ML24011A133

Sdaa Chapter 10, Steam and Power Conversion System, SE with No Open Items

ML24011A133
Person / Time
Issue date:	03/11/2024
From:	Hayden T NRC/NRR/DNRL/NRLB
To:	NuScale
Haydne, T., NRR/DNRL,
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ML24011A132	List: ML24011A133
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Download: ML24011A133 (36)

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10 STEAM AND POWER CONVERSION SYSTEM

This chapter of the safety evaluation report (SER) documents th e U.S. Nuclear Regulatory Commission (NRC) staffs review of Chapter 10, Steam and Power Conversion System, of the NuScale Power, LLC (the applicant), US460 Standard Design Appro val Application (SDAA),

Part 2, Final Safety Analysis Report (FSAR). The staffs regu latory findings documented in this report are based on Revision 1 of the SDAA, dated October 31, 2 023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23306A033). In the SDAA, the applicant used the English system of measure to provide the pre cise parameter values. Where appropriate, the NRC staff converted these values for presentat ion in this safety evaluation to the International System (SI) units of measure based on the NRC s standard convention. In these cases, the SI converted value is approximate and is prese nted first, followed by the applicant-provided parameter value in English units within pare ntheses. If only one value appears in either SI or English units, it is directly quoted fr om the SDAA and not converted.

The steam and power conversion sy stem removes thermal energy from the reactor coolant system and transfers it to the main turbine generator where it is converted into electric energy.

The main elements of the steam and power conversion system incl ude the main steam, turbine generator, turbine bypass, air-cooled condenser, condensate pol ishing, feedwater treatment, condensate and feedwater, and auxiliary boiler systems. In gene ral, the steam and power conversion systems are not safety related and are not required for safe shutdown. However, the main steam and main feedwater system have piping that penetrate s the containment and components that directly interface with safety-related structur es, systems, and components (SSCs). The failure of these components can adversely impact pl ant safety and the plants ability to achieve a safe shutdown. In addition, failure of som e system equipment may result in the potential for internal flooding or the creation of missiles that may have the potential to adversely impact SSCs important to safety.

Using a graded approach, the staffs primary review of the powe r conversion systems focused on the safety-related piping and the system components that are part of the main steam and main feedwater systems that support containment isolation and t he operation of the safety-related decay heat removal system (DHRS). However, as in dicated above, the staff also recognizes that the failure of some portion of the power conver sion system may have the potential to adversely impact SSCs that are important to safety. Since the power conversion system functions are not relied on to support the safety-relate d and safe-shutdown aspects of the NuScale plant, the staff reviewed the design to verify that failure of the non-safety-related, non-risk-significant systems will not adversely affect the plan ts ability to achieve and maintain safe shutdown or result in excessive releases of radioactivity to the environment.

In addition, although applicants for standard design approvals (SDA) are not required to submit plans for an initial test program, Regulatory Guide (RG) 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, acknowledges that SDA applicants have previously submitted initial test program plans to assist a future combined license (COL) applicant referencing the SDA in meeting the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 52.79(a)(28). In each subsection of this chapter, the staff lists related tests from the initial test program that were evaluated as part of the SDAA re view and reviewed in FSAR Section 14.2, Initial Plant Test Program.

10-1 10.1 Summary Description

FSAR Section 10.1, Summary Description, contains an introduct ory account of the steam and power conversion system and summarizes the protective features incorporated in its design.

FSAR Figure 10.1-1, Power Conversion System Block Flow Diagram, provides a high-level flow diagram of the system. FSAR Figure 10.1-2, Heat Balance D iagram, shows the steam/condensate flow stream for the power conversion system. F SAR table 10.1.-1 and 10.1-2, provides representative values for the pressure, temperature, enthalpy, and flow at various point of the power conversion system. FSAR table 10.3-1 shows s ome of the major power and conversion system operating parameters at rated thermal power. These parameters, along with the secondary -side heat balance diagram, provide the basis for the power conversion system design.

FSAR sections 10.2 through 10.4.9 describe the main elements of the steam and power conversion system. Sections 10.2 through 10.4.9 of this SER doc ument the staffs review.

10.2 Turbine Generator

10.2.1 Introduction

The NuScale power plant comprises up to six individual NuScale power modules (NPMs), each of which has its own turbine generator and its own turbine cont rol system.

The turbine generator system (TGS) is not safety related. The T GS converts the energy of the steam produced in the steam generators (SGs) into mechanical sh aft power and then into electrical energy. The TGS is not credited for mitigation of de sign-basis events (DBEs) and has no safe-shutdown functions, but a failure of the TGS may result in the generation of turbine missiles with the potential to adversely affect SSCs that are i mportant to safety.

10.2.2 Summary of Application

FSAR Section 10.2, Turbine Generator, describes the TGS. The information provided includes the TGS design bases, system descriptions, component descriptio ns, and TGS control and protection systems. FSAR Section 14.2, Initial Plant Test Progr am, contains information on TGS initial testing (Test # 29, Turbine Generator System). There are no proposed inspections, tests, analyses, and acceptance criteria (ITAAC) or specific te chnical specification (TS) requirements associated with the TGS.

10.2.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are discussed in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Section 10.2, Turbine Generator, Revision 3, issued March 2007, and are summarized b elow:

• 10 CFR Part 50, Domestic Licensing of Production and Utilizat ion Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 4, Environmental and dynamic effects design ba ses, requires, in part, the protection of SSCs important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles.

10-2 The NRC staffs review of the TGS review interfaces with the re view of certain other matters addressed in other SRP sections, identified in Section I, Area s of Review, of SRP Section 10.2.

10.2.4 Technical Evaluation

Since the TGS is not a safety-related or risk-significant syste m, the staff conducted its review using the enhanced safety-focus review approach as discussed in the SRP Introduction, Part 2.

The inability of the TGS to perform its normal intended functio n will have no direct effect on plant safety or the ability of the plant to achieve and maintai n a safe-shutdown condition.

Therefore, the staffs review of the TGS focused on how the sys tem design would preclude TGS failure from adversely affecting SSCs important to safety and h ow compliance with GDC 4 will be ensured in events in which TGS failures may result in the ej ection of turbine missiles due to excessive turbine overspeed.

The staffs evaluation of the TGS is based on the information i n the applicants SDAA. A general description of the TGS is provided in SDAA Part 2, Section 10.2.2, System Description, as well as in FSAR Figure 10.2-1, Turbine Generator System Schematic, and TGS design parameters appear in FSAR Table 10.2-1, Turbine Generator Design Details. The staff followed the guidance in SRP Section 10.2 in its review of the TGS.

10.2.4.1 GDC 4, Environmental and Dynamic Effects Design Bases

Although the TGS is not safety related, missiles generated by t urbine failure can adversely affect the integrity of SSCs important to safety. To satisfy GD C 4, and as discussed in section 3.5.1.3 of this safety evaluation, the turbine must hav e a low probability of rotor failure to minimize the likelihood that turbine missiles will affect SSCs important to safety. Alternatively, the applicant may propose to install barriers, or take credit f or existing structures or features such as barriers, to reduce or eliminate turbine missile hazard s to equipment, provided that the barriers meet acceptance criteria described in SRP Section 3.5. 3, Barrier Design Procedures.

The arrangement and the orientation of the TGS relative to thes e essential SSCs are also to be considered in the overall minimization of turbine missiles.

The staff reviewed Section 3.5.1.3, Turbine Missiles, and Sec tion 3.5.2, Structures, Systems, and Components to be Protected from External Missiles, of the NuScale FSAR to determine if GDC 4 compliance was achieved based on plant layout or the inco rporation in the plant design of missile barriers to protect essential SSCs from turbine miss iles. FSAR section 3.5.2 states that the reactor building (RXB) and control building (CRB) belo w the 30 foot above-grade threshold are designed to withstand all design-basis missiles d iscussed in the FSAR, Section 3.5.1.3 and Section 3.5.1.4, Missiles Generated by Tor nadoes and Extreme Winds which includes turbine missiles.

As indicated in FSAR Section 3.5.1.3, the NuScale plant uses ba rriers in lieu of turbine rotor integrity and turbine missile generation probability as the bas is for justifying adequate protection from turbine missiles. Since the protection of SSCs from turbin e missiles will no longer depend on minimizing turbine missile generation probability and the ov erspeed protection system to prevent destructive overspeed conditions, review of the overspe ed protection system as it relates to ensuring compliance with GDC 4 is not necessary. Sec tion 3.5.1.3 of this SER includes the staffs review of compliance with GDC 4 as it rela tes to turbine missiles.

Based on the information provided in FSAR Sections 10.2, 3.5.1. 3, and 3.5.2, the staff finds that the NuScale design complies with GDC 4, with respect to the des ign and operation of the

10-3 turbine generator, because all SSCs important to safety are hou sed in the RXB and the CRB, and the applicant indicated that these buildings are designed t o protect against turbine missiles.

Section 3.5.1.3 of this SER includes the staffs review of comp liance with GDC 4 as it relates to turbine missiles.

10.2.5 Initial Test Program

The preoperational test related to the TGS is TGS Test # 29, wh ich ensures that the various design aspects related to the TGS are implemented. The test is performed in accordance with FSAR Table 14.2-29, Test # 29 Turbine Generator System. Secti on 14.2 of this SER documents the staffs evaluation of the initial test program fo r the NuScale SDAA.

10.2.6 Conclusion

Based on its review and as discussed above in the technical eva luation section, the staff concludes that the applicant has met the requirements of GDC 4 for FSAR Section 10.2.

10.3 Main Steam System

10.3.1 Main Steam System

10.3.1.1 Introduction

The main steam system (MSS) transfers steam produced in the SGs to the TGS. The NuScale design defines the MSS as only the portions from the flanges im mediately downstream of the containment system (CNTS) main steam isolation valves (MSIVs) u p to the turbine stop valves.

Portions of the main steam piping inside containment are identi fied as part of the SGs, and the portion upstream of the main steamline flange, including the MS IVs and main steam isolation bypass valves (MSIBVs), is identified as part of the CNTS. Rega rdless of how the NuScale design defines the MSS, the staff performed its review consiste nt with the system boundaries defined in NuScale Design Specific Review Standard (DSRS), Sect ion 10.3, Main Steam Supply System, Revision 0, issued June 2016 (ML15355A322). For the purposes of this review, the staff considers the MSS to extend from the outlet o f the reactor pressure vessel steam plenum (on the secondary side of the SGs) up to and inclu ding the turbine stop valves.

Such a system includes the containment isolation valves, connec ted piping that is 6.4 centimeters (2.5 inches) in nominal diameter or larger, and the steamline to the decay heat removal system (DHRS) up to the DHRS actuation valves.

10.3.1.2 Summary of Application

FSAR Section 10.3, Main Steam System, contains the MSS design basis and system and component descriptions, as well as system operation, inspection s, and testing information. The portion of the MSS located inside containment is described in F SAR Section 5.4, Reactor Coolant System Component and Subsystem Design, and FSAR Sectio n 6.2, Containment Systems, discusses MSS SSCs associated with containment isolat ion.

There is no proposed system ITAAC for the MSS. However, in SDAA Part 8, License Conditions; Inspections, Tests, Analyses & Acceptance Criteria (ITAAC), the applicant proposes ITAAC for the MSIVs, secondary MSIVs, and secondary MS IBVs. Section 14.3.7 of this SER evaluates these ITAAC.

10-4 FSAR section 14.2 (Test # 24, Main Steam System) provides inf ormation on MSS initial testing. Specific TS requirements associated with the MSS are p rovided in FSAR Chapter 16, Technical Specifications, Revision 0, and SDAA Part 4, US460 Generic Technical Specifications, Volumes 1 and 2, Revision 0.

10.3.1.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are discussed in NuScale DSRS section 10.3 and are sum marized below. Review interfaces with other SRP sections also can be found in Item I, Review Interfaces, of DSRS section 10.3.

• GDC 2, Design bases for protection against natural phenomena, requires that SSCs important to safety be designed to withstand the effects of nat ural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiche s without loss of capability to perform their safety functions.

• GDC 4, Environmental and dynamic effects design bases, requi res that SSCs important to safety be designed to accommodate the effects of a nd to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

• GDC 5, Sharing of structures, systems, and components, requi res that SSCs important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to per form their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

• 10 CFR 50.63, Loss of all alternating current power, require s that a nuclear power plant have the ability to withstand for a specified duration and reco ver from a station blackout (SBO) as defined in 10 CFR 50.2.

• GDC 34, Residual heat removal, requires that a system to rem ove residual heat be provided.

