ML18190A509

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LLC Response to NRC Request for Additional Information No. 468 (Erai No. 9420) on the NuScale Design Certification Application
ML18190A509
Person / Time
Site: NuScale
Issue date: 07/09/2018
From: Rad Z
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-0718-60804
Download: ML18190A509 (13)
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RAIO-0718-60804 July 09, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information No.

468 (eRAI No. 9420) on the NuScale Design Certification Application

REFERENCE:

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

468 (eRAI No. 9420)," dated May 10, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's response to the following RAI Question from NRC eRAI No. 9420:

15-17 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions on this response, please contact Paul Infanger at 541-452-7351 or at pinfanger@nuscalepower.com.

Sincerely, Zackary Za Z ckary W. Rad

Director, Di t Regulatory R l t Affairs Aff i NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Rani Franovich, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9420 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0718-60804 :

NuScale Response to NRC Request for Additional Information eRAI No. 9420 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9420 Date of RAI Issue: 05/10/2018 NRC Question No.: 15-17 Regulatory Basis General Design Criterion (GDC) 1 in 10 CFR Part 50, Appendix A, requires structures, systems, and components (SSCs) important to safety to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

The NRC regulations in 10 CFR 50.2 define safety-related, in part, as SSCs that prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or §100.11 of that chapter, as applicable. In addition, 10 CFR 52.47(a)(2)(iv) provides equivalent siting and safety analysis offsite dose guidelines for new nuclear power plant standard design certifications.

Introduction In RAI 8744, Question 15.02.08-4, RAI 9205, Question 15-3, and RAI 9237, Question 15.06.03-3 the NRC staff requested the applicant to provide additional information justifying the credit of nonsafety-related components for design basis accident (DBA) mitigation. The applicant's response referred to the guidance in RG 1.206 which specifies that nonsafety-related components may be used as backup protection to mitigate transient or accident conditions; or NUREG-0138 which details the acceptable means by which nonsafety-related components can be credited for DBA mitigation. The staff found the applicant's response generally acceptable; however, the staff also determined that more information is required in order to find that the applicant is appropriately meeting the guidance provided in NUREG-0138, precedence, and overall find that the applicant's FSAR Tier 2, Chapter 15 analyses are in compliance with the applicable regulations.

Discussion The NRC staff agrees that NUREG-0138 (November 1976), "Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3, 1976, Memorandum from Director, NRR to NRR Staff," in Issue No. 1, "Treatment of Non-Safety Grade Equipment in Evaluations of Postulated Steam Line Break Accidents," discusses the acceptance of reliance on specific nonsafety-related valves as part of the mitigation of secondary line breaks.

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In particular, NUREG-0138 defines the issue as follows:

In evaluating the consequences of postulated breaks of steam lines the current staff position (SRP 10.3) states that the design should preclude the blowdown of more than one steam generator, assuming a concurrent single component failure, and assuming that the turbine stop and control valves remain functional. Provided that these valves and their control systems are designed for closure under the postulated conditions, and because they are high quality components, the staff does not require that they be designed to the requirements for safety-related equipment.

Regarding the reliance on non-safety grade equipment[1] to mitigate steam line break accidents, NUREG-0138 includes the following general NRC staff position:

For loss-of-coolant accidents (LOCA) involving a spontaneous rupture of the primary system boundary, where significant damage to the fuel and a major release of fission products are potential consequences, the most stringent quality and design requirements, including seismic qualification, are imposed on those systems needed to prevent and cope with a LOCA.

However, for accidents involving spontaneous failures of secondary system piping not part of the primary system boundary, where the potential consequences are significantly lower, less stringent requirements are imposed on the quality and design of the systems needed to cope with such secondary system ruptures. This approach results, in the staff's judgment, in a proper weighing of consequences and safety requirements in order to assure a balanced level of safety over the entire spectrum of postulated design basis accidents.

In NUREG-0138, the NRC staff discusses the reliance on non-safety grade valves, such as turbine stop, control, and intercept valves to mitigate the consequences of a steam line break.

For example, the staff indicates in NUREG-0138 that the continued reliability of these components over the life of the plant is assured by frequent (generally weekly) inservice tests.

