mhi's second responses to us-apwr dcd rai no. …art mitsubishi heavy industries, ltd. 16-5,...

ArtMITSUBISHI HEAVY INDUSTRIES, LTD.

16-5, KONAN 2-CHOME, MINATO-KUTOKYO, JAPAN

March 10, 2009

Document Control DeskU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

Attention: Mr. Jeffrey A. Ciocco

Docket No. 52-021MHI Ref: UAP-HF-09080

Subject: MHI's Second Responses to US-APWR DCD RAI No.150-1635 REVISION I

References: 1) "Request for Additional Information No. 150-1635 Revision 1, SRP Section:17.04 - Reliability Assurance Program (RAP), Application Section: 17.4Reliability Assurance Program," dated January 9, 2009.

2) Letter MHI Ref: UAP-HF-09046 from Y. Ogata (MHI) to U.S. NRC, "MHI'sResponses to US-APWR DCD RAI No.150-1635," dated February 6, 2009

With this letter, Mitsubishi Heavy Industries, Ltd. ("MHI") transmits to the U.S. NuclearRegulatory Commission ("NRC") a document as listed in Enclosures.

Enclosed are the second responses to the RAIs contained within Reference 1. In the initialresponses submitted with Reference 2, MHI committed to submit responses to RAI #17-04-19,#17-04-20, #17-04-23, #17-04-24 and #17-04-30 within 60 days after RAI issue date.

Please contact Dr. C. Keith Paulson, Senior Technical Manager, Mitsubishi Nuclear EnergySystems, Inc. if the NRC has questions concerning any aspect of the submittals. His contactinformation is below.

Sincerely,

Yoshiki Ogata,General Manager-APWR Promoting DepartmentMitsubishi Heavy Industries, LTD.

Enclosure:

1. Responses to Request for Additional Information No. 150-1635 Revision 1.

CC: J. A. CioccoC. K. Paulson

Contact InformationC. Keith Paulson, Senior Technical ManagerMitsubishi Nuclear Energy Systems, Inc.300 Oxford Drive, Suite 301

IY2R&

Monroeville, PA 15146E-mail: ck [email protected]: (412) 373-6466

Docket No. 52-021MHI Ref: UAP-HF-09080

Enclosure 1

UAP-HF-09080Docket Number 52-021

Responses to Request for Additional Information No. 150-1635Revision 1

March 2009

/1

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

3/10/2008

US-APWR Design Certification

Mitsubishi Heavy Industries

Docket No.52-021

NO.150-1635 REVISION 0RAI NO.:

SRP SECTION: 17.04 - Reliability Assurance Program (RAP)

APPLICATION SECTION: 17.4 Reliability Assurance Program

DATE OF RAI ISSUE: 1/9/2008

QUESTION NO.: 17-04-19

In Table 17.4-1 ("Risk Significant SSCs") of the US-APWR DCD, Revision 1, many risk-significantSSCs are identified through text descriptions only (i.e., specific component identification numbers arenot provided). As a result, it is unclear as to what specific components are in RAP. For example, underthe Component Cooling Water System (CCWS) in Table 17.4-1, Item 6, it is not clear as to what specificcomponents are included under the description "SSCs that compose CCW boundary." Another example,under the Fire Suppression System (FSS) in Table 17.4-1, Item 7, the description "FSS -CCWSBoundary Motor Operated Valves (ACWCH-1A/B, 2A/B, 3A/B, 4A/B, 6A/B, 7A/B,8A/B)" is clearer(assuming these are the applicable valves) than the current description "FSS - CCWS Boundary MotorOperated Valves [TBD]."

The staff requests that the applicant more clearly describe the risk-significant SSCs in Table17.4-1of the US-APWR DCD by using text descriptions and specific component identification numbers, whenapplicable.

ANSWER:

As noted above, in Table 17.4-1 there are some SSCs for which the identification numbers(component IDs) have not been assigned.

Nevertheless, most of these SSCs have PRA IDs, which are the identification number named infault trees of PRA model. And by these PRA lDs, SSCs can be identified in US-APWR ProbabilisticRisk Assessment, MUAP-07030(R1).

For SSCs labeled as "[TBD]" in Table 17.4-1,o PRA ID

17-04-19-1

* Figure number of US-APWR Probabilistic Risk Assessment, MUAP-07030(R1) containing theSSC

" Component IDare tabulated as follows.

Figure No. and Pageof US-APWR

# SSCs PRA ID Probabilistic Risk Component IDAssessment, MUAP-

07030(R1)1 Accumulator injection systr _________ ________ ________

5 Piping of discharge lines train A l [from accumulator tank I Figure 6A.2-1 To be determinedthrough D I to RCS cold leg piping) (Page 6A.2-21) ___oea__er _ne

2. Charging injection systemi~ ________ ________________

______________ (Pgue 6A.4-21)22 Charging flow control orifice OR02 Figure 6A.4-1 To be determined

3 Component cooling Wter system I CCWS) ________________

21 Charging injection Pump Cooling ACWCH5A (B) To be determinedLine Check ValvesCharging injection pump cooling Figure 6A.10-1

22 discharge line motor operated ACWCH6A (B) (Page.6A.10-18) To be determinedvalves

26 CCWS - fire suppression system ACWCH2A (B)boundary motor operated valves ACWCH4A (B) To be determined

41 Containment system4 ______________ ____________

1I Containment vessel - Figure 6A15.3-1 To be determined2 Hydrogen ignition system -! (Page.6A.15.3-10) To be determined

6 Emergency feedwatersystem (EFAS)13 EFW pit discharge line piping (from EFW pits to motor-

drive / turbine-drive EFW To be determinedpumps _

14 EFW pit discharge line tie-line [between EFW pitpiping discharge line A and B: on To be determined

these lines there arePW2A(B))

15 A-D-emergency feedwater line (from the EFW pump to To be determined, A(B,C,D) piping CN)

16 T/D pump steam supply line (from CNtoT/D pumps:piping on these lines there are Figure 6A.5-1 To be determined

TS2A(B,C,D),TS3 (Page.6A.5-35)A(B,C,D) and TS1 A(B))

17 Minimum/Full flow line piping [from EFW pumpdischarge lines to EFW

pits: on these lines,there are ORTWAA(ABBA,BB), TVVT/MW7 AA To be determined

(AB BA,BB), TW6/MW6AA(AB BA,BB),

EFW01A(B) andEFW02A(B)_ I

17-04-19-2



07030(Ri)7 Emergency power source (EPS)

480V AC motor control center1 buses EPS-4ESBA(B,C,D)2 480VAC load center buses - EPS-4LCA(B,C,D)3 6.9kV buses - EPS-6ESBA(B,C,D)4 125V DC buses train A and D - EPS-DCA(D)5 125V DC buses train B and C - EPS-DCB(C)6 120V buses train A-D - EPS-VITA(B,C,D)7 Swing MCC incomer circuit EPS-4SB(D)1

breakers8 Batteries BA1A(B,C,D) EPS- BA1A(B,C,D)

6.9kV AC bus incomer circuit 6EA(B,C,D), EPS-6EA(B,C,D)breakers 6HA(B,C,D) EPS-6HA(B,C,D)

10 Gas turbine discharge circuit GTBA(B,C,D) Figure 6A.11-1 EPS- GTBA(BCD)10breakers (Page.6A. 11-53,54)ES-TABCD

Circuit breakers between 6.9kV11 bus and 6.9kV/480V safety power 41A(B,C,D) EPS-41A(B,C,D)

transformers12 MCC bus incomer circuit breakers 4JA(B,C,D) EPS- 4JA(B,C,D)

13 Circuit breakers between 125V VIT4A(BCD) EPS-VIT4A(BCD)DC bus and Inverter Vl4(,,DP-VT4 _BCD

14 Class 1E gas turbine generators GTG EPS- GTA(B,C,D)15 Gas turbines generator sequencers - To be determined16 Inverters - EPS- INVA(B,C,D)17 Main transformers _ EPS- MTF

18 6.9kV/480V safety power EPS- 4PTA(B,C,D)transformers8' Alternative AC power sources (Permanent bus)

1 Non-class 1 E gas turbine P1(2) ACC GTG EPS- GTP1 (2)1generators

2 480V permanent buses To be determined3 6.9kV permanent buses To be determined

Circuit breakers between 6.9kV4 bus and 6.9kV/480V power 41P1(2) EPS- 41P1 (2)

transformer5 Batteries To be determined

6Gas turbine generator discharge Figure 6A.11-16 circuit breakers GTBPI (2) (Page.6A.11-53,54) EPS- GTBP1(2)

7 AAC selector circuit breakers EPS- 4AA(B,C,D)Circuit breakers between 125V

8 DC bus and Inverter EPS-VIT4P1(2)

9 Inverters EPS- INVP1 (2)

10 Gas turbines generator sequencers To be determined

11 6.9kV/480V power transformers EPS-4PTP1(2)

17-04-19-3



07030(R1)

9 Non-essential chilled water system

1 Non-essential chilled water ACWCH7A (B), Figure 6A.10-1system - CCWS boundary motor ACWCH8A (B) (Page.6A.10-18) To be determinedoperated valves

C0 Fire suppression systems (ISS)1 FSS pump discharge motor FS-V-2 To be determined

operated valve2 FSS pump discharge flow meter FS-F-1 To be determined3 Reactor cavity injection line orifice FS-O-2 To be determined4 FSS piping (from tank to tie line Cfrom Raw water tank

piping) to tie line piping: on this Fige 6A.15.4-1 To be determined(Page.6A.15.4-12) T edtrieline there are FS-V-1,FS-V-2 and FS-F-i1

5 Raw water tank To be determined6 FSS pump discharge manual FSVI To be determined

valve7 FSS - CCWS Boundary motor ACWCH1A (B) Figure 6A.10-1 To be determined

_ operated valves ACWCH3A (B) (Page.6A. 10-18)II1 High head safety injection system _________________ _______________

12 Piping (from safety injection Figure 6A.1-1 Tpumps to CN) (Page.6A.1-34)

15 Instrumentation and control QI&C) s.ystem________ Figure 6A .13-2 T ed t r i eContainment pressure sensors " (Page.6A.13-32) To be determined

17 Main feedwater systeem(MFWS) ____________________________ __...._____..._......___

1 Main feedwater system

18 Main steam supply system (MSS )

3 Main steam line piping Figure 6A.6-1

(from SG to turbine) (Page.6A.6-25) To be determinedFigure 6A.6-2

(Page.6A.6-26)21 Containment spray / residual heat removal (CS/RHIR), system9 RWSP discharge line isolation 9007A (B,C,D) Figure 6A.14.3-1 CSS-MOV-

valves (Page.6A.14.3-12) 001A(B,C,D)12 Piping (from RWSP through

CS/RHR pumps and Figure 6A.3-1heat exchangers to (Page.6A.3-64)

cold leg piping or CN)2..2 Refueling water storage system (RWS) ______ _......

1 Refueling water storage pit - To be determined(RWSP) sump strainers

. N eactor protection system (RPS) ...___..._...._.... _ _ _

1 Reactor trip breakers Figue.6A.14.12-36) To be determined

17-04-19-4



07030(R1)

2 Control rod (rod injection) -JTo be determined

S Chilled water system (VW S) . ... . . . . . . . . . . .. . ...............

1 Chiller units train B and C - Figure 6A.14.4-1 VWS-PEQ-001B(C)2 Pumps train B and C - (Page.6A.14.4-12) VWS-PPP-001B(C)25 Essential service water system (E iWS) __________

14 Piping of train B and C [TBD] [from ESW pumps to Figure 6A.9 -2 To be determined15 Piping of train A and D [TBD] I discharged water pit) (Page.6A.9-47) To be determined(For piping, the information to identify the locations in the systems is supplied in the column of "PRA ID"of this list.)

