RAIO-0719-66362, 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 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com 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
RAIO-0719-66362 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :
NuScale Response to NRC Request for Additional Information eRAI No. 9681
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.
- 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.
- 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.
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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.
- 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.
- 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.
- 22. ITAAC 7 and 8 in Table 3.16-1: To make the scope of the ITAAC consistent among the DC, ITA, and AC.
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.
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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.
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NuScale Tier 1 Definitions Tier 1 1.1-1 Draft Revision 3 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.
NuScale Tier 1 Definitions Tier 1 1.1-2 Draft Revision 3 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
NuScale Tier 1 Definitions Tier 1 1.1-3 Draft Revision 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.
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.
NuScale Tier 1 Definitions Tier 1 1.1-4 Draft Revision 3 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.
NuScale Tier 1 General Provisions Tier 1 1.2-2 Draft Revision 3 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:
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.
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 General Provisions Tier 1 1.2-3 Draft Revision 3 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.
NuScale Tier 1 NuScale Power Module Tier 1 2.1-3 Draft Revision 3 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.
NuScale Tier 1 NuScale Power Module Tier 1 2.1-4 Draft Revision 3 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
NuScale Tier 1 NuScale Power Module Tier 1 2.1-5 Draft Revision 3 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.
NuScale Tier 1 NuScale Power Module Tier 1 2.1-13 Draft Revision 3 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 Code Class 1, 2 and 3 piping systems listed in Table 2.1-1 comply with ASME Code Section III requirements.
An inspection will be performed of the NuScale Power Module ASME Code Class 1, 2 and 3 as-built piping system Design Reports for systems listed in Table 2.1-1 required by ASME Code Section III.
The ASME Code Section III Design Reports (NCA-3550) exist and conclude that the NuScale Power Module ASME Code Class 1, 2 and 3 as-built piping systems listed in Table 2.1-1 meet the requirements of ASME Code Section III.
2.
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.
An inspection will be performed of the NuScale Power Module ASME Code Class 1, 2, and 3 as-built component Data Reports for components listed in Table 2.1-2 required by ASME Code Section III.
ASME Code Section III Data Reports for the NuScale Power Module ASME Code Class 1, 2, and 3 components listed in Table 2.1-2 and interconnecting piping exist and conclude that the requirements of ASME Code Section III are met.
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.
An inspection will be performed of the NuScale Power Module ASME Code Class CS as-built component Data Reports for components listed in Table 2.1-2 required by ASME Code Section III.
ASME Code Section III Data Reports for the NuScale Power Module ASME Code Class CS components listed in Table 2.1-2 exist and conclude that the requirements of ASME Code Section III are met.
4.
Safety-related SSC are protected against the dynamic and environmental effects associated with postulated failures in high-and moderate-energy piping systems.
An inspection and analysis will be performed of the as-built high-and moderate-energy piping systems and protective features for the safety-related SSC.
Protective features are installed in accordance with the as-built Pipe Break Hazard Analysis Report and safety-related SSC are protected 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 Code Class 2 piping systems listed in Table 2.1-1 and interconnected equipment nozzles are evaluated for LBB.
An analysis will be performed of the ASME Code Class 2 as-built piping systems listed in Table 2.1-1 and interconnected equipment nozzles.
The as-built LBB analysis for the ASME Code Class 2 piping systems listed in Table 2.1-1 and interconnected equipment nozzles is bounded by the as-designed LBB analysis.
6.
The RPV beltline material has a Charpy upper-shelf energy of 75 ft-lb minimum.
A vendor test will be performed of the Charpy V-Notch specimen of the RPV beltline material.
An ASME Code Certified Material Test Report exists and concludes that the initial RPV beltline material Charpy upper-shelf energy is 75 ft-lb minimum.
7.
The CNV serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.
A leakage test will be performed of the pressure containing or leakage-limiting boundaries, and CIVs.
The leakage rate for local leak rate tests (Type B and Type C) for pressure containing or leakage-limiting boundaries and CIVs meets the requirements of 10 CFR Part 50, Appendix J.
8.
Containment isolation valve cClosure times for CIVs listed in Table 2.1-3 limit potential releases of radioactivity.
A test will be performed of the automatic CIVs listed in Table 2.1-3.
Each CIV listed in Table 2.1-3 travels from the full open to full closed position in less than or equal to the time listed in Table 2.1-3 after receipt of a containment isolation signal.
NuScale Tier 1 NuScale Power Module Tier 1 2.1-14 Draft Revision 3 9.
The length of piping listed in Table 2.1-1 shall be minimized between the containment penetration and the associated outboard CIVs.
An inspection will be performed of the as-built piping listed in Table 2.1-1 between containment penetrations and associated outboard CIVs.
The length of piping between each containment penetration and its associated outboard CIV is less than or equal to the length identified in Table 2.1-1.
10.
The CNTS containment electrical penetration assemblies listed in Table 2.1-3 are sized to power their design loads.
i.
An analysis will be performed of the CNTS as-designed containment electrical penetration assemblies listed in Table 2.1-3.
i.
An electrical rating report exists that defines and identifies the required design electrical rating to power the design loads of each CNTS containment electrical penetration assembly listed in Table 2.1-3.
ii.
An inspection will be performed of CNTS as-built containment electrical penetration assembliesy listed in Table 2.1-3.
ii.
The electrical rating of each CNTS containment electrical penetration assembly listed in Table 2.1-3 is greater than or equal to the required design electrical rating as specified in the electrical rating report.
11.
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 instrumentNot used.
An inspection will be performed of the MPS Class 1E as-built instrumentation and control current-carrying circuitsNot used.
i.
Physical separation between redundant divisions of MPS 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.
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 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.
An inspection will be performed of the as-built RPV surveillance capsule holders.
Four surveillance capsule holders are installed in the RPV beltline region at locations where the capsules will be exposed to a neutron flux consistent with the objectives of the RPV surveillance program.approximately 90 degree intervals.
Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 NuScale Power Module Tier 1 2.1-15 Draft Revision 3 13.
The remotely-operated CNTS containment isolation valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.
A test will be performed of the CNTS safety-relatedremotely-operated CNTS containment isolation valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each remotely-operated CNTS containment isolation valve listed in Table 2.1-2 strokes fully open and fully closed by remote operation under preoperational temperature, differential pressure, and flow conditions.
14.
The ECCS safety-related valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.
A test will be performed of the ECCS safety-related valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each ECCS safety-related valve listed in Table 2.1-2 strokes fully open and fully closed by remote operation under preoperational temperature, differential pressure, and flow conditions.
15.
The DHRS safety-related valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.
A test will be performed of the DHRS safety-related valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each DHRS safety-related valve listed in Table 2.1-2 strokes fully open and fully closed by remote operation under preoperational temperature, differential pressure, and flow conditions.
16.
Not used.
Not used.
Not used.
17.
Not used.
Not used.
Not used.
18.
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.
A test will be performed of the CNTS safety-related hydraulic-operated valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each CNTS safety-related hydraulic-operated valve listed in Table 2.1-2 fails to (or maintains) its safety-related position on loss of motive power under preoperational temperature, differential pressure, and flow conditions.
19.
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.
A test will be performed of the ECCS safety-related RRVs and RVVs listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each ECCS safety-related RRV and RVV listed in Table 2.1-2 fails to (or maintains) its safety-related positionopen on loss of electrical power to its corresponding trip valve under preoperational temperature, differential pressure, and flow conditions.
20.
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.
A test will be performed of the DHRS safety-related hydraulic-operated valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each DHRS safety-related hydraulic-operated valve listed in Table 2.1-2 fails to (or maintains) its safety-related positionopen on loss of motive power under preoperational temperature, differential pressure, and flow conditions.
21.
The CNTS safety-related check valves listed in Table 2.1-2 change position under design-basis temperature, differential pressure, and flow conditions.
A test will be performed of the CNTS safety-related check valves listed in Table 2.1-2 under preoperational temperature, differential pressure, and flow conditions.
Each CNTS safety-related check valve listed in Table 2.1-2 strokes fully open and closed (under forward and reverse flow conditions, respectively) under preoperational temperature, differential pressure, and flow conditions.
Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 NuScale Power Module Tier 1 2.1-16 Draft Revision 3 22.
- i. 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 ii. A CNTS containment electrical penetration assembly is rated to withstand the maximum fault current for its circuits without a circuit interrupting device.Not used.
- i. An analysis will be performed of the CNTS as-built containment electrical penetration assembly.Not used.
- i. A circuit interrupting device coordination analysis exists and concludes that the current carrying capability for each CNTS containment electrical penetration assembly listed 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. An analysis of the CNTS containment penetration maximum fault current exists and concludes the fault current is less than the current carrying capability of the CNTS containment electrical penetrationNot used.
23.
The CNV serves as an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment.
A preservice design pressure leakage test of the CNV will be performed.
No water leakage is observed at CNV bolted flange connections.
24.
The NPM lifting fixture supports its rated load.
A rated load test will be performed of the NPM lifting fixture.
The NPM lifting fixture supports a load of at least 150 percent of the manufacturer's rated capacity.
25.
The NPM lifting fixture is constructed to provide assurance that a single failure does not result in the uncontrolled movement of the lifted load.
An inspection will be performed of the as-built NPM lifting fixture.
The NPM lifting fixture is single-failure-proof.
26.
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.
An inspection will be performed of each ECCS valve, CIV, and DHRS actuation valve listed in Table 2.1-2, and associated hydraulic line.
A report exists and concludes each ECCS valve, CIV, and DHRS actuation valve listed in Table 2.1-2, and the associated hydraulic line, is installed in accordance with its associated installation specification.
Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Chemical and Volume Control System Tier 1 2.2-1 Draft Revision 3 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.
NuScale Tier 1 Chemical and Volume Control System Tier 1 2.2-4 Draft Revision 3 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 system ASME Code Class 3 piping listed in Table 2.2-1system complies with the ASME Code Section III.
An inspection will be performed of the chemical and volume control system ASME Code Class 3 as-built piping system Design Report required by ASME Code Section III for piping listed in Table 2.2-1.
The ASME Code Section III Design Report (NCA-3550) exists and concludes that the chemical and volume control listed in Table 2.2-1system ASME Code Class 3 as-built piping system meets the requirements of ASME Code Section III.
2 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 An inspection will be performed of the chemical and volume control system ASME Code Class 3 as-built component Data Reports required by ASME Code Section III for components listed in Table 2.2-2.
ASME Code Section III Data Reports for the chemical and volume control system ASME Code Class 3 components listed in Table 2.2-2 and interconnecting piping exist and conclude that the requirements of ASME Code Section III are met.
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.
A test will be performed of the chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valves listed in Table 2.2-2 under preoperational temperature, differential pressure, and flow conditions.
Each chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valve listed in Table 2.2-2 strokes fully open and fully closed by remote operation under preoperational temperature, differential pressure, and flow conditions.
4 Not used.
Not used.
Not used.
5 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.
A test will be performed of the chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valves listed in Table 2.2-2 under preoperational temperature, differential pressure and flow conditions.
