NRC-94-4194, Forwards Responses to NRC Requests for Addl Info Re AP600
| ML20070E307 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 07/08/1994 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-94-4194 DCP-NRC0127, DCP-NRC127, NTD-NRC-94-4194, NUDOCS 9407140216 | |
| Download: ML20070E307 (131) | |
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I NTD-NRC-94-4194 DCP/NRC0127 :
I Docket No.: STN 52-003 July 8,1994 Document Control Desk ,
U.S. Nuclear Regulatory Commission Washington, D.C 20555 ;
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A'lTENTION: R.W.BORCHARDT (
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SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL !
INFORMATION ON THE AP600 I l
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Dear Mr. Borchardt:
Enclosed are three copic: of the ' Westinghouse resymses to NRC requests for additional information on the AP600 from your letters of April 15,1994, April 29,1994, May 2,1994, May 18,1994, May 23,1994 and May 24,1994. Ia addition, a revision of a response previously submitted is .
provided. ,
A listing of the NRC requests for additional infortnation respmded to in this letter is contained in Attachment A.
These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Kenyon's !
copy.
If you have any questions on this mateiial, please contact Mr. Brian A. McIntyre at 412-374-4334.
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Nicholas J. Liparulo, M ager
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Nuclear Safety Regulatory And Licensing Activities
/nja i Enclosure ec: B. A. McIntyre - Westinghouse I T. Kenyon - NRR
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NTD-NRC-94-4194 ATTACHMENT A '
AP600 RAI RESPONSES SUBMITTED JULY 8,1994 RAI No. Issue 100.012R01: Up-to-date P& ids 210.032 : SSAR Section 3.7.3.9 210.041 : SSAR Section 3.6.2.3.1 210.044 : SSAR section 3.6.2.4 & 3.6.2.4.2 ;
I 210.045 : SSAR section 3.6.2.4.2 210.049 : SSAR section 3.7.3.8.2.1 ;
210.051 : SSAR section 3.7.3.9 210.067 : Design / requirements of pressure-relieving devices 210.078 : WCAP-13054, SSAR Tables 3.9-5, 3.9-6, 3.9-7, 3.9-8 260.023 : Startup and/or preoperational testing 410.167 : Inclusion of drain tanks in Figure 9.3.5-1 410.184 : Fission product control 410.186 : CWS design description / interface requirements 410.134 : Sumps in SFS area ,
41 .2.!2 : Seismic category of sperit fuel facility 410.239 : HVAC site-specific interface requirements ;
410.252 : MSSS conformance with SRP 410.256 : Hydrogen buildup in main condenser ,
440.067 : RHR system operation during mid-loop operation 1 440.072 : Action items from GL 88-17 440.097 : IRWST spill flow <!uring DVI line break 480.053 : Cont pressure instrument line penetration RG 1.11 ;
480.054 : Type C testing of RHR suction isolation valves 480.065 : Potential for drain clogging from coatings I
480.073 : Closure time for containment isolation valves 720.158R02: PRA-based seismic margins method i r
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NTD-NRC-94-4194 ATTACIBfENT A '
AP600 RAI RESPONSES SUBMITTED JULY 8,1994 t
RAI No. Issue 920.002R01: Listing of vital eauioment l
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I NRC REQUEST FOR ADDITIONAL INFORMATION l
jE ' 4) l Response Revision 1 =
l Question 100.12 Several piping and instrumentation diagrams (P&lD) are not self-contained. For example, the passive core cooling f system P&lDs in Figures 6.3-1 through 6.3-4 of the SSAR contain many interfaces with other systems whose P&lDs !
are not provided. Provide a complete set of up-to-date P& ids for the AP600 design [
Response: (Revision 1) l The following P & ids are provided as an enclosure to letter NTD-NRC-94-4190 as D-size plots. P & ids for i other systems will be submitted at the same time as Revision 2 of the AP60J Standard Safety Analysis Report I (SSAR). l Document No. Be.v Document Description l
BDS M6 001 I P & ID STEAM GENERATOR BLOWDOWN '
CCS M6 001 6 COMPONENT COOUNG WATER SYSTEM P & ID j CCS M6 002 6 COMPONENT COOUNG WATER SYSTEM P & ID CCS M6 003 6 COMPONENT COOUNG WATER SYSTEM P & ID CCS M6 004 6 COMPONENT COOUNG WATER SYSTEM P & ID ,
CDS M6 001 G CONDENSATE SYSTEM P & ID CWS M6 001 F CIRCULATING WATER SYSTEM P & ID !
