ML24215A103
| ML24215A103 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 08/02/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24215A000 | List:
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| References | |
| LO-169995 | |
| Download: ML24215A103 (1) | |
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Response to SDAA Audit Question Question Number: A-5.1.3.4-1 Receipt Date: 09/15/2023 Question:
Please clarify the ASME Code classifications for the SGs, SG supports, and SG tube supports.
The NRC staff noted apparent inconsistencies in statements about the design in the Final Safety Analysis Report (FSAR), such as the following:
Section 5.1.3.4 states that the SG supports are reactor vessel internals (RVI) sub-assemblies.
Sections 3.9 and 5.3.1.1 state that the SG supports and SG tube supports are RVIs.
Section 5.3 states that the SGs and SG tube supports are RVIs.
Section 5.4.1.5 states SG supports and SG tube supports are designated as ASME Code Section III, Division 1, Subsection NG Internal Structures.
Based on these examples it appears that the SG supports and SG tube supports are designed as RVIs according to Subsection NG. The staff notes that it is not appropriate to identify the SGs as RVIs since not all parts of the SGs are designed to Subsection NG. For example, Section 5.4.1.3.1 states that the SG tubes are constructed to ASME Code Section III, Subsection NB.
In addition, please identify any exceptions to the ASME Code requirements in the design of the SG supports and SG tube supports.
A follow-up question was issued on November 17, 2023:
The response to Audit Issue A-5.1.3.4-1 states that the steam generator (SG) components, except for the SG tubes and the upper SG supports, are classified as reactor vessel internals (RVI). In addition, the response states, The remaining components of the SG are classified as RVI and are constructed to ASME Code Section III, Subsection NG as internal structures. The response also states that the statement in SDAA Section 5.3 describes what the RPV contains NuScale Nonproprietary NuScale Nonproprietary
and includes the SGs and SG tube supports in the reactor vessel internals because the SGs, except for the SG tubes and upper SG support, are RVI. No changes were proposed to the SDAA.
Based on the response, the SG tubes and upper SG supports are not classified as RVI because the tubes are part of the reactor coolant pressure boundary (RCPB) and the upper SG supports are welded to the reactor pressure vessel (RPV) upper steam plenum shell. The SG consists of other components that are part of the RCPB (e.g., integral steam plenum), constructed to ASME Code Section III, Subsection NB (e.g., Class 1 feed plenum components, Class 2 steam plenum components), constructed to ASME Code Section III, Subsection NC (e.g., inlet flow restrictors),
or welded to the inner surface of the RPV (e.g., lower SG supports). Therefore, the statements that, other than the SG tubes and upper SG supports, SG components are RVI and are constructed to ASME Code Section III, Subsection NG appear to be incorrect.
- a. Given that the lower SG supports are also welded to the RPV, please discuss why they are considered RVI.
- b. The following statements in the SDAA appear to be incorrect because SDAA Section 5.4.1.2 indicates that the SG supports consist of lower and upper SG supports, and based on the response, the upper SG supports are not classified as RVI and are constructed to ASME Code Section III, Subsection NB. Please discuss revising the SDAA to clarify these statements. The staff notes that if the lower SG supports are not to be considered as RVI, then it would not be appropriate to state SG supports are RVI.
SDAA Section 5.1.3.4 states, The RVI sub-assemblies include the core support assembly, lower riser assembly, upper riser assembly, SG supports, instrumentation guide tubes, core support block assembly, and flow diverter. In addition, this statement does not include the SG tube supports.
SDAA Section 5.2 states, The RPV contains the reactor core, reactor vessel internals (including the SGs and SG tube supports), pressurizer, and reactor coolant volume. The RPV is supported laterally and vertically by the CNV. As written, the parenthetical identifies the SGs and SG tube supports as RVI. In addition, as previously discussed, there are SG components other than the SG tubes and upper SG supports that should not be considered RVI and are not constructed to ASME Code Section III, Subsection NG.
SDAA Section 5.3.1.1 states, The reactor vessel internals, including the SG supports and SG tube supports (shown in Figure 5.4-5), are fabricated in accordance with ASME BPVC Section III, NG-4000.
SDAA Section 5.4.1.5 states, The SG supports and SG tube supports are designated as ASME BPVC,Section III, Subsection NG Internal Structures.
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SDAA Section 3.9.3.1.2 states, The SG supports and SG tube supports are designated as internal structures in accordance with ASME Section III, Subsection NG.
SDAA Section 3.9.5 states the SG supports are a sub assembly of the RVI assembly.
- c. SDAA Section 5.4.1.2 states that the upper SG supports are welded to the PZR baffle plate (SDAA Section 5.4.5.4, constructed to ASME Code Section III, Subsection NB) and inner surface of the RPV and the lower SG supports are welded to the inner surface of the RPV. In accordance with ASME Code,Section III, Subparagraph NB1132.2(d), the welds connecting the SG supports to the RPV are considered part of the RPV; therefore, the welds are to be classified as ASME Code Class 1 and conform to ASME Code, Subsection NB. Please discuss where the classification of the welds connecting the lower and upper SG supports to the RPV is addressed in the SDAA.
A follow-up question was issued on January 30, 2024:
A. DCA Section 5.4.1.5 states, in part, The SG supports and SG tube supports are designated as BPVC,Section III, Subsection NG Internal Structures. The design, fabrication, construction, and testing of the SG supports and SG tube supports, including weld materials, follow all requirements of BPVC,Section III, Subsection NG,.
As proposed in the revised response to Audit Issue A-5.1.3.4-1, SDAA Section 5.4.1.5 would state, The SG upper supports are designated as ASME BPVC,Section III, Subsection NF Class 1. The SG lower supports and SG tube supports are designated as ASME BPVC,Section III, Subsection NG. The design, fabrication, construction, and testing of the SG supports and SG tube supports, including weld materials does not adversely affect the integrity of the core support structures. Therefore, it is unclear in the SDAA whether the design, fabrication, construction, and testing of the SG supports and SG tube supports follow all Subsection NF (upper SG supports) and NG (lower SG supports and SG tube supports) requirements. In addition, the NRC staff noted that ED-102539, Rev. 0 indicates that the upper SG supports are Subsection NG.
Please revise SDAA Section 5.4.1.5 to clearly state whether the design, fabrication, construction, and testing of the upper SG supports, including weld materials, follow all requirements of BPVC,Section III, Subsection NF, and the lower SG supports and SG tube supports, including weld materials, follow all requirements of BPVC,Section III, Subsection NG.
If not all requirements are being followed, please add a description and justification of the exceptions to Subsection NF and/or Subsection NG to the SDAA.
