ML18295A787

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LLC Response to NRC Request for Additional Information No. 499 (Erai No. 9564) on the NuScale Design Certification Application
ML18295A787
Person / Time
Site: NuScale
Issue date: 10/22/2018
From: Wike J
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-1018-62231
Download: ML18295A787 (26)
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Text

RAIO-1018-62231 October 22, 2018 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.

499 (eRAI No. 9564) on the NuScale Design Certification Application

REFERENCE:

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

499 (eRAI No. 9564)," dated August 21, 2018 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 Questions from NRC eRAI No. 9564:

  • 05.04.02.01-15
  • 05.04.02.01-16
  • 05.04.02.01-17
  • 05.04.02.01-18
  • 05.04.02.01-19 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, Ji):;tu~

Manager, Licensing NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Bruce Bavol, NRC, OWFN-8G9A Enclosure 1: NuScale Response to NRC Request for Additional Information eRAI No. 9564 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-1018-62231 :

NuScale Response to NRC Request for Additional Information eRAI No. 9564 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9564 Date of RAI Issue: 08/21/2018 NRC Question No.: 05.04.02.01-15 In Question 05.04.02.01-6 in Request for Additional Information (RAI) 9231, the NRC staff requested clarification of the design requirements for the steam generator (SG) tube support structures and how they are addressed in the Tiers 1 and 2 materials of the final safety analysis report (FSAR). In the response to Question 05.04.02.01-6 dated February 12, 2018 (ADAMS Accession No. ML18043B174), NuScale stated that the SG tube supports are classified as internal structures and are constructed in accordance with ASME Code,Section III, Subsection NG as a guide. The response further states that NuScale is applying the requirements in ASME Code,Section III, Subarticles NG-2400, NG-3200, NG- 4300, and NG-4400 to the SG tube support design and fabrication. Additionally, the response stated that NuScale is applying the requirements in ASME Code,Section III, Article NG-5000 as follows: surface examinations (liquid penetrant examinations) will be performed for major joint welds in the SG tube supports, and visual inspections during fabrication will be performed for other welds.

The NRC staff observed in the response to 05.04.02.01-6 in RAI 9231 that NuScale now refers to the lower and upper SG supports as "SG supports" and the SG tube support assemblies that span the full height of the SG tube bundle that include the tabs as "SG tube supports." The NRC staff uses this new terminology in this RAI.

The response to 05.04.02.01-6 in RAI 9231 did not provide sufficient information regarding the meaning of the phrase "as a guide" when describing how ASME Code,Section III, Subsection NG is applied to the SG supports and SG tube supports in the NuScale design. The NRC staff notes that secondary components of current SGs in the United States, including SG tube support structures, are designed in accordance with ASME Code,Section III, Subsections NB, NC, or NG. No current SGs in the United States use ASME Code,Section III, Subsection NG "as a guide." Therefore, it is unclear why the SG supports and SG tube supports would not be designed in accordance with ASME Code,Section III, Subsections NB, NC, or NG without NuScale Nonproprietary

exception given that SG tube integrity is no less important for the NuScale SG design than it is for current SGs in the United States. If the SG supports and SG tube supports were to fail they could impact the integrity of the SG tubes which are a part of the reactor coolant pressure boundary (RCPB). The use of ASME Code,Section III, Subsections NB, NC, or NG provides the staff a reasonable basis to conclude, in accordance with General Design Criterion (GDC) 14, that the probability of abnormal RCPB leakage from the failure of secondary components is an extremely low probability event.

To ensure that the SG supports and SG tube supports are designed to meet the requirements of GDC 1, 4, 14, 15, and 31 in Appendix A of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR), as they relate to maintaining the integrity of the RCPB, the NRC staff requests the following:

a. Revise Sections 3.2.2, 3.9.3.1.2, and 5.4.1.5 in Tier 2 of the FSAR to state that the SG supports and SG tube supports are designed, fabricated, constructed, tested, and inspected in accordance with either ASME Code,Section III, Subsection NB, NC, or NG.
b. If necessary, revise Table 3.2-1 in Tier 2 of the FSAR to reflect the classification information of the SG supports and SG tube supports.

NuScale Response:

The objective of this request for additional information (RAI) response is to provide the NRC greater clarity into the steam generator (SG) tube support design requirements. The NRC states in the request for additional information (RAI) that,"it is unclear why the SG tube supports are not designed in accordance with ASME Code,Section III, Subsections NB, NC, or NG without exception." The NuScale FSAR states that the SG supports and SG tube supports are designed using ASME Section III NG as a guide. The discussion below focuses on each of the American Society of Mechanical Engineers (ASME), Boiler Pressure Vessel Code (BPVC),

Section III subsections referenced by the NRC and explains how subsections NB and NC are not applicable, and how subsection NG is applied to the NuScale design.

The NuScale steam generator (SG) tube supports are not designed in accordance with the ASME BPVC,Section III, subsection NB, because there is no reasonable application of the ASME Code or past precedence to support such a Code classification. In general, the subsection NB Code is applied to items which form part of the reactor coolant pressure NuScale Nonproprietary

boundary (RCPB), such as pressure vessels, SG tubes, reactor coolant pumps and piping. No other design certification applications recently approved (AP1000) or currently under review (US-APWR) have classified SG tube supports as ASME subsection NB components.

The NuScale SG tube supports are not designed in accordance with ASME BPVC,Section III, subsection NC, because they are located within a subsection NB pressure vessel. In general, subsection NC is applied to items which form part of the secondary pressure boundary or are located within a secondary system. For example, the US-APWR design (currently under NRC review), has designed their SG tube supports as subsection NC. The US-APWR design is a "conventional" design, in which the SG is contained within its own, dedicated, subsection NC pressure vessel. However, as the NuScale SG tube supports are not located in a secondary pressure vessel, subsection NC was not applied.

Based on being located within a reactor vessel and because of their interface with other components which are designated as subsection NG items (such as the NuScale reactor vessel internals upper riser), it was concluded that subsection NG was the most appropriate design code to apply to the SG tube supports. The AP1000 design has also classified their SG tube supports as subsection NG components. It is noted that within the AP1000 DCD (Table 3.2-3),

that the SG tube bundle support assembly is designated as subsection NG, but is not identified as a "core support structure".

