ML19200A248

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

ML19200A248
Person / Time
Site:	NuScale
Issue date:	07/19/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0719-66362
Download: ML19200A248 (123)

Text

RAIO-0719-66362 July 19, 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 Response to NRC Request for Additional Information No.

522 (eRAI No. 9681) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

522 (eRAI No. 9681)," dated May 30, 2019 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's 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 Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Cayetano Santos, NRC, OWFN-8H12 : NuScale 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-0719-66362 :

NuScale 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 NuScale Nonproprietary

the ITAAC comply with 10 CFR 52.47(b)(1):

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.

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.

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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.

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:

RAI 9681 Question 14.03-03 included two items to be addressed prior to a list of proposed revisions. This section address the two items.

NuScale Nonproprietary

NRC Item 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.

NuScale Response to NRC Item 1:

ITAAC are not necessary to verify the functionality or the location of the MPS MCR isolation switches. The MCR isolation switches, when repositioned, function to isolate the MPS manual actuation switches in the main control room. This function does not establish and maintains safe shutdown conditions, and failure of this function does not prevent these conditions from being established. While there is not an explicit Design Commitment for the function, the functionality of these switches will be verified as part of Table 2.5-7 ITAAC 02.05.01. Accordingly, ITAAC 02.05.29 has been deleted.

NRC Item 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.

NuScale Response to NRC Item 2:

Design Commitments associated with the following ITAAC were revised to provide alignment between the Tier 1 Design Descriptions and the associated ITAAC tables:

• Table 2.1-4 o 02.01.01 o 02.01.08 o 02.01.13 through 15 o 02.01.18 through 21 NuScale Nonproprietary

o 02.01.24

• Table 2.2-3 o 02.02.03

• Table 2.6-1 o 02.06.02 o 02.06.03

• Table 3.1-2 o 03.01.01

• Table 3.3-1 o 03.03.03

• Table 3.6-2 o 03.06.02

• Table 3.13-1 o 03.13.04

• Table 3.16-1 o 03.16.01 through 05 Where the difference between the Design Commitment in the Design Description section of Tier 1 and the Design Commitment in the ITAAC table is exclusively the result of defining an abbreviation during the first use of the term in the document, the difference is intentional, and therefore no changes were made. For example, Section 2.1 contains the following Design Commitment:

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

In the associated ITAAC table, Table 2.4-1, the Design Commitment is written as:

The NuScale Power Module ASME Code Class 2 piping systems and interconnected equipment nozzles are evaluated for LBB.

The term leak-before-break is used in its entirety for the first usage, while the associated abbreviation is used thereafter.

RAI 9681 Additional Explanations The RAI included a set of additional explanations for the basis of the Staffs proposed revisions.

This section describes how each of the additional explanations were dispositioned.

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1. The proposed definition for Approved design, and the proposed definition revisions for the terms ASME Code and Reconciliation were incorporated into Tier 1, Section 1.1.

The following terms were arranged alphabetically, as suggested, but unmodified:

o Common or Shared ITAAC o Module-Specific ITAAC o Type Test

2. Incorporated the proposed explanatory material into Tier 1, Section 1.2.4.

3. Proposed changes incorporated. Tier 2, Table 14.3-1 was also revised to be consistent with the incorporated changes.

4. Proposed changes incorporated.

5. Proposed changes incorporated.

6. Proposed changes partially incorporated. The proposed revision for Tier 1 Table 2.5-7 ITAAC 02.05.06 was slightly altered. The concern was that the Design Commitment, as proposed, could be misinterpreted to mean communications independence exists between the separation groups and their associated divisions. To avoid this potential confusion, the Design Commitment was divided into two distinct parts, which align with the Acceptance Criteria.

7. Proposed changes incorporated.

8. The proposed changes for the Design Commitment and Acceptance Criterion were partially incorporated. The proposed change to the ITA was unnecessary, and therefore not incorporated to remain consistent with similar ITA in Table 2.5-7. The proposed changes included changing "one of its protection channels" to "any of its protection channels." This makes the Design Commitment false, as placing all channels in bypass (which is not procedurally permitted) would prohibit the performance of a safety-related function. Additionally, the term "protection channel" was replaced with "separation group" for technical accuracy.

9. Proposed changes incorporated.

10. Proposed changes incorporated.

11. Proposed changes incorporated.

12. Proposed changes incorporated.

13. Proposed changes incorporated. Tier 2, Table 14.3-2 was also revised to be consistent with the incorporated changes.

14. Proposed changes incorporated.

15. Proposed changes incorporated.

16. Proposed changes incorporated.

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17. Proposed changes incorporated. Tier 2, Table 14.3-2 was also revised to explicitly state the RBC wet hoist is included in the scope in support of deleting ITAAC 03.10.08.

18. Proposed changes incorporated. Tier 2, Table 14.3-2 was also revised to delete 03.10.08 wording.

19. Proposed changes incorporated.

20. Proposed changes incorporated.

21. Proposed changes incorporated.

22. Proposed changes incorporated.

23. Proposed changes incorporated.

24. Proposed changes incorporated.

25. Proposed changes incorporated.

26. Proposed changes incorporated.

RAI 9681 Proposed Changes In addition to the items to address and the additional explanations sections of the RAI, multiple sections and tables from Tier 1 were provided with proposed changes. This section describes how each of these proposed changes were dispositioned.

Multiple ITAAC were identified with table numbers referenced in the Acceptance Criteria, but not the associated Design Commitment and ITA. The proposal was made to add the table numbers to the Design Commitment and ITA, and that proposal was incorporated, for the following ITAAC:

• Table 2.1-4 o 02.01.14 o 02.01.15 o 02.01.18 through 21

• Table 2.8-2 o 02.08.01 o 02.08.06 o 02.08.08

• Table 3.14-2 o 03.14.01 o 03.14.03 To ensure consistency, all ITAAC were reviewed to verify the use table numbers was applied to the Design Commitment and ITA when used in the Acceptance Criterion. Although not NuScale Nonproprietary

specifically proposed by the RAI, and as a result of this review, table numbers were added to the following ITAAC:

• Table 2.1-4 o 02.01.01 through 03 o 02.01.05 o 02.01.08 through 10 o 02.01.13

• Table 2.2-3 o 02.02.01 through 03 o 02.02.05

• Table 2.5-7 o 02.05.11 o 02.05.13

• Table 2.7-2 o 02.07.01 o 02.07.02

• Table 2.8-2 o 02.08.02 through 05 o 02.08.07 o 02.08.09

• Table 3.1-2 o 03.01.02 o 03.01.03

• Table 3.2-2 o 03.02.01

• Table 3.6-2 o 03.06.02

• Table 3.14.2 o 03.14.02 NuScale Nonproprietary

• Table 3.17-2 o 03.17.01 o 03.17.02

• Table 3.18-2 o 03.18.01 o 03.18.02 The following proposed wording changes provided in the RAI, but not specifically identified in the additional explanations section of the RAI, were incorporated:

• Tier 1, Section 1.2.4, last paragraph

• Table 2.1-4 o 02.01.14 o 02.01.15 o 02.01.18 through 21 o 02.01.24

• Table 2.2-3 o 02.02.05

• Table 2.3-1 o 02.03.01 o 02.03.02, only the ITA change

• Table 2.5-7 o 02.05.18 through 20 o 02.05.27

• Table 2.8-2 o 02.08.01 o 02.08.06 o 02.08.07 o 02.08.08

• Table 3.4-1 o 03.04.02 o 03.04.03 NuScale Nonproprietary

• Table 3.6-2 o 03.06.02

• Table 3.7-1 o 03.07.02, only the Design Commitment change

• Table 3.10-1 o 03.10.04 through 06 o 03.10.09

• Table 3.11-2 o 03.11.07 (Corresponding changes also made to Section 3.11.1, and Tier 2 Table 14.3.2.)

Not all of the changes proposed in the RAI were incorporated. The following is a list of those proposed changes, and a brief description as to why they were not incorporated:

• Tier 1, Section 1.2.4: The word and was not added between design commitments and inspections because it is unnecessary.

• Table 2.3-1, ITAAC 02.03.02: The acceptance criteria was not revised to add pressure instrumentation. The acceptance criteria already specifies CES inlet pressure instrumentation (PIT-1001/PIT-1019).

• Table 3.1-2, ITAAC 03.01.01: The Acceptance Criteria was not revised as proposed.

The proposed phrase meets the air exfiltration assumed could be interpreted to mean equals the air exfiltration assumed. This ITAAC verifies the value used in the analysis was conservative. Therefore, demonstrating actual air exfiltration is less than the assumed value used in the analysis is the appropriate Acceptance Criteria. Additionally, the Design Commitment in this table was revised from meets the assumptions to does not exceed the assumptions to align with the wording used in the Design Commitment associated with this ITAAC in Section 3.1.1, Design Description.

• Table 3.7-1, ITAAC 03.07.02: The proposed change to the Acceptance Criteria results in the elimination of an explicit requirement to test the installed FPS pumps. As such, incorporating the proposed change is inappropriate.

• Table 3.14-2:

o ITAAC 03.14.01: The proposed change is to use the term Seismic Qualification Report instead of seismic record form. No basis was provided for this proposed change. Additionally, the same change was neither proposed for the second Acceptance Criterion for the same ITAAC (03.14.01), nor for the Acceptance NuScale Nonproprietary

Criteria of 02.08.01, which also uses the term seismic record form. The proposed change was not incorporated to maintain consistency.

o ITAAC 03.14.03: The proposed change was to remove the phrase used for processing gaseous radioactive waste from all three areas of the ITAAC. This phrase was specifically included in the ITAAC wording to clearly establish that not all RW-IIa are within scope, and some RW-IIa components were intentionally excluded from Table 3.14-1.

During the review of Tier 1 material associated with this RAI, the following additional changes were identified and incorporated:

• Tier 1, Table 2.3-1: An editorial hyphen was added.

• Tier 1, Section 2.7: The equipment IDs were removed from Section 2.7.1, Design Description, and Table 2.7-2. The equipment IDs were redundant to Table 2.7-1, and therefore unnecessary once reference to the table was added to the Design Commitments, ITA, and Acceptance Criteria. Tier 2 Table 14.2-38, Chemical and Volume Controls System Test #38, and Tier 2 Table 14.3-1 were revised to reflect the deletion of ITAAC 02.07.03 and 02.07.04.

• Tier 1, Section 2.8:

o The scope of equipment portion of the Design Commitments associated with ITAAC 02.08.01 through 03 were deleted from Section 2.8.1, Design Description, and Table 2.8-2 because it is redundant to the information in the Tier 1 System Description. This information remains in the System Description portion of Tier 1 Section 2.8.1 and Tier 2 Table 14.3-1.

o The words located in a harsh environment were deleted from the Design Commitments associated with ITAAC 02.08.09 in Section 2.8.1 and Table 2.8-2 because they were redundant to words already included in the Design Commitments.

• Tier 1, Section 3.8: Because the actions for alternative (remote) shutdown are taken in the MPS equipment rooms in the Reactor Building and not at operator workstations in the remote shutdown station (RSS), ITAAC to verify normal and emergency illumination levels at the RSS are not required. Table 3.8-1, ITAAC 03.08.03 verifies emergency lighting for the activities in the MPS equipment rooms. In this same table, ITAAC 03.08.01 and 03.08.02 were revised to remove the RSS reference. Section 3.8.1 and Tier 2 Table 14.3.2 were revised to remove the reference to the RSS. Tier 2, Table 14.2-60, Plant Lighting System Test #60, was revised to remove the ITAAC reference from the RSS lighting component level tests.

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• Tier 1, Section 3.9: The equipment IDs were removed from Section 3.9.1, Design Description, and Table 3.9-2. The equipment IDs were redundant to Table 3.9-1, and therefore unnecessary once reference to the table was added to the Design Commitments, ITA, and Acceptance Criteria. Tier 2 Tables 14.2-9, Auxiliary Boiler System Test # 9, and 14.2-36, Gaseous Radioactive Waste System Test #36, and Tier 2 Table 14.3-2 were revised to reflect the deletion of ITAAC 03.09.05, 03.09.06 and 03.09.09.

• Tier 1, Section 3.14:

o The scope of equipment portion of the Design Commitments associated with ITAAC 03.14.01 and 03.14.02 were deleted from Section 3.14.1, Design Description, and Table 3.14-2 because it is redundant to the information in the Tier 1 System Description. This information remains in the System Description portion of Tier 1 Section 3.14.1 and Tier 2 Table 14.3-2.

o Five module specific ITAAC were relocated to Table 3.14-2. This change was made based on the Tier 1 Section 1.1 definition for Common or Shared ITAAC, which includes analyses or other generic design and qualification activities that are identical for each NPM (e.g., environmental qualification of equipment).

For a multi-module plant, satisfactory completion of a common or shared ITAAC for the lead NPM shall constitute satisfactory completion of the common or shared ITAAC for associated NPMs. To ensure future users understand these five ITAAC are only required to be performed once, they were moved from Tier 1 Section 2 tables to Tier 1 Section 3 tables. Their corresponding Tier 2 Table 14.3-1 descriptions were moved to Tier 2 Table 14.3-2, accordingly. The following is the list of ITAAC that were moved and their new ITAAC numbers:

§ Table 02.01-4 ITAAC 02.01.22 is now Table 03.14-2 ITAAC 03.14.04

§ Table 02.08-2:

• ITAAC 02.08.03 is now Table 03.14-2 ITAAC 03.14.05

• ITAAC 02.08.05 is now Table 03.14-2 ITAAC 03.14.06

• ITAAC 02.08.06 is now Table 03.14-2 ITAAC 03.14.07

• ITAAC 02.08.08 is now Table 03.14-2 ITAAC 03.14.08

• Tier 1, Section 3.17: The equipment IDs were removed from Section 3.17.1, Design Description, and Table 3.17-2. The equipment IDs were redundant to Table 3.17-1, and therefore unnecessary once reference to the table was added to the Design Commitments, ITA, and Acceptance Criteria. Tier 2 Table 14.2-24, Balance-of-Plant Drain System Test #24, and Tier 2 Table 14.3-2 were revised to reflect the deletion of ITAAC 03.17.03 and 03.17.04.

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• Tier 1, Section 3.18: The equipment IDs were removed from Section 3.18.1, Design Description, and Table 3.18-2. The equipment IDs were redundant to Table 3.18-1, and therefore unnecessary once reference to the table was added to the Design Commitments, ITA, and Acceptance Criteria. Tier 2 Table 14.2-24, Balance-of-Plant Drain System Test #24, and Tier 2 Table 14.3-2 were revised to reflect the deletion of ITAAC 03.18.03.

Impact on DCA:

Tier 1 Chapters 1, 2, 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

NuScale Tier 1 Definitions 1.1 Definitions The definitions below apply to terms that may be used in the design descriptions and associated Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC).

Acceptance Criteria refers to the performance, physical condition, or analysis result for structures, systems, and components (SSC), or program that demonstrates that the design commitment is met.

Analysis means a calculation, mathematical computation, or engineering or technical evaluation. Engineering or technical evaluations could include, but are not limited to, comparisons with operating experience or design of similar SSC.

RAI 14.03-3 Approved design means the design as described in the updated final safety analysis report (UFSAR), including any changes to the final safety analysis report (FSAR) since submission to the NRC of the last update of the FSAR.

As-built means the physical properties of an SSC following the completion of its installation or construction activities at its final location at the plant site. In cases where it is technically justifiable, determination of physical properties of the as-built SSC may be based on measurements, inspections, or tests that occur prior to installation, provided that subsequent fabrication, handling, installation, and testing do not alter the properties.

RAI 14.03-3, RAI 14.03.03-8 ASME Code meansSection III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, as endorsed in 10 CFR 50.55aincorporated by reference in 10 CFR 50.55a with specific conditions or in accordance with relief granted or alternatives authorized by the NRC pursuant to 10 CFR 50.55a, unless a different section of the ASME Code is specifically referenced.

ASME Code Data Report means a document that certifies that a component or system is constructed in accordance with the requirements of the ASME Code. This data is recorded on a form approved by the ASME.

RAI 14.03-3 Common or Shared ITAAC means ITAAC that are associated with common or shared SSC and activities that support multiple NPMs. This includes (1) SSC that are common or shared by multiple NPMs, and for which the interface and functional performance requirements between the common or shared SSC and each NPM are identical, or (2) analyses or other generic design and qualification activities that are identical for each NPM (e.g., environmental qualification of equipment). For a multi-module plant, satisfactory completion of a common or shared ITAAC for the lead NPM shall constitute satisfactory completion of the common or shared ITAAC for associated NPMs.

Component, as used for reference to ASME Code components, means a vessel, concrete containment, pump, pressure relief valve, line valve, storage tank, piping system, or core support structure that is designed, constructed, and stamped in accordance with the rules of the ASME Code. ASME Code Section III classifies a metal containment as a vessel.

Tier 1 1.1-1 Draft Revision 3

NuScale Tier 1 Definitions Design Commitment means that portion of the design description that is verified by ITAAC.

Design Description means that portion of the design that is certified. Design descriptions consist of a system description, system description tables, system description figures, and design commitments. System description tables and system description figures are only used when appropriate. The system description is not verified by ITAAC; only the design commitments are verified by ITAAC. System description tables and system description figures are only verified by ITAAC if they are referenced in the ITAAC table.

Inspect or Inspection means visual observations, physical examinations, or reviews of records based on visual observation or physical examination that compare (a) the SSC condition to one or more design commitments or (b) the program implementation elements to one or more program commitments, as applicable. Examples include walkdowns, configuration checks, measurements of dimensions, or nondestructive examinations. The terms, inspect and inspection, also apply to the review of Emergency Planning ITAAC requirements to determine whether ITAAC are met.

ITAAC are those Inspections, Tests, Analyses, and Acceptance Criteria identified in the combined license that if met by the licensee are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Atomic Energy Act, as amended, and the Commission's rules and regulations.

RAI 14.03-3 Module-Specific ITAAC means ITAAC that are associated with SSC that are specific to and support operation of a single, individual NuScale Power Module. Module-specific ITAAC shall be satisfactorily completed for each NuScale Power Module.

NuScale Power Module (NPM) is a collection of systems, sub-systems, and components that together constitute a modularized, movable, nuclear steam supply system. The NPM is composed of a reactor core, a pressurizer, and two steam generators integrated within a reactor pressure vessel and housed in a compact steel containment vessel.