These component IDs will be updated in Table17.4 incorporating the discussion of expert panel(See the Attachment to this RAI response, page 17.4-18-20, 31, 36). This will be done by the nextrevision of the US-APWR DCD.

As noted in the answer to Question No.17-04-20, identification numbers provided in US-APWRProbabilistic Risk Assessment (PRA) will be revised to component IDs in the next revision.Consequently the SSC identification numbers will be consistent in US-APWR Probabilistic RiskAssessment (PRA) and US-APWR DCD.

Impact on DCD

List of risk significant SSCs will be revised as noted above considering the discussion of expert panel inthe response to RAI 17.04-19. (See the Attachment to this RAI response, page 17.4-17-19, 30, 35)

Impact on COLA

There is no impact on COLA from this RAI.

Impact on PRA

There is no impact on PRA from this RAI.

17-04-19-5


3/10/2008



Docket No. No. 52-021

RAI NO.: NO.150-1635 REVISION 0

SRP SECTION: 17 04 - Reliability Assurance Program (RAP)



QUESTION NO. : 17.04-20

In general, the component identification numbers provided in Table 17.4-1 ("Risk Significant SSCs")of the US-APWR DCD, Revision 1, are not consistent with the component identification numbers usedin the US-APWR Probabilistic Risk Assessment (PRA), MUAP-07030(RO). For example, RefuelingWater Storage System (RWS) check valve 012A(B) in Table 17.4-1 of the US-APWR DCD, Revision 1,corresponds to check valve 006A(B) in the US-APWR PRA, MUAP-07030(R0).

The staff requests that the applicant make consistent the component identification numbersprovided in the US-APWR DCD (for example, the component identification numbers in Table 17.4-1 andChapter 19) and the component identification numbers used in the US-APWR PRA, MUAP-07030(RO).

ANSWER:

SSCs' identification numbers provided in Table 17.4 are the component IDs in P&ID. And those ofUS-APWR Probabilistic Risk Assessment (PRA), MUAP-07030(RO) are PRA IDs used in the fault treesof PRA models.

The lists showing the correspondence between component IDs and PRA IDs has been submitted as"Conversion table for P&ID to PRA components numbers" in UAP-HF-08052. And component lDs alsocan be corresponded to PRA IDs based on Table 6-2 and Table 6-3 of US-APWR Probabilistic RiskAssessment (PRA), MUAP-07030(R1).

In the next revision of the US-APWR Probabilistic Risk Assessment (PRA), MUAP-07030, PRA IDswill be revised to component IDs. Consequently the SSC identification numbers in both documents willbe consistent.

17-04-20-1

Impact on DCD

There is no impact on DCD from this RAI.

Impact on COLA


Impact on PRA


17-04-20-2


3/10/2008









The applicant did not include in Table 17.4-1 of the US-APWR DCD, Revision 1, the SSCs (e.g.,valves/orifices/coolers) necessary to provide component cooling water (CCW) cooling to the High HeadSafety Injection (HPI) pumps. Since the HPI pumps are considered risk-significant in Table 17.4-1, itwould suggest the SSCs necessary to provide CCW cooling to these pumps may also be risk-significant.

The staff requests that the applicant include in Table 17.4-1 of the US-APWR DCD the SSCs (e.g.,valves/orifices/coolers) necessary to provide CCW cooling to the HPI pumps. Otherwise, provide thebasis for not including these SSCs in Table 17.4-1 of the US-APWR DCD (include in the basis adiscussion of the associated risk importance measures from the various PRA models, consideration ofdeterministic methods, e.g., defense-in-depth, consideration of seismic margins analysis, and the expertpanel's deliberation for not including these SSCs in D-RAP).

ANSWER:

Among the SSCs that compose CCW boundary with the HPI pumps, the risk significant SSCsbased on the risk importance (e.g. RAW>2, FV>0.005) are listed with PRA IDs and component IDs asfollows.

SSCs composing CCW boundary with the PRA ID Component IDHPI pumps

A(B,C,D)-Sl pump motor outlet manual valve 133A(B,C,D) [NCS-VLV-114A(B,C,D)]A(B,C,D)-SI oil cooler outlet manual valve 160A(B,C,D) [NCS-VLV-115A(B,C,D)]A(B,C,D)-SI pump outlet manual valve 161A(B,C,D) [NCS -VLV- 116A(B,C,D)]A(B,C,D)-SI pump outlet manual valve 132A(B,C,D) T [NCS -VLV- 119A(B,C,D)]

17-04-23-1

In Table 17.4-1 of the US-APWR DCD, Revision 1, these SSCs are listed as "SSCs that composeCCW boundary" in Item 6 of Component cooling water system (CCWS).

These SSCs will be included respectively in Table 17.4-1 incorporating the discussion of expertpanel (See the Attachment to this RAI response, page 17.4-11, 12). This will be done by the nextrevision of the US-APWR DCD.

On the other hand, the following SSCs also composing CCW boundary with the HPI pumps are notidentified risk significant by the risk importance.

SSCs composing CCW boundary with the P IDPRA IDComponent IDHPI pumps

A(B,C,D)- SI pump outlet orifice 1260A(B,C,D) [NCS-FE-1270A(B,C,D)]A(B,C,D)- SI pump motor outlet orifice 1266A(B,C,D) [NCS-FE-1274A(B,C,D)]A(B,C,D)- SI pump inlet manual valve CCW0002A (B,C,D) [NCS-VLV-111 A(B,C,D)]A(B,C,D)- Oil cooler [TBD]

The treatment of these SSCs will be discussed in the expert panel.

Impact on DCD

List of risk significant SSCs will be revised as noted above considering the discussion of expert panel inthe response to RAI 17.04-23. (See the Attachment to this RAI response, pagel7.4-11, 10)

Impact on COLA


Impact on PRA


17-04-23-2


3/10/2008









The applicant did not include in Table 17.4-1 of the US-APWR DCD, Revision 1, the SSCs (e.g.,valves/orifices/coolers) necessary to provide component cooling water (CCW) cooling to theContainment Spray/Residual Heat Removal (CS/RHR) pumps and heat exchangers. Since the CS/RHRpumps and heat exchangers are considered risk-significant in Table 17.4-1, it would suggest the SSCsnecessary to provide CCW cooling to these pumps and heat exchangers may also be risk-significant.

The staff requests that the applicant include in Table 17.4-1 of the US-APWR DCD the SSCs (e.g.,valvesYorifices/coolers) necessary to provide CCW cooling to the CS/RHR pumps and heat exchangers.Otherwise, provide the basis for not including these SSCs in Table 17.4-1 of the US-APWR DCD(include in the basis a discussion of the associated risk importance measures from the various PRAmodels, consideration of deterministic methods, e.g., defense-in-depth, consideration of seismicmargins analysis, and the expert panel's deliberation for not including these SSCs in D-RAP).

ANSWER:

Among the SSCs that compose CCW boundary with the CS/RHR pumps and heat exchangers, therisk significant SSCs based on the risk importance (e.g. RAW>2, FV>0.005) with PRA IDs andcomponent IDs as follows.

SSCs composing CCW boundary with the CS/RHR PRA ID SSC IDpumps and heat exchangers

A(B,C,D)- CS/RHR pump outlet manual valve 187A(B,C,D) [NCS-VLV-131A(B,C,D)]A(B,C,D)- CS/RHR pump motor outlet manual valve 183A(B,C,D) [NCS-VLV-128A(B,C,D)]

17-04-24-1

A(B,C,D)- CS/RHR pump outlet orifice 1244A(B,C,D) [NCS-FE1246A(B,C,D)]A(B,C,D)- CS/RHR pump motor outlet orifice 1246A(B,C,D) [NCS-FE1250A(B,C,D)]A(B,C,D)- CS/RHR pump inlet manual valve CCW0003A(B,CD) [NCS-VLV-125A(B,C,D)]A(B,C,D)- CS/RHR heat exchanger inlet manual valve 107A(B,C,D) [NCS-VLV-141A(B,C,D)]A(B,C,D)- CS/RHR heat exchanger outlet orifice 1242A(B,C,D) [NCS-FE1242A(B,C,D)]A(B,C,D)-CS/RHR CS/RHR heat exchanger outlet 113A(BCD) [NCS-VLV-144A(BCD)]manual valveA(B,C, D)- CS/RHR heat exchanger outlet valve 114A(B,C,D) [NCS-MOV-145A(B,C,D)]


These SSCs will be included respectively in Table 17.4-1 incorporating the discussion of expertpanel (See the Attachment to this RAI response, page 17.4-12, 13). This will be also done by the nextrevision of the US-APWR DCD.

Impact on DCD

List of risk significant SSCs will be revised as noted above considering the discussion of expert panel inthe response to RAI 17.04-24. (See the Attachment to this RAI response, page 17.4-12, 13)

* Impact on COLA


Impact on PRA


17-04-24-2


3/10/2008









The applicant did not include in Table 17.4-1 of the US-APWR DCD, Revision 1, the SSCs (e.g.,valves/orifices/coolers) necessary to provide component cooling water (CCW) cooling to the ChargingInjection System (CHI) pumps. Since the CHI pumps are considered risk-significant in Table 17.4-1, itwould suggest the SSCs necessary to provide CCW cooling to these pumps may also be risk-significant.

The staff requests that the applicant include in Table 17.4-1 of the US-APWR DCD the SSCs (e.g.,valves/orifices/coolers) necessary to provide CCW cooling to the CHI pumps. Otherwise, provide thebasis for not including these SSCs in Table 17.4-1 of the US-APWR DCD (include in the basis adiscussion of the associated risk importance measures from the various PRA models, consideration ofdeterministic methods, e.g.,defense-in-depth, consideration of seismic margins analysis, and the expertpanel's deliberation for not including these SSCs in D-RAP).

ANSWER:

Among the SSCs that compose CCW boundary with CHI pumps, the risk significant SSCs based onthe risk importance (e.g. RAW>2, FV>0.005) are listed with PRA IDs and component IDs as follows.

SSCs composing CCW boundary with the CHI P ID SSC IDpumps

A(B)- seal water heat exchanger inlet manual valve 224A(B) [NCS-VLV-311A(B)]A(B)- CHI oil cooler inlet manual valve 225A(B) [NCS-VLV-312A(B)]A(B)- CHI pump motor inlet manual valve 226A(B) [NCS-VLV-301A(B)]

17-04-30-1


These SSCs will be included respectively in Table 17.4-1 incorporating the discussion of expertpanel (See the Attachment to this RAI response, page 17.4-13). This will be also done by the nextrevision of the US-APWR DCD.

On the other hand, the following SSCs composing also CCW boundary with the HPI pumps are notidentified risk significant by the risk importance.

SSCs composing CCW boundary with the CHISSscmpsn pmsjPRA ID SSC ID

pumpsA(B)- CHI pump motor line orifice FE1257(8) [NCS-FE-1266(7)]

A(B)- CHI oil cooler line orifice FE1254(5) [NCS- FE -1260(1)]A(B)- CHI oil cooler OILA(B) [TBD]A(B)- seal water heat exchanger SWCA(B) [TBD]

The treatment of these SSCs will be discussed in the expert panel.

