part of the containment pressure boundary. Leakage criteria for these seals is small compared to the other Type B boundaries due to the smaller size of the seals. 5.3.5 Containment Vessel Flange The CNV closure flange is a large double O-ring design (~45-foot circumference). This seal maintains the containment boundary between upper and lower CNV assemblies (see Figure 3-2). The CNV closure flange leakage limit for the CLIP is estimated to be 0.4-0.5 SCFH based on the linear seal length and performance of off-the-shelf metal seals. © Copyright 2019 by NuScale Power, LLC 41

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.3.6 Bolting The CNV bolting design for all EPAs, ports and manways is in accordance with ASME Section III, Division 1, NB-3231. Calculations were performed to verify the number and cross-sectional area of bolts required to withstand containment design pressure and maintain gasket reaction for leak tightness. The calculations were performed in accordance with ASME Boiler and Pressure Vessel Code Section III, Appendix E. The calculations determined the quantity, size, and spacing of the bolting for the ASME Code Class 1 flanges. The CNV bolted-closure design and preload design requirements ensure that Type B flange seals, including EPAs, remain sealed at design pressure. Flanges are as-found tested in accordance with 10 CFR 50, Appendix J before removal for refueling outage activities. The COL holders administrative controls are used during reassembly, including dual torque verification and QC hold points, to ensure EPAs, ports, manways, and flange seals are reassembled with fasteners at the correct torque. An as-left Type B test on the penetration seal verifies leakage is within the CLIP limits. 5.4 Type C Local Leak Rate Testing The PSCIVs are tested using either the pressure-decay or flow-makeup method. For the pressure-decay method the test volume is pressurized with air or nitrogen. These test methods are described in Section 5.3. Pressure to the PSCIV is applied in the same direction as the pressure is applied when the valve is required to perform its safety function. 5.4.1 Type C Test Method The PSCIVs are local leak rate tested using the pressure-decay method or the flow-makeup method at a pressure not less than Pa, which has been calculated to be less than 1,000 psia. Each CIV to be tested is closed by normal means without any preliminary exercising or adjustments (see Section 5.4.2). This closure can be the periodic closed stroke required as part of the IST Program. Piping is drained and vented as needed and a test volume is established that when pressurized, produces a differential pressure across the valve. The valve is then prepared for testing by removing the normal insert and replacing it with a test blank insert in the valve body (Figure 5-1). The test blank is closed to the CNV to establish the pressure boundary for the test in the same direction as would be required for the valve to perform its safety function. Test equipment is installed on the test port located between the test blank and the inboard ball. The valves are aligned so that a vent path is established downstream of the tested valve. The valve is then tested with air or nitrogen using either the pressure-decay or flow-makeup method as specified in the CLIP. When valve testing is completed, the test equipment is vented and the valves realigned. The tested valve is opened and the second CIV closed. The test alignment for the second CIV is established. The test equipment is repressurized and the second valve tested. © Copyright 2019 by NuScale Power, LLC 42

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.4.2 Test Considerations

As-found testing is a test performed to see how a component would perform if called upon in an accident scenario. It is the first actuation of the component that has been sitting in standby since the last time it was tested or for normal operation.

Technical specification limiting condition for operation criteria generally involve as-found values.

As-left testing is the final performance of a surveillance test or calibration of a component to determine its functional performance prior to placing it back into service. Technical specification surveillance test criteria generally involve as-left values. Appendix J LLRT requires as-found testing of Type B and C penetrations when entering a refueling outage and as-left testing when reassembling Type B penetrations or performing post-work testing on a PSCIV if maintenance was required that affected the leak function of the valve.
Preconditioning occurs when a component is exercised, adjusted, or otherwise manipulated prior to as-found testing performance. ASME and NRC requirements do not allow preconditioning for the performance of any as-found testing. For Type C penetrations, the requirements of the IST program and the CLIP must be balanced.

As-found stroke time testing is required for IST and an as-found LLRT is required for Appendix J. Both PSCIV valves in the same valve body present a challenge to the test engineer and the operations personnel to meet the requirements of both programs while achieving accurate as-found test results. 5.4.3 Primary System Containment Isolation Valves The PSCIV design provides double seals and a means to detect and measure leakage at bolted connections and stem packing that form part of the normal operating pressure boundary. The valve design also provides for Appendix J, Type C testing by the use of normal and test inserts that allow the pressurization of either side of the valves (see Figure 5-1). The normal and test inserts are provided with double O-ring seals to ensure the integrity of the test connection. A test port allows the Type C test rig to be connected to the valve body. Pressure can be applied between the test insert and the ball operator. This pressure is applied from the CNV direction for Type C testing. © Copyright 2019 by NuScale Power, LLC 43

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 5-1 Primary system containment isolation valves Type C testing 5.4.4 Secondary System Containment Isolation Valves There are no Type C leak test requirements for the feedwater isolation valves, MSIVs, and main steam isolation bypass valves. These valves do have specific leakage criteria for DHRS operability. Leak testing of these valves is in accordance with the technical specifications and the IST Program to maintain DHRS operability. There are two thermal relief valves on the steam generator lines inside containment. These valves provide thermal overpressure protection during potential solid water conditions during module startup and shutdown. These valves are part of a closed system inside containment and therefore are included in the licensee's CLIP (i.e. require an Appendix J, Type C leakage test) to ensure system leakage integrity. 5.5 Containment Leakage Rate Test Program The COL holders CLIP contains the following attributes:

Apply limits established in the plant design basis and the technical specifications to establish LLRT criteria to ensure all penetrations meet the preservice and periodic limit of (0.60) La at Pa for the combined leakage rate of all penetrations and valves subject to Type B and C testing.
Perform Type B LLRT testing in accordance with the ISI Program frequency.
Perform Type C LLRT testing in accordance with the IST Program frequency.
Document results of applicable ISI on the CNTS.
Document results of as-found and as-left Type B and C LLRTs.
Document post-work testing results on Type B and C pressure boundaries.
Analyze adverse conditions for generic considerations. All Type B seals are the same double O-ring design, and all Type C valves are the identical, two-inch,

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 hemispherical cartridge valve design. Additionally, each site employing the NuScale design will have 12 identical containment systems.

Maintain records to produce a periodic leakage test summary report to the NRC in accordance with 10 CFR 50, Appendix J.

5.5.1 Containment Leakage Limits The leakage rates of penetrations and valves subject to Type B and C testing are combined in accordance with 10 CFR 50, Appendix J. The combined leakage rate for all penetrations and valves subject to Type B and C testing shall be less than (0.60) La at Pa. Design basis containment leakage limits:

La = 0.20 weight percent, dry air inside containment at design pressure
La = approximately 17.5 SCFH, dry air inside containment at peak containment accident pressure
Pa = peak containment accident pressure, given in FSAR Tier 2, Section 6.2.1 The CLIP limits are derived from the design basis limits to meet (0.60) La for LLRT. If repairs are required to meet CLIP limits, the results are reported in a separate summary to the NRC in accordance with Reference 7.1.3, including the structural conditions of the components that contributed to the failure. As each Type B or C test, or group of tests, is completed, the combined total leakage rate is revised to reflect the latest results. Thus, a reliable summary of containment leak tightness is maintained current. Leakage rate limits and the criteria for the combined leakage results are described in the plant technical specifications.