RAI 14.03-3 Reconciliation or Reconciled means the identification, assessment, and disposition of differences between an approved design featurea design feature as described in the Updated Final Safety Analysis Report and an as-built plant design feature. For ASME Code piping systems, it is the reconciliation of differences between the approved designdesign as described in the UFSAR and the as-built piping system. For structural features, it is the reconciliation of differences between the approved designdesign as described in the UFSAR and the as-built structural feature.

Report, as used in the ITAAC table Acceptance Criteria column, means a document that verifies that the acceptance criteria of the subject ITAAC have been met and references the supporting documentation. The report may be a simple form that consolidates all of the necessary information related to the closure package for supporting successful completion of the ITAAC.

RAI 14.03-3 Tier 1 1.1-2 Draft Revision 3

NuScale Tier 1 Definitions Common or Shared ITAAC means ITAAC that are associated with common or shared SSC and activities that support multiple NPMs. This includes (1) SSC that are common or shared by multiple NPMs, and for which the interface and functional performance requirements between the common or shared SSC and each NPM are identical, or (2) analyses or other generic design and qualification activities that are identical for each NPM (e.g., environmental qualification of equipment). For a multi-module plant, satisfactory completion of a common or shared ITAAC for the lead NPM shall constitute satisfactory completion of the common or shared ITAAC for associated NPMs.

RAI 03.07.02-24S1 Safe Shutdown Earthquake (SSE) Ground Motion is the site-specific vibratory ground motion for which safety-related SSC are designed to remain functional. The SSE for a site is a smoothed spectra developed to envelop the ground motion response spectra. The SSE is characterized at the free ground surface. A combined license (COL) applicant may use the SSE for design of site-specific SSC.

System Description (Tier 1) includes

• a concise description of the system's or structure's safety-related functions, nonsafety-related functions that support safety-related functions, and certain nonsafety risk-significant functions.

• a listing of components required to perform those functions.

• identification of the system safety classification.

• the system components general locations.

The system description may include system description tables and figures.

Test means actuation or operation, or establishment of specified conditions, to evaluate the performance or integrity of as-built SSC, unless explicitly stated otherwise, to determine whether ITAAC are met.

Tier 1 means the portion of the design-related information contained in the generic Design Control Document that is approved and certified by the design certification rule (Tier 1). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 includes:

• definitions and general provisions

• design descriptions

• ITAAC

• significant site parameters

• significant interface requirements RAI 14.03-3 Type Test means a test on one or more sample components of the same type and manufacturer to qualify other components of the same type and manufacturer. A type test is not necessarily a test of an as-built SSC.

Tier 1 1.1-3 Draft Revision 3

NuScale Tier 1 Definitions Top-Level Design Features means the principal performance characteristics and physical attributes that are important to performing the safety-related and certain nonsafety-related functions of the plant.

RAI 14.03-3 Module-Specific ITAAC means ITAAC that are associated with SSC that are specific to and support operation of a single, individual NuScale Power Module. Module-specific ITAAC shall be satisfactorily completed for each NuScale Power Module.Type Test means a test on one or more sample components of the same type and manufacturer to qualify other components of the same type and manufacturer. A type test is not necessarily a test of an as-built SSC.

Tier 1 1.1-4 Draft Revision 3

NuScale Tier 1 General Provisions attributes depicted on these figures, provided that the top-level design features discussed in the design description pertaining to the figure are not adversely affected. Valve position indications shown on system description figures do not represent a specific operational state.

The figure legends in Tier 2 Section 1.7 are used to interpret Tier 1 system description figures.

1.2.4 Implementation of Inspections, Tests, Analyses, and Acceptance Criteria Design commitments, inspections, tests, analyses, and acceptance criteria are provided in ITAAC tables with the following format:

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria RAI 14.03-3 Each commitment in the Design Commitment column of the ITAAC tables has one or more associated requirements for inspections, tests or analyses specified in the Inspections, Tests, Analyses column. Each inspection, test, or analysis has an associated acceptance criterion in the third column of the ITAAC tables that demonstrate that the Design Commitment in the first column has been met.

Inspections, tests, or analyses may be performed by the licensee or by its authorized vendors, contractors, or consultants.

Inspections, tests, or analyses may be

• performed by more than a single individual or group.

• implemented through discrete activities separated by time.

• performed at any time prior to fuel load, including before issuance of the combined license for those ITAAC that do not require as-built equipment.

• performed at a location other than the construction site.

Additionally, inspections, tests, or analyses may be performed as part of other activities such as construction inspections and preoperational testing. Therefore, inspections, tests, or analyses need not be performed as a separate or discrete activity.

If an acceptance criteria does not specify the temperature, pressure, or other conditions under which an inspection or test must be performed, then the inspection or test conditions are not constrained.

RAI 14.03-3 Many of the Acceptance Criteria state that a report or analysis exists and concludes that...

When these words are used, it indicates that the ITAAC for that Design Commitment will be met when it is confirmed that appropriate documentation exists and the documentation shows that the Design Commitment is met.

Tier 1 1.2-2 Draft Revision 3

NuScale Tier 1 General Provisions For the acceptance criteria, appropriate documentation may be a single document or a collection of documents that show that the stated acceptance criteria are met. Examples of appropriate documentation include:

• design reports

• test reports

• inspection reports

• analysis reports

• evaluation reports

• design and manufacturing procedures

• certified data sheets

• commercial grade dedication procedures and records

• quality assurance records

• calculation notes

• equipment qualification data packages RAI 14.03-3 Conversion or extrapolation of test results from the test conditions to the design conditions may be necessary to satisfy an ITAAC. Suitable justification should be provided for, and applicability of, any necessary conversions or extrapolations of test results necessary to satisfy an ITAAC.

1.2.5 Acronyms and Abbreviations The acronyms and abbreviations contained in Tier 2 Table 1.1-1 are applicable to Tier 1.

Tier 1 1.2-3 Draft Revision 3

NuScale Tier 1 NuScale Power Module

• The CNTS supports the ECCS by providing structural support of the trip and reset valves for the ECCS reactor vent valves (RVVs) and reactor recirculation valves (RRVs).

• The CNTS supports the RCS by closing the CIVs for pressurizer spray, chemical and volume control system (CVCS) makeup, CVCS letdown, and RPV high point degasification when actuated by module protection system (MPS) for RCS Isolation.

• The CNTS supports the RXB by providing a barrier to contain mass, energy, and fission product release by closure of the CIVs upon containment isolation signal.

• The CNTS supports the DHRS by closing CIVs for main steam and feedwater and opening DHRS actuation valves when actuated by MPS for DHRS operation.

• The ECCS supports the RCS by opening the ECCS reactor vent valves and RRVs when their respective trip valve is actuated by MPS.

• The DHRS supports the RCS by opening the DHRS actuation valves on a DHRS actuation signal.

• The CNTS supports the MPS by providing MPS actuation instrument information signals through the CNV.

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

• The CNTS supports the Reactor Building crane (RBC) by providing lifting attachment points that the RBC can connect to so that the NPM can be lifted.

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

• The CNTS supports the SGS by providing structural support for the SGS piping.

• The CNTS supports the CRDS by providing structural support for the CRDS piping.

• The CNTS supports the RCS by providing structural support for the RCS piping.

• The CNTS supports the feedwater system (FWS) by providing structural support for the FWS piping.

Design Commitments RAI 14.03-3

• The NPMNuscale Power Module American Society of Mechanical Engineers (ASME)

Code Class 1, 2 and 3 piping systems listed in Table 2.1-1 comply with ASME Code Section III requirements.

RAI 14.03-3

• 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.

Tier 1 2.1-3 Draft Revision 3

NuScale Tier 1 NuScale Power Module

• 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

• The CIV cClosure 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.

• Physical separation exists between the redundant divisions of the MPS Class 1E instrumentation and control current-carrying circuits, and between Class 1E instrumentation and control current-carrying circuits and non-Class 1E instrumentation and current-carrying circuits. The scope of this commitment includes the cables from the NPM disconnect box to the instrument.

RAI 14.03-3

• 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 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.The CNTS safety-related valves change position under design differential pressure.

RAI 14.03-3

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

RAI 14.03-3

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

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

NuScale Tier 1 NuScale Power Module

• The CNTS safety-related 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.design differential pressure.

RAI 14.03-3

• The ECCS safety-related 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.design differential pressure.

RAI 14.03-3

• The DHRS safety-related 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.design differential pressure.

RAI 14.03-3

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

RAI 08.01-1S1, RAI 08.01-2, RAI 14.03-3

• A CNTS containment electrical penetration assembly is rated to withstand fault currents for the time required to clear the fault from its power source, or a CNTS containment electrical penetration assembly is rated to withstand the maximum fault 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 3

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 14.03.03-3S1, 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. Containment isolation valve cClosure A test will be performed of the Each CIV listed in Table 2.1-3 travels times for CIVs listed in Table 2.1-3 limit automatic CIVs listed in Table 2.1-3. from the full open to full closed potential releases of 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 2.1-13 Draft Revision 3

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 assembliesy 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. Physical separation exists between the An inspection will be performed of the i. Physical separation between redundant divisions of the MPS Class MPS Class 1E as-built instrumentation redundant divisions of MPS Class 1E instrumentation and control and control current-carrying 1E instrumentation and control current-carrying circuits, and between circuitsNot used. current-carrying circuits is Class 1E instrumentation and control provided by a minimum current-carrying circuits and non-Class separation distance, or by barriers 1E instrumentation and current- (where the minimum separation carrying circuits. The scope of this distances cannot be maintained),

commitment includes the cables from or by a combination of separation the NPM disconnect box to the distance and barriers.

instrumentNot used. ii. Physical separation between MPS Class 1E instrumentation and control current-carrying circuits and non-Class 1E instrumentation and control current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained),

or by a combination of separation distance and barriersNot 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 RPV surveillance program.approximately surveillance program. 90 degree intervals.

Tier 1 2.1-14 Draft Revision 3

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

13. The remotely-operated CNTS A test will be performed of the CNTS Each remotely-operated CNTS containment isolation valves listed in safety-relatedremotely-operated CNTS containment isolation valve listed in Table 2.1-2 change position under containment isolation valves listed in Table 2.1-2 strokes fully open and fully design-basis temperature, differential Table 2.1-2 under preoperational closed by remote operation under pressure, and flow conditions. temperature, differential pressure, and preoperational temperature, flow conditions. differential pressure, and flow conditions.

14. The ECCS safety-related valves listed in A test will be performed of the ECCS Each ECCS safety-related valve listed in Table 2.1-2 change position under safety-related valves listed in Table 2.1-2 strokes fully open and fully design-basis temperature, differential Table 2.1-2 under preoperational closed by remote operation under pressure, and flow conditions. temperature, differential pressure, and preoperational temperature, flow conditions. differential pressure, and flow conditions.

15. The DHRS safety-related valves listed A test will be performed of the DHRS Each DHRS safety-related valve listed in Table 2.1-2 change position under safety-related valves listed in in Table 2.1-2 strokes fully open and design-basis temperature, differential Table 2.1-2 under preoperational fully closed by remote operation pressure, and flow conditions. temperature, differential pressure, and under preoperational temperature, flow conditions. differential pressure, and flow conditions.

16. Not used. Not used. Not used.

17. Not used. Not used. Not used.

18. The CNTS safety-related A test will be performed of the CNTS Each CNTS safety-related hydraulic-operated valves listed in safety-related hydraulic-operated hydraulic-operated valve listed in Table 2.1-2 fail to (or maintain) their valves listed in Table 2.1-2 under Table 2.1-2 fails to (or maintains) its safety-related position on loss of preoperational temperature, safety-related position on loss of electrical power under design-basis differential pressure, and flow motive power under preoperational temperature, differential pressure, and conditions. temperature, differential pressure, and flow conditions. flow conditions.

19. The ECCS safety-related RRVs and RVVs A test will be performed of the ECCS Each ECCS safety-related RRV and RVV listed in Table 2.1-2 fail to (or maintain) safety-related RRVs and RVVs 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 electrical power to their temperature, differential pressure, and positionopen on loss of electrical corresponding trip valves under flow conditions. power to its corresponding trip valve design-basis temperature, differential under preoperational temperature, pressure, and flow conditions. differential pressure, and flow conditions.

20. The DHRS safety-related hydraulic- A test will be performed of the DHRS Each DHRS safety-related hydraulic-operated valves listed in Table 2.1-2 safety-related hydraulic-operated operated valve listed in Table 2.1-2 fail to (or maintain) their safety-related valves listed in Table 2.1-2 under fails to (or maintains) its safety-related position on loss of electrical power preoperational temperature, positionopen on loss of motive power under design-basis temperature, differential pressure, and flow under preoperational temperature, differential pressure, and flow conditions. differential pressure, and flow conditions. conditions.

21. The CNTS safety-related check valves A test will be performed of the CNTS Each CNTS safety-related check valve listed in Table 2.1-2 change position safety-related check valves listed in listed in Table 2.1-2 strokes fully open under design-basis temperature, Table 2.1-2 under preoperational and closed (under forward and reverse differential pressure, and flow temperature, differential pressure, and flow conditions, respectively) under conditions. flow conditions. preoperational temperature, differential pressure, and flow conditions.

Tier 1 2.1-15 Draft Revision 3

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

22. i. A CNTS containment electrical i. An analysis will be performed of the i. A circuit interrupting device penetration assembly is rated to CNTS as-built containment electrical coordination analysis exists and withstand fault currents for the time penetration assembly.Not used. concludes that the current carrying required to clear the fault from its capability for each CNTS containment power source. electrical penetration assembly listed OR in Table 2.1-3 is greater than the analyzed fault currents for the time required to clear the fault from its power source.

OR ii. A CNTS containment electrical ii. An analysis of the CNTS containment penetration assembly is rated to penetration maximum fault current withstand the maximum fault current exists and concludes the fault current for its circuits without a circuit is less than the current carrying interrupting device.Not used. capability of the CNTS containment electrical penetrationNot 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.

24. The NPM lifting fixture supports its A rated load test will be performed of The NPM lifting fixture supports a load rated load. the NPM lifting fixture. of at least 150 percent of the manufacturer's rated capacity.

25. The NPM lifting fixture is constructed An inspection will be performed of the The NPM lifting fixture is single-failure-to provide assurance that a single as-built NPM lifting fixture. proof.

failure does not result in the uncontrolled movement of the lifted load.

26. The ECCS valves, CIVs, and DHRS An inspection will be performed of A report exists and concludes each actuation valves listed in Table 2.1-2, each ECCS valve, CIV, and DHRS ECCS valve, CIV, and DHRS actuation and their associated hydraulic lines, actuation valve listed in Table 2.1-2, valve listed in Table 2.1-2, and the are installed such that each valve can and associated hydraulic line. associated hydraulic line, is installed in perform its safety function. accordance with its associated installation specification.

Tier 1 2.1-16 Draft Revision 3

NuScale Tier 1 Chemical and Volume Control System 2.2 Chemical and Volume Control System 2.2.1 Design Description

System Description

The scope of this section is the chemical and volume control system (CVCS). The system purifies reactor coolant, manages reactor coolant chemistry, provides reactor coolant inventory injection and discharge, and supplies spray flow to the pressurizer to reduce the reactor coolant system pressure. The CVCS is nonsafety-related. Each NuScale Power Module (NPM) has its own module-specific CVCS. The Reactor Building houses all CVCS equipment.

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

• The CVCS supports the RCS by isolating dilution sources.

Design Commitments RAI 14.03-3

• The chemical and volume control system American Society of Mechanical Engineers (ASME) Code Class 3 piping listed in Table 2.2-1 complies with the ASME Code Section III.

RAI 14.03-3

• The chemical and volume control system ASME Code Class 3 components listed in Table 2.2-2 conform to the rules of construction of ASME Code Section III.

RAI 14.03-3

• The chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valves listed in Table 2.2-2 change position under design-basis temperature, differential pressure, and flow conditions.design differential pressure.

RAI 14.03-3

• The chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valves listed in Table 2.2-2 perform their function to fail to (or maintain) their position on loss of motive power under design-basis temperature, differential pressure, and flow conditions.fail to or maintain their safety-related position on loss of motive power under design differential pressure.

2.2.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.2-3 contains the inspections, tests, and analyses for the CVCS.

Tier 1 2.2-1 Draft Revision 3

NuScale Tier 1 Chemical and Volume Control System RAI 14.03-3 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 2.2-1system complies system ASME Code Class 3 as-built the chemical and volume control listed with the ASME Code Section III. piping system Design Report in Table 2.2-1system ASME Code Class 3 required by ASME Code Section III for as-built piping system meets the piping listed in Table 2.2-1. 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 in Table 2.2-2 perform their function isolation valves listed in Table 2.2-2 performs its function to fail to (or to fail to (or maintain) their position under preoperational temperature, maintain) its positionperforms fails on loss of motive power under differential pressure and flow closed on loss of motive power under design-basis temperature, differential conditions. preoperational temperature, differential pressure, and flow conditions. pressure, and flow conditions.

Tier 1 2.2-4 Draft Revision 3

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

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

2. The CES pressure instrumentation A test will be performed of the CES The CES detects a pressure increase in supports RCS leakage detection. pressure instrumentation. the CES inlet pressure instrumentation (PIT-1001/PIT-1019), which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour.

Tier 1 2.3-2 Draft Revision 3

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

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

system conceptfunctional specification phase system requirementsdesign phase system designprototype development phase equipment requirements specification phase hardware planning phase hardware requirements phase hardware design phase.

software planning phase software requirements phase software design phase systemoftware implementation phase software configuration phase system testing phase system installation and checkout phase

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

RAI 14.03-3

• Physical separation exists (i) between eachthe redundant separation groups of the MPS Class 1E instrumentation and control current-carrying circuits,and (ii) between each divisions 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 eachthe redundant separation groups of the MPS Class 1E instrumentation and control circuits, and(ii) between each divisions of the MPS Class 1E instrumentation and control circuits, and (iii) between Class 1E Tier 1 2.5-3 Draft Revision 3

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

• Communications independence exists between Separation Groups A, B, C, and Dredundant separation groups and divisions of the Class 1E MPS.

RAI 14.03-3

• 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.

RAI 14.03-3

• The MPS automatically initiates a reactor trip signal for reactor trip functions listed in Table 2.5-1.

RAI 14.03-3

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

• The MPS automatically actuates a reactor trip.

RAI 14.03-3

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

• The MPS manually actuates a reactor trip.