The staff also states that the NRC conducted a survey of the reliability of these valves at operating light water reactors. The staff found no control system failures and few incidents where the valves did not fully close. Based on its review, the staff concluded that the reliability of these valves is of the same order of magnitude as that accepted for nuclear safety-grade components. Regarding feedwater isolation, the staff states that the rationale for reliance on "non-safety grade" feedwater components is similar to that presented for steam line valves. In NUREG-0138, the staff states its belief that it is acceptable to rely on these non-safety grade components in the steam and feedwater systems because their design and performance are compatible with the accident conditions for which they are called upon to function.

NUREG-0138 is referenced in other NRC and industry documents discussing reliance on nonsafety-related components as part of the mitigation of accidents. For example, in NUREG-1793 (September 2004), "Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design," the NRC staff states in Chapter 15, "Transient and Accident Analyses,"

Section 15.1.2, "Non-Safety-Related Systems Assumed in the Analysis," that crediting nonsafety-related backup systems and components in the design-basis analyses is acceptable NuScale Nonproprietary

for several reasons, including operating data that show that the turbine stop and control valves are reliable, and taking credit for the turbine valves in the design-basis analyses for backup protection is consistent with the staff position stated in NUREG-0138. In a request for additional information dated November 16, 2009 (ADAMS Accession No. ML093140231) regarding a proposed change to Technical Specifications for the auxiliary feedwater system at the Point Beach nuclear power plant, the NRC staff notes that NUREG-0138 allows a licensee to take credit for nonsafety-grade components in the main feedwater line, even though they are not designed seismic Category I, to perform a backup isolation function in certain accident scenarios, because the staff does not require that an earthquake be assumed to occur coincident with the postulated main steamline break. In its November 16, 2009, letter, the NRC staff further indicates that NUREG-0138 prescribes that in order to rely on these nonsafety-grade components, their design and performance must be compatible with the accident conditions for which they are called upon to be credited, and the reliability of these valves is of the same order of magnitude as that accepted for nuclear safety-grade components. In its reply dated December 16, 2009 (ADAMS Accession No. ML093510809), the Point Beach licensee provides information supporting its determination that the design and performance of the main feedwater regulating valves, which would be downgraded to nonsafety-related, will be compatible with the accident conditions for which the valves will be credited. These NRC and licensee documents indicate that the application of NUREG-0138 requires that the nonsafety-related components to be credited for main steam line or feedwater line breaks must be determined to be reliably capable of performing their intended function.

In its response to RAI 8744, Question 15.02.08-4, the NuScale design certification applicant states that the nonsafety-related FW supply check valves to be credited for a line break are listed in FSAR Tier 2, Section 3.2, "Classification of Structures, Systems, and Components,"

Table 3.2-1 of the same title, with requirements for augmented quality, designed to Seismic Category I and included in the inservice testing program. The NRC staff notes that Table 3.2-1 lists Feedwater Supply Check Valve (without an identifying valve number) as Seismic Category I with "Technical Specification Surveillance for operability and in-service testing" in the "Augmented Design Requirements" column. The "QA Program Applicability" column specifies AQ-S with Note 2 indicating that AQ-S "indicates that the pertinent requirements of 10 CFR 50 Appendix B are applicable to SSC classified as seismic category II in accordance with the quality assurance program."

In its response to RAI 9205, Question 15-3, the applicant provides information on the three nonsafety-related components that are credited in the NuScale Power Module Chapter 15 safety analyses (nonsafety-related feedwater check valves, nonsafety-related feedwater regulating valves, and nonsafety-related secondary (backup) main steam isolation valves (MSIV) and bypass valves). The information provided details the applicant's position on why the crediting of each valve is appropriate for DBA mitigation. As mentioned above, the staff generally finds the applicant's arguments acceptable; however, additional information is required to assure the applicant is appropriately meeting the guidance in NUREG-0138.

In its response to RAI 9237, Question 15.06.03-3, the applicant provided information related to NuScale Nonproprietary

the crediting of the nonsafety-related secondary MSIVs for a steam generator tube failure (SGTF). The response explains that due to the augmented design requirements identified, precedence from prior design certifications, and statements in RG 1.206, it is appropriate to credit the secondary MSIVs to mitigate the effects of an SGTF. However, the staff needs additional information in order to reach a safety finding on the applicant's proposal.