I CVS M6 001 7 CHEMICAL & VOLUME CONTROL SYSTEM P & ID .[
CVS M6 002 7 CHEMICAL & VOLUME CONTROL SYSTEM P & ID l
1 DOS M6 001 D STANDBY DIESEL AND AUXIUARY BOILER FUEL OIL SYSTEM P & ID DWS M6 001 E DEMINERAUZED WATER TRANSFER AND STORAGE SYSTEM P & ID FPS M6 001 3 FIRE PROTECTION SYSTEM P & ID i FPS M6 002 3 FIRE PROTECTION SYSTEM P & ID !
FPS M6 003 3 FIRE PROTEC110N SYSTEM P & ID !1 FPS M6 004 3 FIRE PROTECTION SYSTEM P & ID i FPS M6 005 3 FIRE PROTECTION SYSTEM P & ID i FPS M6 006 3 FIRE PROTECTION SYSTEM P & ID !
FWS M6 001 H FEEDWATER SYSTEM P & ID i MSS M6 001 G MAIN STEAM SYSTEM P & ID 100.12(R1)-1 l
NRC REQUEST FOR ADDITIONAL INFORMATION i@ "pji Response Revision 1
. e MSS M6 002 F MAIN STEAM SYSTEM EXTRACTION P & ID PCS M6 001 6 PASSIVE CONTAINMENT COOLING SYSTEM P & ID PSS M6 001 6 PRIMARY SAMPLING SYSTEM P & ID PXS M6 001 8 PASSIVE CORE COOUNG SYSTEM P & ID PXS M6 002 8 PASSIVE CORE COOUNG SYSTEM P & ID PXS M6 003 8 PASSIVE CORE COOUNG SYSTEM P & ID PXS M6 004 8 PASSIVE CORE COOUNG SYSTEM P & ID RCS M6 001 8 REACTOR COOLANT SYSTEM P & ID RCS M6 002 8 REACTOR COOLANT SYSTEM P & ID RCS M6 003 8 REACTOR COOLANT SYSTEM P & ID RNS M6 001 5 NORMAL RESIDUAL HEAT REMOVAL SYSTEM P & ID SFS M6 001 8 SPENT FUEL PIT COOUNG SY. TEM P & ID SGS M6 001 7 STEAM GENERATOR SYSTEM P & ID SGS M6 002 7 STEAM GENERATOR SYSTEM P & ID SWS M6 001 F SERVICE WATER SYSTEM P & ID VAS M6 001 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 002 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 003 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 004 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 005 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 006 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 007 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VAS M6 008 4 RADIOLOGICALLY CONTROLLED AREA HVAC P & ID VBS M6 001 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VBS M6 002 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VBS M6 003 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VBS M6 004 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VBS M6 005 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VBS M6 006 3 NI NON-RADIOACTIVE VENTILATION SYSTEM P & ID VES M6 001 5 MAIN CONTROL ROOM EMERGENCY HABITABIUTY P & ID VFS M6 001 4 CONTAINMENT AIR FILTRATION SYSTEM P & ID 100.12(R1)-2 3 WeStingh0USB
NRC REQUEST FOR ADDITIONAL INFORMATION e mJ Responso Revision 1 El
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VFS M6 002 4 CONTAINMENT AIR FILTRATION SYSTEM P & ID VHS M6 001 K HEALTH PHYSICS AND HOT M ACHINE SHOP HVAC SYSTEM P & ID VRS M6 001 K RADWASTE BUILDING HVAC SYSTEM P & ID VRS M6 002 B RADWASTE BUILDING HVAC SYSTEM P & ID VWS M6 001 4 CENTRAL CHILLED WATER SYSTEM P & ID VWS M6 002 4 CENTRAL CHILLED WATER SYSTEM P & ID VWS M6 003 4 CENTRAL CHILLED WATER SYSTEM P & ID VWS M6 004 4 CENTRAL CHILLED WATER SYSTEM P & ID VWS M6 005 4 CENTRAL CHILLED WATER SYSTEM P & ID VWS M6 006 4 CENTRAL CHILLED WATER SYSTEM P & ID