Impact on DCD

List of risk significant SSCs will be revised as noted above considering the discussion of expert panel inthe response to RAI 17.04-30. (See the Attachment to this RAI response, page 17.4-13)

Impact on COLA


Impact on PRA


17-04-30-2

Attach

17. QUALITY ASSURANCE AND US-APWR Design Control DocumentRELIABILITY ASSURANCE

CONTENTS

Page

17.0 QUALITY ASSURANCE AND RELIABILITY ASSURANCE .................... 17.1-1

17.1 . Quality Assurance During the Design Phase ............................................. 17.1-1

17.2 Quality Assurance During the Construction and Operations Phase .......... 17.2-1

17.3 Quality Assurance Program ....................................................................... 17.3-1

17.4 Reliability Assurance Program .................................................................... 17.4-1

17.4.1 New Section 17.4 in the Standard Review Plan ............................... 17.4-1

17 .4 .2 Introductio n ....................................................................................... 17 .4-1

17 .4 .3 S co pe ................................................................................................ 17 .4 -2

17.4.4 Quality Controls ................................................................................ 17.4-2

17.4.5 Integration into Existing Operational Programs ................................ 17.4-3

17.4.6 Operating Experience ....................................................................... 17.4-4

17 .4 .7 D -R A P .............................................................................................. 17 .4 -4

17.4.7.1 SSCs Identification ........................................................................ 17.4-4

17.4.7.2 Expert Panel ............................................... :-*--*---** ..... 111*111,17.4-5

17.4.7.3 Phase I D-RAP Implementation and SSCs included ..................... 17.4-5

17.4.8 ITAAC for the D-RAP ...................................................................... 17.4-41

17.4.9 Combined License Information ....................................................... 17.4-41

17.4.10 References ..................................................................................... 17.4-42

17.5 Quality Assurance Program Description ..................................................... 17.5-1

17.5.1 Combined License Information ......................................................... 17.5-1

17.5.2 R eferences ....................................................................................... 17.5-1

17.6 Description of the Applicant's Program for Implementation of 10 CFR 50.65,the Maintenance Rule ................................................................................ 17.6-1

Tier 2 17-1 Revision 1


17.6.1 Com bined License Inform ation ......................................................... 17.6-1

Tier 2 17-11 Revision 1

17. QUALITY ASSURANCE ANDRELIABILITY ASSURANCE

US-APWR Design Control Document

TABLES

Paqe

Table 17.4-1 Risk significant SSCs .......................................................................... 17.4-6

Tier 2 17-iii

Revision I

Tier 2 17-iii Revision I


ACRONYMS AND ABBREVIATIONS

AAC alternative ACac alternating currentCAP corrective action programCCF common cause failureCCW component cooling waterCCWS component cooling water systemCDF core damage frequencyCFR Code of Federal RegulationsCOL Combined LicenseCOLA Combined License ApplicationCS containment sprayCSS containment spray systemCVCS chemical volume control systemDAS diverse actuation systemdc direct currentDCD Design Control DocumentD-RAP design reliability assurance programDVI direct vessel injectionECCS emergency core cooling systemEFW emergency feedwaterEFWP emergency feedwater pitEFWS emergency feedwater systemEJ engineering judgeEP expert panelEPS emergency power sourceESF engineered safety featuresESW essential service waterESWS essential service water systemFIRE FIRE eventFLOOD FLOOD eventFSS fire suppression systemsFV Fussell VeselyFVW Fussell Vesely worthHSIS human-system interface systemHVAC heating, ventilation, and air conditioning

Tier 2 17-iv Revision I


ACRONYMS AND ABBREVIATIONS

I&C instrumentation and controlITAAC inspection, test, analyses, and acceptance criteriakV kilovoltLOCA loss-of-coolant accidentLOOP loss of offsite powerLPSD low power and shut down operationM/D motor drivenMCC motor control centerMFWS main feedwater systemMHI Mitsubishi Heavy Industries, Ltd.MOV motor operated valveMSS main steam supply systemNESH Nuclear Energy Systems HeadquartersNRC U.S. Nuclear Regulatory CommissionO-RAP operational reliability assurance programPAM postaccident monitoringPCMS plant control and monitoring systemPRA probabilistic risk assessmentQA quality assuranceQAP quality assurance programQAPD quality assurance program descriptionRAP reliability assurance programRAW risk achievement worthRCP reactor coolant pump

RCS reactor coolant systemRG Regulatory GuideRHR residual heat removalRHRS residual heat removal systemRPS reactor protection systemRRW risk reduction worthRTNSS regulatory treatment of non-safety-related systemsRWAT refueling water auxiliary tankRWS refueling water storageRWSP refueling water storage pitRWSS refueling water storage system

Tier 2 17-v Revision 1


ACRONYMS AND ABBREVIATIONS (Continued)

SBO station blackoutSDV safety depressurization valve

SFP spent fuel pitSFPCS spent fuel pit cooling and purification system

SG steam generatorSGTR steam generator tube ruptureSIS safety injection systemSRP Standard Review PlanSSC structure, system, and componentT/D turbine drivenVCT volume control tankVWS chilled water systemWMS waste management system

Tier 2 17-vi

Revision I

Tier 2 17-vi Revision I


17.0 QUALITY ASSURANCE AND RELIABILITY ASSURANCE

Quality Assurance Program Description (QAPD) as described in Sections 17.1, 17.2,17.3 and 17.5 of this chapter of DCD is applicable for Quality Assurance (QA) duringdesign certification.

17.1 Quality Assurance During the Design Phase

For quality assurance during the design certification phase, see Section 17.5.

The Combined License (COL) Applicant is responsible for development a QualityAssurance Program applicable to its activities during design other than the DesignCertification.

Tier 2 17.1-1

Revision I

Tier 2 17.1-1 Revision I



17.2 Quality Assurance During the Construction and Operations Phase

The COL Applicant is responsible for development of the construction and operationalphase Quality Assurance Program.

Tier 2 17.2-1 Revision ldep080708R0



17.3 Quality Assurance Program

The General Manager of Nuclear Energy Systems Headquarters (NESH) is responsiblefor the Design Certification Activities of US-APWR. The major design activities areperformed by the Nuclear Energy Systems Engineering Center engineers. QA Programcontrols governing the activities are specified in QAPD (PQD-HD-1 9005 Rev.--42) (Ref LSPLA 1676-039]

17.4-2 Ref 17.5.5-4).

Subcontractors of the Nuclear Energy Systems Engineering Center performing designactivities in support of the US-APWR are also required to follow QAPD (PQD-HD-19005Rev.42). 11 SPLA 1676-039

For the quality assurance program description during the design certification phase, seeSection 17.5.

The COL applicant is responsible for development a Quality Assurance ProgramDescription during design other than the Design Certification, construction and operationphase.

Tier 2 17.3-1

Revision I

Tier 2 17.3-1 Revision 1


17.4 Reliability Assurance Program

This section presents the US-APWR reliability assurance program (RAP).

17.4.1 New Section 17.4 in the Standard Review Plan

As noted in Item E of SECY 95-132 (Ref. 17.4-1), an applicant for design certificationshould establish the scope, purpose, objective, and essential elements of an effective D-RAP and would implement those portions of the D-RAP that apply to design certification.A COL Applicant is responsible for augmenting and completing the remainder of the D-RAP to include any site-specific design information and identify the risk-significant SSCs.Once the site-specific D-RAP is established and the risk-significant SSCs are identified,the procurement, fabrication, construction, and preoperational testing can beimplemented in accordance with the COL holder's D-RAP or other programs and wouldbe verified using the inspections, test, analyses and acceptance criteria (ITAAC) process.

17.4.2 Introduction

The purposes of the US-APWR RAP are to provide reasonable assurance that: 1) theUS-APWR is designed, constructed, and operated in a manner that is consistent with theassumptions and risk insights for the risk-significant SSCs, 2) the risk-significant SSCs ILPLA 1474-0 1Jdo not degrade to an unacceptable level during plant operations, 3) the frequency oftransients that challenge SSCs is minimized, and 4) the SSCs function reliably whenchallenged. An additional goal is to facilitate communication between the probabilisticrisk assessment (PRA), the design, and the ultimate COL activity.

The PRA evaluates the US-APWR design response to a spectrum of initiating events toensure that plant damage has a very low probability and that risk to the public isminimized. Risk significant SSCs for the US-APWR design control document (DCD) areidentified and made available to the design organization.

The US-APWR D-RAP process is implemented in several phases. Phase I, the DesignCertification phase, collects system information and develops a system model. Thissystem information and model is used as input to the design phase PRA, an operatingexperience review, and a review for external events. The goal of the RAP during thisstage is to ensure that the reactor design meets the purposes above, through the design,procurement, fabrication, construction and preoperational testing activities and programs.The results of each of these activities are provided to an expert panel (EP) whichidentifies risk significant items using probabilistic, deterministic, and other methods forinclusion in the program. Phase II, the site-specific phase, introduces the plant's site-specific information to the D-RAP process. During Phase II, the site-specific SSCs arecombined with the US-APWR design SSCs into a list for the specific plant. Phase III, thelast phase of the D-RAP, implements the procurement, fabrication, construction, andpreoperational testing. The site-specific list of SSCs is also provided as an input to theoperational phase of RAP (O-RAP) which addresses the specific plant operation andmaintenance activities. The designer, MHI, is responsible for Phase I of the D-RAP. Theobjective during this stage is to ensure that the reliability for the SSCs within the scopeof the RAP is maintained during plant operations. Phases II and III of the D-RAP and theO-RAP are the responsibility of the COL Applicant. The COL Applicant will specify the



policy and implement procedures to address the specific plant operation and

maintenance activities associated with the risk-significant SSCs identified by the D-RAP.

17.4.3 Scope

The US-APWR D-RAP identifies risk-significant SSCs and provides risk insights andreliability assumptions for aspects of plant operation, maintenance, and performancemonitoring to be addressed to ensure safe, reliable plant operation or mitigate planttransients or other events that could present a risk to the public. The risk-significantSSCs are identified using PRA, deterministic, or other methods of analysis, includingindustry experience, and EPs.

17.4.4 Quality Controls

a. Organization

The MHI is responsible for Phase I of the D-RAP.

General Manager, US-APWR project: The General Manager, US-APWR project isoverall responsible for the establishment of and implementation of the US-APWR D-RAP.In this regard, the General Manager or his designated representative is responsible toassure all affected organizations are aware of the D-RAP, its purpose, and therequirements herein.

General Manager, Reactor and Plant Safety: The General Manager, Reactor and PlantSafety, is responsible for the use of the PRA results and risk insights for the EP, and forthe conduct and coordination of the EP. The Reactor and Plant Safety organizationincludes the risk and reliability organization.

General Manager, QA: The General Manager, QA is responsible to assure properimplementation of QA program elements. This includes design control, procedures andinstructions, records, corrective actions and audits pertaining to the D-RAP.

General Managers, Design Engineering: The General Managers, Design Engineering,are responsible to implement this D-RAP and specifically to assure that the US-APWR isdesigned consistent with the reliability assumptions and insights of the PRA for risk-significant SSCs.

The risk and reliability organization is responsible to ask the related design engineeringsections to review key assumptions and to feed back their comments to ensure keyassumptions are realistic and achievable.

The risk and reliability organization is responsible to provide the RAP related inputs inthe design process by participating in the design change process.

The risk and reliability organization is also responsible to involve in the design review.

b. Design Control



The list of risk-significant SSCs for the D-RAP and its key assumptions shall bemaintained by the risk and reliability organization. The list and changes thereof shall beapproved by the EP and be provided to design engineering and QA staff working on theUS-APWR project.