B. SDAA Section 5.4.1.2 discusses SG supports and SG tube supports. It is the NRC staffs understanding that the SG supports consist of the upper and lower SG supports, NuScale Nonproprietary NuScale Nonproprietary
and the SG tube supports are the assemblies between each column of tubes. The revised response to Audit Issue A-5.1.3.4-1 proposes to change SDAA 3.9.5 to include SG tube supports and tube support assemblies, lower SG supports, support clips, and backing strips to the list of RVI sub-assemblies. It also proposed to change SDAA Section 5.1.3.4 to state, in part, The RVI sub-assemblies includelower SG supportsSG tube supports and tube support assemblies; lower SG supports, support clips, and backing strips It is unclear to the NRC staff what the difference is between SG tube supports and tube support assemblies. In addition, the NRC staff noted that lower SG supports appears twice in the proposed changes to SDAA Section 5.1.3.4.
Please describe the difference between SG tube supports and tube support assemblies in the proposed changes to SDAA Sections 3.9.5 and 5.1.3.4. Please also consider removing the duplicate lower SG tube supports in the proposed revision to 5.1.3.4.
A follow-up question was issued on February 27th:
The proposed markup for SDAA Section 3.9.3.1.2 in the response to Audit Issue A-5.1.3.4-1 states, Remaining portions of the RVI are designated as internal structures and constructed to ASME Subsection NG. In addition, the proposed markup states, The lower SG supports and SG tube supports are designated as internal structures in accordance with ASME Section III, Subsection NG.
During the clarification call on Audit Issue A-5.1.3.4.1-1 on February 7, 2024, NuScale noted that they use the term construction as it is used in the ASME Code. The staff notes that NCA-9200 of the 2017 Edition defines construction as an all-inclusive term comprising materials, design, fabrication, examination, testing, inspection, and certification required in the manufacture and installation of an item.
Therefore, based on the proposed statements in SDAA Section 3.9.3.1.2, the FSAR appears to clearly state that the design, fabrication, construction, and testing of the lower SG supports and SG tube supports meet the requirements of ASME Section III, Subsection NG.
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The proposed markup for SDAA Section 3.9.3.1.2 in the response to Audit Issue A-5.1.3.4-1 states, The upper SG supports are designated as Class 1 supports in accordance with ASME Section III, Subsection NF. However, it does not have a similar statement regarding its construction. Given the language used for the lower SG supports and the SG tube supports and the discussion of the use of the term construction, adding a similar statement regarding the construction of the upper SG supports would add clarity to the FSAR.
The proposed markup of SDAA Section 3.9.5 in the response to Audit Issue A-5.1.3.4-1 states, Although conformance with ASME Section III, Subsection NG is not mandatory, internal structures are constructed to not adversely affect the integrity of the core support structures in accordance with NG-1122(c). The staff notes that the proposed markup of SDAA Section 5.3.1.2 states, The design, fabrication, construction, and testing of the SG supports and SG tube supports, including weld materials does not adversely affect the integrity of the core support structures. Based on the February 7, 2024, clarification call, this language was used because it meets ASME Code Section III, Subsection NG-1122(c). Therefore, should it state lower SG supports and SG tube supports, since the upper SG supports are designated as Subsection NF?
Response
The lower steam generator (SG) supports are designated as internal structures within the reactor vessel internals (RVI). The lower SG supports are constructed to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPVC)Section III, Subsection NG.
The upper SG supports are designated as Class I supports and are constructed to ASME BPVC Section III, Subsection NF.
The SG tubes are part of the reactor coolant pressure boundary (RCPB). The SG tubes are constructed to ASME BPVC Section III, Subsection NB.
The SG tube supports are designated as internal structures and are constructed to ASME BPVC Section III, Subsection NG.
Response to Follow-up questions received on November 17th, 2023:
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Follow-up question a)
The lower SG supports are located within the RPV and are designated as internal structures.
The upper SG supports are also located within the RPV and could also be designated as internal structures; however, to support manufacturing schedules, the upper SG supports are classified as Class I supports. The upper SG supports are welded to the pressurizer baffle plate and to the inner surface of the reactor pressure vessel (RPV), as stated in Standard Design Approval Application (SDAA) Section 5.4.1.2.
The classification of SG components for the NuScale Power Module (NPM-20) maintains all minimum ASME BPVC requirements, as indicated in the attached markups to the SDAA and considers requirements based on function and weld location, as appropriate.
The markups attached to this response incorporate changes to address this response and also to address an internal Corrective Action Program item.
Follow-up question b)
The upper SG supports are classified as Class I supports and are not considered internal structures; they are constructed to ASME BPVC Section III, Subsection NF. The lower SG supports are designated as internal structures and are constructed to ASME BPVC Section III, Subsection NG. Both upper SG supports and lower SG supports provide support to the SG and are discussed in the Steam Generator Tube Supports and Steam Generator Supports description in SDAA Section 5.4.1.2. The description in SDAA Section 5.4.1.2 does not discuss ASME BPVC Section III requirements for different supports; those discussions are in SDAA Section 3.9, as noted in SDAA Section 5.4.1.2; in SDAA Section 5.3.1.1; and in SDAA Section 5.4.1.5.
Section 5.1.3.4 and Section 5.1.3.7 of SDAA contain markups attached to this response to clarify RVI components.
The statement ascribed to SDAA Section 5.2 are actually in SDAA Section 5.3. The markups attached to this response remove the confusing statement from SDAA Section 5.3.
Section 5.3.1.1 of SDAA contains markups attached to this response to clarify and correct the ASME BPVC designations of the different supports.
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Section 5.4.1.5 of SDAA contains markups attached to this response to clarify the SG materials requirements.
Section 3.9.3.1.2 of SDAA contains markups attached to this response to clarify and correct the SG load combination and stress limit requirements.
Section 3.9.5 of SDAA contains markups attached to this response to clarify and correct the RVI requirements.
Follow-up question c)
As stated in SDAA Section 5.4.1.5, The SGS materials forming the RCPB, including weld materials, conform to fabrication, construction, and testing requirements of ASME BPVC,Section III, Subsection NB. Table 5.4-3 lists the materials for weld filler metals for SG supports, which comply to the appropriate requirements for RPV welds.
Response to Follow-up questions received on January 30th, 2024 and February 27th, 2024 A clarification call was held on February 7th, to discuss the follow-up questions received on January 30th. Topics of discussion were the ASME designations for the upper and lower steam generator supports, and steam generator tube supports; the term construction as defined by the ASME Code; and the configuration of the tube support assembly.