• 10 CFR 20.1406, Minimization of contamination, as it relates to the design features that will facilitate eventual decommissioning, requires that SDA app licants describe how facility design will minimize, to the extent practicable, conta mination of the facility and the environment, facilitate eventual decommissioning, and minim ize, to the extent practicable, the generation of radioactive waste.

10.3.1.4 Technical Evaluation

The staff reviewed the MSS design, described in FSAR section 10.3, in accordance with guidance in DSRS section 10.3, to evaluate whether the design c omplies with the regulatory requirements listed in section 10.3.1.3 of this SER.

Each NPM has two SGs and a dedicated MSS. Each of the two SGs h as two steamlines that combine before exiting containment and ultimately terminate in the turbine generator building (TGB) at the turbine stop valve. FSAR figure 10.1-1 provides a flow diagram of the system, and FSAR Table 10.3-4, Classification of Structures, Systems, and Components, presents

10-5 information on the location, saf ety and system classifications, and any augmented design requirement for MSS components.

The primary function of the MSS is to transport steam from the SGs to the TGS. Other functions of the MSS include delivering steam to the gland seal regulator, delivering steam directly to the condenser through turbine bypass, collecting the drainage conde nsed in the main steam piping and delivering it to the condensate collection tank, and transp orting extraction steam from the turbine to the feedwater heaters. The safety-related portions o f the MSS are the piping and valves between each reactor pressure vessel and the MSIVs. The remainder of the MSS, including the turbine generator, is not safety related. Under a ccident conditions, the CNTS isolates the SGs and the safety-related portion of the system f rom the portion that is not safety related. As specified in SDAA section 10.3.2, the MSS piping is protected from overpressure by Main Steam Safety Valves. Overpressure protection is discussed in SDAA section 5.2.2 and reviewed in the corresponding section of this SER.

An evaluation of the MSS abnormal and anticipated operational o ccurrences is described in FSAR Chapter 15, Transient and Accident Analysis, where the N uScale responses to postulated accidents are considered, including an evaluation of a main steamline break, feedwater line break, and SG tube rupture. Chapter 15 of this S ER provides the staffs evaluation of the transient and accident analyses for the MSS.

In FSAR Section 10.3.3, Safety Evaluation, the applicant prov ided its evaluation of the MSS and how it complies with the requirements of the GDC identified in the regulatory basis for this section of the SER. Following is the staffs review of complian ce with these GDC.

10.3.1.4.1 GDC 2, Design Bases for Protection against Natural Phenomena

The staff reviewed the MSS to evaluate whether the design compl ies with the requirements of GDC 2, with respect to its design for protection against the ef fect of natural phenomena such as earthquakes, tornados, and hurricanes. Protection against inte rnal flooding is evaluated in Section 3.4.1 of this SER. Compliance with the requirements of GDC 2 is based on the MSS being designed to withstand the effects of natural environmenta l phenomena without losing the ability to perform its safety function and on adherence to Regu latory Position C.1 of RG 1.29, Seismic Design Classification for Nuclear Power Plants, Revis ion 6, issued July 2021, for the safety-related portion of the system, and Regulatory Position C.2 for the portions of the system that are not safety related. The staff reviewed FSAR sections 5.4, 6.2, and 10.3 to determine whether the portions of the main steamline important to safety are protected against natural phenomena.

FSAR figure 10.1-1 indicates that the main steamline, including piping and valves between each SG and flange immediately downstream of the MSIVs (identified a s removable spool piece in figure 10.1-1), is classified as seismic 1 and located inside t he RXB. FSAR Section 10.3.3 states that the portions of the MSS downstream of the MSIVs to the secondary MSIVs are designed to remain functional during and after a safe-shutdown earthquake and meet the guidelines of RG 1.29. It also states that the RXB is designed as an engineered barrier to withstand a postulated design -basis missile as discussed in Se ction 3.7. This design satisfies the criteria of GDC 2 by the proper design and use of missile b arriers to protect essential SSCs against potential missiles generated by tornado or hurricane wi nds.

NuScale DSRS section 10.3, subsection III, item 4, indicates th at the essential portions of the MSS should be designed to Quality Group B or seismic Category I requirements. The staff

10-6 reviewed FSAR section 6.2, figure 10.1-1, and table 10.3-4 and confirmed that SSCs important to safety, including the main steam isolation and backup main s team valves and the spool piece from which the system is disconnected during refueling, are loc ated in the RXB. In FSAR Section 6.2.4.2.1, General Description, the MSIVs, the main s team backup isolation valves, and the feedwater isolation valves are classified as single sec ondary system containment isolation valves and as seismic Category I, Quality Group B, co mponents capable of remote operation from the control room, as described in SDAA section 6.2.4.2.2.3. The secondary MSIV and the main steam backup isolation valves are also seismi c Category I components as indicated in FSAR Table 10.4-4, Classification of Structures, Systems, and Components.

Based on its review, the staff finds that the MSS design compli es with GDC 2, because the RXB is designed to protect essential SSCs against potential missile s generated by tornado or hurricane winds, and the essential portions of the MSS are desi gned to seismic Category I requirements, in accordance with the guidance of NuScale DSRS s ection 10.3.

10.3.1.4.2 GDC 4, Environmental and Dynamic Effects Design Bases

The staff reviewed the MSS to evaluate whether the design compl ies with the requirements of GDC 4, as related to adverse environmental phenomena and dynamic effects associated with possible fluid flow instabilities, including steam hammer and e ffects of pipe breaks. Compliance with the requirements of GDC 4 is based on the important-to-saf ety SSCs being adequately protected from environmental phenomena or the dynamic effects t hat may result from such phenomena, including water and steam hammer, pipe whip, and jet impingement.

The staff reviewed the NuScale design to evaluate whether the f unctions important to safety will be maintained in the event of adverse environmental phenomena a nd dynamic effects. FSAR section 10.3.3 states that the portions of the MSS downstream o f the MSIVs to the secondary MSIVs are protected from pipe whip and jet impingement forces r esulting from breaks in nearby systems (including the MSS of adjacent power modules) by the pi ping design layout. The portions of the MSS downstream of the MSIVs to the secondary MS IVs are physically separated from safety-related systems in the RXB using walls and other re straints and have no adverse impacts on safety functions. The staff reviewed the above information and finds that the portions of the main steamlines and MSS that are subject to protection u nder GDC 4 are located inside the RXB, a seismic Category I structure designed for wind and m issile loads, and therefore are acceptable.

Further, regarding the GDC 4 requirements, the staff reviewed t he applicants consideration of steam and water hammer effects on the MSS. FSAR section 10.3.3 states that the design uses drain pots, line sloping, and drain valves to minimize the effe cts of dynamic loads and water hammer. These design features are consistent with those identif ied in NUREG-0927, Evaluation of Water Hammer Occurrence in Nuclear Power Plants, Revision 1, issued March 1984, as being effective in minimizing the frequency and severity of water and steam hammer occurrences. Therefore, the staff finds that the MSS SSC s important to safety are adequately protected from adverse environmental phenomena and d ynamic effects associated with possible fluid flow instabilities.

10.3.1.4.3 GDC 5, Sharing of Structures, Systems, and Components

GDC 5 contains provisions restricting the sharing of SSCs impor tant to safety between nuclear power units. The applicant stated that each NPM has SGs and a d edicated MSS. There are no important-to-safety components in the MSS that are shared among NPMs; therefore, the loss of

10-7 components in one MSS does not significantly impair the ability of other NPMs to perform their safety functions. Therefore, the requirements of GDC 5 are met.

10.3.1.4.4 GDC 34, Residual Heat Removal

FSAR Section 3.1.4.5, Criterion 34 Residual Heat Removal, sta tes that the power provisions of GDC 34 are not applicable to the NuScale design. NuScale adopt ed certain principal design criteria (PDC) including PDC 34, which states:

A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other resid ual heat from the reactor core at a rate such that specified acceptable fuel desi gn limits and the design conditions of the reactor coolant pressure boundary [RCP B] are not exceeded.

Suitable redundancy in components and features, and suitable in terconnections, leak detection, and isolation capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single f ailure.

Sections 5.4 and 8.3 of this SER contain the staffs evaluation of the designs compliance with PDC 34.

The staff reviewed the MSS design against the decay and residua l heat removal safety function requirements of PDC 34. The applicant stated that the decay and residual heat removal safety function required by PDC 34 is performed by the DHRS flowpath, and the containment isolation function of the CNTS is performed by the MSIVs and the feedwate r isolation valves (FWIVs).

Consistent with PDC 34, the secondary MSIVs that are not safety related downstream of the MSIVs are credited as backup isolation components if an MSIV fa ils to close and provide additional assurance that the blowdown of a second SG is limite d if a steamline breaks upstream of the MSIV. As discussed above in section 10.3.1.4 of this SE, the staff found that the treatment of the secondary MSIVs and secondary MSIBVs is co nsistent with the previous staff position on the treatment of equipment that is not safety related when applying single-failure criteria for steamline break accidents as shown in NURE G-0138, Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3, 19 76 Memorandum from Director, NRR to NRR Staff, issued November 1976 (ML13267A423).

Based on its review as discussed above, the staff finds that th e MSS, including the steamlines between the SGs and disconnect flange, conforms to the requirem ents of PDC 34 with respect to the system function of transferring residual and sensible he at from the reactor coolant system, assuming a single failure.

10.3.1.4.5 10 CFR 50.63, Loss of All Alternating Current Power

The staff reviewed the MSS capability to supply steam to the DH RS for the removal of decay heat during an SBO. Successful operation of the DHRS relies on the MSIVs ability to isolate steam; this forms part of the DHRS flowpath and pressure bounda ry. In addition, secondary MSIVs that are not safety related are provided as a backup to t he MSIVs. Both the MSIVs and secondary MSIVs fail closed during an SBO.

The staff finds this acceptable because the safety-related main steam components are designed such that they perform their safety function, and the system ha s sufficient capability to cope with

10-8 an SBO. Therefore, the staff concludes that the requirements of 10 CFR 50.63 are met.

Section 8.4 of this SER further documents the staff evaluation of the SBO event.

10.3.1.4.6 10 CFR 20.1406, Minimization of Contamination

The regulation in 10 CFR 20.1406 requires, in part, that each s tandard design approval applicant describe how the facility design and procedures for o peration will minimize, to the extent practicable, contamination of the facility and environme nt, as well as the generation of radioactive waste. The MSS is not normally a significant radiat ion hazard in a pressurized-water reactor (PWR), and it is only in the unlikely event of a primar y to secondary system leak or SG tube failure that the steam would become contaminated. FSAR Tab le 12.3-26, Regulatory Guide 4.21 Design Features for Main Steam System, lists the de sign features specific to the MSS for the minimization of contamination. Examples of these fe atures include the use of corrosion-resistant materials, proper chemistry controls, steam line radiation monitors, and collection and control of fluid leaks by the radioactive waste drain system and balance-of-plant drain system (BPDS). Also, a minimum of two barriers are provided between clean systems (nonradioactive systems), such as the nitrogen distribution sys tem and the auxiliary boiler system (ABS), and the MSS to prevent cross-contamination.

The staff reviewed FSAR section 10.3 and Section 12.3, Radiati on Protection Design Features, as they relate to prevention and minimization of con tamination. Because the NuScale FSAR design provides adequate measures for early leak detection and controls in the MSS design to minimize contamination as described above, the staff concludes that the system as described in the FSAR conforms to 10 CFR 20.1406.

10.3.1.4.7 Additional Design Aspects Reviewed

For the DHRS to function properly, the main steamlines are isol ated, and steam is routed to the DHRS passive condensers. The staf f reviewed the capability of the main steamlines to isolate in the event of a postulated break in a main steamline, assuming a single active failure. In the event an MSIV fails to close, the NuScale design incorporates s eismic Category I secondary MSIVs and secondary MSIBVs, neither of which are safety related. The TS includes surveillances for operability and inservice testing of the seco ndary MSIVs. The staff notes that the treatment of the secondary MSIVs and secondary MSIBVs is co nsistent with the previous staff position on the treatment of equipment that is not safety related when applying single-failure criteria for steamline break accidents, as shown in NUREG-0138. Therefore, the staff finds this acceptable. Section 5.4 of this SER contains t he staffs evaluation of the DHRS.