5.5.2 Test Frequency Schedules for performance of periodic Type B tests are specified in the COL holders IST Program and periodic Type C tests are also specified. Provisions for reporting test results are described in the COL holders leak rate testing program. Conditional testing is in accordance with the COL holders procedures, but includes Type B or C testing anytime repair, replacement, or modification to a containment pressure boundary takes place. Type B tests are performed during reactor shutdown or refueling, or other convenient intervals, but in no case at intervals greater than two years (as specified in the COL holders ISI Program) per 10 CFR 50, Appendix J. Type C tests are performed during reactor shutdown or refueling, but in no case at intervals greater than two years (as specified in the COL holders IST Program. 5.5.3 Test Results and Reporting Requirements The CLIP program reporting requirements are pursuant to 10 CFR 50, Appendix J, Option A (V) (Reference 7.1.3). Preoperational and periodic tests are documented in a summary report that is made available for inspection, upon request, at the plant site. The summary report includes, at a minimum: © Copyright 2019 by NuScale Power, LLC 45

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1

a schematic arrangement of the leakage rate measurement system
the instrumentation used
the supplemental test method
the test program selected as applicable to the preoperational test
the subsequent periodic tests The report contains an analysis and interpretation of the leakage rate test data for the Type B and C test results and the applicable ISI results to the extent necessary to demonstrate the acceptability of the containment leakage rate in meeting acceptance criteria.

For each periodic test, leakage test results from Type B and C testing are included in the summary report. The summary report contains an analysis and interpretation of the Type B and C test results and the applicable ISI results that were performed since the last inspection interval (usually the last refueling outage). Leakage test results from Type B and C testing that failed to meet the CLIP acceptance criteria are included in a separate accompanying summary report that includes an analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements are included. 5.5.4 Special Testing Requirements Any major modification or replacement of components that is part of the containment boundary performed after preoperational leakage rate testing are followed by a Type B or C test as applicable for the area affected by the modification. The measured leakage from the test is included in the summary report. 5.5.5 Multiple NuScale Power Module Testing Multiple-NPM testing does not impact test frequencies of the CLIP nor the test frequencies of either the ISI or IST Programs. Risk-informed methods are not available to initial ISI or IST Programs, yet multiple-NPM testing is a factor in CLIP, inservice inspection, and IST:

Generic considerations of adverse conditions not only potentially affect similar components in the affected NPM, consideration must be given to similar components in all NPMs.
A NuScale 12-NPM plant plans for a nominal six refueling outages annually. This will provide a rapid accumulation of performance history for the CLRT, inservice inspection, and IST Programs. With NRC approval, risk-informed methods could be applied sooner than a traditional one-unit or two-unit reactor design.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 5.6 Type A Testing Challenges As discussed in the NuScale Design Certification Application (DCA), Part 7, Exemption Requests, NuScale is requesting an exemption from the requirements of GDC 52 and Appendix J Type A tests which specify the design for and performance of preoperational and periodic integrated leak rate testing at containment design pressure. However, NuScale has reviewed the requirements of GDC 52 and Appendix J Type A testing to assess the potential of performing integrated leakage testing within the NuScale design. The inherent safety feature of a small metal containment in direct contact with the UHS presents unique challenges to performing ILRT in the NuScale Power Plant design. At the conclusion of a normal shutdown for refueling, the CNV is filled with water to provide heat transfer during reactor refueling by filling the containment with water up to a level near the reactor pressurizer baffle plate. The heat transfer across the reactor vessel wall into the containment filled with water and through the containment wall into the UHS water provides cooling for the fuel in the RPV (Figure 5-2). The high heat transfer ability of the system coupled with the changing decay heat from the core, as well as the UHS heat transfer that is coupled to the rest of the NPMs in the UHS pool, creates a highly variable temperature system. 5.6.1 Temperature To ensure temperature variations are detected and offset, high-precision sensors both in the top of the pressurizer (Figure 5-2) and in the containment gas space are provided. If RPV water level were lower than the baffle plate, then additional area under the baffle would need to be individually instrumented. Sensors, more accurate and located in different locations than the normal plant temperature sensors inside the RPV and CNV, would be needed to monitor temperature changes of coolant in the RPV and the CNV. While the exact number of additional sensor required is not known, including the additional permanently installed sensors in the NuScale design would significantly increase the number of sensors for the CNV and add more signal leads to those already required. Permanently installed sensors may not be in the optimal locations for a given test. Differing conditions (water level, air and water temperature) than those for the design test will require review and possible reconfiguration of instrument numbers and locations to provide meaningful CILRT test results. © Copyright 2019 by NuScale Power, LLC 47

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Figure 5-2 Reactor pressure vessel, containment vessel, ultimate heat sink temperature gradients 5.6.2 Temperature Change During a CILRT, a coincident undetected temperature change of the gas volume would result in an uncompensated change in pressure of magnitude similar to the allowable pressure change associated with test leak rate limit (Lt). From the Combined Gas Law, a 0.1ºF temperature rise will raise CNV pressure proportionately.

                                        =                                                    Equation 5-1

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Rearranging the equation to determine P1

                                            =                                                  Equation 5-2 For the NuScale design, P2 at 1,000 psia, T1 at 559.7 R (100 degrees F), T2 (one tenth of a degree change) 559.6 R (99.9 degrees F) provides a pressure of:

559.7

                                         = 1000          = 1000.179 559.6 The equivalent average temperature change in a large standard plant design with a 60 psia test pressure would have a resulting pressure of:

559.7

                                           = 60          = 60.011 559.6 Thus, for the NuScale design, a 0.1 degree F undetected temperature increase in the gas volume, the pressure would rise 0.179 psi. The allowable pressure change to meet the leakage criteria for the NuScale design is approximately 0.06 psia. Therefore, a 0.1 degree F change in average temperature of the gas in the CNT results in three times the pressure change for the maximum allowable leak rate at 1,000 psia. Changes in average temperature of the fluid inside containment have a similar, although less pronounced, impact. This emphasizes the need for very accurate temperature measurement in order to obtain a representative average temperature of the CNT atmosphere during ILRT testing. The also highlights a challenge in obtaining accurate pressure measurement, as high-precision gauges available for field installation are typically accurate within 0.01%

of full scale, or 0.1 psia for a 1,000 psia measurement. The pressure change of 0.011 psi for a large standard plant is more than 16 times smaller, and would not be expected to cause failure of an ILRT. 5.6.3 Instrumentation Sensors to measure dew point temperature or relative humidity are not currently included in the NuScale containment or RPV instrumentation. Multiple dew point sensors to perform the CILRT are needed in various regions and elevations:

near the top of the CNV where the insulated head is above the reactor pool surface
mid-height of the gas volume
just above the top of the internal CNV water level to ensure these different environments are monitored
inside the pressurizer and possibly under the baffle plate to monitor the RPV gas space Dew point sensors capable of withstanding 1,000 or 500 psia are uncommon, but available. However, the most accurate sensors (e.g., gravimetric, chilled mirror hygrometer) are not suitable for field installation and usually are designed for an