ASME Subsection NG, subarticle NG-1120, "Definition of structures and the application of these rules to them", describes two paths to apply this code; "Core support structures" or "Internal structures". Based on both the function of the NuScale SG tube supports and their location (above and remote from the core), there is no basis to designate the NuScale SG tube supports as core support structures. Therefore, as discussed in the response to 05.04.02.01-6 in RAI 9231, by the definition of the ASME Code, the SG tube supports are internal structures. Per subarticle NG-1120, for internal structures, the requirements of subsection NG only apply as stipulated. As such, NuScale (in full compliance with the ASME subsection NG Code) has stipulated the applicable requirements of subsection NG (as explained in the response to 05.04.02.01-6 in RAI 9231). Therefore, NuScale disagrees with the NRC characterization of this approach as an "exception" to subsection NG, because this approach fully meets the requirements of subsection NG. The NuScale SG tube supports are designed, constructed, tested and inspected in accordance with ASME Section III, subsection NG.

In order to convey that the SG tube supports were not core supports and therefore would be designed and fabricated in accordance with the provisions of subarticle NG-1120 for internal structures, NuScale broadly described this approach as using subsection NG, "as a guide" (since NG only applies "as stipulated" in the ASME code). The subsection NG requirements for NuScale Nonproprietary

the SG tube supports as stipulated by NuScale were provided in the response to RAI 9231-05.04.02.01-6 , and are reiterated here:

  • NG-2400
  • NG-3200
  • NG-4300 and NG-4400
  • NG-5000 surface examinations (PT examinations) for major joint welds and visual inspections for other welds (such as spacers).

In order to facilitate the clarity of the NuScale design and enhance the NRCs understanding of what (if any) different rules or requirements are applied to the SG tube supports based on their classification as "internal structures", as compared to if they were designated as "core supports," NuScale is providing the following discussion to explain the application of subsection NG to the SG tube support assembly (SG tube supports and upper and lower SG supports).

The SG tube support assembly is designed, constructed, fabricated, tested and inspected in accordance with subsection NG, in a manner identical to core supports, with the following exceptions.

The welds between the upper and lower SG supports to the RPV shell and integral steam plenum are classified as subsection NB welds and are part of the Class 1 (NB) vessel. These welds are designed, constructed, tested and inspected in accordance with ASME Subsection NB. Therefore, subsection NG does not apply to these major welds and the internal structure classification of the SG support assembly has no implications for these welds.

The SG tube supports assemblies are "internal structures" in accordance with subarticle NG-1121, therefore Code stamping of these assemblies as "core supports" per NG-8000, will not be performed.

Fabrication will be in accordance with NG-4000, except that an N-Certificate holder is not required. The SG tube support fabricator shall have a 10 CFR 50, Appendix B quality program.

Inspections are performed in accordance with NG-5000 and performed consistent with the weld quality factors (per NG-3350) used for analysis to satisfy NG-3000 stress limits. As previously indicated, the inspections consist of surface exams (PT) for segment welds in the SG tube supports and visual inspection for the spacers. However, these inspection requirements must be reconciled with quality factors used in the final analysis (not yet performed) and are subject to change as permitted by subsection NG.

NuScale Nonproprietary

One of the most important factors in SG tube design is to address fluid elastic instability (FEI).

FEI performance is strongly correlated with the design of SG tube supports, including thier damping characteristics. NuScale has stipulated FEI design requirements for the SG tube/tube support assembly as a design requirement in the applicable ASME component design specification. Likewise, design requirements have been stipulated to address vortex shedding, turbulent buffeting and flutter/gallop as it relates to the SG tube/SG tube support assembly. As discussed, these stipulations are outside the requirements of the ASME Code (but is based on guidance in ASME, non-mandatory Appendix N). In order to support these evaluations, NuScale has fabricated prototypic SG tube supports and is using them to test a full-scale SG tube bundle. This testing includes flow tests to characterize FEI, vortex shedding and turbulent buffeting. NuScale considers the stipulated flow-induced vibration (FIV) design analysis and in-progress testing as critical means of demonstrating the design of the SG tube supports fully meet all applicable General Design Criteria.

In summary, NuScale considers the overall design, testing, inspection and fabrication approach for the SG tube/SG tube support assembly is adequate and is commensurate with assuring an extremely low probability of failure of either the SG tubes or the SG tube support assemblies.

This is based on selection of appropriate ASME Code requirements, stipulation of additional design requirements, testing and implementation of a Steam Generator Program.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9564 Date of RAI Issue: 08/21/2018 NRC Question No.: 05.04.02.01-16 In Question 05.04.02.01-6.d in RAI 9231, the NRC staff requested confirmation that there are no tube support plate-to-tube support plate welds in the NuScale design and requested Table 5.2-7 in Tier 2 of the FSAR be revised to that effect. In response to Question 05.04.02.01-6.d, NuScale confirmed that there are no tube support plate-to-tube support plate welds in the design, and stated that there are welds between sections of the SG tube supports and that Table 5.2-7 was revised accordingly.

In the response to Question 05.04.02.01-6.d, NuScale noted the inservice inspection examination method for the welds between sections of the SG tube supports in Table 5.2-7 as general visual. For these welds, Table 5.2-7 was also revised to include a note that the general visual exams are an augmented exam and that the welds are examined where accessible. The NRC staff understands the phrase "augmented exam" to mean an exam is being performed beyond what is required. The NRC staff notes that the inservice inspection examination information for the SG supports and the assemblies that span the full height of the SG tube bundle (i.e., the tabs) is not included in Table 5.2-7.

The NRC staff does not understand the general visual examination methodology and what is meant by the phrase "examine where accessible." The NRC staff notes that current SGs in the United States follow industry guidelines for secondary side inspections (i.e., Nuclear Energy Institute (NEI) 97-06, Electric Power Research Institute (EPRI) PWR Steam Generator Examination Guidelines, and EPRI Steam Generator Integrity Assessment Guidelines).