DSRS section 10.3, subsection III, item 6.C, states that the re viewer should verify that MSIVs and other shutoff valves can clos e against maximum steam flow. The applicant stated in FSAR Section 6.2.4.3, Design Evaluation, that all Containment Isol ation Valves (CIVs), which include the primary-side CIVs and the secondary-side CIVs, are required to isolate their flowpath with the required stroke time against the flow generated during line break conditions. These flow rates are derived from high-energy line break and safety analys is calculations and represent the maximum steam flow conditions for the flowpath. FSAR section 6. 2.4.3 states that MSIVs are capable of stopping fully developed pipe break flow for steam c onditions. The basis for these flow conditions is to bound the expected range of flows. FSAR S ection 10.3.2.1.2, Secondary Main Steam Isolation Valves, states that the secondary MSIVs a re capable of closing in steam conditions. Based on this information, the staff concludes that the MSIVs and the secondary MSIVs are capable of closing against maximum expected steam flo w conditions.

10-9 10.3.1.5 Initial Test Program

The preoperational test related to the MSS is the TGS test # 24, which tests design functions of the TGS. The test is performed by the COL holder in accordance with FSAR Table 14.2-24, Test # 24 Main Steam System. Section 14.2 of this SER documen ts the staffs evaluation of the initial test program for the standard design approval revie w.

Safety-related active components in the MSS are designed to be tested during plant operation.

FSAR Section 6.2.1.1.2, Design Features, states that the CNTS components (which include the MSIVs and MSIBVs) are designed to meet the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI inspection r equirements for Class 1, Class 2, and Class 3 main condensers (MCs), including the prese rvice inspection requirements.

FSAR Section 10.3.4, Inspections and Tests, states that the p ortion of the main steamlines that are not safety related are inspected and tested in accorda nce with the requirements of ASME B31.1, Power Piping. FSAR Section 3.9.6, Functional Des ign, Qualification, and In-service Testing Programs for Pumps, Valves, and Dynamic Rest raints, and Section 6.6, In-service Inspection and Testing of Class 2 and 3 Systems and Components, describe periodic inservice inspection and inservice testing of ASME Sec tion III, Class 2 and 3, components.

10.3.1.6 Conclusion

Based on its review as discussed above, the staff concludes tha t the MSS for the NuScale design satisfies the relevant requirements for the MSS as descr ibed in section 10.3.1.3 of this SER.

10.3.2 Steam and Feedwater System Materials

10.3.2.1 Introduction

To minimize flow-accelerated corrosion (FAC), the NuScale FSAR provides the design and fabrication codes, water chemistry and the compatibility of mat erials with the environment for portions of the power conversion system. These portions of the power conversion system include portions of the MSS and FWS that are not safety related, and consist of the TGS, including the turbine bypass system and the turbine gland seali ng system, the MSS, including extraction steam, the condensate and feedwater system (FWS), in cluding the condensate polisher skid and resin regeneration system(CPS), and the ABS.

10.3.2.2 Summary of Application

The materials used to minimize FAC described in FSAR Section 10.3.6, Steam and Feedwater System Materials, are for the portions of the MSS and CFWS tha t are not safety related, which include the following:

• The portions of the MSS that are not safety related, as descri bed in FSAR section 10.3, and extend from the flange immediately downstream of the safety -related MSIVs to the inlet of the turbine generator vendor package.

• The portions of the CFWS that are not safety related as descri bed in FSAR Section 10.4.6, Condensate and Feedwater System, and extend f rom the entrance of the MC (section 10.4.1, Air-Cooled Condensers) to the flange immediately upstream of the SG FWIVs. The CPS is another associated subsystem in the po rtions of the CFWS

10-10 that are not safety related (section 10.4.5, Condensate Polish er Skid and Resin Regeneration System).

• The TGS (which is entirely not safety related) as described in FSAR, section 10.2. Other associated subsystems that are part of the TGS are the turbine gland sealing system (TGSS) (FSAR section 10.4.3) and turbine bypass system (TBS) (F SAR section 10.4.4).

• The ABS (which is entirely not safety related) as described in FSAR Section 10.4.7, Auxiliary Boiler System.

The quality group for the portions of the MSS, the portions of the CFWS, the TGS, the ABS, and their associated subsystems that are not safety related is Qual ity Group D, and the piping is designed and constructed to ASME B31.1.

The materials are generally selected to minimize the impact of FAC. Erosion and corrosion are also minimized by the use of a Secondary Water Chemistry Control Program, which is described and reviewed in section 10.4.6 of this SER. The Secon dary Water Chemistry Control Program protects the safety-related steam generator system (SGS ) and DHRS from contamination originating in the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not safety related.

The piping design and layout considerations of the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not safety related will me et the guidance in Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, dat ed May 2, 1989, and Electric Power Research Institute (EPRI) NSAC-202L, Revision 3, Recomme ndations for an Effective Flow-Accelerated Corrosion Program, issued August 2007, to min imize erosion and corrosion (including FAC). The staff notes that historically, documents s uch as GL 89-08 have referred to FAC as erosion/corrosion. Therefore, the terms FAC and erosio n/corrosion are used interchangeably throughout this section of the SER.

COL Item 10.3-2 states that the COL applicant will provide a de scription of the FAC monitoring program based on GL 89-08 and the latest revision to EPRI NSAC-202L at the time of the COL application.

There are no proposed ITAAC, and no specific TS requirements as sociated with the main steam and feedwater materials.

10.3.2.3 Regulatory Basis

The following NRC regulations contain the relevant requirements for this review:

• GDC 1, Quality standards and records, and 10 CFR 50.55a, Co des and standards, require that SSCs important to safety shall be designed, fabric ated, erected, and tested to quality standards commensurate with the importance of the sa fety functions to be performed.

• GDC 35, Emergency core cooling, states that a system to prov ide abundant emergency core cooling shall be provided. The system safety fun ction shall be to transfer heat from the reactor core following any loss of react or coolant at a rate such that (1) fuel and clad damage that could interfere with continu ed effective core cooling is prevented and (2) clad metal-water reaction is limited to negli gible amounts.

10-11

• 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nu clear Power Plants and Fuel Reprocessing Plants, Criterion XIII, Handling, Storage a nd Shipping, requires that measures be established to control the handling, storage, shipping, cleaning, and preservation of materials and equipment to prevent damage or de terioration.

• 10 CFR 50.65, Requirements for monitoring the effectiveness o f maintenance at nuclear power plants, requires that power reactor licensees mo nitor the performance or condition of SSCs against licensee-established goals in a manne r sufficient to provide reasonable assurance that such SSCs can fulfill their intended functions.

The guidance in SRP Section 10.3.6, Steam and Feedwater System Materials, Revision 3, issued March 2007, lists the acceptance criteria adequate to me et the above requirements, as well as review interfaces with other SRP sections.

The following documents provide additional criteria or guidance in support of the SRP acceptance criteria to meet the above requirements:

• GL 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning

• EPRI NSAC-202L, Recommendations for an Effective Program Aga inst Erosive Attack, Revision 4

10.3.2.4 Technical Evaluation

10.3.2.4.1 System Design and Code of Construction

The NRC staff considered the system design and code of construc tion, the materials selection and fabrication, water chemistry and the FAC program for the po rtions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not safety r elated.

SRP section 10.3.6 is based on the use of ASME Code,Section II I, Class 2 and Class 3 components. However, the portions of the NuScale MSS, the porti ons of the CFWS, the TGS, the ABS, and their associated subsystems that are not safety re lated are designed to ASME B31.1. The selection of ASME B31.1 as the code of construc tion is consistent with the recommendations in RG 1.26, Quality Group Classifications and Standards for Water-, Steam-,

and Radioactive-Waste-Containing Components of Nuclear Power Pl ants, for Quality Group D components. ASME B31.1 provides requirements for piping design, construction (including welding procedure qualifications, welder qualifications, fabric ation methods, welding preheat, and interpass temperatures) and examination. Section 3.2 of thi s SER presents the staffs review of the adequacy of system classifications. Based on the staffs review of the adequacy of the system classification in section 3.2 of this SER, the staff finds the use of ASME B31.1 is acceptable for the design of the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not safety related.

The staff also reviewed SRP section 10.3.6 related to the requi rement of fracture toughness testing. The SRP states that fracture toughness testing is used to meet GDC 35, so that the steam and feedwater system integrity can be maintained to allow the systems to fulfill their safety functions of removing dec ay heat and supplying steam to engineered safety feature pumps. Since the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are reviewed in this section of the SER are not safety related, the staff finds the standard toughness requirements in ASME B31.1 to be acceptable and that additional fracture toughness requirements are not required, considering that the systems are not safety related.

10-12 NuScale stated that the piping design and layout considerations of the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not sa fety related will meet the guidance in GL 89-08 and EPRI NSAC-202L to minimize erosion/cor rosion (including FAC).

EPRI NSAC-202L provides guidance on design changes to minimize the rate of FAC damage to the piping systems. Since NuScale will use past operating exper ience and guidance to mitigate FAC using GL 89-08 and NSAC-202L, the staff finds that the desi gn provides reasonable assurance that FAC will be minimized to ensure that the piping systems will perform their functions.

10.3.2.4.2 Material Selection and Water Chemistry

For effective control of FAC, several key areas of information are needed, including design, water chemistry, and materials. While the specific grades of pi ping components have not been selected, NuScale stated that the portions of the MSS, the port ions of the CFWS, the TGS, the ABS, and their associated subsystems that are not safety relate d will be constructed with materials that are resistant to FAC, such as chromium-molybdenu m (Cr-Mo) steel, or stainless steel, consistent with the guidance in EPRI NSAC-202L. Cr-Mo st eels and stainless steels have an extensive history in steam and power conversion systems, and the material is suitable for steam and elevated temperature water service if controls are pr ovided to prevent material degradation. Cr-Mo steels, such as SA-355 Grade P11 or P22, are listed in EPRI NSAC-202L as FAC-resistant alloys. The staff finds the material for these systems to be acceptable because Cr-Mo steel or stainless will be used, which has acceptable res istance to FAC and is consistent with EPRI NSAC-202L. In addition to the materials (Cr-Mo steel or stainless steels) specified for these systems, FAC can be controlled by predicting FAC rates in accordance with EPRI NSAC-202L to ensure that these piping systems will be able to p erform their functions.

Therefore, the staff finds that the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems that are not safety related will be constructed with Cr-Mo steels or stainless steels, which are FAC-resistant materials and therefore are acceptable and will meet 10 CFR 50.65 provisions related to accounting for material selection and fab rication using industrywide operating experience.

The staff reviewed the guidance in SRP section 10.3.6 related t o cleaning and handling of safety-related materials in RG 1.28, Quality Assurance Program Criteria (Design and Construction). Since the portions of the MSS, CFWS, TGS, ABS, and their associated subsystems reviewed in this section of the SER are not safety r elated, these quality assurance requirements are not applicable. However, chemical contaminatio n originating in the portions of the MSS, the portions of the CFWS, the TGS, the ABS, and their associated subsystems that are not safety related affect the safety-related portions of th e DHRS, CNTS, and SGS.

Therefore, to mitigate any potential impact on the safety-relat ed portions of the DHRS, CNTS, and SGS, the NuScale design includes a Secondary Water Chemistr y Control Program, and the use of appropriate material for the DHRS, CNTS, and SGS provide s protection to these safety-related systems. The staff reviewed the Secondary Water Chemistry Control Program and safety-related materials in sections 10.4.5 and 6.1.1 of th is SER, respectively.

10.3.2.4.3 Flow-Accelerated Corrosion

SRP section 10.3.6, Item III.3, states that EPRI NSAC-202L prov ides acceptable methods to minimize FAC. The use of EPRI NSAC-202L is also acceptable beca use that document is endorsed in Section XI.M17, Flow-Accelerated Corrosion, of NU REG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, issued December 201 0.

10-13 NuScale stated that the piping design of the portions of the MS S, the portions of the CFWS, the TGS, the ABS, and their associated subsystems that are not safe ty related meets the guidance of GL 89-08 and EPRI NSAC-202L to reduce FAC. EPRI NSAC-202L pr ovides guidance for nuclear power plants to implement an effective program to detec t and mitigate FAC. This includes the identification of recommended FAC program tasks wi th key steps of identifying and ranking susceptible systems, performing FAC predictive analysis, selecting and scheduling components for inspection, performing inspections, evaluating i nspection data, assessing worn components, and repairing or repl acing components. In addition, GL 89-08 requires that a long-term FAC program be in place using a demonstrated analysis method, such as EPRI NSAC-202L. Therefore, the staff finds meeting the guidance in G L 89-08 and EPRI NSAC-202L for minimizing the occurrence of FAC acceptable, and the 10 CFR 50.65 provisions related to accounting for FAC industrywide operating experience are satisf ied.