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 application involving flow past the sensor. Without dew point or relative humidity measurement, the effect of evaporative increases in vapor pressure would have to be approximated. Ensuring proper placement, accuracy, and calibration will be difficult, and meeting all ANSI/ANS 56.8 criteria appears insufficient to provide accurate test results and essentially not possible. Allowance for uncertainties due to these effects are included in establishing acceptance criteria. These allowances apply even if the plant conditions causing the need for them do not occur during the test (i.e., even if no temperature or dew point variations actually exist, the data would have to be adjusted on the assumption that the lack of ability to sense such variations was the result of insufficient monitoring capability). 5.6.4 Allowable Leak Rate The allowable leak rates of large PWRs are typically above 1 SCFM with many being around 5 SCFM. The test acceptance criterion for NuScale is approximately 0.226 SCFM at 1,000 psia. As a result, a CILRT for a NPM must have a monitoring accuracy that is 27 times better than commonly used. Because large PWRs sometimes have difficulty meeting acceptance criteria for stability or accuracy, the challenge for NuScale is even greater. The acceptance criterion for passing a CILRT is 75 percent of the design leak rate. One-third to one-half of the margin is needed for actual leakage with the rest needed for operational margin to allow for degradation in future LLRT results without requiring immediate repair. Because the combination of uncertainty allowances may result in reducing the acceptable leak rate result to less than half of the 0.75 of La, it is likely NuScale CILRTs will fail repeatedly on assumed and actual data uncertainty and need to be re-performed at a considerably higher rate than existing plant CILRTs Normal operational instrumentation is insufficient in accuracy and coverage to be useful for a CILRT. More accurate sensors are needed for a NuScale CILRT because the leak rate that needs to be detected is about 1/30 of that for a large PWR, and they must work at 1,000 psia. This presumes that such instrumentation is either permanently or temporarily installed inside the module, rather than being outside with sensors inserted through external test points. 5.6.5 Alternate Testing Arrangement Considered Testing under dry containment conditions was also considered. Such conditions could be achieved through a complete core offload, or by completing the test with the ECCS valves closed and decay heat removal through the normal operating pathways (i.e. the SGs and, to a lesser extent, the CVCS). A full core offload would eliminate impacts from core heat of the NPM being tested, but would not eliminate the heat transfer from the reactor pool to the containment gas space. Numerous temperature and humidity measurements throughout the containment would © Copyright 2019 by NuScale Power, LLC 50

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 still be required, which presents the same challenges already discussed. This also defeats the purpose of performing an as-found and as-left containment leakage test, as this cannot reasonably be performed in the NuScale design if a full core offload is first required. Testing with the ECCS valves closed would require that reactor pressure be safely above the containment test pressure, due to the passive nature of the ECCS valves which will begin to open when CNV pressure is close to or above RPV pressure. Reactor pressure would, therefore, have to be greater than approximately 1,100 psia, with a corresponding pressurizer temperature of approximately 556 degrees F. This scenario would present an unacceptable negative safety impact to ECCS operation and introduce additional significant impacts to containment gas space temperature. Testing in these conditions also would not eliminate the challenges already presented. 5.6.6 Integrated Leak Rate Testing Assessment Conclusions NuScale has reviewed the requirements of GDC 52 and Appendix J Type A testing to assess the potential of performing integrated leakage testing within the NuScale design. The inherent safety features of the NPM limit the ability of the NuScale design to conform with Appendix J Type A testing acceptance criteria and limit the effectiveness of Type A tests for the NuScale design. The heat transfer mechanisms and high heat transfer ability of the NPM creates a variable temperature and pressure atmosphere within containment. The prescriptive Appendix J Type A testing requirements and acceptance criteria are impractical for the NuScale design. The temperature and pressure impacts on Type A testing and associated acceptance criteria for the NuScale design, increases the likelihood of inaccurate results, false test failures, and multiple testing iteration requirements. Application of Type A testing requirements to the NuScale containment would likely yield inaccurate leakage results due to the limited effectiveness of Type A acceptance criteria when applied to the NuScale design. The evaluation of the bolted flange connections discussed in Section 3.2 provides reasonable assurance that the Type B measured leakage of the bolted flange connection will be representative of leakage at design basis conditions. This evaluation supports that the containment design does not require Type A testing to measure leakage while under pressure. Accessibility constraints within containment, and the installation of a large quantity of additional CNV permanent or temporary instrumentation for Type A testing, would expose occupational radiation workers to unnecessary radiation doses to support testing without a commensurate safety benefit. This unnecessary exposure would be required to support installation, maintenance, and calibration of the equipment necessary to perform Type A tests. Conformance with GDC 52 and Appendix J Type A testing requirements is impractical for the NuScale design. The CLIP, supported by the NuScale design, provides leakage integrity assurance for the NuScale containment. © Copyright 2019 by NuScale Power, LLC 51

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 6.0 Material Selection and Aging Degradation Leakage Rate Test Program The NuScale containment design includes material selection that supports leakage integrity assurance. Potential degradation can be identified by inspection and examination prior to the formation of containment leakage pathways. 6.1 Material Selection and Operating Conditions Table 6-1 lists the CNV pressure-retaining materials. Nonpressure-retaining materials are not listed or discussed because they are not involved in maintaining the leak tightness for the CNV. Most of the CNV is immersed in the reactor pool water during plant operation. The CNV operating temperature is 100 degrees F. Although the minimum specified pool water temperature is 40 degrees F, typical pool water temperature is approximately 100 degrees F under operating conditions. During most of the plant operating period, the CNV interior is evacuated of water vapor or non-condensable gases to maintain a vacuum atmosphere by the CES. The internal CNV pressure is about 0.037 psia under normal operating conditions. This ensures that water vapor leaked into the CNV does not condense and collect at the bottom of the CNV. During the plant shutdown process, the CNV is flooded with reactor pool water when the operating condition is in the safe shutdown mode and the RCS coolant temperature drops below 300 degrees F. Only a small portion of the CNV is part of the RCPB: the portion in contact with RCS coolant during plant operation. The CNV components in contact with RCS coolant are the following nozzles and their safe ends:

CVC injection nozzle (CNV6)
CVC discharge nozzle (CNV13)
pressurizer spray nozzle (CNV7)
RPV high point degasification nozzle (CNV14)

The nozzles are SA-508 Grade 3 Class 2 cladded with stainless steel on the inside and outside surfaces. The safe ends are nickel-base Alloy 690 and are attached to the nozzles with nickel-base weld filler metals Alloy 52/152. 6.1.1 Pool Water Chemistry The NuScale reactor pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in the EPRI Primary Water Chemistry Guidelines (Reference 7.1.25). The NuScale reactor pool water chemistry parameters are listed in Table 6-2. No target limit and monitoring frequency are set for lithium, hydrogen peroxide, magnesium, calcium, or aluminum. Lithium is not expected to be added in the pool and hydrogen peroxide addition is only required as needed. The control limit and monitoring requirement for calcium, magnesium, and aluminum will be specified by the fuel vendor. For pH and conductivity, control limits are not included. Only the monitoring frequency is specified. Because the pool pH is primarily determined by boron concentration, the control limit is established for boron, rather than pH and conductivity. The remainder of © Copyright 2019 by NuScale Power, LLC 52