During the clarification call, NuScale agreed to clarify the meaning and usage of tube support assembly in the final safety analysis report.
After the clarification call, an additional follow-up question was received on February 27th. While the Section 5.4.1.5, Steam Generator Materials, statement of no adverse impact is derived from subsection NG, it is true for both the upper and lower steam generator supports. However, as noted by the staff, the statement suggests the upper supports are designated Subsection NG. In response, changes in Section 5.4.1.5 confirm the construction requirements for the upper steam generator supports, and changes to the existing statement specify the lower steam generator supports.
The attached markup shows the FSAR changes for the tube support assembly and the steam generator support construction requirements.
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Markups of the affected changes, as described in the response, are provided below:
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NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-27 Draft Revision 2 mounting assemblies are analyzed to ensure minimum safety factors of five for material ultimate strength and three for material yield strength, and are maintained for dual-load-path loading conditions considering the dynamic load factor specified.
3.9.3.1.2 Load Combinations and Stress Limits Audit Question A-5.1.3.4-1 The RPV is a Seismic Category I, ASME Section III, Class 1 component. Load combinations and stress limits for the Class 1 RPV componentsshell and head are presented in Table 3.9-3. Load combinations for RPV piping and valve nozzles are in Table 3.9-4. Load combinations for RPV bolted flange connections are in Table 3.9-5. Load combinations for Class 1 supports are in Table 3.9-6.
The CNV is a Seismic Category I component. The ASME classification of the CNV and its supports is described in Section 3.8.2. Load combinations and stress limits for the CNV and its supports are in Section 3.8.2.
Audit Question A-5.1.3.4-1 The RVI are Seismic Category I components. Portions of the RVI that perform a core support function are Class CS components in accordance with ASME Section III, Subsection NG. Remaining portions of the RVI are designated as internal structures; however, they are designed using NG-3000 as a guide and constructed to ASME Subsection NG. Load combinations and stress limit are in Table 3.9-7.
Audit Question A-5.1.3.4-1 The lower SG supports, upper SG supports, SG tube supports, and support clipsSG supports and SG tube supports are Seismic Category I components.
The lower SG supports and SG tube supports are designated as internal structures in accordance with ASME Section III, Subsection NG. The upper SG supports are designated as Class I supports in accordance with ASME Section III, Subsection NF. The load combinations and stress limit are consistent with those in Table 3.9-7.
Portions of the CRDM providing a RCPB function are ASME Code Class 1, Seismic Category I components. The CRDM pressure housing is a Class 1 appurtenance per ASME BPVC,Section III, NCA-1271. The load combinations and stress limits are in Table 3.9-8. The CRDM seismic supports located on the RPV and CNV head are ASME Code Class 1, Seismic Category I component supports.
The DHRS condensers are Seismic Category I components and are classified as ASME Section III, Class 2 components. The condenser supports are classified as ASME Section III, Subsection NF, Class 2 supports. Load combinations and stress limits are presented in Table 3.9-9 and Table 3.9-10.
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-39 Draft Revision 2 and measured. The testing of the prototype included CRA drop time and misalignment testing and wear susceptibility assessment is described in Section 4.2.4.2.3.
The CRDS design has an Operability Assurance Program. The CRDS Operability Assurance Program testing is a series of tests designed to demonstrate the life cycle performance and endurance, including wear characteristics, of the CRDS in a representative operational environment.
Audit Item A-3.9.4-3 The testing verifies the performance of the CRDS components under a broad range of conditions of temperature, pressure, and flow representative of design conditions. The tests also demonstrate the acceptability of the design to meet the seismic and dynamic conditions that are expected based on the seismic and dynamic analyses. The life cycle testing is conducted under simulated normal operating conditions to demonstrate acceptable mechanical operation of the CRDM and ensure that anticipated mechanical actuations (steps, reactor trips, CRA coupling/decoupling cycles) representative of the 60-year design life of the system can be achieved without a failure that would prevent the CRDM from performing its safety-related function.
This series of tests is intended to demonstrate acceptable performance of the CRDS with respect to wear, functioning times, latching, and the ability to overcome a stuck rod, meeting system design requirements.
COL Item 3.9-5:
An applicant that references the NuScale Power Plant US460 standard design will implement a control rod drive system Operability Assurance Program that meets the requirements described in Section 3.9.4.4 and provide a summary of the testing program and results.
A series of production tests are performed on each CRDM that verifies the integrity of the pressure housing and the function of the CRDM. These tests include a hydrostatic test in accordance with the ASME BPVC,Section III, Division I, Subsection NB.
The as-built CRDMs are subject to pre-operational testing that verifies the sequencing of the operating coils and verifies the design requirements are met for insertion, withdrawal, and drop times. A description of the initial startup test program is provided in Section 14.2.
In accordance with the technical specifications, the CRDMs are subjected periodically to partial-movement checks to demonstrate the operation of the CRDM and acceptable core power distribution. In addition, drop tests of the CRA are performed as specified in Technical Specification Surveillance Requirement 3.1.4.3 to verify the ability to meet trip time requirements.
3.9.5 Reactor Vessel Internals Audit Question A-5.1.3.4-1
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-40 Draft Revision 2 The RVI assembly is comprised of several sub-assemblies located inside the RPV.
The RVI support and align the reactor core system, which includes the CRAs, support and align the control rod drive rods, and include the guide tubes that support and house the in-core instrumentation (ICI). In addition to performing these support and alignment functions, the RVI channel reactor coolant from the reactor core to the SG and back to the reactor core.
Audit Question A-5.1.3.4-1 The RVI primary functions are to:
provide structures to support, properly orient, position, and seat the fuel assemblies to maintain the fuel in an analyzed geometry to ensure core cooling capability and physics parameters are met under all modes of operational and accident conditions provide support and properly align the CRDS without precluding full insertion of control rods under all modes of operational and accident conditions provide the flow envelope to promote natural circulation of the RCS fluid with consideration given to minimizing pressure losses and bypass leakage associated with the RVI, and to the flow of coolant to the core during refueling operations improve neutron economy provide instrument guide tubes in support of the in-core instrumentation (ICI) provide structural support of the SG tubes The RVI assembly is comprised of the following sub assemblies and items:
core support assembly (CSA) lower riser assembly Audit Question A-5.1.3.4-1 upper riser assembly (URS) flow diverter ICI and riser level sensor guide tubes CSA mounting brackets Audit Question A-5.1.3.4-1 SG tube supports and tube support assemblies Audit Question A-5.1.3.4-1 lower SG supports, support clips, and backing strip Audit Question A-5.1.3.4-1 The selection, design, fabrication, installation, and inspection of reactor internals meet the criteria of 10 CFR 50.55a. The design and construction of the core support structures comply with ASME BPVC,Section III, Division 1, Subsection NG. For internal structures, Although conformance with ASME Section III, Subsection NG is not mandatory; however, internal structures are designed using NG-3000 as a guide and are constructed to not adversely affect the integrity of the core support structures, in accordance with NG-1122(c). The RVI materials including base materials and weld
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-41 Draft Revision 2 filler materials are discussed in Section 4.5.2 and are designed to minimize the number of welds and bolted interfaces within the high neutron flux regions.