FSAR Section 10.3.6.3, Flow-Accelerated Corrosion, includes C OL Item 10.3-2 which makes a COL applicant responsible for developing the site-specific Flow -Accelerated Corrosion Monitoring Program based on GL 89-08 and the latest revision of EPRI NSAC-202L at the time of the application. Conforming FAC programs established through out the domestic nuclear fleet allow licensees to identify, monitor, and mitigate FAC-related damage in advance of piping failure. The staff finds the COL item acceptable, as it will en sure that a COL applicant will develop an FAC program in accordance with the applicable guidan ce to meet pertinent requirements of 10 CFR 50.65 related to accounting for industry wide operating experience.

10.3.2.5 Inspections, Tests, Analyses, and Acceptance Criteria

There are no ITAAC required for the steam and feedwater system materials.

10.3.2.6 Technical Specifications

There are no TS requirements associated with the materials of t he portions of the MSS, portions of the CFWS, the TGS, the ABS, and their associated subsystems that are not safety related.

Required TS for the portions of the MSS, the portions of the CF WS, the TGS, the ABS, and their associated subsystems are evaluated in the related system secti ons in this chapter of the SER.

Therefore, the staff finds the lack of TS requirements associat ed with the materials for these systems that are not safety-related is acceptable, in accordanc e with 10 CFR 50.36, Technical specifications.

10.3.2.7 Conclusion

Based on its review of the steam and feedwater materials, the s taff concludes that the materials to be used for the portions of the MSS, the portions of the CFW S, the TGS, the ABS, and their associated subsystems that are not safety related are acceptabl e and meet appropriate regulatory requirements as identified in section 10.3.2.3 of th is SER.

10.4 Other Features of Steam and Power Conversion System

10.4.1 Air-Cooled Condensers

10.4.1.1 Introduction

Each NPM has an air-cooled condenser in which turbine exhaust s team is directly condensed by the air and the condensate is returned to the SG as feedwate r through the condensate and feedwater system (FWS). The Air Cooled Condenser System (ACCS), also referred to as the

10-14 main condenser, is made up of the air-cooled condensers and its subsystem, i.e., the condenser air removal system (CARS). Each NPM has a condenser that provid es adequate capacity for the FWS.

10.4.1.2 Summary of Application

FSAR section 10.4.1 contains the ACCS design basis, system, and component descriptions, as well as information on system operation, inspections, and testi ng. FSAR Figure 10.4-1, Air Cooled Condenser System Piping and Instrumentation Drawing, pr ovides a graphical representation of the ACCS. The ACCS receives steam from the MS S through the turbine exhaust piping and delivers condensate to the FWS through the c ondensate collection tank.

FSAR figure 10.1-1 shows the general connections and interfaces with these systems. The ACCS operates under a vacuum maintained by the CARS, which is d escribed in FSAR Section 10.4.2, Condenser Air Removal System. CARS radiation monitoring instrumentation provided at the discharge of CARS monitors the radiation releas es from the ACCS.

FSAR section 14.2 (Test # 07, Air Cooled Condenser System) pr ovides information on the ACCS initial testing. No proposed ITAAC and no specific TS requ irements are associated with the ACCS.

10.4.1.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are provided in SRP Section 10.4.1, Main Condensers, and are summarized below:

• GDC 4, requires, in part, that SSCs important to safety be des igned to accommodate the effects of and to be compatible with the environmental conditio ns associated with normal operation, maintenance, testing, and postulated accidents, incl uding loss-of-coolant accidents.

• GDC 60, Control of releases of radioactive materials to the e nvironment, requires, in part, that provisions be included in the nuclear power unit des ign to control suitably the release of radioactive materials in gaseous and liquid effluent s during normal operation, including AOOs.

• GDC 64, Monitoring radioactivity releases, requires, in part, that provisions be included for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including AOOs, and from po stulated accidents.

• 10 CFR 20.1406, requires, in part, that the design features th at will facilitate eventual decommissioning and minimize, to the extent practicable, contam ination of the facility and the environment, facilitate eventual decommissioning, and m inimize, to the extent practicable, the generation of radioactive waste.

10.4.1.4 Technical Evaluation

The staff reviewed the ACCS design, described in FSAR section 1 0.4.1, in accordance with guidance in SRP section 10.4.1, to evaluate compliance with the regulatory requirements listed in section 10.4.1.3 of this SER. Because the ACCS transports se condary-side fluids outside the RXB and into and through components outside in the yard, the st aff also reviewed the system to evaluate whether the design complies with 10 CFR 20.1406.

10-15 10.4.1.4.1 GDC 4, Environmental and Dynamic Effects Design Bases

The staff reviewed the design of the ACCS to evaluate whether t he design complies with the requirements of GDC 4. The staffs review was performed to veri fy that the system was appropriately protected against environmental and dynamic effec ts or that a failure of the ACCS and the resulting discharging fluid (i.e., flooding) would not adversely affect SSCs important to safety. In FSAR Section 10.4.1.1, Design Basis, the applicant stated that the MC serves no safety-related functions, is not credited for mitigation of a d esign-basis accident (DBA) and has no safe shutdown functions. Also, in FSAR Section 10.4.1.3, Sa fety Evaluation, the applicant stated that flooding resulting from a failure of the Condensate Collection Tank does not prevent operation of a safety-related system because no such systems ar e located in the TGB.

Since the ACCS is not credited for performing any accident miti gation, not relied on for safe shutdown, and serves no safety-related function, if the condens er loses its ability to perform its intended design function, nuclear safety will not be affected. Therefore, external missiles, pipe whip, and jet impingement would not affect nuclear safety. In a ddition, in regard to flooding that could occur as a result of a failure of the condensate collecti on tank, the design has provisions to prevent unacceptable flooding of areas containing safety-rel ated equipment. FSAR Section 3.4.1, Internal Flood Protection for Onsite Equipment Failures, states the following:

Water from tanks and piping that are seismic category III and n ot protected against tornadoes or hurricanes are potential flooding sources outside the buildings. There are no large tanks or water sources near entra nces to the RXB and CRB. The site is graded to transport water away from these buildings.

Therefore, failure of equipment outside the CRB and RXB cannot cause internal flooding.

Based on the above review, the staff finds that the ACCS design complies with GDC 4, because the performance of the systems intended function is not necess ary for nuclear safety or safe shutdown, and the potential flooding due to a failure of the AC CS does not result in adverse effects on SSCs important to safety.

10.4.1.4.2 GDC 60, Control of Releases of Radioactive Materials to the Environment, and GDC 64, Monitoring Radioactivity Releases

The staff reviewed the design of the ACCS to evaluate whether t he design complies with the requirements of GDC 60, with respect to control of release of r adioactive materials, and GDC 64, with respect to the monitoring of radioactive releases. Com pliance with GDC 60 and GDC 64 requires provisions to be included in the nuclear power unit design to monitor and suitably control the release of radioactive materials during normal oper ation, including AOOs, and postulated accidents. Meeting these requirements provides a lev el of assurance that the release of radioactive materials in gaseous and liquid effluents from t he ACCS during normal operation, including AOOs, and postulated accidents is kept as low as is r easonably achievable, in accordance with 10 CFR Part 50, Appendix I, Numerical Guides f or Design Objectives and Limiting Conditions for Operation to Meet the Criterion As Low as is Reasonable Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reacto r Effluents.

In FSAR section 10.4.1.3, the applicant indicated that the cont rol of radioactive materials to the atmosphere is covered in FSAR Section 10.4.2, Condenser Air Re moval System, which describes how gases in the condenser are removed by the CARS, w hich controls release of the ACCS radioactive effluents to the environment. FSAR section 10. 4.2 states that the CARS

10-16 maintains a vacuum on the condenser to remove gases. Removed ga ses pump through water separator tanks and vent to the atmosphere. The CARS and gland seal steam system exhausts have direct, unfiltered pathways out of the TGB to the atmosphe re. Process and effluent radiation monitoring instrumentation for the ACCS is provided f or the CARS common vent air evacuation line, two common liqui d ring vacuum pumps shared among six ACCS units and six steam jet air ejector skids, one for each of the six ACCS units. The radiation ACCS process flow and identification of the radiation monitoring of the ACCS is s hown in FSAR Figure 11.5-1b, Radioactive Effluent Flow Paths with Process and Effluent Radi ation Monitors, and described in FSAR Table 11.5-1, Process and Effluent Radiation Monitorin g Instrumentation Characteristics.

Based on the above review, the staff finds that the ACCS design complies with GDC 60 and 64, because the design includes provisions to monitor and suitably control the release of radioactive materials from the ACCS during normal operation.

10.4.1.4.3 10 CFR 20.1406, Minimization of Contamination

The regulations in 10 CFR 20.1406 require, in part, that each a pplicant describe how the facility design and procedures for operation will minimize, to the exten t practicable, contamination of the facility and the environment, as well as the generation of radioactive waste.

Primary-to-secondary leakage through the SG tubes has the poten tial to introduce radioactive material into the MMS, and the contamination can then be transf erred to the condenser through the turbine exhaust or the TBS. FSAR Table 11.1-5, Secondary C oolant Design Basis Source Term, and FSAR Table 11.1-7, Secondary Coolant Realistic Sour ce Term, show, respectively, the secondary coolant design basis and realistic radiation source terms.

The staff reviewed the design of the ACCS to evaluate whether t he design complies with the requirements of 10 CFR 20.1406. The staff finds that the ACCS i ncludes design features that address the provisions of RG 4.21, Minimization of Contaminati on and Radioactive Waste Generation: Live-Cycle Planning, as described in FSAR Table 12.3-13, Regulatory Guide 4.21 Design Features for Air Cooled Condenser System. Section 12.3 of this SER presents the general review of NuScale conformance with RG 4.21.

10.4.1.5 Initial Test Program

The preoperational test related to the ACCS is test # 07, which ensures that the various design aspects related to the ACCS are implemented. The test is perfor med in accordance with FSAR Table 14.2-7, Test # 07, Air Cooled Condenser System. Section 14.2 of this SER documents the staff evaluation of the initial test program for the standa rd design approval review.

10.4.1.6 Conclusion

Based on its review of the ACCS as described above, the staff c oncludes the design is acceptable because it meets the appropriate regulatory requirem ents as stated in section 10.4.1.3 of this SER.

10.4.2 Condenser Air Removal System

10.4.2.1 Introduction

The staff reviewed FSAR Section 10.4.2, Condenser Air Removal System, in accordance with guidance in SRP Section 10.4.2, Main Condenser Evacuation Syst em, to evaluate whether the

10-17 CARS is designed and built to establish and maintain ACCS vacuu m and to monitor for radioactive material. CARS radiation monitoring instrumentation provided at the discharge of CARS monitors the radiation releases from the ACCS.

10.4.2.2 Summary of Application

FSAR section 10.4.2 contains a general description of the CARS system and specifies that its primary functions are to reduce dissolved oxygen in the feedwater and to maintain ACCS vacuum condition during plant startup, cooldown, and normal ope rating conditions by removing air and non-condensable gases from the MCs. The CARS subsystem has a non-safety-related with augmented requirements function to provide post-accident i nstrumentation to monitor variables such as radioactivity, the status of safety-related e quipment, and the status of fission product barriers.

FSAR section 14.2 (Test # 07) provides information on the initi al testing of the CARS. No proposed ITAAC and no specific TS requirements are associated w ith the CARS safety evaluation.

10.4.2.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are identified in SRP section 10.4.2 and are summarize d below:

• GDC 60requires, in part, that provisions be included in the CA RS design to suitably control the release of radioactive materials in gaseous and liq uid effluents during normal operation, including AOOs.

• GDC 64, requires, in part, that provisions be included in the nuclear power unit design for monitoring the effluent discharge paths and the plant envir ons for radioactivity that may be released from normal operations, including AOOs, and fro m postulated accidents.

SRP section 10.4.2, item I, also indicates review interfaces wi th other SRP sections.

10.4.2.4 Technical Evaluation

The CARS is used to remove air from the ACCS and, therefore, es tablishes a vacuum. To prevent loss of condenser vacuum, each ACCS is provided with tw o 100 percent capacity CARS in parallel. Each NPM has a dedicated secondary system wit h independent condenser air ejector systems. If one system is unavailable because of mainte nance or lost during normal operation, the redundant system is started. The components in t he CARS are not shared among other NPMs; therefore, the failure of the CARS does not impair the ability of other NPMs to perform their safety functions. A failure of CARS results in an increase in pressure in the MC to which CARS is connected. The loss of ACCS vacuum is an AOO and is discussed in FSAR Section 15.2.3, Loss of Condenser Vacuum.