The risk and reliability organization shall ensure that the design engineers are providedthe list of risk-significant SSCs for the D-RAP and its key assumption. The designengineers shall take into account the list of the risk-significant SSCs for the D-RAP andits key assumptions in their design activities and give some feedback to the risk andreliability organization in order to ensure that the key assumptions are realistic andachievable, if necessary.

c. Procedures and Instructions

General Manager, US-APWR project or his designated representative has prepared theprocedures and instructions used in implementation of the D-RAP. General Manager,US-APWR project is responsible for development and verification of implementation ofthe D-RAP, and for assuring all affected MHI organizations are aware of the D-RAP.

d. Records

Records related to the D-RAP which are required to be maintained include the following:

- List of Risk-Significant SSCs

- EP meeting minutes/summaries

- Other quality assurance program records in accordance with the US-APWRQAPD (Ref. 17.4-2) for design certification.L SPLA 1676-0379

e. Corrective action

Deficiencies identified where design documents address SSC reliability assumptionswhich are not compatible with the reliability assumptions of the PRA, or are notachievable or are unrealistic shall be entered into the corrective action program (CAP)system and addressed appropriately. The CAP utilized to support the QAPD can beused to implement the corrective actions related to the RAP.

f. Audit

Audit plans shall include for consideration, sampling the effectiveness of implementationof RAP implementation procedure. Audits shall consider several key aspects of the RAPincluding the identification of risk-significant SSCs, whether design and procurementinformation is consistent with the risk insights from the PRA, and whether assumedequipment reliability is determined to be practicable or achievable.

17.4.5 Integration into Existing Operational Programs

The US-APWR D-RAP is a source to other administrative and operational programs.Certain risk-significant SSCs identified in the D-RAP are included in existing operationalprograms such as the technical specifications surveillance requirements and provide




assurance that the reliability values assumed in the PRA will be maintained throughoutthe plant life. The O-RAP implements the measures that yield the significantimprovements in the PRA through the plant's existing programs for maintenance or QA.Implementation of the Maintenance Rule requirements contained in 10CFR50.65 (Ref.17.4-23) is an example of how the plant could address the enhanced treatment of certain ISPLA 1676-039SSCs in the O-RAP. Per SECY 95-132, the COL Applicant may meet most of theobjectives of the O-RAP via existing programs such as maintenance rule, in-servicetesting, and QA. The COL Applicant must address non-safety risk significant SSCs.

17.4.6 Operating Experience

Consideration and use of operating experience is vital to the overall objective of theD-RAP. Operating experience is considered along with various PRA analytical andimportance measures when developing a comprehensive risk analysis. The EPconsiders component operating history and industry operating experience when it can beapplied to assessing risk significance. For example, operating experience indicates thatmotor driven and turbine driven pumps may have different reliability.

The review of operating experience investigates situations where previous failures ofcomponents in similar design applications have led to functional failures of SSCs. Thereview of operating experiences is not limited to hardware failure but also extends tosituations where human performance led to functional failures of SSCs of a similarsystem design. As an example, the US-APWR design improves reliability and eliminatesrequired operator actions to switch over from injection to recirculation typical inconventional PWRs.

17.4.7 D-RAP

As discussed in Section 17.4.2, Phase I of the D-RAP includes the initial identification ofSSCs to be included in the program, implementation of the aspects applicable to designefforts, and definition of the scope, requirements, and implementation options to beincluded in the later phases.

17.4.7.1 SSCs Identification

During the US-APWR design phase, risk significant SSCs are identified for inclusion inthe scope of the D-RAP. A list of risk significant SSCs is developed and controlled as adesign input for consideration during the design phase. The list of risk significant SSCsis initially based on the results of the PRA and the EP. For further discussion on PRA,refer to Chapter 19, Section 19.1, of this DCD. The PRA is used to identify risksignificant SSCs based on risk achievement worth (RAW) and Fussell-Vesely Worth(FVW). For further information, see Chapter 19, Section 19.1.7.4 of this DCD. The list ofrisk significant SSCs identified during the design phase is updated when the plant-specific PRA is developed. In addition to the PRA input, information from operatingexperience of Japanese design plants, as well as US industry experience is consideredfor identification of risk significant SSCs. A third source in the D-RAP process foridentifying risk significant SSCs is the use of an EP consisting of representatives fromDesign Engineering, PRA, as well as other highly qualified individuals with operations,and maintenance experience who are independent of the PRA Section. The EP alsoI I SPLA 1474-006



reviews the categorization of SSCs determined to be not risk significant (NRS) fromquantified PRA results (e.g., technical adequacy of the basis used in the categorization,review of defense-in-depth implications, review of safety margin implications). As part ofthe D-RAP process, the PRA analytical results, operating experience, and an EPprocess are combined to develop a comprehensive list of risk significant SSCs.

17.4.7.2 Expert Panel

An EP, consisting of !gh__uesentatives=_f=Riati d• "A^R... ass fro - I4k SPLA 1474007

E-rn4iinfel at least one person with design engineering experience, at least one personwith PRA experience, at least one person with operations and maintenance experience.and at least one person with quality assurance experience, is responsible for the finalselection of the SSCs included in the D RAP. Industrywopefating=experien=cewhen=it=ca anbe=app=•iedýto=assessin=risk=sig=nificane•=and=engneerting=judgment=are=empleyed=iI fSPLA 1474-0081

e ng4h&oSSs4- he4-RAR. Industry operating experience and useof the Expert Panel are used as the part of deterministic approach and other processes,and engineering judgment are employed in considering the addition of SSCs to the D-IRAP. Ench ,oting Momber of tP_ d th. , o, o of , duct.,-, _nd

U.- f-J y= R . The level of education and experience of voting memberlof the RAP EP is defined in the Expert Panel Implementinq Procedure for US-APWR

SPLA 1474-014]

Reliability Assurance Program as follows:

A person who has graduated science and technology university or who hasidentical educational background, and who has more than 10 years of experiencein the specific area of Nuclear Power Plant, such as design, or has identicalexperience.

or

A person who has graduated high school or who has identical educationalbackground, and who has more than 15 years of experience in the specific areaof Nuclear Power Plant, such as design, or has identical experience.

17.4.7.3 Phase I D-RAP Implementation and SSCs included

The implementation of the Phase I D-RAP is the responsibility of MHI as it applies to thereactor design process. The SSCs included in this phase are listed in Table 17.4-1.

Tier 2 17.4-5

Revision I


Table 17.4-1 Risk significant SSCs (sheet I of 34)

# Systems, Structures and Rationale 1 Insights and AssumptionsComponents (SSCs)

1 Accumulator injection system

1 Discharge line secondary isolation RAW/CCF The accumulator provides safety injection function forcheck valves train A through D refill and re-flooding of the reactor vessel following a loss[VLV-102A (B,C,D)] of coolant accident (LOCA). Also provides negative

2 Boundary check valves train A RAW/CCF reactivity to shutdown the reactor.through D (Discharge line)[VLV-103A (B,C,D)] Single failure of any SSCs listed here has potential to

3 Discharge line isolation motor RAW cause failure of its dedicated train to inject coolant tooperated valves train A through D RCS.[VLV-101A (B,C,D)]

4 Discharge line orifices train A through RAWD[RO06A (B,C,D)]

5 Piping of discharge lines train A RAWthrough D[TBD]

6 A-D-Accumulators EJ[SIS-CTK-001A (B,C,D)]

m C

>0

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Table 17.4-1 Risk significant SSCs (sheet 2 of 34)-I

-4

4.1

Systems, Structures andComponents (SSCs) Rationale(1) Insights and Assumptions

Charging line[AOV-146][FCV-1 38][AOV-1 59]

2 RCP seal cooling injection line air RAWoperated valves[FCV-140][AOV-1 65]

3 Auxiliary spray injection line air RAW/LPSDoperated valve[AOV-1 55]

4 A,B-Charging pumps RAW/CCF/LPSD[CVS-RPP-001A (B)]

5 Volume control tank discharge line RAW/LPSDcheck valve[VLV-1 25]

6 Volume control tank discharge line RAW/LPSDmotor operated valves[LCV-1 21 B][LCV-1 21 C]

7 RWS refueling water auxiliary tank RAW(L2)discharge line change valves1.L .V-i.21..(E)I _.

7-8 RWS refueling *water auxiliary tank RAW/LPSDdischarge line check valve[VLV-595]

I H U UlItIIHdi VUIUIIIU LUFILnUI sysLUITI IkUVI, ) ImIdIIldI~nIs

appropriate volume and quality of reactor coolant for theprimary reactor coolant system, adjusts boronconcentration for the chemical shim control, and suppliesseal water to the reactor coolant pump seals, anddisposes borated water .discharged from the primaryreactor coolant system.

RCP seal water injection provided by the CVCS is anessential function to prevent RCP seal LOCA under lossof CCW conditions. When loss of CCW occurs, either thefire suppression system or the non-essential chilledwater system is connected to the charging pump coolingline. Thus, the RCP seal water injection is maintainedunder loss of CCW conditions.

Since CVCS is not completely separated in trains, largeexternal leak from SSCs that result in loss of inventory isassumed to result in degradation or failure of the system.Accordingly, SSCs that has the potential of large leak arerisk significant.

SSCs that have potential to cause common causefailures among multiple trains are also important. Suchcommon cause failure results in loss of redundant SSCs.

-4

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0>-

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0

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89 RWS refueling water auxiliary tankdischarge line manual valve[VLV-5911

RAW/LPSD

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Table 17.4-1 Risk significant SSCs (sheet 3 of 34)

# Systems, Structures and Rationale~1 ) Insights and AssumptionsComponents (SSCs)

-91O Charging pump minimum flow line RAW/LPSDcheck valves During low power and shutdown operation, CVCS[VLV-129A (B)] provides RCS make up function. On low VCT level,

4-011 Charging pump discharge line check RAW/LPSD suction is switched from the VCT to the refueling watervalves auxiliary tank, which is supplied by the refueling water[VLV-131A (B)] storage pit.

4412 Charging line containment isolation RAW/LPSDcheck valve Low-pressure letdown line isolation valves are[VLV-1 53] automatically closed and the CVCS is isolated from the

42-13 Charging line isolation check valve RAW/LPSD RHRS with receiving the RCS loop low-level signal to[VLV-1 60] prevent loss of RCS inventory at mid-loop operation.

4.314 Charging line boundary isolation RAW/LPSD When these valves are not closed, loss of a RCScheck valve inventory is prevented by manually closing the air-[VLV-1 611] operated valve at the downstream of these valves.