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 the NuScale reactor pool water chemistry parameters: boron, chloride, fluoride, sulfate, silica, TSS, and gamma isotopic activity are tabulated in Table 6-2 with their target limits and monitoring frequency. 6.1.2 Reactor Coolant System Coolant Chemistry NuScale follows the EPRI Primary Water Chemistry Guidelines. Limits for chemical species are provided in Table 6-3. These reactor coolant chemistry specifications conform to the recommendations of Regulatory Guide 1.44 (Reference 7.1.24). The RCS water chemistry is controlled to minimize corrosion of RCS surfaces and to minimize corrosion product transport during normal operation. The CVCS provides the means for adding chemicals through charging flow and for removing chemicals through dilution or purification. For reactivity control, boric acid is added as a soluble neutron poison. The concentration of boric acid is varied throughout reactor operation as needed for reactivity control. To maintain the alkalinity of the coolant, lithium hydroxide enriched with the lithium-7 isotope is added to the coolant. Slight alkalinity is maintained in accordance with the recommendations of the EPRI Primary Water Chemistry Guidelines. This chemical is chosen for its compatibility with boric acid, stainless steel, zirconium alloys, and nickel-base alloys. Lithium hydroxide is added to the coolant through the charging flow of the CVCS. It is removed from the coolant by the purification systems of the CVCS or reduced in concentration by dilution. The coolant pH is determined based on the recommendations in the EPRI PWR Primary Water Chemistry Guidelines and fuel vendor limits. To maintain the reducing environment in the coolant, dissolved hydrogen is added to the coolant. Hydrogen is chosen for its compatibility with the aqueous environment and its ability to suppress radiolytic oxygen generation during normal operation. Dissolved hydrogen is added to the coolant by direct injection of high pressure gaseous hydrogen into the CVCS charging flow. During startup, oxygen is removed by a combination of mechanical degasification by the CES and by chemical degasification using hydrazine. Hydrazine is an effective oxygen scavenger at low temperatures and is added to the coolant by the charging flow of the CVCS. © Copyright 2019 by NuScale Power, LLC 53

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Table 6-1 Containment vessel pressure-retaining materials Item Material CNV Vessel CNV top head cover SA-182 Grade F304/304L CNV top head and upper CNV shell SA-508 Grade 3 Class 2 (cladded with Type 309L/308L) Lower CNV shell and CNV bottom head SA-965 Grade FXM-19 Nozzles and Access Ports Nozzles and access ports SA-508 Grade 3 Class 2 (cladded with Type 309L/308L) Safe-ends for nozzles SB-166 or SB-167 UNS N06690 Covers for access ports SA-240 Type 304/304L Pressure-Retaining Bolting Studs and nuts for CNV main closure flange between upper SB-637 UNS N07718 CNV and lower CNV Studs and nuts for closure covers in upper CNV shell SB-637 UNS N07718

Pressurizer heater access ports
SG inspection ports
CRDM access port Studs and nuts for closure covers in CNV top head SA-564 Grade 630 Condition H1100
CNV top head cover
PZR heater power covers
CNV head manway cover
CRDM power cover
Instrumentation and controls (I&C) Division 1 covers
I&C Channel A through B covers
RP Group 1 & 2 covers Weld Filler Metals Low alloy steel weld filler metals for SA-508 Grade 3 Class 2 SFA-5.5, SFA-5.23, SFA-5.28, or SFA-5.29, classifications compatible with low alloy steels SA-508 Stainless steel weld filler metals for joining SA-965 Grade SFA-5.4: E209 FXM-19 SFA-5.9: ER209 Stainless steel weld filler metals for cladding SA-508 Grade SFA-5.4: E308L, E309L 3 Class 2 SFA-5.9: ER308L, ER309L, EQ308L, EQ309L SFA-5.22: E308L, E309L Note: One-layer of Type 309L cladding is used for CNV interior surface. At least two layers of cladding (first layer is Type 309L and subsequent layer(s) is Type 308L) are used for CNV exterior surface or flange sealing surface.

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Table 6-2 Target limits and monitoring frequency for pool water Water Chemistry Monitor Normal Range Unit Monitor Method Parameters Frequency pH N/A N/A 1/week pH meter Conductivity N/A N/A 1/week Conductivity meter Boron 1,800 ppm 1/week ICP-MS/Titration Chloride <0.15 ppm 1/month IC Fluoride <0.15 ppm 1/month IC Sulfate <0.15 ppm 1/month IC Silica <1 ppm 1/month UV-Vis TSS 1.0 ppm 1/month Gravimetric Procedure Gamma Isotopic 0.001 µCi/gram 1/week Gamma Spectroscopy Activity Table 6-3 Reactor coolant system coolant chemistry Normal Operating Parameter (units) RG 1.44 Limit Range Chloride (ppm) 0.05 0.15 Fluoride (ppm) 0.05 0.15 Dissolved Oxygen (ppm) 0.005 0.10 Sulfate (ppm) 0.05 N/A Hydrogen (cc/kg) 25 - 50 N/A Boron (ppm) 0 - 2,000 N/A 6.2 Aging Degradation Assessment This section assesses the following aging degradations for the CNV pressure boundary materials:

fatigue
boric acid corrosion
primary water stress corrosion-cracking (PWSCC)
stress corrosion-cracking (SCC) of austenitic stainless steels
SCC of pressure-retaining bolting materials
irradiation embrittlement of lower CNV 6.2.1 Fatigue The pressure-retaining components of CNV are analyzed for fatigue in accordance with applicable subsections of ASME Section III. For the CNV nozzles and their safe ends

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 that are in contact with RCS coolant during normal plant operation, the fatigue analysis considers the environmental effects in accordance with RG 1.207 (Reference 7.1.26) and NUREG/CR-6909 (Reference 7.1.27). The CNV components that are part of RCPB are described in Section 6.1. The SA-508 Grade 3 Class 2 nozzles are cladded with stainless steel. Therefore, only Alloy 690 safe ends and Alloy 52/152 safe end welds in contact with RCS coolant during normal operation are subjected to the environmental correction factor per RG 1.207. Qualification of high-heat input cladding processes if used on SA-508 Grade 3 Class 2 complies with RG 1.43 to demonstrate the processes do not induce excessive intergranular separations under the cladding. This prevents possible propagation of pre-existing flaws in SA-508 Grade 3 Class 2 base metal from fatigue loading during plant operation. Based on the above consideration, cracking of CNV pressure-retaining components due to fatigue loading is unlikely during the design lifetime. 6.2.2 Boric Acid Corrosion Exposure of carbon steel or low-alloy steels to borated pool water or reactor coolant can lead of loss of materials by boric acid corrosion. The pressure-retaining low-alloy steels are SA-508 Grade 3 Class 2 and are compatible with low-alloy steel welds. To prevent boric acid corrosion, the exposed low-alloy steel surfaces are cladded with austenitic stainless steel by weld overlay. The interior surfaces are cladded with one layer of Type 309L, while the exterior surfaces and flange sealing surfaces are cladded with one layer of Type 309L followed by at least one additional layer of Type 308L. The remaining pressure-retaining materials are either stainless steels or nickel-based alloys. Because all CNV surfaces in contact with borated pool water or reactor coolant are corrosion-resistant stainless steels or nickel-based alloys, boric acid corrosion is not an applicable aging degradation mechanism for the CNV. 6.2.3 Primary Water Stress Corrosion-Cracking Alloy 600 and its weld Alloy 82/182 are susceptible to PWSCC when exposed to high-purity deaerated, hydrogenated water at elevated temperatures. The PWSCC of Alloy 600/82/182 is enhanced when high tensile stress, high temperature, and a susceptible microstructure are present simultaneously. The CNV components that are part of the RCPB are described in Section 3.0. Only these Alloy 690 safe ends and their Alloy 52/152 welds are in contact with reactor coolant during plant operation. Alloy 690 for fabricating the safe ends is thermally treated to maximize resistance to PWSCC. Extensive laboratory testing and PWR operating experience have confirmed that Alloy 690/52/152 are highly resistant to PWSCC (Reference 7.1.13). The NuScale primary chemistry follows EPRI Primary Water Chemistry Guidelines which also minimize PWSCC. Therefore, PWSCC of Alloy 690 safe ends and Alloy 52/152 safe end welds in the CNV is unlikely. © Copyright 2019 by NuScale Power, LLC 56