10.4.2.4.1 GDC 60, Control of Releases of Radioactive Materials to the Environment, and GDC 64, Monitoring Radioactivity Releases

The exhaust of the condenser air removal system (CARS) vacuum p umps is monitored for radioactive effluents and is serviced by the process sampling s ystem. FSAR Figure 7.1-2, Post-Accident Monitoring General Arrangement Drawing, notes t hat the CARS is equipped for

10-18 integrated sampling with a PING monitor to detect particulate /iodine/noble gases/argon-41 gamma and has provisions for representative grab samples throug h the takeoff and return path for monitoring process flow. If required, operators in the main control room can manually shut down and isolate the CARS in response to an abnormal plant cond ition. The alarm setpoints, control room monitoring capability, and operator response in ac cordance with site procedures enables the operation of the system and monitoring and control of radiation releases. Therefore, the staff finds that the CARS complies with GDC 60 and 64.

10.4.2.5 Initial Test Program

The preoperational test related to the CARS for standard design approval is Test # 07, which includes a test to verify that the system can maintain main con denser vacuum pressure, provide steam to the condenser deaerator as indicated by steam flow and supply gland seal steam to the turbine generator at design pressures. This test is perform ed is performed by the COL holder in accordance with FSAR table 14.2-7. Section 14.2 of th is SER documents the staff evaluation of the initial test program for the standard design review.

10.4.2.6 Conclusion

Based on its review of the CARS as described above, the staff c oncludes that the design is acceptable because it meets the appropriate regulatory requirem ents as stated in section 10.4.2.3 of this SER.

10.4.3 Turbine Gland Sealing System

10.4.3.1 Introduction

The staff reviewed FSAR Section 10.4.3, Turbine Gland Sealing System, in accordance with guidance in SRP Section 10.4.3, Turbine Gland Sealing System. The TGSS is a turbine subsystem that provides a source of sealing steam to prevent ai r leakage into the turbine under vacuum and steam leakage out of the turbine under pressure duri ng certain load conditions.

10.4.3.2 Summary of Application

FSAR section 10.4.3 contains a general description of the TGSS and specifies that the system (1) prevents air leakage into the turbine under vacuum, (2) pre vents steam leakage out of the turbine under pressure, and (3) provides for the use of redunda nt steam supplies and controlling devices.

10.4.3.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are provided in SRP section 10.4.3 and are summarized below:

• GDC 60requires, in part, that the TGSS design provides for the control of releases of radioactive materials to the environment.

• GDC 64requires, in part, that the TGSS design provide for moni toring releases of radioactive materials to the environment during normal operatio n, including AOOs and postulated accidents.

SRP section 10.4.3 also indicates review interfaces with other SRP sections.

10-19 10.4.3.4 Technical Evaluation

The staff reviewed FSAR section 10.4.3, in accordance with SRP section 10.4.3, to evaluate compliance with the regulatory requirements listed in section 1 0.4.3.3 of this SER. The applicant stated in FSAR Section 10.4.3.3,Safety Evaluation, that the TGSS has no safety-related functions, is not credited for mitigation of a DBA, and has no safe-shutdown functions; therefore, it is not required to operate during or after a DBA.

The staff reviewed the information in the FSAR on the TGSS incl uding the system description and information in FSAR Sections 11.3, Gaseous Waste Managemen t System, and 11.5, Process and Effluent Radiation Monitoring Instrumentation and Sampling System. The TGSS steam exhausts have direct, unfiltered pathways out of the TGB to the atmosphere. FSAR table 11.5-1 shows that process radiation monitoring for the gl and steam outlet is provided at the turbine generator skid common exhaust vent point. Hence, th e release to the turbine building from the turbine gland system is monitored.

FSAR Section 11.5.1.2, Effluent Instrumentation Alarm Setpoint s, states that All process and effluent monitors provide local and Main Control Room (MCR) ind ication of radiation at each location and provide an alarm function in the MCR when predeter mined thresholds are exceeded. Thus, design features are in place to control and mo nitor releases of radioactive materials in the effluents of the TGSS; accordingly, the staff finds these sampling and monitoring provisions for the TGS meet the requirements of GDC 60 and GDC 64, respectively, as they relate to control and monitoring of the releases of the radioactive materials to the environment.

10.4.3.5 Initial Test Program

Preoperational tests related to the TGSS for standard design ap proval include test # 29, which ensures that the various design aspects related to the TGS are implemented. These tests are performed by the COL holder in accordance with FSAR table 14.2-29. Section 14.2 of this SER documents the staff evaluation of the initial test program for the standard design approval review.

10.4.3.6 Conclusion

Based on its review of the TGSS above, the staff concludes that the design is acceptable because it meets the appropriate regulatory requirements as sta ted in section 10.4.3.3 of this SER.

10.4.4 Turbine Bypass System

10.4.4.1 Introduction

The staff reviewed FSAR Section 10.4.4, Turbine Bypass System, in accordance with guidance in SRP Section 10.4.4, Turbine Bypass System. The TB S is a turbine subsystem that provides main steam directly from the SGs to the air-coole d condenser in a controlled manner to remove heat from the NPM.

10.4.4.2 Summary of Application

FSAR section 10.4.4 contains a general description of the TBS. The TBS consists of a line connected to the main steam combined header with a regulating v alve and an inline

10-20 desuperheater discharging to the ACCS. The turbine bypass valve dumps steam from the main steam header through the desuperheater to the condenser. The va lve is capable of throttling the full bypass flow from the turbine to the condenser without requ iring actuation of the main steam safety valve.

FSAR section 14.2 (Test # 29) provides information on the initi al testing of the TBS. No proposed ITAAC or specific TS requirements are associated with the TBS.

10.4.4.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are provided in SRP section 10.4.4 and summarized belo w:

• GDC 4, requires, in part, that SSCs important to safety be des igned to accommodate the effects of and to be compatible with the environmental conditio ns associated with normal operation, maintenance, testing, and postulated accidents, incl uding loss-of-coolant accidents.

• GDC 34, requires, in part, the ability to use the system for s hutting down the plant during normal operations. The operation of the TBS eliminates the need to rely solely on safety systems, which are required to meet the redundancy and power so urce requirements of this criterion.

SRP section 10.4.4 also indicates review interfaces with other SRP sections.

10.4.4.4 Technical Evaluation

The staff reviewed FSAR section 10.4.4, in accordance with SRP section 10.4.4, to evaluate compliance with the regulatory requirements listed in section 1 0.4.4.3 of this SER. The applicant stated that the TBS is part of the TGS, has no safety-related f unction, is not credited for mitigation of a DBE, and has no safe-shutdown functions. The ap plicant also stated that the TBS is designed to satisfy GDC 4, and its failure will not affe ct any SSCs important to safety.

10.4.4.4.1 GDC 4, Environmental and Dynamic Effects Design Bases

The staff reviewed the design of the TBS to evaluate compliance with the requirements of GDC 4. Conformance to GDC 4 requires that failure of the TBS ca used by a pipe break or malfunction of the system should not adversely affect essential systems or components that are necessary for safe shutdown or accident prevention or mitigatio n.

FSAR Section 10.4.4.1, Design Basis, states that the TBS serv es no safety function, is not risk-significant, is not credited for mitigation of a DBS, and has no safe-shutdown functions.

FSAR section 10.4.3.3 states that the TGB does not contain safe ty-related equipment. The staff reviewed the TGS design in FSAR section 10.2 and verified that the piping and valves associated with the TBS are located completely within the turbi ne building and that the turbine building does not contain any SSCs important to safety.

Based on the above discussion, the staff finds that the TBS mee ts the GDC 4 criterion as it relates to the adverse effects of a pipe break or malfunction o n those components of the system

10-21 necessary for safe shutdown or accident prevention or mitigatio n since the turbine building does not contain such components.

10.4.4.4.2 GDC 34, Residual Heat Removal

The staff reviewed the TBS to evaluate whether the design compl ies with the requirements of GDC 34 as related to the ability to use the system to shut down the plant during normal operations by removing residual heat without using the turbine generator. The TBS is not credited for compliance with GDC 34. The decay and residual hea t removal safety function per GDC 34, as modified by NuScales PDC 34 and as discussed in sec tion 10.3 of this SER, is performed by the DHRS, which is a passive design that consists of two independent trains, each capable of performing the system safety function in the event o f a single failure. Section 5.4.3 of this SER covers the staffs review of the DHRS and compliance w ith PDC 34. Since the DHRS is used to comply with PDC 34 and the TBS is not credited for c ompliance, the staff finds that GDC 34 is not applicable to this system.

10.4.4.5 Initial Test Program

Preoperational tests related to the TBS for standard design app roval include the test # 29Test and # 63, Hot Functional Testing, which ensure that the vari ous design aspects of the TBS are implemented. These tests are performed in accordance with F SAR table 14.2-29 and FSAR Table 14.2-63, Hot Functional Testing. Section 14.2 of this S ER documents the staff evaluation of the initial test program for the standard design approval review.

10.4.4.6 Conclusion

Based on its review of the TBS above, the staff concludes that the design is acceptable because it meets the appropriate regulatory requirements as sta ted in section 10.4.4.3 of this SER.

10.4.5 Condensate Polisher Skid and Resin Regeneration System

10.4.5.1 Introduction

A condensate polisher skid in the TGB is part of the FWS for ea ch NPM. The purpose of the condensate polisher skid is to treat and clean feedwater to rem ove corrosion products and ionic impurities. The cation resin used in the condensate polisher sk id is regenerated in the CPS, shared among all NPMs. The CPS is designed to receive the catio n resin from the condensate polisher skids, regenerate the resin, and return it to the cond ensate polisher skids. FSAR Section 10.3.5, Water Chemistry, describes the water quality requirements, but the evaluation of that information is included here in this SER because the co ndensate polisher skid and resin regeneration systems are a principal means of controlling secon dary water chemistry.

10.4.5.2 Summary of Application

FSAR section 10.4.5 contains the system design basis and descri ptions of the condensate polisher skid and the CPS, system operation, components, instru mentation, and the inspection and testing provisions. FSAR Section 10.4.5.1, Design Basis, states that the condensate polisher skid and CPS serve no safety-related functions, are no t risk-significant, are not credited for mitigation of a DBA, and have no safe-shutdown functions. F SAR section 10.3.5 is also reviewed with section 10.4.5 in this SER section because it des cribes the Secondary Water Chemistry Control Program, including the chemistry requirements, methods of treatment,

10-22 monitoring, and sampling, and contaminant ingress sources and r esponse. FSAR Section 9.3.2, Process Sampling System, describes collection of representati ve samples of the secondary water. FSAR section 11.5 addresses detection of radioactivity i n the secondary coolant (for example, from a primary-to-secondary leak). FSAR Section 10.4.8, Feedwater Treatment System, describes chemical additions to the feedwater as part of the Secondary Water Chemistry Control Program.

FSAR figure 10.1-1 shows the locations of the condensate polish er skid and CPS, as well as the feedwater treatment system (FWTS) chemical injection.

FSAR section 14.2, Test # 27, Condensate Polisher Resin Regene ration System, and Test #

72, Primary and Secondary System Chemistry, provides informat ion on initial testing. SDAA Part 4, Section 5.5.5, Secondary Water Chemistry, contains TS requirements associated with the Secondary Water Chemistry Control Program.

10.4.5.3 Regulatory Basis

The relevant regulatory requirements for this area of review an d the associated acceptance criteria are summarized below.

• GDC 14, Reactor coolant pressure boundary, in Appendix A to 10 CFR Part 50, requires that the RCPB be designed, fabricated, erected, and te sted to ensure an extremely low probability of abnormal leakage, rapidly propagat ing failure, and gross rupture.

NUREG-800 SRP Section 10.4.6, Condensate Cleanup System, iden tifies acceptance criteria that are adequate to meet applicable requirements for the conde nsate polisher skid and CPS.

10.4.5.4 Technical Evaluation

10.4.5.4.1 Secondary Water Chemistry

Secondary water chemistry is focused on preventing corrosion in SGs, condensers, piping, and other components. Principal param eters that must be controlled are pH, dissolved oxygen, and impurity ion concentrations, including sodium, chloride, and su lfate ions.