4415 RCP seal water injection line RAWboundary isolation check valves[VLV-182A (B,C,D)]

4-516 RCP seal water injection line RAWsecondary isolation check valves[VLV-181A (B,C,D)]

4-617 RCP seal water injection line third RAWisolation check valves[VLV-1 79A (B,C,D)]

4-18 Charging line containment isolation RAW/LPSDmotor operated valve

-1 [MOy- 52]

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# Systems, Structures and Rationaie(1 ) Insights and AssumptionsComponents (SSCs)

4-819 Charging line containment isolation RAW/LPSD The "Insights and Assumptions" for these SSCs aremotor operated valve described on the previous page.[MOV-1 51]

4-920 RCP seal water injection line RAWcontainment isolation motoroperated valves[MOV-178A (B,C,D)]

2-021 Charging line orifice RAW/LPSD[FE-1 38]

2422 Charging flow control orifice RAW/LPSD[TBDi

2223 RCP seal water injection line orifices RAW[FE-160A (B,C,D)]

2-324 Regenerative heat exchanger RAW/LPSD[CHX-001]

2425 Charging pump minimum flow line RAW/LPSDmanual valves[VLV-130A (B)]

2526 Charging pump discharge line RAW/LPSDmanual valves[VLV-1 32A (B)]

2.627 Charging pump discharge line cross RAW/LPSDtie-line manual valve[VLV-1 33]

2-728 Charging pump suction line manual RAW/LPSDvalves[VLV-1 26A (B)]

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# Systems, Structures and Rationale~') Insights and AssumptionsComponents (SSCs)

2-29 Charging line manual valves RAW/LPSD The "Insights and Assumptions" for these SSCs are[VLV-145] described on the previous page.[VLV-147]

2ý930 Charging line by-pass line manual RAW/LPSDvalve[VLV-144]

30,31 RCP seal water injection line manual RAWvalves[VLV-1 64][VLV-1 66][VLV-1 68][VLV-170B][VLV-1 71 B][VLV-1 73]

34-32 RCP seal water injection by-pass RAWline manual valve[VLV-1 63]

3-2-33 RCP seal water injection line manual RAWvalves[VLV-180A (B,C,D)]

3334 RCP seal water injection line needle RAWvalves[VLV-177A (B,C,D)]

3435 Low-pressure letdown line air LPSDoperated valve[HCV-102]

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Systems, Structures and n 1 IComponents (SSCs) Rationale Insights and Assumptions

3 Componentcooling water system (CCWS)

1 CCW pump discharge line check RAW/CCF/LPSD The component cooling water system (CCWS) transfervalves heat from plant safety-related components to the[VLV-016A (B,C,D)) essential service water system (ESWS). This system

2 A-D-Component cooling water FV/RAW/CCF supports various safety and non-safety mitigationpumps /LPSID systems. Accordingly, reliability of CCWS emergency[NCS-RPP-001A (B,C,D)] feedwater system (EFWS) has significant impact on risk.

3 A-D-Component cooling water heat RAW/CCF/LPSD CCWS has four trains, each having a component coolingexchangers water pump and a component cooling water heat[NCS-RHX-001A (B,C,D)] exchanger. Two trains compose a subsystem, which

shares a supply / return header and a surge tank.4 CCW pump discharge cross tie-line RAW/CCF/LPSD

motor operated valves SSCs that have either of the following characteristics are[MOV-020A (B,C,D)] risk significant.

5 CCW pump suction line cross tie-line RAW/CCF/LPSD - SSCs that have potential to cause common causemotor operated valves failures among multiple trains. Common cause[MOV-007A (B,C,D)] failure of such system will result in loss of multiple

trains.6 SSCs that compose CCW boundary RAW/EJ/LPSD - SSCs that have potential to cause large external7 Safety Injection pump motor outlet RAW(L2) leak are risk significant. Since the two trains that

manual valve compose a subsystem are not physically isolated,[NCS-VLV-114A(B.C.D)I large external leak from SSCs that result in loss of

8 Safety Injection oil cooler outlet RAW(L2) inventory is assumed to result in degradation ormanual valve failure of two trains.[NCS-VLV-115A(Bý .C.D)]

9 Safety Injection pump outlet manual RAW(L2)valve[NCS -VLV- 116A(BCD_

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10 Safety Injection pump outlet manual RAW(L2)valve These valves are used (opened) to provide alternative[NCS-VLV- 119A(..Pu] CCW from the fire suppression system or the non-

:7 CS/RHR heat exchanger discharge FV/RAW/CCF essential chilled water system to the charging pump11 line motor operated valves /LPSD cooling line under loss of CCW events. These are

[MOV-145A (B,CD)] important SSCs at loss of CCW events to prevent RCP12 CS/RHR pump outlet manual valve RAW(L2)/LPSD seal LOCA.

[NCS-VLV-1 31A(BC,D)]13 CS/RHR pump motor outlet manual

valve RWL2)/LPSD

[NCS-VLV-1 28A(BCDV14 CS/RHR pump outlet orifice RAW(L2)/LPSD

[NCS-FE 1246A(BC.Q,D15 CS/RHR pump motor outlet orifice RAW(L2)/LPSD

[NCS-FE1 250A(B.CD)]16 CS/RHR pump inlet manual valve RAW(L2)/LPSD

[NCS-VLV-1 25A(BC,D)17 CS/RHR heat exchangqer inlet

manual valve RAW(L2)/LPSD[NCS-VLV-141A(BCD)]

18 CS/RHR heat exchanger outlet RAW(L2)/LPSDorifice[NOS-FE1242A(BC.QD)I

19 CS/RHR CS/RHR heat exchanger RAW(L2/LPSDoutlet manual valve[NCS-VLV-144A(BC0D)]

;Umi-U,F

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20 CS/RHR heat exchanger outlet valve RAW/FV.RAW(L2)/[NCS-MOV-145A(B,C,D)] LPSD/

FV, RAW(FLOOD)/RAW(FIRE)

8 Charging injection Pump Cooling RAW/CCF/LPSDLine Check Valves

2_ [TBD]

9 Charging injection pump cooling RAW/CCF/LPSD22 discharge line motor operated valves

[TBD]23 CHI seal water heat exchanger inlet LPSD

manual valve -A(B)[NCS-VLV-311A(B)]

24 CHI oil cooler inlet manual valve-A(B) LPSD[NCS-VLV-312A(B)]

25 CHI pump motor inlet manual valve - LPSD

[NCS-VLV-301A(B)]4-0 CCWS _ fire suppression system RAW/CCF/LPSD26 boundary motor operated valves

[TBD]4-14 CCWS - RWSP line boundary check RAW/LPSD27 valves

[VLV-065A (B)]-1-2 CCWS - RWSP line boundary RAW/LPSD28 manual valves

[VLV-066A (B)]

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4 ;Containment system1 Containment vessel EJ The containment vessel is designed to completely

[TBD] enclose the reactor and reactor coolant system and toensure that essentially no leakage of radioactivematerials to the environment would result even if a majorfailure of the reactor coolant system were to occur.

[TBD] Hydrogen ignition system are provided for protectionagainst possible detonation following a core damage

accident to meet the requirement of 10CFR50.34(f) and1 OCFR50.44(c).

5 Containment isolation system-

1 Instrument air system check valve RAW(L2) In the case of core damage accident, the containment[VLV-003] isolation valve is important to prevent radionuclide

I_ releases to the environment.

m Cr->

C>

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Systems, Structures andComponents (SSCs) Rationale(1 ) Insights and Assumptions

EFW pit discharge[VLV-008A (B)1

2 A(D)-emergency feedwater pump RAW/LPSDactuation valves[EFS-MOV-103A(D) ]

3 B,C-Emergency feedwater pumps RAW/CCF/LPSD[EFS-RPP-001B (C)]

4 A,D-Emergency feedwater pumps FV/RAWICCF/LPSD[EFS-RPP-001A (D)]

5 Feedwater line check valves RAW/CCF/LPSD[VLV-018A (B,C,D)]

6 EFW pump discharge line check RAW/CCF/LPSDvalves[VLV-012A (B,C,D)]

7 Minimum/Full flow line check valves RAW/LPSD[VLV-020A (B,C,D)][VLV-022A (B,C,D)]

8 Minimum/Full flow line manual valves RAW/LPSD[VLV-021A (B,C,D)][VLV-023A (B,C,D)]

9 A-D-emergency feedwater control RAW/LPSDvalves[EFS-MOV-017A (B,C,D))

1U UeIYUlHlU IUy UUWdICwl r sy5LeiI k-r-vvo) suppuuii

feedwater to the steam generators in order to removereactor decay heat and RCS residual. This system isrequired after all initiating events exceeding large andmedium LOCA. Accordingly, reliability of EFW systemhas significant impact on risk.Two trains share one emergency feedwater pit, whichhas 50% capacity to perform cold shutdown. Large leakfrom SSCs or failure that result in degradation of watersupply from EFW pit will lead to lack of EFW. In this casemanual action to supply feedwater from SecondaryDemineralizer Water Tank is required.SSCs that have either of the following characteristics arerisk significant.- SSCs that have potential to cause common cause

failures among multiple trains. Common causefailure of such system will result in loss of multipletrains.

- SSCs that have potential to cause large leak orfailure that result in degradation of water supply fromEFW pit will lead are risk important. If such failureoccurs, manual action to supply feedwater fromsecondary demineralizer water tank will be required.

-.1

mc

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CO

CO)

0

0I

0

00

0

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10 A-D-emergency feedwater isolationvalves[EFS-MOV-019A (B,C,D)]

RAW



11 A-D-emergency feedwaterorifices[FE3716,3726,3736,3746]

line RAW (FLOOD) The "Insights and Assumptions" for these SSCs aredescribed on the previous page.

12 A-D-emergency feedwater line tie- RAW/CCF(FLOOD)line valves [EFS-MOV-014A (B,C,D)]_

13 EFW pit discharge line piping RAW/LPSD[TBDI

14 EFW pit discharge line tie-line piping RAW(FLOOD)[TBD]

15 A-D-emergency feedwater line RAW(FLOOD)A(B,C,D) piping[TBD]

16 T/D pump steam supply line piping RAW/LPSD[TBD]

17 Minimum/Full flow line piping RAW/LPSD[TBD]

18 A,B-Emergency feedwater pits RAW/LPSD[EFS-RPT-001A(B)]

19 Minimum/Full flow line manual valves RAW/LPSD[VLV-026A (B)]

20 EFW pump suction line manual RAW/LPSDvalves[VLV-009A (B,C,D)]

_c

-<C,

X DmcE- >

CU--I

co

c

--U

CACn-C z

00

z m

000

0D

0-.

21 EFW pump discharge line manualvalves[VLV-013A (B,C,D)]

RAW/LPSD

-I

"4


# Systems, Structures and RationalendsComponents (SSCs) Insights and Assumptions

22 EFW pit discharge line manual RAW/LPSD The "Insights and Assumptions" for these SSCs arevalves described on the previous page.[VLV-007A (B)]

23 Secondary demineralizer water tank RAW/LPSDdischarge line manual valves[VLV-006A (B)]

24 Secondary demineralizer water tank RAW(FLOOD)discharge line check valve[VLV-005]

25 EFW pit water level transmitter 1(2,3,4)[EFS-LT-3760, 3761, 3770, 3771]

m

F

COCO,C

zm

-JL"4

joC

C-

CD,

C

zm

z

C

Co

0

0

0

00

C)C

CD'9,

0

0) F- ~ .- 0 .

:-00C1 C C DS 00.:bg

0 (A


# Systems, Structures and Rationale~') Insights and AssumptionsComponents (SSCs) ________________o

'7 Emergency power source (EPS)

1 480V AC motor control center (MCC) RAW/LPSD The EPS consists of four separate trains. Each safetybuses train consists of one 6.9kV AC medium voltage bus andF[D EPS_-4ESBA(,C _D.)] 480V AC low voltage buses (Load Centers, Motor

2 480V AC load center buses Control Centers). Each AC medium voltage bus connects[-T-BD EPS-4LCA(B,CD)] to class 1E gas turbine generator. This system supports

3 6.9kV buses RAW/EJ/LPSD various safety mitigation systems and therefore, reliability[T•D EPS-6ESBA(B.C.D)] of the EPS system has significant impact on risk.

4 125V DC buses train A and D RAW/LPSD[_[T9B EPS-_EPS-DCAD)1 Since the EPS consists of four separate trains, single

5 125V DC buses train B and C RAW(L2) failure in trains not significantly impact risk. However,[TBD EPS- EPS-DCBOC)11 failure of multiple trains is have significant impact on risk.

