ML19283E530

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LLC Supplemental Response to NRC Request for Additional Information No. 522 (Erai No. 9681) on the NuScale Design Certification Application
ML19283E530
Person / Time
Site: NuScale
Issue date: 10/10/2019
From: Melton M
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-1019-67575
Download: ML19283E530 (63)
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Text

RAIO-1019-67575 October 10, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Supplemental Response to NRC Request for Additional Information No. 522 (eRAI No. 9681) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 522 (eRAI No. 9681)," dated July 15, 2019
2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 522 (eRAI No.9681)," dated July 19, 2019 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9681:

14.03-3 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions on this response, please contact Nadja Joergensen at 541-452-7338 or at njoergensen@nuscalepower.com.

Sincerely, Michael Melton Manager, Licensing NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Cayetano Santos, NRC, OWFN-8H12 Michael Dudek, NRC, OWFN-8H12 : NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9681 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-1019-67575 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9681 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9681 Date of RAI Issue: 05/30/2019 NRC Question No.: 14.03-3 Please see the attachment to this Request for Additional Information.

Title 10, Section 52.47(b)(1) of the Code of Federal Regulations (CFR) requires that a design certification application contain the proposed inspections, tests, analyses, and acceptance criteria (ITAAC) that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a facility that incorporates the design certification has been constructed and will operate in accordance with the design certification, the provisions of the Atomic Energy Act of 1954, as amended (AEA),

and the NRC's rules and regulations. For the ITAAC to be "sufficient," (1) the inspections, tests, and analyses (ITA) must clearly identify those activities necessary to demonstrate that the acceptance criteria (AC) are met; (2) the AC must state clear design or performance objectives demonstrating that the Tier 1 design commitments (DCs) are satisfied; (3) the ITA and AC must be consistent with each other and the Tier 1 DC; (4) the ITAAC must be capable of being performed and satisfied prior to fuel load; and (5) the ITAAC, as a whole, must provide reasonable assurance that, if the ITAAC are satisfied, the facility has been constructed and will be operated in accordance with the design certification, the AEA, and the NRC's rules and regulations.

The staff has reviewed all DCD Rev 2, Tier 1 ITAAC tables and Chapter 1 of Tier 1 against these objectives, and in light of NRC guidance, Commission policy, and lessons learned from plants that are currently under construction that are in the process of implementing ITAAC.

Based on this review, the staff has compiled the attached list of proposed ITAAC wording changes. The applicant is requested to make these changes in the Tier 1 ITAAC tables and in Chapter 1 of Tier 1, or otherwise show that the ITAAC comply with 10 CFR 52.47(b)(1).

Additionally, the applicant is requested to address the following items, or otherwise show that the ITAAC comply with 10 CFR 52.47(b)(1):

NuScale Nonproprietary

1. ITAAC 29 in Table 2.5-7 verifies that the MCR isolation switches are located in the remote shutdown station but it does not verify the functionality of the switches. Please explain how ITAAC 29 verifies that the MCR isolation switches actually isolate the manual MCR switches from the MPS in case of fire. If ITAAC 29 does not verify the functionality of the MCR isolation switches, please explain what changes to the existing ITAAC in Tier 1 would be necessary to verify the functionality of the MCR isolation switches through ITAAC. If the applicant believes that ITAAC are not necessary to verify the functionality of the MCR isolation switches, please explain this and please explain why an ITAAC is, nonetheless, necessary to verify the location of the MCR isolation switches.
2. The design commitments listed in the design descriptions of DCA Part 2, Tier 1 are not consistent with the design commitments in the corresponding ITAAC tables. Although not identified in the attachment, the design commitments in the design descriptions of DCA Part 2, Tier 1 should be revised to be consistent with the design commitment in the ITAAC tables.

Additional explanations for the basis of the staff's proposed revisions in the attachment are provided below:

1. Tier 1, Section 1.1: Propose adding a definition of "approved design" to clarify what this term refers to. Without a definition, it is not clear who the approver is or when the design is considered approved (at certification or when the ITAAC is closed?). To provide clarity and flexibility, the staff proposes to define the "approved design" in terms of the updated final safety analysis report.
2. Tier 1, Section 1.2.4: Propose adding explanatory material consistent with past design certifications as applied to the NuScale design.
3. ITAAC 12 in Table 2.1-4: To resolve the use of the ambiguous word, "approximately" in the AC.
4. ITAAC 22 in Table 2.1-4: To clarify the applicability of the ITAAC to the assemblies and to add consideration of overload currents.

NuScale Nonproprietary

5. ITAAC 1 and 2 in Table 2.3-1: To make the scope of the ITA and AC consistent with the DC.
6. ITAAC 3, 4, and 6 in Table 2.5-7: To clarify the applicability of physical separation, electrical isolation, and communications independence in the DC and ITAAC.
7. ITAAC 15 in Table 2.5-7: To clarify the DC and make the DC consistent with the AC.
8. ITAAC 21 in Table 2.5-7: To clarify the DC and resolve an inconsistency between the DC and AC.
9. ITAAC 2, 3, and 4 in Table 2.7-2: The DCs for ITAAC 2 to 4 relate to a single Chemical Volume and Control System (CVCS) high radiation signal, but the AC for each ITAAC cover all 3 CVCS radiation signals. The proposed changes consolidate ITAAC 2 to 4 so that the scope of the DC matches the scope of the AC.
10. ITAAC 1 in Table 3.4-1: To resolve an inconsistency between the DC and AC.
11. ITAAC 4 in Table 3.4-1: The DC is actually an ITA. The staff's proposed revisions correct this.
12. ITAAC 2 in Table 3.5-1: To remove an unnecessary conditional statement in the DC and to clarify what the "approved" analysis is.
13. ITAAC 3 in Table 3.7-1: To clarify in the AC the alternative shutdown capability referred to in the DC.
14. ITAAC 4, 5, and 6 in Table 3.9-2: See explanation for ITAAC 2, 3, and 4 in Table 2.7-2.
15. ITAAC 8 and 9 in Table 3.9-2: See explanation for ITAAC 2, 3, and 4 in Table 2.7-2.
16. ITAAC 1, 2, and 3 in Table 3.10-1: To resolve inconsistencies between the DC and AC.

NuScale Nonproprietary

17. ITAAC 7 in Table 3.10-1: The DC is actually an ITA. The staff's proposed revisions correct this and make it consistent with the AC.
18. ITAAC 8 in Table 3.10-1: This ITAAC could be deleted if the proposed revisions to ITAAC 7 in Table 3.10-1 are incorporated as shown in the attachment since the scope of the revised ITAAC 7 would encompass the scope of ITAAC 8.
19. ITAAC 10 in Table 3.10-1: To resolve inconsistencies between the DC and AC.
20. ITAAC 5 in Table 3.11-2: To remove unnecessary and ambiguous qualifying language in the AC.
21. ITAAC 2 in Table 3.12-2: To remove unnecessary and ambiguous qualifying language in the AC.
22. ITAAC 7 and 8 in Table 3.16-1: To make the scope of the ITAAC consistent among the DC, ITA, and AC.
23. ITAAC 9 in Table 3.16-1: To clarify the scope of the ITA.
24. ITAAC 10 in Table 3.16-1: To make the scope of the ITA and AC consistent with the DC.
25. ITAAC 2, 3, and 4 in Table 3.17-2: See explanation for ITAAC 2, 3, and 4 in Table 2.7-2.
26. ITAAC 2 and 3 in Table 3.18-2: See explanation for ITAAC 2, 3, and 4 in Table 2.7-2.

NuScale Response:

Staff provided observations and comments regarding the NuScale response to request for additional information (RAI) 9681. Each of the comments and observations are listed below, and are immediately followed by the actions taken by NuScale to address them.

Table 2.1-4 ITAAC #12 NuScale Nonproprietary

RAI 9681 response added the language requested except for the phrase, "the objectives of" in the design commitment (DC). This phrase was included in the acceptance criteria (AC). The design commitment in Section 2.1 is also lacking this phrase.

The DC for ITAAC 02.01.12 was revised to align with the AC. Both the DC and AC now include the phrase "with the objectives of the RPV surveillance program." The DC in Section 2.1 was also revised to exactly match the revised Inspections, Test, Analyses, and Acceptance Criteria (ITAAC) DC.

Table 2.2-3 ITAAC #1 The phrase "listed in Table 2.2-1" was added to the AC but it appears to be in the wrong location. The wrong "system" appears to have been deleted in the AC.

The AC for ITAAC 02.02.01 was revised to correct both errors.

Table 2.3-1 ITAAC #2 Per RAI response, staff comment in the AC was not incorporated because "The acceptance criteria was not revised to add pressure instrumentation. The acceptance criteria already specifies 'CES inlet pressure instrumentation (PIT-1001/PIT-1019)' " The AC is still unclear.

Staff requests revised ITAAC language. DC - "The CES inlet pressure instrumentation supports RCS leakage detection" ITA - "A test will be performed of the CES inlet pressure instrumentation" AC - "The CES inlet pressure instrumentation detects a pressure increase which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour" During the public meeting to discuss the NRC Staff's observations and comments, NuScale explained that the requested changes would introduce inconsistencies between the revised ITAAC language and ITAAC 02.03.01. To prevent these inconsistencies, both ITAAC 02.03.01 and ITAAC 02.03.02 were revised to incorporated the requested changes.

Table 2.5-7 ITAAC #1 There appear to be several cases in which the text for design commitments listed in Section 2.5.1 do not match DC text in ITAAC table (see design commitments for 1i, viii, ix). The text for design commitments listed in Section 2.5.1 are listed in a different order than the DC in the ITAAC Table.

NuScale Nonproprietary

Design Commitments in both Section 2.5.1 and Table 2.5-7 were revised to ensure alignment of the text. Additionally, the order of the DCs in Section 2.5.1 was modified to match the order in Table 2.5-7.

Table 2.5-7 ITAAC #21 Response to RAI 9681 incorporates this as item number xiii in ITAAC #1 in Table 2.5-7. The DC uses the term "separation channels." The AC uses the term "separation groups."

Consistent terms should be used. Staff requests that the DC be revised to "The MPS is capable of performing its safety-related functions when any one of its separation groups is out of service."

The term "separation channels" was changed to "separation groups" in the DC for ITAAC 02.05.01xiii. and the corresponding Design Commitment in Section 2.5.1.

Table 3.1-2 ITAAC #1 The DC is mathematically inconsistent with the AC because "does not exceed" and "is less than" do not mean the same thing. If the measured value equals the assumed value, then the DC is met, but the AC is not. Staff requests revised ITAAC language. DC - "The air exfiltration out of the CRE is less than or equal to the assumptions used to size the CRHS inventory and supply flow rate." AC - "The air exfiltration measured by tracer gas testing is less than or equal to the CRE air infiltration rate assumed in the dose analysis."

The phrase "does not exceed" was revised to "is less than or equal to" in the DC for ITAAC 03.01.01 and the corresponding DC in Section 3.3.1. Additionally, the AC for ITAAC 03.01.01 was revised to modify "less than" to "less an equal to".

Table 3.9-2 ITAAC #2 There is a slight discrepancy between the DC in the ITAAC Table and the design commitment listed in Section 3.9.1. In the table the DC states ". . CRHS automatically respond to the high-radiation signals . . . " whereas in Section 3.9.1 it states ". . CRHS automatically respond to the CRVS high radiation signals. . . "

The ITAAC 03.09.02 DC was revised to align with the corresponding DC in Section 3.9.1.

Table 3.9-2 ITAAC #10 NuScale Nonproprietary

In the AC for other ITAAC (e.g., ITAAC #7 and #8 in Table 3.9-2), NuScale replaced "a" real or simulated signal with "the" real or simulated signal. However, for this ITAAC this change was not made.

The article "a" was revised to "the" in the ITAAC 03.09.10 AC.

Table 3.14-2 ITAAC #1 The change to Seismic Qualification Report was suggested by NRC staff in RAI 9681. The edit should have been applied throughout the ITAAC (i.e. 02.08.01 and the rest of 03.14.01) and was suggested to ensure consistency with ASME QME-1 terminology and to be clear in what the AC required.

The AC of ITAAC 02.08.01 and 03.14.01 were revised to incorporate the requested change.

Additionally, the corresponding ITAAC discussions in Tier 2 Tables 14.3-1 and 14.3-2 were revised to align with the revised AC.

Table 3.17-2 ITAAC #2 This ITAAC refers to Table 3.17-1. In Table 3.17-1 "BPDS" should be identified in the second row, under the column for "Variable Monitored" The entry for "0A condensate polishing system regeneration skid waste effluent" in Table 3.17-1 was revised to "BPDS 0A condensate polishing system regeneration skid waste effluent".

Additionally, the Table 3.18-1 entry for "0B condensate polishing system regeneration skid waste effluent" was revised to "BPDS 0B condensate polishing system regeneration skid waste effluent" for consistency.

Table 14.3-2 In Table 14.3-2 the entry for ITAAC number 03.14.08 references Section 5.4.2 twice for the decay heat removal system (DHRS); however, it is DCA Part 2, Tier 2, Section 5.4.3, that covers the DHRS. Per the public meeting on 9/4/19, staff understands that NuScale plans to move this ITAAC back to Table 2.8-2.