The staff reviewed the information in FSAR sections 10.3.5 and 10.4.5, against the requirements of GDC 14. GCD 14 applies to the condensate polish er skid and CPS because they support secondary water quality to prevent corrosion-relat ed failure of the RCPB, specifically the SG tubing. (CPS is the abbreviation for the resin regeneration system that serves the condensate polisher skid.) Water purity can also aff ect the accumulation of deposits on the secondary side of the SG tubing, which may in turn affec t the ability to perform the required tube inspections. As described in SRP section 10.4.6, an acceptable method of compliance with GDC 14, as it relates to maintaining an extreme ly low probability of abnormal leakage, rapidly propagating failure, or gross rupture of the R CPB, is for the applicant to meet the guidelines in the latest version of the EPRI report series, PWR Secondary Water Chemistry Guidelines. In addition to protecting SG tubes and the condens ate and feedwater components, secondary water chemistry is also selected with the goal of min imizing corrosion and performance losses for all secondary system components. The EPR I Guidelines are applicable because the NPM design of secondary-side materials (including S G tubing) and water chemistry for which there is extensive operating experience and for which the EPRI Guidelines were developed.

10-23 The EPRI Guidelines provide several criteria for the Secondary Water Chemistry Control Program, including sampling frequency and other sampling requir ements, guidelines for continuously monitoring water chemistry parameters, and operati ng limits for impurities and additives. The EPRI Guidelines include associated actions to be performed if limits (action levels) are exceeded. Although the staff does not perform revi ews or issue safety evaluations of EPRI water chemistry guidelines, EPRI Guidelines are recogni zed as representing the industrys best practices in water chemistry control. The Guide lines are reviewed regularly by industry and updated periodically. All PWR licensees are commit ted to following the EPRI secondary water chemistry guidelines through their commitment t o Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines. FSAR Section 5.4.1.6, Steam Generator Program, states that the Steam Generator Program is based on N EI 97-06.

FSAR section 10.3.5 addresses the secondary water chemistry req uirements for SG water and feedwater. FSAR Tables 10.3-3a, Steam Generator Sample (Wet La yup) (Reactor Coolant System 200°F), 10.3-3b, Feedwater Sample (Reactor Coolant S ystem > 200°F to <15%

reactor power), 10.3-3c, Feedwater Sample (15% reactor power ), 10.3-3d, Condensate Sample (15% reactor power, and 10.3-3e, Steam Generator Fill Water (initial fill subsequent to a shutdown), provide the values for certain secondary water chemistry parameters.

The NRC staff reviewed the information in FSAR section 10.3.5 t o determine if the applicants proposed secondary water chemistry parameters provide reasonabl e assurance that the requirements of GDC 14 will be met with respect to secondary wa ter chemistry. The staff compared the parameters monitored and the allowable values to t he threshold values for which corrective action is required in the EPRI Guidelines referenced in the FSAR and the latest version of the EPRI Guidelines. Based on this review, the staff determined that the water chemistry parameters in the application are consistent with the values in the EPRI Guidelines, which represent industry best practices for secondary water che mistry control.

FSAR section 10.3.5 includes COL Item 10.3-1, which makes a COL applicant responsible for developing the site-specific Secondary Water Chemistry Control Program, based on the applicable revisions of NEI 97-06 and the EPRI Secondary Water Chemistry Guidelines at the time of the application. The staff finds the COL item acceptabl e because it is appropriate for a COL applicant to account for site-specific factors in designing a secondary water chemistry program that meets the appropriate industry guidelines.

As stated in FSAR Section 10.3.5.1, Treatment and Monitoring, an all-volatile treatment, such as ammonium hydroxide or an amine, is used for pH control, and hydrazine is added to control dissolved oxygen. The NRC staff finds this acceptable because t hese chemicals are widely used for these purposes in operating plants and their use is al lowed by and addressed in the EPRI Guidelines. The chemical additions to the secondary water, for pH control and oxygen scavenging, are part of the FWTS, discussed in FSAR section 10. 4.8 and this SER.

In addition to providing acceptable water quality treatment to support RCPB integrity, adequate instrumentation and sampling must be provided to verify the eff ectiveness of the treatment system in order to meet GDC 14 and the guidance in SRP section 10.4.6. As described in FSAR Section 10.3.5.2, Sampling, Section 9.3.2.2.1, General Descr iption, and Table 9.3.2-3, Secondary Sampling System Normal Sample Points, the secondary sampling system is designed to sample the parameters specified by the Secondary Wa ter Chemistry Program using continuous and grab samples. FSAR Section 10.4.5.5, Instrument ation, states that instrumentation for pressure, pressure drop, temperature, level, and flow rate are provided for monitoring process conditions. FSAR Section 10.4.5.2, System D escription, states that the

10-24 system is designed such that the temperature at the condensate polishers does not exceed the design temperature for the resin during normal operation or pla nned transients. FSAR Section 10.3.5.3, Contaminant Ingress, states that the conden sate collection tank, condensate polisher discharge, condensate pump discharge, and demineralize d water are monitored for conductivity, chemistry, and oxygen to detect contamination fro m oxygen and ionic species. The staff finds the design for instrumentation and sampling accepta ble because it is consistent with the EPRI Guidelines as recommended in SRP section 10.4.6.

10.4.5.4.2 Condensate Cleanup Capacity

The NRC staff reviewed the cleanup capacity of the condensate p olisher skid and CPS with respect to compliance of the design with GDC 14. Based on SRP s ection 10.4.6, the system should have adequate capacity to provide effluent of the necess ary purity. FSAR section 10.4.5.2 describes the condensate polisher skid and the CPS. The condensate polisher skid is designed with the capacity to treat 100 percent of the FWS flow for each module to bring the water quality within specifications during startup and main tain water quality during operation.

The CPS is shared among the NPMs and has the capacity to regene rate the resin in each of the condensate polisher skids in cycle (one at a time).

The condensate polisher skid removes corrosion products and ion ic impurities from secondary water by passing it through condensate inlet filters, condensat e polisher trains, and resin filters.

The condensate polisher trains contain cation and mixed-bed (ca tion and anion) deionizers for removing ionic impurities. The CPS regenerates spent resin in t he lead cation vessel and is used to replace the mixed-bed resin in the condensate polisher skid. The CPS includes chemical skids, chemical storage, and neutralization tanks and pumps. FSAR Table 10.4-1, Condensate Polisher Resign Regeneration System Resin Tanks, l ists the number and sizes of resin tanks in the CPS.

In FSAR Section 10.4.5.3, System Operation, the applicant sta tes that design features ensure that in the event of condenser tube leaks, concentrations of co ntaminants are limited to allowable values until the FWS is isolated. FSAR section 10.3.5.3 discusses condenser leaks and other potential sources of contamination and how the contam ination is detected. This includes monitoring or sample analyses at the condensate collec tion tank, condensate pump discharge, condensate polisher discharge, and the demineralized water. Condensate polishers are used in the event of upset chemistry conditions, and the al lowable values follow the EPRI Guidelines as discussed and tabulated in FSAR section 10.3.5. T he design features actions to detect contamination in the secondary water and return the wate r to conformance with the Secondary Water Chemistry Control Program requirements.

SRP section 10.4.6 recommends that the CPS be connected to the radioactive waste disposal systems to allow disposal of spent resin or regenerant solution s when necessary. FSAR section 10.4.5.3 states that the water drained from the exhaust ed resin vessel is discharged to the BPDS and monitored for contamination. Figure 9.3.3-2, Bala nce-of-Plant Drain System Diagram, shows schematically t he flow of the regenerant solution to the BPDS tanks and liquid radwaste system. FSAR Section 11.4, Solid Waste Management Sys tem, provides more information about the management of resins that require process ing as contaminated waste, including condensate polisher demineralizer resins.

Based on the design to establish and maintain the specified wat er quality at full FWS flow, respond to ingress of contaminants, and dispose of spent resin or regenerant solutions, the staff finds that the system meets the criteria for condensate cleanup capacity as recommended in SRP section 10.4.6.

10-25 10.4.5.4.3 Materials of Construction

The NRC staff reviewed the materials of construction of the con densate polisher skid and CPS with respect to compliance of the design with GDC 14. SRP secti on 10.4.6 recommends that materials of construction be compatible with the service enviro nment. In FSAR section 10.4.5.2, the applicant states that materials resistant to corrosion, ero sion, and flow-accelerated corrosion (FAC), as discussed in FSAR section 10.3.5, are used for compon ents in which loss of material could occur. FSAR section 10.3.5 states that this applies to co mponents and piping exposed to wet steam, flashing liquid flow, or turbulent single-phase flow in which loss of material could occur. FSAR section 10.3.5 also says emphasis is placed on excl uding copper and copper alloy pipe, valves, and components from the secondary chemistry envir onment to eliminate copper transport to the SGs and allow a higher feedwater pH to reduce iron corrosion and transport to the SGs.

Requirements for the materials are described in FSAR section 10.3.6, which classifies the condensate and feedwater system, including the condensate polis hing skid, as Quality Group D and design code ASME B31.1 (Power Piping). The ASME B31.1 cod e has requirements for design, materials, fabrication, erection, testing, inspection, operation, and maintenance of piping and piping systems and associated components, including require ments for a corrosion allowance. FSAR section 10.3.6.3 states that provisions to mini mize FAC include the selection of Cr-Mo or stainless steel consistent with EPRI FAC guidelines. COL Item 10.3-2 requires a COL applicant to provide a description of its FAC program. Base d on the design code, the materials specified, and the design of the condensate polisher skid and CPS to maintain the water quality according to the Secondary Water Chemistry Progra m, the staff finds that the design provides reasonable assurance that the materials of cons truction will be compatible with the service environment and enable the system to perform its functions.

10.4.5.5 Initial Test Program

Section 14.2 of this SER evaluates preoperational test # 27, de scribed in FSAR section 14.2, Table 14.2-27, Test # 27 Condensate Polisher Resin Regeneratio n System, and preoperational test # 72 in FSAR Section 14.2, Table 14.2-72, Test # 72 Primary and Secondary System Chemistry. These preoperational tests are per formed by the COL holder.

10.4.5.6 Technical Specifications

SDAA Part 4 includes TS 5.5.5, Secondary Water Chemistry Progr am, which provides an administrative program for the monitoring and control of second ary water chemistry to inhibit SG tube degradation. In SDA Part 2, Chapter 16, Technical Specifi cations, the applicant stated that the generic TS were developed with the content typified in NUREG-1431, Standard Technical SpecificationsWestinghouse Plants, Revision 5, issu ed September 2021, and NUREG-1432, Standard Technical Spec ificationsCombustion Engin eering Plants, Revision 5, issued September 2021. The staff reviewed NuScale T S 5.5.5 for applicability to the condensate polisher skid and CPS. The staff determined that TS 5.5.5 provides appropriate administrative requirements for control and monitoring of secon dary water chemistry based on consistency with the Standard Technical Specifications for oper ating reactor designs.

Chapter 16 of this SER contains the staff evaluation of the TS.

10-26 10.4.5.7 Conclusion

Based on its review of the secondary water chemistry, condensat e polisher skid, and CPS above, the staff concludes that the design is acceptable becaus e it meets the appropriate regulatory requirements as stated in section 10.4.5.3 of this S ER.

10.4.6 Condensate and Feedwater System

10.4.6.1 Introduction

The FWS provides feedwater at the required temperature, pressur e, and flow rate to the SGs.

As depicted in the NuScale power conversion system block diagra m (FSAR figure 10.1-1),

condensate is pumped from the condensate collection tank by the condensate pumps; passes through the CPS, the gland steam condenser, and the low-and in termediate-pressure feedwater heaters to the feedwater pumps; and then is pumped th rough the high-pressure feedwater heaters into the tube side of the helical coil SG. Ea ch NPM is supplied with a separate FWS, not shared with other NPMs.

The NuScale FSAR describes the FWS boundaries as extending from the ACCS to the downstream flange on the removable spool piece between the feed water regulating valve (FWRV) and feedwater isolation valve (FWIV). As discussed in se ction 10.4.6.4 of this SER, the applicants description of the FWS boundaries is not consistent with the FWS boundaries described in DSRS section 10.4.7. Specifically, NuScale has not included, as part of the FWS, the portions of the system between the SG and the SG FWIV, whic h functionally serve as part of the FWS. For the purposes of this review, the staff consider s the FWS to extend from the outlet of the MC up to the tube side inlet of the SG, which inc ludes the portion of the system from the SG to the SG FWIV as part of the FWS per guidance in D SRS Section 10.4.7, Condensate and Feedwater System.

10.4.6.2 Summary of Application

FSAR Section 10.6, Condensate and Feedwater System, contains the FWS design basis, system description, and the results of the FWS failure modes an d effects analysis. The FWS is not safety related and is primarily located within the TGB and RXB, with the exception of some piping and the condensate storage tank located outside.