RAI 14.03-3

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

• 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.

Tier 1 2.5-4 Draft Revision 3

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

• 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 2.1-3predefined safe state for that separation group or division.

• 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.

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

RAI 14.03-3

• 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.function as required when associated conditions are met.

RAI 14.03-3

• 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.function as required when associated conditions are met.

RAI 14.03-3

• 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.The MPS overrides function as required when associated conditions are met.

RAI 14.03-3

• The MPS is capable of performing its safety-related functions when any one of its separation channels is out of serviceplaced in maintenance bypass.

• The MPS operational bypasses are indicated in the MCR.

• 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

• 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 under design conditions.

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

• The MCR isolation switches that isolate the manual MCR switches from MPS in case of a fire in the MCR are located in the remote shutdown station (RSS).

Tier 1 2.5-5 Draft Revision 3

NuScale Tier 1 Module Protection System and Safety Display and Indication System RAI 14.03-3 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 ConceptFunctional quality process composed of System ConceptFunctional Specification Phase satisfies the the following system design Specification Phase. requirements of the System lifecycle phases, with each ConceptFunctional Specification phase having outputs which Phase.

satisfy the requirements of that ii.b. An analysis will be performed of ii.b. The output documentation of the phase. the output documentation of the MPS RequirementsDesign Phase i.a. System Concept Functional System RequirementsDesign satisfies the requirements of the Specification Phase Phase. System RequirementsDesign Phase.

i.b. System RequirementsDesign iii.c.An analysis will be performed of iii.c.The output documentation of the Phase the output documentation of the MPS DesignPrototype Development

• System Prototype Development System DesignPrototype Phase satisfies the requirements of Phase Development Phase. the System DesignPrototype

• Equipment Requirements Development Phase.

Specification Phase ivi.d.An analysis will be performed of ivi.d.The output documentation of the

• Hardware Planning Phase the output documentation of the MPS ImplementationEquipment

• Hardware Requirements Phase System Requirements Specification Phase

• Hardware Design Phase ImplementationEquipment satisfies the requirements of the

• Software Planning Phase Requirements Specification System ImplementationEquipment

• Software Requirements Phase Phase. Requirements Specification Phase.

i.c. SystemSoftware Design Phase i.d. SystemSoftware vi.e.An analysis will be performed of vi.e.The output documentation of the Implementation Phase the output documentation of the MPS TestHardware Planning Phase

• Software Configuration Phase System TestHardware Planning satisfies the requirements of the i.e. System Testing Phase Phase. System TestHardware Planning i.f. System Installation and Phase.

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

vii. An analysis will be performed of vii. The output documentation of the the output documentation of the MPS Hardware Design Phase satisfies Hardware Design Phase. the requirements of the Hardware Design Phase.

viii. An analysis will be performed of viii. The output documentation of the the output documentation of the MPS Software Planning Phase Software Planning Phase. satisfies the requirements of the Software Planning Phase.

ix. An analysis will be performed of ix. The output documentation of the the output documentation of the MPS Software Requirements Phase Software Requirements Phase. satisfies the requirements of the Software Requirements Phase.

x. An analysis will be performed of x. The output documentation of the the output documentation of the MPS Software Design Phase satisfies Software Design Phase. the requirements of the Software Design Phase.

Tier 1 2.5-16 Draft Revision 3

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 xi. An analysis will be performed of xi. The output documentation of the the output documentation of the MPS Software Implementation Phase Software Implementation Phase. satisfies the requirements of the Software Implementation Phase.

xii. An analysis will be performed of xii. The output documentation of the the output documentation of the MPS Software Configuration Phase Software Configuration Phase. satisfies the requirements of the Software Configuration Phase.

xiii. An analysis will be performed of xiii. The output documentation of the the output documentation of the MPS Testing Phase satisfies the System Testing Phase. requirements of the System Testing Phase.

xiv. An analysis will be performed of xiv. The output documentation of the the output documentation of the MPS Installation Phase satisfies the System Installation Phase. requirements of the System Installation 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 Class 1E MPS. provided.

iii.b. Communications iii.b. Communications independence independence exists between between Division I and II of the Class Division 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.

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.

Tier 1 2.5-17 Draft Revision 3

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

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 channels 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.

xv. Two of the four separation xv. An inspection will be performed xv. Separation groups A & C and groups and one of the two of the as-built MPS. Division I of RTS and ESFAS utilize a divisions of RTS and ESFAS will different programmable technology utilize a different from separation groups B & D and programmable technology. Division II of RTS and ESFAS.

2. Protective measures are provided to A test will be performed on the access Protective measures restrict modification restrict modifications to the MPS control features associated with MPS to the MPS tunable parameters without tunable parametersNot used. tunable parametersNot used. proper configuration and authorizationNot used.

Tier 1 2.5-18 Draft Revision 3

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

3. Physical separation exists (i) An inspection will be performed of i. Physical separation between between the redundanteach the MPS Class 1E as-built redundanteach separation groups separation groups of the MPS Class instrumentation and control current- and divisions of the MPS Class 1E 1E instrumentation and control carrying circuits. instrumentation and control current-current-carrying circuits, and(ii) carrying circuits is provided by a between each divisions of the MPS minimum separation distance, or by Class 1E instrumentation and barriers (where the minimum control current-carrying circuits, and separation distances cannot be (iii) between Class 1E maintained), or by a combination of instrumentation and control separation distance and barriers.

current-carrying circuits and non- ii. Physical separation between each Class 1E instrumentation and division of the MPS Class 1E control current-carrying circuits. instrumentation and control current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers.

iii. Physical separation between MPS Class 1E instrumentation and control current-carrying circuits and non-Class 1E instrumentation and control current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers.

4. Electrical isolation exists (i) between An inspection will be performed of i. Class 1E electrical isolation devices the redundanteach separation the MPS Class 1E as-built are installed between redundanteach groups of the MPS Class 1E instrumentation and control circuits. separation groups and divisions of instrumentation and control circuits, the MPS Class 1E instrumentation and and(ii) between each divisions of control circuits.

the MPS Class 1E instrumentation ii. Class 1E electrical isolation devices and control circuits, and (iii) are installed between each division of between Class 1E instrumentation the MPS Class 1E instrumentation and and control circuits and non-Class control circuits.

1E instrumentation and control iii. Class 1E electrical isolation devices circuits to prevent the propagation are installed between MPS Class 1E of credible electrical faults. instrumentation and control circuits and non-Class 1E instrumentation and control circuits.

Tier 1 2.5-19 Draft Revision 3

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. Communications independence A test will be performed of the Class Communications independence between exists between redundant 1E MPSNot used. redundant separation groups and separation groups and divisions of divisions of the Class 1E MPS is the Class 1E MPSNot used. providedNot 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. The MPS automatically initiates a A test will be performed of the A reactor trip signal is automatically reactor trip signalNot used. MPSNot used. initiated for each reactor trip function listed in Table 2.5-1Not used.

9. The MPS automatically initiates an A test will be performed of the An ESF actuation signal is automatically ESF actuation signalNot used. MPSNot used. initiated for each ESF function listed in Table 2.5-2Not used.

10. The MPS automatically actuates a A test will be performed of the The RTBs open upon an injection of a reactor trip.Not used. MPS.Not used. single simulated MPS reactor trip signal.Not used.

11. The MPS automatically actuates the A test will be performed of the MPS. The ESF equipment automatically engineered safety feature 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. The MPS manually actuates a A test will be performed of the The RTBs open when a reactor trip is reactor trip.Not used. MPS.Not used. manually initiated from the main control room.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 2.5-3Table 2.5-2 when manually initiated.

14. The reactor trip logic fails to a safe A test will be performed of the Loss of electrical power in a separation state such that loss of electrical MPS.Not used. group results in a trip state for that power to a MPS separation group separation group.Not used.

results in a trip state for that separation group.Not used.

15. The ESFs logic fails to a safe state A test will be performed of the Loss of electrical power in a separation such that loss of electrical power to MPS.Not used. group results in an actuation state for a MPS separation group results in a that separation group.Not used.

predefined safe state for that separation group.Not used.

Tier 1 2.5-20 Draft Revision 3

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

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 room. 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. The RTBs are installed and arranged An inspection will be performed of The RTBs have the proper connections for in order to successfully accomplish the as-built RTBs, including the the shunt and undervoltage trip the reactor trip function under connections for the shunt and mechanisms and auxiliary contacts, and design conditions.Not used. undervoltage trip mechanism and are arranged as shown in Figure 2.5-2 to auxiliary contacts.Not used. successfully accomplish the reactor trip function.Not used.

28. Two of the four separation groups An inspection will be performed of Separation groups A & C and Division I of and one of the two divisions of RTS the as-built MPS.Not used. RTS and ESFAS utilize a different and ESFAS will utilize a different programmable technology from programmable technology.Not separation groups B & D and Division II of used. RTS and ESFAS.Not used.

29. The MCR isolation switches that An inspection will be performed of The MCR isolation switches are located in isolate the manual MCR switches the location of the as-built MCR the remote shutdown station.

from MPS in case of a fire in the MCR isolation switches.

are located in the remote shutdown station.

Tier 1 2.5-22 Draft Revision 3

NuScale Tier 1 Neutron Monitoring System 2.6 Neutron Monitoring System 2.6.1 Design Description

System Description

The scope of this section is the neutron monitoring system (NMS). The NMS is a safety-related system. Each NuScale Power Module has its own module-specific NMS. The Reactor Building houses all NMS equipment.

The NMS monitors the neutron flux level of the reactor core by detecting neutron leakage from the core. The NMS measures neutron flux as an indication of core power and provides safety-related inputs to the module protection system.

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

• The NMS supports the module protection system by providing neutron flux data for various reactor trips.

Design Commitments

• Electrical isolation exists between the NMS Class 1E circuits and connected non-Class 1E circuits to prevent the propagation of credible electrical faults.

RAI 14.03-3

• Physical separation exists between the redundant divisions of the NMS Class 1E instrumentation and control current-carrying circuits, and 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 between the redundant divisions of the NMS Class 1E instrumentation and control circuits, and as well as between Class 1E instrumentation and control circuits and non-Class 1E instrumentation and control circuits to prevent the propagation of credible electrical faults.

2.6.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.6-1 contains the inspections, tests, and analyses for the NMS.

Tier 1 2.6-1 Draft Revision 3

NuScale Tier 1 Radiation Monitoring Module Specific 2.7 Radiation Monitoring Module Specific 2.7.1 Design Description

System Description

The scope of this section is automatic actions of various systems based on radiation monitoring. Automatic actions of systems based on radiation monitoring are nonsafety-related functions. The components actuated by these automatic radiation monitoring functions are contained in module-specific systems.

Design Commitments RAI 14.03-3

• The containment evacuation system (CES) automatically responds to athe CES high radiation signal from CES-RT-1011listed in Table 2.7-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The chemical and volume control system (CVCS) automatically responds to the CVCS and auxiliary boiler system (ABS)a high radiation signal from CVC-RT-3016signals listed in Table 2.7-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The CVCS automatically responds to a high radiation signal from 6A-AB-RT-0142 to mitigate a release of radioactivity.

RAI 14.03-3

• The CVCS automatically responds to a high radiation signal from 6B-AB-RT-0141 to mitigate a release of radioactivity.

2.7.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 2.7-2 contains the inspections, tests, and analyses for the radiation monitoring - module-specific automatic actions.

Tier 1 2.7-1 Draft Revision 3

NuScale Tier 1 Radiation Monitoring Module Specific RAI 14.03-3 Table 2.7-2: Radiation Monitoring - Module-Specific Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The CES automatically responds to A test will be performed of the CES Upon initiation of a real or simulated athe CES high radiation signal from high radiation signal listed in CES high radiation signal listed in CES-RT-1011listed in Table 2.7-1 to Table 2.7-1. Table 2.7-1, the CES automatically mitigate a release of radioactivity. aligns/actuates the identified components to the positions identified in the table.

2. The CVCS automatically responds to A test will be performed of the CVCS Upon initiation of athe real or the CVCS and ABSa high radiation and ABS high radiation signals listed in simulated CVCS and ABS high signals listed in Table 2.7-1 from Table 2.7-1. radiation signals listed in Table 2.7-1, CVC-RT3016 to mitigate a release of the CVCS automatically aligns/

radioactivity. actuates the identified component(s) to the position identified in the table.

3. The CVCS automatically responds to a A test will be performed of the CVCS Upon initiation of a real or simulated high radiation signal from 6A-AB-RT- high radiation signal. CVCS high radiation signal listed in 0142 to mitigate a release of Table 2.7-1, the CVCS automatically radioactivity. aligns/actuates the identified component to the position identified in the table.

4. The CVCS automatically responds to a A test will be performed of the CVCS Upon initiation of a real or simulated high radiation signal from 6B-AB-RT- high radiation signal. CVCS high radiation signal listed in 0141 to mitigate a release of Table 2.7-1, the CVCS automatically radioactivity. aligns/actuates the identified component to the position identified in the table.

Tier 1 2.7-3 Draft Revision 3

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). 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 features:

RAI 14.03-3, 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 design requirement.

RAI 14.03-3, 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).

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. The scope of equipment for this design commitment is module-specific, Tier 1 2.8-1 Draft Revision 3

NuScale Tier 1 Equipment Qualification Class 1E equipment located within a harsh environment, and module-specific, nonsafety-related equipment with an augmented equipment qualification design requirement located within the boundaries of the NuScale Power Module.

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

• The non-metallic parts, materials, and lubricants used in module-specific mechanical equipment 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. The scope of equipment for this design commitment is module-specific, safety-related mechanical equipment, and module-specific, nonsafety-related mechanical 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.)

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 14.03-3

• The Class 1E digital equipment 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

• The safety-related valves are functionally designed and qualified to perform their safety-related function under the full range of fluid flow, differential pressure, electrical conditions, and temperature 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 14.03-3

• The safety-related decay heat removal system (DHRS) passive condensers 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 located in a harsh environmentlisted 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.

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 3

NuScale Tier 1 Equipment Qualification RAI 08.01-1S1, RAI 14.03-3, 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 seismic qualification record form I equipment listed in Table 2.8-1, combination of type test and exists and concludes that the including its associated supports and analysis will be performed of the module-specific Seismic Category I anchorages, withstands design basis module-specific Seismic Category I equipment listed in Table 2.8-1, seismic loads without loss of its equipment listed in Table 2.8-1, including its associated supports function(s) during and after an SSE. including its associated supports and anchorages, will withstand the The scope of equipment for this and anchorages. design basis seismic loads and design commitment is module- ii. An inspection will be performed of perform its function(s) during and specific, safety-related equipment, and the module-specific Seismic after an SSE.

module-specific, nonsafety-related Category I as-built equipment ii. The module-specific Seismic equipment that has one of the listed in Table 2.8-1, including its Category I equipment listed in following design features: associated supports and Table 2.8-1, including its

• Nonsafety-related mechanical and anchorages. associated supports and electrical equipment located within anchorages, is installed in its the boundaries of the NuScale design location in a Seismic Power Module that has an Category I structure in a augmented Seismic Category I configuration bounded by the design requirement. equipments seismic qualification

• Nonsafety-related mechanical and record form.

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.)

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. The connection assemblies. ii. The module-specific electrical scope of equipment for this design equipment listed in Table 2.8-1, commitment is module-specific, Class including associated connection 1E equipment located within a harsh assemblies, are installed in their environment, and module-specific, design location in a configuration nonsafety-related equipment with an bounded by the EQ record form.

augmented equipment qualification design requirement located within the boundaries of the NuScale Power Module.

Tier 1 2.8-17 Draft Revision 3

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

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 perform their the non-metallic parts, materials, and materials, and lubricants used in function up to the end of their lubricants used in module-specific module-specific mechanical qualified life in the design basis harsh mechanical equipment.Not used. equipment listed in Table 2.8-1 environmental conditions (both perform their function up to the end of internal service conditions and their qualified life under the design external environmental conditions) basis harsh environmental conditions experienced during normal (both internal service conditions and operations, AOOs, DBAs, and post- external environmental conditions) accident conditions. The scope of specified in the qualification record equipment for this design form.Not used.

commitment is module-specific, safety-related mechanical equipment, and module-specific, nonsafety-related mechanical 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.)Not used.

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 A type test, analysis, or a combination An EQ record form exists and performs its safety-related function of type test and analysis will be concludes that the Class 1E digital when subjected to the design basis performed of the Class 1E digital equipment listed in Table 2.8-1 electromagnetic interference, radio equipment.Not used. withstands the design basis frequency interference, and electrical electromagnetic interference, radio surges that would exist before, during, frequency interference, and electrical and following a DBA.Not used. surges that would exist before, during, and following a DBA without loss of safety-related function.Not used.

Tier 1 2.8-18 Draft Revision 3

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

6. The safety-related valves 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 safety-related perform their safety-related function the safety-related valves.Not used. valves listed in Table 2.8-1 are capable under the full range of fluid flow, of performing their safety-related differential pressure, electrical function under the full range of fluid conditions, temperature conditions, flow, differential pressure, electrical and fluid conditions up to and conditions, temperature conditions, including DBA conditions.Not used. and fluid conditions up to and including 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 valves 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 valves 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.

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

required to 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 2.8-19 Draft Revision 3

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

• The air exfiltration out of the control room envelope (CRE) does not exceed 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.

Tier 1 3.1-1 Draft Revision 3

NuScale Tier 1 Control Room Habitability RAI 14.03-3 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 A test will be performed of the CRE. The air exfiltration measured by tracer meetsdoes not exceed the gas testing is less than the CRE air assumptions used to size the CRHS infiltration rate assumed in the dose inventory and the supply flow rate. 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 3.1-3 Draft Revision 3

NuScale Tier 1 Normal Control Room Heating Ventilation and Air Conditioning System 3.2 Normal Control Room Heating Ventilation and Air Conditioning System 3.2.1 Design Description

System Description

The scope of this section is the normal control room HVAC system (CRVS). The CRVS serves the entire Control Building (CRB) and the access tunnel between the CRB and the Reactor Building (RXB). The CRVS is a nonsafety-related system. The CRVS supports up to 12 NuScale Power Modules. The CRB houses all CRVS equipment.

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

• The CRVS supports the CRB by providing isolation of the control room envelope (CRE) from the surrounding areas and outside environment via isolation dampers.