The Table 14.3-2 discussion for ITAAC 03.14.08 was revised to correctly reference Section 5.4.3 in both instances. As noted, several ITAAC were transferred from Tier 1 Chapter 2 to Tier 1 Chapter 3 as part of the initial RAI 9681 response, but have restored to their original location NuScale Nonproprietary

as part of this supplemental response. The following ITAAC, associated DCs, and Tier 2 Table 14.3-1 discussions were reinstated:

  • Table 2.1-4, ITAAC 02.01.22
  • Table 2.8-2, ITAAC 02.08.03
  • Table 2.8-2, ITAAC 02.08.05
  • Table 2.8-2, ITAAC 02.08.06
  • Table 2.8-2, ITAAC 02.08.08 The following ITAAC, associated DCs, and Tier 2 Table 14.3-2 discussions were removed:
  • Table 3.14-2, ITAAC 03.14.04
  • Table 3.14-2, ITAAC 03.14.05
  • Table 3.14-2, ITAAC 03.14.06
  • Table 3.14-2, ITAAC 03.14.07
  • Table 3.14-2, ITAAC 03.14.08 Table 14.3-3h In Tables 2.8-1 and 14.3-3h, both titled Module Specific Mechanical and Electrical/I&C Equipment, under Control Rod Drive System, Rod Position Indication (RPI) Coils it states there are 24 total, but should this be 32 total (16 control rod assemblies x 2 RPI trains/assembly)?

The Rod Position Indication (RPI) Coils entries in Tables 2.8-1 and 14.3-3h were revised from "24 total" to "32 total".

An additional change was made to Radioactive Waste Building ITAAC 03.12.03 and its associated ITAAC discussion in Tier 2 Table 14.3-2. The changes made to the ITA and AC establish consistency with similar Reactor Building and Control Building ITAAC, 03.11.06 and 03.13.04, respectively.

Changes were also made to test abstracts in the Initial Test Program. The changes include:

  • Table 14.2-47, Emergency Core Cooling System Test #47, System Level Test #47-1, was revised to remove test objectives and acceptance criteria associated with containment vessel pressure and temperature. These parameters are not required to demonstrate the functional operability of the emergency core cooling system as a whole, and were therefore deleted.

NuScale Nonproprietary

  • Table 14.2-107, Remote Shutdown Workstation Test #107, was restored at the request of the Staff. This change permits future reviewers to more easily identify how the remote shutdown station functionality is demonstrated via factory and site acceptance testing.

Changes were made to Table 14.3-2 ITAAC No. 03.11.14 and ITAAC No. 03.12.01 to include additional relevant requirements in the discussion section.

Impact on DCA:

Tier 1 Chapters 2 and 3, and Tier 2 Chapter 14 Sections 14.2 and 14.3 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Tier 2 NuScale Final Safety Analysis Report RAI 09.01.04-1, RAI 09.05.01-6, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.02-1, RAI 14.03.02-2, RAI 14.03.03-1, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.03-8, RAI 14.03.07-1, RAI 14.03.08-1S1, RAI 14.03.09-1, RAI 14.03.09-2, RAI 14.03.09-3, RAI 14.03.12-2, RAI 14.03.12-3, RAI 18-46S1 Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, X Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, to demonstrate that air exfiltration from the CRE is controlled. RG 1.197 allows two options for CRE testing; either integrated testing (tracer gas testing) or component testing. Section 6.4 Control Room Habitability, describes the testing requirements for the CRE habitability program. Section 6.4 provides the maximum air exfiltration allowed from the CRE.

In accordance with Table 14.2-18, a preoperational test using the tracer gas test method demonstrates that the air exfiltration from the CRE does not exceed the assumed unfiltered leakage rate provided in Table 6.4-1: Control Room Habitability System Design Parameters 14.3-59 for the dose analysis. Tracer gas testing in accordance with ASTM Certified Design Material and Inspections, Tests, Analyses, and E741 will be performed to measure the unfiltered in-leakage into the CRE with the control room habitability system (CRHS) operating.

03.01.02 CRH The CRHS valves are tested by remote operation to demonstrate the X capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18, a preoperational test demonstrates that each CRHS valve listed in Tier 1 Table 3.1-1 (Table 14.3-4a) strokes fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate Acceptance Criteria design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

Draft Revision 4

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.04 RXB Section 12.3, Radiation Protection Design Features, provides the X design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone and access requirements for normal operation and post-accident conditions, and to demonstrate compliance with 10 CFR 50.49, GDC 4, and PDC 19, GDC 61, 10 CFR 50.34(f)(2)(vii), and other relevant requirements.

Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed of the RXB radiation barriers to verify wall materials and thicknesses. The required thicknesses are specified in Table 12.3-6. Attenuation capabilities are determined based on wall materials and thicknesses, and an analysis and report will conclude that attenuation capabilities are greater than or equal to the approved design.

14.3-80 03.11.05 RXB Section 12.3.2.2, Design Considerations, provides the design bases X Certified Design Material and Inspections, Tests, Analyses, and for radiation shielding. Radiation shielding is provided to meet the radiation zone requirements for normal operation and control room access requirements for post-accident conditions. Radiation attenuating doors must meet or exceed the radiation attenuation capability of the wall within which they are installed.

An ITAAC inspection is performed to verify that the RXB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.

Acceptance Criteria Draft Revision 4

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.08 RXB Section 3.6, Protection against Dynamic Effects Associated with X X Postulated Rupture of Piping, provides the design bases and criteria for the analysis required to demonstrate that safety-related SSC are not impacted by the adverse effects of a high-and moderate-energy pipe failure within the plant.

An ITAAC inspection is performed to verify that the as-built protective features located in the RXB outside the reactor pool bay credited in the reconciled Pipe Break Hazards Analysis Report (such as pipe whip restraints, pipe whip or jet impingement barriers, jet impingement shields, or guard pipe) have been installed in accordance with design drawings of sufficient detail to show the existence and location of the protective hardware. The as-built inspection is intended to verify that changes to postulated pipe failure locations and protective features or protected equipment made during construction do not adversely affect the safety-related 14.3-84 functions of the protected equipment.

Certified Design Material and Inspections, Tests, Analyses, and 03.12.01 RWB Section 12.3, Radiation Protection Design Features, provides the X design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone requirement for normal operation and post-accident conditions and to demonstrate conformance with GDC 61, RG 4.21, and RG 8.8, and other relevent requirements. Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed of the RWB radiation barriers to verify wall materials and thicknesses. The required thicknesses are specified in Table 12.3-7. Attenuation capabilities are determined Acceptance Criteria based on wall materials and thicknesses, and an analysis and report will conclude that attenuation capabilities are greater than or equal to the approved design.

Draft Revision 4

NuScale Tier 1 NuScale Power Module

  • The Nuscale Power Module ASME Code Class 1, 2, and 3 components listed in Table 2.1-2 conform to the rules of construction of ASME Code Section III.

RAI 14.03-3

  • The Nuscale Power Module ASME Code Class CS components listed in Table 2.1-2 conform to the rules of construction of ASME Code Section III.
  • Safety-related structures, systems, and components (SSC) are protected against the dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.

RAI 14.03-3

  • The Nuscale Power Module ASME Code Class 2 piping systems listed in Table 2.1-1 and interconnected equipment nozzles are evaluated for leak-before-break (LBB).

RAI 14.03.03-8

  • The RPV beltline material has a Charpy upper-shelf energy of 75 ft-lb minimum.
  • The CNV serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.

RAI 14.03-3

  • Closure times for CIVs listed in Table 2.1-3 limit potential releases of radioactivity.

RAI 14.03-3

  • The length of piping listed in Table 2.1-1 shall be minimized between the containment penetration and the associated outboard CIVs.

RAI 08.01-1S1, RAI 14.03-3

  • The CNTS containment electrical penetration assemblies listed in Table 2.1-3 are sized to power their design loads.

RAI 14.03-3, RAI 14.03-3S1

  • The RPV is provided with surveillance capsule holders to hold a capsule containing RPV material surveillance specimens at locations where the capsules will be exposed to a neutron flux consistent with the objectives of the RPV surveillance program.

RAI 14.03-3

  • The remotely-operated CNTS containment isolation valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The ECCS valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The DHRS valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The CNTS hydraulic-operated valves listed in Table 2.1-2 fail to (or maintain) their safety-related position on loss of electrical power under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The ECCS RRVs and RVVs listed in Table 2.1-2 fail to (or maintain) their safety-related position on loss of electrical power to their corresponding trip valves under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3 Tier 1 2.1-4 Draft Revision 4

NuScale Tier 1 NuScale Power Module

  • The DHRS hydraulic-operated valves listed in Table 2.1-2 fail to (or maintain) their safety-related position on loss of electrical power under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The CNTS check valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.

RAI 14.03-3S1

  • Each CNTS containment electrical penetration assembly listed in Table 2.1-3 is rated either (i) to withstand fault and overload currents for the time required to clear the fault from its power source, or (ii) to withstand the maximum fault and overload current for its circuits without a circuit interrupting device.

RAI 14.03-3, RAI 14.03.07-1

  • The NPM lifting fixture supports its rated load.

RAI 14.03.07-1

  • The NPM lifting fixture is constructed to provide assurance that a single failure does not result in the uncontrolled movement of the lifted load.

RAI 14.03.03-5S3

  • The ECCS valves, CIVs, and DHRS actuation valves listed in Table 2.1-2, and their associated hydraulic lines, are installed such that each valve can perform its safety function.

2.1.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.1-4 contains the inspections, tests, and analyses for the NPM.

Tier 1 2.1-5 Draft Revision 4

NuScale Tier 1 NuScale Power Module RAI 06.02.06-22, RAI 06.02.06-23, RAI 08.01-1, RAI 08.01-1S1, RAI 08.01-2, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.03-3S1, RAI 14.03.03-4S1, RAI 14.03.03-5S3, RAI 14.03.03-6S1, RAI 14.03.03-7S1, RAI 14.03.03-8, RAI 14.03.03-11S1, RAI 14.03.07-1 Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The NuScale Power Module ASME An inspection will be performed of the The ASME Code Section III Design Code Class 1, 2 and 3 piping systems NuScale Power Module ASME Code Reports (NCA-3550) exist and listed in Table 2.1-1 comply with ASME Class 1, 2 and 3 as-built piping system conclude that the NuScale Power Code Section III requirements. Design Reports for systems listed in Module ASME Code Class 1, 2 and 3 as-Table 2.1-1 required by ASME Code built piping systems listed in Section III. Table 2.1-1 meet the requirements of ASME Code Section III.
2. The NuScale Power Module ASME An inspection will be performed of the ASME Code Section III Data Reports for Code Class 1, 2, and 3 components NuScale Power Module ASME Code the NuScale Power Module ASME listed in Table 2.1-2 conform to the Class 1, 2, and 3 as-built component Code Class 1, 2, and 3 components rules of construction of ASME Code Data Reports for components listed in listed in Table 2.1-2 and Section III. Table 2.1-2 required by ASME Code interconnecting piping exist and Section III. conclude that the requirements of ASME Code Section III are met.
3. The NuScale Power Module ASME An inspection will be performed of the ASME Code Section III Data Reports for Code Class CS components listed in NuScale Power Module ASME Code the NuScale Power Module ASME Table 2.1-2 conform to the rules of Class CS as-built component Data Code Class CS components listed in construction of ASME Code Section III. Reports for components listed in Table 2.1-2 exist and conclude that the Table 2.1-2 required by ASME Code requirements of ASME Code Section III Section III. are met.
4. Safety-related SSC are protected An inspection and analysis will be Protective features are installed in against the dynamic and performed of the as-built high- and accordance with the as-built Pipe environmental effects associated with moderate-energy piping systems and Break Hazard Analysis Report and postulated failures in high- and protective features for the safety-related SSC are protected moderate-energy piping systems. safety-related SSC. against or qualified to withstand the dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.
5. The NuScale Power Module ASME An analysis will be performed of the The as-built LBB analysis for the ASME Code Class 2 piping systems listed in ASME Code Class 2 as-built piping Code Class 2 piping systems listed in Table 2.1-1 and interconnected systems listed in Table 2.1-1 and Table 2.1-1 and interconnected equipment nozzles are evaluated for interconnected equipment nozzles. equipment nozzles is bounded by the LBB. as-designed LBB analysis.
6. The RPV beltline material has a Charpy A vendor test will be performed of the An ASME Code Certified Material Test upper-shelf energy of 75 ft-lb Charpy V-Notch specimen of the RPV Report exists and concludes that the minimum. beltline material. initial RPV beltline material Charpy upper-shelf energy is 75 ft-lb minimum.
7. The CNV serves as an essentially leak- A leakage test will be performed of the The leakage rate for local leak rate tight barrier against the uncontrolled pressure containing or leakage- tests (Type B and Type C) for pressure release of radioactivity to the limiting boundaries, and CIVs. containing or leakage-limiting environment. boundaries and CIVs meets the requirements of 10 CFR Part 50, Appendix J.
8. Closure times for CIVs listed in A test will be performed of the Each CIV listed in Table 2.1-3 travels Table 2.1-3 limit potential releases of automatic CIVs listed in Table 2.1-3. from the full open to full closed radioactivity. position in less than or equal to the time listed in Table 2.1-3 after receipt of a containment isolation signal.

Tier 1 2.1-11 Draft Revision 4

NuScale Tier 1 NuScale Power Module Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

9. The length of piping listed in An inspection will be performed of the The length of piping between each Table 2.1-1 shall be minimized as-built piping listed in Table 2.1-1 containment penetration and its between the containment penetration between containment penetrations associated outboard CIV is less than or and the associated outboard CIVs. and associated outboard CIVs. equal to the length identified in Table 2.1-1.
10. The CNTS containment electrical i. An analysis will be performed of i. An electrical rating report exists penetration assemblies listed in the CNTS as-designed that defines and identifies the Table 2.1-3 are sized to power their containment electrical penetration required design electrical rating to design loads. assemblies listed in Table 2.1-3. power the design loads of each CNTS containment electrical penetration assembly listed in Table 2.1-3.

ii. An inspection will be performed of ii. The electrical rating of each CNTS CNTS as-built containment containment electrical penetration electrical penetration assemblies assembly listed in Table 2.1-3 is listed in Table 2.1-3. greater than or equal to the required design electrical rating as specified in the electrical rating report.