FSAR section 14.2 (Test # 25, Condensate and Feedwater System ) provides information on the FWS initial testing. There are no proposed ITAAC related to the FWS; however, in SDAA Part 8, Section 2.4, Equipment Qualification, the applicant p roposed ITAAC for the following FWS equipment: the feedwater supply check valves, the FWIV, and the FWRV. Section 14.3.7 of this SER evaluates these ITAAC.

The TS associated with the FWS are provided in SDAA, Part 4, U S460 Generic Technical Specifications, section 3.7.2, Feedwater Isolation.

10.4.6.3 Regulatory Basis

The relevant requirements for this area of review and the assoc iated acceptance criteria are provided in section 10.4.7 of the NuScale DSRS and are summariz ed below:

10-27

• GDC 2 requires, in part, that important-to-safety portions of the FWS be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perf orm their safety functions.

• GDC 4 requires, in part, that structures, systems and components impo rtant to safety be appropriately protected against the dynamic effects associated with possible fluid flow instabilities (e.g., water hammer) during normal plant operatio n, as well as during upset or accident conditions.

• GDC 5 requires, in part, that structures, systems and components impo rtant to safety not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety func tions.

• GDC 44, Cooling water, as related, in part, to the following :

- the capability to transfer heat loads from the reactor system to an ultimate heat sink under both normal operating and accident conditions.

- suitable redundancy of components and features for onsite ele ctric power systems so that, under accident conditions, the safety function s can be performed assuming a single active component failure (which may be coincident with the loss of offsite power for certain events)

GDC 45, Inspection of cooling water system, requires, in par t, that the cooling water system shall be designed to permit periodic in-service inspecti on of important system components and equipment.

GDC 46, Testing of cooling water system, requires, in part, that the cooling water system shall be designed to permit appropriate functional testi ng of the system and components to ensure structural integrity and leak-tightness, o perability and performance of active components, and capability of the integra ted system to function as intended during normal, shutdown, and accident conditions.

10 CFR 20.1406, requires, in part, that the design will facili tate eventual decommissioning and minimize, to the extent practicable, contam ination of the facility and the environment, and minimize, to the extent practicable, t he generation of radioactive waste.

DSRS section 10.4.7 also indicates review interfaces with other SRP sections.

10.4.6.4 Technical Evaluation

The staff reviewed the FWS described in the FSAR in accordance with the review procedure in DSRS section 10.4.7. As indicated in section 10.4.6.1 of this S ER, the applicants description of the FWS boundaries is not consistent with the FWS boundaries de scribed in DSRS section 10.4.7. Thus, SSCs that are functionally part of the FWS are no t identified as part of the FWS in the FSAR. The staff performed its review based on the system bo undaries identified in the DSRS as described in section 10.4.6.1 of this SER. Use of the D SRS system boundaries ensures that the SSCs that are functionally part of the FWS and located inside the RXB and containment are properly designed to ensure that they perform t heir safety-related functions under all normal and accident conditions and that their failure will not adversely impact other important-to-safety SSCs. The results of the staffs review are provided below.

10-28 10.4.6.4.1 GDC 2, Design Bases for Protection against Natural Phenomena

The staff reviewed the FWS to evaluate whether the design compl ies with the requirements of GDC 2, with respect to its design for protection against the ef fects of natural phenomena such as earthquakes, tornados, hurricanes, and floods. Compliance wi th the requirements of GDC 2 is based on the FWS being designed to withstand the effects of natural environmental phenomena without losing the ability to perform its safety func tion and on adherence to Regulatory Position C.1 in RG 1.29, Revision 4, for the safety-related portion of the system and Regulatory Position C.2 for the portions of the system that are not safety related.

RG 1.29, Regulatory Position C.1.d, states that the pertinent q uality assurance requirements of Appendix B to 10 CFR Part 50 shall apply to all activities affe cting the safety-related function of those portions of the steam and feedwater systems of PWRs exten ding from and including the secondary side of the SG up to and including the outermost CIVs, and connecting piping of a nominal size of 6.35 centimeters (2.5 inches) or larger, up to and including the first valve that is either normally closed or capable of automatic closure during a ll modes of normal operation.

The staff reviewed FSAR figure 10.1-1 and table 10.4-4 and conf irmed that SSCs important to safety, including the feedwater isolation and check valves as w ell as the regulating valves and the spool piece from which the system is disconnected during re fueling, are all designed to seismic Category I and thus comply with Regulatory Position C.1 in RG 1.29.

FSAR Section 3.2, Classification of Structures, Systems and Co mponents, categorizes SSCs based on safety importance and other considerations. FSAR table 10.4-4 gives the location, safety classification, and seismic category for the FWS. All of the FWS SSCs located in the TGB are listed as non-safety-related, Quality Group D, and seismic Category III, except for the FWRVs, feedwater supply check valves, and the FWRV limit switch, which are seismic Category I.

10.4.6.4.2 GDC 4, Environmental and Dynamic Effects Design Bases

The staff reviewed the FWS to evaluate whether the design compl ies with the requirements of GDC 4, as related to dynamic effects associated with possible f luid flow instabilities, including water hammer and effects of pipe breaks. Compliance with the re quirements of GDC 4 is based on identification of SSCs important to safety that need to be p rotected from dynamic effects, including internally and externally generated missiles, pipe wh ip, and jet impingement due to high-and moderate-energy missiles and water hammer.

The FWS CIVs perform the safety-related function of containment isolation. The FWIVs provide for isolation of feedwater and support DHRS operation by isolat ing the DHRS from the FWS.

FSAR Section 10.4.6.3, Safety Evaluation, states that dynamic effects such as water hammer can be generated by FWIV or FWRV closure and opening, check val ve closure, or pump start and stop. The section also states that the potential for water hammer in the FWS is minimized by design features such as pipe slope, the use of available dra ins before startup, and adjustment of valve closure timing. The applicant then conclud es that the FWS piping arrangement and valve characteristics ensure water hammer loads are below SG design limits.

FSAR Section 13.5, Plant Procedures, states that Administrat ive and operating procedures are utilized by the operating organization (plant staff) to ens ure that routine operating, off-normal, and emergency activities are conducted in a safe ma nner.

The design aspect related to water hammer prevention was examin ed by the staff as part of the SDA FSAR review, while the pr ocedures will be evaluated during the staffs review of the COL

10-29 application as site-specific items. Based on the staffs review of the FWS design as described in FSAR section 10.4.7, the staff finds the NuScale FWS design to be consistent with guidance in NUREG-0927, Evaluation of Water Hammer Occurrence in Nuclear P ower Plants, Revision 1, issued March 1984, and in compliance with GDC 4.

10.4.6.4.3 GDC 5, Sharing of Structures, Systems, and Components

The staff reviewed the design of the FWS to evaluate compliance with the requirements of GDC 5 with respect to shared systems among NPMs. Compliance wit h GDC 5 requires that provisions be included in the nuclear power unit design to ensu re that an event in one NPM does not adversely impact the ability of the shared systems to perform their safety functions in other NPM units or their ability to safely achieve and maintain safe shutdown. The FWS is not shared among NPMs; therefore, the failure of FWS or components in the FWS will not impair the ability of SSCs important to safety in other NPMs to perfor m their safety functions.

Therefore, the requirement of GDC 5 for sharing of systems betw een units is satisfied.

10.4.6.4.4 GDC 44, Cooling Water

The staff reviewed the design of the FWS to evaluate compliance with the requirements of GDC 44. The DHRS performs this function for the NPM. The feed water system does have a connection with the DHRS but does not have the safety function to transfer heat under accident conditions and, therefore, GDC 44 is not applicable to the CFWS. The description of the DHRS system is in FSAR, Section 5.4.3, and the staffs evaluation of the DHRS system is in Section 5.4.4 of this report.

10.4.6.4.5 GDC 45, Inspection of Cooling Water System, and GDC 46, Testing of Cooling Water System

The staff reviewed the FWS design to evaluate whether design pr ovisions are provided for periodic inspections of components, as required by GDC 45 and p eriodic functional testing of the system and components, as re quired by GDC 46. Using the enh anced safety-focus review approach, the staffs review focused on the SSCs that are funct ionally part of the FWS that perform or support feedwater and containment isolation function s. The isolation functions are important to nuclear safety because FWS isolation is required t o establish and maintain the pressure boundary for the DHRS and thus preserve DHRS inventory and ensure proper operation of the safety-related DHRS.

Even though the feedwater system SSCs are relied on for feedwat er system isolation, which is necessary for DHRS operation, the applicant concluded that GDC 45 and GDC 46 are not applicable to the FWS because the applicant does not consider t he FWS a cooling system as described in GDC 44. The applicants position is based, in part, on the applicants defining the boundaries of the feedwater system differently than the way it is defined in guidance (DSRS section 10.4.7). In the NuScale FSAR, the feedwater system SSCs relied on for feedwater system isolation, which is neces sary for DHRS operation, were not included as part of the FWS, thus leaving important-to-safety SSCs, generally covered in thi s system review, out of scope of the system. The staffs position is that, because the SSCs that are functionally part of the FWS are credited for isolating the FWS and providing and maintainin g the pressure boundary, the provisions for inspection and testing called for in GDC 45 and 46 are applicable. As discussed below, however, the staff determined that the FWS design satisf ies the requirements of GDC 45 and GDC 46.

10-30 The FWS components that perform the isolation functions are the FWIV and the feedwater isolation check valves. These valves are considered CIVs and ar e discussed in FSAR section 6.2. The valves are located such that there is sufficie nt access to allow for inservice inspection; maintenance; repair; and 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, t ype C testing. These valves are included in the plants TS, and testing of the FWIV is also performed according to TS 3.7.2, Feedwater Isolation.

The staff confirmed that the FWS SSCs were included as part of the inservice testing program in Chapter 3, Design of Systems, Structure, Components, and Eq uipment, of the FSAR. FSAR Table 3.9-16, Active Valve List, identifies components that a re subject to preservice and inservice testing plans. Among the valves that are included for the CNTS are the feedwater isolation and the feedwater check valves, and the identified fu nctions of the valves are feedwater isolation, containment isolation, and decay heat remo val boundary. Since provisions permit periodic inservice inspection of the feedwater isolation and check valves, and they are included in the inservice test program, as indicated in FSAR ta ble 3.9-16, the staff finds that the NuScale design provides for periodic inspection and testing and therefore satisfactorily addresses the requirements of GDC 45 and 46.

10.4.6.4.6 Compliance with 10 CFR 20.1406, Minimization of Contamination

The regulations in 10 CFR 20.1406 require, in part, that each S DA applicant describe how the facility design and procedures for operation will minimize, to the extent practicable, contamination of the facility and the environment, as well as m inimize the generation of radioactive waste. Primary-to-secondary leakage through the SG tubes has the potential to introduce radioactive material into the FWS. The applicant stat ed, in FSAR section 10.4.6.3, that the FWS design satisfies the requirements of 10 CFR 20.1406, mi nimizing the contamination of the facility and the environment in accordance with RG 4.21.

The staff reviewed the design of the FWS to evaluate whether th e design complies with the requirements of 10 CFR 20.1406. In FSAR section 10.4.6.3, the a pplicant stated that main steam and condensate monitoring with MSS and FWS isolation capa bilities minimizes the contamination and release to the environment, and that the FWS drains to the BPDS, which discharges to the radioactive waste drain system should the FWS become contaminated. The staff finds that the applicant has also addressed the FWS desig n features for compliance with RG 4.21 in FSAR Table 12.3-23, Regulatory Guide 4.21 Features for Condensate and Feedwater System. Section 12.3 of this SER includes the genera l review of NuScale compliance with RG 4.21.

Based on the above discussion, the staff concludes that the FWS, as described in the FSAR, complies with 10 CFR 20.1406.

10.4.6.5 Initial Test Program

Preoperational tests related to the FWS for standard design app roval include the FWS tests

(#25 and #28). These tests are performed by the COL holder in a ccordance with FSAR Table 14.2-25, Test # 25 Condensate and Feedwater System, and FSAR Table 14.2-28, Test # 28 Feedwater Heater Vents and Drains System. Section 14.2 of this SER documents the staff evaluation of the initial test program for the standard design approval review.

10-31 10.4.6.6 Conclusion

The staff finds the FWS design acceptable because it meets appl icable regulatory requirements, including GDC 2, regarding protection from natural phenomena; G DC 4, on protection against missiles and effects of pipe break; GDC 5, on shared systems; G DC 45, on inspections; GDC 46, on periodic testing; and 10 CFR 20.1406.