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 6.2.4 Stress Corrosion-Cracking of Austenitic Stainless Steels The austenitic stainless steels for the CNV pressure-retaining components are the following:

SA-182 Grade F304/304L for CNV top head cover
SA-240 Type 304/304L for covers of various access ports
SA-965 Grade FXM-19 and E209/ER209 welds The Type 304/304L and FXM-19 austenitic stainless steel base metals are used in the solution-annealed condition. The maximum carbon content is limited to 0.03 percent for Type 304/304L and 0.04 percent for FXM-19. If water quenching is not used following solution annealing, non-sensitization is verified by ASTM A262 Practice A or E. In accordance with RG 1.44, if these materials are exposed to 800 to 1500 degrees F during heat treating or processing other than welding for more than one hour subsequent to solution annealing, ASTM Practice A or E is retested to confirm non-sensitization in the as-built condition.

There are no pressure-retaining austenitic stainless steel welds with Type 3XX weld filler metals. The circumferential welds between SA-965 Grade FXM-19 components in the lower CNV are joined with E209/ER209 weld filler metals. Therefore, pressure-retaining austenitic stainless steel welds in the CNV are limited to E209/ER209. The pressure-retaining austenitic stainless steels or welds are only in contact with pool water during plant operation or NPM movement. As described in Section 6.1, the pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Type 304/304L austenitic stainless steel has been used for spent fuel pool liners and spent fuel pool racks with excellent operating experience. Occasionally, transgranular stress corrosion-cracking (TGSCC) has been observed in Type 304 piping when exposed to borated water near ambient temperatures in PWRs. This has been attributed to sensitization, elevated chloride concentration, and high residual stresses from welding (Reference 7.1.14). However, TGSCC is unlikely to occur in CNV Type 304/304L closure covers based on the following considerations:

Sensitization of Type 304/304L is prevented by limiting carbon content to 0.03 percent maximum. Nonsensitization is verified by ASTM A262 Practice A or E in accordance with RG 1.44 as described above.
The pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Deleterious species in the water are monitored to be below acceptable limits. Halogens such as chloride and fluoride are maintained below 0.15 ppm.

SA-965 Grade FXM-19 (UNS S20910) is used for the lower CNV (Figure 6-1). This material is commonly referred to as XM-19 or Nitronic 50. Due to its higher yield strength and better SCC resistance than Type 304, XM-19 has been used extensively in boiling water reactor internals. According to Reference 7.1.15, there has been no failure or cracking of XM-19 after more than 25 years of service. © Copyright 2019 by NuScale Power, LLC 57

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 The use of XM-19 in borated PWR water is limited to components in pumps and valves. There have been no reports of SCC of XM-19 in such applications. Compared to Type 304/304L, XM-19 contains slightly more chromium and nickel, which enhances its SCC resistance. Because XM-19 contains small amounts of niobium and tantalum, it is more resistant to sensitization than Type 304 or 316. Therefore, XM-19 provides superior corrosion resistance than Type 304 and Type 316. The E209/ER209 weld filler metals and SA-965 Grade FXM-19 forgings have almost the same chemical composition. For the lower CNV use, the maximum carbon content of SA-965 Grade FXM-19 and E209/ER209 weld filler metals is restricted to 0.04 percent In addition, the weld filler metals contain 5 to 20 FN delta ferrite in accordance with RG 1.31 (Reference 7.1.16) to prevent microfissuring. Residual delta ferrite in E209/ER209 welds is beneficial in preventing SCC. Therefore, SA-965 Grade FXM-19 and E209/ER209 welds have superior corrosion resistance than Type 304/304L, and SCC is even less likely for the lower CNV than in the Type 304/304L closure covers. 6.2.5 Stress Corrosion-Cracking of Pressure-Retaining Bolting Materials The pressure-retaining bolting materials in the CNV are the following:

SB-637 UNS N07718 (also known as Alloy 718)
SA-564 Grade 630 (also known as Type 17-4PH), H1100 Alloy 718 is used for studs and nuts for the CNV main closure flange between the upper CNV and the lower CNV, and for various closure covers in the upper CNV shell. Type 17-4PH is used for studs and nuts for various closure covers in the CNV top head.

The pressure-retaining bolting materials are in contact with the pool water only during plant operation or NPM movement. As described in Section 6.1, the pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Alloy 718 is an austenitic, precipitation-hardenable alloy whose composition is adjusted to enable strengthening by heat treatment. Alloy 718 has been used inside PWRs due to its excellent SCC resistance in primary water, although intergranular SCC of Alloy 718 has been reported (Reference 7.1.16). The Alloy 718 bolting for the CNV is submerged in the pool water during plant operation. Intergranular SCC of nickel-base alloys is unlikely at the pool water temperature of 100 degrees F. Alloy 718 contains at least 50 percent nickel. Alloys containing more than 30 percent nickel are extremely resistant to chloride-induced TGSCC (Reference 7.1.18). Therefore, SCC of Alloy 718 bolting in the CNV is unlikely. Type 17-4PH is a martensitic precipitation hardenable stainless steel. Type 17-4PH in the H900 condition is relatively susceptible to SCC. However, laboratory SCC testing showed that Type 17-4PH in the overaged H1100 condition is much more resistant to SSC (Reference 7.1.19). Type 17-4PH for CNV bolting is used in the overaged H1100 condition. There have been no reports of SCC of Type 17-4PH in the H1100 condition in PWR applications. Failure of Type 17-4PH in the H1100 condition due to thermal embrittlement has been reported after exposure to temperatures above 500 degrees F © Copyright 2019 by NuScale Power, LLC 58

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 (Reference 7.1.20). However, thermal aging embrittlement is not a concern for 17-4PH used for the CNV because the operating temperature is only 100 degrees F. The pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Deleterious species in the water are monitored to be below acceptable limits. Halogens such as chloride and fluoride are maintained below 0.15 ppm. Therefore, SCC of Type 17-4PH bolting in the CNV is unlikely. 6.2.6 Irradiation Embrittlement of Lower Containment Vessel Of the CNV pressure-retaining components, the portion of the lower CNV near the beltline region has a 57-EFPY (effective full power years) peak fluence exceeding 1E+17 n/cm2, E > 1 MeV. The 57-EFPY fluence of the SA-508 Grade 3 Class 2 and compatible low alloy steel welds in the CNV is below 1E+17 n/cm2, E > 1 MeV. Therefore, the SA-965 Grade FXM-19 forgings (Figure 6-1) and the circumferential E209/ER209 welds between the SA-965 Grade FXM-19 forgings are assessed for irradiation embrittlement. The following 57-EFPY peak fluence base is used to assess CNV irradiation embrittlement.