Therefore, the NRC staff requests the following to ensure that the SG tube supports meet the requirements of GDC 1, 4, 14, 15, 31, and 32 in Appendix A of 10 CFR Part 50, 10 CFR 50.55a, 10 CFR 50.36, 10 CFR 50.65, and Appendix B in 10 CFR Part 50 as they relate to implementing a SG program and maintaining the integrity of the RCPB:

NuScale Nonproprietary

a. Revise Section 5.4.1.6 in Tier 2 of the FSAR to describe the general visual examination that will be performed on the welds between sections of the SG tube supports with an emphasis on how the general visual examination meets industry guidelines on secondary side inspections.

This description should include a discussion of the meaning of the phrase "examine where accessible."

b. Revise Table 5.2-7 to include the inservice inspection examination category and method for the lower and upper supports and the assemblies that span the full height of the SG tube bundle (i.e., the tabs). If necessary, revise Section 5.4.1.6 to include a description of the examination methodology with an emphasis on how it meets industry guidelines on secondary side inspections.

NuScale Response:

a. The NuScale Final Safety Analysis Report (FSAR), Section 5.4.1.6 describes the Steam Generator (SG) Program. The SG program includes secondary side visual inspection. FSAR Table 5.2-7 includes American Society of Mechanical Engineers (ASME)Section XI (Rules for Inservice Inspection of Nuclear Power Plant Components) inspections and augmented inspections. Since the NuScale SGs are located within the reactor pressure vessel (RPV)

(ASME Subsection NG, Internal structures is applied - see response to RAI 9436 - 05.04.02.01-15), there are no ASME Section XI inspection requirements. Therefore, any visual inspections performed as part of the SG program are neither ASME Section XI inspections nor "augmented inspections", and therefore are conducted to different criteria.

The industry guidelines for SG secondary side inspections are provided in Section 10.5 of the Electric Power Research Institute, "Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines". These guidelines, as they apply to shell side SG components (including SG tube supports), are focused on foreign object search and retrieval. Per the industry guidelines, the minimum regions to be inspected for loose or foreign objects include the "shell-to-tube bundle annulus region" and the "tube lanes". These regions directly correspond to the NuScale design, as there is an annulus between the innermost and outermost tube columns and between the riser and RPV respectively and there are tube lanes between adjacent helical columns in the SG. Figure 2-7 in the Comprehensive Vibration Assessment Program (CVAP) Technical Report (TR-0716-50439) provides the best representation of these features. Therefore, these guidelines are implemented without modification, as they apply to the shell side of the NuScale SG.

NuScale Nonproprietary

FSAR Section 5.4.16 currently states that the NuScale SG Program follows the EPRI guidance and NEI 97-06. Since no deviations from the industry guidelines for SG shell side inspections are proposed for the NuScale design, no change is necessary to FSAR section 5.4.1.6.

The criteria for performance of Table 5.2-7 augmented visual inspections were described in FSAR Section 5.2.4 and directs the reader to the ASME Boiler Pressure Vessel Code (BPVC)

Section XI. However, since there are no ASME Section XI required inspections and the steam generator industry guidelines provide requirements for inspection of the primary side, these inspections have been removed.

b. As discussed in the response to part a), there are no ASME Section XI required inspections, associated with Subsection NG, Internal Structures. Therefore, the inspections identified in FSAR Table 5.2-7 have been eliminated.

Impact on DCA:

FSAR Table 5.2-7 has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary RAI 04.05.02-2, RAI 05.04.02.01-6, RAI 05.04.02.01-16 Table 5.2-7: Reactor Vessel Internals Inspection Elements Description Location Examination Examination Notes Category Method Core Support Components Reflector Block - Bottom Core Support B-N-3 VT-3 Assembly Reflector Block Intermediate Core Support B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1.

Exam will be of the interior surface, checking for a gap developing between reflector blocks.

Reflector Block Top Core Support B-N-3 VT-3 Assembly Reflector Block Alignment Pins Core Support B-N-3 VT-1 Inspection only required Assembly when reflector blocks are removed for another reason Core Barrel Core Support B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1 of accessible surfaces Lower core Plate Core Support B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1 of accessible surfaces Upper Core Plate Lower Riser B-N-3 VT-3 Assembly Lower Core Plate Alignment Pins Core Support B-N-3 VT-3 Assembly Upper Support Block Core Support B-N-2 VT-1 Assembly Core Barrel to Lower Core Plate Core Support B-N-2 VT-1 Assembly Fuel Pins Lower Riser B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1 Fuel Pins Caps Lower Riser B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1 Fuel Pin Capture Weld Lower Riser B-N-2 VT-1 Assembly Shared Fuel Pins and Nuts Core Support B-N-3 VT-1 Required VT-3 Assembly augmented to VT-1 Lower Riser to Upper Core Plate Lower Riser B-N-3 VT-3 Assembly ICIGT Bottom Flag ICIGT 1 to Upper Core Lower Riser B-N-3 VT-1 Required VT-3 Plate Assembly augmented to VT-1 Internal Structures ICI Guide Tubes Lower Riser N/A VT-3 Assembly ICI Guide Tubes (Upper) Upper Riser N/A VT-3 Assembly ICI Guide Tubes Bottom Flag Lower Riser N/A VT-1 Required VT-3 Assembly augmented to VT-1 Tier 2 5.2-40 Draft Revision 3

NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary Table 5.2-7: Reactor Vessel Internals Inspection Elements (Continued)

Description Location Examination Examination Notes Category Method Brace to Pipe Upper Riser N/A VT-3 Assembly Pipe to Cap Upper Riser N/A VT-3 Assembly ICI Centering Plate 1-12 Upper Riser N/A VT-3 Assembly ICIGT Link to Upper Riser Hanger Ring Upper Riser N/A VT-3 Assembly Injection to RPV11 Upper Riser N/A VT-3 ICIGT to Integrated Steam Plenum PZR N/A VT-3 Pressurizer Spray Nozzle PZR Spray Nozzle to N/A VT-3 Safe End (RPV14-RPV15)

Surveillance Capsule Core Support N/A VT-3 Assembly Steam Generator Col 01-21 Tube Support Top/Bottom SG N/A General Visual Augmented exam.

Sections to Tube Support middle Section Examine where accessible.

Col 01-21 Outer Spacer to Tube Support SG N/A General Visual Augmented exam.

Examine where accessible.

Col 01-21 Middle Spacer to Tube Support SG N/A General Visual Augmented exam.