10.4.7 Auxiliary Boiler System

10.4.7.1 Introduction

The staff reviewed FSAR Section 10.4.7, Auxiliary Boiler Syste m. There is no specific SRP section applicable to the review of the ABS. However, the staff appropriately used similar regulatory requirements for similar systems in SRP Chapter 10, Steam and Power Conversion System, such as the TGSS for this area of review. The ABS is d esigned to supply steam to systems where main steam is not available or not preferred.

10.4.7.2 Summary of Application

FSAR section 10.4.7 contains a general description of the ABS, as well as relevant information on the ABS design, including the design bases, instrumentation, and the inspection and testing program.

FSAR section 14.2 (Test # 06, Au xiliary Boiler System) provid es information on initial testing of the ABS. There are no proposed ITAAC, or specific TS requiremen ts associated with the ABS.

10.4.7.3 Regulatory Basis

There is no specific SRP section applicable to the review of th e ABS. Therefore, based on similar systems, the staff used the following relevant regulato ry requirements for this area of review:

• GDC 2, requires, in part, that SSCs important to safety be des igned to withstand the effects of natural phenomena such as earthquakes, tornadoes, hu rricanes, and floods; this includes consideration of the failure of SSCs that are not safety related due to natural phenomena such as earthquakes, tornadoes, hurricanes, a nd floods, which could adversely affect SSCs that are important to safety.

• GDC 4, requires, in part, that SSCs important to safety shall be appropriately protected against the dynamic effects of external missiles, internal miss iles, pipe whip, and jet impingement forces associated with pipe break; this includes co nsideration of a failure of the ABS due to a pipe break or malfunction that could adversely affect essential systems or components necessary for safe shutdown or accident preventio n or mitigation.

• GDC 5 requires, in part, that SSCs important to safety not be shared among the NPMs

• GDC 60, requires, in part, the ability of the auxiliary steam system design to control releases of radioactive materials to the environment.

• GDC 64, requires, in part, that provisions be included in the nuclear power unit design for monitoring the effluent discharge paths and the plant envir ons for radioactivity that

10-32 may be released from normal operations, including AOOs, and fro m postulated accidents.

• 10 CFR 20.1406, requires, in part, that the design features wi ll minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the g eneration of radioactive waste.

10.4.7.4 Technical Evaluation

The ABS is not safety related. The ABS is a non-seismic system designed to supply steam to systems where main steam is not available or not preferred. The primary functions of the ABS are to provide steam to the turbine gland seals, support the AC CS by supplying steam to the deaerator during startup, and support the condensate polishing resin regeneration system. The ABS is not credited for mitigation of a DBE and has no safe-shu tdown functions.

10.4.7.4.1 GDC 2, Design Bases for Protection against Natural Phenomena

Compliance with the requirements of GDC 2 is based on the fact that the ABS is designed to withstand the effects of postulated natural phenomena, includin g earthquakes, such that it would not result in the loss of the capability of SSCs importan t to safety to perform their safety functions. FSAR Section 10.4.10.3, Safety Evaluation, states that the ABS serves no safety function, is not credited for mitigation of a DBA, and has no s afe-shutdown functions. In FSAR table 10.4-4, the seismic classification for the ABS is listed as Seismic Category III, and the location is the TGB. Since the ABS is not safety related, or ri sk significant, and the TGB does not contain any seismic Category I SSCs, the staff finds that t he seismic Category III categorization of the ABS is appropriate, and the ABS complies with GDC 2.

10.4.7.4.2 GDC 4, Environmental and Dynamic Effects Design Bases

The staff also reviewed the ABS design to verify that the syste m was protected against environmental and dynamic effects or that a failure of the ABS and the resulting discharging fluid (i.e., flooding) would not adversely affect SSCs importan t to safety as stated in GDC 4.

Compliance with the requirements of GDC 4 is based on the deter mination that failures of the ABS due to pipe break or malfunction would not adversely affect any of the plants SSCs important to safety. FSAR table 10.4-4 identifies the location of the ABS as the turbine building.

Because there are no safety-related SSCs that can potentially b e affected by the dynamic effects of an ABS failure, the staff finds that the design of t he ABS meets the requirements of GDC 4.

10.4.7.4.3 GDC 5, Sharing of Structures, Systems, and Components

Since the ABS serves all NPMs, the staff also reviewed the desi gn of the ABS to evaluate compliance with the requirements of GDC 5 with respect to share d systems. Compliance with GDC 5 requires that provisions be included in the nuclear power unit design to ensure that an event in one NPM does not significantly impair the ability of i mportant-to-safety SSCs in any other NPM units to perform their safety functions, including th e ability to safely achieve and maintain safe shutdown. The applicant stated in FSAR section 10.4.7.3 that there are no safety-related components in the ABS that are shared among NPMs ; therefore, failure of the ABS does not significantly impair the ability of other NPMs to perform their safety functions.

10-33 The staff evaluated the design of the ABS, and as discussed in the previous section, finds that failure of this system would not significantly impair the abili ty of SSCs important to safety to perform their safety function. Therefore, the staff finds that the ABS design meets the requirements of GDC 5.

10.4.7.4.4 GDC 60, Control of Releases of Radioactive Materials to the Environment, and GDC 64, Monitoring Radioactivity Releases

The staff also reviewed the design of the ABS to evaluate wheth er the design complies with the requirements of GDC 60 and GDC 64 with respect to control and m onitoring of radioactive releases. Compliance with GDC 60 and GDC 64 requires that the n uclear power unit design include provisions to monitor and suitably control the release of radioactive materials during normal operation, including AOOs and postulated accidents. Meet ing these requirements provides a level of assurance that the release of radioactive m aterials in gaseous and liquid effluents from the ABS during normal operation, including AOOs and postulated accidents, is kept as low as is reasonably achievable, in accordance with 10 CFR Part 50, Appendix I.

FSAR figure 11.5-1b shows that effluent radiation monitoring is provided for the ABS. As stated in FSAR Section 11.5-4, Effluent and Process Monitoring Off No rmal Radiation Conditions, if high radiation is detected in t he ABS skid vents, skid drains, or steam header drains, then the auxiliary boiler superheater skid outlet valve closes, auxiliar y boiler skid to superheater skid valve closes, the module-specif ic main steam to auxiliary boiler header valves close, and the MCR receives an alarm.

Based on the above, the staff finds that the ABS complies with GDC 60 and 64.

10.4.7.4.5 10 CFR 20.1406, Minimization of Contamination

The staff also reviewed the design of the ABS to evaluate wheth er the design complies with the requirements of 10 CFR 20.1406 with respect to minimizing conta mination of the facility and the environment. The staff finds that the ABS includes design featu res that address the provisions of RG 4.21, as described in FSAR Table 12.3-12, Regulatory Gui de 4.21 Design Features for Auxiliary Boiler System. Section 12.3 of this SER describes th e general review of NuScale conformance with RG 4.21.

Based on its review, the staff concludes that the ABS design co mplies with 10 CFR 20.1406.

10.4.7.5 Initial Test Program

Preoperational tests related to the ABS for standard design app roval include the test # 6. The test is performed by the COL holder in accordance with FSAR Tab le 14.2-6, Test # 06, Auxiliary Boiler System. Section 14.2 of this SER documents th e staff evaluation of the initial test program.

10.4.7.6 Conclusion

Based on its review of the ABS described above, the staff concl udes that the design is acceptable because it meets the appropriate regulatory requirem ents as stated in section 10.4.7.3 of this SER.

10.4.8 Feedwater Treatment System

10-34 10.4.8.1 Introduction

The purpose of the feedwater treatment system (FWTS) is to main tain secondary water quality in combination with the condensate polisher skid (FSAR section 10.4.5) and process sampling system (FSAR section 9.3.2) by providing chemical addition and feedwater sampling. The FWTS provides chemical injection capability for maintaining fee dwater pH and dissolved oxygen limits to meet the requirements of the Secondary Water Chemistr y Control Program (FSAR section 10.3.5).

10.4.8.2 Summary of Application

FSAR section 10.4.8 contains the system design basis and a gene ral description of the system equipment and operation. FSAR Section 10.4.8.1, Design Bases, states that the FWTS serves no safety-related functions, is not risk-significant, is not cr edited for mitigation of a DBA, and has no safe-shutdown functions. The FWTS provides for chemical inje ction to support water chemistry and the condensate polisher skid and resin regenerati on system, which are discussed in FSAR sections 10.3.5 and 10.4.5, respectively. FSAR section 9.3.2 describes collection of representative samples of the secondary water. Section 10.4.8 i ncludes COL Item 10.4-1, which requires a COL applicant to provide a secondary water chemistry analysis showing that FWTS size, materials, and capacities satisfy the requirements of the Secondary Water Chemistry Control Program described in section 10.3.5 and that the system is compatible with the chemicals used.

FSAR figure 10.1-1 shows the location of the FWTS chemical inje ction, as well as the condensate polisher skid and CPS.

FSAR section 14.2, Test # 26, Feedwater Treatment System, and Test # 72, provide information on initial testing.

10.4.8.3 Regulatory Basis

The relevant regulatory requirement for this area of review is GDC 14. FSAR sections 10.3.5 and 10.4.5 describe how the secondary water chemistry control s upports compliance with GDC 14. The FWTS supports secondary water chemistry by enabling chemical injection.

GDC 14, Reactor coolant pressure boundary, requires that the RCPB be designed, fabricated, erected, and tested to ensure an extremely low probability of a bnormal leakage, rapidly propagating failure, and gross rupture.

10.4.8.4 Technical Evaluation

The purpose of the FWTS is to control corrosion and erosion of feedwater system components by providing for chemical injection to maintain feedwater pH an d dissolved oxygen concentration. The staff reviewed the information in FSAR secti on 10.4.8, in conjunction with the information in FSAR sections 10.3.5 and 10.4.5, with respect to the requirements of GDC 14.

GDC 14 is applicable to the FWTS since the system is designed t o perform chemical injection to help maintain secondary water chemistry. Secondary water chemis try is controlled, in part, to prevent corrosion-induced failure of the RCPB, specifically the SG tubing. The FWTS was reviewed in conjunction with the secondary water chemistry and condensate polisher skid sections of the FSAR because these systems and programs work to gether to maintain secondary water chemistry and to help prevent corrosion-induced failure of the RCPB.

10-35 The staffs review of FSAR section 10.4.8 focused on the abilit y of the FWTS to provide the appropriate chemical injection that allows for the control of c ertain secondary water chemistry parameters. FSAR sections 10.3.5 and 10.4.5 describe how mainta ining acceptable secondary water chemistry will demonstrate compliance with GDC 14 with re spect to corrosion-induced failure of the RCPB.

FSAR section 10.4.8.2, System Description, states that chemic al injection points are provided downstream of the FWS condensate pumps, and there is separate e quipment for pH control and oxygen scavenger injection. Figure 10.1-1 shows the FWTS injection location schematically. FSAR section 10.4.8.3, System Operation, state s that the Process Sampling System (PSS) monitors the feedwat er, and the FWTS makes chemica l additions to keep feedwater chemistry within limits. The FWTS chemical addition p umps use the PSS readings to determine the chemical injection requirements, which are contro lled by the plant control system and module control system. These chemical injections are availa ble for normal operation and for cleanup modes during startup. The staff finds this design accep table based on having provisions and equipment for monitoring the feedwater chemistry and injecting chemicals for pH and dissolved oxygen control.

FSAR section 10.4.8 includes COL Item 10.4-1, which makes a COL applicant responsible for providing a secondary water chemistry analysis that must show t hat the size, materials, and capacity of the FWTS equipment and components satisfy the water quality requirements of the Secondary Water Chemistry Control Program described in FSAR sec tion 10.3.5, and that the system is compatible with the feedwater chemicals. The staff fi nds it acceptable for the COL applicant to be responsible for the details of the FWTS because the design will be required to meet the requirements of the Secondary Water Chemistry Control Program described in section 10.3.5, while allowing the COL applicant flexibility in the choice of chemicals and equipment to meet those requirement s. The flexibility to customize the design is acceptable because the EPRI Guidelines that form the basis for the Seconda ry Water Chemistry Control Program allow for more than one way to maintain pH and dissolve d oxygen limits.

10.4.8.5 Initial Test Program

Section 14.2 of this SER contains the evaluation of preoperatio nal test # 26 in FSAR section 14.2, Table 14.2-26, Test # 26 Feedwater Treatment System, an d preoperational test # 72 in FSAR section 14.2, table 14.2-72. These preoperational tests ar e performed by the COL holder.

10.4.8.6 Conclusion

Based on its review of the FWTS described above, the staff conc ludes that the design is acceptable because it meets the requirements of GDC 14.

10-36