• The CRVS supports the CRB by maintaining the CRB at a positive pressure relative to the RXB and the outside atmosphere to control the ingress of potentially airborne radioactivity from the RXB or the outside atmosphere to the CRB.

• The CRVS supports the highly reliable DC power system by providing ventilation to maintain airborne hydrogen concentrations below the allowable limits.

• The CRVS supports the normal DC power system by providing ventilation to maintain airborne hydrogen concentrations below allowable limits.

Design Commitments RAI 14.03-3

• The CRVS air-operated CRE isolation dampers listed in Table 3.2-1 perform their function to fail to the closed position on loss of motive power under design basis temperature, differential pressure, and flow conditions.

• The CRVS maintains a positive pressure in the CRB relative to the outside environment.

• The CRVS maintains the hydrogen concentration levels in the CRB battery rooms containing batteries below one percent by volume.

3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.2-2 contains the inspections, tests, and analyses for the CRVS.

Tier 1 3.2-1 Draft Revision 3

NuScale Tier 1 Normal Control Room Heating Ventilation and Air Conditioning System RAI 14.03-3 Table 3.2-2: Normal Control Room Heating Ventilation and Air Conditioning Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The CRVS air-operated CRE isolation A test will be performed of the air- Each CRVS air-operated CRE isolation dampers listed in Table 3.2-1 perform operated CRE isolation dampers listed damper listed in Table 3.2-1 performs their function to fail to the closed in Table 3.2-1 under preoperational its function to fail to the closed position on loss of motive power temperature, differential pressure and position on loss of motive power under design basis temperature, flow conditions. under preoperational temperature, differential pressure, and flow differential pressure, and flow conditions. conditions.

2 The CRVS maintains a positive A test will be performed of the CRVS The CRVS maintains a positive pressure in the CRB relative to the while operating in the normal pressure of greater than or equal to outside environment. operating alignment. 1/8 inches water gauge in the CRB relative to the outside environment, while operating in the normal operating alignment.

3 The CRVS maintains the hydrogen A test will be performed of the CRVS The airflow capability of the CRVS concentration levels in the CRB battery while operating in the normal maintains the hydrogen concentration rooms containing batteries below one operating alignment. levels in the CRB battery rooms percent by volume. containing batteries below one percent by volume.

Tier 1 3.2-3 Draft Revision 3

NuScale Tier 1 Reactor Building Heating Ventilation and Air Conditioning System 3.3 Reactor Building Heating Ventilation and Air Conditioning System 3.3.1 Design Description

System Description

The scope of this section is the Reactor Building HVAC system (RBVS). The RBVS is designed to remove radioactive contaminants from the exhaust streams of the Reactor Building (RXB) general area, the Radioactive Waste Building (RWB) general area, and the Annex Building. The RBVS is a nonsafety-related system. The RBVS supports up to 12 NuScale Power Modules. The RXB and the RWB house the RBVS equipment.

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

• The RBVS supports the RXB by maintaining the RXB at a negative pressure relative to the outside atmosphere to control the movement of potentially airborne radioactivity from the RXB to the environment.

• The RBVS supports the RWB by maintaining the RWB at a negative ambient pressure relative to the outside atmosphere to control the movement of potentially airborne radioactivity from the RWB to the environment.

• The RBVS supports the highly reliable DC power system by providing ventilation to maintain airborne hydrogen concentrations below allowable limits.

• The RBVS supports the normal DC power system by providing ventilation to maintain airborne hydrogen concentrations below allowable limits.

Design Commitments

• The RBVS maintains a negative pressure in the RXB relative to the outside environment.

• The RBVS maintains a negative pressure in the RWB relative to the outside environment.

RAI 14.03-3

• The RBVS maintains the hydrogen concentration levels in the RXB battery rooms containing batteries below one percent by volume.

3.3.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.3-1 contains the inspections, tests, and analyses for the RBVS.

Tier 1 3.3-1 Draft Revision 3

NuScale Tier 1 Fuel Handling Equipment System 3.4 Fuel Handling Equipment System 3.4.1 Design Description

System Description

The scope of this section is the fuel handling equipment (FHE) system. The FHE system is designed to support the periodic refueling of the reactor as well as movement of control rods and other radioactive components within the reactor core, refueling pool, and spent fuel pool. The FHE system is a nonsafety-related system. The FHE system supports up to 12 NuScale Power Modules (NPMs). The Reactor Building houses all FHE system equipment.

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

• The FHE system supports the reactor fuel assembly by providing structural support during handling of fuel.

Design Commitments RAI 14.03-3

• The single-failure-proof fuel handling machine (FHM) main and auxiliary hoists are constructed to provide assurance that a failure of a single hoist mechanism component does not result in the uncontrolled movement of the lifted loadThe fuel handling machine (FHM) main and auxiliary hoists are single-failure-proof in accordance with the approved design.

• The FHM main hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.

• The FHM auxiliary hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.

RAI 14.03-3

• Single-failure-proofThe FHM welds are inspectedcomply with the American Society of Mechanical Engineers NOG-1 Code.

• The FHM travel is limited to maintain a water inventory for personnel shielding with the pool level at the lower limit of the normal operating low water level.

RAI 09.01.04-1

• The new fuel jib crane hook movement is limited to prevent carrying a fuel assembly over the fuel storage racks in the spent fuel pool.

3.4.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.4-1 contains the inspections, tests, and analyses for the FHE system.

Tier 1 3.4-1 Draft Revision 3

NuScale Tier 1 Fuel Handling Equipment System RAI 09.01.04-1, RAI 14.03-3 Table 3.4-1: Fuel Handling Equipment System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The single-failure-proof FHM main and An inspection will be performed of the The FHM main and auxiliary hoists are auxiliary hoists are constructed to as-built FHM main and auxiliary hoists. single-failure-proofA report exists and provide assurance that a failure of a concludes that the FHM main and single hoist mechanism component auxiliary hoists are single-failure-proof does not result in the uncontrolled in accordance with the approved movement of the lifted loadThe FHM design.

main and auxiliary hoists are single-failure-proof in accordance with the approved design.

2. The FHM main hoist is capable of A rated load test will be performed of The FHM main hoist lifts, supports, lifting and supporting its rated load, the FHM main hoist. holds with the brakes, and transports a holding the rated load, and load of at least 125 percent of the transporting the rated load. manufacturers rated capacity.

3. The FHM auxiliary hoist is capable of A rated load test will be performed of The FHM auxiliary hoist lifts, supports, lifting and supporting its rated load, the FHM auxiliary hoist. holds with the brakes, and transports a holding the rated load, and load of at least 125 percent of the transporting the rated load. manufacturers rated capacity.

4. Single-failure-proofThe FHM welds are An inspection will be performed of the The results of the non-destructive inspectedcomply with the American as-built FHM welds. examination of the FHM welds comply Society of Mechanical Engineers NOG- with American Society of Mechanical 1 Code. Engineers NOG-1 Code.

5. The FHM travel is limited to maintain a A test will be performed of the FHM The FHM maintains at least 10 feet of water inventory for personnel gripper mast limit switches. water above the top of the fuel shielding with the pool level at the assembly when lifted to its maximum lower limit of the normal operating height with the pool level at the lower low water level. limit of the normal operating low water level.

6. The new fuel jib crane hook A test will be performed of new fuel jib The new fuel jib crane interlocks movement is limited to prevent crane interlocks. prevent the crane from carrying a fuel carrying a fuel assembly over the fuel assembly over the spent fuel racks.

storage racks in the spent fuel pool.

Tier 1 3.4-2 Draft Revision 3

NuScale Tier 1 Fuel Storage System 3.5 Fuel Storage System 3.5.1 Design Description

System Description

The scope of this section is the fuel storage system. The fuel storage system consists of the fuel storage racks in the spent fuel pool (SFP) that can store either spent fuel assemblies or new fuel assemblies. The fuel storage system is a nonsafety-related system. The fuel storage system supports up to 12 NuScale Power Modules (NPMs). The Reactor Building houses all fuel storage system equipment.

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

• The fuel storage system supports the reactor fuel assembly system by providing mechanical support for storage of new and spent fuel in a wet storage location.

• The fuel storage system supports the reactor fuel assembly system by providing neutron absorption to ensure subcriticality during storage of new and spent fuel.

• The fuel storage system supports the control rod assembly system by providing mechanical support for storage of control rods in fuel assemblies.

Design Commitments

• The fuel storage system American Society of Mechanical Engineers (ASME) Code Class NF components conform to the rules of construction of ASME Code Section III.

• The fuel storage racks maintain an effective neutron multiplication factor (k-effective) within the following limits at a 95 percent probability, 95 percent confidence level when loaded with fuel of the maximum reactivity to assure subcriticality during plant life, including normal operations and postulated accident conditions:

RAI 14.03-3 If credit for soluble boron is taken, k-effective must not exceed 0.95 if flooded with borated water, and k-effective must not exceed 1.0 if flooded with unborated water.

RAI 14.03-3 k-effective must not exceed 1.0 if flooded with unborated water 3.5.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.5-1 contains the inspections, tests, and analyses for the fuel storage system.

Tier 1 3.5-1 Draft Revision 3

NuScale Tier 1 Fuel Storage System RAI 14.03-3 Table 3.5-1: Fuel Storage System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The fuel storage system ASME Code An inspection will be performed of the ASME Code Section III Data Reports for Class NF components conform to the fuel storage system ASME Code Class the fuel storage system ASME Code rules of construction of ASME Code NF as-built component Data Reports Class NF fuel storage racks exist and Section III. required by ASME Code Section III. conclude that the requirements of ASME Code Section III are met.

2 The fuel storage racks maintain an An inspection will be performed of the The as-built fuel storage racks, effective neutron multiplication factor as-built fuel storage racks, their including any neutron absorbers, and (k-effective) within the following limits configuration in the SFP, and the their configuration within the SFP at a 95 percent probability, 95 percent associated documentation. conform to the design values for confidence level when loaded with materials and dimensions and their fuel of the maximum reactivity to tolerances, as shown to be acceptable assure subcriticality during plant life, in the approved fuel storage criticality including normal operations and analysis described in the UFSAR.

postulated accident conditions:

• If credit for soluble boron is taken, k-effective must not exceed 0.95 if flooded with borated water, and

• k-effective must not exceed 1.0 if flooded with unborated water.

Tier 1 3.5-2 Draft Revision 3

NuScale Tier 1 Ultimate Heat Sink RAI 14.03-3

• The UHS Code Class 3 components listed in Table 3.6-1 conform to the rules of construction of ASME Code Section III.

RAI 14.03-3

• The spent fuel pool, refueling pool, reactor pool, and dry dock piping and connections are located to prevent the drain down of the SFP and reactor pool water level below the minimum safety water level.

3.6.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.6-2 contains the inspections, tests, and analyses for the UHS.

Tier 1 3.6-2 Draft Revision 3

NuScale Tier 1 Ultimate Heat Sink RAI 14.03-3 Table 3.6-2: Ultimate Heat Sink Piping System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The ultimate heat sink ASME Code An inspection will be performed of the The ASME Code Section III Design Class 3 piping system listed in ultimate heat sink ASME Code Class 3 Report (NCA-3550) exists and Table 3.6-1 complies with ASME Code as-built piping system listed in concludes that the ultimate heat sink Section III requirements. Table 3.6-1 Design Report required by ASME Code Class 3 as-built piping ASME Code Section III. system listed in Table 3.6-1 meets the requirements of ASME Code Section III.

2 The UHS Code Class 3 components An inspection will be performed of the The ASME Code Section III Data Report listed in Table 3.6-1 conform to the UHS ASME Code Class 3 as-built for the UHS ASME Code Class 3 rules of construction of ASME Code component Data Report for the components listed in Table 3.6-1 and Section III. components listed in Table 3.6-1 interconnecting piping exists and required by ASME Code Section III. concludes that the requirements of ASME Code Section III are met.

23 The spent fuel pool, refueling pool, An inspection will be performed of the There are no gates, openings, drains, reactor pool, and dry dock piping and as-built SFP, RFP, reactor pool and dry or piping within the SFP, RFP, reactor connections are located to prevent the dock piping and connections. pool, and dry dock that are below 80 ft drain down of the SFP and reactor building elevation (55 ft pool level) as pool water level below the minimum measured from the bottom of the SFP safety water level. and reactor pool.

Tier 1 3.6-4 Draft Revision 3

NuScale Tier 1 Fire Protection System 3.7 Fire Protection System 3.7.1 Design Description

System Description

The scope of this section is the fire protection system (FPS). The FPS is comprised of the equipment and components that provide early fire detection and suppression to limit the spread of fires. The FPS is a nonsafety-related system that supports up to 12 NuScale Power Modules (NPMs). The FPS equipment is located throughout the plant site.

The FPS includes the following equipment:

• fire water storage tanks, motor and diesel driven fire pumps, jockey pump, distribution piping, valves, and fire hydrants

• automatic fire detection, fire alarm notification, and fire suppression systems, including fire water supply and distribution systems

• manual firefighting capability, including portable fire extinguishers, standpipes, hydrants, hose stations, and fire department connections The FPS performs the following nonsafety-related system functions that are verified by Inspections, Tests, Analyses, and Acceptance Criteria:

• The FPS supports the Reactor Building by providing fire prevention, detection, and suppression.

• The FPS supports the Radioactive Waste Building by providing fire prevention, detection, and suppression.

• The FPS supports the Control Building by providing fire prevention, detection, and suppression.

Design Commitments

• Two separate firewater storage tanks provide a dedicated volume of water for firefighting.

RAI 14.03-3

• The FPS has a sufficient number of fire pumps to satisfy the flow demand for any FPS connected to the pumpsprovide the flow demand for the largest sprinkler or deluge system plus an additional 500 gpm for fire hoses assuming failure of the largest fire pump or loss of off-site power.

• Safe-shutdown can be achieved assuming that all equipment in any one fire area (except for the main control room (MCR) and under the bioshield) is rendered inoperable by fire damage and that reentry into the fire area for repairs and operator actions is not possible. An alternative shutdown capability that is physically and electrically independent of the MCR exists. Additionally, smoke, hot gases, or fire suppressant cannot migrate from the affected fire area into other fire areas to the extent that they could adversely affect safe-shutdown capabilities, including operator actions.

Tier 1 3.7-1 Draft Revision 3

NuScale Tier 1 Fire Protection System RAI 09.05.01-6, RAI 14.03-3 Table 3.7-1: Fire Protection System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 Two separate firewater storage tanks An inspection will be performed of the Each firewater storage tank provides a provide a dedicated volume of water as-built firewater storage tanks. usable water volume dedicated for for firefighting. firefighting that is greater than or equal to 300,000 gallons.

2 The FPS has a sufficient number of fire i. An analysis will be performed of i. A report exists and concludes that pumps to provide the design flow the as-built fire pumps. the fire pumps can provide the requirements to satisfy the flow ii. A test will be performed of the fire flow demand for the largest demand for the largest sprinkler or pumps. sprinkler or deluge system plus an deluge system plus an additional additional 500 gpm for fire hoses 500 gpm for fire hoses assuming assuming failure of the largest fire failure of the largest fire pump or loss pump or loss of off-site power.

of off-site power. ii. Each fire pump delivers the design flow to the FPS, while operating in the fire-fighting alignment.

3 Safe-shutdown can be achieved A safe-shutdown analysis of the as- A safe-shutdown analysis report exists assuming that all equipment in any built plant will be performed, and concludes that:

one fire area (except for the MCR and including a post-fire safe-shutdown

• Safe-shutdown can be achieved under the bioshield) is rendered circuit analysis. assuming that all equipment in any inoperable by fire damage and that one fire area (except for the MCR and reentry into the fire area for repairs under the bioshield) is rendered and operator actions is not possible. inoperable by fire and that reentry An alternative shutdown capability into the fire area for repairs and that is physically and electrically operator actions is not possible independent of the MCR exists.

• Smoke, hot gases, or fire suppressant Additionally, smoke, hot gases, or fire cannot migrate from the affected suppressant cannot migrate from the fire area into other fire areas to the affected fire area into other fire areas extent that they could adversely to the extent that they could adversely affect safe-shutdown capabilities, affect safe-shutdown capabilities, including operator actions.

including operator actions.

• An independent alternative shutdown capability that isMPS equipment rooms within the Reactor Building used as the alternative shutdown capability are physically and electrically independent of the MCR exists.

4 A plant FHA considers potential fire A FHA of the as-built plant will be A FHA report exists and concludes hazards and ensures the fire performed. that:

protection features in each fire area

• Combustible loads and ignition are suitable for the hazards. sources are accounted for, and

• Fire protection features are suitable for the hazards they are intended to protect against.

Tier 1 3.7-3 Draft Revision 3

NuScale Tier 1 Plant Lighting System 3.8 Plant Lighting System 3.8.1 Design Description

System Description

The scope of this section is the plant lighting system (PLS). The PLS is a nonsafety-related system and supports up to 12 NuScale Power Modules (NPMs). The PLS provides artificial illumination for the entire plant: buildings (interior and exterior), rooms, spaces, and all outdoor areas of the plant. The PLS consists of normal and emergency lighting and includes miscellaneous non-lighting loads as required.

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

RAI 14.03-3

• The PLS supports the Reactor Building (RXB) by providing normal lighting.

RAI 14.03-3

• The PLS supports the RXBReactor Building by providing emergency lighting.

RAI 14.03-3

• The PLS supports the RXB by providing emergency lighting for the remote shutdown station (RSS).

• The PLS supports the Control Building by providing normal lighting.

• The PLS supports the Control Building by providing emergency lighting in the main control room (MCR).

Design Commitments RAI 14.03-3

• The PLS provides normal illumination of the operator workstations and auxiliary panels in the MCR and the operator workstations in the RSS.

RAI 14.03-3

• The PLS provides emergency illumination of the operator workstations and auxiliary panels in the MCR and the operator workstations in the RSS.

RAI 14.03-3

• Eight-hour battery-pack emergency lighting fixtures provide illumination for post-fire safe shutdown (FSSD) activities performed by operators outside the MCR and remote shutdown station (RSS) where post-FSSD activities are performed.

3.8.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.8-1 contains the inspections, tests, and analyses for the PLS.

Tier 1 3.8-1 Draft Revision 3

NuScale Tier 1 Plant Lighting System RAI 14.03-3 Table 3.8-1: Plant Lighting System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The PLS provides normal illumination i.A test will be performed of the MCR i.The PLS provides at least 100 foot-of the operator workstations and operator workstations and auxiliary candles illumination at the MCR auxiliary panels in the MCR and panel illumination. operator workstations and at least 50 operator workstations in the RSS. ii.A test will be performed of the RSS foot-candles at the auxiliary panels.

operator workstations illumination. ii.The PLS provides at least 100 foot-candles illumination at the RSS operator workstations.