CNV beltline forgings: 5.5E+18 n/cm2, E > 1 MeV
CNV beltline welds: 2.1E+18 n/cm2, E > 1 MeV The above values are conservative because the actual 57-EFPY peak CNV fluence values are expected to be lower. To enable comparison of irradiation embrittlement data originated from different reactor types, fluence is usually converted to average number of displacements per atom (dpa). A typical conversion factor for light water reactors is 1 dpa = 6.7E+20 n/cm2 per MRP-175 (Reference 7.1.21). Using this conversion factor, the dpa equivalent of the lower CNV bounding peak fluence is the following:
CNV beltline forgings: 5.5E+18 n/cm2, E > 1 MeV = 0.0082 dpa
CNV beltline welds: 2.1E+18 n/cm2, E > 1 MeV = 0.0031 dpa Based on extensive irradiated Type 3XX fracture toughness data, MRP-175 proposed the following screening fluence for irradiation embrittlement in PWR reactor internals:
wrought austenitic stainless steels 1.5 dpa
austenitic stainless steel welds or CASS 1 dpa However, NUREG/CR-7027 (Reference 7.1.22) recently proposed the following more conservative threshold fluence for irradiation embrittlement in Type 3XX austenitic stainless steels than the MRP-175 screening fluence:
wrought austenitic stainless steels = 0.5 dpa
austenitic stainless steel welds or CASS = 0.3 dpa Although XM-19 and E209/ER209 welds contain higher manganese and nitrogen content than Type 3XX, there is no data to indicate such differences have a pronounced

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 effect on irradiation embrittlement. Fracture toughness testing of irradiated solution annealed XM-19 was recently performed under an EPRI-DOE program (Reference 7.1.23). After 0.28 dpa at 340 degrees C (644 degrees F), fracture toughness was found to be JIc = 198 kJ/m2 (KJc = 212 MPam) when tested at 289 degrees C (552 degrees F). This XM-19 value is well within the scatter band of Type 3XX austenitic stainless steels reviewed by MRP-175 or NUREG/CR-7027. Therefore, XM-19 irradiation embrittlement behavior is similar to Type 3XX, at least up to 0.28 dpa. The bounding 57-EFPY peak fluence is 0.0082 dpa for the CNV beltline base metal and 0.0031 dpa for the CNV beltline welds. These bounding fluence values are tiny fractions of either MRP-175 screening fluence or NUREG/CR-7027 threshold fluence for irradiation embrittlement. Therefore, loss of fracture toughness in the lower CNV SA-965 Grade FXM-19 forgings or the E209/ER209 welds from neutron irradiation during the design lifetime is negligible. © Copyright 2019 by NuScale Power, LLC 60

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 7.0 References 7.1.1 U.S. Code of Federal Regulations, Codes and Standards, Section 50.55a, Part 50, Chapter I, Title 10, Energy, (10 CFR 50.55a). 7.1.2 U.S. Code of Federal Regulations, General Design Criteria for Nuclear Power Plants, Appendix A, Part 50, Chapter I, Title 10, Energy, (10 CFR 50 Appendix A). 7.1.3 U.S. Code of Federal Regulations, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Appendix J, Part 50, Chapter I, Title 10, Energy, (10 CFR 50 Appendix J). 7.1.4 U.S. Nuclear Regulatory Commission, Performance-Based Containment Leak-Test Program, Regulatory Guide 1.163, Rev. 0, September 1995. 7.1.5 U.S. Nuclear Regulatory Commission, Standard Review Plan, PWR Dry Containments, Including Subatmospheric Containments, NUREG-0800, Section 6.2.1.1.A, Rev. 3, March 2007. 7.1.6 U.S. Nuclear Regulatory Commission, Containment Leakage Testing, NuScale DSRS Section 6.2.6, Rev. 0, June 2016, ML15356A388. 7.1.7 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III Division 1, Rules for Construction of Nuclear Facility Components, 2013 edition, New York, NY. 7.1.8 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components, 2013 edition, New York, NY. 7.1.9 American National Standards Institute/American Nuclear Society, Containment Isolation Provisions for Fluid Systems After a LOCA, ANSI/ANS 56.2, 1984, La Grange Park, IL. 7.1.10 American National Standards Institute/American Nuclear Society, Containment System Leakage Testing Requirements, ANSI/ANS 56.8, 1994, La Grange Park, IL. 7.1.11 Institute of Electrical and Electronics Engineers, IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations, IEEE Standard 317-1983 (R2003), New York, NY. 7.1.12 Nuclear Energy Institute, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, NEI 94-01, Rev. 3, June 2011. 7.1.13 Electric Power Research Institute, Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI #1009801, 2004, Palo Alto, CA. © Copyright 2019 by NuScale Power, LLC 62

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 7.1.14 NRC Information Notice 2011-04: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors. 7.1.15 Electric Power Research Institute, EPRI Materials Management Matrix Project: Economic Simplified Boiling Water Reactor Degradation Matrix, Rev. 0, EPRI

                  #1016332, 2008, Palo Alto, CA.

7.1.16 U.S. Nuclear Regulatory Commission, Control of Ferrite Content in Stainless Steel Weld Metal, Regulatory Guide 1.31, Rev. 4, October 2013. 7.1.17 McIlree, A. R., Degradation of High Strength Austenitic Alloys X-750, 718 and A286 in Nuclear Power Systems, 1st International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE, 1984. 7.1.18 U.S. Nuclear Regulatory Commission, A Review of Stress Corrosion Cracking of High-Level Nuclear Waste Container Materials, CNWRA 92-021, August 1992. 7.1.19 Rowland, M. C. and W. R. Smith, Sr., Precipitation-Hardening Stainless Steels in Water-Cooled Reactors, Nuclear Engineering, (January 1962), pp. 14-22. 7.1.20 Xu, H. and S. Fyfitch, Aging Embrittlement Modeling of Type 17-4 PH at LWR Temperature, Proceedings of the 10th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 3-9, 2001. 7.1.21 Electric Power Research Institute, Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175), EPRI #1012081, 2005, Palo Alto, CA. 7.1.22 U.S. Nuclear Regulatory Commission, Degradation of LWR Core Internal Materials due to Neutron Irradiation, NUREG/CR-7027, December 2010. 7.1.23 Teysseyre, S., et al., Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X-750 and XM-19 Exposed to BWR Environments, Presentation at International Light Water Reactor Materials Reliability Conference and Exhibition, August 1-4, 2016, Chicago, Illinois. 7.1.24 U.S. Nuclear Regulatory Commission, Control of the Processing and Use of Stainless Steel, Regulatory Guide 1.44, Rev. 1, March 2011. 7.1.25 Electric Power Research Institute, "Pressurized Water Reactor Primary Water Chemistry Guidelines," Technical Report 3002000505, Rev. 7, April 2014, Palo Alto, CA. 7.1.26 U.S. Nuclear Regulatory Commission, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors, Regulatory Guide 1.207, March 2007. © Copyright 2019 by NuScale Power, LLC 63

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 7.1.27 U.S. Nuclear Regulatory Commission, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, NUREG/CR-6909, Rev. 1, Draft Report for Comment, March 2014. © Copyright 2019 by NuScale Power, LLC 64