Each chemical and volume control system ASME Code Class 3 air-operated demineralized water system supply isolation valve listed in Table 2.2-2 performs its function to fail to (or maintain) its positionperforms fails closed on loss of motive power under preoperational temperature, differential pressure, and flow conditions.
NuScale Tier 1 Containment Evacuation System Tier 1 2.3-2 Draft Revision 3 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 supports RCS leakage detection.
A test will be performed of the CES level instrumention.
The CES level instrumentation detects 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 supports RCS leakage detection.
A test will be performed of the CES pressure instrumentation.
The CES detects a pressure increase in 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.
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-3 Draft Revision 3 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
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-4 Draft Revision 3 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.
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-5 Draft Revision 3 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).
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-16 Draft Revision 3 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 are implemented using a quality process composed of the following system design lifecycle phases, with each phase having outputs which satisfy the requirements of that phase.
i.a.
System Concept Functional Specification Phase i.b. System RequirementsDesign Phase
- System Prototype Development Phase
- Equipment Requirements Specification Phase
- Hardware Planning Phase
- Hardware Requirements Phase
- Hardware Design Phase
- Software Planning Phase
- Software Requirements Phase i.c.
SystemSoftware Design Phase i.d. SystemSoftware Implementation Phase
- Software Configuration Phase i.e.
System Testing Phase i.f.
System Installation and Checkout Phase i.a.
An analysis will be performed of the output documentation of the System ConceptFunctional Specification Phase.
i.a.
The output documentation of the MPS ConceptFunctional Specification Phase satisfies the requirements of the System ConceptFunctional Specification Phase.
ii.b. An analysis will be performed of the output documentation of the System RequirementsDesign Phase.
ii.b. The output documentation of the MPS RequirementsDesign Phase satisfies the requirements of the System RequirementsDesign Phase.
iii.c.An analysis will be performed of the output documentation of the System DesignPrototype Development Phase.
iii.c.The output documentation of the MPS DesignPrototype Development Phase satisfies the requirements of the System DesignPrototype Development Phase.
ivi.d.An analysis will be performed of the output documentation of the System ImplementationEquipment Requirements Specification Phase.
ivi.d.The output documentation of the MPS ImplementationEquipment Requirements Specification Phase satisfies the requirements of the System ImplementationEquipment Requirements Specification Phase.
vi.e.An analysis will be performed of the output documentation of the System TestHardware Planning Phase.
vi.e.The output documentation of the MPS TestHardware Planning Phase satisfies the requirements of the System TestHardware Planning Phase.
vi.f. An analysis will be performed of the output documentation of the System Installation and CheckoutHardware Requirements Phase.
vi.f. The output documentation of the MPS Installation and CheckoutHardware Requirements Phase satisfies the requirements of the System Installation and CheckoutHardware Requirements Phase.
vii. An analysis will be performed of the output documentation of the Hardware Design Phase.
vii. The output documentation of the MPS Hardware Design Phase satisfies the requirements of the Hardware Design Phase.
viii. An analysis will be performed of the output documentation of the Software Planning Phase.
viii. The output documentation of the MPS Software Planning Phase satisfies the requirements of the Software Planning Phase.
ix. An analysis will be performed of the output documentation of the Software Requirements Phase.
ix. The output documentation of the MPS Software Requirements Phase satisfies the requirements of the Software Requirements Phase.
x.
An analysis will be performed of the output documentation of the Software Design Phase.
x.
The output documentation of the MPS Software Design Phase satisfies the requirements of the Software Design Phase.
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-17 Draft Revision 3 xi. An analysis will be performed of the output documentation of the Software Implementation Phase.
xi. The output documentation of the MPS Software Implementation Phase satisfies the requirements of the Software Implementation Phase.
xii. An analysis will be performed of the output documentation of the Software Configuration Phase.
xii. The output documentation of the MPS Software Configuration Phase satisfies the requirements of the Software Configuration Phase.
xiii. An analysis will be performed of the output documentation of the System Testing Phase.
xiii. The output documentation of the MPS Testing Phase satisfies the requirements of the System Testing Phase.
xiv. An analysis will be performed of the output documentation of the System Installation Phase.
xiv. The output documentation of the MPS Installation Phase satisfies the requirements of the System Installation Phase.
ii.
Protective measures are provided to restrict modifications to the MPS tunable parameters.
ii.
Test will be performed on the access control features associated with MPS tunable parameters.
ii.
Protective measures restrict modification to the MPS tunable parameters without proper configuration and authorization.
iii.a. Communications independence exists between Separation Groups A, B, C, and D Class 1E MPS.
iii.b. Communications independence exists between Division I and II of the Class 1E MPS.
iii.
A test will be performed of the Class 1E MPS.
iii.a. Communications independence between Separation Groups A, B, C, and D of the Class 1E MPS is provided.
iii.b. Communications independence between Division I and II of the Class 1E MPS is provided.
iv.
The MPS automatically initiates a reactor trip signal for reactor trip functions listed in Table 2.5-1.
iv.
A test will be performed of the MPS.
iv.
Reactor trip signal is automatically initiated for each reactor trip function listed in Table 2.5-1.
v.
The MPS automatically initiates an ESF actuation signal for ESF functions listed in Table 2.5-2.
v.
A test will be performed of the MPS.
v.
An ESF actuation signal is automatically initiated for each ESF function listed in Table 2.5-2.
vi.
The MPS automatically actuates a reactor trip.
vi.
A test will be performed of the MPS.
vi.
The RTBs open upon an injection of a single simulated MPS reactor trip signal.
vii.
The MPS manually actuates a reactor trip.
vii.
A test will be performed of the MPS.
vii.
The RTBs open when a reactor trip is manually initiated from the main control room.
viii. The reactor trip logic fails to a safe state such that loss of electrical power to a MPS separation group results in a trip state for that separation group.
viii. A test will be performed of the MPS.
viii. Loss of electrical power in a separation group results in a trip state for that separation group.
ix.
The ESFs logic fails to a safe state such that loss of electrical power to a MPS separation group results in a safe state listed in Table 2.1-3.
ix.
A test will be performed of the MPS.
ix.
Loss of electrical power in a separation group results in the safe state listed in Table 2.1-3.
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
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-18 Draft Revision 3 x.
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.
x.
A test will be performed of the MPS.
x.
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. The operating bypass is automatically removed when the interlock condition is no longer satisfied.
xi.
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.
xi.
A test will be performed of the MPS.
xi.
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. The operating bypass is automatically removed when the permissive condition is no longer satisfied.
xii.
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.
xii.
A test will be performed of the MPS.
xii.
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.
xiii. The MPS is capable of performing its safety-related functions when any one of its separation channels is out of service.
xiii. A test will be performed of the MPS.
xiii. The MPS performs its safety-related functions if any one of its separation groups is out of service.
xiv. The RTBs are installed and arranged as shown in Figure 2.5-2 in order to successfully accomplish the reactor trip function.
xiv. An inspection will be performed of the as-built RTBs, including the connections for the shunt and undervoltage trip mechanism and auxiliary contacts.
xiv. The RTBs have the proper connections for the shunt and undervoltage trip mechanisms and auxiliary contacts, and are arranged as shown in Figure 2.5-2 to successfully accomplish the reactor trip function.
xv.
Two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.
xv.
An inspection will be performed of the as-built MPS.
xv.
Separation groups A & C and Division I of RTS and ESFAS utilize a different programmable technology from separation groups B & D and Division II of RTS and ESFAS.
2.
Protective measures are provided to restrict modifications to the MPS tunable parametersNot used.
A test will be performed on the access control features associated with MPS tunable parametersNot used.
Protective measures restrict modification to the MPS tunable parameters without proper configuration and authorizationNot used.
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
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-19 Draft Revision 3 3.
Physical separation exists (i) between the redundanteach 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.
An inspection will be performed of the MPS Class 1E as-built instrumentation and control current-carrying circuits.
i.
Physical separation between redundanteach separation groups and divisions of the MPS 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.
ii.
Physical separation between each division of the MPS 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.
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 the redundanteach 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 instrumentation and control circuits and non-Class 1E instrumentation and control circuits to prevent the propagation of credible electrical faults.
An inspection will be performed of the MPS Class 1E as-built instrumentation and control circuits.
i.
Class 1E electrical isolation devices are installed between redundanteach separation groups and divisions of the MPS Class 1E instrumentation and control circuits.
ii.
Class 1E electrical isolation devices are installed between each division of the MPS Class 1E instrumentation and control circuits.
iii. Class 1E electrical isolation devices are installed between MPS Class 1E instrumentation and control circuits and non-Class 1E instrumentation and control circuits.
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
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-20 Draft Revision 3 5.
Electrical isolation exists between the EDSS-MS subsystem non-Class 1E circuits and connected MPS Class 1E circuits to prevent the propagation of credible electrical faults.
i.
A type test, analysis, or a combination of type test and analysis will be performed of the Class 1E isolation devices.
i.
The Class 1E circuit does not degrade below defined acceptable operating levels when the non-Class 1E side of the isolation device is subjected to the maximum credible voltage, current transients, shorts, grounds, or open circuits.
ii.
An inspection will be performed of the MPS Class 1E as-built circuits.
ii.
Class 1E electrical isolation devices are installed between the EDSS-MS Subsystem non-Class 1E circuits and connected MPS Class 1E circuits.
6.
Communications independence exists between redundant separation groups and divisions of the Class 1E MPSNot used.
A test will be performed of the Class 1E MPSNot used.
Communications independence between redundant separation groups and divisions of the Class 1E MPS is providedNot used.
7.
Communications independence exists between the Class 1E MPS and non-Class 1E digital systems.
A test will be performed of the Class 1E MPS.
Communications independence between the Class 1E MPS and non-Class 1E digital systems is provided.
8.
The MPS automatically initiates a reactor trip signalNot used.
A test will be performed of the MPSNot used.
A reactor trip signal is automatically initiated for each reactor trip function listed in Table 2.5-1Not used.
9.
The MPS automatically initiates an ESF actuation signalNot used.
A test will be performed of the MPSNot used.
An ESF actuation signal is automatically initiated for each ESF function listed in Table 2.5-2Not used.
10.
The MPS automatically actuates a reactor trip.Not used.
A test will be performed of the MPS.Not used.
The RTBs open upon an injection of a single simulated MPS reactor trip signal.Not used.
11.
The MPS automatically actuates the engineered safety feature equipment to perform its safety-related function listed in Table 2.5-2.
A test will be performed of the MPS.
The ESF equipment automatically actuates to perform its safety-related function listed in Table 2.5-2 upon an injection of a single simulated MPS signal.
12.
The MPS manually actuates a reactor trip.Not used.
A test will be performed of the MPS.Not used.
The RTBs open when a reactor trip is manually initiated from the main control room.Not used.
13.
The MPS manually actuates the ESF equipment to perform its safety-related function listed in Table 2.5-2.
A test will be performed of the MPS.
The MPS actuates the ESF equipment to perform its safety-related function listed in Table 2.5-3Table 2.5-2 when manually initiated.
14.