2 The PLS provides emergency i.A test will be performed of the MCR i.The PLS provides at least 10 foot-illumination of the operator operator workstations and auxiliary candles of illumination at the MCR workstations and auxiliary panels in panel illumination. operator workstations and auxiliary the MCR and operator workstations in ii.A test will be performed of the RSS panels when it is the only MCR lighting the RSS. operator workstations illumination. system in operation.

ii.The PLS provides at least 10 foot-candles at the RSS operator workstations when it is the only RSS lighting system in operation.

3 Eight-hour battery-pack emergency A test will be performed of the eight- Eight-hour battery-pack emergency lighting fixtures provide illumination hour battery-pack emergency lighting lighting fixtures illuminate their for post-FSSD activities performed by fixtures. required target areas to provide at operators outside the MCR and RSS least one foot-candle illumination in where post-FSSD activities are the areas outside the MCR or RSS performed. where post-FSSD activities are performed.

Tier 1 3.8-2 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 3.9 Radiation Monitoring - NuScale Power Modules 1 - 12 3.9.1 Design Description

System Description

The scope of this section is automatic actions of various systems based on radiation monitoring (RM). Automatic actions of systems based on RM are nonsafety-related functions. The systems actuated by these automatic RM functions are shared by NuScale Power Modules (NPMs) 1-12.

Design Commitments RAI 14.03-3

• The normal control room HVAC system (CRVS) automatically responds to athe CRVS high-radiation signals from 00-CRV-RT-0503, 00-CRV-RT-0504, and 00-CRV-RT-0505upstream of the CRVS filter unit listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The CRVS and the control room habitability system (CRHS) automatically respond to athe CRVS high-radiation signals from 00-CRV-RT-0510 and 00-CRV-RT-0511downstream of the CRVS filter unit listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The Reactor Building HVAC system (RBVS) automatically responds to athe RBVS high-radiation signals from 00-RBV-RE-0510, 00-RBV-RE-0511, and 00-RBV-RE-0512listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The gaseous radioactive waste system (GRWS) automatically responds to athe GRWS high-radiation signals from 00-GRW-RIT-0046listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The GRWS automatically responds to a high-radiation signal from 00-GRW-RIT-0060 to mitigate a release of radioactivity.

RAI 14.03-3

• The GRWS automatically responds to a high-radiation signal from 00-GRW-RIT-0071 to mitigate a release of radioactivity.

RAI 14.03-3

• The liquid radioactive waste system (LRWS) automatically responds to athe LRWS high-radiation signals from 00-LRW-RIT-0569 and 00-LRW-RIT-0571listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The auxiliary boiler system (ABS) automatically responds to athe ABS high-radiation signals from 00-AB-RT-0153listed in Table 3.9-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The ABS automatically responds to a high-radiation signal from 00-AB-RT-0166 to mitigate a release of radioactivity.

Tier 1 3.9-1 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 RAI 14.03-3

• The pool surge control system (PSCS) automatically responds to athe PSCS high-radiation signal from 00-PSC-RE-1003listed in Table 3.9-1 to mitigate a release of radioactivity.

3.9.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.9-2 contains the inspections, tests, and analyses for radiation monitoring NPMs 1-12.

Tier 1 3.9-2 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 RAI 14.03-3 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 athe real or athe CRVS high-radiation signals from high-radiation signals listed in simulated CRVS high-radiation signals 00-CRV-RT-0503, 00-CRV-RT-0504, and Table 3.9-1. upstream of the CRVS filter unit listed 00-CRV-RT-0505upstream of the CRVS in Table 3.9-1, the CRVS automatically filter unit listed in Table 3.9-1 to aligns/actuates the identified mitigate a release of radioactivity. 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 athe real or respond to athe high-radiation signals high-radiation signals listed in simulated CRVS high-radiation signals from 00-CRV-RT-0510 and 00-CRV-RT- Table 3.9-1. downstream of the CRVS filter unit 0511downstream of the CRVS filter listed in Table 3.9-1, the CRVS and the unit listed in Table 3.9-1 to mitigate a CRHS automatically align/actuate the release of radioactivity. 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 athe real or athe RBVS high-radiation signals from high-radiation signals listed in simulated RBVS high-radiation signals 00-RBV-RE-0510, 00-RBV-RE-0511, and Table 3.9-1. listed in Table 3.9-1, the RBVS 00-RBV-RE-0512listed in Table 3.9-1 to automatically aligns/actuates the mitigate a release of radioactivity. 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 athe real or athe GRWS high-radiation signals from high-radiation signals listed in simulated GRWS high-radiation signals 00-GRW-RIT-0046listed in Table 3.9-1 Table 3.9-1. listed in Table 3.9-1, the GRWS to mitigate a release of radioactivity. automatically aligns/actuates the identified components to the positions identified in the table.

5 The GRWS automatically responds to a A test will be performed of the GRWS Upon initiation of a real or simulated high-radiation signal from 00-GRW- high-radiation signalsNot Used. GRWS high-radiation signals listed in RIT-0060 to mitigate a release of Table 3.9-1, the GRWS automatically radioactivityNot Used. aligns/actuates the identified components to the positions identified in the tableNot Used.

6 The GRWS automatically responds to a A test will be performed of the GRWS Upon initiation of a real or simulated high-radiation signal from 00-GRW- high-radiation signalsNot Used. GRWS high-radiation signals listed in RIT-0071 to mitigate a release of Table 3.9-1, the GRWS automatically radioactivityNot Used. aligns/actuates the identified components to the positions identified in the tableNot Used.

7 The LRWS automatically responds to A test will be performed of the LRWS Upon initiation of athe real or athe LRWS high-radiation signals from high-radiation signals listed in simulated LRWS high-radiation signals 00-LRW-RIT-0569 and 00-LRW-RIT- Table 3.9-1. listed in Table 3.9-1, the LRWS 0571listed in Table 3.9-1 to mitigate a automatically aligns/actuates the release of radioactivity. identified components to the positions identified in the table.

8 The ABS automatically responds to A test will be performed of the ABS Upon initiation of athe real or athe ABS high-radiation signals from high-radiation signals listed in simulated ABS high-radiation signals 00-AB-RT-0153listed in Table 3.9-1 to Table 3.9-1. listed in Table 3.9-1, the ABS mitigate a release of radioactivity. automatically aligns/actuates the identified components to the positions identified in the table.

Tier 1 3.9-6 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 Table 3.9-2: Radiation Monitoring - NuScale Power Modules 1-12 Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 9 The ABS automatically responds to a A test will be performed of the ABS Upon initiation of a real or simulated high-radiation signal from 00-AB-RT- high-radiation signalNot Used. ABS high-radiation signal listed in 0166 to mitigate a release of Table 3.9-1, the ABS automatically radioactivityNot Used. aligns/actuates the identified components to the positions identified in the tableNot Used.

10 The PSCS automatically responds to A test will be performed of the PSCS Upon initiation of a real or simulated athe PSCS high-radiation signal from high-radiation signal listed in PSCS high-radiation signal listed in 00-PSC-RE-1003listed in Table 3.9-1 to Table 3.9-1. Table 3.9-1, the PSCS automatically mitigate a release of radioactivity. aligns/actuates the identified components to the positions identified in the table.

Tier 1 3.9-7 Draft Revision 3

NuScale Tier 1 Reactor Building Crane 3.10 Reactor Building Crane 3.10.1 Design Description

System Description

The scope of this section is the Reactor Building crane (RBC). The RBC is a bridge crane that rides on rails anchored to the Reactor Building. The bridge crane can travel the length of the reactor pool, refueling pool, and the dry dock. The RBC is nonsafety-related and supports up to 12 NuScale Power Modules (NPMs). The Reactor Building houses all RBC equipment.

The RBC includes the following:

• RBC with auxiliary hoist RAI 14.03.07-1

• below-the-hook lifting devices, including the module lifting adapter (MLA) and the wet hoist The RBC performs the following risk-significant system function that is verified by Inspections, Tests, Analyses, and Acceptance Criteria:

• The RBC supports the NuScale Power Module by providing structural support and mobility while moving from refueling, inspection and operating bay.

Design Commitments RAI 14.03-3

• The single-failure-proof RBC main hoist is constructed to provide assurance that a failure of a single hoist mechanism does not result in the uncontrolled movement of the lifted loadThe RBC main hoist is single-failure-proof in accordance with the approved design.

RAI 14.03-3

• The single-failure-proof RBC auxiliary hoists are constructed to provide assurance that a failure of a single hoist mechanism does not result in the uncontrolled movement of the lifted loadThe RBC auxiliary hoists are single-failure-proof in accordance with the approved design.

RAI 14.03-3

• The single-failure-proof RBC wet hoist is constructed to provide assurance that a failure of a single hoist mechanism does not result in the uncontrolled movement of the lifted loadThe RBC wet hoist is single-failure-proof in accordance with the approved design.

• The RBC main hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.

• The RBC auxiliary hoists are capable of lifting and supporting their rated load, holding the rated load, and transporting the rated load.

• The RBC wet hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.

Tier 1 3.10-1 Draft Revision 3

NuScale Tier 1 Reactor Building Crane RAI 14.03-3

• Load path RBC welds are inspectedAll RBC weld joints whose failure could result in the drop of a critical load comply with the American Society of Mechanical Engineers NOG-1 Code.

RAI 14.03-3

• Load path RBC wet hoist welds are inspected.

RAI 14.03.07-1

• The MLA is capable of supporting its rated load.

RAI 14.03-3, RAI 14.03.07-1

• The MLA is single-failure-proof in accordance with the approved design.

3.10.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.10-1 contains the inspections, tests, and analyses for the RBC.

Tier 1 3.10-2 Draft Revision 3

NuScale Tier 1 Reactor Building Crane RAI 14.03-3, RAI 14.03.07-1 Table 3.10-1: Reactor Building Crane Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 The single-failure-proof RBC main An inspection will be performed of the The RBC main hoist is single-failure-hoist is constructed to provide as-built RBC main hoist. proofA report exists and concludes assurance that a failure of a single that the RBC main hoist is single-hoist mechanism does not result in the failure-proof in accordance with the uncontrolled movement of the lifted approved design.

loadThe RBC main hoist is single-failure-proof in accordance with the approved design.

2 The single-failure-proof RBC auxiliary An inspection will be performed of the The RBC auxiliary hoists are single-hoists are constructed to provide as-built RBC auxiliary hoists. failure-proofA report exists and assurance that a failure of a single concludes that the RBC auxiliary hoists hoist mechanism does not result in the are single-failure-proof in accordance uncontrolled movement of the lifted with the approved design.

loadThe RBC auxiliary hoists are single-failure-proof in accordance with the approved design.

3 The single-failure-proof RBC wet hoist An inspection will be performed of the The RBC wet hoist is single-failure-is constructed to provide assurance as-built RBC wet hoist. proofA report exists and concludes that a failure of a single hoist that the RBC wet hoist is single-failure-mechanism does not result in the proof in accordance with the uncontrolled movement of the lifted approved design.

loadThe RBC wet hoist is single-failure-proof in accordance with the approved design.

4 The RBC main hoist is capable of lifting A rated load test will be performed of The RBC main hoist lifts, supports, and supporting its rated load, holding the RBC main hoist. holds with the brakes, and transports a the rated load, and transporting the load of at least 125 to 130 percent of rated load. the manufacturers rated capacity.

5 The RBC auxiliary hoists are capable of A rated load test will be performed of The RBC auxiliary hoists lift, support, lifting and supporting their rated load, the RBC auxiliary hoists. hold with the brakes, and transport a holding the rated load, and load of at least 125 to 130 percent of transporting the rated load. the manufacturers rated capacity.

6 The RBC wet hoist is capable of lifting A rated load test will be performed of The RBC wet hoist lifts, supports, holds and supporting its rated load, holding the RBC wet hoist. with the brakes, and transports a load the rated load, and transporting the of at least 125 to 130 percent of the rated load. manufacturers rated capacity.

7 Load path RBC welds are inspectedAll An inspection will be performed of the The results of the non-destructive RBC weld joints whose failure could as-built RBC weld joints whose failure examination of the RBC welds joints result in the drop of a critical load could result in the drop of a critical whose failure could result in the drop comply with the American Society of load. of a critical load comply with American Mechanical Engineers NOG-1 Code. Society of Mechanical Engineers NOG-1 Code.

8 Load path RBC wet hoist welds are An inspection will be performed of the The results of the non-destructive inspectedNot Used. as-built RBC wet hoistNot Used. examination of the RBC wet hoist welds comply with American Society of Mechanical Engineers NOG-1 CodeNot Used.

Tier 1 3.10-3 Draft Revision 3

NuScale Tier 1 Reactor Building Crane Table 3.10-1: Reactor Building Crane Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 9 The MLA is capable of supporting its i. A rated load test will be performed i. The MLA single load path rated load. of the MLA single load path elements support a load of at least elements. 300 percent of the manufacturer's rated capacity.

ii. A rated load test will be performed of the MLA dual load path ii. The MLA dual load path elements elements. support a load of at least 150 percent of the manufacturer's rated capacity.

10 The MLA is single-failure-proof in An inspection will be performed of the A report exists and concludes that the accordance with the approved design. as-built MLA. MLA is single-failure-proof in accordance with the approved design.

Tier 1 3.10-4 Draft Revision 3

NuScale Tier 1 Reactor Building RAI 14.03-3

• Non-Seismic Category I SSC located where there is a potential for adverse interaction with the RXB or a Seismic Category I SSC exists in the RXB will not impair the ability of Seismic Category I SSC to perform their safety functions during or following a safe shutdown earthquake (SSE).

RAI 14.03.03-1

• Safety-related SSC are protected against the dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.

3.11.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.11-2 contains the inspections, tests, and analyses for the RXB.

Tier 1 3.11-2 Draft Revision 3

NuScale Tier 1 Reactor Building RAI 14.03-3, RAI 14.03.02-3, RAI 14.03.03-1, RAI 14.03.03-11S1 Table 3.11-2: Reactor Building Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 Fire and smoke barriers provide An inspection will be performed of the The following RXB fire and smoke confinement so that the impact from RXB as-built fire and smoke barriers. barriers exist in accordance with the internal fires, smoke, hot gases,or fire fire hazards analysis, and have been suppressants is contained within the qualified for the fire rating specified in RXB fire area of origin. the fire hazards analysis:

• fire-rated doors

• fire-rated penetration seals

• fire-rated dampers

• fire-rated walls, floors, and ceilings

• smoke barriers 2 Internal flooding barriers provide An inspection will be performed of the The following RXB internal flooding confinement so that the impact from RXB as-built internal flooding barriers. barriers exist in accordance with the internal flooding is contained within internal flooding analysis report and the RXB flooding area of origin. have been qualified as specified in the internal flooding analysis report:

• flood resistant doors

• curbs and sills

• walls

• water tight penetration seals

• National Electrical Manufacturer's Association enclosures 3 The Seismic Category I RXB is An inspection will be performed of the The RXB floor elevation at ground protected against external flooding in RXB as-built floor elevation at ground entrances is higher than the maximum order to prevent flooding of safety- entrances. external flood elevation.

related SSC within the structure.

4 The RXB includes radiation shielding An inspection will be performed of the The thickness of RXB radiation barriers for normal operation and as-built RXB radiation shielding shielding barriers is greater than or post-accident radiation shielding. barriers. equal to the required thickness specified in Table 3.11-1.

5 The RXB includes radiation An inspection will be performed of the The RXB radiation attenuating doors attenuating doors for normal as-built RXB 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 in accordance with the within which they are installed. approved door schedule design.

6 The RXB is Seismic Category I and i. An inspection and analysis will be i. A design report exists and maintains its structural integrity under performed of the as-built RXB. concludes that the deviations the design basis loads. ii. An inspection will be performed of between the drawings used for the as- built RXB. construction and the as-built RXB have been reconciled, and the RXB maintains its structural integrity under the design basis loads and that all demand to capacity ratios are less than 1.0 (i.e. D/C < 1.0).

ii. The dimensions of the RXB critical sections conform to the approved design.

Tier 1 3.11-7 Draft Revision 3

NuScale Tier 1 Reactor Building Table 3.11-2: Reactor Building Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 7 Non-Seismic Category I SSC located An inspection and analysis will be A report exists and concludes that the where there is a potential for adverse performed of the as-built non-Seismic Non-Seismic Category I SSC located interaction with the RXB or a Seismic Category I SSC located where there is a where there is a potential for adverse Category I SSC exists in the RXB will potential for adverse interaction with interaction with the RXB or a Seismic not impair the ability of Seismic the RXB or a Seismic Category I SSC in Category I SSC exists in the RXB will Category I SSC to perform their safety the RXB. not impair the ability of Seismic functions during or following a SSE. Category I SSC to perform their safety functions during or following an SSE as demonstrated by one or more of the following criteria:

• Seismic Category I SSC are isolated from non-Seismic Category I SSC, so that interaction does not occur.

• Seismic Category I SSC are analyzed to confirm that the ability to perform their safety functions is not impaired as a result of impact from non-Seismic Category I SSC.

• A non-Seismic Category I restraint system designed to Seismic Category I requirements is used to assure that no interaction occurs between Seismic Category I SSC and non-Seismic Category I SSC.

8 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- safety-related SSC are protected moderate-energy piping systems. related SSC located in the RXB outside against or qualified to withstand the the Reactor Pool Bay. dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.

Tier 1 3.11-8 Draft Revision 3

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

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 in accordance with the within which they are installed. approved door schedule design.