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Appendix A. Containment Isolation Summary Figures The following figures are provided to show the isolation valves and closed loop piping systems which form part of the containment pressure boundary for the GDC 55, 56 and 57 piping penetrations of the CNV. Collectively, these figures identify all of the fluid service penetrations of the NuScale containment. Table A-1 Simplified figures based on the following piping and instrumentation diagrams A-1 NP12-01-A013-M-PD-3450 Containment system A-2 NP12-01-A030-M-PD-3451 Steam generator system A-3 NP12-01-B030-M-PD-1028 Decay heat removal system © Copyright 2019 by NuScale Power, LLC 65

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 All CIVs are located on top of the CNV head. All PSCIVs are Type C tested. The SSCIVs are the MSIVs / bypass and FWIVs located on GDC 57 SGS lines. The SSCIVs are leak tested per technical specifications for DHRS operability. Figure A-1 Containment system © Copyright 2019 by NuScale Power, LLC 66

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 The SGS is a GDC 57 closed loop. main steam, feedwater, and decay heat removal lines inside CNV up to and including the MSIVs and FWIVs are ASME Code Class 2. All SG lines (SG tubes) inside RPV are ASME Code Class 1. Figure A-2 Steam generator system © Copyright 2019 by NuScale Power, LLC 67

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 The entire DHRS boundary is ASME Code Class 2. The DHRS meets ANSI/ANS 56.2 criteria for a closed loop outside containment. Figure A-3 Decay heat removal system © Copyright 2019 by NuScale Power, LLC 68

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 A.1 Type B Containment Penetrations Penetration(1) Quantity of EPA Sheath Component(2) Nominal Size (Opening) Identification Modules CNV8 I&C Division 1 NPS 3 1 CNV9 I&C Division 2 NPS 3 1 Pressurizer Heater Power (Elect-CNV15 NPS 12 9 1) Pressurizer Heater Power (Elect-CNV16 NPS 12 9 2) CNV17 I&C Channel A NPS 8 4 CNV18 I&C Channel B NPS 8 4 CNV19 I&C Channel C NPS 8 4 CNV20 I&C Channel D NPS 8 4 CNV24 Head Manway NPS 18 N/A CNV25 CRDM Access Opening 67 inch N/A CNV26 Manway 38 inch N/A CNV27 SG Inspection Port 1 38 inch N/A CNV28 SG Inspection Port 2 38 inch N/A CNV29 SG Inspection Port 3 38 inch N/A CNV30 SG Inspection Port 4 38 inch N/A © Copyright 2019 by NuScale Power, LLC 69

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration(1) Quantity of EPA Sheath Component(2) Nominal Size (Opening) Identification Modules CNV31 Pressurizer Port 1 44 inch N/A CNV32 Pressurizer Port 2 44 inch N/A Reactor vent valve (RVV) Trip/Reset CNV33 NPS 3(4) N/A A(3) CNV34 RVV Trip/Reset B(3) NPS 3(4) N/A Reactor recirculation valve Trip/Reset CNV35 NPS 3(4) N/A A(3) Reactor recirculation valve Trip/Reset CNV36 NPS 3(4) N/A B(3) CNV37 CRDM Power NPS 18 16 CNV38 Rod position indication Group 1 NPS 10 4 CNV39 Rod position indication Group 2 NPS 10 4 CNV40 RVV Trip/Reset C-1(3) NPS 3(4) N/A CNV41 RVV Trip C-2(3) NPS 3(4) N/A

                   -                             CNV closure flange                             170 inch                                    N/A (1)  The penetration ID number CNV21 is not used.

(2) Type B test at containment design pressure. (3) RCPB valve. (4) CNV33-36 and CNV40 are 3-inch penetrations for ECCS trip and reset valves. Each penetration has two bolted connections (trip and reset valve) that each require a Type B test. CNV41 is a 3-inch penetration for a single ECCS trip valve. This penetration has one bolted connection. © Copyright 2019 by NuScale Power, LLC 70

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 A.2 Containment Piping Penetrations Penetration CNV1 CNV2 CNV3 CNV4 Identification Main Steam Main Steam Feedwater Feedwater Main Steam Main Steam Description isolation bypass isolation bypass isolation valve 1 isolation valve 2 isolation valve 1 isolation valve 2 valve 1 valve 2 Containment GDC 57 GDC 57 GDC 57 GDC 57 GDC 57 GDC 57 Basis Fluid Water Water Steam Steam Steam Steam Nominal Size NPS 5 (outlet) NPS 5 (outlet) NPS 12 NPS 2 NPS 12 NPS 2 (Opening) NPS 4 (inlet) NPS 4 (inlet) CIV Connection (Nozzle to Safe NPS 5 NPS 5 NPS 12 NPS 2 NPS 12 NPS 2 End) Type C No, TS leak test No, TS leak test No, TS leak test No, TS leak test No, TS leak test No, TS leak test Leakage Test Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter-Valve Type turn ball turn ball turn ball turn ball turn ball turn ball Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close Operator Type from onboard from onboard from onboard from onboard from onboard from onboard nitrogen nitrogen nitrogen nitrogen nitrogen nitrogen accumulator accumulator accumulator accumulator accumulator accumulator CNV head CNV head CNV head CNV head CNV head CNV head Location inboard inboard inboard(1) inboard(1) inboard(1) inboard(1) Normal Open Open Open Closed Open Closed Position Shutdown Closed Closed Closed Closed Closed Closed Position Safety Function Closed Closed Closed Closed Closed Closed Position © Copyright 2019 by NuScale Power, LLC 71

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV1 CNV2 CNV3 CNV4 Identification Failure Position Closed Closed Closed Closed Closed Closed Primary Automatic Automatic Automatic Automatic Automatic Automatic Actuation Secondary Remote manual Remote manual Remote manual Remote manual Remote manual Remote manual Actuation Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Power Source accumulator accumulator accumulator accumulator accumulator accumulator Design 2100 psia 2100 psia 2100 psia 2100 psia 2100 psia 2100 psia Pressure Design 650 F 650 F 650 F 650 F 650 F 650 F Temperature Seismic I I I I I I Category Design Code Section III, NC Section III, NC Section III, NC Section III, NC Section III, NC Section III, NC Valve (1) Single CIV. © Copyright 2019 by NuScale Power, LLC 72

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV5 CNV6 CNV7 Identification RCCW return RCCW return CVC injection CVC injection Pressurizerspray Pressurizerspray inboard outboard inboard outboard inboard outboard Description containment containment containment containment containment containment isolation valve isolation valve isolation valve isolation valve isolation valve isolation valve Containment GDC 57(1) GDC 57(1) GDC 55 GDC 55 GDC 55 GDC 55 Basis Fluid Water Water Water Water Water Water Nominal Size NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 (Opening) CIV Connection (Nozzle to Safe NPS 4 NPS 4 NPS 4 NPS 4 NPS 4 NPS 4 End) Type C Yes Yes Yes Yes Yes Yes Leakage Test Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter-Valve Type turn ball turn ball turn ball turn ball turn ball turn ball Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close Operator Type from onboard from onboard from onboard from onboard from onboard from onboard nitrogen nitrogen nitrogen nitrogen nitrogen nitrogen accumulator accumulator accumulator accumulator accumulator accumulator CNV Head CNV Head CNV Head CNV Head CNV Head CNV Head Location Inboard Outboard Inboard Outboard Inboard Outboard Normal Open Open Open Open Open Open Position Shutdown Closed Closed Closed Closed Closed Closed Position Safety Function Closed Closed Closed Closed Closed Closed Position Failure Position Closed Closed Closed Closed Closed Closed © Copyright 2019 by NuScale Power, LLC 73