Audit Question A-5.1.3.4-1 Section 3.1 discusses the design's general compliance with GDCs. Descriptions below describe compliance for reactor vessel internals. The design complies with 10 CFR 50.55a and the relevant requirements of the following General Design Criteria of 10 CFR 50, Appendix A:
Audit Question A-5.1.3.4-1 GDC 1 and 10 CFR 50.55a, as they relates to reactor internals; the reactor internals are designed to quality standards commensurate with the importance of the safety-related functions to be performed. Section 17.5 describes the Quality Assurance Program Description.The RVI components are Seismic Category I and designed to meet ASME BPVC,Section III, Division 1, Subsection NG Code requirements.
GDC 2, as it relates to reactor internals; the reactor internals are designed to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety-related functions for core cooling and control rod insertion. Pursuant to GDC 2, mechanical components are designed to withstand the loads generated by natural phenomena as discussed Section 3.1.1.
GDC 4, as it relates to reactor internals; reactor internals are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operations, maintenance, testing, and postulated pipe ruptures, including LOCA. Dynamic effects associated with postulated pipe ruptures, such as guillotine breaks of primary piping that cause asymmetric loading effects, are excluded from the design basis when analyses demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
GDC 10, as it relates to reactor internals; reactor internals are designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of AOOs.
3.9.5.1 Design Arrangements Figure 3.9-1 through Figure 3.9-4 show the RVI subassemblies with components that comprise the RVI.
Each of the RVI sub-assemblies is described in more detail below.
Audit Question A-5.1.3.4-1 During all operating conditions, the reactor core is supported by the core support structures of the CSA (core barrel and lower core plate) and upper core plateand lower riser assembly. The deadweight and other mechanical loads from the fuel are transferred to the upper and lower core plates. Under seismic and other accident conditions, the CSA transfers loads to the RPV through the CSA mounting brackets at the bottom of the RPV.
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-42 Draft Revision 2 The CSA includes the core barrel, upper support blocks, lower core plate, shared fuel pins and nuts, and reflector blocks (Figure 3.9-4).
Audit Question A-5.1.3.4-1 The core barrel is a continuous cylinder with no welds. The upper support blocks are fastened to the core barrel. The lower core plate, which is welded to the bottom of the core barrel, supports and aligns the bottom end of the fuel assemblies. The CSA mounting brackets are attached to the RPV bottom head.
Audit Question A-5.1.3.4-1 The reflector blocks contain no welds. The reflector blocks are aligned by reflector block alignment pins and stacked on the lower core plate inside the core barrel.
The shape of the reflector block assembly closely conforms to the shape of the peripheral fuel assemblies and constrains lateral movement of the fuel assemblies and minimizes the reactor coolant flow that bypasses the fuel assemblies.
The fuel is surrounded by a heavy neutron reflector. The heavy reflector, termed reflector blocks, reflects neutrons back into the core to improve fuel performance.
The heavy reflector provides the core envelope and directs the flow through the core. The heavy reflector blocks do not provide support to the core and is classified as an internal structure. During seismic and other accident events the heavy reflector limits lateral movement of the fuel assemblies and transfers potential loads to the core barrel assembly.
Audit Question A-5.1.3.4-1 The reflector blocks contain no welds. The reflector blocks are aligned by reflector block alignment pins and stacked on the lower core plate inside the core barrel.
The shape of the reflector block assembly closely conforms to the shape of the peripheral fuel assemblies and constrains lateral movement of the fuel assemblies and minimizes the reactor coolant flow that bypasses the fuel assemblies.
Audit Question A-5.1.3.4-1 The CSA sits upon the CSA mounting brackets, which are attached to the RPV bottom head.
A flow diverter is attached to the RPV bottom head, under the CSA, as shown in Figure 3.9-1. This flow diverter eases the transition of the reactor coolant flow from the downward flow outside the core barrel to upward flow through the fuel assemblies. The flow diverter reduces flow turbulence and recirculation and minimizes flow related pressure loss in this region.
The lower riser assembly includes the lower riser, the upper core plate, CRA guide tube assemblies, CRA guide tube support plate, and ICI guide tubes (Figure 3.9-3). The lower riser assembly is located immediately above the CSA and is aligned with and supported on the CSA by the four upper support blocks.
The lower riser channels the reactor coolant flow exiting the reactor core upward toward the upper riser, and separates this flow from the flow outside the lower riser that is returning from the SGs, with the exception of flow paths in the riser
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-43 Draft Revision 2 shell that permit a small amount of reactor coolant to bypass the top of the riser and flow into the downcomer region.
The upper core plate, which is attached to the upper support blocks, supports and aligns the top end of the fuel assemblies. The CRA guide tubes are fastened to the upper core plate, extend upward, where they are fastened to the CRA guide tube support plate. These guide tubes house the portion of the CRAs that extend above the top of the reactor core.
The ICI guide tubes are supported at the top by the CRA guide tube support plate, and at the bottom by the upper core plate, both of which maintain alignment of the ICI guide tubes with their respective fuel assemblies.
Audit Question A-5.1.3.4-1 The upper riser assembly includes the upper riser shell, a series of CRDS support plates, a transition shell, a hanger plate, control rod drive shaft sleeves and a bellows assembly. The upper riser assembly also accepts and positions the RCS injection piping. The CRDS support plates provide support for the control rod drive shafts, the ICI guide tubes, riser level sensor guide tube, and the upper riser shell.
The transition shell provides the flow interface between the upper riser assembly and lower riser assembly. The ICI guide tubes extend from their respective penetrations in the RPV top head downward through the PZR space, the upper riser, and the lower riser to their respective fuel assemblies. The portion of the riser level sensor and ICI guide tubes extend from the RPV upper head penetration to the bottom of the upper riser assembly. The portion of the ICI guide tubes extending from the RPV upper head penetrations to the bottom of the upper riser assembly is depicted in Figure 3.9-2. There is a bellows assembly in the lower portion of the upper riser (Figure 3.9-2). The bellows assembly allows for differential vertical thermal expansion between the upper riser assembly (URS)A and RPV and additionally provides a downward compressive load on the lower riser assembly to minimize leakage flow between the assemblies.