3 The RWB is an RW-IIa structure and An inspection and analysis will be A design report exists and concludes maintains its structural integrity under performed of the as-built RW-IIa RWB. that the deviations between the the design basis loads. 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 3.12-4 Draft Revision 3

NuScale Tier 1 Control Building RAI 14.03-3, RAI 14.03.02-3 Table 3.13-1: Control Building Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 Fire and smoke barriers provide An inspection will be performed of the The following CRB fire and smoke confinement so that the impact from CRB as-built fire and smoke barriers. barriers exist in accordance with the internal fires, smoke, hot gases, or fire fire hazards analysis, and have been suppressants is contained within the qualified for the fire rating specified in CRB fire area of origin. the fire hazards analysis:

• fire-rated doors

• fire-rated penetration seals

• fire-rated dampers

• fire-rated walls, floors, and ceilings

• smoke barriers 2 Internal flooding barriers provide An inspection will be performed of the The following CRB internal flooding confinement so that the impact from CRB as-built internal flooding barriers. barriers exist in accordance with the internal flooding is contained within internal flooding analysis report and the CRB flooding area of origin. have been qualified as specified in the internal flooding analysis report:

• flood resistant doors

• walls

• water tight penetration seals

• National Electrical Manufacturer's Association (NEMA) enclosures 3 The Seismic Category I CRB is An inspection will be performed of the The CRB floor elevation at ground protected against external flooding in CRB as-built floor elevation at ground entrances is higher than the maximum order to prevent flooding of safety- entrances. external flood elevation.

related SSC within the structure.

4 The CRB at Elevation 120-0 and i. An inspection and analysis will be i. A design summary report exists below (except for the elevator shaft, performed of the as-built CRB. and concludes that the deviations the stairwells, and the fire protection ii. An inspection will be performed of between the drawings used for vestibule which are Seismic Category the as-built CRB at Elevation 120- construction and the as-built CRB II) and below is Seismic Category I and 0 and below. have been reconciled, and the CRB maintains its structural integrity under at Elevation 120-0 and below the design basis loads. (except for the elevator shaft, the stairwells, and the fire protection vestibule) maintains its structural integrity under the design basis loads and that all demand to capacity ratios are less than 1.0 (i.e.

D/C < 1.0).

ii. The dimensions of the CRB critical sections conform to the approved design.

Tier 1 3.13-3 Draft Revision 3

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

System Description

RAI 14.03-3, RAI 14.03.03-3S1, RAI 14.03.03-4S1, RAI 14.03.03-6, RAI 14.03.03-6S1, RAI 14.03.03-7, RAI 14.03.03-7S1 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 (the only common, safety-related equipment), 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 a path for makeup water to the ultimate heat sink (UHS).

• 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. 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 functions:

RAI 14.03-3, 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, RAI 14.03.03-6, RAI 14.03.03-7 Provides a path for makeup water to the UHS RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7 Provides containment of UHS water RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7 Monitors UHS water level RAI 14.03-3, RAI 14.03.03-6, RAI 14.03.03-7 Tier 1 3.14-1 Draft Revision 3

NuScale Tier 1 Equipment Qualification - Shared Equipment

• 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. The scope of equipment for this design commitment is the nonsafety-related equipment that provides monitoring of the UHS water level and the non-safety related electrical equipment on the fuel handling machine and reactor building crane used to physically support irradiated fuel.

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

• 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

• 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

• 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

• 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 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 Table 3.14-2 contains the inspections, tests, and analyses for EQ -- shared equipment.

Tier 1 3.14-2 Draft Revision 3

NuScale Tier 1 Equipment Qualification - Shared Equipment RAI 14.03-3, 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 seismic qualification record form equipment listed in Table 3.14-1, combination of type test and exists and concludes that the including its associated supports and analysis will be performed of the common Seismic Category I anchorages, withstands design basis common Seismic Category I equipment listed in Table 3.14-1, seismic loads without loss of its equipment listed in Table 3.14-1, including its associated supports function(s) during and after a safe including its associated supports and anchorages, will withstand the shutdown earthquake. The scope of and anchorages. design basis seismic loads and equipment for this design perform its function during and commitment is common, safety- after a safe shutdown earthquake.

related equipment, and common, ii. An inspection will be performed of ii. The common Seismic Category I nonsafety-related equipment that the common Seismic Category I as- equipment listed in Table 3.14-1, provides one of the following built equipment listed in including its associated supports nonsafety-related functions: Table 3.14-1, including its and anchorages, is installed in its

• Provides physical support of associated supports and design location in a Seismic irradiated fuel (fuel handling anchorages. Category I structure in a machine, spent fuel storage racks, configuration bounded by the reactor building crane, and module equipments seismic qualification lifting adaptor) record form.

• Provides a path for makeup water to the UHS

• Provides containment of UHS water

• Monitors UHS water level

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.

The scope of equipment for this ii. An inspection will be performed of ii. The common electrical equipment design commitment is the common as-built electrical listed in Table 3.14-1, including its nonsafety-related equipment that equipment listed in Table 3.14-1, connection assemblies, is installed provides monitoring of the UHS water including its connection in its design location in a level and the non-safety related assemblies. configuration bounded by the EQ electrical equipment on the fuel record form.

handling machine and reactor building crane used to physically support irradiated fuel.

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 3.14-6 Draft Revision 3

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 3.14-7 Draft Revision 3

NuScale Tier 1 Human Factors Engineering 3.15 Human Factors Engineering 3.15.1 Design Description

System Description

The human factors engineering (HFE) program design process is employed to design the control rooms and the human-system interfaces (HSIs) and associated equipment while relating the high-level goal of plant safety into individual, discrete focus areas for the design.

The HFE and control room design team establish design guidelines, define program-specific design processes, and verify that the guidelines and processes are followed. The scope of the HFE program includes the following:

• location and accessibility requirements for the control rooms and other control stations

• layout requirements of the control rooms, including requirements regarding the locations and design of individual displays and panels

• basic concepts and detailed design requirements for the information displays, controls, and alarms for HSI control stations

• coding and labeling conventions for control room components and plant displays

• HFE design requirements and guidelines for the screen-based HSI, including the actual screen layout and the standard dialogues for accessing information and controls

• requirements for the physical environment of the control rooms (e.g., lighting, acoustics, heating, ventilation, and air conditioning)

• HFE requirements and guidelines regarding the layout of operator workstations and work spaces

• corporate policies and procedures regarding the verification and validation of the design of HSI RAI 14.03-3, RAI 18-43 The HFE program applies to the design of the main control room (MCR) and the remote shutdown station. The HSI of the technical support center, the emergency operations facility, and local control stations (LCS) are derivatives of the main control room (MCR) HSI.

The design of local control stationLCS is accomplished concurrently with the applicable system design and follows guidelines established by the HFE and control room design team.

Design Commitments RAI 14.03-3

• The MCR design incorporates HFE principles that reduce the potential for operator error.

RAI 18-46S1

• The as-builtconfiguration of the MCR HSI is consistent with the final design specificationsverified and validated by the integrated system validation testas reconciled by the Design Implementation Implementation Plan.

Tier 1 3.15-1 Draft Revision 3

NuScale Tier 1 Physical Security System 3.16 Physical Security System 3.16.1 Design Description

System Description

The NuScale Power Plant physical security system design provides the capabilities to detect, assess, impede and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment.

Design Commitments RAI 14.03-3

• Vital equipment within the Reactor Building (RXB) and Control Building (CRB) will be located only within a vital area.

RAI 14.03-3

• Access to vital equipment within the RXB and CRB will requires passage through at least two physical barriers.

RAI 14.03-3

• The external walls, doors, ceilings, and floors in the main control room (MCR), and central alarm station (CAS), and the last access control function for access to the protected area will be bullet-resistant.

RAI 14.03-3

• An access control system will be installed and designed for use by individuals who are authorized access to vital areas within the RXB and CRBnuclear island and structures without escort.

RAI 14.03-3

• Unoccupied vital areas within the RXB and CRBnuclear island and structures will be designed with locking devices and intrusion-detection devices that annunciate in the CAS.

RAI 14.03-3

• The CAS will be located inside the protected area and will be designed so that the interiors is not visible from the perimeter of the protected area.

RAI 14.03-3

• Security alarm devices in the Reactor Building (RXB) and Control Building (CRB),

including transmission lines to annunciators, will be tamper-indicating and self-checking, and alarm annunciation indicates the type of alarm and its location.

RAI 14.03-3

• Intrusion- detection and assessment systems forin the RXB and CRB will be designed to provide visual display and audible annunciation of alarms in the CAS.

RAI 14.03-3

• Intrusion detection systems' recording equipment will record onsite security alarm annunciations with the nuclear island and structures, including each alarm, false alarm, alarm check, and tamper indication and the type of alarm, location, alarm circuit, date, and time.

Tier 1 3.16-1 Draft Revision 3

NuScale Tier 1 Physical Security System RAI 14.03-3

• Emergency exits inthrough the vital area boundaries within the RXB and CRBnuclear island and structures will be alarmed with intrusion- detection devices and are secured by locking devices that allow prompt egress during an emergency.

• The CAS will have landline telephone service with the control room and local law enforcement authorities.

• The CAS will be capable of continuous communication with on-duty security force personnel.

• Non-portable communications equipment in the CAS will remain operable from an independent power source in the event of the loss of normal power.

3.16.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.16-1 contains the inspections, tests, and analyses for physical security system.

Tier 1 3.16-2 Draft Revision 3

NuScale Tier 1 Physical Security System RAI 14.03-3 Table 3.16-1: Physical Security System Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. Vital equipment will be located only All vital equipment locations will be Vital equipment is located only within within a vital area. inspected. a vital area.

2. Access to vital equipment requires All vital equipment physical barriers Vital equipment is located within a passage through at least two physical will be inspected. protected area such that access to the barriers. vital equipment requires passage through at least two physical barriers.

3. The external walls, doors, ceilings, and Type test, analysis, or a combination of A report exists and concludes that the floors in the MCR and CAS will be type test and analysis of the external walls, doors, ceilings, and floors in the bullet-resistant. walls, doors, ceilings, and floors in the MCR and CAS are bullet-resistant.

MCR and CAS, will be performed.

4. An access control system will be The access control system will be The access control system is installed installed and designed for use by tested. and provides authorized access to vital individuals who are authorized access areas within the nuclear island and to vital areas within the nuclear island structures only to those individuals and structures without escort. with authorization for unescorted access.

5. Unoccupied vital areas within the Tests, inspections, or a combination of Unoccupied vital areas within the nuclear island and structures will be tests and inspections of unoccupied nuclear island and structures are designed with locking devices and vital areas' intrusion detection locked and alarmed and intrusion is intrusion detection devices that equipment and locking devices will be detected and annunciated in the CAS.

annunciate in the CAS. performed.

6. The CAS will be located inside the The CAS will be inspected. The CAS is located inside the protected area and will be designed so protected area, and the interior of the that the interior is not visible from the alarm station is not visible from the perimeter of the protected area. perimeter of the protected area.

7. Security alarm devices in the RXB and All security alarm devices and Security alarm devices, within the CRB, including transmission lines to transmission lines in the RXB and CRB nuclear island and structuresin the annunciators, will be tamper- will be tested. RXB and CRB including transmission indicating and self-checking, and lines to annunciators, are tamper-alarm annunciation indicates the type indicating and self-checking; an of alarm and its location. automatic indication is provided when failure of the alarm system or a component thereof occurs or when the system is on standby power; the alarm annunciation indicates the type of alarm and location.

8. Intrusion detection and assessment Intrusion detection and assessment The intrusion detection systems, systems within the nuclear island and systems in the RXB and CRB will be within the nuclear island and structuresin the RXB and CRB will be tested. structuresin the RXB and CRB provide designed to provide visual display and a visual display and audible audible annunciation of alarms in the annunciation of all alarms in the CAS.

CAS.

9. Intrusion detection systems' recording The intrusion detection systems' Intrusion detection systems' recording equipment will record security alarm recording equipment in the RXB and equipment is capable of recording annunciations within the nuclear CRB will be tested. each security alarm annunciation island and structures including each within the nuclear island and alarm, false alarm, alarm check, and structures, including each alarm, false tamper indication, and the type of alarm, alarm check, and tamper alarm, location, alarm circuit, date, and indication and the type of alarm, time. location, alarm circuit, date, and time.

Tier 1 3.16-3 Draft Revision 3

NuScale Tier 1 Physical Security System Table 3.16-1: Physical Security System Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

10. Emergency exits through the vital area Tests, inspections, or a combination of Emergency exits through the vital area boundaries within the nuclear island tests and inspections of emergency boundaries within the nuclear island and structures will be alarmed with exits through vital area boundaries and structures are alarmed with intrusion detection devices and within within the nuclear island and intrusion detection devices and the nuclear island and structures are structures will be performed. secured by locking devices that allow secured by locking devices that allow prompt egress during an emergency.

prompt egress during an emergency.

11. The CAS will have a landline telephone Tests, inspections, or a combination of The CAS is equipped with landline service with the control room and local tests and inspections of the CAS's telephone service with the control law enforcement authorities. landline telephone service will be room and local law enforcement performed. authorities.

12. The CAS will be capable of continuous Tests, inspections, or a combination of The CAS is capable of continuous communication with on-duty security tests and inspections of the CAS's communication with on-duty force personnel. continuous communication watchmen, armed security officers, capabilities will be performed. armed responders, or other security personnel who have responsibilities within the physical protection program and during contingency response events.

13. Non-portable communications Tests, inspections, or a combination of All nonportable communication equipment in the CAS will remain tests and inspections of the devices in the CAS remain operable operable from an independent power nonportable communications from an independent power source in source in the event of the loss of equipment will be performed. the event of the loss of normal power.

normal power.

Tier 1 3.16-4 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 6 3.17 Radiation Monitoring - NuScale Power Modules 1 - 6 3.17.1 Design Description

System Description

The scope of this section is automatic actions of various systems based on radiation monitoring. Automatic actions of systems based on radiation monitoring are nonsafety-related functions. The systems actuated by these automatic radiation monitoring functions are shared by NuScale Power Modules (NPMs) 1 through 6.

Design Commitments RAI 14.03-3

• The containment flooding and drain system (CFDS) automatically responds to athe CFDS high-radiation signal from 6A-CFD-RT-1007listed in Table 3.17-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The balance-of-plant drain system (BPDS) automatically responds to athe BPDS high-radiation signals from 6A-BPD-RIT-0552listed in Table 3.17-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The BPDS automatically responds to a high-radiation signal from 6A-BPD-RIT-0529 to mitigate a release of radioactivity.

RAI 14.03-3

• The BPDS automatically responds to a high-radiation signal from 6A-BPD-RIT-0705 to mitigate a release of radioactivity.

3.17.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.17-2 contains the inspections, tests, and analyses for radiation monitoring --

NuScale Power Modules 1 - 6.

Tier 1 3.17-1 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 6 RAI 14.03-3 Table 3.17-2: Radiation Monitoring - Inspections, Tests, Analyses, and Acceptance Criteria for NuScale Power Modules 1-6 No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The CFDS automatically responds to A test will be performed of the CFDS Upon initiation of a real or simulated athe CFDS high-radiation signal from high-radiation signal listed in CFDS high-radiation signal listed in 6A-CFD-RT-1007listed in Table 3.17-1 Table 3.17-1. Table 3.17-1, the CFDS automatically to mitigate a release of radioactivity. aligns/actuates the identified components to the positions identified in the table.

2. The BPDS automatically responds to A test will be performed of the BPDS Upon initiation of athe real or athe BPDS high-radiation signals from high-radiation signals listed in simulated BPDS high-radiation signals 6A-BPD-RIT-0552listed in Table 3.17-1 Table 3.17-1. listed in Table 3.17-1 the BPDS to mitigate a release of radioactivity. automatically aligns/actuates the identified components to the positions identified in the table.

3. The BPDS automatically responds to a A test will be performed of the BPDS Upon initiation of a real or simulated high-radiation signal from 6A-BPD-RIT- high-radiation signal. BPDS high-radiation signal listed in 0529 to mitigate a release of Table 3.17-1, the BPDS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

4. The BPDS automatically responds to a A test will be performed of the BPDS Upon initiation of a real or simulated high-radiation signal from 6A-BPD-RIT- high-radiation signal. BPDS high-radiation signal listed in 0705 to mitigate a release of Table 3.17-1, the BPDS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

Tier 1 3.17-3 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 7 - 12 3.18 Radiation Monitoring - NuScale Power Modules 7 - 12 3.18.1 Design Description

System Description

The scope of this section is automatic actions of various systems based on radiation monitoring. Automatic actions of systems based on radiation monitoring are nonsafety-related functions. The systems actuated by these automatic radiation monitoring functions are shared by NuScale Power Modules (NPMs) 7 through 12.

Design Commitments RAI 14.03-3

• The containment flooding and drain system (CFDS) automatically responds to athe CFDS high-radiation signal from 6B-CFD-RT-1007listed in Table 3.18-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The balance-of-plant drain system (BPDS) automatically responds to athe BPDS high-radiation signals from 6B-BPD-RIT-0551listed in Table 3.18-1 to mitigate a release of radioactivity.

RAI 14.03-3

• The BPDS automatically responds to a high-radiation signal from 6B-BPD-RIT-0530 to mitigate a release of radioactivity.

3.18.2 Inspections, Tests, Analyses, and Acceptance Criteria Table 3.18-2 contains the inspections, tests, and analyses for radiation monitoring of NuScale Power Modules 7 - 12.

Tier 1 3.18-1 Draft Revision 3

NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 7 - 12 RAI 14.03-3 Table 3.18-2: Radiation Monitoring Inspections, Tests, Analyses, and Acceptance Criteria For NuScale Power Modules 7 - 12 No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The CFDS automatically responds to A test will be performed of the CFDS Upon initiation of a real or simulated athe CFDS high-radiation signal from high-radiation signal listed in CFDS high-radiation signal listed in 6B-CFD-RT-1007listed in Table 3.18-1 Table 3.18-1. Table 3.18-1, the CFDS automatically to mitigate a release of radioactivity. aligns/actuates the identified components to the positions identified in the table.

2. The BPDS automatically responds to A test will be performed of the BPDS Upon initiation of athe real or athe BPDS high-radiation signals from high-radiation signals listed in simulated BPDS high-radiation signals 6B-BPD-RIT-0551listed in Table 3.18-1 Table 3.18-1. listed in Table 3.18-1, the BPDS to mitigate a release of radioactivity. automatically aligns/actuates the identified components to the positions identified in the table.

3. The BPDS automatically responds to a A test will be performed of the BPDS Upon initiation of a real or simulated high-radiation signal from 6B-BPD-RIT- high-radiation signal. BPDS high-radiation signal listed in 0530 to mitigate a release of Table 3.18-1, the BPDS automatically radioactivity. aligns/actuates the identified components to the positions identified in the table.

Tier 1 3.18-3 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 14.03-3 Table 14.2-9: Auxiliary Boiler System Test # 9 Preoperational test is required to be performed once.