Examine where accessible.

Tier 2 5.2-42 Draft Revision 3

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9564 Date of RAI Issue: 08/21/2018 NRC Question No.: 05.04.02.01-17 In Question 05.04.02.01-8.a to RAI 9231, the NRC staff requested information about the design requirements for the SG inlet flow restrictors. In the response to Question 05.04.02.01-8.a, NuScale stated that the SG inlet flow restrictors are designed, fabricated, constructed, tested, and inspected as non-structural attachments and revised the FSAR accordingly. The NRC staff observed that Table 3.2-1 in Tier 2 of the FSAR notes the Quality Group as N/A for the SG inlet flow restrictors.

In response to Question 03.09.05-2 to RAI 8901, NuScale stated that the SG inlet flow restrictors and associated hardware are non-pressure boundary items and are not inside or integral to the RCPB and therefore are not "reactor internals" (ADAMS Accession No. ML17284A092). The response further indicated that there is no ASME design code associated with the SG inlet flow restrictors.

It is unclear to the NRC staff why there is no ASME code associated with the design, fabrication, construction, testing, and inspection of the SG inlet flow restrictors given they are a first-of-a-kind design, are mounted to the ASME Code Class 1 feed plenum tubesheets (forms part of the RCPB), and extend inside the ASME Code Class 1 SG tubes (forms part of the RCPB). If the SG inlet flow restrictors were to fail they could impact the integrity of the SG tubes and feed plenum tubesheets (integrity of the RCPB), and flow through the SG tubes. The NRC staff notes that secondary components of current SGs in the United States are designed in accordance with ASME Code,Section III, Subsections NB, NC, or NG. The use of ASME Code,Section III, Subsections NB, NC, or NG provides the staff a reasonable basis to conclude, in accordance with GDC 14, that the probability of abnormal RCPB leakage from the failure of secondary components is an extremely low probability event.

NuScale Nonproprietary

To ensure the SG inlet flow restrictors satisfy the requirements of 10 CFR Part 50, Appendix A, GDC 1, 4, 14, 15, 30, and 31, as they relate to ensuring the integrity of the RCPB, please provide the design, fabrication, construction, testing, and inspection criteria for the SG inlet flow restrictors and a discussion of how the criteria will ensure integrity of the RCPB.

NuScale Response:

The Steam Generator (SG) flow restrictors are designated as safety-related components based on their function to support the Decay Heat Removal System (i.e. perform residual heat removal function) and are located within the secondary side of the SGs. Based on these considerations, and Regulatory Guide 1.26 (Table 1), application of American Society of Mechanical Engineers (ASME) Boiler Pressure Vessel Code (BPVC)Section III, Subsection NC is required. Therefore, the code classification of the SG tube inlet flow restrictors (including associated mounting hardware) has been changed to ASME Section III, Subsection NC. The SG flow restrictors (including associated mounting hardware) are designed, fabricated, constructed, tested and inspected in accordance with ASME Section III, Subsection NC.

Changes to FSAR Section 5.4.1.5 are provided to implement the described change.

Impact on DCA:

FSAR Section 5.4.1.5 has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design RAI 05.04.02.01-17 The SG inlet flow restrictors are designed, fabricated, constructed, tested, and inspected as non-structural attachments to the RPV. The SG inlet flow restrictors are designed, fabricated, constructed, tested and inspected in accordance with the ASME BPVC,Section III, Subsection NC.

RAI 05.04.02.01-11 Refer to Section 5.2.3 for additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 1 components.

Refer to Section 5.2.3.4.2 for cleaning and cleanliness controls for the SGs. Refer to Section 6.6 for additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 2 components.

Threaded fasteners are described in Section 3.13.

5.4.1.6 Steam Generator Program The SG program monitors the performance and condition of the SGs to ensure they are capable of performing their intended functions. The program provides monitoring and management of tube degradation and degradation precursors that permit preventative and corrective actions to be taken in a timely manner, if needed. The SG program is described in the plant technical specifications and is a part of the overall ISI program. The program implements applicable portions of Section XI of the BPVC and specifically addresses 10 CFR 50.55a(b)(2)(iii). Appendix B to 10 CFR 50 applies to implementation of the SG program.

The NuScale SG Program follows NEI 97-06 and EPRI guidance (Reference 5.4-2).

Application of established commercial SG Program requirements to the NuScale design are appropriate based on the historical causes of SG tube degradation and the features of the NuScale SG design. The NuScale design incorporates design improvements necessary to restrict SG tube degradation and has additional design features that reduce the risk of SG tube degradation compared to existing PWR designs.

Historically, significant SG tube degradation has occurred in the operating PWR SG fleet due to various corrosion mechanisms, including wastage and both primary and secondary side stress corrosion cracking. These corrosion mechanisms were related to materials selection, plant chemistry control, and control of the ingress of impurities and corrosion products to the SGs. In the operating fleet, detrimental SG corrosion has been effectively mitigated based on use of A690TT SG tubing, application of EPRI primary and secondary plant chemistry control, and design of condensate systems (including extensive use of polishing resin beds and improved materials). These improvements have been implemented in the NuScale design.

RAI 05.04.02.01-2 In addition to chemistry and materials considerations, where the NuScale design is equivalent to the existing PWR fleet, there are two areas where the NuScale design has reduced SG tube degradation risk. The NuScale SG tube wall thickness is thicker than existing designs (see Table 5.4-2) based on incorporation of a substantial degradation Tier 2 5.4-13 Draft Revision 3

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9564 Date of RAI Issue: 08/21/2018 NRC Question No.: 05.04.02.01-18 In the response to Question 05.02.01.01-7 of RAI 9335, NuScale stated that the material of the thermal relief valves is in Section 5.4.1.5 in Tier 2 of the FSAR. The NRC staff did not find the material of the thermal relief valves identified in Section 5.4.1.5 or in Table 5.4-3. Therefore, to ensure the materials comply with 10 CFR 50.55a, please revise Section 5.4.1.5 and Table 5.4-3 to add the materials of the thermal relief valves, or alternatively describe in Section 5.4.1.5 and Table 5.4-3 where the materials are identified in the FSAR.