The upper riser assembly is located immediately above the lower riser assembly and extends upward to the pressurizer (PZR) baffle plate. It channels the reactor coolant exiting the lower riser upwards and into the space above the top of the upper riser shell and below the PZR baffle plate. Reactor coolant flow then turns downward through the annular space outside of the upper riser and inside of the RPV where the SG helical tube bundles are located. Flow paths are located in the upper riser to permit a small amount of reactor coolant to bypass the top of the riser and flow into the SG tube bundle region. These flow paths ensure sufficient boron concentration remains in the reactor coolant during DHRS-driven riser uncovery conditions following non-LOCA transients, while not introducing structural integrity and fatigue concerns.
NuScale evaluated the potential for acoustic noise from vortices that may be formed at the upper riser holes, and other potential flow-induced vibration effects of flow through the holes onto the SG tubes. The riser flow holes are not expected to produce vortices due to the flow through the holes. If, however, vortices form, they do not coincide with a relevant acoustic mode of the riser. Additionally, the
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-44 Draft Revision 2 fluid passing through the riser holes produces minimal forces on the nearest SG tube column, and the frequency of the normal operation flow through the holes does not coincide with a predominant structural mode of the adjacent SG tubes.
Therefore, the riser flow holes do not introduce structural integrity concerns due to FIV.
Audit Question A-5.1.3.4-1 The RVIupper riser assembly is fastened to the pressurizer baffle plate at the hanger plate. The CRDS shaft sleeves connect the hanger plate to the rest of the upper riser assembly.
During refueling and maintenance outages the upper riser assembly stays attached to the upper section of the NPM (upper CNV, upper RPV, and SG) while providing physical access for inspection to the RPV, SG, and upper riser assembly components.
The set of upper CRDS support plates in the upper riser assembly, in conjunction with the CRA guide tube support plate, CRA guide tubes, and upper core plate in the lower riser assembly align and provide lateral support for the control rod drive shafts. These component geometries ensure adequate alignment of the CRDS with the fuel assemblies and permit full insertion of control rods under design-basis events (DBEs).
3.9.5.2 Loading Conditions Section 3.9.1 describes acceptable analytical methods for Seismic Category I components and supports designated ASME BPVC,Section III, Division 1, Class CS. The plant and system operating transient conditions, including postulated seismic events and DBE, that provide the basis for the design of the RVI are provided in Section 3.9.3. Section 3.9.2 addresses the CVAP including the preoperational vibration test program plan for the RVI that is consistent with RG 1.20.
COL Item 3.9-6:
An applicant that references the NuScale Power Plant US460 standard design will develop a Reactor Vessel Internals Reliability Program to address industry identified aging degradation mechanism issues.
COL Item 3.9-7:
An applicant that references the NuScale Power Plant US460 standard design will provide a summary of reactor core support structure American Society of Mechanical Engineers (ASME) service level stresses, deformation, and cumulative usage factor values for each component and each operating condition in conformance with ASME Boiler and Pressure Vessel Code Section III Subsection NG.
3.9.5.3 Design Bases The RVI core support structures and internal structures are designed for the service loadings and load combinations shown in Table 3.9-7. The method of
NuScale Final Safety Analysis Report Summary Description NuScale US460 SDAA 5.1-4 Draft Revision 2 5.1.3.2 Reactor Coolant System Piping The RCS piping external to the RPV consists of the following lines:
two PZR spray line branches from a common spray header RPV high point degasification line RCS injection line RCS discharge line The RCS injection line has branch lines that connect to each of the ECCS valve reset valves.
Further description of the RCS piping is in Section 5.4, RCS Component and Subsystem Design.
5.1.3.3 Pressurizer The PZR is integral to the RPV and occupies the volume inside the RPV above the PZR baffle plate. The RCS components in the PZR volume are the PZR spray nozzles and the PZR heater assemblies. The PZR provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions at a temperature greater than THot for pressure control of the RCS during steady-state operations and transients. Maintaining the saturated conditions higher than THot ensures the reactor coolant remains subcooled during normal operations. The PZR also serves as a surge volume. The PZR controls reactor coolant pressure within the permitted operating range for normal operating transients without actuating the RSVs.
Further description of the PZR is in Section 5.4.5, Pressurizer.
5.1.3.4 Reactor Vessel Internals The RVI contain several sub-assemblies that provide support and alignment for the core, the control rod assemblies, the control rod drive shafts, and the instrument guide tubes. Additionally, the RVI channels reactor coolant flow from the reactor core to the SGs and back within the RPV.
Audit Question A-5.1.3.4-1 The RVI sub-assemblies include the core support assembly;, lower riser assembly;, upper riser assembly;, lower SG supports;, in-core instrumentation and riser level sensorinstrumentation guide tubes;, core support block assembly mounting brackets;, SG tube supports and tube support assemblies; lower SG supports, support clips, and backing strip; and flow diverter.
Audit Question A-5.1.3.4-1 Further description of the RVI are in Section 3.9, Mechanical Systems and Components, and Section 5.3, Reactor Vessel, and in Figure 5.4-5.
NuScale Final Safety Analysis Report Summary Description NuScale US460 SDAA 5.1-5 Draft Revision 2 5.1.3.5 Reactor Safety Valves Two direct spring operated RSVs connect to the top of the RPV upper head and discharge directly to the CNV. These valves are part of the RCPB and provide overpressure protection as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
Further description of the RSVs are in Section 5.2.2, Overpressure Protection.
5.1.3.6 Emergency Core Cooling Valves The ECCS valves consist of two RVVs and two RRVs. The RVVs connect to the upper head portion of the RPV and discharge the PZR steam space directly to the CNV. The RRVs connect to the RPV shell just above the main closure flange.
When opened, they permit recirculation of water in the CNV back into the RPV and ultimately through the core. The ECCS valves are a part of the RCPB and function during emergency core cooling operation. The RVVs also provide overpressure protection during operations at low temperature conditions.
Further description of the RVVs and RRVs are in Section 6.3, Emergency Core Cooling System.
5.1.3.7 Steam Generators Audit Question A-5.1.3.4-1 The SG system is an integral part of the RPV comprised of the SG tubes, SG tube supports, steam and feedwater piping inside containment, SG tube inlet flow restrictors, feed plenums, and steam plenums. The design contains two independent, but intertwined, SGs located inside the RPV that facilitate heat transfer to the secondary coolant system and provide redundancy for the DHRS.