""T T? 4 CEN'm AL C"!LLED "'*JER SYSm' n & !D VXS M6 001 J ANNEX / AUXILIARY BUILDING NON-RADIOACTIVE HVAC SYSTEM (S) P & ID VXS M6 002 K ANNEX /AUXIUARY BUILDING NON-RADIOACTIVE HVAC SYSTEM (S) P & ID VXS M6 003 J ANNEX /AUXIUARY BUILDING NON-RADIOACTIVE HVAC SYSTEM (S) P & ID VZS M6 001 F DIESEL GENERATOR BUILDING VENTILATION SYSTEM P & ID WGS M6 001 3 GASEOUS RADWASTE SYSTEM P & ID WLS M6 001 3 UQUID RADWASTE SYSTEM P & ID WLS M6 002 3 UQUID RADWASTE SYSTEM P & ID WLS M6 003 3 UQUID RADWASTE SYSTEM P & ID WLS M6 004 3 UQUID RADWASTE SYSTEM P & iD WRS M6 001 2 RADIOACT1VE WASTE DRAIN SYSTEM P & ID WRS M6 002 2 RADIOACTIVE WASTE DRAIN SYSTEM P & ID WRS M6 003 2 RADIOACTIVE WASTE DRAIN SYSTEM P & ID WSS M6 001 F SPENT RESIN PROCESSING SYSTEM P & ID WSS M6 002 G SPENT RESIN PROCESSING SYSTEM P & ID ZOS M6 001 F ONSITE STANDBY POWER SOURCE SYSTEM DG UNIT A P & ID ZOS M6 002 F ONSITE STANDBY POWER SOURCE SYSTEM DG UNIT B P & ID SSAR Revision: NONE W westinghouse 1 .12(R193
NRC REQUEST FOR ADDITIONAL INFORMATION i
Question 210.32 j i
The response to Q210.12 dated November 30, 1992 references EPRI NP-6153, " Seismic Analysis of Multiply !
Supported Piping Systems," as the basis for combining the results of the modal spectra analysis and seismic anchor l motion by the square root sum of the squares (SRSS) method. The staff has not endorsed EPRI NP-6153 and does j not agree that this report provides an adequate technical basis for using the SRSS method. The staff's position i remains as stated in Q210.12, i.e., the responses due to the inertia effect and SAM should be combined by the ,
absolute sum method (see Section 3.9.2.ll.2.g of the SRP). Revise Section 3.7.3.9 of the SSAR to reflect this staff position. In addition, either revise or delete the exception to Section 3.9.2.II.2.g of the SRP in WCAP-13054. !
Response: l The SSAR will be revised as shown below to reflect the Staff's position. The exception to Section 3.9.2.ll.2.g of j the SRP will be revised in the next revision of WCAP-13054, i SSAR Revision:
i In Subsection 3.7.3.9 revise the third and fourth paragraphs under Seismic Anchor Motions as follows: i The results of the modal spectra analysis (multiple input or envelope) are combined with the results fmm l
seismic anchor motion by the :-q== -- : :r c' 6: 1,x=: absolute sum methods, t P.i ; 5:=d r- E 5 "r '.2.f ' ' e' ^.SCE St=id i S4 (":'. ._..r 3). i i
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NRC REQUEST FOR ADDITIONAL INFORMATION l
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Question 210.41 Section 3.6.2.3.1 of the SSAR " Jet Impingement " states that if a simplified static analysis is performed instead of a dynamic analysis, the jet impingement force is multiplied by a dynamic load factor of 1.2 to 2.0, depending upon the time variance of thejet load. The staff's position, which agrees with Section 7.3 of ANSI /ANS 58.2-1988 (Ref. 4 in Section 3.6.4 of the SSAR) is that this load factor should be 2.0. Either revise Section 3.6.2.3.1 to reflect this staff position, or provide a more detailed basis for a 1.2 factor.