NuScale Response:

NuScale modified the Final Safety Analysis Report (FSAR) Section 5.4.1.2 to indicate that the pressure-retaining materials of the thermal relief valve are specified in Table 6.1-3.

Impact on DCA:

FSAR Section 5.4.1.2 and Table 6.1-3 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design create a continuous support path through the columns. The circumferential spacing of the tube supports is optimized to provide the minimum possible tube free span lengths, while still accommodating the transition of the tubes to the steam and feedwater plena.

RAI 05.04.02.01-6 The SG tubes are supported for vibration and seismic loads by vertical bars that extend through the tube bundle from the feed to the steam plena. As shown on Figure 5.4-6, the tube support assemblies are attached to upper SG supports that are welded to the inner surface of the RPV and also interface with lower SG supports that are welded to the inner surface of the RPV. The SG tube support assemblies in the SG provides contact with each tube at eight separate circumferential locations. The use of 8 sets of tube support assemblies limits the unsupported tube lengths, which ensures SG tube modal frequencies are sufficiently high to preclude unacceptable flow-induced vibration.

RAI 05.04.02.01-6 As shown in Figure 5.4-6, the lower SG supports permit thermal growth and provide lateral support of the tube supports.

Inlet Flow Restrictors RAI 05.04.02.01-8 A flow restriction device at the inlet to each tube ensures secondary-side flow stability and precludes density wave oscillations. The SG tube inlet flow restrictors provide the necessary secondary-side pressure drop for flow stability. The flow restrictors are mounted on a plate in each feed plenum that is attached to the secondary-side face of the tubesheets with stud bolts to avoid attaching the restrictors directly to the tube.

The flow restrictor stud bolts are welded to the tubesheet at each mounting location.

Mounting plate spacers hold the flow restrictor mounting plate off the surface of the tubesheet (see Figure 5.4-5). Spacers are located at each mounting plate attachment point. As shown in Figure 5.4-8, the individual flow restrictors extend into the tubes and are removable to support SG inspection, cleaning, tube plugging, or other maintenance and repair activities. The flow restrictor bolts are located at the center of the flow restrictor bolt assembly. The flow restrictor bolt runs the length of the assembly and holds the flow restrictor subcomponent. The flow restrictor bolts or nuts and the flow restrictor stud bolts or nuts include a locking feature to minimize the potential for loose parts generation.

Thermal Relief Valves RAI 05.02.01.01-7, RAI 05.04.02.01-18 To establish desired SG and DHRS chemistry during startup and shutdown, the SG and DHRS are flushed to the condenser, creating a water solid condition. Unintended containment isolation during these flushing evolutions could result in overpressure conditions caused by changes in fluid temperature. A single thermal relief valve is located on each feedwater line upstream of the tee that supplies the feed plenums (see Figure 5.4-9) to provide overpressure protection during shutdown conditions for the secondary side of the SGs, feedwater and steam piping inside containment, and the Tier 2 5.4-6 Draft Revision 2

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design DHRS when the secondary system is water solid and the containment is isolated. The thermal relief valves are spring-operated, balanced-bellows relief valves that vent directly into the containment. The thermal relief valves are classified Quality Group B and designedare designed, fabricated, constructed, tested and inspected as Class 2 in accordance with Section III of the BPVC and are Seismic Category I components. The pressure-retaining materials of thermal relief valves are specified in accordance with the materials identified in Table 6.1-3.

The thermal relief valves provide investment protection for the secondary system components during shutdown conditions and are not credited for safety-related overpressure protection for these systems during operation. Overpressure protection during operation is provided by system design pressure and the RSVs as described in Section 5.2.2.

Main Steam and Feedwater Plena Vent and Drain Valves Manual valves allow draining the main steam and feedwater plena prior to cover removal to facilitate outage maintenance and testing. The valves are used for maintenance only and are normally closed and capped.

Compatibility of Steam Generator Tubing with Primary and Secondary Coolant The chemistry of the primary and secondary water is controlled in accordance with industry guidelines suitably modified to address the unique NPM design and to ensure compatibility with the primary and secondary coolant. Section 5.2.3 describes the compatibility aspects of the reactor coolant chemistry that provide corrosion protection for stainless steels and nickel alloys, including SG components exposed to the reactor coolant. Section 6.1 describes the compatibility aspects of the secondary coolant chemistry that provide corrosion protection for stainless steels and nickel alloys, including the SG components exposed to the secondary system coolant and Section 10.3.5 describes the secondary water quality control program which is in accordance with Nuclear Energy Institute (NEI) 97-06 (Reference 5.4-1).

Copper deposits are a major source of SG corrosion products in nuclear plants with copper alloys in their secondary system. To minimize internal SG tube corrosion, low-melting point metals such as lead, antimony, cadmium, indium, mercury, zinc, bismuth, copper, tin, and their alloys and high sulfur materials; with the exception of strong acid cation resin; are excluded from use in reactor coolant primary system components and secondary system components.

Estimated radioactivity design limits for the secondary side of the SGs during normal operation and the basis are addressed in Section 11.1.2. The radiological effects associated with an SG tube failure are provided in Section 15.0.3.8.2.

5.4.1.3 Performance Evaluation A single RCS natural circulation flow loop is entirely contained within the RPV, thereby eliminating distinct RCS piping loops and the associated potential for a large pipe break (i.e., large break loss-of-coolant accident [LOCA]) event. This design, combined with the intertwined SGs tube bundle configuration, eliminates the potential for Tier 2 5.4-7 Draft Revision 2