The auxiliary boiler system (ABS) is described in Section 10.4.10 and 11.5.2.2.14 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

None The ABS functions verified by other tests are:

The auxiliary boiler supports the nonsafety-related CPS Test #30-1 condensate polishing system (CPS) by supplying steam for resin regeneration.

The auxiliary boiler supports the turbine nonsafety-related CAR Test #32-1 generator by supplying gland seal steam.

The auxiliary boiler supports the FWS by nonsafety-related CAR Test #32-1 supplying steam to the condenser for sparging when necessary.

The auxiliary boiler supports the module nonsafety-related TG Test #33-1 heatup system (MHS) by supplying steam for heating reactor coolant at startup and shutdown.

Prerequisites

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

ii. Verify a pump curve test has been completed for the auxiliary boiler pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each auxiliary boiler remotely- Operate each valve from the MCR and MCR display and local, visual operated valve can be operated local control panel (if design has local observation indicate each valve fully remotely. valve control). opens and fully closes.

ii. Verify each auxiliary boiler air- Place each valve in its non-safe position. MCR display and local, visual operated valve fails to its safe Isolate and vent air to the valve. observation indicate each valve fails to position on loss of air. its safe position.

iii. Verify each auxiliary boiler air- Place each valve in its non-safe position. MCR display and local, visual operated valve fails to its safe Isolate electrical power to each air- observation indicate each valve fails to position on loss of electrical power to operated valve. its safe position.

its solenoid.

iv. Verify each auxiliary boiler low Align the ABS to allow for pump MCR display and local, visual pressure boiler feedwater pump can operation. observation indicate each pump starts be started and stopped remotely. Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify each auxiliary boiler high Stop and start each pump from the MCR. MCR display and local, visual pressure boiler feedwater pump can observation indicate each pump starts be started and stopped remotely. and stops.

Align the ABS to allow for pump Audible and visible water hammer are operation. not observed when the pump starts.

vi. Verify the speed of each auxiliary Align the ABS to provide a flow path to MCR display indicates the speed of each boiler high pressure boiler feedwater operate a selected AB variable-speed variable speed pump obtains both pump can be manually controlled. pump. minimum and maximum pump speeds.

Vary the auxiliary boiler pump speed from minimum to maximum speed from the MCR.

Tier 2 14.2-33 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-9: Auxiliary Boiler System Test # 9 (Continued) vii. Verify the ABS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler flash tank i. auxiliary boiler flash tank vent radioactivity. vent. isolation valve is closed.

ii. auxiliary boiler high pressure steam supply isolation valves are closed.

[ITAAC 03.09.08]

(i.and ii.)

viii. Verify the ABS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler high auxiliary boiler high pressure to low radioactivity. pressure to low pressure steam supply. pressure steam supply pressure control valve is closed.

[ITAAC 03.09.089]

ix. Verify each ABS instrument is Initiate a single real or simulated The instrument signal is displayed on an available on an MCS or PCS display. instrument signal from each ABS MCS or PCS display, or is recorded by (Test not required if the instrument transmitter. the applicable control system historian.

calibration verified the MCS or PCS display.)

System Level Test Test Objective Test Method Acceptance Criteria None Tier 2 14.2-34 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 09.03.03-1S2, RAI 14.02-6, RAI 14.02-6S1, RAI 14.03-3 Table 14.2-24: Balance-of-Plant Drain System Test # 24 Preoperational test is required to be performed to support sequence of construction turnover of the BPDS system.

BPDS system is described in Section 9.3.3 and 11.5.2.2.15 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The BPDS supports the condensate nonsafety-related Test #24-1 polisher demineralizers, the three Test #24-7 cooling tower chemical addition systems, and the DWS reverse osmosis units by providing a means to collect and transfer chemical wastes to either the LRWS or to the UWS.

2. The BPDS supports the two TGBs, the nonsafety-related Test #24-1 two diesel generators, the auxiliary Test #24-7 boiler, the combustion turbine, the Central Utility Building, and the diesel driven firewater pump by providing a means to collect, treat, and transfer the waste water to the either the LRWS or to the UWS.

3. The BPDS supports the CRB floor nonsafety-related Test #24-1 drains by providing a means to Test #24-7 collect, treat, and transfer the waste water to the UWS.

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 Test Objective Test Method Acceptance Criteria

i. Verify each BPDS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each BPDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each BPDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each BPDS pump can be Align the BPDS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the pump speed of each BPDS Vary the speed of each pump from the MCR display indicates the speed of each variable-speed pump can be MCR and local control panel (if design pump varies from minimum to manually controlled. has local pump control). maximum speed.

vi. Verify each BPDS instrument is Initiate a single real or simulated The instrument signal is displayed on an available on an MCS or PCS display. instrument signal from each BPDS MCS or PCS display, or is recorded by the (Test not required if the instrument transmitter. applicable control system historian.

calibration verified the MCS or PCS display.)

Tier 2 14.2-64 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drain System Test # 24 (Continued)

System Level Test #24-2 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north chemical waste water sump i. The north chemical waste water to mitigate a release of radioactivity. pump in operation. Initiate a real or sump pump stops.

simulated high radiation signal on the ii. North chemical waste collection 60A CPS regeneration skid waste sump to BPDS collection tank effluent. isolation valve is closed.

Repeat the test for each pump. iii. North chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.02]

(i through iii)

System Level Test #24-3 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a south chemical waste water i. The pump stops.

to mitigate a release of radioactivity. sump pump in operation. Initiate a real or ii. South chemical waste collection simulated high radiation signal on the sump to BPDS collection tank 60B CPS regeneration skid waste effluent. isolation valve is closed.

Repeat the test for each pump. iii. South chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.18.02]

(i through iii)

System Level Test #24-4 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north waste water sump pump in i. The north waste water sump pump to mitigate a release of radioactivity. operation. Initiate a real or simulated stops.

high radiation signal in the BPDS north ii. North waste water sump discharge TGB floor drains. to BPDS collection tank isolation Repeat the test for each pump. valve is closed.

iii. North waste water sump discharge to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.023]

(i thorugh iii)

System Level Test #24-5 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a south waste water sump pump in i. The south waste water sump pump to mitigate a release of radioactivity. operation. Initiate a real or simulated stops.

high radiation signal in the BPDS south ii. South waste water sump discharge TGB floor drains. to BPDS collection tank isolation Repeat the test for each pump. valve is closed.

iii. South waste water sump discharge to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.18.023]

(i through iii)

Tier 2 14.2-66 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drain System Test # 24 (Continued)

System Level Test #24-6 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north waste water sump pump in i. The north chemical waste water to mitigate a release of radioactivity. operation. Initiate a real or simulated sump pump stops.

high radiation signal in the BPDS ii. North chemical waste collection auxiliary blowdown cooler condensate. sump to BPDS collection tank Repeat the test for each pump. isolation valve is closed.

iii. North chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.024]

(i through iii)

System Level Test #24-7 Test Objective Test Method Acceptance Criteria Verify BPDS automatically controlled Align each BPDS sump or tank to allow MCR displays and local, visual pumps, in sumps and tanks without a fire water in a selected sump or tank to be observation verifies the following:

water removal pump, start and stop pumped to its design location. If the i. The primary pump starts on HI level automatically and transfer liquid waste sump fill rate in the following test and transfers water to its design to its design location. method is insufficient for automatic start location in the LRWS or UWS system.

of the alternate pump, the primary pump ii. The alternate pump starts on HI-HI may be temporarily removed from level.

service to allow an increase in the sump iii. Both primary and alternate pumps level. stop on LO level.

i. Verify that Pump #1 is set to the iv. The primary pump starts on HI level.

primary pump and Pump #2 is set to alternate. Fill the selected sump or v. The alternate pump starts on HI-HI tank until a HI water level is obtained level.

to start the primary pump.

ii. Continue filling the sump or tank until a HI-HI level starts the alternate pump.

iii. Stop filling the sump or tank to allow the primary and alternate pumps to stop on LO level.

iv. Change pump controls to make Pump #2 the primary pump and Pump #1 the alternate pump, and refill the sump or tank until the primary pump starts on HI level.

v. Continue filling the sump or tank until a HI-HI level starts the alternate pump.

Note: Pump #1 and Pump #2 are not the actual names of the pumps; these names are used to differentiate between the two pumps.

Tier 2 14.2-67 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 14.03-3 Table 14.2-36: Gaseous Radioactive Waste System Test # 36 Preoperational test is required to be performed once.

The GRWS is described in Section 11.3 and 11.5.2.2.6 and the functions verified by this test or another preoperational test are:

System Function System Function Categorization Function Verified by Test #

1. The GRWS supports the LRWS by nonsafety-related Test #36-1 receiving and / or collecting potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.

2. The GRWS supports the CES by nonsafety-related Test #36-1 receiving and / or collecting CES Test #41-2 potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.

3. The NDS supports the GRWS by nonsafety-related Test #36-1 providing nitrogen for purging of the NDS Test #15 component-level tests GRWS.

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 Test Objective Test Method Acceptance Criteria

i. Verify each GRWS remotely-operated Operate each valve from the (main MCR display and local, visual valve can be operated remotely. control room) MCR and local control observation indicate each valve fully panel (if design has local valve control). opens and fully closes.

ii. Verify each GRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each GRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify GRWS valves automatically i. Initiate a real or simulated high GRWS MCR display and local, visual operate to maintain vessel volume. moisture separator level. observation indicate the following:

ii. Initiate a real or simulated low GRWS i. The moisture separator drain valve is moisture separator level. open.

ii. The moisture separator drain valve is closed.

v. Verify GRWS inlet isolation valves Simulate a GRWS inlet stream oxygen MCR display and local, visual automatically close and nitrogen concentration high signal. observation indicate the following:

purge valve opens on high inlet i. The inlet stream isolation valves are stream oxygen concentration. closed.

ii. The nitrogen purge valve is open.

vi. Verify GRWS isolates upon loss of Simulate a loss of RWBVS exhaust flow. MCR display and local, visual RWBV exhaust flow. observation indicate the GRWS isolation valves are closed.

Tier 2 14.2-89 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Gaseous Radioactive Waste System Test # 36 (Continued) vii. Verify radiation isolation of GRWS i. Initiate a real or simulated GRWS MCR display and local, visual charcoal decay beds upon detection train A decay bed discharge flow observation indicate the following:

of decay bed discharge flow high high radiation signal. i. GRWS train A charcoal decay bed radiation level. ii. Initiate a real or simulated GRWS discharge isolation valve is closed.

train B decay bed discharge flow high [ITAAC 03.09.04]

radiation signal.

ii. GRWS train B charcoal decay bed discharge isolation valve is closed.

[ITAAC 03.09.045]

viii. Verify radiation isolation of GRWS Initiate a real or simulated GRWS MCR display and local, visual discharge to the RBVS exhaust upon discharge to the RBVS exhaust high observation indicate the GRWS detection of a high radiation level. radiation signal. discharge to the RBVS exhaust isolation valves are closed.

[ITAAC 03.09.046]

ix. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a GRWS grab sample through the grab sampling device. obtained.

device indicated on the GRWS piping and instrumentation diagram.

x. Verify each GRWS instrument is Initiate a single real or simulated The instrument signal is displayed on an available on an MCS or PCS display. instrument signal from each GRWS MCS or PCS display, or is recorded by the (Test not required if the instrument transmitter. applicable control system historian.

calibration verified the MCS or PCS display.)

System Level Test #36-1 Test Objective Test Method Acceptance Criteria Verify GRWS can process a gaseous waste i. Align GRWS to receive gaseous waste i. The gaseous waste stream is stream and nitrogen stream. from a gaseous waste stream. successfully processed through the Process the gaseous waste stream following processes:

through the gaseous waste process.

• gas cooler ii. Align GRWS charcoal drying heater to

• moisture separator receive nitrogen from NDS.

• charcoal drying heater Process nitrogen through the

• charcoal guard bed charcoal drying process.

• charcoal decay beds

• RWB exhaust ii. Nitrogen is successfully processed through the charcoal drying heater.

Tier 2 14.2-90 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 14.03-3, RAI 14.03.03-3S1, RAI 14.03.03-4S1, RAI 14.03.03-6S1, RAI 14.03.03-7S1 Table 14.2-38: Chemical and Volume Control System Test # 38 Preoperational test is required to be performed for each NPM.

The CVCS is described in Section 9.3.4 and 11.5.2.2.11 and the functions verified by this test, other preoperational tests and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The CVCS supports the RCS by nonsafety-related Test #38-1 providing primary coolant makeup. Ramp Change in Load Demand Test #100

2. The CVCS supports the RCS by nonsafety-related Test #38-1 providing primary coolant letdown. Ramp Change in Load Demand Test #100

3. The CVCS supports the RCS by nonsafety-related Test #38-2 providing pressurizer spray flow for Ramp Change in Load Demand RCS pressure control. Test #100

4. The CVCS supports the RCS by nonsafety-related Test #38-3 changing the boron concentration of the primary coolant.

5. The BAS supports the CVCS by nonsafety-related Test #38-3 providing uniformly mixed borated water on demand.

6. The LRWS supports the CVCS by nonsafety-related Test #38-1 receiving and processing primary LRWS Test #35-2 coolant from CVCS letdown.

The CVCS functions verified by other tests are:

The CVCS supports emergency core nonsafety-related MPS Test #63-6TGS Test #33-1 cooling system (ECCS) valves by providing water to reset the ECCS valves.

The CVCS supports the RCS by heating nonsafety-related TGS Test #33-1 primary coolant.

The CVCS supports the RCS by isolating safety-related MPS Test #63-6 dilution sources.

The CVCS supports the RCS by providing nonsafety-related MPS Test #63-11 primary coolant makeup in beyond design basis events.

Prerequisites

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

ii. Verify a pump curve test has been completed and approved for the CVCS pumps.

iii. Component Level Tests iv., v., and vi. must be performed under preoperational test conditions that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CVCS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CVCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each observation indicate each valve fails to electrical power to its solenoid. air-operated valve. its safe position.

iii. Verify each CVCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position Tier 2 14.2-94 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Chemical and Volume Control System Test # 38 (Continued) xiii. Verify ion exchanger isolation on Initiate a simulated high MCR display and local, visual non-regenerative heat exchanger non-regenerative heat exchanger outlet observation indicate the following:

high outlet temperature to protect temperature signal. i. CVCS purification bypass diverting plant equipment. valve is in the bypass position.

ii. Mixed bed ion exchanger A inlet isolation valves (2) are closed.

iii. Auxiliary ion exchanger inlet isolation valve is closed.

iv. Cation exchanger inlet isolation valve is closed.

xiv. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler steam flow i. CVCS module heatup system 60A radioactivity. to the 60A MHS heat exchanger. heat exchanger inlet and outlet isolation valves are closed.

[This component-level test is required to be performed once for each CVCS associated with the MHS 60A heat exchanger.]

[ITAAC 02.07.023]

xv. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler steam flow i. CVCS module heatup system 60B radioactivity. to the 60B MHS heat exchanger. heat exchanger inlet and outlet isolation valves are closed.

[This component-level test is required to be performed once for each CVCS associated with the MHS 60B heat exchanger.]

[ITAAC 02.07.024]

xvi. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the RCS discharge flow to the i. chemical and volume control RCS radioactivity. regenerative heat exchanger. discharge to process sampling isolation valve closed.

[This component-level test is required to be performed once for each CVCS.]

[ITAAC 02.07.02]

xvii.Verify each CVCS instrument is Initiate a single real or simulated The instrument signal is displayed on an available on an MCS or PCS display. instrument signal from each CVCS MCS or PCS display, or is recorded by the (Test not required if the instrument transmitter. applicable control system historian.

calibration verified the MCS or PCS display.)

Tier 2 14.2-96 Draft Revision 3

NuScale Final Safety Analysis Report Initial Plant Test Program RAI 14.03-3 Table 14.2-60: Plant Lighting System Test # 60 Preoperational test is required to be performed once.

The plant lighting system (PLS) is described in Section 9.5.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. PLS supports the CRB by providing nonsafety-related component-level test i.

normal lighting.

2. The PLS supports the CRB by nonsafety-related component-level test ii.

providing emergency lighting in the main control room.

3. The PLS supports the RXB by nonsafety-related component-level test i.

providing normal lighting.

4. The PLS supports the RXB by nonsafety-related component-level test ii.

providing emergency lighting for the remote shutdown station.

5. The PLS supports the RXB by nonsafety-related component-level test iii.

providing emergency lighting for post-fire safe-shutdown activities outside of the MCR and RSS.

Prerequisites N/A (Note: Component level test iii. supports ITAAC and the requirements of NFPA 804.)

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify the PLS provides normal With normal MCR and RSS lighting in i. a. The PLS provides at least 100 illumination of the MCR and RSS service, measure the light at each MCR foot-candles illumination at the operator workstations, and the MCR and RSS workstation. MCR operator workstations and safety display information panel. at least 50 foot-candles at the MCR auxiliary panels.

[ITAAC 03.08.01]

i. b. The PLS provides at least 100 foot-candles illumination at the RSS operator workstations.

[ITAAC 03.08.01]

ii. The PLS provides emergency With MCR and RSS emergency ii. a. The PLS provides at least 10 illumination of the MCR and RSS illumination in service, measure the light foot-candles of illumination at operator workstations and the MCR at each MCR and RSS workstation and the MCR operator workstations safety display information panel. MCR safety display information panel. and the RSSMCR safety display informationauxiliary panels.

[ITAAC 03.08.02]

ii. b. The PLS provides at least 10 foot-candles at the RSS operator workstations.

[ITAAC 03.08.02]

iii. Verify the eight-hour battery pack With no AC power available, measure the iii. The required target areas are emergency lighting fixtures provide light at each eight-hour battery pack illuminated to provide at least one illumination for post-fire safe- emergency lighting fixture target area. foot-candle illumination in the areas shutdown activities performed by outside the MCR or RSS where post-operators outside the MCR and RSS. fire safe-shutdown activities are performed.

[ITAAC 03.08.03]

Tier 2 14.2-142 Draft Revision 3

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 14.03.03-3S1, 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 3

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 02.01.12 NPM Section 5.3.1.6, Material Surveillance, discusses the use of specimen X capsules installed in specimen guide baskets.

An ITAAC inspection is performed to verify that the correct number of guide baskets are attached to the outer surface of the core barrel at about the mid height of the core support assembly at approximately 90-degree intervalslocations where the capsules will be exposed to a neutron flux consistent with the objectives of the RPV surveillance program.