Response
The SSAR is consistent with ANSI /ANS 58.2-1988. The dynamic load factor between 1.2 and 2.0 is used when the system response can be adequately represented by a one-degree-of-freedom elastic / plastic system. An example of this method is the Approximate Design Methods described in " Introduction to Structural Dynamics" by J. M.
Biggs. (Reference 9)
SSAR Revision:
The third paragraph of Subsection 3.6.2.3.1 will be revised as follows.
If simplified static analysis is performed instead of a dynamic analysis, the preceding jet load (F T ) is multiplied by a dynamic load factor. For an equivalent static analysis of the target structure, the jet impingement force is multiplied by a dynamic load factor of 1.2 to 2.0, depending upon the time variance of the jet load and the elastic / plastic behavior of the target. This factor assumes that the target can be represented as essentially a one-degree-of-freedom system, l
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'l NRC REQUEST FOR ADDITIONAL. INFORMATION I c :m
' Question 210.44 l t
i Section 3.6.2.4 of the SSAR, " Protective Assembly Design Criteria," states that auxiliary guard pipes provide l additional confidence that pipes will not leak into the annulus between the containment wall and the shield building.
l' This implies that these guard pipes are identical to those in the containment penetration area break exclusion zone which are discussed in Section 3.6.2.1.1.4. However, Section 3.6.2.4.2 of the SSAR. " Auxiliary Guard Pipes," f provides design criteria for these guard pipes that is not consistent with the criteria in Section 3.6.2.1.1.4, and is !
unacceptable for guard pipes in the break exclusion zones. Revise Sections 3.6.2.4 and 3.6.2.4 2 to more clearly define the difference, if any, between auxiliary guard pipes and those in the break exclusion zones, and identify .
more specifically where auxiliary guard pipes will be used in the AP600. !
Response: !
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The SSAR will be revised as shown below to clarify the design basis for different types of guard pipes.
SSAR Revision: i Revise the paragraph for subsection 3.6.2.4 as follows: '
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In addition to pipe whip restraints, other protective devices are designed to protect against the effects of postulated pipe ruptures. Barriers and shields are designed to protect against jet impingement. ^ r"!!:rj ;:=d
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Guard pipes lisi the break exclusion 2ones provide additional confidence that pipes will not leak into the annulus
. between the containment vessel and the shield building.
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Revise the paragraph for subsection 3.6.2.4.2 as follows:
(The use of guard pipes has been minimized by plant arrangement and routing of high<nergy piping. Whece
. y = rd. ;=d;?px =: dx!; d t "::=d dy==I: =d :n inn- '! :ff=:;. of pr^'"'-ed i . ' c' S: ;
- :'n:d p!p:. ^ =:!!:rj ;=dp?pr =: " ::!y :' 'n =r = ! gn:!: rq : un: _:: = h: =::".d. 3:
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dn!;n. q=!!'j ;;==, =d ingx:!:: rz x;un" 'c- t:;=:d;!;x =: i :f an t: =q:;;;x.. : f= ^ 91E Ced:. F=:!: "', C!= 2 p';I;;. Guard pipes in the c'ontainment annulus ~amas"of the break exclusion zones are j designed as described in SSAR subsection 3.6.2.1.1.4. Other guard pipes sae: designed and constructed to the same ASME rules as the enclosed process piper l f
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NRC REQUEST FOR ADDITIONAL INFORMATION g-- uig Question 210.45 Section 3.6.2.4.2 of the SSAR states that auxiliary guard pipes will be constructed in accordance with the rules for ASME Section 111, Class 3 piping. Because of potential in-senice inspection problems, the staff discourages the use of auxiliary guard pipes. However, if the response to Q210.44 indicates that they will be used, the staff's position is that they should be constructed to the same ASME rules as those required for the enclosed piping.
Revise Section 3.6.2.4.2 to reflect this staff position.
Response
Guard pipes that are not in the break exclusion zones are constructed to the same ASME rules as those required for the enclosed piping.