6 120V buses train A-D RAW(L2/ FIRE) Accordingly, SSCs that have potential to cause common[-[T- EPS-VITA(B.C.D)] cause failures among multiple trains are risk significant

7 Swing MCC incomer circuit breakers RAW/CCF/LPSD[T-D EPS-4SB D)LD]

8 Batteries RAW/CCF/LPSD[T-BED EPS-BA1A(BC.D)]

9 6.9kV AC bus incomer circuit FV/RAW/CCF/LPSDbreakers

__[TBD EPS-6HA(B.CD)10 Gas turbine discharge circuit RAW/CCF/LPSD

breakers FV/CCF(FIRE)[T-99 EPS- GTBA(B,CD)]

-L,,4

mc

>I

COCD,

0>m Z

CC,,

00

00-

-0CD 0. 55 @ x

(D CD 0 0CD : ::

0 CL =



11 Circuit breakers between 6.9kV bus RAW/CCF/LPSD The "Insights and Assumptions" for these SSCs areand 6.9kV/480V safety power described on the previous page.transformers[Ti-D EPS- 41A(B.CD)I]

12 MCC bus incomer circuit breakers RAW/CCF/LPSD[-T-B.D EPS- 4JA(BC,D)]

13 Circuit breakers between 125V DC RAW/CCF/LPSDbus and Inverter[T-B9 EPS- VIT4A (B.CD)]

14 Class 1 E gas turbine generators FV/RAW/CCF[TBD EPS- GTA B _DjI /LPSD)

15 Gas turbines generator sequencers RAW/CCF/LPSD[TBD] FV(FIRE)

16 Inverters RAW/CCF/LPSD[T-RD EPS- INVA(B.C.D)]

17 Main transformers RAW(L2)[T-BD EPS- MTF]

18 6.9kV/480V safety power RAW/LPSDtransformers[TED EPS- 4PTA(B.C, D)]

;;0ma)F

C,C;U

z0m

C

I-

C-)

z0m

z

C

-,u

0

0

0

CD

:3O4

'D CD _ L

CD CD~ 0 0FCDl

0 0) C(D


Systems, Structures andComponents (SSCs) Rationale(

1) Insights and Assumptions

Non-class 1 E gas turbine generators[TBD EPS- GTPI(2A] /LPSD

2 480V permanent buses RAW(L2)[TBD]

3 6.9kV permanent buses RAW(L2)[TBD]

4 Circuit breakers between 6.9kV bus RAW(L2)and 6.9kV/480V power transformer[T-Bg EPS- 4 1P1(2)]

5 Batteries RAW/CFF/LPSD[TBD]

6 Gas turbine generator discharge RAW/CCF/LPSDcircuit breakers[T-9D EPS- GTBPI (2?A

7 AAC selector circuit breakers RAW/CCF/LPSD[-TED EBP.S- 4_A.A(8C D)]________

8 Circuit breakers between 125V DC RAW/CCF/LPSDbus and InverterF[D EPS:1 VT4EP1(2)]_

9 Inverters RAW/CCF/LPSD[TBD EPS- INVP1(2)]

10 Gas turbine generator sequencers RAW/CCF/LPSD

Two non-safety buses called "Permanent bus", which isconnected to Alternative AC (AAC), which consists ofnon-class 1E gas turbine generators respectively. Eachnon-class 1E gas turbine generators is manuallyconnected to two safety medium voltage buses viaselector circuit under the occurrence of loss of safety ACpower. The AAC is a countermeasure against stationblackout events.

SSCs that have potential to cause failures that degradethe availability to supply AAC power to safety mediumvoltage are risk significant.

Systems for the mitigation of core damage accident areconnected to permanent bus.

-UJO

mcF->

F:-4>i0> CD

C z

m Z

C

CD

0

(1)

00

0

0

CD0-

11 6.9kV/480V power transformers[TE-D EPS-4PTP1(2)]

RAW/LPSD

CO 0. <05 .0 0

C CD -a :IE~CM.-L'.0 ~

:3.2c Lin


Systems, Structures andComponents (SSCs) Rationale Insights and Assumptionsm Cr->

9 Non-essential chilled water system01-1 Fr-<

1 Non-essential chilled water system - RAW/LPSD In the case of loss of component cooling water events, =i>-< CCCWS boundary motor operated non-essential chilled water system or fire suppression Cn

valves system provides alternative component cooling water to caCo5;[TBD] charging pumps in order maintain RCP seal water C

injection. ;UThese SSCs are risk significant because large external

mleak from these valves result in loss of alternativecomponent cooling water from both non-essential chilledwater system and fire suppression system. On the otherhand, failure of other SSCs of this system affects only",_ the non-essential chilled water system itself.

CCn

0

0

00


# Systems, Structures and Rationale~ 1 ) Insights and Assumptions

Components (SSCs) Ratinale(_) InsightsandAssumptions

0 . Fire suppression systems (FSS)

1 FSS pump discharge motor operatedvalve[TBD]

FV(L2)/RAW(L2)

2 FSS pump discharge flow meter RAW(L2)[TBD]

3 Reactor cavity injection line orifice RAW(L2)[TBD]

4 FSS piping (from tank to tie line RAW(L2)piping)[TBD]

5 Raw water tank RAW(L2)[TBD]

In the case of core damage accident, Fire SuppressionSystems (FSS) injects water from Raw Water Tank intothe reactor cavity via the direct injection line by the firewater pumps.

The containment spray system and/or safety injectionsystem perform the reactor cavity flooding through thedrain line at loop compartment to prevent core-concreteinteraction when the reactor vessel is failed. The Firesuppression system performs as alternative function forthe reactor cavity flooding.

In the case of loss of component cooling water events,fire suppression system or non-essential chilled watersystem provides alternative component cooling water tocharging pumps in order maintain RCP seal waterinjection.

Large external leak from these valves result in loss ofalternative component cooling water from both non-essential chilled water system and fire suppressionsystem. On the other hand, external leak from otherSSCs degrade the fire suppression system but the non-essential chilled water system is still available foralternative component cooling. Therefore these valvesare risk significant SSCs in preventing core damage.

M .0

m CF>

>1

-CA)>Ci

Ci)ca~;U z>0m

0 >m Z

a

CiMU

0

00

CD

6 FSS pump discharge manual valve[TBD]

RAW(L2)

7 FSS - CCWSoperated valves[TBD]

Boundary motor RAW/LPSD



11 High head safety injection system

1 Safety injection pump discharge FV/RAW/CCF/LPSD In the case of LOCA, high head safety injection systemcheck valves injects coolant from refueling water storage pit (RWSP)[VLV-004A (B,C,D)] into the reactor vessel via the Direct Vessel Injection

2 Safety injection pump outlet orifices RAW(FLOOD) (DVI) line by the safety injection pumps. This system is1A(B,C,D) also essential for bleed and feed operation.[FE962(963,964,965)]

3 Minimum flow line orifices 3 ALC, RAW(FLOOD) Since this system consists of four independent trains,(D) failure of one train does not have significant impact on[FE972(973,974,975)] risk. However, failures of SSCs that impact multiple

4 Containment isolation check valves RAW/CCF/LPSD trains are risk significant.[VLV-01OA (B,C,D)]

5 Containment isolation motor RAW(FLOOD) SSCs that have either of the following characteristics areoperated valves FV(FLOOD) risk significant.[MOV-01 1 A(B,C,D)] - SSCs that have potential to cause common cause

6 RV injection line orifices RAW(FLOOD) failures among multiple trains. Common cause(between VLV-012 A(B,C,D) and failure of such system will result in loss of multiple

MOV-001 1 A(B,C,D)) trains.7 Injection line secondary isolation RAW/CCF/LPSD - SSCs that have potential to cause loss of RWSP

check valves inventory out side the containment due to large[VLV-012A (B,C,D)] external leaks. Loss of RWSP inventory impacts not

8 Injection line boundary check valves RAW/CCF/LPSD only all four trains of high head safety injection[VLV-013A (B,C,D)] system but also other systems that use RWSP as

9 A-'D-Safety injection pumps FV/RAWICCF/LPSD water source.

[SIS-RPP-001A (B,C,D)] I10 Containment isolation motor RAW

operated valves FV(FLOOD)[MOV-009A (B,C,D)]

-A

74mC

=i-C,,

>Cn

0>

m Z

Q

I0

--U

0

0

000

CD

CD 0 0.0 CL

X" "

-I

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11 GontainIe•t. I--sla4-----tQ- RAW/LPSD The "Insights and Assumptions" for these SSCs areojaerate4-valves FV(FLOOD) described on the previous page.Safety injection pump suctionIsolation valves[MOV-001A(BCD)]

12 Piping RAW/LPSD13 Minimum flow line orifices RAW(FLOOD)

(next to VLV-L023 A(B,C,D))14 Minimum flow line manual valves RAW(FLOOD)

[VLV-024 A(B,C,D)]15 Minimum flow line manual valves RAW(FLOOD)

[VLV-023 A(B,C,D)]16 A(BC,D)-Hot leg recirculation line RAW(FLOOD)

isolation valves[MOV-014 A(BCD)]

m CE: >

->-CO

> C

zmZ

:3CD 0CD (DC

CD 0

0

1>0

CD 0 (n0- M --m-•<ci, o.

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CDU3

0

0O

e-

30a)

0D

0



12j Heating, ventilation, and air conditioning (HVAC) system

B,C-Emergency feedwater pump RAW/CCF/LPSD EFW M/D pump room fans maintain room temperatureroom fans FV(FLOOD) when pumps are running. EFW M/D pumps are assumed[VRS-RFN-401 B,C] to be unavailable within the mission time without room

cooling due to high room temperature.

HVAC systems of other rooms are considered not to berisk significant for the following reasons.- HVAC of emergency gas turbine room

Gas turbine units itself has function to intake outerair to remove heat out to atmosphere. Accordingly,HVAC is considered not essential to maintain gasturbine function.

- HVAC of ESF room (RHR/CSS pump, SI pump)According to room temperature analysis, roomtemperature will not exceeds limit of the systemduring the mission time regardless of availability ofHVAC.

- HVAC of classl E electric power room (Class 1 EI&C, switch gear, battery, battery charger)This system is running during normal operation andcontinues to run after initiating events. Reliability ofnormally operating HVAC systems are considered tobe high and failure of this system is unlikely to occurduring the mission time.

- HVAC of EFW T/D pump roomSince T/D driven EFW pump room can operateunder high room temperature conditions, they areassumed to be available regardless of room coolingdurinq the mission time.

mcr-->

->

-C,,

C,,Ci,>

O

>0

0 >m Z

CO

CD

0

0

0

0


Systems, Structures andComponents (SSCs) Rationale(1) Insights and Assumptions

Containment fan cooler[VCS-CAH-001A (B,C,D)]

I ttlillJUldtUlV ;UIILIUI U1 k.oUrlL•IIIIITIU1I V•bb5 l d1aTmoUbpnere

is judged important by experts from a point of view ofkeeping function of safety components in Containment

Main control room air handling unit[VRS-RAH-101A (B,C,D)]

I U3111}Jr:ldLUlU U.UIIIIUI U1 II1dlll t.;UIILIUI IUUIII idtlllUb5[Jll:l 1b5

judged important by experts from the viewpoint of.operator habitability during an accident.

-'I

-4>mCE: >

F;

>0

m.Ci

CD

-o

zm

0>

0

C

0

CD

0


Systems, Structures and Rationale~1 ) Insights and AssumptionsComponents (SSCs) Rta_)IihsnAsmio

15 lnstrumentation and control (I&C) system

1 Permanent bus low voltage signal RAW/CCF This software provides start signal to non-class 1E gassoftware turbine generator. Under SBO, This software must

operate in order to backup of the safety bus by AACpower source. ..........