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV5 CNV6 CNV7 Identification Primary Automatic Automatic Automatic Automatic Automatic Automatic Actuation Secondary Remote manual Remote manual Remote manual Remote manual Remote manual Remote manual Actuation Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Power Source accumulator accumulator accumulator accumulator accumulator accumulator Design 2100 psia(2) 2100 psia(2) 2100 psia 2100 psia 2100 psia 2100 psia Pressure Design 650 F(2) 650 F(2) 650 F 650 F 650 F 650 F Temperature Seismic I I I I I I Category Design Code Section III, NB(2) Section III, NB(2) Section III, NB Section III, NB Section III, NB Section III, NB Valve (1) RCCW/CRDS lines in containment meet the criteria of GDC 57; however, these lines are not credited as a containment boundary. (2) Valve provides a containment boundary and is classified as ASME Code Class 2 with a minimum design pressure and temperature requirement equivalent to the CNTS (1,000 psia, 550 F). However, all PSCIVs are designed to ASME Section III NB with a design pressure and temperature requirement equivalent to the RCS (2100 psia, 650 F). © Copyright 2019 by NuScale Power, LLC 74

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV10 CNV11 CNV12 Identification Containment Containment Containment Containment RCCW supply RCCW supply evacuation evacuation flooding inboard flooding outboard inboard outboard Description inboard outboard containment containment containment containment containment containment isolation valve isolation valve isolation valve isolation valve isolation valve isolation valve Containment GDC 56 GDC 56 GDC 56 GDC 56 GDC 57(1) GDC 57(1) Basis Noncondensable Noncondensable Fluid Water Water Water Water gas and vapor gas and vapor Nominal Size NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 (Opening) CIV Connection (Nozzle to Safe NPS 4 NPS 4 NPS 4 NPS 4 NPS 4 NPS 4 End) Type C Yes Yes Yes Yes Yes Yes Leakage Test Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter-Valve Type turn ball turn ball turn ball turn ball turn ball turn ball Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close nitrogen to close Operator Type from onboard from onboard from onboard from onboard from onboard from onboard nitrogen nitrogen nitrogen nitrogen nitrogen nitrogen accumulator accumulator accumulator accumulator accumulator accumulator CNV head CNV head CNV head CNV head CNV head CNV head Location inboard outboard inboard outboard inboard outboard Normal Open Open Closed Closed Open Open Position Shutdown Closed Closed Closed Closed Closed Closed Position Safety Function Closed Closed Closed Closed Closed Closed Position © Copyright 2019 by NuScale Power, LLC 75

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV10 CNV11 CNV12 Identification Failure Position Closed Closed Closed Closed Closed Closed Primary Automatic Automatic Automatic Automatic Automatic Automatic Actuation Secondary Remote manual Remote manual Remote manual Remote manual Remote manual Remote manual Actuation Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Nitrogen Power Source accumulator accumulator accumulator accumulator accumulator accumulator Design 2100 psia(1) 2100 psia(1) 2100 psia(1) 2100 psia(1) 2100 psia(1) 2100 psia(1) Pressure Design 650 F(1) 650 F(1) 650 F(1) 650 F(1) 650 F(1) 650 F(1) Temperature Seismic I I I I I I Category Design Code Section III, NB(2) Section III, NB(2) Section III, NB(2) Section III, NB(2) Section III, NB(2) Section III, NB(2) Valve (1) Valve provides a containment boundary and is classified as ASME Code Class 2 with a minimum design pressure and temperature requirement equivalent to the CNTS (1,050 psia, 550 F). However, all PSCIVs are designed to ASME Section III NB with a design pressure and temperature requirement equivalent to the RCS (2100 psia, 650 F). © Copyright 2019 by NuScale Power, LLC 76

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV13 CNV14 CNV22 CNV23 Identification RPV high point RPV high point Decay heat CVC discharge CVC discharge Decay heat degasification degasification removal system inboard outboard removal system Description inboard outboard Train 2 containment containment train 1 condensate containment containment condensate isolation valve isolation valve penetration isolation valve isolation valve penetration Containment GDC 55 GDC 55 GDC 55 GDC 55 GDC 57(1) GDC 57(1) Basis Noncondensable Noncondensable Fluid Water Water Water Water gas and steam gas and steam Nominal Size NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 NPS 2 (Opening) CIV Connection (Nozzle to Safe NPS 4 NPS 4 NPS 4 NPS 4 N/A N/A End) Type C Yes Yes Yes Yes N/A N/A Leakage Test Wedged, quarter- Wedged, quarter- Wedged, quarter- Wedged, quarter-Valve Type N/A N/A turn ball turn ball turn ball turn ball Hydraulic to open, Hydraulic to open, Hydraulic to open, Hydraulic to open, nitrogen to close nitrogen to close nitrogen to close nitrogen to close Operator Type from onboard from onboard from onboard from onboard N/A N/A nitrogen nitrogen nitrogen nitrogen accumulator accumulator accumulator accumulator CNV head CNV head CNV head CNV head Location N/A N/A inboard outboard inboard outboard Normal Open Open Closed Closed N/A N/A Position Shutdown Closed Closed Closed Closed N/A N/A Position Safety Function Closed Closed Closed Closed N/A N/A Position © Copyright 2019 by NuScale Power, LLC 77

NuScale Containment Leakage Integrity Assurance Technical Report TR-1116-51962-NP Rev. 1 Penetration CNV13 CNV14 CNV22 CNV23 Identification Failure Position Closed Closed Closed Closed N/A N/A Primary Automatic Automatic Automatic Automatic N/A N/A Actuation Secondary Remote manual Remote manual Remote manual Remote manual N/A N/A Actuation Nitrogen Nitrogen Nitrogen Nitrogen Power Source N/A N/A accumulator accumulator accumulator accumulator Design 2100 psia 2100 psia 2100 psia 2100 psia 2100 psia 2100 psia Pressure Design 650 F 650 F 650 F 650 F 650 F 650 F Temperature Seismic I I I I I I Category Design Code Section III, NB Section III, NB Section III, NB Section III, NB Section III, NC Section III, NC Valve (1) DHRS lines have the attributes of both a closed loop inside and outside of containment. An exemption is provided to clarify the system design within GDC 57 criteria. CNV3, CNV4, CNV22 and CNV23 are the DHRS lines that penetrate containment. The DHRS becomes a closed system outside containment when the FWIVs and MSIVs shut, creating the DHRS boundary. The test for this system is the leakage test of the FWIVs, MSIVs and main steam bypass valves in accordance with the ASME OM for Category A valves. © Copyright 2019 by NuScale Power, LLC 78

LO-0419-65333 : Affidavit of Zackary W. Rad, AF-0419-65334 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the method by which NuScale develops its containment leakage integrity assurance. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled NuScale Containment Leakage Integrity Assurance. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies AF-0419-65334 Page 1 of 2

upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on May 28, 2019. Zackary W. Rad AF-0419-65334 Page 2 of 2}}

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