SSAR Revision:
Please see the response to RAI 210.44 for SSAR revisions, i
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l NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.49 Section 3.7.3.8.2.1 of the SSAR states that when run pipe is decoupled from the analytical model of the branch pipe, the connection point is considered to be anchored for seismic inertia analysis of the branch pipe. The response spectra for this analytical anchor are the spectra at the building floor location corresponding to run pipe supports near the connection point. The staff believes that the response spectra at the run pipe suppons may not be a conservative assumption when compared with the actual configuration before decoupling. Revise this section to either change this assumption, or provide a more detailed basis for the assumption.
Response
The Staff has previously accepted the decoupling of the run pipe for the equivalent static load method of analysis of small diameter branch piping systems. In this method, there are no amplified seismic response spectra at the run pipe connection point. The AP600 position extends this method to dynamic analysis of larger diameter branch piping (3 inches and above). This approach is valid when the SSE inertial displacement of the run pipe (or equipment nozzle) is limited to 1 inch in each of three coordinate directions. The basis for this limit is as follows.
The maximum displacement that is expected in the supported piping system is 4 inches based on the 4.5 Srn stress lim;t in the proposed ASME Code revision. In order to preclude significant inertial amplificatwn by the run pipe, the run pipe frequency should be less than one half the branch pipe frequency. Since the inertial displacement is inversely proportional to the square of the frequency, the run pipe displacement should be less than one-fourth of the branch pipe displacement:
DELTA (RUN) = DELTA (BRANCH)*(FREQ(BRANCH)/FREQ(RUN))**2 Therefore if the run pipe (equipment) displacement is less than 1 inch and the branch pipe meets the ASME Code stress limit there will not be significant SSE inertial amplification.
SSAR Revision:
Revise the third paragraph of Subsection 3.7.3.8.2.1 as follows:
When the run pipe is decoupled from the analytical model of the branch pipe, the connection point is considered to be anchored for seismic inertia analysis. The response spectra for this analytical anchor are the spectra at the building floor location corresponding to mn pipe supports near the connection point. The motions of the connection point are determined from a separate seismic inertia analysis of the run pipe. These motions are applied as static anchor motions to the branch pipe. This criterion is accepted industry practice. For example, a three-inch branch pipe can be decoupled from a ten-inch or a larger run pipe. When this approach is used the safe' shutdown earthquake inertial displacement of the tun pipe bruch location (or equipment nozzle) is limited to 1 inch in each of three coordinate dinctions in order to minimize the amplification effects of the run pipe (or equipment) on the building floor response spectra.
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e NRC REQUEST FOR ADDITIONAL INFORMATION j i
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i Question 210.51 l E
Section 3.7.3.9 of the SSAR, " Combination of Support Responses," states that the effect of relative seismic anchor displacements are obtained by either using the worst combination of peak displacements or by proper representation r of the relative phasing characteristics associated with different support inputs. Provide more details relative to how l proper representation is obtained. Identify and justify any deviations from the guidelines in Section 3.9.2.ll.2.g i of the SRP. In addition, either revise or delete the exception to Section 3.9.2.II.2.g of the SRP in WCAP-13054, j if applicable. j
Response
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The AP600 position is the same as the previous position accepted by the NRC in other applications . When all the ;
component supports are in the same structure, the relative seismic anchor motions is small and the effects are I neglected. This is applicable to all building structures and to those supplemental steel frames that are rigid in .l comparison to the components. Supplemental steel frames that are flexible can have significant seismic anchor i motions which are considered. When the components supports are in different structures, the relative seismic l anchor motion between the structures is taken to be out-of-phase and the effects are considered. ;
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This position is different than that in SRP 3.9.211.2.g which requires an analysis for relative seismic anchor motions j when the supports are in the same building. '
SSAR Revision: ,
Reivise the third paragraph of Subsection 3.7.3.9 as Follows: .I The effect of relative seismic anchor displacements is obtained e.ther by using the worst combination of the peak -l displacements or by proper representation of the relative phasing characteristics associated with different support inputs. For components supported by the interior concrete building, the seismic motions at all elevations above the l
basemat are taken to be in phase. When all thebe t supports are in the same structure, the relative seismic j anchor motions is small and the effects are neglected 87his is applicable'to all building structures and to those
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j supplemental steel frames that a rigid in' comparison to the componentsM? Supplemental steel frames that.'are i flexible can have significant seismic anchor motions which are sonsidered. (When thehj q=ha supports lare in different structures, the relative seismic anchor motion between the structures .is takenito be out-of-phase _ and the effects are considered. ;
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NRC REQUEST FOR ADDITIONAL INFORMATION 50; Question 210.67 The design and analysis requirements for ASME Class 1, 2, and 3 pressure-relieving devices discussed in Sections 3.9.3.3 and 10.3.2 of the SSAR do not appear to be consistent with staff positions on this issue. To be acceptable, such installations should be designed in accordance with ASME Section III, Appendix 0, " Rules for the Design of Safety Valve Installations," as supplemented by the additional criteria in Section 3.9.3, Section 11.2 of the SRP. Revise Sections 3.9.3.3 and 10.3.2 to be consistent with this position. In addition, delete the reference to ANSI /ASME B 31.1, Appendix 2 in Section 10.3.2.2.2.