2 Component cooling water system RAW/CCF SSCs that have potential to cause common cause failure

train isolation signal software of signals are risk significant since such failure may3 SG isolation signal software RAW/CCF result in loss of total system function.4 Engineered safety features actuation RAW/CCF

signal software (P,S) EFW T/D pump start signals are risk significant since5 SG(EFW) isolation signals RAW/CCF such failure results in loss of one of two available EFW6 Main steam line isolation signal RAW/CCF pumps under, SBO and loss of EFW room cooling

software conditions.7 Black out signal software RAW/CCF8 CCW start signals RAW(L2,FLOOD) Reliability of signals other than "S signal" is assumed to9 Containment pressure sensors RAW(L2)/CCF(L2) have same reliability with "P signal".

[TBD]10 A-D-Emergency feed water pump RAW

start signals11 EFW pump start signal software RAW/CCF12 Diverse actuation system EJ The unreliability of this system is assumed to be 0.01.

m C

-> CF-<C

C,,CO)

>mZ

CnCO)

C >0

z m

0>

0

C)Cn


Systems, Structures and Rationale~') Insights and AssumptionsComponents (SSCs)

16 Waste management system, (.WMS)1 Refueling water storage (RWS) RAW Large External leak of the boundary check valve results

system - WMS line boundary check in loss of inventory from the RWS system. Systems thatvalve relies on the RWS as water source is affected by this[VLV-037] failure mode.

17 Main feedwater system (MFWS)

1 Main feedwater system RAW The Main feedwater system is credited as a function tosecondary side cooling during general transients, which

Tdoes not involve loss of main feedwater.

m C

=1>

Cn~

0>

C,,

C >

>0z m

0>

0

C,

CD

)6I

Table 17.4-1 Risk significant SSCs (sheet 24 of 34)-I

-1

-'4

(01

to

5.

Systems, Structures andComponents (SSCs) Rationale(i) Insights and Assumptions

A-D-Main steam isolation valves[NMS-AOV-515A (B,C,D)1 FV/CCF(FIRE)

2 A-D-Main steam bypass isolation RAW(L2)valves[NMS-HCV-3615,3625,3635,3645]

3 Main steam line piping RAW4 Main steam line isolation check RAW(FIRE)

valve s A(B,C and D)[VLV-516A(B,C and D)]

5 Al -A2-Main steam safety valves RAW(L2)Bl-B2-Main steam safety valvesC1 -C2-Main steam safety valvesD1-D2-Main steam safety valves[NMS-VLV-509A (B,C,D)][NMS-VLV-510A (B,C,D) ]

IVidlil .tfiedIll IbUlIdlUll VdIVU I1UIdLb Lil: IUP i UIUU OL.dIf 1

Generator (SG) at the Steam Generator Tube Rupture(SGTR). In case of secondary line break, main steamisolation is required to prevent unlimited steam release.Main steam line piping is required to be intact to isolatethe ruptured SG at SGTR events.

-4

m CEo >

=-<

0>

mZCz

c ,

CII

0

0

0

,,,>

00

0>

6 A,B,C,D,E,F,G,H,J,K,L, M,N,P,Q-Turbine bypass valves[NMS-TCV-500A(B,C,D,E,F,G,H,J,K,L,M,N,P.Q) 1

RAW(L2)


Systems, Structures andComponents (SSCs) Rationale(') Insights and Assumptions

J-O)--Od1ULy U1valves[RCS-MOV-1 17A(B)]

FV/CCF(FLOOD,FIRE)oarety uepressurizauon vaives kouxiopen during bleed and feed operation.

2 [A(B) -Safety depressurization valves RAW(FLOOD,FIRE)I [RCS-MOV-1 16 A(B) I

Pressurizer safety valves releases RCS pressure incase of high RCS pressure. Failure of safety valves tore-close results in loss of primary coolant.3 A-D-Pressurizer safety valves

[RCS-VLV-1 20][RCS-VLV-1 21][RCS-VLV-1 22]I'I•.q-\/I \/-19*i1

RAW

MJD

m C

->-C,,

C4mZ-

CD

Ci

C Z

0,,-.

00C

:3

MD

Depressurization valves[RCS-MOV-1 181[RCS-MOV-1 191

In the case of core damage accident, depressurizationof the reactor coolant system is required to prevent highpressure melt ejection and direct containment heating.

-ICD

74

CD

<.


# Systems, Structures and Rationale~1 ) Insights and AssumptionsComponents (SSCs) Ratinale___ InsightsandAssumptions

21 Containment spray / residual heat removal (CS/RHR) system

1 Heat exchanger bypass valves RAW/LPSD The Containment Spray / Residual Heat Removal[FCV-604] (CS/RHR) System consists of four independent trains.[FCV-636] The CS/RHR System has the following four functions.

2 RHR line heat exchanger discharge RAW/LPSD a. Containment Sprayair operated valves b. Alternative Core Cooling[FCV-603] c. RHR Operation during operating modes 4 , 5 and 6..[FCV-633]

3 Pump suction line check valves RAW/CCF/LPSD Since CS/RHR system consists of four independent[VLV-004A (B,C,D)] trains, failure of one train does not have significant

4 RHR line containment isolation check RAW/CCF/LPSD impact on risk. However, failures of SSCs that impactvalves multiple trains are risk significant.[VLV-022A (B,C,D)]

5 RHR line containment isolation motor RAW/CCF/LPSD SSCs that have either of the following characteristics areoperated valves risk significant.[MOV-021A (B,C,D)] SSCs that have potential to cause common cause

6 A-D-Containment spray/residual RAW/CCF/LPSD failures among multiple trains. Common causeheat removal pumps FV(FLOOD) failure of such system will result in loss of multiple[RHS-RPP-001A (B,C,D)] trains.

7 A~D-Containment spray/residual RAW/CCF/LPSD - SSCs that have potential to cause loss of RWSP

heat removal heat exchangers inventory out side the containment due to large

[RHS-RHX-001A (B,C,D)] external leaks. Loss of RWSP inventory impacts not

8 RHR line boundary check valves RAW/LPSD only all four trains of CS/RHR system but also other

[VLV-028A (B,C,D)] systems that use RWSP as water source.

9 RWSP discharge line isolation valves RAW[-T.BD-CS S-MOV-001A(BC,D)]

-4

r"

mCD

->

cn

i,>

<CD

CD

>C)

0

e-

S0

0

CD

,a.C03 X jCM- 0 D0 00

CD0 C


# Systems, Structures and Rationale~l) Insights and AssumptionsComponents (SSCs) R

10 CS line containment isolation motoroperated valves[MOV-004A (B,C,D)]

RAWFV(FLOOD)

The "Insights and Assumptions" for these SSCs aredescribed on the previous page.

11 A-D-CS line check valves RAW/CCF(FLOOD)[VLV-005A(B,C,D)]

12 Piping RAW[TBD]

13 CS line heat exchanger discharge RAWmanual valves[VLV-002A (B,C,D)]

14 Minimum flow line manual valves RAW[VLV-13A (BC,D)]

15 CS/RHR - spent fuel pit boundarymanual valves (discharge line)rVLV-031A (D))

RAW

m C

a) -

F-<Cl

CO>0C0>

c

CA)

>0

Co

0

U)

0

0

0

(D*

- h - +... . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .

16 From FSS to CSS tievalve[VLV-0 121

line check RAW(L2) These valves are required to open toinjection from FSS to the spray header.

perform firewater

.4-17 From FSS to

operated valve[CSS-MOV-011]

CSS tie line motor FV(L2)/RAW(L2)

18 CS/RHR - spent fuel pit boundarymanual valves (suction line)[VLV-034A (D)]

RAW/LPSD These valves are required to open to performgravitational injection from the spent fuel pit to the RCSwhen RCS is atmospheric pressure at LPSD operation.

19 CS/RHR - spent fuel pit boundarymanual valves (suction line)IVLV-33A(D)1

LPSD


# Systems, Structures and Rationale~l Insights and AssumptionsComponents (SSCs)

20 CS/RHR pump hot leg suction LPSD Failure of these valves result in loss of RHR during LPSDisolation valves[MOV-001A(B,C,D)][MOV-002A(B,C,D)]

21 RCS cold leg injection line motor LPSDoperated valves [MOV-026A(B,C,D)]

22 RCS cold leg injection line check LPSDvalves[VLV-027A(B,C,D)][VLV-028A(B,CD)]

-L"4

mC

U-

4>

-ZSO)

SO)

ci

>C

C,

CO

0

0>m Z

-I0

0)

00

Table 17.4-1 Risk significant SSCs (sheet 29 of 34)Systems, Structures and Rationale~1 ) Insights and Assumptions

Components (SSCs)

22 • 1 Refueling water storage system (RWS)

1 Refueling water storage pit (RWSP)sump strainers[TBD]

FV/RAW/CCF

-LC4

CA

5.M

0o

2 Refueling water storage pit RAW[RWS-CPT-001]

3 Refueling water recirculation pump RAW/LPSDsuction line manual valves[VLV-006A (B)]

4 Refueling water recirculation pump RAW/LPSDdischarge line check valves[VLV-012A (B)]

5 Refueling water recirculation pump RAW/LPSDdischarge line manual valves[VLV-01 3A (B)]

6 RWSP discharge line containment RAW/LPSDisolation motor operated valves[MOV-002][MOV-004]

7 A,B-Refueling water recirculation RAW/LPSDpumps[RWS-RPP-001A (B)]

8 RWSP discharge line manual valve RAW/LPSD___[VLV-001]

The RWSP is the source of borated water forcontainment spray and safety injection. During LPSDoperation, RWSS has the following functions.a. Refill refueling water auxiliary tank (RWAT) for RCS

injection via charging pumps.b. Refill SFP for gravitational injection to RCS.

SSCs that have either of the following characteristics arerisk significant.- SSCs that have potential to cause common cause

failures among multiple trains. Sump strainers havepotential of sump screen, which may occur inmultiple trains.

- SSCs that have potential to cause resulting loss ofRWSP inventory out side the containment due tolarge external leaks are risk significant, since suchfailure impacts all systems that use RWSP as watersou rce.

SSCs that have potential to cause failure to supplyRWSP water to RWAT or SFP during LPSD operationare also considered risk significant.

7,4

m CE0 >

>-

-C,,

CO

CO

C z

0

0>

C,

Cn

('-I

09 Refueling water recirculation pumpsuction cross tie line manual valverVLV-0051

RAW/LPSD



10 Refueling water recirculation pump RAW/LPSD The "Insights and Assumptions" for these SSCs aredischarge cross tie line manual described on the previous page.valve[VLV-014]

11 Refueling water storage auxiliary tank LPSD[RWS-OTK-0021

4412 Refueling water auxiliary tank inlet RAW/LPSDline manual valve[VLV-052]

4t213 Refueling water auxiliary tank RAW/LPSDdischarge line manual.valve[VLV-101]

4-314 Refueling water auxiliary tank LPSDsuction line manual valves[VLV-021][VLV-051]

15 RWSAT line orifice LPSD[TBD: downstream side of VLV-0211

4416 RWSP suction line containment LPSDisolation air operated valve[AOV-022]

-L

-4

>-,1

0>

CD>C,

-u

0

cn

C z

0

-,,

0>0

0

-CD

CD

-4

i.,

0

5.