Request For reliance on nonsafety-related components to mitigate a DBA in the NuScale accident analysis, consistent with the guidance in NUREG-0138, the NRC staff requests that the NuScale design certification applicant provide the following information:

(1) Identifying numbers for the applicable nonsafety-related components, (2) Type of components (e.g., if a check valve, then what type - swing check or nozzle check, etc.),

(3) Performance history and operating experience for the applicable components and their application in the NuScale design, (4) Detailed design and qualification requirements to be applied to the components, including the associated electrical design and qualification requirements where it is applicable (i.e.,

backup isolation valves that require electrical signals to actuate),

(5) Preservice and inservice testing, and Technical Specification requirements to be applied to the components, (6) Planned modifications to the NuScale Design Certification application, such as Part 2 (FSAR Tier 1 and Tier 2), and Part 4 (Technical Specifications), to specify the design, qualification, ITAAC, preservice and inservice testing, and Technical Specification requirements for the applicable components, and (7) Clarification of the reference to AQ-S in the QA Program Applicability column in NuScale FSAR Tier 2, Table 3.2-1, for the applicable nonsafety-related valves to be credited.

In addition, for reliance on the nonsafety-related secondary MSIV to mitigate an SGTF event, as required by the definition of safety-related in 10 CFR 50.2, the applicant must ensure that the estimated offsite doses do not exceed the requirements of 10 CFR 52.47(a)(2)(iv) if the secondary MSIV fails to close. Therefore, the applicant is requested to:

(8) Perform an analysis that demonstrates 10 CFR 52.47(a)(2)(iv) criteria are met assuming the secondary MSIV fails to close for the SGTF event.

With regards to the feedwater regulating valve (FRV), which is credited to close within a specified time in the containment response analysis, the imposed closure time in the analysis NuScale Nonproprietary

needs to be clearly reflected in the FSAR. Although Technical Specification Surveillance Requirement 3.7.2.1 indicates a surveillance to verify the closure time is within limits, that value should be specified in the FSAR. Accordingly, in addition to the aforementioned augmented quality considerations, the staff requests that NuScale:

(9) Provide a closure time for the FRV corresponding to that assumed in the analysis in the FSAR.

[1] In the past, mechanical equipment classified as safety-related was referred to as meeting safety-grade qualification requirements, such as seismic, environmental, and functional requirements.

NuScale Response:

The NuScale responses to the NRCs request for additional information (RAI) is provided in nine separate responses. Each question response is provided below.

(1) and (2)

The nonsafety-related components used to mitigate a design basis accident (DBA) in the NuScale accident analysis are identified in the table below.

Tag No. Name Type MS-AOV-1003 Backup Main Steam Isolation Valve air operated, gate valve MS-AOV-2003 (MSIV)

MS-AOV-1004 Backup Main Steam Isolation Bypass air operated, gate valve MS-AOV-2004 Valve (MSIBV)

FW-CKV-1007 Backup Feedwater Check Valve nozzle check valve FW-CKV-2007 (FWCV)

FW-FCV-1006 air operated, flow control Feedwater Regulating Valve (FWRV)

FW-FCV-2006 valve (3)

NuScale does not have performance history and operating experience for these components identified in (1) and (2). Equipment requirement specifications for these valves are not currently developed. These specifications will be developed during the detailed design phase. NuScale employs good engineering practice to use industry operating history when selecting designs and use existing designs that have a proven history of reliability when determining component designs.

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(4)

NuScale system design documentation requires that the Backup MSIVs and Backup MSIBVs must function under the same steam and liquid flow rate conditions under which the safety-related MSIVs are required to operate.

NuScale system design documentation require that the FWRV must meet similar functional requirements as the safety-related Feedwater Isolation Valve (FWIV). The nonsafety-related Backup FWCV design documentation specifies requirements for the Backup FWCV similar to the safety-related FWCV that is integral to the FWIV body.