Response
The SSAR will be revised to specify that the ASME Code,Section III Appendix 0 rules for safety valve installations apply to AP600.
SSAR Revision:
Revise Subsection 3.9.3.3 as follows 3.9.3.3 Design and Installation Criteria of Class 1, and-2, and 3 Pressure Relieving Devices The design of pressure relieving valves comply with the requirements of ASME Code,Section III, Appendix O, " Rules for the Design of Safety Valve lastallations.' When there is more than one valve on the same run of pipe, the sequence of valve openings is based on the anticipated sequence of valve opening. .%is sequence is determined by the set point pressures or control system logic. De applicable stress limits are satisfied for all components in the piping run and connecting systems including supports. He reaction forces and moments are based on a dynamic load factor of 2.0 unless a dynamic structural analysis is performed to calculate these forces and moments.
Revise the fourth paragraph of Subsection 10.3.2.2.2 as follows The piping and valve arrangement minimizes the loads on the attachment, and analysis confirms the design by use of guidelines in-AN@ASME E" ' ippen'!!- 2, "" "andcug "u!: . ^- - P^ j;;n " c^'i:y Wh .
W.tn!! n " ASME Code,Section III, Appendix 0, ' Rules for the Design of Safety Valve Installations.'
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NRC REQUEST FOR ADDITIONAL INFORMATION
= m;a Ouestion 210.78 Revision I to WCAP-13054 lists exceptions to Sections C.I.3.3(a) and C.I.3.3(b) of Appendix A to Section 3.9.3 of the SRP, that state that all pipe break loads are classified as Service Level D. The staff does not agree with these exceptions. Pipe breaks other than a LOCA or main-steam /feedwater pipe break are defined as design basis pipe breaks (DBPB) in Section 3.9.3 of the SRP and should be designed to Service Level C limits. Revise WCAP-13054 to delete these exceptions and revise Tables 3.9-5, 3.9-6, 3.9-7, and 3.4-8 of the SSAR to include Sustained Loads + DBPB under *1evel C Service."
Response
The NRC Staff has previously accepted the classification of all pipe ruptures as Service Level D. The use oflevel D stress limits ensures pressure boundary integrity while allowing some permanent distortion of the system and provides adequate protection for these postulated pipe breaks.
SSAR Revision: NONE 21038-1 W westinghouse ;
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- !1 NRC REQUEST FOR ADDITIONAL INFORMATION i 1
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4 Question 260.23 Section 14.2.1 of the SSAR states that preoperational and/or startup testing is performed on those systems that:
- a. Are relied upon for safe shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe condition for an extended shutdown period;
- b. Are relied upon for safe shutdown and cooldown of the reactor under transient and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown period following such conditions;
- c. Are relied upon for establishing conformance with safety limits or limiting conditions for operation;
- d. Are classified as engineered safety features actuation systems (ESFAS) or are relied upon to support operation of engineered safety features actuation systems within design limits;
- e. Are assumed to function during an accident or for which credit is taken in the accident analysis and in the probabilistic risk assessment (PRA); and
- f. Are used to process, store, control, or limit the release of radioactive materials.