23 ,f Reactor protection system (RPS)

1 Reactor trip breakers RAW/CCF These systems are necessary to provide negative[TBD] reactivity for plan t trip.

2 Control rod (rod injection) FV/RAW/CCF[TBD]

'24 Chle ae sse VS

1 Chiller units train B and C FV/RAW/CCF/LPSD The safety related water system supplies chilled water to[T-BE.VWS-PEQ-001B(C)] safety related HVAC systems.

SSCs that have potential to cause common causefailures among trains B and C are risk significant since

2 Pumps train B and C RAW/CCF/LPSD such failures results in loss room cooling in M/D EWF[T4Q-B W- pump area.

SSCs that compose train A and D are not risk significant

because the PRA assumes only the M/D EFW pumps tobe dependent on room cooling during the mission time.

-h

mc-4

mZ>

Cn

S0

z m

0>

0

0

CD

a

X

0

0

x oI.-0 (A CD .0

.• ."4 DM•_


Systems, Structures and Rationale~1 ) Insights and AssumptionsComponents (SSCs)

25 Essential service water system (ESWS)

1 Pump discharge line check valves RAW/CCF/LPSD The essential service water system (ESWS) transfers[VLV-502A (B,C,D)] heat from the CCW system as Ultimate Heat Sink (UHS).

2 Essential service water pump motor RAW/CCF/LPSD This system supports the CCW system, which supportscooling line check valves various safety and non-safety mitigation systems.[VLV-602A (B,C,D)] Accordingly, reliability of CCWS EFW system has

3 A-D-Essential service water pump FV/RAW/CCF/LPSD significant impact on risk.[EWS-OPP-001A (B,C,D)] Si

4 Al-E-1-Essential-serviGe_-water-pump RAW/LPSD Since ESWS consists of four independent trains, failureoutlet Strainers of one train does not have significant impact on risk.[FWS OSR 00A (9)] However, failures of SSCs that impact multiple trainsA (BC,D) -CCW heat exchanger inlet have risk significant impact on risk. Accordingly, SSCsstrainers[TBD] that have potential to cause common cause failures

5 Al-D1-Essential service water pump RAW/LPSD among multiple trains are risk significant.

outlet strainersA2-QD2 Essential SerVice water pump

[EWS-OSR-001A (B,C,D)][EWS QSR 002A (ICnD)4

6 Valves located in essential service RAW/LPSDwater pump motor cooling line oftrain B & C[VLV-601B (C)]

7 ESW pump motor cooling line valves RAW(L2)of train A & D[VLV-601A (D)]

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Systems, Structures andComponents (SSCs) Rationale Insights and Assumptions

8 4fesleeated-%R cscential ScieFwatef ESW pump motor cooling linetransmitters of train A.B & LC and D[FT-2060.2061.2062 and 20631

The "Insights and Assumptions" for these SSCs aredescribed on the previous page.

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9 ESW pump motor cooling line orifices RAW/LPSDof train AXB.C and D [TBD]

910 Main piping orifices of train B and RAW/LPSDDC[FE2025, FE2026]

4-01_1 Main piping orifices of train A and D RAW(L2)[FE2024, FE2027]

4412 Main piping valves of train B and C RAW/LPSD[MOV-503B (C)][VLV-506B (C)][VLV-507B (C)][VLV508B (C)][VLV-509B (C)][VLV-511B (C)][VLV-514B (C)][VLV-517B (C)][VLV-520B (C)]

1-213 Main piping valves of train A and D RAW(L2)[MOV-503A (D)][VLV506A (D)][VLV-507A (D)][VLV508A (D)][VLV-509A (D)][VLV-51 1A (D)][VLV-514A (D)][VLV-517A (D)][VLV-520A (D)]

4-314 Piping of train B and C [TBD] RAW/LPSD

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4-415 Piping of train A and D [TBD] RAW(L2)

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26 pit cooling and purification system (SFPCS)

1 RWS - SFP inlet line boundary RAW/LPSD Large External leak of valves that form boundarycheck valves between RWS result in loss of inventory of the RWS[VLV-027] system. Accordingly, systems that relies on the RWS as

2 RWS - SFP inlet line manual valve RAW/LPSD water source is affected by failure of these valves.[VLV-028]

3 RWS - SFP demineralizer line RAW During RCS is atmospheric pressure at LPSD operation,boundary manual valves the spent fuel pit is used as water source of gravitational[VLV-103A (B)] injection in case loss of decay heat removal function

4 RWS - SFP inlet line manual valves LPSD occurs. SSCs associated with gravitational injection line[VLV-029] are considered to be risk significant.[VLV-015][VLV-017]

5 Spent fuel pit LPSD[RPT-001]

6 A-D-Spent fuel pit strainers LPSD[SFS-RSR-001A (B,C,D)]

7 Spent fuel pit discharge line manual LPSDvalves[VLV-021A(D)]

8 Spent fuel pit discharge cross tie-line LPSDmanual valve

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Table 17.4-1 Risk significant SSCs (sheet 35 of 34)--I

-40.

Notes:1. Definition of Rationale Terms:

CCF = Common Cause FailureFV = Fussell-VeselyRAW = Risk Achievement WorthFV(L2) = Fussell-Vesely for L2RAW(L2) = Risk Achievement Worth for L2

CCF(L2) = Common Cause Failure for L2LPSD =Low Power and Shut Down OperationEJ = Engineering JudgeFLOOD = FLOOD EventFIRE = FIRE EventEP = Expert Panel

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17.4.8 ITAAC for the D-RAP

Tier 1 ITAAC are proposed to verify that the D-RAP provides reasonable assurance thatthe design of SSCs within the scope of the RAP is consistent with their assumed designreliability. The list of risk-significant SSCs for ITAAC will be prepared by introducing theplant's site-specific information to the list shown in Table 17.4-1 in the Phase II of the D-RAP. The ITAAC acceptance criteria are established to ensure that the estimatedreliability of each as-built SSC is at least equal to the assumed design reliability and thatindustry experience including operations, maintenance, and monitoring activities wereassessed in estimating the reliability of these SSCs.

17.4.9 Combined License Information

COL 17.4(1)

COL 17.4(2)

The COL Applicant shall be responsible for the development andimplementation of the Phases II and Ill of the D-RAP, including QA SPLA 1676-036requirements. In the Phase II, the plant's site-specific informationI

should be introduced to the D-RAP process and the site-specific risk- I SPLA 1474-0117significant SSCs should be combined with the US-APWR design risk-significant SSCs into a list for the specific plant. Phase II is performed SPLA 1676-0361during the COL application phase and updated/maintained during theCOL license holder phase. In the Phase Ill, procurement, fabrication,construction, and test specifications for the SSCs within the scope ofthe RAP should ensure that significant assumptions, such asequipment reliability, are realistic and achievable. The QArequirements should be implemented during the procurement,fabrication, construction, and pre-operation testing of the SSCs withinthe scope of the RAP. Phase Ill is performed during the COL license I SPLA 1676-036holder phase and prior to initial fuel loading. The COL applicant willpropose a method by which it will incorporate the objectives of thereliability assurance program into other programs for design oroperational errors that degrade nonsafety-related, risk-significantSSCs.

The COL Applicant shall be responsible for the development andimplementation of the O-RAP, in which the RAP activities should beintegrated into the existing operational program (i.e., MaintenanceRule, surveillance testing, in-service inspection, in-service testing, andQA). The O-RAP should also include the process for providingcorrective actions for design and operational errors that degrade non-safety-related SSCs within the scope of the RAP. A description of the SPLA 1676-036proposed method for developing/integrating the operational RAP intooperating plant programs (e.g., maintenance rule, guality assurance)is performed during the COL application phase. Thedevelopment/integration of the operational RAP is performed durinqthe COL license holder phase and prior to initial fuel loading. All SSCs SPLA 1474-010identified as risk-significant within the scope of the D-RAP should becategorized as high-safety-significant (HSS) within the scope of initialMaintenance Rule.

Tier 2 17.4-41

Revision I



17.4.10 References

17.4-1 "Policy and Technical Issues Associated with the Regulatory Treatment ofNon-Safety Systems (RTNSS) in Passive Plant Design," SECY 95-132, U.S.Nuclear Regulatory Commission, Washington, DC, May 1995.

17.4.2 "Quality Assurance Program (QAP) Descripon For Design Certification of [SPLA 1676-039

the US-APWR (PQD-HD-19005 Rev.2)" 1

17.4-23 'Requirements for Monitoring the Effectiveness of Maintenance at NuclearLSPL 1676-039i

Power Plants,' "Domestic Licensing of Production and Utilization Facilities,"Ener.gy. Title 10, Code of Federal Regulations, Part 50.65, U.S. NuclearRegulatory Commission, Washington, DC.

Tier 2 17.4-42

Revision I



17.5 Quality Assurance Program Description

For the Design Certification phase, the MHI-NESH US-APWR Project Quality AssuranceProgram (QAP) is the top-level policy document that establishes the quality assurancepolicy and assigns major functional responsibilities for plants designed by MHI-NESH.The QAP describes the methods and establishes QAP and administrative controlrequirements, described in "Quality Assurance Program (QAP) Description For DesignCertification of the US-APWR (PQD-HD-19005 Rev.f2)" (Ref 17.5.5-4), that meet 10 [SPLA 1676-039

CFR Part 50, Appendix B and 10 CFR Part 52. The QAP is based on the requirementsof ASME NQA-1-1994, "Quality Assurance Requirements for Nuclear FacilityApplications," Parts I and II, as specified in Ref.17.5.5-4.

The MHI QAPD for the Design Certification Phase has been prepared on the basis of theNRC approved QAP template (NEI, 06-14A Rev.4 and earlier revisions) (Ref 17.5.5-3)prepared by the Nuclear Energy Institute and has been evaluated against the SRP. TheMHI QAPD provides the QAP controls implemented. MHI performed the comparison ofSRP (Mar. 2007) (Ref 17.5.5-2) and draft SRP (Sept. 2006) (Ref 17.5.5-1) which wasused as a reference for the MHI QAPD and determined that there is no impact to theMHI QAPD.

Business policies of MHI-NESH establish high level responsibilities and authority forcarrying out administrative functions which are outside the scope of the QAP.

Procedures establish practices for certain activities which are common to all MHI-NESHorganizations performing those activities such that the activity is controlled and carriedout in a manner that meets QAP requirements. Organization specific proceduresestablish detailed implementation requirements and methods, and may be used toimplement the business policies of MHI-NESH or be unique to particular functions orwork activities.

The COL applicant is responsible for development a Quality Assurance Program

Description during design other than the Design Certification, construction and operation.


COL 17.5(1) The COL applicant shall develop and implement the design other thanthe Design Certification, construction and operational QAP that alsocovers the activities described in Section 17.5.

17.5.2 References

17.5.5-1 "Draft Standard Review Plan (SRP) 17.5 dated September 22, 2006"

17.5.5-2 "Standard Review Plan (SRP) 17.5 March 2007"

17.5.5-3 "Quality Assurance Program Description (NEI 06-14A Rev.4 and earlierversions)"

17.5.5-4 "Quality Assurance Program (QAP) Description For Design Certification ofthe US-APWR (PQD-HD-19005 Rev.-2)" I SPLA 1676-039

Tier2 175-1ReviionI




17.6 Description of the Applicant's Program for Implementation of 10 CFR 50.65,the Maintenance Rule

The COL Applicant is responsible for development of the program for implementation of10 CFR 50.65, the Maintenance Rule.


COL 17.6(1) The COL applicant develops and implements the program forimplementation of 10 CFR 50.65, the Maintenance Rule.

Tier 2 17.6-1

Revision I


mhi's second responses to us-apwr dcd rai no. …art mitsubishi heavy industries, ltd. 16-5,...

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