The SGs are a once-through, helical-coil design, with primary-side reactor coolant outside the tubes and secondary-side fluid inside the tubes. On the primary side, the reactor coolant flows downward across the outside of the tubes, transferring heat to the fluid inside the SG tubes. On the secondary side, preheated feedwater enters the two feed plenums at the bottom of each SG, flows up the helical tubes where it is heated, boiled, and superheated, and exits the two steam plenums at the top of each SG. The DHRS connects to the steam and feed piping to permit use of the SG to remove decay heat from the primary coolant.
Further description of the SGs are in Section 5.4.1, Steam Generators.
5.1.4 System Evaluation To optimize performance of the NPM over the range of power levels within the design basis, steady state values for primary and secondary side parameters are established as a function of reactor power. The following criteria are considered to define optimal performance: maximizing electrical generation, using support systems efficiently, and providing margin to analytical limits. The following parameters are determined:
primary coolant temperatures, primary coolant flow rates, PZR water level, SG water
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-1 Draft Revision 2 5.3 Reactor Vessel A NuScale Power Module (NPM) consists of a reactor core, two steam generators (SGs),
and a pressurizer all contained within a single reactor pressure vessel (RPV), with a containment vessel (CNV) that surrounds the RPV. The NPM includes the piping located between the RPV and the CNV.
Audit Question A-5.1.3.4-1 The RPV is a pressure retaining vessel component of the reactor coolant system (RCS).
Section 5.1 and Section 5.2 describe the RCS and reactor coolant pressure boundary (RCPB). The RPV metal vessel that forms part of the RCPB is a barrier to the release of fission products. The RPV contains the reactor core, reactor vessel internals (including the SGs and SG tube supports), pressurizer, and reactor coolant volume. The RPV is supported laterally and vertically by the CNV. The RPV provides support and attachment locations for the control rod drive mechanisms (CRDMs), the CRDM seismic support structure, pressurizer heater bundles, in-core instrumentation, SG system piping, RCS piping, reactor safety valves, reactor vent valves, and reactor recirculation valves. The RPV is certified and stamped in accordance with Article NCA-8000 of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC)Section III.
The reactor vessel is in Figure 5.3-1 and design parameters are in Table 5.3-1.
5.3.1 Reactor Vessel Materials 5.3.1.1 Material Specifications The materials and applicable specifications used in the RPV and appurtenances are in Table 5.2-3.
Selection and fabrication of the RPV materials maintains RCPB integrity for the plant design lifetime. Selection of bolting materials, pressure retaining base materials, and weld filler materials are from the ASME BPVC Section II and comply with Article NB-2000 of ASME BPVC Section III (Reference 5.3-1). The austenitic stainless steel portion of the lower RPV has superior ductility and is less susceptible to the effects of neutron and thermal embrittlement, eliminating the need to calculate fracture toughness according to the requirements of 10 CFR 50, Appendix G. The ferritic low alloy steel of the upper RPV meets the fracture toughness requirements of 10 CFR 50, Appendix G. Reference 5.3-7 provides further details regarding the resistance to neutron and thermal embrittlement capability of the austenitic stainless steel material used in the lower RPV.
The RCPB materials comply with the relevant requirements of the following regulations:
GDC 1 and 30. The RPV design, fabrication, and testing meets ASME BPVC Class 1 in accordance with the Quality Assurance Program described in Chapter 17, Quality Assurance and Reliability Assurance.
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-2 Draft Revision 2
GDC 4. The RPV design and fabrication is compatible with environmental conditions of the reactor coolant and containment atmosphere (Reference 5.3-7).
GDC 14 and 31. The RPV design and fabrication has sufficient margin to assure the RCPB behaves in a non-brittle manner and minimizes the probability of rapidly propagating fracture and gross rupture of the RCPB (Reference 5.3-7).
GDC 15. The RPV design, fabrication, and testing meets ASME BPVC Class 1 requirements. Therefore, the design conditions of the RCPB are not exceeded during any condition of normal operation, including anticipated operational occurrences. (Reference 5.3-7).
GDC 32. Inspection of the RCPB is in Section 5.2.4. Section 5.3.1.6 discusses that a material surveillance program for the RPV is not required.
The design supports an exemption from the requirements of 10 CFR 50.60, which includes an exemption from 10 CFR 50, Appendix H (Reference 5.3-7).
10 CFR 50, Appendix G. The RPV materials meet applicable fracture toughness acceptance criteria. However, the design supports an exemption from the requirements of 10 CFR 50.60, which includes an exemption from 10 CFR 50, Appendix G (Reference 5.3-7). Section 5.3.1.5 provides further details.
Audit Question A-5.1.3.4-1 The RPV fabrication is in accordance with the requirements of ASME BPVC Section III, NB-4000. The reactor vessel internals, including the SG supports and SG tube supports (shown in Figure 5.4-5), are fabricated in accordance with ASME BPVC Section III, NG-4000. The RPV supports and CRDM seismic support structure fabrication is in accordance with ASME BPVC Section III, NF-4000.
5.3.1.2 Special Processes Used for Manufacture and Fabrication of Components Forged low alloy steel and austenitic stainless steel form the RPV assembly shells that surround the reactor core, pressurizer, and SGs. Forgings form the various required geometries with a minimum amount of welding.
Section 5.2.3, RCPB Materials, addresses the upper RPV cladding.
Measures are taken to prevent sensitization of austenitic stainless steel materials during component fabrication. Heat treatment parameters comply with ASME BPVC Section II. Water quenching cools the austenitic stainless steel materials to avoid carbide formation at the grain boundaries; alternatively, cooling through the sensitization temperature range occurs quickly enough to avoid carbide formation at the grain boundaries. When means other than water quenching are used, corrosion testing in accordance with Practice A or E of American Society for Testing and Materials (ASTM) A262 (Reference 5.3-3) verifies nonsensitization of the base material.
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-11 Draft Revision 2 The SG weld filler metals are in Table 5.4-3 and are in accordance with ASME BPVC Section II, Part C.
Audit Question A-5.1.3.4-1 The SG upper supports are designated as ASME BPVC,Section III, Subsection NF Class 1. The SG lower supports and SG tube supports are designated as ASME BPVC,Section III, Subsection NG. Internal Structures. The design, fabrication, construction, and testing of the SG supports and SG tube supports, including weld materials does not adversely affect the integrity of the core support structures.
The SG piping structural supports, including weld materials, conform to fabrication and construction requirements of ASME BPVC,Section III, Subsection NF. The SG piping structural support materials are in Table 5.4-3.