11. Not used. Not used. Not used.
12. The RPV is provided with surveillance An inspection will be performed of the Four surveillance capsule holders are capsule holders to hold a capsule as-built RPV surveillance capsule installed in the RPV beltline region at containing RPV material surveillance holders. locations where the capsules will be specimens at locations where the exposed to a neutron flux consistent capsules will be exposed to a neutron with the objectives of the RPV flux consistent with the objectives of surveillance program.

the RPV surveillance program.

13. The remotely-operated CNTS A test will be performed of the Each remotely-operated CNTS containment isolation valves listed in remotely-operated CNTS containment containment isolation valve listed in Table 2.1-2 change position under isolation valves listed in Table 2.1-2 Table 2.1-2 strokes fully open and fully design-basis temperature, differential under preoperational temperature, closed by remote operation under pressure, and flow conditions. differential pressure, and flow preoperational temperature, conditions. differential pressure, and flow conditions.
14. The ECCS valves listed in Table 2.1-2 A test will be performed of the ECCS Each ECCS valve listed in Table 2.1-2 change position under design-basis valves listed in Table 2.1-2 under strokes fully open and fully closed by temperature, differential pressure, and preoperational temperature, remote operation under flow conditions. differential pressure, and flow preoperational temperature, conditions. differential pressure, and flow conditions.
15. The DHRS valves listed in Table 2.1-2 A test will be performed of the DHRS Each DHRS valve listed in Table 2.1-2 change position under design-basis valves listed in Table 2.1-2 under strokes fully open and fully closed by temperature, differential pressure, and preoperational temperature, remote operation under flow conditions. differential pressure, and flow preoperational temperature, conditions. differential pressure, and flow conditions.
16. Not used. Not used. Not used.
17. Not used. Not used. Not used.

Tier 1 2.1-12 Draft Revision 4

NuScale Tier 1 NuScale Power Module Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

18. The CNTS hydraulic-operated valves A test will be performed of the CNTS Each CNTS hydraulic-operated valve listed in Table 2.1-2 fail to (or maintain) hydraulic-operated valves listed in listed in Table 2.1-2 fails to (or their safety-related position on loss of Table 2.1-2 under preoperational maintains) its safety-related position electrical power under design-basis temperature, differential pressure, and on loss of motive power under temperature, differential pressure, and flow conditions. preoperational temperature, flow conditions. differential pressure, and flow conditions.
19. The ECCS RRVs and RVVs listed in A test will be performed of the ECCS Each ECCS RRV and RVV listed in Table 2.1-2 fail to (or maintain) their RRVs and RVVs listed in Table 2.1-2 Table 2.1-2 fails to (or maintains) its safety-related position on loss of under preoperational temperature, safety-related position on loss of electrical power to their differential pressure, and flow electrical power to its corresponding corresponding trip valves under conditions. trip valve under preoperational design-basis temperature, differential temperature, differential pressure, and pressure, and flow conditions. flow conditions.
20. The DHRS hydraulic-operated valves A test will be performed of the DHRS Each DHRS hydraulic-operated valve listed in Table 2.1-2 fail to (or maintain) hydraulic-operated valves listed in listed in Table 2.1-2 fails to (or their safety-related position on loss of Table 2.1-2 under preoperational maintains) its safety-related position electrical power under design-basis temperature, differential pressure, and on loss of motive power under temperature, differential pressure, and flow conditions. preoperational temperature, flow conditions. differential pressure, and flow conditions.
21. The CNTS check valves listed in A test will be performed of the CNTS Each CNTS check valve listed in Table 2.1-2 change position under check valves listed in Table 2.1-2 under Table 2.1-2 strokes fully open and design-basis temperature, differential preoperational temperature, closed (under forward and reverse pressure, and flow conditions. differential pressure, and flow flow conditions, respectively) under conditions. preoperational temperature, differential pressure, and flow conditions.
22. Each CNTS containment electrical An analysis will be performed of each For each CNTS containment electrical penetration assembly listed in CNTS as-built containment electrical penetration assembly listed in Table 2.1-3 is rated either (i) to penetration assembly listed in Table 2.1-3, either (i) a circuit withstand fault and overload currents Table 2.1-3.Not used. interrupting device coordination for the time required to clear the fault analysis exists and concludes that the from its power source, or (ii) to current carrying capability for the withstand the maximum fault and CNTS containment electrical overload current for its circuits without penetration assembly is greater than a circuit interrupting device.Not used. the analyzed fault and overload currents for the time required to clear the fault from its power source, or (ii) an analysis of the CNTS containment electrical penetration maximum fault and overload current exists and concludes the fault and overload current is less than the current carrying capability of the CNTS containment electrical penetration.Not used.
23. The CNV serves as an essentially A preservice design pressure leakage No water leakage is observed at CNV leaktight barrier against the test of the CNV will be performed. bolted flange connections.

uncontrolled release of radioactivity to the environment.

Tier 1 2.1-13 Draft Revision 4

NuScale Tier 1 Chemical and Volume Control System RAI 14.03-3, RAI 14.03-3S1 Table 2.2-3: Chemical and Volume Control System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The chemical and volume control An inspection will be performed of The ASME Code Section III Design Report system ASME Code Class 3 piping the chemical and volume control (NCA-3550) exists and concludes that listed in Table 2.2-1 complies with the system ASME Code Class 3 as-built the chemical and volume control ASME Code Section III. piping Design Report required by systemlisted in Table 2.2-1 ASME Code ASME Code Section III for piping Class 3 as-built piping systemlisted in listed in Table 2.2-1. Table 2.2-1 meets the requirements of ASME Code Section III.

2 The chemical and volume control An inspection will be performed of ASME Code Section III Data Reports for system ASME Code Class 3 the chemical and volume control the chemical and volume control system components listed in Table 2.2-2 system ASME Code Class 3 as-built ASME Code Class 3 components listed in conform to the rules of construction component Data Reports required by Table 2.2-2 and interconnecting piping of ASME Code Section III ASME Code Section III for exist and conclude that the components listed in Table 2.2-2. requirements of ASME Code Section III are met.

3 The chemical and volume control A test will be performed of the Each chemical and volume control system ASME Code Class 3 chemical and volume control system system ASME Code Class 3 air-operated air-operated demineralized water ASME Code Class 3 air-operated demineralized water system supply system supply isolation valves listed demineralized water system supply isolation valve listed in Table 2.2-2 in Table 2.2-2 change position under isolation valves listed in Table 2.2-2 strokes fully open and fully closed by design-basis temperature, differential under preoperational temperature, remote operation under preoperational pressure, and flow conditions. differential pressure, and flow temperature, differential pressure, and conditions. flow conditions.

4 Not used. Not used. Not used.

5 The chemical and volume control A test will be performed of the Each chemical and volume control system ASME Code Class 3 chemical and volume control system system ASME Code Class 3 air-operated air-operated demineralized water ASME Code Class 3 air-operated demineralized water system supply system supply isolation valves listed demineralized water system supply isolation valve listed in Table 2.2-2 its in Table 2.2-2 perform their function isolation valves listed in Table 2.2-2 function to fail to (or maintain) its to fail to (or maintain) their position under preoperational temperature, position performs on loss of motive on loss of motive power under differential pressure and flow power under preoperational design-basis temperature, differential conditions. temperature, differential pressure, and pressure, and flow conditions. flow conditions.

Tier 1 2.2-4 Draft Revision 4

NuScale Tier 1 Containment Evacuation System 2.3 Containment Evacuation System 2.3.1 Design Description

System Description

The scope of this section is the containment evacuation system (CES). Water vapor and non-condensable gases are removed from the containment vessel by the CES. The water vapor is collected and condensed into the CES sample vessel where it is monitored using level and temperature instrumentation. The CES pressure instrumentation and sample vessel level instrumentation is used to quantify and trend leak rates in the containment.

The CES is a nonsafety-related system. Each NuScale Power Module (NPM) has its own module-specific CES. The Reactor Building houses all CES equipment.

The CES performs the following nonsafety-related system functions that are verified by Inspections, Tests, Analyses, and Acceptance Criteria:

Design Commitments RAI 14.03-3S1

  • The CES sample vessel level instrumentation supports RCS leakage detection.
  • The CES inlet pressure instrumentation supports RCS leakage detection.

2.3.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.3-1 contains the inspections, tests, and analyses for the CES.

Tier 1 2.3-1 Draft Revision 4

NuScale Tier 1 Containment Evacuation System RAI 14.03-3, RAI 14.03-3S1 Table 2.3-1: Containment Evacuation System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The CES sample vessel level A test will be performed of the CES The CES sample vessel level instrumentation supports RCS leakage sample vessel level instrumention. instrumentation detects a level detection. increase in the CES sample vesseltank, which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour.
2. The CES inlet pressure instrumentation A test will be performed of the CES The CES inlet pressure instrumentation supports RCS leakage detection. inlet pressure instrumentation. detects a pressure increase in the CES inlet pressure instrumentation, which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour.

Tier 1 2.3-2 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System The primary purpose of the SDIS is to provide accurate, complete and timely information pertinent to MPS status and information displays. The SDIS provides display panels of MPS post-accident monitoring variables to support manually controlled protective actions if required.

The SDIS performs the following nonsafety-related system function that is verified by Inspections, Tests, Analyses, and Acceptance Criteria:

  • The SDIS supports the main control room (MCR) by providing displays of PAM Type B and Type C variables.

Design Commitments RAI 14.03-3S1

  • The MPS design and software are implemented using a quality process composed of the following system designsoftware lifecycle phases, with each phase having outputs whichthat satisfy the requirements of that phase:

system concept phase system requirements phase system design phase system implementation phase system test phase system installation and checkout phase

  • Protective measures are provided to restrict modifications to the MPS tunable parameters.

RAI 14.03-3S1

  • Communications independence exists between Separation Groups A, B, C, and D of the Class 1E MPS.

RAI 14.03-3S1

  • Communications independence exists between Divisions I and II of the Class 1E MPS.

RAI 14.03-3S1

RAI 14.03-3S1

  • The MPS automatically initiates an ESF actuation signal for ESF functions listed in Table 2.5-2.

RAI 14.03-3S1

RAI 14.03-3S1

RAI 14.03-3S1

  • The reactor trip logic fails to a safe state such that loss of electrical power to a MPS separation group results in a trip state for that separation group.

RAI 14.03-3S1 Tier 1 2.5-3 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System

  • The ESFs logic fails to a safe state such that loss of electrical power to a MPS separation group results in a safe state listed in Table 2.1-3.

RAI 14.03-3S1

  • The MPS interlocks listed in Table 2.5-4 automatically establish an operating bypass for the specified reactor trip or ESF actuations when the interlock condition is met, and the operating bypass is automatically removed when the interlock condition is no longer satisfied.

RAI 14.03-3S1

  • The MPS permissives listed in Table 2.5-4 allow the manual bypass of the specified reactor trip or ESF actuations when the permissive condition is met, and the operating bypass is automatically removed when the permissive condition is no longer satisfied.

RAI 14.03-3S1

  • The O-1 Override listed in Table 2.5-4 is established when the manual override switch is active and the RT-1 interlock is established. The Override switch must be manually taken out of Override when the O-1 Override is no longer needed.

RAI 14.03-3S1

  • The MPS is capable of performing its safety-related functions when any one of its separation groups is out of service.

RAI 14.03-3S1

  • The reactor trip breakers (RTBs) are installed and arranged as shown in Figure 2.5-2 in order to successfully accomplish the reactor trip function.

RAI 14.03-3S1

  • Two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.

RAI 14.03-3

  • Physical separation exists (i) between each separation group of the MPS Class 1E instrumentation and control current-carrying circuits, (ii) between each division of the MPS Class 1E instrumentation and control current-carrying circuits, and (iii) between Class 1E instrumentation and control current-carrying circuits and non-Class 1E instrumentation and control current-carrying circuits.

RAI 14.03-3

  • Electrical isolation exists (i) between each separation group of the MPS Class 1E instrumentation and control circuits, (ii) between each division of the MPS Class 1E instrumentation and control circuits, and (iii) between Class 1E instrumentation and control circuits and non-Class 1E instrumentation and control circuits to prevent the propagation of credible electrical faults.
  • Electrical isolation exists between the highly reliable DC power system-module-specific (EDSS-MS) subsystem non-Class 1E circuits and connected MPS 1E circuits to prevent the propagation of credible electrical faults.

RAI 14.03-3, RAI 14.03-3S1

  • Communications independence exists between Separation Groups A, B, C, and D of the Class 1E MPS.

RAI 14.03-3, RAI 14.03-3S1

  • Communications independence exists between Divisions I and II of the Class 1E MPS.
  • Communications independence exists between the Class 1E MPS and non-Class 1E digital systems.

Tier 1 2.5-4 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System RAI 14.03-3, RAI 14.03-3S1

RAI 14.03-3, RAI 14.03-3S1

  • The MPS automatically initiates an ESF actuation signal for ESF functions listed in Table 2.5-2.

RAI 14.03-3S1

RAI 14.03-3

  • The MPS automatically actuates the ESF equipment to perform its safety-related function listed in Table 2.5-2.

RAI 14.03-3S1

RAI 14.03-3

  • The MPS manually actuates the ESF equipment to perform its safety-related function listed in Table 2.5-2.

RAI 14.03-3S1

  • The reactor trip logic fails to a safe state such that loss of electrical power to an MPS separation group or division results in a trip state for that separation group or division.

RAI 14.03-3, RAI 14.03-3S1

  • The ESFs logic fails to a safe state such that loss of electrical power to an MPS separation group or division results in a safe state listed in Table 2.1-3.

RAI 14.03-3S1

  • An MPS signal, once initiated (automatically or manually), results in an intended sequence of protective actions that continue until completion, and requires deliberate operator action in order to return the safety systems to normal.

RAI 14.03-3S1

  • The MPS response times from sensor output through equipment actuation for the reactor trip functions and engineered safety featureESF functions are less than or equal to the value required to satisfy the design basis safety analysis response time assumptions.