To be consistent with the guidance of Regulatory Position (RP) C.1 of Regulatory Guide (RG) 1.68 (Rev. 2, August 1978), the staff believes that paragraphs e, d. and e above should be revised as follows:
- c. Are relied upon for establishing conformance with safety limits or limiting conditions for operation tJJat.
will be included in the facility technical specifications;
- d. Are classified as engineered safety features actuation systems (ESFAS) or are relied upon to support 55 ensure operation of engineered safety features actuation systems within design limits;
- c. Are assumed to function or for which credit is taken in the accident analysis of the facility. as described in the SS AR. and/or in its desien-specific probabilistic ri3k assessment (PRA);
Revise Section 14.2.1 of the SSAR accordingly.
In addition, Section 14.2.1 or 14.2.8 of the SSAR should be revised to identify, if applicable, any startup tests that are to be performed to demonstrate the operability of structures, systems and components that are not considered essential to meet the criteria of RP C.1 of RG 1.68 fRev. 2. August 1978).
Response
To be consistent with the guidance of Regulatory Position (RP) C.1 of Regulatory Guide (RG) 1.68 (Rev. 2, August 1978), paragraphs e, d and e will revised as shown below.
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.l NRC REQUEST FOR ADDITIONAL INFORMATION at:::
l There are no tests which demonstrate the operability of structures, systems and components that are not considered essential to meet the criteria of RP C.1 of RG 1.68 (Rev. 2, August 1978). Therefore, a revision to l
Section 14.2.1 or 14.2.8 is not applicable.
SSAR Revision:
Revise the sixth paragraph of Subsection 14.2.1 as follows:
Precerational and/or startup testing is performed on those systems that:
- a. Are relied upon for safe shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe condition for an extended shutdown period;
- b. Are relied upon for safe shutdown and cooldown of the reactor under transient and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown period following such conditions;
- c. Are relied upon for establishing conformance with safety limits or limiting conditions for operation that will be included in the facility technical specifications;
- d. Are classified as engineered safety features actuation systems (ESFAS) or are relied upon to support or ensure operation of engineered safety features actuation systerns within design limits;
- e. Are assumed to function or for which credit is taken in the accident analysis of the AP600 as described in the SSAR, or in the AP600 probabilistic risk assessment (PRA);
- f. Are used to process, store, control, or limit the release of radioactive materials.
260.23-2 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION EM di.1
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!Sl Question 410.167 Why does Section 9.3.5 of the SSAR discuss sumps and drain tanks and Figure 9.3.5-1 only shows sumps? Include the drain tanks in the figure.
Response
The drain tanks referenced in Subsection 9.3.5 are included in the waste water system which is described in Subsection 9.2.9 Refer to the response for RAI 410.163 for the update of Figure 9.2.9-1 to include drain tanks.
SSAR Subsection 9.3.5.2.1 of the equipment and floor drainage system includes a reference to Figur s 9.2.9-1 for the waste water system components.
SSAR Revision: NONE l
1 410.1 GL1 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.184 Section 6.5.3 of the SSAR does not provide information on system and component descriptions for the fission product control systems. If there is no such system for AP600, Section 6.5.3 should be either rewritten to explain which systems or components will perform the fission product control function. or deleted.
Response
There are no active systems to provide fission product control in containment following a postulated accident. The containment atmosphere is depleted of elemental iodine and particulates as a result of the natural removal processes as discussed in SSAR section 6.5.2,15.6.5.3.2 and in the response to RAI 450.8 Revision 1. The fission control mechanisms and the limited containment leakage from the containment result in offsite doses less than the guideline values of 10 CFR 100.
SSAR Revision:
Revise Subsection 6.5.3 (add a paragraph) as follows:
6.5.3 Fission Product Control Systems The containment atmosphere is depleted of elemental iodine and particulates as a result of the natural removal processes as discussed in SSAR section 6.5.2 and 15.6.5.3.2. ' No active fission product control systems are required in the' AP600 design to meet regulatory requirements. ~ The fission control mechanisms and the limited containment leakage from the containment result in offsite doses less than the guideline values of 10 CFR 100.