NuScale Final Safety Analysis Report Engineered Safety Feature Materials RAI 05.02.03-1, RAI 05.04.02.01-18, RAI 06.01.01-1, RAI 06.01.01-1S1, RAI 06.01.01-2, RAI 06.01.01-3, RAI 06.01.01-4 Table 6.1-3: Pressure Retaining Materials for RCPB and ESF Valves Bodies SA-182 (Note 1) Grade F304, F304L, F304LN, F316, F316L, F316LN SA-351 (Note 2) Grade CF3, CF3A, CF3M, CF8, CF8A, CF8M SA-479 (Note 1) Type 304, 304L, 304LN, 316, 316L, 316LN Bonnets SA-182 (Note 1) Grade F304, F304L, F304LN, F316, F316L, F316LN SA-240 (Note 1) Type 304, 304L, 304LN, 316, 316L, 316LN SA-351 (Note 2) Grade CF3, CF3A, CF3M, CF8, CF8A, CF8M SA-479 (Note 1) Type 304, 304L, 304LN, 316, 316L, 316LN Discs SA-182 (Note 1) Grade F304, F304L, F304LN, F316, F316L, F316LN SA-351 (Note 2) Grade CF3, CF3A, CF3M, CF8, CF8A, CF8M SA-479 (Note 1) Type 304, 304L, 304LN, 316, 316L, 316LN, XM-19 SA-564 Type 630 Condition H1100 or H1150 SB-637 UNS N07718 Stems SA-479 (Note 1) Type 304, 304L, 304LN, 316, 316L, 316LN, XM-19 SA-564 Type 630 Condition H1100 or H1150 SB-637 UNS N07718 Pressure Retaining Studs, Bolts, and Screws SA-193 (Note 3) Grade B8, B8A, B8M, B8MA, B8R, B8RA, B8S, B8SA SA-453 (Note 4) Grade 660 Class A or B SA-564 Type 630 Condition H1100 SB-637 (Note 54) UNS N07718 Pressure Retaining Nuts SA-193 (Note 3) Grade B8, B8A, B8M, B8MA, B8R, B8RA, B8S, B8SA SA-194 Grade 8, 8A, 8M, 8MA, 8R, 8RA, 8S, 8SA SA-453 (Note 4) Grade 660 Class A or B SA-564 Type 630 Condition H1100 SB-637 (Note 54) UNS N07718 Filler Metals for Pressure Retaining Welds SFA-5.4 (Note 65) E308, E308L, E309, E309L, E316, E316L SFA-5.9 (Note 65) ER308, ER308L, E309, E309L, ER316, ER316L

1. Carbon is limited to 0.03 percent maximum for unstablized Type 3XX that are welded or exposed to temperature range between 800 °F and 1500 °F subsequent to final solution anneal.
2. Carbon is limited to 0.03 percent maximum. Delta ferrite is limited to 20 percent maximum, except to 14 percent maximum for CF3M and CF8M.
3. B8A, B8MA, B8R, and B8RA can only be used for Class 1 valves.
4. SA-453 Grade 660 is not used for pressure-retaining bolting exposed to RCS or pool water during operation or refuel, except for the specific bolting design, fabrication, and installation that are optimized to prevent stress corrosion cracking.
54. Solution treatment temperature range prior to precipitation hardening treatment is restricted to 1800 °F to 1850 °F.
65. Carbon is limited to 0.03 percent maximum. The ferrite number is in the range of 5FN to 20FN except Type 316 and Type 316L are in the range of 5FN to 16FN.

Tier 2 6.1-11 Draft Revision 2

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9564 Date of RAI Issue: 08/21/2018 NRC Question No.: 05.04.02.01-19 In Question 05.04.02.01-14.b to RAI 9231, the NRC staff requested that an appropriate program element that meets the intent of the maintenance of the SG secondary-side integrity program element in NEI 97- 06 be added to Combined License (COL) Information Item 5.4-1. The purpose of the maintenance of SG secondary-side integrity program element in NEI 97-06 is to monitor secondary-side SG components that are susceptible to degradation.

In response to Question 05.04.02.01-14.b, NuScale added "shell side integrity and accessibility assessment" and "steam plant corrosion product deposition assessment" to COL Information Item 5.4-1 to meet the intent of the SG secondary-side integrity program element in NEI 97-06.

However, it is unclear to the NRC staff how the proposed SG program elements meet the purpose of the SG secondary-side integrity program element from NEI 97-06.

The NRC staff requests Section 5.4.1.6 in Tier 2 of the FSAR be revised to describe how the proposed elements meet the intent of the NEI 97-06 program element, or alternatively use different terminology to describe the SG program element (e.g., "maintenance of SG shell side integrity" or "monitoring of SG shell side components that are susceptible to degradation"). This is requested to ensure GDC 32, 10 CFR 50.55a, 10 CFR 50.36, 10 CFR 50.65, and Appendix B in 10 CFR Part 50 are met by ensuring implementation of a SG program to maintain the structural and leakage integrity of the SG tubes.

NuScale Response:

The NRC has requested additional clarification for the use of the terms shell side integrity and accessibility assessment and steam plant corrosion product deposition assessment, as it relates to secondary side maintenance and integrity of the steam generators (SGs). Upon NuScale Nonproprietary

further review, NuScale is providing additional clarification and modifying the previous response to request for additional information (RAI) 9231, question 05.04.02.01-14.b, as described below.

NEI 97-06, Steam Generator Program Guidelines, includes broad guidance for maintenance of SG secondary side integrity by performing secondary side visual inspections. More specific guidance for performance of these inspections is provided in EPRI, Steam Generator Management Program: Pressurized Water Reactor Steam Generator Examination Guidelines.

Both of these documents are the basis for the NuScale SG secondary side maintenance and inspection program. In a traditional pressurized water reactor (PWR) SG, the purpose of secondary side integrity inspections is to assess the shell side of the SG for loose parts and/or foreign objects .

Regarding the term shell side integrity and accessibility assessment; for the NuScale design, SG shell side integrity is assessed by a visual inspection for loose parts and/or foreign objects, which is directly analogous to how shell side inspections are performed in typical PWR SGs, and is best described as a "shell side integrity assessment". The wording of COL item 5.4-1 has been changed to reflect this assessment. In the NuScale design the SG can be accessed for inspection between columns of tubes and at gaps in the riser and reactor pressure vessel (RPV) shell to inspect the full height of the SG. Visual inspections may be performed directly or by using remote cameras, boroscopes, fiber optic probes or other means. The extent of accessibility to the shell side of the NuScale design is considered to be approximately equivalent to existing PWR SGs.