RAI 14.03-3, RAI 14.03-3S1

  • The MPS interlocks listed in Table 2.5-4 automatically establish an operating bypass for the specified reactor trip or ESF actuations when the interlock condition is met, and the operating bypass is automatically removed when the interlock condition is no longer satisfied.

RAI 14.03-3, RAI 14.03-3S1

  • The MPS permissives listed in Table 2.5-4 allow the manual bypass of the specified reactor trip or ESF actuations when the permissive condition is met, and the operating bypass is automatically removed when the permissive condition is no longer satisfied.

RAI 14.03-3, RAI 14.03-3S1

  • The O-1 Override listed in Table 2.5-4 is established when the manual override switch is active and the RT-1 interlock is established. The Override switch must be manually taken out of Override when the O-1 Override is no longer needed.

Tier 1 2.5-5 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System RAI 14.03-3, RAI 14.03-3S1

  • The MPS is capable of performing its safety-related functions when any one of its separation channels is out of service.

RAI 14.03-3S1

  • The MPS operational bypasses are indicated in the MCR.

RAI 14.03-3S1

  • The MPS maintenance bypasses are indicated in the MCR.
  • The MPS self-test features detect faults in the system and provide an alarm in the MCR.
  • The PAM Type B and Type C displays are indicated on the SDIS displays in the MCR.
  • The controls located on the operator workstations in the MCR operate to perform important human actions (IHAs).

RAI 14.03-3, RAI 14.03-3S1

  • The reactor trip breakers (RTBs) are installed and arranged as shown in Figure 2.5-2 in order to successfully accomplish the reactor trip function.

RAI 14.03-3S1

  • Two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.

2.5.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.5-7 contains the inspections, tests, and analyses for the MPS and SDIS.

Tier 1 2.5-6 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System RAI 14.03-3, RAI 14.03-3S1 Table 2.5-7: Module Protection System and Safety Display and Indication System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. i. The MPS design and software i.a. An analysis will be performed of i.a. The output documentation of the are implemented using a the output documentation of the MPS Concept Phase satisfies the quality process composed of System Concept Phase. requirements of the System Concept the following system design Phase.

lifecycle phases, with each i.b. An analysis will be performed of i.b. The output documentation of the phase having outputs which the output documentation of the MPS Requirements Phase satisfies the satisfy the requirements of that System Requirements Phase. requirements of the System phase. Requirements Phase.

i.a. System Concept Phase i.c. An analysis will be performed of i.c. The output documentation of the i.b. System Requirements Phase the output documentation of the MPS Design Phase satisfies the i.c. System Design Phase System Design Phase. requirements of the System Design i.d. System Implementation Phase Phase.

i.e. System Test Phase i.d. An analysis will be performed of i.d. The output documentation of the i.f. System Installation and the output documentation of the MPS Implementation Phase satisfies Checkout Phase System Implementation Phase. the requirements of the System Implementation Phase.

i.e. An analysis will be performed of i.e. The output documentation of the the output documentation of the MPS Test Phase satisfies the System Test Phase. requirements of the System Test Phase.

i.f. An analysis will be performed of i.f. The output documentation of the the output documentation of the MPS Installation and Checkout Phase System Installation and Checkout satisfies the requirements of the Phase. System Installation and Checkout Phase.

ii. Protective measures are ii. Test will be performed on the ii. Protective measures restrict provided to restrict access control features modification to the MPS tunable modifications to the MPS associated with MPS tunable parameters without proper tunable parameters. parameters. configuration and authorization.

iii.a. Communications iii. A test will be performed of the iii.a. Communications independence independence exists between Class 1E MPS. between Separation Groups A, B, C, Separation Groups A, B, C, and and D of the Class 1E MPS is D of the Class 1E MPS. provided.

iii.b. Communications iii.b. Communications independence independence exists between between Division I and II of the Class Divisions I and II of the Class 1E 1E MPS is provided.

MPS.

iv. The MPS automatically initiates iv. A test will be performed of the iv. Reactor trip signal is automatically a reactor trip signal for reactor MPS. initiated for each reactor trip trip functions listed in function listed in Table 2.5-1.

Table 2.5-1.

v. The MPS automatically initiates v. A test will be performed of the v. An ESF actuation signal is an ESF actuation signal for ESF MPS. automatically initiated for each ESF functions listed in Table 2.5-2. function listed in Table 2.5-2.

vi. The MPS automatically vi. A test will be performed of the vi. The RTBs open upon an injection of a actuates a reactor trip. MPS. single simulated MPS reactor trip signal.

vii. The MPS manually actuates a vii. A test will be performed of the vii. The RTBs open when a reactor trip is reactor trip. MPS. manually initiated from the main control room.

Tier 1 2.5-14 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System Table 2.5-7: Module Protection System and Safety Display and Indication System Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria viii. The reactor trip logic fails to a viii. A test will be performed of the viii. Loss of electrical power in a safe state such that loss of MPS. separation group results in a trip electrical power to a MPS state for that separation group.

separation group results in a trip state for that separation group.

ix. The ESFs logic fails to a safe ix. A test will be performed of the ix. Loss of electrical power in a state such that loss of electrical MPS. separation group results in the safe power to a MPS separation state listed in Table 2.1-3.

group results in a safe state listed in Table 2.1-3.

x. The MPS interlocks listed in x. A test will be performed of the x. The MPS interlocks listed in Table Table 2.5-4 automatically MPS. 2.5-4 automatically establish an establish an operating bypass operating bypass for the specified for the specified reactor trip or reactor trip or ESF actuations when ESF actuations when the the interlock condition is met. The interlock condition is met, and operating bypass is automatically the operating bypass is removed when the interlock automatically removed when condition is no longer satisfied.

the interlock condition is no longer satisfied.

xi. The MPS permissives listed in xi. A test will be performed of the xi. The MPS permissives listed in Table 2.5-4 allow the manual MPS. Table 2.5-4 allow the manual bypass bypass of the specified reactor of the specified reactor trip or ESF trip or ESF actuations when the actuations when the permissive permissive condition is met, condition is met. The operating and the operating bypass is bypass is automatically removed automatically removed when when the permissive condition is no the permissive condition is no longer satisfied.

longer satisfied.

xii. The O-1 Override listed in xii. A test will be performed of the xii. The O-1 Override listed in Table 2.5-4 Table 2.5-4 is established when MPS. is established when the manual the manual override switch is override switch is active and the RT-1 active and the RT-1 interlock is interlock is established. The Override established. The Override switch must be manually taken out switch must be manually taken of Override when the O-1 Override is out of Override when the O-1 no longer needed.

Override is no longer needed.

xiii. The MPS is capable of xiii. A test will be performed of the xiii. The MPS performs its safety-related performing its safety-related MPS. functions if any one of its separation functions when any one of its groups is out of service.

separation groupschannels is out of service.

xiv. The RTBs are installed and xiv. An inspection will be performed xiv. The RTBs have the proper arranged as shown in of the as-built RTBs, including the connections for the shunt and Figure 2.5-2 in order to connections for the shunt and undervoltage trip mechanisms and successfully accomplish the undervoltage trip mechanism auxiliary contacts, and are arranged reactor trip function. and auxiliary contacts. as shown in Figure 2.5-2 to successfully accomplish the reactor trip function.

Tier 1 2.5-15 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System Table 2.5-7: Module Protection System and Safety Display and Indication System Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

5. Electrical isolation exists between i. A type test, analysis, or a i. The Class 1E circuit does not degrade the EDSS-MS subsystem non-Class combination of type test and below defined acceptable operating 1E circuits and connected MPS Class analysis will be performed of the levels when the non-Class 1E side of 1E circuits to prevent the Class 1E isolation devices. the isolation device is subjected to propagation of credible electrical the maximum credible voltage, faults. current transients, shorts, grounds, or open circuits.

ii. An inspection will be performed ii. Class 1E electrical isolation devices of the MPS Class 1E as-built are installed between the EDSS-MS circuits. Subsystem non-Class 1E circuits and connected MPS Class 1E circuits.

6. Not used. Not used. Not used.
7. Communications independence A test will be performed of the Class Communications independence between exists between the Class 1E MPS and 1E MPS. the Class 1E MPS and non-Class 1E digital non-Class 1E digital systems. systems is provided.
8. Not used. Not used. Not used.
9. Not used. Not used. Not used.
10. Not used. Not used. Not used.
11. The MPS automatically actuates the A test will be performed of the MPS. The ESF equipment automatically engineered safety featureESF actuates to perform its safety-related equipment to perform its function listed in Table 2.5-2 upon an safety-related function listed in injection of a single simulated MPS signal.

Table 2.5-2.

12. Not used. Not used. Not used.
13. The MPS manually actuates the ESF A test will be performed of the MPS. The MPS actuates the ESF equipment to equipment to perform its safety- perform its safety-related function listed related function listed in Table 2.5-2. in Table 2.5-2 when manually initiated.
14. Not used. Not used. Not used.
15. Not used. Not used. Not used.
16. An MPS signal once initiated A test will be performed of the MPS i. Upon initiation of a real or simulated (automatically or manually), results reactor trip and engineered safety MPS reactor trip signal listed in in an intended sequence of features signals. Table 2.5-1, the RTBs open, and the protective actions that continue RTBs do not automatically close when until completion, and requires the MPS reactor trip signal clears.

deliberate operator action in order ii. Upon initiation of a real or simulated to return the safety systems to MPS engineered safety feature normal. actuation signal listed in Table 2.5-2, the ESF equipment actuates to perform its safety-related function and continues to maintain its safety-related position and perform its safety-related function when the MPS engineered safety feature actuation signal clears.

17. The MPS response times from A test will be performed of the MPS. The MPS reactor trip functions listed in sensor output through equipment Table 2.5-1 and ESFs functions listed in actuation for the reactor trip Table 2.5-2 have response times that are functions and ESF functions are less less than or equal to the design basis than or equal to the value required safety analysis response time to satisfy the design basis safety assumptions.

analysis response time assumptions.

Tier 1 2.5-17 Draft Revision 4

NuScale Tier 1 Module Protection System and Safety Display and Indication System Table 2.5-7: Module Protection System and Safety Display and Indication System Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

18. Not used. Not used. Not used.
19. Not used. Not used. Not used.
20. Not used. Not used. Not used.
21. Not used. Not used. Not used.
22. MPS operational bypasses are A test will be performed of the MPS. Each operational MPS manual or indicated in the MCR. automatic bypass is indicated in the MCR.
23. MPS maintenance bypasses are A test will be performed of the MPS. Each maintenance bypass is indicated in indicated in the MCR. the MCR.
24. The MPS self-test features detect A test will be performed of the MPS. A report exists and concludes that:

faults in the system and provide an

  • Self-testing features verify that faults alarm in the main control roomMCR. requiring detection are detected.
  • Self-testing features verify that upon detection, the system responds according to the type of fault.
  • Self-testing features verify that faults are detected and responded within a sufficient timeframe to ensure safety function is not lost.
  • The presence and type of fault is indicated by the MPS alarms and displays.
25. The PAM Type B and Type C displays An inspection will be performed for The PAM Type B and Type C displays are indicated on the SDIS displays in the ability to retrieve the as-built PAM listed in Table 2.5-5 are retrieved and the MCR. Type B and Type C displays on the displayed on the SDIS displays in the SDIS displays in the MCR. MCR.
26. The controls located on the A test will be performed of the The IHAs controls provided on the operator workstations in the MCR controls on the operator workstations operator workstations in the MCR operate to perform IHAs. in the MCR. perform the functions listed in Table 2.5-6.
27. Not used. Not used. Not used.
28. Not used. Not used. Not used.

Tier 1 2.5-18 Draft Revision 4

NuScale Tier 1 Equipment Qualification 2.8 Equipment Qualification 2.8.1 Design Description

System Description

The scope of this section is equipment qualification (EQ) of equipment specific to each NuScale Power Module. Equipment qualification applies to safety-related electrical and mechanical equipment and safety-related digital instrumentation and controls equipment.

RAI 14.03.03-6, RAI 14.03.03-7 Additionally, this section applies to a limited population of module-specific, nonsafety-related equipment that has augmented Seismic Category I or environmental qualification requirements. The nonsafety-related equipment in this section has one of the following design features:

RAI 14.03.03-6, RAI 14.03.03-7

  • Nonsafety-related mechanical and electrical equipment located within the boundaries of the NuScale Power Module that has an augmented Seismic Category I or environmental qualification design requirement.

RAI 14.03.03-6, RAI 14.03.03-7

  • Nonsafety-related mechanical and electrical equipment that performs a credited function in Chapter 15 analyses (secondary main steam isolation valves (MSIV),

feedwater regulating valves (FWRV) and secondary feedwater check valves).

Design Commitments RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7

  • The module-specific Seismic Category I equipment listed in Table 2.8-1, including its associated supports and anchorages, withstands design basis seismic loads without loss of its function(s) during and after a safe shutdown earthquake (SSE).

RAI 08.01-1S1, RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7

  • The module-specific electrical equipment located in a harsh environment listed in Table 2.8-1, including associated connection assemblies, withstand the design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and post-accident conditions, and performs its function for the period of time required to complete the function.

RAI 14.03-3S1

  • The non-metallic parts, materials, and lubricants used in module-specific mechanical equipment listed in Table 2.8-1 perform their function up to the end of their qualified life in the design basis harsh environmental conditions (both internal service conditions and external environmental conditions) experienced during normal operations, AOOs, DBAs, and post-accident conditions.

RAI 14.03-3

  • The Class 1E computer-based instrumentation and control systems listed in Table 2.8-1 located in a mild environment withstand design basis mild environmental conditions without loss of safety-related functions.

RAI 14.03-3S1 Tier 1 2.8-1 Draft Revision 4

NuScale Tier 1 Equipment Qualification

  • The Class 1E digital equipment listed in Table 2.8-1 performs its safety-related function when subjected to the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA.