The reactor trip logic fails to a safe state such that loss of electrical power to a MPS separation group results in a trip state for that separation group.Not used.
A test will be performed of the MPS.Not used.
Loss of electrical power in a separation group results in a trip state for that separation group.Not used.
15.
The ESFs logic fails to a safe state such that loss of electrical power to a MPS separation group results in a predefined safe state for that separation group.Not used.
A test will be performed of the MPS.Not used.
Loss of electrical power in a separation group results in an actuation state for that separation group.Not used.
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
NuScale Tier 1 Module Protection System and Safety Display and Indication System Tier 1 2.5-22 Draft Revision 3 24.
The MPS self-test features detect faults in the system and provide an alarm in the main control room.
A test will be performed of the MPS.
A report exists and concludes that:
- Self-testing features verify that faults 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 are indicated on the SDIS displays in the MCR.
An inspection will be performed for the ability to retrieve the as-built PAM Type B and Type C displays on the SDIS displays in the MCR.
The PAM Type B and Type C displays listed in Table 2.5-5 are retrieved and displayed on the SDIS displays in the MCR.
26.
The controls located on the operator workstations in the MCR operate to perform IHAs.
A test will be performed of the controls on the operator workstations in the MCR.
The IHAs controls provided on the operator workstations in the MCR perform the functions listed in Table 2.5-6.
27.
The RTBs are installed and arranged in order to successfully accomplish the reactor trip function under design conditions.Not used.
An inspection will be performed of the as-built RTBs, including the connections for the shunt and undervoltage trip mechanism and auxiliary contacts.Not used.
The RTBs have the proper connections for the shunt and undervoltage trip mechanisms and auxiliary contacts, and are arranged as shown in Figure 2.5-2 to successfully accomplish the reactor trip function.Not used.
28.
Two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.Not used.
An inspection will be performed of the as-built MPS.Not used.
Separation groups A & C and Division I of RTS and ESFAS utilize a different programmable technology from separation groups B & D and Division II of RTS and ESFAS.Not used.
29.
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.
An inspection will be performed of the location of the as-built MCR isolation switches.
The MCR isolation switches are located in the remote shutdown station.
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
NuScale Tier 1 Neutron Monitoring System Tier 1 2.6-1 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring Module Specific Tier 1 2.7-1 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring Module Specific Tier 1 2.7-3 Draft Revision 3 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 athe CES high radiation signal from CES-RT-1011listed in Table 2.7-1 to mitigate a release of radioactivity.
A test will be performed of the CES high radiation signal listed in Table 2.7-1.
Upon initiation of a real or simulated CES high radiation signal listed in Table 2.7-1, the CES automatically aligns/actuates the identified components to the positions identified in the table.
2.
The CVCS automatically responds to the CVCS and ABSa high radiation signals listed in Table 2.7-1 from CVC-RT3016 to mitigate a release of radioactivity.
A test will be performed of the CVCS and ABS high radiation signals listed in Table 2.7-1.
Upon initiation of athe real or simulated CVCS and ABS high radiation signals listed in Table 2.7-1, the CVCS automatically aligns/
actuates the identified component(s) to the position identified in the table.
3.
The CVCS automatically responds to a high radiation signal from 6A-AB-RT-0142 to mitigate a release of radioactivity.
A test will be performed of the CVCS high radiation signal.
Upon initiation of a real or simulated CVCS high radiation signal listed in Table 2.7-1, the CVCS automatically aligns/actuates the identified component to the position identified in the table.
4.
The CVCS automatically responds to a high radiation signal from 6B-AB-RT-0141 to mitigate a release of radioactivity.
A test will be performed of the CVCS high radiation signal.
Upon initiation of a real or simulated CVCS high radiation signal listed in Table 2.7-1, the CVCS automatically aligns/actuates the identified component to the position identified in the table.
NuScale Tier 1 Equipment Qualification Tier 1 2.8-1 Draft Revision 3 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,
NuScale Tier 1 Equipment Qualification Tier 1 2.8-2 Draft Revision 3 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.
NuScale Tier 1 Equipment Qualification Tier 1 2.8-17 Draft Revision 3 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 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 an 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:
- Nonsafety-related mechanical and electrical equipment located within the boundaries of the NuScale Power Module that has an augmented Seismic Category I design requirement.
- 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.)
i.
A type test, analysis, or a combination of type test and analysis will be performed of the module-specific Seismic Category I equipment listed in Table 2.8-1, including its associated supports and anchorages.
ii.
An inspection will be performed of the module-specific Seismic Category I as-built equipment listed in Table 2.8-1, including its associated supports and anchorages.
i.
A seismic qualification record form exists and concludes that the module-specific Seismic Category I equipment listed in Table 2.8-1, including its associated supports and anchorages, will withstand the design basis seismic loads and perform its function(s) during and after an SSE.
ii.
The module-specific Seismic Category I equipment listed in Table 2.8-1, including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the equipments seismic qualification record form.
2.
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, AOOs, 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, 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.
i.
A type test or a combination of type test and analysis will be performed of the module-specific electrical equipment listed in Table 2.8-1, including associated connection assemblies.
ii.
An inspection will be performed of the module-specific as-built electrical equipment listed in Table 2.8-1, including associated connection assemblies.
i.
An EQ record form exists and concludes that the module-specific electrical equipment listed in Table 2.8-1, including associated connection assemblies, perform their function under the environmental conditions specified in the EQ record form for the period of time required to complete the function.
ii.
The module-specific electrical equipment listed in Table 2.8-1, including associated connection assemblies, are installed in their design location in a configuration bounded by the EQ record form.
NuScale Tier 1 Equipment Qualification Tier 1 2.8-18 Draft Revision 3 3.
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.)Not used.
A type test or a combination of type test and analysis will be performed of the non-metallic parts, materials, and lubricants used in module-specific mechanical equipment.Not used.
A qualification record form exists and concludes that 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 under the design basis harsh environmental conditions (both internal service conditions and external environmental conditions) specified in the qualification record form.Not used.
4.
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.
i.
A type test or a combination of type test and analysis will be performed of the Class 1E computer-based instrumentation and control systems listed in Table 2.8-1 located in a mild environment.
ii.
An inspection will be performed of the Class 1E as-built computer-based instrumentation and control systems listed in Table 2.8-1 located in a mild environment.
i.
An EQ record form exists and concludes that the Class 1E computer-based instrumentation and control systems listed in Table 2.8-1 located in a mild environment perform their function under the environmental conditions specified in the EQ record form.
ii.
The Class 1E computer-based instrumentation and control 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 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.Not used.
A type test, analysis, or a combination of type test and analysis will be performed of the Class 1E digital equipment.Not used.
An EQ record form exists and concludes that the Class 1E digital equipment listed in Table 2.8-1 withstands the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA without loss of safety-related function.Not used.
Table 2.8-2: Equipment Qualification Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Equipment Qualification Tier 1 2.8-19 Draft Revision 3 6.
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, temperature conditions, and fluid conditions up to and including DBA conditions.Not used.
A type test or a combination of type test and analysis will be performed of the safety-related valves.Not used.
A Qualification Report exists and concludes that the safety-related valves listed in Table 2.8-1 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.Not used.
7.
The safety-related relief valves listed in Table 2.8-1 provide overpressure protection.
i.
A vendor test will be performed of each safety-related relief valves listed in Table 2.8-1.
ii.
An inspection will be performed of each safety-related as-built relief valves listed in Table 2.8-1.
i.
An American Society of Mechanical Engineers Code Section III Data Report exists and concludes that the relief valves listed in Table 2.8-1 meet the 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 condensers have the capacity to transfer their design heat load.Not used.
A type test or a combination of type test and analysis will be performed of the safety-related DHRS passive condensers.Not used.
A report exists and concludes that the safety-related DHRS passive condensers listed in Table 2.8-1 have a heat removal capacity sufficient to transfer their design heat load.Not used.
9.
The 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, AOOs, DBAs, and postaccident conditions and performs its function for the period of time required to complete the function.
- i. A type test or a combination of type test and analysis will be performed of the CNTS containment electrical penetration assemblies equipmentlisted in Table 2.8-1 including associated connection assemblies.
ii. An inspection will be performed of the containment CNTS electrical penetration assembles listed in Table 2.8-1, including associated connection assemblies.
- i. An EQ record form exists and concludes that the CNTS electrical penetration assemblies listed in Table 2.8-1, including associated connection assemblies, performs their function under the environmental conditions specified in the EQ record form for the period of time required to complete the function.
ii. The CNTS electrical penetration 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.
Table 2.8-2: Equipment Qualification Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Control Room Habitability Tier 1 3.1-1 Draft Revision 3 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.
NuScale Tier 1 Control Room Habitability Tier 1 3.1-3 Draft Revision 3 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 meetsdoes not exceed the assumptions used to size the CRHS inventory and the supply flow rate.
A test will be performed of the CRE.
The air exfiltration measured by tracer gas testing is less than the CRE air infiltration rate assumed in the dose analysis.
2 The CRHS valves listed in Table 3.1-1 change position under design basis temperature, differential pressure, and flow conditions.
A test will be performed of the CRHS valves listed in Table 3.1-1 under preoperational temperature, differential pressure, and flow conditions.
Each CRHS valve listed in Table 3.1-1 strokes fully open and fully closed by remote operation under preoperational temperature, differential pressure, and flow conditions.
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.
A test will be performed of the CRHS solenoid-operated valves listed in Table 3.1-1 under preoperational temperature, differential pressure and flow conditions.
Each CRHS solenoid-operated valve listed in Table 3.1-1 performs its function to fail open on loss of motive power under preoperational temperature, differential pressure, and flow conditions.
4 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 DBA.
An analysis will be performed of the as-built CRE heat sinks.
A report exists and concludes that 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 DBA.
5 The CRHS maintains a positive pressure in the MCR relative to adjacent areas.
A test will be performed of the CRHS.
The CRHS maintains a positive pressure of greater than or equal to 1/8 inches water gauge in the CRE relative to adjacent areas, while operating in DBA alignment.
NuScale Tier 1 Normal Control Room Heating Ventilation and Air Conditioning System Tier 1 3.2-1 Draft Revision 3 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.
NuScale Tier 1 Normal Control Room Heating Ventilation and Air Conditioning System Tier 1 3.2-3 Draft Revision 3 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 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.
A test will be performed of the air-operated CRE isolation dampers listed in Table 3.2-1 under preoperational temperature, differential pressure and flow conditions.
Each CRVS air-operated CRE isolation damper listed in Table 3.2-1 performs its function to fail to the closed position on loss of motive power under preoperational temperature, differential pressure, and flow conditions.
2 The CRVS maintains a positive pressure in the CRB relative to the outside environment.
A test will be performed of the CRVS while operating in the normal operating alignment.
The CRVS maintains a positive pressure of greater than or equal to 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 concentration levels in the CRB battery rooms containing batteries below one percent by volume.
A test will be performed of the CRVS while operating in the normal operating alignment.
The airflow capability of the CRVS maintains the hydrogen concentration levels in the CRB battery rooms containing batteries below one percent by volume.