Regarding the term steam plant corrosion product deposition assessment; upon further review this term is not directly included within NEI 97-06, Steam Generator Program Guidelines, and has therefore, been deleted from COL items 5.4-1. NuScale has determined that this term is redundant to assessment of degradation, which is included within the SG Program described in COL item 5.4-1. Assessment of degradation more broadly covers assessment of the SG secondary side condition, including identification of inspection plans, to address all types of SG degradation applicable to the NuScale design, including steam plant corrosion product deposition. Therefore, the need for secondary side inspections to assess corrosion product deposition is addressed by the term assessment of degradation, therefore the term steam plant corrosion product depositions assessment does not need to be identified separately within the COL item NuScale Nonproprietary

Impact on DCA:

COL Item 5.4-1 and Table 1.8-2 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Interfaces with Certified Design RAI 01-61, RAI 02.04.13-1, RAI 03.04.01-4, RAI 03.04.02-1, RAI 03.04.02-2, RAI 03.04.02-3, RAI 03.05.01.04-1, RAI 03.05.02-2, RAI 03.06.02-6, RAI 03.06.02-15, RAI 03.06.03-11, RAI 03.07.01-2, RAI 03.07.01-3, RAI 03.07.02-6S1, RAI 03.07.02-6S2, RAI 03.07.02-8, RAI 03.07.02-12, RAI 03.07.02-15S5, RAI 03.08.04-3S2, RAI 03.08.04-23S1, RAI 03.08.04-23S2, RAI 03.08.05-14S1, RAI 03.09.02-15, RAI 03.09.02-48, RAI 03.09.02-67, RAI 03.09.02-69, RAI 03.09.03-12, RAI 03.09.06-5, RAI 03.09.06-6, RAI 03.09.06-16, RAI 03.09.06-16S1, RAI 03.09.06-27, RAI 03.11-8, RAI 03.11-14, RAI 03.11-14S1, RAI 03.11-18, RAI 03.13-3, RAI 04.02-1S2, RAI 05.02.03-19, RAI 05.02.05-8, RAI 05.04.02.01-13, RAI 05.04.02.01-14, RAI 05.04.02.01-19, RAI 06.02.06-22, RAI 06.02.06-23, RAI 06.04-1, RAI 09.01.01-20, RAI 09.01.02-4, RAI 09.01.05-3, RAI 09.01.05-6, RAI 09.03.02-3, RAI 09.03.02-4, RAI 09.03.02-5, RAI 09.03.02-6, RAI 09.03.02-8, RAI 10.02-1, RAI 10.02-2, RAI 10.02-3, RAI 10.02.03-1, RAI 10.02.03-2, RAI 10.03.06-1, RAI 10.03.06-5, RAI 10.04.06-1, RAI 10.04.06-2, RAI 10.04.06-3, RAI 10.04.10-2, RAI 11.01-2, RAI 12.03-55S1, RAI 13.01.01-1, RAI 13.01.01-1S1, RAI 13.02.02-1, RAI 13.03-4, RAI 13.05.02.01-2, RAI 13.05.02.01-2S1, RAI 13.05.02.01-3, RAI 13.05.02.01-3S1, RAI 13.05.02.01-4, RAI 13.05.02.01-4S1, RAI 14.02-7, RAI 19-31, RAI 19-31S1, RAI 19-38, RAI 20.01-13 Table 1.8-2: Combined License Information Items Item No. Description of COL Information Item Section COL Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.1 site-specific plant location.

COL Item 1.1-2: A COL applicant that references the NuScale Power Plant design certification will provide the 1.1 schedules for completion of construction and commercial operation of each power module.

COL Item 1.4-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.4 prime agents or contractors for the construction and operation of the nuclear power plant.

COL Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site- 1.7 specific diagrams and legends, as applicable.

COL Item 1.7-2: A COL applicant that references the NuScale Power Plant design certification will list additional 1.7 site-specific piping and instrumentation diagrams and legends as applicable.

COL Item 1.8-1: A COL applicant that references the NuScale Power Plant design certification will provide a list of 1.8 departures from the certified design.

COL Item 1.9-1: A COL applicant that references the NuScale Power Plant design certification will review and 1.9 address the conformance with regulatory criteria in effect six months before the docket date of the COL application for the site-specific portions and operational aspects of the facility design.

COL Item 1.10-1: A COL applicant that references the NuScale Power Plant design certification will evaluate the 1.10 potential hazards resulting from construction activities of the new NuScale facility to the safety-related and risk significant structures, systems, and components of existing operating unit(s) and newly constructed operating unit(s) at the co-located site per 10 CFR 52.79(a)(31).

The evaluation will include identification of management and administrative controls necessary to eliminate or mitigate the consequences of potential hazards and demonstration that the limiting conditions for operation of an operating unit would not be exceeded. This COL item is not applicable for construction activities (build-out of the facility) at an individual NuScale Power Plant with operating NuScale Power Modules.

COL Item 2.0-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate 2.0 that site-specific characteristics are bounded by the design parameters specified in Table 2.0-1.

If site-specific values are not bounded by the values in Table 2.0-1, the COL applicant will demonstrate the acceptability of the site-specific values in the appropriate sections of its combined license application.

COL Item 2.1-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.1 site geographic and demographic characteristics.

COL Item 2.2-1: A COL applicant that references the NuScale Power Plant design certification will describe 2.2 nearby industrial, transportation, and military facilities. The COL applicant will demonstrate that the design is acceptable for each potential accident, or provide site-specific design alternatives.

COL Item 2.3-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.3 site-specific meteorological characteristics for Section 2.3.1 through Section 2.3.5, as applicable.

COL Item 2.4-1: A COL applicant that references the NuScale Power Plant design certification will investigate 2.4 and describe the site-specific hydrologic characteristics for Section 2.4.1 through Section 2.4.14, except Section 2.4.8 and Section 2.4.10.

COL Item 2.5-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.5 site-specific geology, seismology, and geotechnical characteristics for Section 2.5.1 through Section 2.5.5, below.

Tier 2 1.8-3 Draft Revision 3

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued)

Item No. Description of COL Information Item Section COL Item 5.3-1: A COL applicant that references the NuScale Power Plant design certification will establish 5.3 measures to control the onsite cleaning of the reactor pressure vessel during construction in accordance with Regulatory Guide 1.28.

COL Item 5.3-2: A COL applicant that references the NuScale Power Plant design certification will develop 5.3 operating procedures to ensure that transients will not be more severe than those for which the reactor design adequacy had been demonstrated.

COL Item 5.3-3 A COL applicant that references the NuScale Power Plant design certification will describe their 5.3 reactor vessel material surveillance program consistent with NUREG 0800, Section 5.3.1.