The SG inlet flow restrictors are non-structural attachments to the RPV. The SG inlet flow restrictors design, fabrication, construction, testing, and inspections conform with the ASME BPVC,Section III, Subsection NC.
Section 5.2.3, RCPB Materials, contains additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 1 components. Section 5.2.3.4.2, Cleaning and Contamination Protection Procedures, describes cleaning and cleanliness controls for the SGs.
Section 6.1, Engineered Safety Feature Materials, has additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 2 components.
Section 3.13 describes threaded fasteners.
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 based on NEI 97-06 (Reference 5.4-1) and Regulatory Guide (RG) 1.121 and is documented in the technical specifications.
The program implements applicable portions of Section XI of the ASME BPVC and specifically addresses 10 CFR 50.55a(b)(2)(iii). Appendix B to 10 CFR 50 applies to implementation of the SG program.
Historically, significant SG tube degradation in the operating PWR SG fleet was due to various corrosion mechanisms, including wastage and both primary and secondary side stress corrosion cracking. These corrosion mechanisms relate to materials selection, plant chemistry control, and control of the ingress of impurities and corrosion products to the SGs. In the design, detrimental SG corrosion mitigation is achieved by use of SB-163 UNS N06690 SG tubing, application of EPRI primary and secondary plant chemistry control guidelines, and design of
License Conditions; ITAAC Module-Specific Structures, Systems, and Components Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Design Descriptions and ITAAC NuScale US460 SDAA 24 Draft Revision 2 Audit Question A-5.1.3.4-1 Table 2.1-4: NuScale Power Module Mechanical Equipment Equipment Name Equipment Identifier ASME Code Section III Class Valve Actuator Type Containment Isolation Valve Reactor Coolant System RCS integral RPV/SG/Pressurizer RPV-VSL-0001 1
N/A N/A RVI upper core plate N/A CS N/A N/A RVI core barrel N/A CS N/A N/A RVI reflector blocks N/A CS N/A N/A RVI lower core plate N/A CS N/A N/A RVI uppercore support blocks N/A CS N/A N/A RVI core support mounting brackets N/A CS N/A N/A Upper SG supports N/A 1
N/A N/A RCS reactor safety valves RCS-PSV-0003A RCS-PSV-0003B 1
N/A No SG #1 relief valve SG-RV-0102 2
N/A Yes SG #2 relief valve SG-RV-0202 2
N/A Yes RPV instrument seal assemblies RPV39 RPV40 RPV41 RPV42 1
N/A N/A Emergency Core Cooling System RVVs ECC-POV-0001A ECC-POV-0001B 1
Hydraulic No RVV trip valves ECC-SV-0101A-1 ECC-SV-0101A-2 ECC-SV-0101B-1 ECC-SV-0101B-2 1
Solenoid No RVV reset valves ECC-SV-0103A ECC-SV-0103B 1
Solenoid No RVV valve venturis ECC-FV-0001A ECC-FV-0001B 1
N/A No RRVs ECC-POV-0002A ECC-POV-0002B 1
Hydraulic No RRV trip valves ECC-SV-0102A-1 ECC-SV-0102A-2 ECC-SV-0102B-1 ECC-SV-0102B-2 1
Solenoid No ECCS RRV reset valves ECC-SV-0104A ECC-SV-0104B 1
Solenoid No RRV valve venturis ECC-FV-0002A ECC-FV-0002B 1
N/A No Containment System CNV CNT-VSL-0001 1
N/A N/A RPV high point degas solenoid valve CVC-SV-0404 3
Solenoid No PZR spray flow check valve CVC-CKV-0323 3
N/A No CVCS injection flow check valve CVC-CKV-0329 3
N/A No CVCS discharge air operated valve CVC-AOV-0336 3
Air No CVCS injection inboard CIV CVC-HOV-0331 1
Electro-hydraulic Yes CVCS injection outboard CIV CVC-HOV-0330 1
Electro-hydraulic Yes PZR spray inboard CIV CVC-HOV-0325 1
Electro-hydraulic Yes PZR spray outboard CIV CVC-HOV-0324 1
Electro-hydraulic Yes
NuScale Final Safety Analysis Report Summary Description NuScale US460 SDAA 5.1-4 Draft Revision 2 5.1.3.2 Reactor Coolant System Piping The RCS piping external to the RPV consists of the following lines:
two PZR spray line branches from a common spray header RPV high point degasification line RCS injection line RCS discharge line The RCS injection line has branch lines that connect to each of the ECCS valve reset valves.
Further description of the RCS piping is in Section 5.4, RCS Component and Subsystem Design.
5.1.3.3 Pressurizer The PZR is integral to the RPV and occupies the volume inside the RPV above the PZR baffle plate. The RCS components in the PZR volume are the PZR spray nozzles and the PZR heater assemblies. The PZR provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions at a temperature greater than THot for pressure control of the RCS during steady-state operations and transients. Maintaining the saturated conditions higher than THot ensures the reactor coolant remains subcooled during normal operations. The PZR also serves as a surge volume. The PZR controls reactor coolant pressure within the permitted operating range for normal operating transients without actuating the RSVs.
Further description of the PZR is in Section 5.4.5, Pressurizer.
5.1.3.4 Reactor Vessel Internals The RVI contain several sub-assemblies that provide support and alignment for the core, the control rod assemblies, the control rod drive shafts, and the instrument guide tubes. Additionally, the RVI channels reactor coolant flow from the reactor core to the SGs and back within the RPV.
Audit Question A-5.1.3.4-1 The RVI sub-assemblies include the core support assembly;, lower riser assembly;, upper riser assembly;, SG supports, in-core instrumentation and riser level sensorinstrumentation guide tubes;, core support block assembly mounting brackets;, SG tube supports and tube support assemblies; and lower SG supports, support clips, and backing stripand flow diverter.
Audit Question A-5.1.3.4-1 Further description of the RVI are in Section 3.9, Mechanical Systems and Components, and Section 5.3, Reactor Vessel, and in Figure 5.4-5.
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-5 Draft Revision 2 The SG design permits periodic inspection and testing of critical areas and features to assess their structural and pressure boundary integrity when the NPM is disassembled for refueling as shown in Figure 5.4-3. The internal surface of SG tubes is accessible over their entire length for application of nondestructive examination methods and techniques that are capable of finding the types of degradation that may occur over the life of the tubes. Individual SG tubes may be plugged and, if necessary, stabilized to prevent adverse interaction with non-plugged tubes. Access to the internal (secondary) and external (primary) sides of tubesheets affords opportunity for inspection, and for removal of foreign objects. Figure 5.4-3 and Figure 5.4-4 contain illustrations of the steam and feed plena inspection ports.