NuScale Tier 1 Reactor Building Heating Ventilation and Air Conditioning System Tier 1 3.3-1 Draft Revision 3 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.
NuScale Tier 1 Fuel Handling Equipment System Tier 1 3.4-1 Draft Revision 3 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.
NuScale Tier 1 Fuel Handling Equipment System Tier 1 3.4-2 Draft Revision 3 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 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 FHM main and auxiliary hoists are single-failure-proof in accordance with the approved design.
An inspection will be performed of the as-built FHM main and auxiliary hoists.
The FHM main and auxiliary hoists are single-failure-proofA report exists and concludes that the FHM main and auxiliary hoists are single-failure-proof in accordance with the approved design.
2.
The FHM main hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.
A rated load test will be performed of the FHM main hoist.
The FHM main hoist lifts, supports, holds with the brakes, and transports a load of at least 125 percent of the manufacturers rated capacity.
3.
The FHM auxiliary hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.
A rated load test will be performed of the FHM auxiliary hoist.
The FHM auxiliary hoist lifts, supports, holds with the brakes, and transports a load of at least 125 percent of the manufacturers rated capacity.
4.
Single-failure-proofThe FHM welds are inspectedcomply with the American Society of Mechanical Engineers NOG-1 Code.
An inspection will be performed of the as-built FHM welds.
The results of the non-destructive examination of the FHM welds comply with American Society of Mechanical Engineers NOG-1 Code.
5.
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.
A test will be performed of the FHM gripper mast limit switches.
The FHM maintains at least 10 feet of water above the top of the fuel assembly when lifted to its maximum height with the pool level at the lower limit of the normal operating low water level.
6.
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.
A test will be performed of new fuel jib crane interlocks.
The new fuel jib crane interlocks prevent the crane from carrying a fuel assembly over the spent fuel racks.
NuScale Tier 1 Fuel Storage System Tier 1 3.5-1 Draft Revision 3 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.
NuScale Tier 1 Fuel Storage System Tier 1 3.5-2 Draft Revision 3 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 Class NF components conform to the rules of construction of ASME Code Section III.
An inspection will be performed of the fuel storage system ASME Code Class NF as-built component Data Reports required by ASME Code Section III.
ASME Code Section III Data Reports for the fuel storage system ASME Code Class NF fuel storage racks exist and conclude that the requirements of ASME Code Section III are met.
2 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:
- 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.
An inspection will be performed of the as-built fuel storage racks, their configuration in the SFP, and the associated documentation.
The as-built fuel storage racks, including any neutron absorbers, and their configuration within the SFP conform to the design values for materials and dimensions and their tolerances, as shown to be acceptable in the approved fuel storage criticality analysis described in the UFSAR.
NuScale Tier 1 Ultimate Heat Sink Tier 1 3.6-2 Draft Revision 3 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.
NuScale Tier 1 Ultimate Heat Sink Tier 1 3.6-4 Draft Revision 3 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 Class 3 piping system listed in Table 3.6-1 complies with ASME Code Section III requirements.
An inspection will be performed of the ultimate heat sink ASME Code Class 3 as-built piping system listed in Table 3.6-1 Design Report required by ASME Code Section III.
The ASME Code Section III Design Report (NCA-3550) exists and concludes that the ultimate heat sink ASME Code Class 3 as-built piping system listed in Table 3.6-1 meets the requirements of ASME Code Section III.
2 The UHS Code Class 3 components listed in Table 3.6-1 conform to the rules of construction of ASME Code Section III.
An inspection will be performed of the UHS ASME Code Class 3 as-built component Data Report for the components listed in Table 3.6-1 required by ASME Code Section III.
The ASME Code Section III Data Report for the UHS ASME Code Class 3 components listed in Table 3.6-1 and interconnecting piping exists and concludes that the requirements of ASME Code Section III are met.
23 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.
An inspection will be performed of the as-built SFP, RFP, reactor pool and dry dock piping and connections.
There are no gates, openings, drains, or piping within the SFP, RFP, reactor pool, and dry dock that are below 80 ft building elevation (55 ft pool level) as measured from the bottom of the SFP and reactor pool.
NuScale Tier 1 Fire Protection System Tier 1 3.7-1 Draft Revision 3 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.
NuScale Tier 1 Fire Protection System Tier 1 3.7-3 Draft Revision 3 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 provide a dedicated volume of water for firefighting.
An inspection will be performed of the as-built firewater storage tanks.
Each firewater storage tank provides a usable water volume dedicated for firefighting that is greater than or equal to 300,000 gallons.
2 The FPS has a sufficient number of fire pumps to provide the design flow requirements to satisfy 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.
i.
An analysis will be performed of the as-built fire pumps.
ii.
A test will be performed of the fire pumps.
i.
A report exists and concludes that the fire pumps can 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.
Each fire pump delivers the design flow to the FPS, while operating in the fire-fighting alignment.
3 Safe-shutdown can be achieved assuming that all equipment in any one fire area (except for the 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.
A safe-shutdown analysis of the as-built plant will be performed, including a post-fire safe-shutdown circuit analysis.
A safe-shutdown analysis report exists and concludes that:
- Safe-shutdown can be achieved assuming that all equipment in any one fire area (except for the MCR and under the bioshield) 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 independent of the MCR exists.
4 A plant FHA considers potential fire hazards and ensures the fire protection features in each fire area are suitable for the hazards.
A FHA of the as-built plant will be performed.
A FHA report exists and concludes that:
- Combustible loads and ignition sources are accounted for, and
- Fire protection features are suitable for the hazards they are intended to protect against.
NuScale Tier 1 Plant Lighting System Tier 1 3.8-1 Draft Revision 3 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.
NuScale Tier 1 Plant Lighting System Tier 1 3.8-2 Draft Revision 3 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 of the operator workstations and auxiliary panels in the MCR and operator workstations in the RSS.
i.A test will be performed of the MCR operator workstations and auxiliary panel illumination.
ii.A test will be performed of the RSS operator workstations illumination.
i.The PLS provides at least 100 foot-candles illumination at the MCR operator workstations and at least 50 foot-candles at the auxiliary panels.
ii.The PLS provides at least 100 foot-candles illumination at the RSS operator workstations.
2 The PLS provides emergency illumination of the operator workstations and auxiliary panels in the MCR and operator workstations in the RSS.
i.A test will be performed of the MCR operator workstations and auxiliary panel illumination.
ii.A test will be performed of the RSS operator workstations illumination.
i.The PLS provides at least 10 foot-candles of illumination at the MCR operator workstations and auxiliary panels when it is the only MCR lighting 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 lighting fixtures provide illumination for post-FSSD activities performed by operators outside the MCR and RSS where post-FSSD activities are performed.
A test will be performed of the eight-hour battery-pack emergency lighting fixtures.
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-FSSD activities are performed.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 Tier 1 3.9-1 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 Tier 1 3.9-2 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 Tier 1 3.9-6 Draft Revision 3 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 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.
A test will be performed of the CRVS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated CRVS high-radiation signals upstream of the CRVS filter unit listed in Table 3.9-1, the CRVS automatically aligns/actuates the identified components to the positions identified in the table.
2 The CRVS and the CRHS automatically respond to athe 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.
A test will be performed of the CRVS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated CRVS high-radiation signals downstream of the CRVS filter unit listed in Table 3.9-1, the CRVS and the CRHS automatically align/actuate the identified components to the positions identified in the table.
3 The 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.
A test will be performed of the RBVS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated RBVS high-radiation signals listed in Table 3.9-1, the RBVS automatically aligns/actuates the identified components to the positions identified in the table.
4 The 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.
A test will be performed of the GRWS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated GRWS high-radiation signals listed in Table 3.9-1, the GRWS automatically aligns/actuates the identified components to the positions identified in the table.
5 The GRWS automatically responds to a high-radiation signal from 00-GRW-RIT-0060 to mitigate a release of radioactivityNot Used.
A test will be performed of the GRWS high-radiation signalsNot Used.
Upon initiation of a real or simulated GRWS high-radiation signals listed in Table 3.9-1, the GRWS automatically aligns/actuates the identified components to the positions identified in the tableNot Used.
6 The GRWS automatically responds to a high-radiation signal from 00-GRW-RIT-0071 to mitigate a release of radioactivityNot Used.
A test will be performed of the GRWS high-radiation signalsNot Used.
Upon initiation of a real or simulated GRWS high-radiation signals listed in Table 3.9-1, the GRWS automatically aligns/actuates the identified components to the positions identified in the tableNot Used.
7 The 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.
A test will be performed of the LRWS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated LRWS high-radiation signals listed in Table 3.9-1, the LRWS automatically aligns/actuates the identified components to the positions identified in the table.
8 The 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.
A test will be performed of the ABS high-radiation signals listed in Table 3.9-1.
Upon initiation of athe real or simulated ABS high-radiation signals listed in Table 3.9-1, the ABS automatically aligns/actuates the identified components to the positions identified in the table.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 12 Tier 1 3.9-7 Draft Revision 3 9
The ABS automatically responds to a high-radiation signal from 00-AB-RT-0166 to mitigate a release of radioactivityNot Used.
A test will be performed of the ABS high-radiation signalNot Used.
Upon initiation of a real or simulated ABS high-radiation signal listed in Table 3.9-1, the ABS automatically aligns/actuates the identified components to the positions identified in the tableNot Used.
10 The 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.
A test will be performed of the PSCS high-radiation signal listed in Table 3.9-1.
Upon initiation of a real or simulated PSCS high-radiation signal listed in Table 3.9-1, the PSCS automatically aligns/actuates the identified components to the positions identified in the table.
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
NuScale Tier 1 Reactor Building Crane Tier 1 3.10-1 Draft Revision 3 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.
NuScale Tier 1 Reactor Building Crane Tier 1 3.10-2 Draft Revision 3 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.
NuScale Tier 1 Reactor Building Crane Tier 1 3.10-3 Draft Revision 3 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 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.
An inspection will be performed of the as-built RBC main hoist.
The RBC main hoist is single-failure-proofA report exists and concludes that the RBC main hoist is single-failure-proof in accordance with the approved design.
2 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.
An inspection will be performed of the as-built RBC auxiliary hoists.
The RBC auxiliary hoists are single-failure-proofA report exists and concludes that the RBC auxiliary hoists are single-failure-proof in accordance with the approved design.
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.
An inspection will be performed of the as-built RBC wet hoist.
The RBC wet hoist is single-failure-proofA report exists and concludes that the RBC wet hoist is single-failure-proof in accordance with the approved design.
4 The RBC main hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.
A rated load test will be performed of the RBC main hoist.
The RBC main hoist lifts, supports, holds with the brakes, and transports a load of at least 125 to 130 percent of the manufacturers rated capacity.
5 The RBC auxiliary hoists are capable of lifting and supporting their rated load, holding the rated load, and transporting the rated load.
A rated load test will be performed of the RBC auxiliary hoists.
The RBC auxiliary hoists lift, support, hold with the brakes, and transport a load of at least 125 to 130 percent of the manufacturers rated capacity.