RAI 14.03-3S1

  • The valves listed in Table 2.8-1 are functionally designed and qualified to perform their safety-related function under the full range of fluid flow, differential pressure, electrical, temperature, and fluid conditions up to and including DBA conditions.

RAI 14.03-3

  • The safety-related relief valves listed in Table 2.8-1 provide overpressure protection.

RAI 14.03-3S1

  • The DHRS condensers listed in Table 2.8-1 have the capacity to transfer their design heat load.

RAI 08.01-1S1, RAI 14.03-3

  • The containment system (CNTS) containment electrical penetration assemblies listed in Table 2.8-1, including associated connection assemblies, withstand the design basis harsh environmental conditions experienced during normal operations, AOOs, DBAs, and post-accident conditions, and performs its function for the period of time required to complete the function.

2.8.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.8-2 contains the inspections, tests, and analyses for equipment qualification - module-specific equipment.

Tier 1 2.8-2 Draft Revision 4

Tier 1 NuScale Tier 1 RAI 06.02.04-2, RAI 08.01-1S1, RAI 14.03-3S1, RAI 14.03.03-2, RAI 14.03.03-6, RAI 14.03.03-7 Table 2.8-1: Module Specific Mechanical and Electrical/I&C Equipment Description Location EQ Qualification Seismic Class 1E EQ Environment Program Category I Category(1)

Containment System CNTS I&C Division I Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNTS I&C Division II Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNTS PZR Heater Power #1 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNTS PZR Heater Power #2 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNTS I&C Channel A Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical CNTS I&C Channel B Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical 2.8-3 CNTS I&C Channel C Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical CNTS I&C Channel D Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical CNTS CRD Power Electrical Penetration Assembly RXB - Top of Module Harsh Electrical Yes No A (EPA) RXB - Inside Containment Mechanical CNTS RPI Group #1 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNTS RPI Group #2 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical MS #1 CIV (MSIV #1) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical MS #2 CIV (MSIV #2) RXB - Top of Module Harsh Electrical Yes Yes AB Equipment Qualification Mechanical MS line #1 Bypass Valve (MSIV Bypass #1) RXB - Top of Module Harsh Electrical Yes Yes AB Draft Revision 4 Mechanical MS line #2 Bypass Valve (MSIV Bypass #2) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical FW #1 CIV (FWIV #1) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical

Table 2.8-1: Module Specific Mechanical and Electrical/I&C Equipment (Continued)

Tier 1 NuScale Tier 1 Description Location EQ Qualification Seismic Class 1E EQ Environment Program Category I Category(1)

SG2 Steam Supply CIV/ MS Bypass Isolation Valve RXB - Top of Module Harsh Electrical Yes No A Close Position Sensors (2 Total)

SG2 Steam Supply CIV/ MS Bypass Isolation Valve RXB - Top of Module Harsh Electrical Yes No A Open Position Sensors (2 Total)

Steam Generator System SG Tubes and Tube Supports RXB - Inside Containment N/A N/A Yes N/A N/A Steam Plenums (4 Total) RXB - Inside Containment N/A N/A Yes N/A N/A Feedwater Plenums (4 Total) RXB - Inside Containment N/A N/A Yes N/A N/A Flow Restrictors RXB - Inside Containment N/A N/A Yes N/A N/A Thermal Relief Valves (2 Total) RXB - Inside Containment Harsh Mechanical Yes N/A B Control Rod Drive System Rod Position Indication (RPI) Coils (2432 Total) RXB - Inside Containment Harsh Electrical Yes No B Control Rod Drive Shafts RXB - Inside Containment N/A N/A Yes N/A N/A Control Rod Drive Latch Mechanism RXB - Inside Containment N/A N/A Yes N/A N/A 2.8-7 CRDM Pressure Boundary (Latch Housing, Rod RXB - Inside Containment N/A N/A Yes N/A N/A Travel Housing, Rod Travel Housing Plug)

Control Rod Assembly All components RXB - Inside Containment N/A N/A Yes N/A N/A Neutron Source Assembly Primary and secondary neutron source rodlets RXB - Inside Containment N/A N/A Yes N/A N/A Spider body, hub or coupling housing RXB - Inside Containment N/A N/A Yes N/A N/A Reactor Coolant System Reactor Vessel Internals RXB - Inside Containment N/A N/A Yes N/A N/A Reactor Safety Valve Position Indicators (4 Total) RXB - Inside Containment Harsh Electrical Yes No B Reactor Safety Valves (2 Total) RXB - Inside Containment Harsh Electrical Yes N/A A Mechanical Equipment Qualification Narrow Range Pressurizer Pressure Elements (4 RXB - Inside Containment Harsh Electrical Yes Yes A Total)

Wide Range RCS Pressure Elements (4 Total) RXB - Inside Containment Harsh Electrical Yes Yes A Draft Revision 4 PZR/RPV Level Elements (4 Total) RXB - Top of Module Harsh Electrical Yes Yes A RXB - Inside Containment Narrow Range RCS Hot Leg Temperature RXB - Inside Containment Harsh Electrical Yes Yes A Elements (12 Total)

NuScale Tier 1 Equipment Qualification RAI 08.01-1S1, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.03-6, RAI 14.03.03-7 Table 2.8-2: Equipment Qualification Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The module-specific Seismic Category i. A type test, analysis, or a i. A sSeismic qQualification I equipment listed in Table 2.8-1, combination of type test and Reportrecord form exists and including its associated supports and analysis will be performed of the concludes that the module-anchorages, withstands design basis module-specific Seismic Category I specific Seismic Category I seismic loads without loss of its equipment listed in Table 2.8-1, equipment listed in Table 2.8-1, function(s) during and after an SSE. including its associated supports including its associated supports and anchorages. and anchorages, will withstand the ii. An inspection will be performed of design basis seismic loads and the module-specific Seismic perform its function(s) during and Category I as-built equipment after an SSE.

listed in Table 2.8-1, including its ii. The module-specific Seismic associated supports and Category I equipment listed in anchorages. Table 2.8-1, including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the equipments sSeismic qQualification Reportrecord form.

2. The module-specific electrical i. A type test or a combination of i. An EQ record form exists and equipment located in a harsh type test and analysis will be concludes that the module-environment listed in Table 2.8-1, performed of the module-specific specific electrical equipment listed including associated connection electrical equipment listed in in Table 2.8-1, including associated assemblies, withstand the design basis Table 2.8-1, including associated connection assemblies, perform harsh environmental conditions connection assemblies. their function under the experienced during normal ii. An inspection will be performed of environmental conditions operations, AOOs, DBAs, and post- the module-specific as-built specified in the EQ record form for accident conditions and performs its electrical equipment listed in the period of time required to function for the period of time Table 2.8-1, including associated complete the function.

required to complete the function. connection assemblies. ii. The module-specific electrical equipment listed in Table 2.8-1, including associated connection assemblies, are installed in their design location in a configuration bounded by the EQ record form.

3. The non-metallic parts, materials, and A type test or a combination of type A qualification record form exists and lubricants used in module-specific test and analysis will be performed of concludes that the non-metallic parts, mechanical equipment listed in the non-metallic parts, materials, and materials, and lubricants used in Table 2.8-1 perform their function up lubricants used in module-specific module-specific mechanical to the end of their qualified life in the mechanical equipment listed in equipment listed in Table 2.8-1 design basis harsh environmental Table 2.8-1.Not used. perform their function up to the end of conditions (both internal service their qualified life under the design conditions and external basis harsh environmental conditions environmental conditions) (both internal service conditions and experienced during normal external environmental conditions) operations, AOOs, DBAs, and specified in the qualification record post-accident conditions.Not used. form.Not used.

Tier 1 2.8-11 Draft Revision 4

NuScale Tier 1 Equipment Qualification Table 2.8-2: Equipment Qualification Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

4. The Class 1E computer-based i. A type test or a combination of i. An EQ record form exists and instrumentation and control systems type test and analysis will be concludes that the Class 1E listed in Table 2.8-1 located in a mild performed of the Class 1E computer-based instrumentation environment withstand design basis computer-based instrumentation and control systems listed in mild environmental conditions and control systems listed in Table 2.8-1 located in a mild without loss of safety-related Table 2.8-1 located in a mild environment perform their functions. environment. function under the environmental ii. An inspection will be performed of conditions specified in the EQ the Class 1E as-built computer- record form.

based instrumentation and control ii. The Class 1E computer-based systems listed in Table 2.8-1 instrumentation and control located in a mild environment. systems listed in Table 2.8-1 located in a mild environment are installed in their design location in a configuration bounded by the EQ record form.

5. The Class 1E digital equipment listed A type test, analysis, or a combination An EQ record form exists and in Table 2.8-1 performs its safety- of type test and analysis will be concludes that the Class 1E digital related function when subjected to performed of the Class 1E digital equipment listed in Table 2.8-1 the design basis electromagnetic equipment listed in Table 2.8-1.Not withstands the design basis interference, radio frequency used. electromagnetic interference, radio interference, and electrical surges that frequency interference, and electrical would exist before, during, and surges that would exist before, during, following a DBA.Not used. and following a DBA without loss of safety-related function.Not used.
6. The valves listed in Table 2.8-1 are A type test or a combination of type A Qualification Report exists and functionally designed and qualified to test and analysis will be performed of concludes that the valves listed in perform their safety-related function the valves listed in Table 2.8-1.Not Table 2.8-1 are capable of performing under the full range of fluid flow, used. their safety-related function under the differential pressure, electrical, full range of fluid flow, differential temperature, and fluid conditions up pressure, electrical, temperature, and to and including DBA conditions.Not fluid conditions up to and including used. DBA conditions.Not used.
7. The safety-related relief valves listed in i. A vendor test will be performed of i. An American Society of Table 2.8-1 provide overpressure each safety-related relief valve Mechanical Engineers Code protection. listed in Table 2.8-1. Section III Data Report exists and ii. An inspection will be performed of concludes that the relief valves each safety-related as-built relief listed in Table 2.8-1 meet the valve listed in Table 2.8-1. valves required set pressure, capacity, and overpressure design requirements.

ii. Each relief valve listed in Table 2.8-1 is provided with an American Society of Mechanical Engineers Code Certification Mark that identifies the set pressure, capacity, and overpressure.

Tier 1 2.8-12 Draft Revision 4

NuScale Tier 1 Equipment Qualification Table 2.8-2: Equipment Qualification Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

8. The DHRS condensers listed in A type test or a combination of type A report exists and concludes that the Table 2.8-1 have the capacity to test and analysis will be performed of DHRS condensers listed in Table 2.8-1 transfer their design heat load.Not the DHRS condensers listed in have a heat removal capacity sufficient used. Table 2.8-1.Not used. to transfer their design heat load.Not used.
9. The CNTS containment electrical i. A type test or a combination of i. An EQ record form exists and penetration assemblies listed in type test and analysis will be concludes that the CNTS electrical Table 2.8-1, including associated performed of the CNTS penetration assemblies listed in connection assemblies, withstand the containment electrical penetration Table 2.8-1, including associated design basis harsh environmental assemblies listed in Table 2.8-1 connection assemblies, performs conditions experienced during normal including associated connection their function under the operations, AOOs, DBAs, and assemblies. environmental conditions postaccident conditions and performs ii. An inspection will be performed of specified in the EQ record form for its function for the period of time the containment CNTS electrical the period of time required to required to complete the function. penetration assembles listed in complete the function.

Table 2.8-1, including associated ii. The CNTS electrical penetration connection assemblies. assemblies listed in Table 2.8-1, including associated connection assemblies, are installed in their design location in a configuration bounded by the EQ record form.

Tier 1 2.8-13 Draft Revision 4

NuScale Tier 1 Control Room Habitability 3.1 Control Room Habitability 3.1.1 Design Description

System Description

The scope of this section is the control room habitability system (CRHS). The CRHS provides clean breathing air to the control room envelope and maintains a positive control room pressure during high radiation or loss of offsite power conditions for habitability and control of radioactivity. The CRHS is a nonsafety-related system which supports up to 12 NuScale Power Modules (NPMs). The Control Building houses all CRHS equipment.

The CRHS performs the following nonsafety-related system function that is verified by Inspections, Tests, Analyses, and Acceptance Criteria:

  • The CRHS supports the Control Building by providing clean breathing air to the main control room (MCR) and maintains a positive control room pressure during high radiation or loss of normal AC power conditions.

Design Commitments RAI 14.03-3S1

  • The air exfiltration out of the control room envelope (CRE) does not exceedis less than or equal to the assumptions used to size the CRHS inventory and the supply flow rate.

RAI 14.03-3

  • The CRHS valves listed in Table 3.1-1 change position under design basis temperature, differential pressure, and flow conditions.

RAI 14.03-3

  • The CRHS solenoid-operated valves listed in Table 3.1-1 perform their function to fail open on loss of motive power under design basis temperature, differential pressure, and flow conditions.
  • The CRE heat sink passively maintains the temperature of the CRE within an acceptable range for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a design basis accident (DBA).
  • The CRHS maintains a positive pressure in the MCR relative to the adjacent areas.

3.1.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.1-2 contains the inspections, tests, and analyses for the CRHS.

Table 3.1-1: Control Room Habitability System Mechanical Equipment Equipment Name Failure Position Air supply isolation solenoid valves (2 Total) Open CRE pressure relief isolation valves (2 Total) Open Tier 1 3.1-1 Draft Revision 4

NuScale Tier 1 Control Room Habitability RAI 14.03-3, RAI 14.03-3S1 Table 3.1-2: Control Room Habitability System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The air exfiltration out of the CRE does A test will be performed of the CRE. The air exfiltration measured by tracer not exceedis less than or equal to the gas testing is less than or equal to the assumptions used to size the CRHS CRE air infiltration rate assumed in the inventory and the supply flow rate. dose analysis.