COL Item 5.4-1: A COL applicant that references the NuScale Power Plant design certification will develop and 5.4 implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines, and applicable Electric Power Research Institute steam generator guidelines at the time of the COL application. The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity and accessibility assessment, steam plant corrosion product deposition assessment, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting.

COL Item 6.2-1: A COL applicant that references the NuScale Power Plant design certification will develop a 6.2 containment leakage rate testing program that will identify which option is to be implemented under 10 CFR 50, Appendix J. Option A defines a prescriptive-based testing approach whereas Option B defines a performance-based testing program.

COL Item 6.2-2: A COL applicant that references the NuScale Power Plant design certification will verify that the 6.2 final design of the containment vessel meets the design basis requirement to maintain flange contact pressure at accident temperature, concurrent with peak accident pressure.

COL Item 6.3-1: A COL applicant that references the NuScale Power Plant design certification will describe a 6.3 containment cleanliness program that limits debris within containment. The program should contain the following elements:

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

COL Item 6.4-1: A COL applicant that references the NuScale Power Plant design certification will comply with 6.4 Regulatory Guide 1.78 Revision 1, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release.

COL Item 6.4-2: Not used. 6.4 COL Item 6.4-3: Not used. 6.4 COL Item 6.4-4: Not used. 6.4 COL Item 6.4-5: A COL applicant that references the NuScale Power Plant design certification will specify testing 6.4 and inspection requirements for the control room habitability system and control room envelope integrity testing as specified in Table 6.4-4.

COL Item 6.6-1: A COL applicant that references the NuScale Power Plant design certification will implement an 6.6 inservice testing program in accordance with 10 CFR 50.55a(f).

Tier 2 1.8-10 Draft Revision 3

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design flowrates across the SG tubes in PWR recirculating steam generators as discussed in Section 5.1. This low flow rate reduces the flow energy available to cause flow induced vibration (FIV) wear degradation of SG tubes. Based on the additional tube wall margin and the additional margin against FIV turbulent buffeting wear (the most likely SG tube degradation mechanism), application of the existing PWR SG Program requirements to the NuScale design is appropriate.

RAI 05.04.02.01-6 For SGs in the operating PWR fleet with A690TT SG tubing, the only observed degradation has been wear as a result of flow induced vibration (tube-to-tube or tube-to-support plate) or wear due to foreign objects. With respect to the risk of introduction of foreign objects, the NPM is at no greater risk than existing designs, therefore no deviations from existing SG program guidelines are warranted. From the standpoint of SG tube design, the two significant differences between the NuScale SG design and existing designs is the helical shape of the SG tubing and the SG tube support structure. The helical shape of the SG tubing itself does not represent risk of degradation based on the minimum bend radius of the helical tubing being within the experience base of operating PWR SG designs. The SG tube support design is novel.

However, as discussed in Section 5.4.1.2, it preserves attributes of the existing tube support (plate) designs. Prototypic testing of the SG tube supports is performed to validate acceptable performance (including wear) of the SG tube support design.

Implementation of a typical SG program is appropriate based on evaluation of the design of the SG tube supports.

5.4.1.6.1 Degradation Assessment A SG degradation assessment of the NPM SG identified several potential degradation mechanisms. As observed in the operating PWR fleet, wear is the most likely degradation mechanism. The preliminary SG degradation assessment also identified the potential for several secondary side corrosion mechanisms, including under deposit pitting and intergranular attack based on the once through design with secondary boiling occurring inside the tubes. The estimated growth rates for these potential defects is sufficiently low that the SG tube plugging criteria for the NPM SG is a 40% through wall defect, consistent with the existing PWR SG fleet.

Based on the ability to implement tube plugging criteria consistent with the operating PWR SG fleet, consistent implementation of other elements of the SG Program, including SG inspection frequency, is appropriate.

RAI 05.04.02.01-14, RAI 05.04.02.01-19 COL Item 5.4-1: A COL applicant that references the NuScale Power Plant design certification will develop and implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines," and applicable Electric Power Research Institute steam generator guidelines at the time of the COL application. The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity and accessibility assessment, steam plant corrosion product deposition Tier 2 5.4-14 Draft Revision 3

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design assessment, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting.

5.4.2 Reactor Coolant System Piping 5.4.2.1 Design Basis Pressure-retaining portions of piping that penetrate the RCS form, in part, the RCPB as defined in 10 CFR 50.2 and include the pressurizer spray supply, RCS injection, RCS discharge, and RPV high-point degasification piping.

5.4.2.2 Design Description Each of the RCS lines enter containment through welded penetrations on the containment upper head and contain two containment isolation valves mounted on the outside of the containment as described in Section 6.2.

A single pressurizer spray supply line enters through the containment head. This line branches inside containment into two pressurizer spray supply lines, each of which is welded to a penetration on the RPV upper head with a corresponding spray nozzle inside the RPV near the top of the pressurizer space.

The RPV high-point degasification line is a single line that is routed from the containment upper head to a welded penetration on the RPV upper head.

The RCS injection line is routed from the containment upper head to a welded penetration on the side of the RPV. Inside the RPV, the line continues from the RPV wall through the lower portion of the upper riser assembly and terminates near the center of the riser. Reactor coolant injection flow enters in the central riser above the reactor core. The RCS injection line also contains two branch connections to the ECCS lines that connect to the reset valves of the five ECCS valves: three RVVs and two RRVs.

The RCS discharge line is routed from the containment upper head to a penetration on the side of the RPV at an elevation just below the SGs. This penetration takes suction from the annular region between the RPV wall and the riser. Figure 6.6-1 depicts the RCS piping from the CNV upper head to the respective penetrations on the RPV.

5.4.2.3 Performance Evaluation Section 3.9, Section 3.12, and Section 5.2 provide information regarding the RCS piping criteria, methods, and materials, and include the design, fabrication, and operational provisions to control those factors that contribute to stress-corrosion cracking. The RCS piping supports the functional aspects of the chemical volume and control system (CVCS) as summarized in Section 9.3.4.

5.4.2.4 Tests and Inspections RAI 03.08.02-15S1 Tier 2 5.4-15 Draft Revision 3

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