6 The RBC wet hoist is capable of lifting and supporting its rated load, holding the rated load, and transporting the rated load.
A rated load test will be performed of the RBC wet hoist.
The RBC wet hoist lifts, supports, holds with the brakes, and transports a load of at least 125 to 130 percent of the manufacturers rated capacity.
7 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.
An inspection will be performed of the as-built RBC weld joints whose failure could result in the drop of a critical load.
The results of the non-destructive examination of the RBC welds joints whose failure could result in the drop of a critical load comply with American Society of Mechanical Engineers NOG-1 Code.
8 Load path RBC wet hoist welds are inspectedNot Used.
An inspection will be performed of the as-built RBC wet hoistNot Used.
The results of the non-destructive examination of the RBC wet hoist welds comply with American Society of Mechanical Engineers NOG-1 CodeNot Used.
NuScale Tier 1 Reactor Building Crane Tier 1 3.10-4 Draft Revision 3 9
The MLA is capable of supporting its rated load.
i.
A rated load test will be performed of the MLA single load path elements.
ii. A rated load test will be performed of the MLA dual load path elements.
i.
The MLA single load path elements support a load of at least 300 percent of the manufacturer's rated capacity.
ii.
The MLA dual load path elements support a load of at least 150 percent of the manufacturer's rated capacity.
10 The MLA is single-failure-proof in accordance with the approved design.
An inspection will be performed of the as-built MLA.
A report exists and concludes that the MLA is single-failure-proof in accordance with the approved design.
Table 3.10-1: Reactor Building Crane Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Reactor Building Tier 1 3.11-2 Draft Revision 3 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.
NuScale Tier 1 Reactor Building Tier 1 3.11-7 Draft Revision 3 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 confinement so that the impact from internal fires, smoke, hot gases,or fire suppressants is contained within the RXB fire area of origin.
An inspection will be performed of the RXB as-built fire and smoke barriers.
The following RXB fire and smoke barriers exist in accordance with the fire hazards analysis, and have been qualified for the fire rating specified in 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 confinement so that the impact from internal flooding is contained within the RXB flooding area of origin.
An inspection will be performed of the RXB as-built internal flooding barriers.
The following RXB internal flooding barriers exist in accordance with the internal flooding analysis report and 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 protected against external flooding in order to prevent flooding of safety-related SSC within the structure.
An inspection will be performed of the RXB as-built floor elevation at ground entrances.
The RXB floor elevation at ground entrances is higher than the maximum external flood elevation.
4 The RXB includes radiation shielding barriers for normal operation and post-accident radiation shielding.
An inspection will be performed of the as-built RXB radiation shielding barriers.
The thickness of RXB radiation shielding barriers is greater than or equal to the required thickness specified in Table 3.11-1.
5 The RXB includes radiation attenuating doors for normal operation and for post-accident radiation shielding. These doors have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed.
An inspection will be performed of the as-built RXB radiation attenuating doors.
The RXB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.
6 The RXB is Seismic Category I and maintains its structural integrity under the design basis loads.
i.
An inspection and analysis will be performed of the as-built RXB.
ii.
An inspection will be performed of the as-built RXB.
i.
A design report exists and concludes that the deviations between the drawings used for 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.
NuScale Tier 1 Reactor Building Tier 1 3.11-8 Draft Revision 3 7
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 SSE.
An inspection and analysis will be performed of the as-built non-Seismic Category I SSC located where there is a potential for adverse interaction with the RXB or a Seismic Category I SSC in the RXB.
A report exists and concludes that the 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 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 against the dynamic and environmental effects associated with postulated failures in high-and moderate-energy piping systems.
An inspection and analysis will be performed of the as-built high-and moderate-energy piping systems and protective features for the safety-related SSC located in the RXB outside the Reactor Pool Bay.
Protective features are installed in accordance with the as-built Pipe Break Hazard Analysis Report and safety-related SSC are protected against or qualified to withstand the dynamic and environmental effects associated with postulated failures in high-and moderate-energy piping systems.
Table 3.11-2: Reactor Building Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Radioactive Waste Building Tier 1 3.12-4 Draft Revision 3 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 barriers for normal operation and post-accident radiation shielding.
An inspection will be performed of the as-built RWB radiation shielding barriers.
The thickness of RWB radiation shielding barriers is greater than or equal to the required thickness specified in Table 3.12-1.
2 The RWB includes radiation attenuating doors for normal operation and for post-accident radiation shielding. These doors have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed.
An inspection will be performed of the as-built RWB radiation attenuating doors.
The RWB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.
3 The RWB is an RW-IIa structure and maintains its structural integrity under the design basis loads.
An inspection and analysis will be performed of the as-built RW-IIa RWB.
A design report exists and concludes that the deviations between the drawings used for construction and the as-built RW-IIa RWB have been reconciled and that the as-built RW-IIa RWB maintains its structural integrity under the design basis loads.
NuScale Tier 1 Control Building Tier 1 3.13-3 Draft Revision 3 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 confinement so that the impact from internal fires, smoke, hot gases, or fire suppressants is contained within the CRB fire area of origin.
An inspection will be performed of the CRB as-built fire and smoke barriers.
The following CRB fire and smoke barriers exist in accordance with the fire hazards analysis, and have been qualified for the fire rating specified in 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 confinement so that the impact from internal flooding is contained within the CRB flooding area of origin.
An inspection will be performed of the CRB as-built internal flooding barriers.
The following CRB internal flooding barriers exist in accordance with the internal flooding analysis report and 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 protected against external flooding in order to prevent flooding of safety-related SSC within the structure.
An inspection will be performed of the CRB as-built floor elevation at ground entrances.
The CRB floor elevation at ground entrances is higher than the maximum external flood elevation.
4 The CRB at Elevation 120-0 and below (except for the elevator shaft, the stairwells, and the fire protection vestibule which are Seismic Category II) and below is Seismic Category I and maintains its structural integrity under the design basis loads.
i.
An inspection and analysis will be performed of the as-built CRB.
ii.
An inspection will be performed of the as-built CRB at Elevation 120-0 and below.
i.
A design summary report exists and concludes that the deviations between the drawings used for construction and the as-built CRB have been reconciled, and the CRB at Elevation 120-0 and below (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.
NuScale Tier 1 Equipment Qualification - Shared Equipment Tier 1 3.14-1 Draft Revision 3 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
NuScale Tier 1 Equipment Qualification - Shared Equipment Tier 1 3.14-2 Draft Revision 3 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.
NuScale Tier 1 Equipment Qualification - Shared Equipment Tier 1 3.14-6 Draft Revision 3 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 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 common, safety-related equipment, and common, nonsafety-related equipment that provides one of the following nonsafety-related functions:
- Provides physical support of irradiated fuel (fuel handling machine, spent fuel storage racks, reactor building crane, and module lifting adaptor)
- Provides a path for makeup water to the UHS
- Provides containment of UHS water
- Monitors UHS water level i.
A type test, analysis, or a combination of type test and analysis will be performed of the common Seismic Category I equipment listed in Table 3.14-1, including its associated supports and anchorages.
i.
A seismic qualification record form exists and concludes that the common Seismic Category I equipment listed in Table 3.14-1, including its associated supports and anchorages, will withstand the design basis seismic loads and perform its function during and after a safe shutdown earthquake.
ii.
An inspection will be performed of the common Seismic Category I as-built equipment listed in Table 3.14-1, including its associated supports and anchorages.
ii.
The common Seismic Category I equipment listed in Table 3.14-1, including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the equipments seismic qualification record form.
2.
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, DBA, 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 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.
i.
A type test or a combination of type test and analysis will be performed of the common electrical equipment listed in Table 3.14-1, including its connection assemblies.
i.
An equipment qualification record form exists and concludes that the common electrical equipment listed in Table 3.14-1, including its connection assemblies, performs its function under the environmental conditions specified in the equipment qualification record form for the period of time required to complete the function.
ii.
An inspection will be performed of the common as-built electrical equipment listed in Table 3.14-1, including its connection assemblies.
ii.
The common electrical equipment listed in Table 3.14-1, including its connection assemblies, is installed in its design location in a configuration bounded by the EQ record form.
3.
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.
i.
An inspection and reconciliation analysis will be performed of the as-built RW-IIa components and piping used for processing gaseous radioactive waste listed in Table 3.14-1.
i.
A report exists and concludes that the as-built RW-IIa components and piping used for processing gaseous radioactive waste listed in Table 3.14-1 meet the RW-IIa design criteria.
NuScale Tier 1 Equipment Qualification - Shared Equipment Tier 1 3.14-7 Draft Revision 3 4.
Each CNTS containment electrical penetration assembly listed in Table 2.1-3 is rated either (i) to withstand fault and overload currents for the time required to clear the fault from its power source, or (ii) to withstand the maximum fault and overload current for its circuits without a circuit interrupting device.
An analysis will be performed of each CNTS as-built containment electrical penetration assembly listed in Table 2.1-3.
For each CNTS containment electrical penetration assembly listed in Table 2.1-3, either (i) a circuit interrupting device coordination analysis exists and concludes that the current carrying capability for the CNTS containment electrical penetration assembly is greater than 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 lubricants used in module-specific mechanical equipment listed in Table 2.8-1 perform their function up to the end of their qualified life in the design basis harsh environmental conditions (both internal service conditions and external environmental conditions) experienced during normal operations, AOOs, DBAs, and post-accident conditions.
A type test or a combination of type test and analysis will be performed of the non-metallic parts, materials, and lubricants used in module-specific mechanical equipment listed in Table 2.8-1.
A qualification record form exists and concludes that 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 under the design basis harsh environmental conditions (both internal service conditions and external environmental conditions) specified in the qualification record form.
6.
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.
A type test, analysis, or a combination of type test and analysis will be performed of the Class 1E digital equipment listed in Table 2.8-1.
An EQ record form exists and concludes that the Class 1E digital equipment listed in Table 2.8-1 withstands the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA without loss of safety-related function.
7.
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.
A type test or a combination of type test and analysis will be performed of the valves listed in Table 2.8-1.
A Qualification Report exists and concludes that the valves listed in Table 2.8-1 are capable of performing their safety-related function under the full range of fluid flow, differential pressure, electrical, temperature, and fluid conditions up to and including DBA conditions.
8.
The DHRS condensers listed in Table 2.8-1 have the capacity to transfer their design heat load.
A type test or a combination of type test and analysis will be performed of the DHRS condensers listed in Table 2.8-1.
A report exists and concludes that the DHRS condensers listed in Table 2.8-1 have a heat removal capacity sufficient to transfer their design heat load.
Table 3.14-2: Equipment Qualification - Shared Equipment ITAAC (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Human Factors Engineering Tier 1 3.15-1 Draft Revision 3 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.
NuScale Tier 1 Physical Security System Tier 1 3.16-1 Draft Revision 3 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.
NuScale Tier 1 Physical Security System Tier 1 3.16-2 Draft Revision 3 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.
NuScale Tier 1 Physical Security System Tier 1 3.16-3 Draft Revision 3 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 within a vital area.
All vital equipment locations will be inspected.
Vital equipment is located only within a vital area.