2 The CRHS valves listed in Table 3.1-1 A test will be performed of the CRHS Each CRHS valve listed in Table 3.1-1 change position under design basis valves listed in Table 3.1-1 under strokes fully open and fully closed by temperature, differential pressure, and preoperational temperature, remote operation under flow conditions. differential pressure, and flow preoperational temperature, conditions. differential pressure, and flow conditions.

3 The CRHS solenoid-operated valves A test will be performed of the CRHS Each CRHS solenoid-operated valve listed in Table 3.1-1 perform their solenoid-operated valves listed in listed in Table 3.1-1 performs its function to fail open on loss of motive Table 3.1-1 under preoperational function to fail open on loss of motive power under design basis temperature, differential pressure and power under preoperational temperature, differential pressure, and flow conditions. temperature, differential pressure, and flow conditions. flow conditions.

4 The CRE heat sink passively maintains An analysis will be performed of the A report exists and concludes that the the temperature of the CRE within an as-built CRE heat sinks. CRE heat sink passively maintains the acceptable range for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> temperature of the CRE within an following a DBA. acceptable range for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a DBA.

5 The CRHS maintains a positive A test will be performed of the CRHS. The CRHS maintains a positive pressure in the MCR relative to pressure of greater than or equal to adjacent areas. 1/8 inches water gauge in the CRE relative to adjacent areas, while operating in DBA alignment.

Tier 1 3.1-2 Draft Revision 4

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 RAI 14.03-3, RAI 14.03-3S1 Table 3.9-2: Radiation Monitoring - NuScale Power Modules 1-12 Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The CRVS automatically responds to A test will be performed of the CRVS Upon initiation of the real or simulated the CRVS high-radiation signals high-radiation signals listed in CRVS high-radiation signals upstream upstream of the CRVS filter unit listed Table 3.9-1. of the CRVS filter unit listed in in Table 3.9-1 to mitigate a release of Table 3.9-1, the CRVS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

2 The CRVS and the CRHS automatically A test will be performed of the CRVS Upon initiation of the real or simulated respond to the CRVS high-radiation high-radiation signals listed in CRVS high-radiation signals signals downstream of the CRVS filter Table 3.9-1. downstream of the CRVS filter unit unit listed in Table 3.9-1 to mitigate a listed in Table 3.9-1, the CRVS and the release of radioactivity. CRHS automatically align/actuate the identified components to the positions identified in the table.

3 The RBVS automatically responds to A test will be performed of the RBVS Upon initiation of the real or simulated the RBVS high-radiation signals listed high-radiation signals listed in RBVS high-radiation signals listed in in Table 3.9-1 to mitigate a release of Table 3.9-1. Table 3.9-1, the RBVS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

4 The GRWS automatically responds to A test will be performed of the GRWS Upon initiation of the real or simulated the GRWS high-radiation signals listed high-radiation signals listed in GRWS high-radiation signals listed in in Table 3.9-1 to mitigate a release of Table 3.9-1. Table 3.9-1, the GRWS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

5 Not Used. Not Used. Not Used.

6 Not Used. Not Used. Not Used.

7 The LRWS automatically responds to A test will be performed of the LRWS Upon initiation of the real or simulated the LRWS high-radiation signals listed high-radiation signals listed in LRWS high-radiation signals listed in in Table 3.9-1 to mitigate a release of Table 3.9-1. Table 3.9-1, the LRWS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

8 The ABS automatically responds to the A test will be performed of the ABS Upon initiation of the real or simulated ABS high-radiation signals listed in high-radiation signals listed in ABS high-radiation signals listed in Table 3.9-1 to mitigate a release of Table 3.9-1. Table 3.9-1, the ABS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

9 Not Used. Not Used. Not Used.

10 The PSCS automatically responds to A test will be performed of the PSCS Upon initiation of thea real or the PSCS high-radiation signal listed in high-radiation signal listed in simulated PSCS high-radiation signal Table 3.9-1 to mitigate a release of Table 3.9-1. listed in Table 3.9-1, the PSCS radioactivity. automatically aligns/actuates the identified components to the positions identified in the table.

Tier 1 3.9-4 Draft Revision 4

NuScale Tier 1 Equipment Qualification - Shared Equipment 3.14 Equipment Qualification - Shared Equipment 3.14.1 Design Description

System Description

The scope of this section is equipment qualification (EQ) of equipment shared by NuScale Power Modules 1 through 12, and a limited set of one-time module specific analyses.

RAI 09.01.03-1S1, RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7 This section applies to the safety-related reactor pressure vessel (RPV) support stand and Reactor Building (RXB) over-pressurization vents, and a limited population of common, nonsafety-related equipment that has augmented Seismic Category I or environmental qualification requirements. The nonsafety-related equipment in this section provides one of the following nonsafety-related functions:

RAI 14.03.03-6, RAI 14.03.03-7

  • Provides physical support of irradiated fuel (fuel handling machine, spent fuel storage racks, reactor building crane, and module lifting adapter).

RAI 14.03-3

  • Provides containment of the UHS water.
  • Monitors UHS water level.

RAI 14.03.08-1S1 Additionally, this section applies to the nonsafety-related, RW-IIa components and piping used for processing gaseous radioactive waste.

Design Commitments RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7

  • The common, Seismic Category I equipment listed in Table 3.14-1, including its associated supports and anchorages, withstands design basis seismic loads without loss of its function(s) during and after a safe shutdown earthquake.

RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7

  • The common electrical equipment listed in Table 3.14-1 located in a harsh environment, including its connection assemblies, withstands the design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences, design basis accidents, and post-accident conditions, and performs its function for the period of time required to complete the function.

RAI 14.03-3, RAI 14.03.08-1S1

  • The RW-IIa components and piping used for processing gaseous radioactive waste listed in Table 3.14-1 are constructed to the standards of RW-IIa.

RAI 14.03-3, RAI 14.03-3S1

  • Each containment system (CNTS) containment electrical penetration assembly listed in Table 2.1-3 is rated either (i) to withstand fault and overload currents for the time required to clear the fault from its power source, or (ii) to with withstand the maximum fault and overload current for its circuits without a circuit interrupting device.

RAI 14.03-3, RAI 14.03-3S1 Tier 1 3.14-1 Draft Revision 4

NuScale Tier 1 Equipment Qualification - Shared Equipment

  • The non-metallic parts, materials, and lubricants used in module-specific mechanical equipment listed in Table 2.8-1 perform their function up to the end of their qualified life in the design basis harsh environmental conditions (both internal service conditions and external environmental conditions) experienced during normal operations, anticipated operational occurrences (AOOs), design basis accidents (DBAs),

and post-accident conditions.

RAI 14.03-3, RAI 14.03-3S1

  • The Class 1E digital equipment listed in Table 2.8-1 performs its safety-related function when subjected to the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA.

RAI 14.03-3, RAI 14.03-3S1

  • The valves listed in Table 2.8-1 are functionally designed and qualified to perform their safety-related function under the full range of fluid flow, differential pressure, electrical, temperature, and fluid conditions up to and including DBA conditions.

RAI 14.03-3, RAI 14.03-3S1

  • The decay heat removal system (DHRS) condensers listed in Table 2.8-1 have the capacity to transfer their design heat load.

3.14.2 Inspections, Tests, Analyses, and Acceptance Criteria RAI 14.03-3S1 Table 3.14-2 contains the inspections, tests, and analyses for EQ -- shared equipment.

Tier 1 3.14-2 Draft Revision 4

NuScale Tier 1 Equipment Qualification - Shared Equipment RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.08-1S1 Table 3.14-2: Equipment Qualification - Shared Equipment ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The common Seismic Category I i. A type test, analysis, or a i. A sSeismic qQualification equipment listed in Table 3.14-1, combination of type test and Reportrecord form exists and including its associated supports and analysis will be performed of the concludes that the common anchorages, withstands design basis common Seismic Category I Seismic Category I equipment seismic loads without loss of its equipment listed in Table 3.14-1, listed in Table 3.14-1, including its function(s) during and after a safe including its associated supports associated supports and shutdown earthquake. and anchorages. anchorages, will withstand the design basis seismic loads and perform its function during and after a safe shutdown earthquake.

ii. An inspection will be performed of ii. The common Seismic Category I the common Seismic Category I as- equipment listed in Table 3.14-1, built equipment listed in including its associated supports Table 3.14-1, including its and anchorages, is installed in its associated supports and design location in a Seismic anchorages. Category I structure in a configuration bounded by the equipments sSeismic qQualification Reportrecord form.

2. The common electrical equipment i. A type test or a combination of i. An equipment qualification record listed in Table 3.14-1 located in a harsh type test and analysis will be form exists and concludes that the environment, including its connection performed of the common common electrical equipment assemblies, withstands the design electrical equipment listed in listed in Table 3.14-1, including its basis harsh environmental conditions Table 3.14-1, including its connection assemblies, performs experienced during normal connection assemblies. its function under the operations, anticipated operational environmental conditions occurrences, DBA, and post-accident specified in the equipment conditions and performs its function qualification record form for the for the period of time required to period of time required to complete the function. complete the function.

ii. An inspection will be performed of ii. The common electrical equipment the common as-built electrical listed in Table 3.14-1, including its equipment listed in Table 3.14-1, connection assemblies, is installed including its connection in its design location in a assemblies. configuration bounded by the EQ record form.

3. The RW-IIa components and piping i. An inspection and reconciliation i. A report exists and concludes that used for processing gaseous analysis will be performed of the the as-built RW-IIa components radioactive waste listed in Table 3.14-1 as-built RW-IIa components and and piping used for processing are constructed to the standards of piping used for processing gaseous radioactive waste listed in RW-IIa. gaseous radioactive waste listed in Table 3.14-1 meet the RW-IIa Table 3.14-1. design criteria.

Tier 1 3.14-5 Draft Revision 4

NuScale Tier 1 Equipment Qualification - Shared Equipment Table 3.14-2: Equipment Qualification - Shared Equipment ITAAC (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

4. Each CNTS containment electrical An analysis will be performed of each For each CNTS containment electrical penetration assembly listed in CNTS as-built containment electrical penetration assembly listed in Table 2.1-3 is rated either (i) to penetration assembly listed in Table 2.1-3, either (i) a circuit withstand fault and overload currents Table 2.1-3. interrupting device coordination for the time required to clear the fault analysis exists and concludes that the from its power source, or (ii) to current carrying capability for the withstand the maximum fault and CNTS containment electrical overload current for its circuits without penetration assembly is greater than a circuit interrupting device. the analyzed fault and overload currents for the time required to clear the fault from its power source, or (ii) an analysis of the CNTS containment electrical penetration maximum fault and overload current exists and concludes the fault and overload current is less than the current carrying capability of the CNTS containment electrical penetration.
5. The non-metallic parts, materials, and A type test or a combination of type A qualification record form exists and lubricants used in module-specific test and analysis will be performed of concludes that the non-metallic parts, mechanical equipment listed in the non-metallic parts, materials, and materials, and lubricants used in Table 2.8-1 perform their function up lubricants used in module-specific module-specific mechanical to the end of their qualified life in the mechanical equipment listed in equipment listed in Table 2.8-1 design basis harsh environmental Table 2.8-1. perform their function up to the end conditions (both internal service of their qualified life under the design conditions and external basis harsh environmental conditions environmental conditions) (both internal service conditions and experienced during normal external environmental conditions) operations, AOOs, DBAs, and specified in the qualification record post-accident conditions. form.
6. The Class 1E digital equipment listed A type test, analysis, or a combination An EQ record form exists and in Table 2.8-1 performs its of type test and analysis will be concludes that the Class 1E digital safety-related function when performed of the Class 1E digital equipment listed in Table 2.8-1 subjected to the design basis equipment listed in Table 2.8-1. withstands the design basis electromagnetic interference, radio electromagnetic interference, radio frequency interference, and electrical frequency interference, and electrical surges that would exist before, during, surges that would exist before, during, and following a DBA. and following a DBA without loss of safety-related function.
7. The valves listed in Table 2.8-1 are A type test or a combination of type A Qualification Report exists and functionally designed and qualified to test and analysis will be performed of concludes that the valves listed in perform their safety-related function the valves listed in Table 2.8-1. Table 2.8-1 are capable of performing under the full range of fluid flow, their safety-related function under the differential pressure, electrical, full range of fluid flow, differential temperature, and fluid conditions up pressure, electrical, temperature, and to and including DBA conditions. fluid conditions up to and including DBA conditions.
8. The DHRS condensers listed in A type test or a combination of type A report exists and concludes that the Table 2.8-1 have the capacity to test and analysis will be performed of DHRS condensers listed in Table 2.8-1 transfer their design heat load. the DHRS condensers listed in have a heat removal capacity Table 2.8-1. sufficient to transfer their design heat load.