02.01.13 NPM The CNTS remotely operated CNTS containment isolation valves X are tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test 14.3-21 demonstrates that the CNTS remotely operated CNTS containment Certified Design Material and Inspections, Tests, Analyses, and isolation valves listed in Tier 1 Table 2.1-2 stroke fully open and fully closed by remote operation under preoperational test conditions.

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

02.01.14 NPM The emergency core cooling system (ECCS) safety-related valves X are tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

Acceptance Criteria In accordance with Table 14.2-63, a preoperational test demonstrates that the ECCS safety-related valves listed in Tier 1 Table 2.1-2 stroke fully open and fully closed by remote operation Draft Revision 3 under preoperational test conditions.

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

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 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 The CNTS electrical penetrations listed in Tier 2 Table 2.1-3 may be X Certified Design Material and Inspections, Tests, Analyses, and 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 protection devices whose maximum fault current in these circuits Acceptance Criteria would not damage the electrical penetration assembly if that fault current was available indefinitely. An analysis of a CNTS as-built containment penetration without a circuit interrupting device Draft Revision 3 confirms and concludes in a report that the maximum fault current is less than the current carrying capability of the CNTS containment electrical penetration.Not used.

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 02.05.27 MPS Section 7.0.4.1.2, Reactor Trip System, discusses the arrangement of X the protection system RTBs. Figure 7.0-6: Reactor Trip Breaker Arrangement provides the arrangement of the RTBs.

This ITAAC verifies that the RTBs conform to the arrangement indicated in Tier 1 Figure 2.5-1. In addition, the ITAAC inspection verifies proper connection of the shunt and undervoltage trip mechanisms and other auxiliary contacts.Not used.

02.05.28 MPS Section 7.1.5.1, Application of NUREG/CR-6303 Guidelines, X discusses that two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.

A ITAAC inspection is performed to verify that MPS separation groups A & C and Division I of RTS and ESFAS utilize a different programmable technology from separation groups B & D and 14.3-53 Division II of RTS and ESFAS.Not used.

Certified Design Material and Inspections, Tests, Analyses, and 02.05.29 MPS Section 7.1.3.3, Redundancy in Nonsafety I&C System Design, X discusses that when operators evacuate the MCR and occupy the RSS, two manual isolation switches for the MPS divisions are provided to isolate the MPS manual actuation switches in the MCR to prevent fires in the MCR from causing spurious actuations of associated equipment.

An ITAAC inspection is performed of each MCR isolation switch location to verify that the switch exists in the RSS.

Acceptance Criteria Draft Revision 3

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 02.07.02 Section 11.5.2.2.11, Chemical and Volume Control System, X discusses the operation of the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

In accordance with Table 14.2-38, a preoperational test demonstrates the CVCS and ABS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated CVCS high radiation signal from CVC-RT-30161004, 0A-AB-RIT-1005, and 0B-AB-RIT-1005.

02.07.03 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the auxiliary boiler system (ABS) and the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically 14.3-57 aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-38, a preoperational test demonstrates the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 6A-AB-RT-0142.

02.07.04 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the ABS and the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

In accordance with Table 14.2-38, a preoperational test Acceptance Criteria demonstrates the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation Draft Revision 3 signal from 6B-AB-RT-0141.

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 Section 3.11 presents information to demonstrate that the X 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 non-metallic 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 module-specific, nonsafety-related mechanical equipment that 14.3-60 performs a credited function in Chapter 15 analyses (secondary Certified Design Material and Inspections, Tests, Analyses, and 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 qualification record form.Not used.

Acceptance Criteria Draft Revision 3

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 Section 3.11, Environmental Qualification of Mechanical and X 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 14.3-62 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.Not used.

02.08.06 EQ Section 3.9.6.1, Functional Design and Qualification of Pumps, X 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 3 range of fluid flow, differential pressure, electrical conditions, temperature conditions, and fluid conditions up to and including DBA conditions.Not used.

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

02.08.08 EQ Section 5.4.2, Decay Heat Removal System, discusses that the DHRS X 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 validate that the DHRS passive condensers are capable of meeting 14.3-63 the specified heat transfer performance requirements. Section 5.4.2 discusses the design heat transfer capabilityof the DHR system Certified Design Material and Inspections, Tests, Analyses, and 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.Not used.

Acceptance Criteria Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 09.01.04-1, RAI 09.05.01-6, RAI 14.03-3, 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 14.3-65 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 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 3

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 03.07.01 FP Section 9.5.1.2.6, Fire Protection Design Features, discusses how the X fire protection system (FPS) meets the guidance provided by RG 1.189 and applicable NFPA standards. Two separate dedicated 100 percent capacity freshwater storage tanks are provided.

An ITAAC inspection is performed to verify that the minimum usable water volume of each firewater storage tank is greater than or equal to 300,000 gallons. If the storage tanks are also used as backup water sources for other non-fire emergencies, the ITAAC inspection verifies that the non-fire emergencies cannot drain the tank below the minimum dedicated useable water volume of 300,000 gallons required for firefighting.

03.07.02 FP Section 9.5.1, Fire Protection Program, discusses how the capacity of X each FPS pump is adequate to supply the total flow demand at the pressure required at the pump discharge. Section 9.5.1 provides the 14.3-74 design flow of the fire pumps.

Certified Design Material and Inspections, Tests, Analyses, and

i. An analysis confirms that the as-built fire pumps provide the flow demand for the largest sprinkler or deluge system plus an additional 500 gpm for fire hoses assuming failure of the largest fire pump or loss of off-site power.

ii. In accordance with Table 14.2-25, a preoperational test demonstrates that each fire pump delivers the design flow to the FPS while operating in the fire-fighting alignment.

Acceptance Criteria Draft Revision 3

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 03.07.03 FP Section 9.5.1 discusses that (a) safe-shutdown can be achieved X assuming that all equipment in any one fire area (except for the MCR and containment) is rendered inoperable by fire and that reentry into the fire area for repairs and operator actions is not possible, (b) that smoke, hot gases, or fire suppressant cannot migrate from the affected fire area into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions, and (c) an independent alternative shutdown capability that is physically and electrically independent of the MCR exists.

A safe shutdown analysis of the as-built plant will be performed, including a post-fire safe shutdown circuit analysis performed in accordance with RG 1.189 and NEI 00-01for all possible fire-induced failures that could affect the safe shutdown success path, including multiple spurious actuations.

14.3-75 The safe shutdown analysis will verify that:

Certified Design Material and Inspections, Tests, Analyses, and

• safe shutdown can be achieved assuming that all equipment in any one fire area (except for the MCR and containment) is rendered inoperable by fire and that reentry into the fire area for repairs and operator actions is not possible.

• smoke, hot gases, or fire suppressant cannot migrate from the affected fire area into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions.

• an independent alternative shutdown capability that isMPS equipment rooms within the Reactor Building used as the alternative shutdown capability are physically and electrically Acceptance Criteria independent of the MCR exists.

Draft Revision 3

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 03.07.04 FP Appendix 9A, Fire Hazards Analysis, discusses the methodology and X presents the fire hazards analysis (FHA) for each fire area. The FHA must reflect the as-built configuration of the plant. The FHA is an analysis of the fire hazards, including combustible loading and ignition sources, and analysis of the fire protection features required to mitigate each postulated fire.

An FHA of the as-built plant will be performed in accordance with RG 1.189, as described in Appendix 9A. The FHA will verify (1) combustible loads and ignition sources are accounted for, and (2) fire protection features are suitable for the hazards they are intended for.

03.08.01 PL Section 9.5.3, Lighting Systems, discusses the plant lighting system X X (PLS) which provides normal illumination of the operator workstations and SDIS panels in the MCR and operator workstations in the RSS. The PLS is capable of delivering at least 100 foot-candles 14.3-76 of illumination to the MCR seated operator stations and 50 foot-candles of illumination to the MCR primary operating areas and Certified Design Material and Inspections, Tests, Analyses, and remote and auxiliary operating panels. Lower illumination levels may be used within these areas to ensure more favorable visual conditions, or for areas where critical tasks are not performed.

In accordance with Table 14.2-60, a preoperational test demonstrates that the PLS provides at least 100 foot-candles illumination at the MCR operator workstations and at least 50 foot-candles at the MCR auxiliary panels.:

i. 100 foot-candles illumination at the MCR operator workstations and at least 50 foot-candles at the MCR auxiliary panels.

ii. 100 foot-candles illumination at the RSS operator workstations.

Acceptance Criteria Draft Revision 3

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 03.08.02 PL Section 9.5.3 discusses the PLS which provides emergency X X illumination of the operator workstations and SDIS panels in the MCR and operator workstations in the RSS.

In accordance with Table 14.2-60, a preoperational test demonstrates that the PLS provides at least 10 foot-candles of illumination at the MCR operator workstations and MCR auxiliary panels.:

i. 10 foot-candles of illumination at the MCR operator workstations and MCR auxiliary panels.

ii. 10 foot-candles at the RSS operator workstations.

03.08.03 PL Section 9.5.3 discusses the use of eight-hour battery pack emergency X X lighting fixtures, which provide illumination of at least one foot-candle for post-fire safe shutdown activities outside of the MCR and RSS in accordance with NFPA 804. These units should provide 14.3-77 lighting for:

Certified Design Material and Inspections, Tests, Analyses, and

• areas required for power restoration / recovery to comply with the guidance of RG 1.189, Fire Protection for Nuclear Power Plants.

• areas where normal actions are required for operation of equipment needed during fire; and

• escape or access routes for firefighting and the remote shutdown area.

In accordance with the requirements in Table 14.2-60, a preoperational test demonstrates that eight-hour battery pack emergency lighting fixtures illuminate their required target areas to provide at least one foot-candle illumination in the areas outside the MCR or RSS where post-fire safe-shutdown activities are performed.

Acceptance Criteria Draft Revision 3

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 03.09.04 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the X operation of the gaseous radioactive waste system (GRWS). For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-00461021A, GRW-RIT-1021B, and GRW-RIT-1026.

03.09.05 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the X operation of the GRWS. For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in 14.3-79 the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-0060Not Used.

03.09.06 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the X operation of the GRWS. For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components identified in Tier 1 Acceptance Criteria Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-0071Not Used.

Draft Revision 3

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 03.09.07 RM Section 11.5.2.1.5, Liquid Radioactive Waste System, discusses the X operation of the liquid radioactive waste system (LRWS). For each high radiation signal listed in Tier 1 Table 3.9-1, the LRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-35, a preoperational test demonstrates the LRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated LRWS high radiation signal from 00-LRW-RIT-05691021 and 00-LRW-RIT-05711022.

03.09.08 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the ABS. For each high radiation signal listed in Tier 1 Table 3.9-1, the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

14.3-80 In accordance with Table 14.2-9, a preoperational test demonstrates Certified Design Material and Inspections, Tests, Analyses, and the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 00-AB-RT-0153AB-RIT-1017 and AB-RIT-1017.

03.09.09 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the ABS. For each high radiation signal listed in Tier 1 Table 3.9-1, the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-9, a preoperational test demonstrates the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 00-AB-Acceptance Criteria RT-0166Not Used.

Draft Revision 3

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 03.10.07 RBC Section 9.1.5 discusses that the single failure-proof RBC is classified X as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed to verify that the ASME Type I as-built RBC welds, including wet hoist welds, are nondestructively examined in accordance with the standards of ASME NOG-1.

This ITAAC inspection may be performed any time after manufacture of the single failure proof RBC (at the factory or later).

03.10.08 RBC Section 9.1.5 discusses that the single failure-proof RBC wet hoist is X classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

14.3-84 An ITAAC inspection is performed to verify that the ASME Type I as-built RBC wet hoist welds are nondestructively examined in Certified Design Material and Inspections, Tests, Analyses, and accordance with the standards of ASME NOG-1.

This ITAAC inspection may be performed any time after manufacture of the single failure-proof RBC wet hoist (at the factory or later)Not Used.

03.10.09 RBC Section 9.1.5.2.2 discusses that the MLA is a single-failure-proof X lifting device in accordance with the requirements of ANSI N14.6.

In accordance with ANSI N14.6, and as described in Section 9.1.5.4 the portions of the MLA that are single load path are load tested to 300% (+5%, -0%) of the manufacturer's rating. As part of the rated load test, critical areas of the MLA, including all load-bearing welds, will undergo nondestructive testing as required by ANSI N14.6.

Acceptance Criteria The portions of the MLA that are dual load path are load tested to 150% (+5%, -0%) of the manufacturer's rating in accordance with Draft Revision 3 ANSI N14.6. As part of the rated load test, critical areas of the MLA, including all load-bearing welds, will undergo nondestructive testing as required by ANSI N14.6.

This ITAAC test may be performed any time after manufacture of the MLA (at the factory or later).

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 03.11.07 RXB Section 3.2.1, Seismic Classification, discusses that per RG 1.29, some X SSC that perform no safety-related functions could, if they failed under seismic loading, prevent or reduce the functioning of Seismic Category I SSC.

An ITAAC inspection and analysis is performed to verify that the as-built non-Seismic Category I SSC where there is a potential for adverse interaction with the RXB or a Seismic Category I SSC in the RXB exists will not impair the ability of Seismic Category I SSC to perform their safety functions as demonstrated by one or more of the following criteria:

• Seismic Category I SSC are isolated from non-Seismic Category I SSC so that interaction does not occur.

• Seismic Category I SSC are analyzed to confirm that the ability to perform their safety functions is not impaired as a result of impact 14.3-91 from non-Seismic Category I SSC.

Certified Design Material and Inspections, Tests, Analyses, and

• A non-Seismic Category I restraint system designed to Seismic Category I requirements is used to assure that no interaction occurs between Seismic Category I SSC and non-Seismic Category I SSC.

Acceptance Criteria Draft Revision 3

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 03.14.03 EQ The classification of SSC that contain radioactive waste in accordance X with RG 1.143 is discussed in Section 3.2.1.4.

The scope of the equipment for this design commitment is the nonsafety-related, radioactive waste components and piping which meet both of the following criteria:

• Classified as RW-IIa in accordance with RG 1.143

• Designed for processing gaseous radioactive waste As described in Section 11.2.2.4 for the liquid radioactive waste system (LRWS) and Section 11.3.2.4 for the gaseous radioactive waste system (GRWS), component classification applies to components up to and including the first isolation device. Tier 1 Table 3.14-1 identifies the components and piping for which this ITAAC is applicable.

14.3-102 An ITAAC inspection and reconciliation analysis is performed of the Certified Design Material and Inspections, Tests, Analyses, and as-built LRWS and GRWS RW-IIa components and piping used for processing gaseous radioactive waste to ensure that deviations between the drawings used for construction and the as-built RW-IIa components and piping are reconciled. A report concludes the as-built RW-IIa components and piping meet the design criteria of RG 1.143, RW-IIa.

Acceptance Criteria Draft Revision 3

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 03.14.04 EQ The CNTS electrical penetrations listed in Tier 1 Table 2.1-3 may be X 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.

Section 8.3.1.2.5 Containment Electrical Penetration Assemblies discusses electrical penetration assemblies that are not equipped with 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 14.3-103 containment penetration without a circuit interrupting device confirms and concludes in a report that the maximum fault current is Certified Design Material and Inspections, Tests, Analyses, and less than the current carrying capability of the CNTS containment electrical penetration.

Acceptance Criteria Draft Revision 3

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 03.14.05 EQ Section 3.11 presents information to demonstrate that the X 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 14.3-104 module-specific, nonsafety-related mechanical equipment that performs a credited function in Chapter 15 analyses (secondary main Certified Design Material and Inspections, Tests, Analyses, and 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 qualification record form.

Acceptance Criteria Draft Revision 3

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 03.14.06 EQ Section 3.11, Environmental Qualification of Mechanical and X 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 14.3-105 electrical surges that would exist before, during, and following a DBA without loss of safety-related function.

Certified Design Material and Inspections, Tests, Analyses, and 03.14.07 EQ Section 3.9.6.1, Functional Design and Qualification of Pumps, Valves, X 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.

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 Acceptance Criteria Qualification Report concludes that safety-related valves are capable of performing their safety-related function under the full range of fluid flow, differential pressure, electrical conditions, temperature conditions, and fluid conditions up to and including DBA conditions.

Draft Revision 3

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 03.14.08 EQ Section 5.4.2, Decay Heat Removal System, discusses that the DHRS X 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 validate that the DHRS passive condensers are capable of meeting the specified heat transfer performance requirements. Section 5.4.2 discusses the design heat transfer capability of the DHR system 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.

03.15.01 HFE Section 18.11, Design Implementation, describes the implementation of HFE aspects of the plant design.

14.3-106 The Design Implementation activities verify that the final MCR is Certified Design Material and Inspections, Tests, Analyses, and consistent with the verified and validated design resulting from the HFE design process. An ITAAC inspection is performed to verify that the as-built configuration of main control room HSI is consistent with the final as-designed HSI configuration. As used here, the final as-designed HSI configuration is the COL holders configuration-controlled design, which includes changes made subsequent to integrated system validation under a licensees configuration control process and includes resolution of human engineering discrepancies.An ITAAC inspection is performed to verify that the as-built configuration of main control room HSI is consistent with the as-designed configuration of main control room HSI as modified by the Integrated System Validation Report.

Acceptance Criteria Draft Revision 3

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 03.17.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the balance-of-plant drain system (BPDS). For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 60A-BPD-RIT-10100552, 0A-BPD-RIT-1001, and 00-BPD-RIT-1034.

03.17.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components 14.3-114 identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6A-BPD-RIT-0529.

03.17.04 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a Acceptance Criteria preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high Draft Revision 3 radiation signal from 6A-BPD-RIT-0705.

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 03.18.01 RM Section 11.5.2.2.9, Containment Flooding and Drain System, X discusses the operation of the containment flooding and drain system (CFDS). For each high radiation signal listed in Tier 1 Table 3.18-1, the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-42, a preoperational test demonstrates the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated CFDS high radiation signal from 60B-CFD-RT-1007.

03.18.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in 14.3-115 the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 60B-BPD-RIT-10100552 and 0B-BPD-RIT-1001.

03.18.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the Acceptance Criteria components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6B-BPD-RIT-0529.

Draft Revision 3 Note:

1. References to Sections, Figures and Tables in Table 14.3-2 refer to Tier 2 unless the reference specifically states Tier 1 Sections, Figures or Tables.