2.
Access to vital equipment requires passage through at least two physical barriers.
All vital equipment physical barriers will be inspected.
Vital equipment is located within a protected area such that access to the vital equipment requires passage through at least two physical barriers.
3.
The external walls, doors, ceilings, and floors in the MCR and CAS will be bullet-resistant.
Type test, analysis, or a combination of type test and analysis of the external walls, doors, ceilings, and floors in the MCR and CAS, will be performed.
A report exists and concludes that the walls, doors, ceilings, and floors in the MCR and CAS are bullet-resistant.
4.
An access control system will be installed and designed for use by individuals who are authorized access to vital areas within the nuclear island and structures without escort.
The access control system will be tested.
The access control system is installed and provides authorized access to vital areas within the nuclear island and structures only to those individuals with authorization for unescorted access.
5.
Unoccupied vital areas within the nuclear island and structures will be designed with locking devices and intrusion detection devices that annunciate in the CAS.
Tests, inspections, or a combination of tests and inspections of unoccupied vital areas' intrusion detection equipment and locking devices will be performed.
Unoccupied vital areas within the nuclear island and structures are locked and alarmed and intrusion is detected and annunciated in the CAS.
6.
The CAS will be located inside the protected area and will be designed so that the interior is not visible from the perimeter of the protected area.
The CAS will be inspected.
The CAS is located inside the protected area, and the interior of the alarm station is not visible from the perimeter of the protected area.
7.
Security alarm devices in the RXB and CRB, including transmission lines to annunciators, will be tamper-indicating and self-checking, and alarm annunciation indicates the type of alarm and its location.
All security alarm devices and transmission lines in the RXB and CRB will be tested.
Security alarm devices, within the nuclear island and structuresin the RXB and CRB including transmission lines to annunciators, are tamper-indicating and self-checking; an 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 systems within the nuclear island and structuresin the RXB and CRB will be designed to provide visual display and audible annunciation of alarms in the CAS.
Intrusion detection and assessment systems in the RXB and CRB will be tested.
The intrusion detection systems, within the nuclear island and structuresin the RXB and CRB provide a visual display and audible annunciation of all alarms in the CAS.
9.
Intrusion detection systems' recording equipment will record security alarm annunciations within 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.
The intrusion detection systems' recording equipment in the RXB and CRB will be tested.
Intrusion detection systems' recording equipment is capable of recording each security alarm annunciation within 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.
NuScale Tier 1 Physical Security System Tier 1 3.16-4 Draft Revision 3 10.
Emergency exits through the vital area boundaries within the nuclear island and structures will be alarmed with intrusion detection devices and within the nuclear island and structures are secured by locking devices that allow prompt egress during an emergency.
Tests, inspections, or a combination of tests and inspections of emergency exits through vital area boundaries within the nuclear island and structures will be performed.
Emergency exits through the vital area boundaries within the nuclear island and structures are alarmed with intrusion detection devices and secured by locking devices that allow prompt egress during an emergency.
11.
The CAS will have a landline telephone service with the control room and local law enforcement authorities.
Tests, inspections, or a combination of tests and inspections of the CAS's landline telephone service will be performed.
The CAS is equipped with landline telephone service with the control room and local law enforcement authorities.
12.
The CAS will be capable of continuous communication with on-duty security force personnel.
Tests, inspections, or a combination of tests and inspections of the CAS's continuous communication capabilities will be performed.
The CAS is capable of continuous communication with on-duty watchmen, armed security officers, armed responders, or other security personnel who have responsibilities within the physical protection program and during contingency response events.
13.
Non-portable communications equipment in the CAS will remain operable from an independent power source in the event of the loss of normal power.
Tests, inspections, or a combination of tests and inspections of the nonportable communications equipment will be performed.
All nonportable communication devices in the CAS remain operable from an independent power source in the event of the loss of normal power.
Table 3.16-1: Physical Security System Inspections, Tests, Analyses, and Acceptance Criteria (Continued)
No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 6 Tier 1 3.17-1 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 1 - 6 Tier 1 3.17-3 Draft Revision 3 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 athe CFDS high-radiation signal from 6A-CFD-RT-1007listed in Table 3.17-1 to mitigate a release of radioactivity.
A test will be performed of the CFDS high-radiation signal listed in Table 3.17-1.
Upon initiation of a real or simulated CFDS high-radiation signal listed in Table 3.17-1, the CFDS automatically aligns/actuates the identified components to the positions identified in the table.
2.
The 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.
A test will be performed of the BPDS high-radiation signals listed in Table 3.17-1.
Upon initiation of athe real or simulated BPDS high-radiation signals listed in Table 3.17-1 the BPDS automatically aligns/actuates the identified components to the positions identified in the table.
3.
The BPDS automatically responds to a high-radiation signal from 6A-BPD-RIT-0529 to mitigate a release of radioactivity.
A test will be performed of the BPDS high-radiation signal.
Upon initiation of a real or simulated BPDS high-radiation signal listed in Table 3.17-1, the BPDS automatically aligns/actuates the identified components to the positions identified in the table.
4.
The BPDS automatically responds to a high-radiation signal from 6A-BPD-RIT-0705 to mitigate a release of radioactivity.
A test will be performed of the BPDS high-radiation signal.
Upon initiation of a real or simulated BPDS high-radiation signal listed in Table 3.17-1, the BPDS automatically aligns/actuates the identified components to the positions identified in the table.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 7 - 12 Tier 1 3.18-1 Draft Revision 3 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.
NuScale Tier 1 Radiation Monitoring - NuScale Power Modules 7 - 12 Tier 1 3.18-3 Draft Revision 3 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 athe CFDS high-radiation signal from 6B-CFD-RT-1007listed in Table 3.18-1 to mitigate a release of radioactivity.
A test will be performed of the CFDS high-radiation signal listed in Table 3.18-1.
Upon initiation of a real or simulated CFDS high-radiation signal listed in Table 3.18-1, the CFDS automatically aligns/actuates the identified components to the positions identified in the table.
2.
The 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.
A test will be performed of the BPDS high-radiation signals listed in Table 3.18-1.
Upon initiation of athe real or simulated BPDS high-radiation signals listed in Table 3.18-1, the BPDS automatically aligns/actuates the identified components to the positions identified in the table.
3.
The BPDS automatically responds to a high-radiation signal from 6B-BPD-RIT-0530 to mitigate a release of radioactivity.
A test will be performed of the BPDS high-radiation signal.
Upon initiation of a real or simulated BPDS high-radiation signal listed in Table 3.18-1, the BPDS automatically aligns/actuates the identified components to the positions identified in the table.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-33 Draft Revision 3 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 condensate polishing system (CPS) by supplying steam for resin regeneration.
nonsafety-related CPS Test #30-1 The auxiliary boiler supports the turbine generator by supplying gland seal steam.
nonsafety-related CAR Test #32-1 The auxiliary boiler supports the FWS by supplying steam to the condenser for sparging when necessary.
nonsafety-related CAR Test #32-1 The auxiliary boiler supports the module heatup system (MHS) by supplying steam for heating reactor coolant at startup and shutdown.
nonsafety-related TG Test #33-1 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-operated valve can be operated remotely.
Operate each valve from the MCR and local control panel (if design has local valve control).
MCR display and local, visual observation indicate each valve fully opens and fully closes.
ii.
Verify each auxiliary boiler air-operated valve fails to its safe position on loss of air.
Place each valve in its non-safe position.
Isolate and vent air to the valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iii. Verify each auxiliary boiler air-operated valve fails to its safe position on loss of electrical power to its solenoid.
Place each valve in its non-safe position.
Isolate electrical power to each air-operated valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iv. Verify each auxiliary boiler low pressure boiler feedwater pump can be started and stopped remotely.
Align the ABS to allow for pump operation.
Stop and start each pump from the MCR.
MCR display and local, visual observation indicate each pump starts and stops.
Audible and visible water hammer are not observed when the pump starts.
v.
Verify each auxiliary boiler high pressure boiler feedwater pump can be started and stopped remotely.
Align the ABS to allow for pump operation.
Stop and start each pump from the MCR. MCR display and local, visual observation indicate each pump starts and stops.
Audible and visible water hammer are not observed when the pump starts.
vi. Verify the speed of each auxiliary boiler high pressure boiler feedwater pump can be manually controlled.
Align the ABS to provide a flow path to operate a selected AB variable-speed pump.
Vary the auxiliary boiler pump speed from minimum to maximum speed from the MCR.
MCR display indicates the speed of each variable speed pump obtains both minimum and maximum pump speeds.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-34 Draft Revision 3 vii. Verify the ABS automatically responds to mitigate a release of radioactivity.
Initiate a real or simulated high radiation signal for the auxiliary boiler flash tank vent.
MCR display verifies the following:
i.
auxiliary boiler flash tank 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 responds to mitigate a release of radioactivity.
Initiate a real or simulated high radiation signal for the auxiliary boiler high pressure to low pressure steam supply.
MCR display verifies the following:
auxiliary boiler high pressure to low pressure steam supply pressure control valve is closed.
[ITAAC 03.09.089]
ix. Verify each ABS instrument is available on an MCS or PCS display.
(Test not required if the instrument calibration verified the MCS or PCS display.)
Initiate a single real or simulated instrument signal from each ABS transmitter.
The instrument signal is displayed on an MCS or PCS display, or is recorded by the applicable control system historian.
System Level Test Test Objective Test Method Acceptance Criteria None Table 14.2-9: Auxiliary Boiler System Test # 9 (Continued)
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-64 Draft Revision 3 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 polisher demineralizers, the three 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.
nonsafety-related Test #24-1 Test #24-7 2.
The BPDS supports the two TGBs, the two diesel generators, the auxiliary 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.
nonsafety-related Test #24-1 Test #24-7 3.
The BPDS supports the CRB floor drains by providing a means to collect, treat, and transfer the waste water to the UWS.
nonsafety-related Test #24-1 Test #24-7 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 valve can be operated remotely.
Operate each valve from the MCR and local control panel (if design has local valve control).
MCR display and local, visual observation indicate each valve fully opens and fully closes.
ii.
Verify each BPDS air-operated valve fails to its safe position on loss of air.
Place each valve in its non-safe position.
Isolate and vent air to the valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iii. Verify each BPDS air-operated valve fails to its safe position on loss of electrical power to its solenoid.
Place each valve in its non-safe position.
Isolate electrical power to each air-operated valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iv. Verify each BPDS pump can be started and stopped remotely.
Align the BPDS to allow for pump operation.
Stop and start each pump from the MCR.
MCR display and local, visual observation indicate each pump starts and stops.
Audible and visible water hammer are not observed when the pump starts.
v.
Verify the pump speed of each BPDS variable-speed pump can be manually controlled.
Vary the speed of each pump from the MCR and local control panel (if design has local pump control).
MCR display indicates the speed of each pump varies from minimum to maximum speed.
vi. Verify each BPDS instrument is available on an MCS or PCS display.
(Test not required if the instrument calibration verified the MCS or PCS display.)
Initiate a single real or simulated instrument signal from each BPDS transmitter.