The NuScale Inservice Testing (IST) Program, as described in the Final Safety Analysis Report (FSAR) Section 3.9.6, Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restriaints, and Table 3.9-17, Valve Augmented Requirements, requires testing of the Backup MSIVs, Backup MSIBVs, Backup FWCVs and FWRVs that is similar to the testing required for the secondary system containment isolation valves (SSCIVs).

Detailed equipment requirement specifications for the Backup MSIVs, Backup MSIBVs, Backup FWCVs and FWRVs will be developed during the detailed design phase.

(5)

The valves identified in response to items (1) and (2) do not meet American Society of Mechanical Engineers (ASME) OM Code ISTA-1100 criteria for preservice and inservice testing; however, the augmented quality requirements of these valves is recognized and included in Augmented Valve Testing Program described in FSAR Section 3.9.6.5, Augmented Valve Testing Program, and Table 3.9-17, Valve Augmented Requirements. Preservice and inservice augmented testing is similar to the SSCIV testing requirements identified in the IST Program in FSAR Table 3.9-16, Valve Inservice Test Requirements per ASME OM Code.

Technical Specification (TS) 3.7.1, Main Steam Isolation Valves (MSIVs), requires two MSIVs (the two MSIVs are the safety-related MSIV and the nonsafety-related backup MSIV) and two MSIBVs (the two MSIBVs are the safety-related MSIBV and the nonsafety-related Backup MSIBV) per steam line to be operable in modes 1, 2 and 3 - when not passively cooled.

Surveillance Requirements (SRs) 3.7.1.2 and 3.7.1.3 are performed to verify MSIV and Backup MSIBV stroke time and valve leakage, respectively, in accordance with the NuScale IST Program.

TS 3.7.2, Feedwater Isolation, requires one FWIV and one FWRV for each steam generator (SG) to be operable in modes 1, 2 and 3 - when not passively cooled. SR 3.7.2.2 and 3.7.2.3 require verification of the FWIV and FWRV stroke time and valve leakage, respectively, in accordance with the IST Program.

The IST Program requires exercise testing of the safety-related FWCV and nonsafety-related Backup FWCVs at each refueling shutdown.

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(6)

The NuScale Design Certification Application (DCA) Parts 2 and 4 recognize the importance of these valves in the NPM design. Each valve is specified in FSAR Tier 2 Section 3.9.6 and Table 3.9-17 and have commensurate testing requirements as the SSCIVs in the IST Program provided in Table 3.9-16. The Backup MSIVs, Backup MSIBVs, and FWRVs have a corresponding TS limiting condition for operability (LCO) and SRs as their safety-related counterpart valves. The nonsafety-related FWCVs have similar exercise test requirements as the safety-related FWCVs.

Equipment requirement specifications for these backup valves will be developed during the detailed design phase, and the specifications will include requirements commensurate with the license basis documents.

(7)

FSAR Table 3.2-1, Classification of Structures, Systems and Components, identifies the applicable requirements for the Secondary (Backup) MSIVs and Secondary (Backup) MSIBVs, FWRVs and FW (Supply) Check Valves under their respective systems. For each valve the QA Program Applicability is identified as "AQ-S." AQ-S is specified as, "the pertinent requirements of 10 CFR 50 Appendix B are applicable to nonsafety-related SSC classified as Seismic Category I or Seismic Category II in accordance with the quality assurance program."

To clarify the intent of this statement with respect to the collection of valves used as backup valves to safety-related valves, NuScale has modified each valve entry with a note 6. Note 6 states that each valve provides nonsafety-related backup isolation to a safety-related isolation device.

(8)

A sensitivity was performed on the limiting steam generator tube failure (SGTF) case presented in FSAR Section 15.6.3, Steam Generator Tube Failure (Thermal Hydraulic). When failure of both the safety related MSIV and the nonsafety-related Backup MSIV was considered, the total mass release increased by about 50 percent before equilibrium pressure was reached between the reactor coolant system and the secondary side. This is primarily driven by the lower design pressure (1000 psi) of the steam piping from the Backup MSIV to the turbine. Following reactor and turbine trip, the main steam safety relief valve lifts until equilibrium conditions are achieved between the primary and secondary below 1000 psi. As presented in FSAR section 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, and FSAR Table 15.0-12, Radiological Dose Consequences for Design Basis Analyses, the SGTF event has significant margin to dose limits (more than 100%), therefore, it is judged that the additional mass release would not challenge normal design basis release limits for this event.