Classifications and Quality Group designations for design, fabrication, construction and testing of SGS components that form part of the RCPB are in Table 3.2-2. Chapter 3 provides detailed information regarding the design basis and qualification of structures, systems, and components based on these classifications and designations. Figure 6.6-1 shows the ASME BPVC Section III, Class 1 and 2 boundaries for the SGS.
Steam Generator Tube Supports and Steam Generator Supports Audit Question A-5.1.3.4-1 The seamless helical coil SG tubing is supported by a series of austenitic stainless steel tube supports and tube support assemblies. The geometric design and materials utilized facilitate fluid flow while minimizing the potential for the generation of corrosive products and buildup. The material and geometry choice precludes two of the most significant historical contributors to tube degradation by the tube supports.
Audit Question A-5.1.3.4-1, Audit Question A-5.4-1, Audit Question A-5.4-6 The tube supports assemblies are between each column of helical tubes as shown in Figure 5.4-6, with a tube support assembly between the outermost column of tubes and the RPV inner wall. The tube support assembly consists of a tube support, an SG tube support spacer, and a socket head cap screw connecting the tube support to the spacer. The SG support spacer sits between the outermost tube support and the RPV inner wall. There is a backing strip between the upper riser assembly and the innermost column of tubes that completes the enclosure of the innermost tubes and functions as the interface between the upper riser assembly and the tube support. The tube support structure is within the primary coolant environment; therefore, no ingress path exists for general corrosion products from the secondary system to deposit on the primary side of the SG. Optimization of the circumferential spacing of the tube supports provides the minimum possible tube free span lengths while still accommodating the transition of the tubes to the steam and FW tube sheets.
Audit Question A-5.1.3.4-1, Audit Question A-5.4-1 The SG supports and SG tube supports provide support for vibration and seismic loads. As shown in Figure 5.4-5, the SG tube supports and tube support assemblies attach to upper SG supports welded to the PZRintegral steam baffle
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-6 Draft Revision 2 plate and inner surface of the RPV, and also interface with lower SG supports welded to the inner surface of the RPV. The use of eight sets of tube supports assemblies limit the unsupported tube lengths, which ensures the SG tubes do not experience unacceptable flow-induced vibration (FIV). Figure 5.4-1 shows two of the eight sets of tube supports.
As shown in Figure 5.4-5, the lower SG supports permit thermal growth and provide lateral support of the tube supports.
Inlet Flow Restrictors The SG inlet flow restrictors are installed in each SG tube at the FW plena locations. Each SG inlet flow restrictor is individually installed and seats against the secondary face of the FW plenum tubesheet and extends into a portion of the hydraulically expanded SG tube within the FW plenum tubesheet. A SG inlet flow restrictor consists of a mandrel, an expanding collet, a flanged sleeve, a locking plate and a hex nut. After the flow restrictor is inserted into the SG tube, the metallic collet on each SG inlet flow restrictor is expanded to seal with the inner diameter of the SG tube. The bearing contact resistance between the expanded collet and tube prevents bypass flow around the flow restrictor as well as the frictional interaction for securing the flow restrictor within the FW plenum.
Secondary side water flows from the FW plenum through a center-flow orifice in the mandrel. The flanged sleeve allows secondary side water from the feed plenum to enter into the space between the sleeve and SG tube to FW plenum tubesheet weld. This secondary side water provides a thermal barrier to the tube-to-tubesheet weld, helping mitigate rapid temperature changes at the weld.
The devices permit in service tube inspections, cleaning, tube plugging, repairs and maintenance activities via installation and removal as needed.
Thermal Relief Valves A single thermal relief valve is on each FW line upstream of the tee that supplies the feed plenums (Figure 5.4-7). The thermal relief valves provide overpressure protection during shutdown conditions for the secondary side of the SGs, FW and steam piping inside containment, and the DHRS when the secondary system is water solid for SG flushing operations and the containment isolation system is actuated. The trapped fluid is subject to heating by core decay heat. The thermal relief valves are spring operated relief valves that vent directly to containment.
The thermal relief valves are classified as Seismic Category I and Quality Group B (ASME Class 2), and designed, fabricated, constructed, tested and inspected in accordance with Section III of the ASME BPVC. The pressure-retaining materials of thermal relief valves are in accordance with the materials identified in Table 6.1-4.
The thermal relief valves protect the secondary system components during off-normal conditions. The system design pressure and the RSVs provide overpressure protection during normal operation. Section 5.2.2, Overpressure Protection, contains details.
NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-11 Draft Revision 2 Welding of the RCPB portions of the SGS with the steam access port components follows procedures qualified in accordance with the applicable requirements of ASME BPVC Section III, Subarticle NB-4300 and Section IX. Welding of the secondary side portions of the SGS constructed to Class 2 follows procedures qualified in accordance with the applicable requirements of ASME BPVC,Section III, Subarticle NC-4300 and Section IX.
The secondary side surfaces of the steam plenum tubesheet, and feed plenum tubesheet use alloy 52/152 cladding. The remaining inside and outside surfaces of the steam plenum and feedwater plenum are Alloy 690 material or low alloy steel clad with austenitic stainless steel.
The SG weld filler metals are in Table 5.4-3 and are in accordance with ASME BPVC Section II, Part C.
Audit Question A-5.1.3.4-1 The SG upper supports are designated and constructed as ASME BPVC,Section III, Subsection NF Class 1. The SG lower supports and SG tube supports are designated as ASME BPVC,Section III, Subsection NG. Internal Structures. The design, fabrication, construction, and testing of lowerthe SG supports and SG tube supports, including weld materials, does not adversely affect the integrity of the core support structures.
The SG piping structural supports, including weld materials, conform to fabrication and construction requirements of ASME BPVC,Section III, Subsection NF. The SG piping structural support materials are in Table 5.4-3.
The SG inlet flow restrictors are non-structural attachments to the RPV. The SG inlet flow restrictors design, fabrication, construction, testing, and inspections conform with the ASME BPVC,Section III, Subsection NC.
Section 5.2.3, RCPB Materials, contains additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 1 components. Section 5.2.3.4.2, Cleaning and Contamination Protection Procedures, describes cleaning and cleanliness controls for the SGs.
Section 6.1, Engineered Safety Feature Materials, has additional description of material compatibility, fabrication and process controls, and welding controls related to the ASME Class 2 components.
Section 3.13 describes threaded fasteners.
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 based on NEI 97-06 (Reference 5.4-1) and Regulatory Guide (RG) 1.121 and is documented in the technical specifications.