The instrument signal is displayed on an MCS or PCS display, or is recorded by the applicable control system historian.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-66 Draft Revision 3 System Level Test #24-2 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds to mitigate a release of radioactivity.
Place a north chemical waste water sump pump in operation. Initiate a real or simulated high radiation signal on the 60A CPS regeneration skid waste effluent.
Repeat the test for each pump.
i.
The north chemical waste water sump pump stops.
ii.
North chemical waste collection sump to BPDS collection tank isolation valve is closed.
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 to mitigate a release of radioactivity.
Place a south chemical waste water sump pump in operation. Initiate a real or simulated high radiation signal on the 60B CPS regeneration skid waste effluent.
Repeat the test for each pump.
i.
The pump stops.
ii.
South chemical waste collection sump to BPDS collection tank isolation valve is closed.
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 to mitigate a release of radioactivity.
Place a north waste water sump pump in operation. Initiate a real or simulated high radiation signal in the BPDS north TGB floor drains.
Repeat the test for each pump.
i.
The north waste water sump pump stops.
ii.
North waste water sump discharge to BPDS collection tank isolation 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 to mitigate a release of radioactivity.
Place a south waste water sump pump in operation. Initiate a real or simulated high radiation signal in the BPDS south TGB floor drains.
Repeat the test for each pump.
i.
The south waste water sump pump stops.
ii.
South waste water sump discharge to BPDS collection tank isolation 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)
Table 14.2-24: Balance-of-Plant Drain System Test # 24 (Continued)
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-67 Draft Revision 3 System Level Test #24-6 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds to mitigate a release of radioactivity.
Place a north waste water sump pump in operation. Initiate a real or simulated high radiation signal in the BPDS auxiliary blowdown cooler condensate.
Repeat the test for each pump.
i.
The north chemical waste water sump pump stops.
ii.
North chemical waste collection sump to BPDS collection tank 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 pumps, in sumps and tanks without a fire water removal pump, start and stop automatically and transfer liquid waste to its design location.
Align each BPDS sump or tank to allow water in a selected sump or tank to be pumped to its design location. If the sump fill rate in the following test method is insufficient for automatic start of the alternate pump, the primary pump may be temporarily removed from service to allow an increase in the sump level.
- i.
Verify that Pump #1 is set to the primary pump and Pump #2 is set to alternate. Fill the selected sump or tank until a HI water level is obtained 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.
MCR displays and local, visual observation verifies the following:
- i.
The primary pump starts on HI level and transfers water to its design location in the LRWS or UWS system.
ii. The alternate pump starts on HI-HI level.
iii. Both primary and alternate pumps stop on LO level.
iv. The primary pump starts on HI level.
- v. The alternate pump starts on HI-HI level.
Table 14.2-24: Balance-of-Plant Drain System Test # 24 (Continued)
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-89 Draft Revision 3 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 receiving and / or collecting potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.
nonsafety-related Test #36-1 2.
The GRWS supports the CES by receiving and / or collecting potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.
nonsafety-related Test #36-1 CES Test #41-2 3.
The NDS supports the GRWS by providing nitrogen for purging of the GRWS.
nonsafety-related Test #36-1 NDS Test #15 component-level tests 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 valve can be operated remotely.
Operate each valve from the (main control room) MCR and local control panel (if design has local valve control).
MCR display and local, visual observation indicate each valve fully opens and fully closes.
ii.
Verify each GRWS air-operated valve fails to its safe position on loss of air.
Place each valve in its non-safe position.
Isolate and vent air to the valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iii. Verify each GRWS air-operated valve fails to its safe position on loss of electrical power to its solenoid.
Place each valve in its non-safe position.
Isolate electrical power to each air-operated valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iv. Verify GRWS valves automatically operate to maintain vessel volume.
i.
Initiate a real or simulated high GRWS moisture separator level.
ii.
Initiate a real or simulated low GRWS moisture separator level.
MCR display and local, visual observation indicate the following:
i.
The moisture separator drain valve is open.
ii.
The moisture separator drain valve is closed.
v.
Verify GRWS inlet isolation valves automatically close and nitrogen purge valve opens on high inlet stream oxygen concentration.
Simulate a GRWS inlet stream oxygen concentration high signal.
MCR display and local, visual observation indicate the following:
i.
The inlet stream isolation valves are closed.
ii.
The nitrogen purge valve is open.
vi. Verify GRWS isolates upon loss of RWBV exhaust flow.
Simulate a loss of RWBVS exhaust flow.
MCR display and local, visual observation indicate the GRWS isolation valves are closed.
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-90 Draft Revision 3 vii. Verify radiation isolation of GRWS charcoal decay beds upon detection of decay bed discharge flow high radiation level.
i.
Initiate a real or simulated GRWS train A decay bed discharge flow high radiation signal.
ii.
Initiate a real or simulated GRWS train B decay bed discharge flow high radiation signal.
MCR display and local, visual observation indicate the following:
i.
GRWS train A charcoal decay bed discharge isolation valve is closed.
[ITAAC 03.09.04]
ii.
GRWS train B charcoal decay bed discharge isolation valve is closed.
[ITAAC 03.09.045]
viii. Verify radiation isolation of GRWS discharge to the RBVS exhaust upon detection of a high radiation level.
Initiate a real or simulated GRWS discharge to the RBVS exhaust high radiation signal.
MCR display and local, visual observation indicate the GRWS discharge to the RBVS exhaust isolation valves are closed.
[ITAAC 03.09.046]
ix. Verify a local grab sample can be obtained from a GRWS grab sample device indicated on the GRWS piping and instrumentation diagram.
Place the system in service to allow flow through the grab sampling device.
A local grab sample is successfully obtained.
x.
Verify each GRWS instrument is available on an MCS or PCS display.
(Test not required if the instrument calibration verified the MCS or PCS display.)
Initiate a single real or simulated instrument signal from each GRWS transmitter.
The instrument signal is displayed on an MCS or PCS display, or is recorded by the applicable control system historian.
System Level Test #36-1 Test Objective Test Method Acceptance Criteria Verify GRWS can process a gaseous waste stream and nitrogen stream.
i.
Align GRWS to receive gaseous waste from a gaseous waste stream.
Process the gaseous waste stream through the gaseous waste process.
ii.
Align GRWS charcoal drying heater to receive nitrogen from NDS.
Process nitrogen through the charcoal drying process.
i.
The gaseous waste stream is successfully processed through the following processes:
gas cooler moisture separator charcoal drying heater charcoal guard bed charcoal decay beds RWB exhaust ii.
Nitrogen is successfully processed through the charcoal drying heater.
Table 14.2-36: Gaseous Radioactive Waste System Test # 36 (Continued)
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-94 Draft Revision 3 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 providing primary coolant makeup.
nonsafety-related Test #38-1 Ramp Change in Load Demand Test #100 2.
The CVCS supports the RCS by providing primary coolant letdown.
nonsafety-related Test #38-1 Ramp Change in Load Demand Test #100 3.
The CVCS supports the RCS by providing pressurizer spray flow for RCS pressure control.
nonsafety-related Test #38-2 Ramp Change in Load Demand Test #100 4.
The CVCS supports the RCS by changing the boron concentration of the primary coolant.
nonsafety-related Test #38-3 5.
The BAS supports the CVCS by providing uniformly mixed borated water on demand.
nonsafety-related Test #38-3 6.
The LRWS supports the CVCS by receiving and processing primary coolant from CVCS letdown.
nonsafety-related Test #38-1 LRWS Test #35-2 The CVCS functions verified by other tests are:
The CVCS supports emergency core cooling system (ECCS) valves by providing water to reset the ECCS valves.
nonsafety-related MPS Test #63-6TGS Test #33-1 The CVCS supports the RCS by heating primary coolant.
nonsafety-related TGS Test #33-1 The CVCS supports the RCS by isolating dilution sources.
safety-related MPS Test #63-6 The CVCS supports the RCS by providing primary coolant makeup in beyond design basis events.
nonsafety-related MPS Test #63-11 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 valve can be operated remotely.
Operate each valve from the MCR and local control panel (if design has local valve control).
MCR display and local, visual observation indicate each valve fully opens and fully closes.
ii.
Verify each CVCS air-operated valve fails to its safe position on loss of electrical power to its solenoid.
Place each valve in its non-safe position.
Isolate electrical power to each air-operated valve.
MCR display and local, visual observation indicate each valve fails to its safe position.
iii. Verify each CVCS air-operated valve fails to its safe position on loss of air.
Place each valve in its non-safe position.
Isolate and vent air to the valve.
MCR display and local, visual observation indicate each valve fails to its safe position
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-96 Draft Revision 3 xiii. Verify ion exchanger isolation on non-regenerative heat exchanger high outlet temperature to protect plant equipment.
Initiate a simulated high non-regenerative heat exchanger outlet temperature signal.
MCR display and local, visual observation indicate the following:
i.
CVCS purification bypass diverting 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 responds to mitigate a release of radioactivity.
Initiate a real or simulated high radiation signal for the auxiliary boiler steam flow to the 60A MHS heat exchanger.
MCR display verifies the following:
i.
CVCS module heatup system 60A 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 responds to mitigate a release of radioactivity.
Initiate a real or simulated high radiation signal for the auxiliary boiler steam flow to the 60B MHS heat exchanger.
MCR display verifies the following:
i.
CVCS module heatup system 60B 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 responds to mitigate a release of radioactivity.
Initiate a real or simulated high radiation signal for the RCS discharge flow to the regenerative heat exchanger.
MCR display verifies the following:
i.
chemical and volume control RCS 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 available on an MCS or PCS display.
(Test not required if the instrument calibration verified the MCS or PCS display.)
Initiate a single real or simulated instrument signal from each CVCS transmitter.
The instrument signal is displayed on an MCS or PCS display, or is recorded by the applicable control system historian.
Table 14.2-38: Chemical and Volume Control System Test # 38 (Continued)
NuScale Final Safety Analysis Report Initial Plant Test Program Tier 2 14.2-142 Draft Revision 3 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 normal lighting.
nonsafety-related component-level test i.
2.
The PLS supports the CRB by providing emergency lighting in the main control room.
nonsafety-related component-level test ii.
3.
The PLS supports the RXB by providing normal lighting.
nonsafety-related component-level test i.
4.
The PLS supports the RXB by providing emergency lighting for the remote shutdown station.
nonsafety-related component-level test ii.
5.
The PLS supports the RXB by providing emergency lighting for post-fire safe-shutdown activities outside of the MCR and RSS.
nonsafety-related component-level test iii.
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 illumination of the MCR and RSS operator workstations, and the MCR safety display information panel.
With normal MCR and RSS lighting in service, measure the light at each MCR and RSS workstation.
i.
a.
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.
[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 illumination of the MCR and RSS operator workstations and the MCR safety display information panel.
With MCR and RSS emergency illumination in service, measure the light at each MCR and RSS workstation and MCR safety display information panel.
ii.
a.
The PLS provides at least 10 foot-candles of illumination at the MCR operator workstations 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 emergency lighting fixtures provide illumination for post-fire safe-shutdown activities performed by operators outside the MCR and RSS.