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(9)

The FWRV closure time is assumed to be 30 seconds. This is described in FSAR Section 15.2.

Impact on DCA:

FSAR Table 3.2-1 been revised as described in the response above and as shown in the markup provided in this response.

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NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components RAI 03.02.01-2, RAI 03.02.01-3, RAI 03.02.02-2, RAI 03.02.02-6, RAI 03.08.02-14, RAI 05.04.02.01-6, RAI 06.02.04-2, RAI 09.02.02-1, RAI 09.02.04-1, RAI 09.02.04-1S1, RAI 09.02.05-1, RAI 09.02.06-1, RAI 09.02.07-4, RAI 09.02.07-5, RAI 09.02.09-2, RAI 09.03.04-5, RAI 09.04.02-1, RAI 10.04.07-2, RAI 11.02-1, RAI 15-17, RAI 19-14 Table 3.2-1: Classification of Structures, Systems, and Components SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

CNTS, Containment System All components (except as listed below) RXB A1 N/A Q None B I

  • CVC Injection & Discharge Nozzles RXB A1 N/A Q None A I
  • CVC PZR Spray Nozzle
  • CVC PZR Spray CIV
  • CVC RPV High Point Degasification Nozzle
  • CVC RPV High Point Degasification CIV
  • RVV & RRV Trip/Reset # 1 & 2 Nozzles
  • RVV Trip 1 & 2/Reset #3 Nozzles
  • CVC Injection & Discharge CIVs
  • NPM Lifting Lugs RXB B1 None AQ-S
  • Top Support Structure
  • Top Support Structure Diagonal Lifting Braces
  • CNV Fasteners RXB A1 N/A Q None N/A I
  • CNV Seismic Shear Lug
  • CNV CRDM Support Frame
  • Containment Pressure Transducer (Narrow Range)
  • Containment Water Level Sensors (Radar Transceiver)
  • SG 1 & 2 Steam Temperature Sensors (RTD)

CNTS CFDS Piping in containment RXB B2 None AQ-S None B II Piping from (CES, CFDS, FWS, MSS, and RCCWS) CIVs to disconnect flange (outside containment) RXB B2 None AQ-S None D I CVCS Piping from CIVs to disconnect flange (outside containment) RXB B2 None AQ-S None C I CIV Close and Open Position Sensors: RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

  • CES, Inboard and Outboard
  • CFDS, Inboard and Outboard
  • CVCS, Inboard and Outboard PZR Spray Line
  • CVCS, Inboard and Outboard RCS Discharge
  • CVCS, Inboard and Outboard RCS Injection
  • CVCS, Inboard and Outboard RPV High-Point Degasification
  • RCCWS, Inboard and Outboard Return and Supply
  • SGS, Steam Supply CIV/MSIVs and CIV/MSIV Bypasses Containment Pressure Transducer (Wide Range) RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I
  • Containment Air Temperature (RTDs) RXB B2 None AQ-S None N/A II
  • SG tubes RXB A1 N/A Q None A I
  • Steam plenums
  • SG tube supports RXB A1 N/A Q None N/A I
  • Upper and lower SG supports
  • Steam piping inside containment RXB A2 N/A Q None B I
  • Thermal relief valves Flow restrictors RXB A2 N/A Q None N/A I Tier 2 3.2-12 Draft Revision 2