Tier 1 3.14-6 Draft Revision 4

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 6 RAI 14.03-3S1 Table 3.17-1: Radiation Monitoring - Automatic Actions for NuScale Power Modules 1 - 6 Variable Monitored Actuated Component(s) Component Action(s)

CFDS containment drain separator 1. CFDS containment drain separator gaseous discharge 1. Close gaseous discharge to Reactor isolation valve Building heating ventilation and air conditioning system BPDS 0A condensate polishing 1. North chemical waste water sump pump A 1. Stop system regeneration skid waste 2. North chemical waste water sump pump B 2. Stop effluent 3. North chemical water sump to BPDS collection tank flow 3. Close control valve

4. North chemical water sump to liquid radioactive waste 4. Close system (LRWS) isolation valve BPDS north turbine building floor 1. North waste water sump pump A 1. Stop drains 2. North waste water sump pump B 2. Stop
3. North waste water sump to BPDS collection tank flow 3. Close control valve
4. North waste water sump to LRWS isolation valve 4. Close BPDS auxiliary blowdown cooler 1. North waste water sump pump A 1. Stop condensate 2. North waste water sump pump B 2. Stop
3. North waste water sump to BPDS collection tank flow 3. Close control valve
4. North waste water sump to LRWS isolation valve 4. Close Tier 1 3.17-2 Draft Revision 4

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 7 - 12 RAI 14.03-3S1 Table 3.18-1: Radiation Monitoring - Automatic Actions For NuScale Power Modules 7 - 12 Variable Monitored Actuated Component(s) Component Action(s)

CFDS containment drain separator 1. CFDS containment drain separator gaseous discharge 1. Close gaseous discharge to Reactor isolation valve Building heating ventilation and air conditioning system BPDS 0B condensate polishing 1. South chemical waste water sump pump A 1. Stop system regeneration skid waste 2. South chemical waste water sump pump B 2. Stop effluent 3. South chemical water sump to BPDS collection tank flow 3. Close control valve

4. South chemical water sump to liquid radioactive waste 4. Close system (LWRS) isolation valve BPDS south turbine building floor 1. South waste water sump pump A 1. Stop drains 2. South waste water sump pump B 2. Stop
3. South waste water sump to BPDS collection tank flow 3. Close control valve
4. South waste water sump to liquid radioactive waste system 4. Close isolation valve Tier 1 3.18-2 Draft Revision 4

Tier 2 NuScale Final Safety Analysis Report RAI 03.02.02-7, RAI 06.02.06-22, RAI 06.02.06-23, RAI 08.01-1S1, RAI 08.01-2, RAI 10.02-3, RAI 10.02.03-1, RAI 10.02.03-2, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.03-3S1, RAI 14.03.03-4S1, RAI 14.03.03-5S3,RAI 14.03.03-6, RAI 14.03.03-6S1, RAI 14.03.03-7, RAI 14.03.03-7S1, RAI 14.03.03-8, RAI 14.03.03-9, RAI 14.03.03-9S1, RAI 14.03.07-1 Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.01 NPM As required by ASME Code Section III NCA-1210, each ASME Code X Class 1, 2 and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550. NCA-3551.1 requires that the drawings used for construction be in agreement with the Design Report before it is certified and be identified and described in the Design Report. It is the responsibility of the N Certificate Holder to furnish a Design Report for each component and support, except as provided in NCA-3551.2 and NCA-3551.3. NCA-3551.1 also requires that the Design Report be certified by a registered professional engineer when it is for Class 1 components and supports, Class CS core 14.3-14 support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components Certified Design Material and Inspections, Tests, Analyses, and designed to Service Loadings greater than Design Loadings. A Class 2 Design Report shall be prepared for Class 1 piping NPS 1 or smaller that is designed in accordance with the rules of Subsection NC. NCA-3554 requires that any modification of any document used for construction, from the corresponding document used for design analysis, shall be reconciled with the Design Report.

An ITAAC inspection is performed of the NuScale Power Module ASME Code Class 1, 2 and 3 as-built piping system Design Report to verify that the requirements of ASME Code Section III are met.

Acceptance Criteria Draft Revision 4

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.21 NPM The CNTS safety-related check valves are tested to demonstrate the X capability to perform their function to transfer open and transfer closed (under forward and reverse flow conditions, respectively) under preoperational temperature, differential pressure, and flow conditions. Check valves are tested in accordance with the requirements of the ASME OM Code, ISTC-5220, Check Valves.

In accordance with Table 14.2-43, a preoperational test demonstrates that the CNTS check valves listed in Tier 1 Table 2.1-2 (Table 14.3-3b) strokes fully open and closed under forward and reverse flow conditions, respectively.

Preoperational test conditions are established that approximate design basis temperature, differential pressure and flow conditions to the extent practicable, consistent with preoperational test limitations.

14.3-24 02.01.22 NPM Not used.The CNTS electrical penetrations listed in Tier 1 X Certified Design Material and Inspections, Tests, Analyses, and Table 2.1-3 may be one of two types, one with or without a circuit interrupting device. An ITAAC confirms that each type of penetration is evaluated to confirm it can withstand its maximum fault current.

A circuit interrupting device coordination analysis confirms and concludes in a report that the as-built containment electrical penetration assembly listed in Tier 1 Table 2.1-3 that has a circuit interrupting device can withstand fault currents for the time required to clear the fault from its power source.

8.3.1.2.5 Containment Electrical Penetration Assemblies discusses electrical penetration assemblies that are not equipped with Acceptance Criteria protection devices whose maximum fault current in these circuits would not damage the electrical penetration assembly if that fault current was available indefinitely. An analysis of a CNTS as-built Draft Revision 4 containment penetration without a circuit interrupting device confirms and concludes in a report that the maximum fault current is less than the current carrying capability of the CNTS containment electrical penetration.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.01 EQ Section 3.10, Seismic and Dynamic Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the Seismic Category I equipment, including its associated supports and anchorages, is qualified by type test, analysis, or a combination of type test and analysis to perform its function under the design basis seismic loads during and after an SSE. The qualification method employed for the Seismic Category I equipment is the same as the qualification method described for that type of equipment in Section 3.10. This method conforms to IEEE-344-2004 and ASME QME-1-2007 (or later editions), as accepted by the NRC staff in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.10.

The scope of equipment for this design commitment is module-specific, safety-related equipment, and module-specific, nonsafety-related equipment that has one of the following design 14.3-52 features:

Certified Design Material and Inspections, Tests, Analyses, and

  • Nonsafety-related mechanical and electrical equipment located within the boundaries of the NuScale Power Module that has an augmented Seismic Category I design requirement.

The ITAAC verifies that: (1) a sSeismic qQualification Reportrecord form exists for each Seismic Category I component type, and (2) the sSeismic qQualification Reportrecord form concludes that the Seismic Category I equipment listed in Tier 1 Table 2.8-1 (Table 14.3-3h), including its associated supports and anchorages, Acceptance Criteria performs its function under the seismic design basis load conditions specified in the sSeismic qQualification Reportrecord form.

Draft Revision 4 After installation in the plant, an ITAAC inspection is performed to verify that the Seismic Category I equipment listed in Tier 1 Table 2.8-1 (Table 14.3-3h), including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the sSeismic qQualification Reportrecord form.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.03 EQ Not used.Section 3.11 presents information to demonstrate that X the non-metallic parts, materials, and lubricants used in mechanical equipment located in a harsh environment are qualified using a type test or a combination of type test and analysis to perform their function up to the end of their qualified life in design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions. Environmental conditions include both internal service conditions and external environmental conditions for the nonmetallic parts, materials, and lubricant. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

The scope of equipment for this design commitment is module-specific, safety-related mechanical equipment, and 14.3-54 module-specific, nonsafety-related mechanical equipment that Certified Design Material and Inspections, Tests, Analyses, and performs a credited function in Chapter 15 analyses (secondary main steam isolation valves (MSIV), feedwater regulating valves (FWRV) and secondary feedwater check valves).

The ITAAC verifies that: (1) an equipment qualification record form or ASME QME-1 report exists for the non-metallic parts, materials, and lubricants used in mechanical equipment designated for a harsh environment, and (2) the qualification record form concludes that the non-metallic parts, materials, and lubricants used in mechanical equipment listed in Tier 1 Table 2.8-1 perform their intended function up to the end of its qualified life under the design basis environmental conditions (both internal service conditions and external environmental conditions) specified in the Acceptance Criteria qualification record form.

Draft Revision 4

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.05 EQ Not used.Section 3.11, Environmental Qualification of Mechanical X and Electrical Equipment, presents information to demonstrate that the Class 1E digital equipment is qualified using a type test, analysis, or a combination of type test and analysis to perform its safety-related function when subjected to electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA. The qualification method employed for Class 1E digital equipment is the same as the qualification method described for that type of equipment in Section 3.11.

The ITAAC verifies that: (1) an equipment qualification record form exists for the Class 1E digital equipment listed in Tier 1 Table 2.8-1, and (2) the equipment qualification record form concludes that the Class 1E digital equipment withstands the design basis electromagnetic interference, radio frequency interference, and 14.3-56 electrical surges that would exist before, during, and following a Certified Design Material and Inspections, Tests, Analyses, and DBA without loss of safety-related function.

02.08.06 EQ Not used.Section 3.9.6.1, Functional Design and Qualification of X Pumps, Valves, and Dynamic Restraints, and Section 3.10.2, Methods and Procedures for Qualifying Mechanical and Electrical Equipment and Instrumentation, discuss that the functional qualification of safety-related valves is performed in accordance with ASME QME-1-2007 (or later edition), as accepted in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.9.6.1. The qualification method employed for the valves agrees with the qualification method described in Section 3.10.2.

Acceptance Criteria The ITAAC verifies that: (1) a Qualification Report exists for the safety-related valves listed in Tier 1 Table 2.8-1, and (2) the Qualification Report concludes that safety-related valves are capable of performing their safety-related function under the full Draft Revision 4 range of fluid flow, differential pressure, electrical conditions, temperature conditions, and fluid conditions up to and including DBA conditions.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.07 EQ Section 3.9.3.2, Design and Installation of Pressure Relief Devices, X discusses that relief valves provide overpressure protection in accordance with the ASME Code Section III.

The ITAAC verifies that: (1) the test for each relief valve listed in Tier 1 Table 2.8-1 (Table 14.3-3h) meets the set pressure, capacity, and overpressure design requirements; and (2) each relief valve listed in Tier 1 Table 2.8-1 (Table 14.3-3h) is provided with an ASME Code Certification Mark that identifies the valve's set pressure, capacity, and overpressure.

02.08.08 EQ Not used.Section 5.4.3, Decay Heat Removal System, discusses that X the DHRS passive condensers provide the safety-related function of transferring their design heat load from the DHRS during shutdown. After manufacture of the DHRS passive condensers, a type test or a combination of type test and analysis is performed to 14.3-57 validate that the DHRS passive condensers are capable of meeting the specified heat transfer performance requirements. Section 5.4.3 Certified Design Material and Inspections, Tests, Analyses, and discusses the design heat transfer capability of the DHRS passive condensers.

The ITAAC verifies that the safety-related passive condensers listed in Tier 1 Table 2.8-1 have a heat removal capacity sufficient to transfer their design heat load.

Acceptance Criteria Draft Revision 4

Tier 2 NuScale Final Safety Analysis Report RAI 09.01.04-1, RAI 09.05.01-6, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.02-1, RAI 14.03.02-2, RAI 14.03.03-1, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.03-8, RAI 14.03.07-1, RAI 14.03.08-1S1, RAI 14.03.09-1, RAI 14.03.09-2, RAI 14.03.09-3, RAI 14.03.12-2, RAI 14.03.12-3, RAI 18-46S1 Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, X Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, to demonstrate that air exfiltration from the CRE is controlled. RG 1.197 allows two options for CRE testing; either integrated testing (tracer gas testing) or component testing. Section 6.4 Control Room Habitability, describes the testing requirements for the CRE habitability program. Section 6.4 provides the maximum air exfiltration allowed from the CRE.

In accordance with Table 14.2-18, a preoperational test using the tracer gas test method demonstrates that the air exfiltration from the CRE does not exceed the assumed unfiltered leakage rate provided in Table 6.4-1: Control Room Habitability System Design Parameters 14.3-59 for the dose analysis. Tracer gas testing in accordance with ASTM Certified Design Material and Inspections, Tests, Analyses, and E741 will be performed to measure the unfiltered in-leakage into the CRE with the control room habitability system (CRHS) operating.

03.01.02 CRH The CRHS valves are tested by remote operation to demonstrate the X capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18, a preoperational test demonstrates that each CRHS valve listed in Tier 1 Table 3.1-1 (Table 14.3-4a) strokes fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate Acceptance Criteria design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

Draft Revision 4

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.14.01 EQ Section 3.10, Seismic and Dynamic Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the Seismic Category I equipment, including its associated supports and anchorages, is qualified by type test, analysis, or a combination of type test and analysis to perform its function under the design basis seismic loads during and after an SSE. The qualification method employed for the Seismic Category I equipment is the same as the qualification method described for that type of equipment in Section 3.10. This method conforms to IEEE-344-2004 and ASME QME-1-2007 (or later editions), as accepted by the NRC staff in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.10.

The scope of equipment for this design commitment is the common, safety-related equipment, and the common, nonsafety-related equipment that provides one of the following nonsafety-related 14.3-93 functions:

Certified Design Material and Inspections, Tests, Analyses, and

  • Provides physical support of irradiated fuel (fuel handling machine, spent fuel storage racks, reactor building crane, and module lifting adapter)
  • Provides a path for makeup water to the UHS
  • Provides containment of UHS water
  • Monitors UHS water level The ITAAC verifies that: (1) a sSeismic qQualification Reportrecord form exists for each Seismic Category I component type, and (2) the seismic qualification record form concludes that the Seismic Category I equipment listed in Tier 1 Table 3.14-1 (Table 14.3-4e),

including its associated supports and anchorages, performs its Acceptance Criteria function under the seismic design basis load conditions specified in the sSeismic qQualification Reportrecord form.

After installation in the plant, an ITAAC inspection is performed to Draft Revision 4 verify that the Seismic Category I equipment listed in Tier 1 Table 3.14-1 (Table 14.3-4e), including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the sSeismic qQualification Reportrecord form.

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 14.03-S1 Table 14.2-47: Emergency Core Cooling System Test # 47 Preoperational test is required to be performed for each NuScale Power Module.

The emergency core cooling system (ECCS) is described in Section 6.3, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The ECCS supports the reactor safety-related Test #47-1 coolant system (RCS) by opening the Module Protection System Test #63-6 ECCS reactor vent valves and reactor recirculation valves when their respective trip valve is actuated by the module protection system (MPS).
2. The ECCS supports the RCS by safety-related Test #47-1 providing recirculated coolant from Module Protection System Test #63-6 the containment to the reactor pressure vessel (RPV) for the removal of core heat.

The ECCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The ECCS supports the RCS by safety-related Module Protection System Test #63-6 providing low temperature overpressure protection (LTOP) for maintaining the reactor coolant pressure boundary.
2. The ECCS supports the containment safety-related Containment System Test #43-1 system (CNTS) by providing a portion of the containment boundary for maintaining containment integrity.
3. The ECCS supports MPS by providing nonsafety related Safety Display and Indication System post accident monitoring instrument Test #66-2 information signals.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests None System Level Test #47-1 Test 47-1 is performed at hot functional testing concurrently with Turbine Generator System Test #33-1 (reference Table 14.2-33) and MPS Test #63-6 to allow testing of ECCS actuation at normal operating pressure and elevated temperatures. Test #33-1 heats the RCS from ambient conditions to the highest temperature achievable by module heatup system (MHS) heating. These hot functional testing conditions provide the highest differential pressure and temperature conditions that can be achieved prior to fuel load.

Test Objective Test Method Acceptance Criteria i.Verify collapsed liquid level remains Ensure the RCS is at normal operating i.RPV riser level remains above the top of above the top of the core during ECCS pressure and at maximum temperature the core.

actuation. achievable by warming the RCS using ii. CNV pressure remains below design MHS heating.

ii. Containment vessel (CNV) pressure pressure identified in Table 6.2-1.

remains below the design pressure i. Manually initiate ECCS from the main iii. CNV temperature remains below limit during ECCS actuation. control room (MCR). design temperature identified in iii. Containment temperature remains ii. Allow RPV riser level and CNV level to Table 6.2-1.

below the design temperature limit become relatively stable.

during ECCS actuation.

Tier 2 14.2-112 Draft Revision 4

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 04.06-2, RAI 14.03-S1 Table 14.2-107: Remote Shutdown Workstation Test # 107Not Used The remote shutdown station (RSS) is described in Section 7.1.1.2.3. Testing associated with the RSS occurs during the performance of factory acceptance testing (FAT) and site acceptance testing (SAT) as described below.

The RSS provides an alternate location to monitor the NuScale Power Module status and operate the module control system (MCS) and plant control system (PCS) during a main control room (MCR) evacuation. The ability to activate the nonsafety MCS and PCS displays and controls at the RSS will be verified during SAT. The ability to isolate the safety-related MCR module protection system (MPS) manual switches using the MCR isolation switches in the RSS as described in Section 7.2.12 will be verified during MPS FAT and SAT.

Refer to Table 14.2-61: Module Control System Test #61 and Table 14.2-62: Plant Control System Test #62 for details regarding MCS and PCS FAT and SAT.

Tier 2 14.2-204 Draft Revision 4

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 03.09.02-10, RAI 04.06-2, RAI 14.03-3S1 Table 14.2-109: List of Test Abstracts Test Number System Abbreviation Test Abstract 1 SFPCS Spent Fuel Pool Cooling System 2 PCUS Pool Cleanup System 3 RPCS Reactor Pool Cooling System 4 PSCS Pool Surge Control System 5 UHS Ultimate Heat Sink 6 PLDS Pool Leakage Detection System 7 RCCWS Reactor Component Cooling Water System 8 CHWS Chilled Water System 9 ABS Auxiliary Boiler System 10 CWS Circulating Water System 11 SCWS Site Cooling Water System 12 PWS Potable Water System 13 UWS Utility Water System 14 DWS Demineralized Water System 15 NDS Nitrogen Distribution System 16 SAS Service Air System 17 IAS Instrument Air System 18 CRHS Control Room Habitability System 19 CRVS Normal Control Room HVAC System 20 RBVS Reactor Building HVAC System 21 RWBVS Radioactive Waste Building HVAC System 22 TBVS Turbine Building HVAC System 23 RWDS Radioactive Waste Drain System 24 BPDS Balance-of-Plant Drain System 25 FPS Fire Protection System 26 FDS Fire Detection System 27 MSS Main Steam System 28 FWS Feedwater System 29 FWTS Feedwater Treatment System 30 CPS Condensate Polishing System 31 HVDS Feedwater Heater Vents and Drains System 32 CARS Condenser Air Removal System 33 TGS Turbine Generator System 34 TLOSS Turbine Lube Oil Storage System 35 LRWS Liquid Radioactive Waste System 36 GRWS Gaseous Radioactive Waste System 37 SRWS Solid Radioactive Waste System 38 CVCS Chemical and Volume Control System 39 BAS Boron Addition System 40 MHS Module Heatup System 41 CES Containment Evacuation System 42 CFDS Containment Flooding and Drain System 43 CNTS Containment System 44 N/A Not Used 45 N/A Not Used 46 RCS Reactor Coolant System 47 ECCS Emergency Core Cooling System 48 DHRS Decay Heat Removal System 49 ICIS In-core Instrumentation System Tier 2 14.2-206 Draft Revision 4

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-109: List of Test Abstracts (Continued)

Test Number System Abbreviation Test Abstract 99 N/A Steam Generator Level Control 100 N/A Ramp Change in Load Demand 101 N/A Step Change in Load Demand 102 N/A Loss of Feedwater Heater 103 N/A 100 Percent Load Rejection 104 N/A Reactor Trip from 100 Percent Power 105 N/A Island Mode Test for the First NuScale Power Module 106 N/A Island Mode Test for Multiple NuScale Power Modules 107 N/A Not UsedRemote Shutdown Workstation 108 N/A NuScale Power Module Vibration Tier 2 14.2-208 Draft Revision 4

NuScale Tier 1 Radioactive Waste Building RAI 14.03-3, RAI 14.03-3S1 Table 3.12-2: Radioactive Waste Building ITAAC No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The RWB includes radiation shielding An inspection and analysis will be A report exists and concludes the barriers for normal operation and performed of the as-built RWB radiation attenuation capability of post-accident radiation shielding. radiation shielding barriers. RWB radiation shielding barriers is greater than or equal to the required attenuation capability of the approved design.

2 The RWB includes radiation An inspection will be performed of the The RWB radiation attenuating doors attenuating doors for normal as-built RWB radiation attenuating are installed in their design location operation and for post-accident doors. and have a radiation attenuation radiation shielding. These doors have a capability that meets or exceeds that radiation attenuation capability that of the wall within which they are meets or exceeds that of the wall installed.

within which they are installed.

3 The RWB is an RW-IIa structure and A reconciliation analysis will be A design summary report exists and maintains its structural integrity under performed of the as-built RW-IIa RWB concludes that (1) the as-built RWB the design basis loads. under the actual design basis loads.An maintains its structural integrity in inspection and analysis will be accordance with the approved design performed of the as-built RW-IIa RWB. under the actual design basis loads, and (2) the in-structure responses for the as-built RWB are enveloped by those in the approved design.the deviations between the drawings used for construction and the as-built RW-IIa RWB have been reconciled and that the as-built RW-IIa RWB maintains its structural integrity under the design basis loads.

Tier 1 3.12-3 Draft Revision 4

Tier 2 NuScale Final Safety Analysis Report RAI 09.01.04-1, RAI 09.05.01-6, RAI 14.03-3, RAI 14.03-3S1, RAI 14.03.02-1, RAI 14.03.02-2, RAI 14.03.03-1, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.03-8, RAI 14.03.07-1, RAI 14.03.08-1S1, RAI 14.03.09-1, RAI 14.03.09-2, RAI 14.03.09-3, RAI 14.03.12-2, RAI 14.03.12-3, RAI 18-46S1 Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, X Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, to demonstrate that air exfiltration from the CRE is controlled. RG 1.197 allows two options for CRE testing; either integrated testing (tracer gas testing) or component testing. Section 6.4 Control Room Habitability, describes the testing requirements for the CRE habitability program. Section 6.4 provides the maximum air exfiltration allowed from the CRE.

In accordance with Table 14.2-18, a preoperational test using the tracer gas test method demonstrates that the air exfiltration from the CRE does not exceed the assumed unfiltered leakage rate provided in Table 6.4-1: Control Room Habitability System Design Parameters 14.3-59 for the dose analysis. Tracer gas testing in accordance with ASTM Certified Design Material and Inspections, Tests, Analyses, and E741 will be performed to measure the unfiltered in-leakage into the CRE with the control room habitability system (CRHS) operating.

03.01.02 CRH The CRHS valves are tested by remote operation to demonstrate the X capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18, a preoperational test demonstrates that each CRHS valve listed in Tier 1 Table 3.1-1 (Table 14.3-4a) strokes fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate Acceptance Criteria design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

Draft Revision 4

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.12.03 RWB The RW-IIa RWB and its design basis loads are discussed in X X Section 3.8.4.1.3, Radioactive Waste Building. Design basis loads for RW-IIa structures as listed in RG 1.143.Guidance for the content and structure of the as-built design report is provided in Standard Review Plan Section 3.8.4, Appendix C.

The scope of this ITAAC is a reconciliation of deviations between the issued for construction drawings that implement the seismic and dynamic analyses and the as-built structures. The design report provides criteria for the reconciliation. Design basis loads for RW-IIa structures as listed in RG 1.143 are:

  • wind
  • tornado 14.3-86
  • tornado missile Certified Design Material and Inspections, Tests, Analyses, and
  • flood
  • precipitation (rain, snow)
  • accidental explosion (fixed facility)
  • accidental explosion (transportation vehicle)
  • malevolent vehicle assault
  • small aircraft crash An ITAAC inspection and reconciliation analysis of the as-built RW-IIa RWB is performed of the as-built RW-IIa RWB to ensure that Acceptance Criteria deviations between the drawings used for construction and the as-built RW-IIa RWB are reconciled and the as-built RW-IIa RWB maintains its structural integrity under the design basis loads in Draft Revision 4 accordance with the approved design under the actual design basis loads, and the in-structure responses for the RWB are enveloped by those in the approved design. The design summary report provides criteria for the reconciliation between design and as-built conditions, as described in Section 3.8.4.5.1.

Tier 2 NuScale Final Safety Analysis Report RAI 14.03-3S1 Table 14.3-3h: Module Specific Mechanical and Electrical/I&C Equipment Equipment Description Location EQ Qualification Seismic Class 1E EQ Identifier Environment Program Category I Category(1)

Containment System CNV-8 CNTS I&C Division I Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNV-9 CNTS I&C Division II Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNV-15 CNTS PZR Heater Power #1 Electrical RXB - Top of Module Harsh Electrical Yes No A Penetration Assembly (EPA) RXB - Inside Containment Mechanical CNV-16 CNTS PZR Heater Power #2 Electrical RXB - Top of Module Harsh Electrical Yes No A Penetration Assembly (EPA) RXB - Inside Containment Mechanical CNV-17 CNTS I&C Channel A Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical CNV-18 CNTS I&C Channel B Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical 14.3-119 CNV-19 CNTS I&C Channel C Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Certified Design Material and Inspections, Tests, Analyses, and Assembly (EPA) RXB - Inside Containment Mechanical CNV-20 CNTS I&C Channel D Electrical Penetration RXB - Top of Module Harsh Electrical Yes Yes A Assembly (EPA) RXB - Inside Containment Mechanical CNV-37 CNTS CRD Power Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNV-38 CNTS RPI Group #1 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical CNV-39 CNTS RPI Group #2 Electrical Penetration RXB - Top of Module Harsh Electrical Yes No A Assembly (EPA) RXB - Inside Containment Mechanical MS-HOV-0101 MS #1 CIV (MSIV #1) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical Acceptance Criteria MS-HOV-0201 MS #2 CIV (MSIV #2) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical MS-HOV-0103 MS line #1 Bypass Valve (MSIV Bypass #1) RXB - Top of Module Harsh Electrical Yes Yes AB Draft Revision 4 Mechanical MS-HOV-0203 MS line #2 Bypass Valve (MSIV Bypass #2) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical FW-HOV-0137 FW #1 CIV (FWIV #1) RXB - Top of Module Harsh Electrical Yes Yes AB Mechanical

Table 14.3-3h: Module Specific Mechanical and Electrical/I&C Equipment (Continued)

Tier 2 NuScale Final Safety Analysis Report Equipment Description Location EQ Qualification Seismic Class 1E EQ Identifier Environment Program Category I Category(1)

RPV7 Steam Plenums (4 Total) RXB - Inside Containment N/A N/A Yes N/A N/A RPV8 RPV9 RPV10 RPV3 Feedwater Plenums (4 Total) RXB - Inside Containment N/A N/A Yes N/A N/A RPV4 RPV5 RPV6 None Flow Restrictors RXB - Inside Containment N/A N/A Yes N/A N/A SG-PSV-1002 Thermal Relief Valves (2 Total) RXB - Inside Containment Harsh Mechanical Yes N/A B SG-PSV-2002 Control Rod Drive System CRDS-ZS- Rod Position Indication (RPI) Coils (324 RXB - Inside Containment Harsh Electrical Yes No B 14.3-124 0001A to 0016A Total)

Certified Design Material and Inspections, Tests, Analyses, and CRDS-ZS-0001B to 0016B None Control Rod Drive Shafts RXB - Inside Containment N/A N/A Yes N/A N/A None Control Rod Drive Latch Mechanism RXB - Inside Containment N/A N/A Yes N/A N/A None CRDM Pressure Boundary (Latch Housing, RXB - Inside Containment N/A N/A Yes N/A N/A Rod Travel Housing, Rod Travel Housing Plug)

Control Rod Assembly None All components RXB - Inside Containment N/A N/A Yes N/A N/A Neutron Source Assembly None Primary and secondary neutron source RXB - Inside Containment N/A N/A Yes N/A N/A Acceptance Criteria rodlets None Spider body, hub or coupling housing RXB - Inside Containment N/A N/A Yes N/A N/A Draft Revision 4 Reactor Coolant System None Reactor Vessel Internals RXB - Inside Containment N/A N/A Yes N/A N/A

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