With no AC power available, measure the light at each eight-hour battery pack emergency lighting fixture target area.
iii. The required target areas are illuminated to provide at least one foot-candle illumination in the areas outside the MCR or RSS where post-fire safe-shutdown activities are performed.
[ITAAC 03.08.03]
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-14 Draft Revision 3 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 Hazard Radiological PRA & Severe Accident FP 02.01.01 NPM As required by ASME Code Section III NCA-1210, each ASME Code 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 support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components 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.
X
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-21 Draft Revision 3 02.01.12 NPM Section 5.3.1.6, Material Surveillance, discusses the use of specimen 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.
X 02.01.13 NPM The CNTS remotely operated CNTS containment isolation valves 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 demonstrates that the CNTS remotely operated CNTS containment 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.
X 02.01.14 NPM The emergency core cooling system (ECCS) safety-related valves 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 demonstrates that the ECCS safety-related 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.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-24 Draft Revision 3 02.01.21 NPM The CNTS safety-related check valves are tested to demonstrate the 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.
X 02.01.22 NPM The CNTS electrical penetrations listed in Tier 2 Table 2.1-3 may be one of two types, one with or without a circuit interrupting device.
An ITAAC confirms that each type of penetration is evaluated to confirm it can withstand its maximum fault current.
A circuit interrupting device coordination analysis confirms and concludes in a report that the as-built containment electrical penetration assembly listed in Tier 1 Table 2.1-3 that has a circuit interrupting device can withstand fault currents for the time required to clear the fault from its power source.
8.3.1.2.5 Containment Electrical Penetration Assemblies discusses electrical penetration assemblies that are not equipped with 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 containment penetration without a circuit interrupting device confirms and concludes in a report that the maximum fault current is less than the current carrying capability of the CNTS containment electrical penetration.Not used.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-53 Draft Revision 3 02.05.27 MPS Section 7.0.4.1.2, Reactor Trip System, discusses the arrangement of 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.
X 02.05.28 MPS Section 7.1.5.1, Application of NUREG/CR-6303 Guidelines, 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 Division II of RTS and ESFAS.Not used.
X 02.05.29 MPS Section 7.1.3.3, Redundancy in Nonsafety I&C System Design, 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.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-57 Draft Revision 3 02.07.02 Section 11.5.2.2.11, Chemical and Volume Control System, 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.
X 02.07.03 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation 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 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 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.
X 02.07.04 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation 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 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 6B-AB-RT-0141.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-60 Draft Revision 3 02.08.03 EQ Section 3.11 presents information to demonstrate that the non-metallic parts, materials, and lubricants used in mechanical equipment located in a harsh environment are qualified using a type test or a combination of type test and analysis to perform their function up to the end of their qualified life in design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions. Environmental conditions include both internal service conditions and external environmental conditions for the 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 performs a credited function in Chapter 15 analyses (secondary main steam isolation valves (MSIV), feedwater regulating valves (FWRV) and secondary feedwater check valves).
The ITAAC verifies that: (1) an equipment qualification record form or ASME QME-1 report exists for the non-metallic parts, materials, and lubricants used in mechanical equipment designated for a harsh environment, and (2) the qualification record form concludes that the non-metallic parts, materials, and lubricants used in mechanical equipment listed in Tier 1 Table 2.8-1 perform their intended function up to the end of its qualified life under the design basis environmental conditions (both internal service conditions and external environmental conditions) specified in the qualification record form.Not used.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-62 Draft Revision 3 02.08.05 EQ Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment, presents information to demonstrate that the Class 1E digital equipment is qualified using a type test, analysis, or a combination of type test and analysis to perform its safety-related function when subjected to electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA. The qualification method employed for Class 1E digital equipment is the same as the qualification method described for that type of equipment in Section 3.11.
The ITAAC verifies that: (1) an equipment qualification record form exists for the Class 1E digital equipment listed in Tier 1 Table 2.8-1, and (2) the equipment qualification record form concludes that the Class 1E digital equipment withstands the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA without loss of safety-related function.Not used.
X 02.08.06 EQ Section 3.9.6.1, Functional Design and Qualification of Pumps, Valves, and Dynamic Restraints, and Section 3.10.2, Methods and Procedures for Qualifying Mechanical and Electrical Equipment and Instrumentation, discuss that the functional qualification of safety-related valves is performed in accordance with ASME QME-1-2007(or later edition), as accepted in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.9.6.1. The qualification method employed for the valves agrees with the qualification method described in Section 3.10.2.
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 range of fluid flow, differential pressure, electrical conditions, temperature conditions, and fluid conditions up to and including DBA conditions.Not used.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-63 Draft Revision 3 02.08.07 EQ Section 3.9.3.2, Design and Installation of Pressure Relief Devices, 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.
X 02.08.08 EQ Section 5.4.2, Decay Heat Removal System, discusses that the DHRS passive condensers provide the safety-related function of transferring their design heat load from the DHRS during shutdown. After manufacture of the DHRS passive condensers, a type test or a combination of type test and analysis is performed to 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 capabilityof 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.Not used.
X Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-65 Draft Revision 3 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 Hazard Radiological PRA & Severe Accident FP 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, 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 for the dose analysis. Tracer gas testing in accordance with ASTM E741 will be performed to measure the unfiltered in-leakage into the CRE with the control room habitability system (CRHS) operating.
X 03.01.02 CRH The CRHS valves 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 14.2-18, 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 design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.
X
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-74 Draft Revision 3 03.07.01 FP Section 9.5.1.2.6, Fire Protection Design Features, discusses how the 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.
X 03.07.02 FP Section 9.5.1, Fire Protection Program, discusses how the capacity of 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 design flow of the fire pumps.
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.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-75 Draft Revision 3 03.07.03 FP Section 9.5.1 discusses that (a) 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, (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.
The safe shutdown analysis will verify that:
- 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 independent of the MCR exists.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-76 Draft Revision 3 03.07.04 FP Appendix 9A, Fire Hazards Analysis, discusses the methodology and 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.
X 03.08.01 PL Section 9.5.3, Lighting Systems, discusses the plant lighting system (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 of illumination to the MCR seated operator stations and 50 foot-candles of illumination to the MCR primary operating areas 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.
X X
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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-77 Draft Revision 3 03.08.02 PL Section 9.5.3 discusses the PLS which provides emergency 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.
X X
03.08.03 PL Section 9.5.3 discusses the use of eight-hour battery pack emergency 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 lighting for:
- 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.
X X
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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-79 Draft Revision 3 03.09.04 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the 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.
X 03.09.05 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the 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 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.
X 03.09.06 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses the 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 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.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-80 Draft Revision 3 03.09.07 RM Section 11.5.2.1.5, Liquid Radioactive Waste System, discusses the 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.
X 03.09.08 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation 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-RT-0153AB-RIT-1017 and AB-RIT-1017.
X 03.09.09 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation 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-RT-0166Not Used.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-84 Draft Revision 3 03.10.07 RBC Section 9.1.5 discusses that the single failure-proof RBC is 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).
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).
X 03.10.08 RBC Section 9.1.5 discusses that the single failure-proof RBC wet hoist is 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).
An ITAAC inspection is performed to verify that the ASME Type I as-built RBC 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 wet hoist (at the factory or later)Not Used.
X 03.10.09 RBC Section 9.1.5.2.2 discusses that the MLA is a single-failure-proof 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.
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 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).
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-91 Draft Revision 3 03.11.07 RXB Section 3.2.1, Seismic Classification, discusses that per RG 1.29, some 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 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.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-102 Draft Revision 3 03.14.03 EQ The classification of SSC that contain radioactive waste in accordance 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.
An ITAAC inspection and reconciliation analysis is performed of the 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.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-103 Draft Revision 3 03.14.04 EQ The CNTS electrical penetrations listed in Tier 1 Table 2.1-3 may be one of two types, one with or without a circuit interrupting device.
An ITAAC confirms that each type of penetration is evaluated to confirm it can withstand its maximum fault current.
A circuit interrupting device coordination analysis confirms and concludes in a report that the as-built containment electrical penetration assembly listed in Tier 1 Table 2.1-3 that has a circuit interrupting device can withstand fault currents for the time required to clear the fault from its power source.
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 containment penetration without a circuit interrupting device confirms and concludes in a report that the maximum fault current is less than the current carrying capability of the CNTS containment electrical penetration.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-104 Draft Revision 3 03.14.05 EQ Section 3.11 presents information to demonstrate that the non-metallic parts, materials, and lubricants used in mechanical equipment located in a harsh environment are qualified using a type test or a combination of type test and analysis to perform their function up to the end of their qualified life in design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions. Environmental conditions include both internal service conditions and external environmental conditions for the nonmetallic parts, materials, and lubricant. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.
The scope of equipment for this design commitment is module-specific, safety-related mechanical equipment, and 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).
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.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-105 Draft Revision 3 03.14.06 EQ Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment, presents information to demonstrate that the Class 1E digital equipment is qualified using a type test, analysis, or a combination of type test and analysis to perform its safety-related function when subjected to electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA. The qualification method employed for Class 1E digital equipment is the same as the qualification method described for that type of equipment in Section 3.11.
The ITAAC verifies that: (1) an equipment qualification record form exists for the Class 1E digital equipment listed in Tier 1 Table 2.8-1, and (2) the equipment qualification record form concludes that the Class 1E digital equipment withstands the design basis electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA without loss of safety-related function.
X 03.14.07 EQ Section 3.9.6.1, Functional Design and Qualification of Pumps, Valves, and Dynamic Restraints, and Section 3.10.2, Methods and Procedures for Qualifying Mechanical and Electrical Equipment and Instrumentation, discuss that the functional qualification of safety-related valves is performed in accordance with ASME QME-1-2007 (or later edition), as accepted in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.9.6.1. The qualification method employed for the valves agrees with the qualification method described in Section 3.10.2.
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 range of fluid flow, differential pressure, electrical conditions, temperature conditions, and fluid conditions up to and including DBA conditions.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-106 Draft Revision 3 03.14.08 EQ Section 5.4.2, Decay Heat Removal System, discusses that the DHRS passive condensers provide the safety-related function of transferring their design heat load from the DHRS during shutdown.
After manufacture of the DHRS passive condensers, a type test or a combination of type test and analysis is performed to 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.
X 03.15.01 HFE Section 18.11, Design Implementation, describes the implementation of HFE aspects of the plant design.
The Design Implementation activities verify that the final MCR is 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.
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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-114 Draft Revision 3 03.17.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the 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.
X 03.17.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the 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 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.
X 03.17.04 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the 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 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-0705.
X 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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP
NuScale Final Safety Analysis Report Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Tier 2 14.3-115 Draft Revision 3 03.18.01 RM Section 11.5.2.2.9, Containment Flooding and Drain System, 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.
X 03.18.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the 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 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.
X 03.18.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the 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 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.
X 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.
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) (Continued)
ITAAC No.
System Discussion DBA Internal/External Hazard Radiological PRA & Severe Accident FP