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

  • Containment sampling system sample panel RXB B2 None AQ
  • Pressure boundary components of any monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.
  • Primary sampling system sample cooler cooling water chillers RXB B2 None AQ Quality Group D D III
  • Combined polisher effluents sample line isolation valve TGB B2 None None None D III
  • Condensate polisher sample line isolation valves
  • Condensate pump discharge sample line isolation valve
  • Condenser hotwell sample line isolation valve
  • Start-up Isolation Valves RXB B2 None AQ-S None D I
  • RXB Steam Traps
  • Technical Specification Surveillance for D I
  • Secondary Main Steam Isolation Bypass Valves (Note 6) operability and in-service testing.
  • Valve Leak Detection
  • Secondary Main Steam Isolation Bypass Valve Close and Open Position Indicators RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I
  • Auxiliary Steam Supply Valve TGB B2 None None None D III
  • Auxiliary Steam Warm-up Valve TGB
  • N2 Injection Isolation Valves RXB
  • Steam Sample Panel Isolation Valve TGB
  • TGB Steam Traps TGB
  • Main Steam Pressure Transmitters RXB, TGB B2 None AQ None N/A III
  • Main Steam Temperature Elements All other components RXB, TGB B2 None None None N/A III FWS, Condensate and Feedwater System All components (except as listed below) TGB, RXB B2 None None None N/A III Feedwater Regulating Valve A/B (Note 6) RXB B2 None AQ-S Technical Specification Surveillance for D I operability and in-service testing.

Feedwater Supply Check Valve (Note 6) RXB B2 None AQ-S Inservice Testing D I Feedwater Regulating Valve Accumulators RXB B2 None AQ Technical Specification Surveillance for D III operability and in-service testing.

Feedwater Regulating Valve A/B Limit Switch RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

  • Condensate Storage Tank (located adjacent to TCB) Yard B2 None None None D III
  • Condensate Storage Tank Makeup Level Control Valve Tier 2 3.2-19 Draft Revision 2

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

((SCB, Security Buildings (Guardhouse)))

  • Security Building Yard B2 None None None N/A III
  • Vehicle inspection sally port

((ANB, Annex Building))

Annex Building Yard B2 None None None N/A III

((DGB, Diesel Generator Building))

Diesel Generator Building Yard B2 None None None N/A III

((CUB, Central Utility Building))

Central Utility Building Yard B2 None None None N/A III

((FWB, Firewater Building))

Firewater Building Yard B2 None None None N/A III CRB, Control Building CRB Structure at EL 120-0 and below (except as discussed below). Yard A1 N/A Q None N/A I

  • CRB Structure above EL 120-0 Yard B2 None AQ-S None N/A II
  • Inside the CRB elevator shaft and two stairwells, full height of structure
  • CRB Fire Protection Vestibule (on East Side of CRB)

MEMS, Metrology and Environmental Monitoring System All components Yard, CRB B2 None AQ IEEE 497-2002 with CORR 1 N/A III COMS, Communication Systems All components Yard for B2 None None None N/A III collection of data CRB for display of results SMS, Seismic Monitoring System All components RXB, CRB B2 None AQ-S None N/A I Note 1: Acronyms used in this table are listed in Table 1.1-1.

Note 2: QA Program applicability codes are as follows:

  • Q = indicates quality assurance requirements of 10 CFR 50 Appendix B are applicable in accordance with the quality assurance program (see Section 17.5).
  • AQ = indicates that pertinent augmented quality assurance requirements for non-safety related SSCs are applied to ensure that the function is accomplished when needed based on that functionality's regulatory requirements. Note that in meeting regulatory guidance, codes, and standards, those applicable SSCs may also have quality assurance requirements invoked by said guidance (e.g., RG 1.26, RG 1.143, IEEE 497, RG 1.189).
  • AQ-S = indicates that the pertinent requirements of 10 CFR 50 Appendix B are applicable to nonsafety-related SSC classified as Seismic Category I or Seismic Category II in accordance with the quality assurance program.
  • None = indicates no specific QA program or augmented quality requirements are applicable.

Note 3: Additional augmented design requirements, such as the application of a Quality Group, radwaste safety, or seismic classification, to nonsafety-related SSC are reflected in the columns Quality Group/Safety Classification and Seismic Classification, where applicable.

Note 4: See Section 3.2.2.1 through Section 3.2.2.4 for the applicable codes and standards for each RG 1.26 Quality Group designation A, B, C, and D. A Quality Group classification per RG 1.26 is not applicable to supports or instrumentation. See Section 3.2.1.4 for a description of RG 1.143 classifications for RW-IIa, RW-IIb, and RW-IIc.

Note 5: Where SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

Note 6: Provides nonsafety-related backup isolation to a safety-related isolation device. See FSAR section 15.0.0.6.6.

Tier 2 3.2-27 Draft Revision 2

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