410.184-1 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 410.186 Section 10.4.5.2 of the SSAR states that the circulating water system (CWS) and cooling tower are applicable to a broad range of sites. On other ALWRs, the heat sinks for the CWS are site dependent. A conceptual design and interface requirements are provided for the normal heat sink and, in some cases. for portions of the CWS that are outside of the design certification scope. The AP600 SSAR did not provide sufficient infonnation on the CWS design or alternative design requirements such as protecting safety-related equipment in the event of failure of the CWS, and kicating the cooling tower far enough from safety-related structures to prevent damage in the event of a cooling tower failure. Provide design descriptions and interface requirements for the CWS as required by 10 CFR Part 52.
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Response
1 As described in SSAR Subsection 10.4.5.2, a reference design is provided for the AP6(X) circulating water system (CWS) and for the cooling tower (which is part of the CWS). The reference design includes a hyperbolic natural draft cooling tower for the heat sink. The Combined License applicant may modify this design to meet site-specific requirements. The Combined License applicant is responsible for determining the system configuration.
The reference design has been evaluated to verify that postulated CWS failures have no adverse impact on any salety-related systems, structures or components. The cooling tower is k)cated sufficiently distant from safety-related structures that its postulated collapse does not affect any safety-related structure. As discussed in SS AR Subsection 10.4.5.2.3. a postulated CWS line break or expansion joint failure within the turbine building has no detrimental effects on safety-related systems, structures or components. A postulated CWS line break in the yard area or a failure of the cooling tower basin has no detrimentid effects on safety-related systems, structures or components.
As addressed in SSAR Subsection 3.4.1.1.1, the site is graded to drain water away from the seismic Category I structures and below grade seismic Category I structures are protected from flooding by waterproofing membranes and waterstops.
SSAR Revision:
- 1. SSAR Subsection 10.4.5.2.3, last paragraph The cooling tower is located sufficiently distant fmm the nuclear ishtnd structures that, in the event of collanse of the cooling tower, there is no potential for damage towlef 4 anktnwtweime-the : := ; . & .:pw44m41afw inwweyequipment, components, or structures required for safe shutdown.
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NRC REQUEST FOR ADDITIONAL INFORMATION
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- 2. Revise SSAR Table 1.8-1 av follows:
Table 1.81 (Sheet 6 of 8)
Summary of AP600 Plant interfaces With Remainder of Plant Item Interface Interface Type Matching Section No. Interface or Sub-Item section 10.1 Design t.nd location of CWS cooling AP6fX) Interface Combined License 10.4.5 tower to avoid adverse impacts on safety- applicant related structures coordination 11.1 Expected release rates of radioactive mate- Site Interface Site specific pa- 11.2 rial from the Liquid Waste System includ- rameters ing:
Location of release points Effluent tempemture Effluent flow rate Size and shape of flow orifices i 1.2 Expected release rates of radioactive mate- Site Interface Site specific pa- 11.3 rials from the Gaseous Waste System rameters including:
Location of release points Height above grade Height relative to adjacent buildings Effluent tempemture Effluent flow rate Effluent velocity Size and shape of flow orifices 410.186-2 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION N
11.3 Expected release rates of radioactive mate- Site Interface Site Specifk 11.4 rial from the Solid Waste System includ- parameters ing:
Location of release points Material types Material qualtities Size and shape of material containers i
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NRC REQUEST FOR ADDITIONAL INFORMATION sH-Ouestion 410.194 Where is flood water routed should the Spent Fuel System (SFS) tail? There are no sumps shown in the SFS area on Figure 1.2-2 of the SSAR.
Response
Flooding in the spent fuel system area (rooms 12272 and 12273) drains to the Level 1 sump via floor drains and/or via flow under the spent fuel system room door to the corridor on Level 2 and then downward via the hatch grating and various other paths to the I.evel 1 sump.
SSAR Revision: NONE 410.194-1 l 3 W85tingh00S8 i i
NhC REQUEST FOR ADDITIONAL INFORMATION dW 2[F Question 410.232 Section 9.1.2.2 of the SSAR states that the spent fuel storage facility is located within the seismic Category I auxiliary building fuel handling area. However, it is not clear that the spent fuel facility itself is a seismic Category I structure. If the spent fuel storage facility is not a seismic Category I structure, provide the rationale for concluding that the design of the facility is in compliance with the guidance of the Standard Review Plan, Regulatory Guides 1.13 and 1.29, and the requirements of General Design Criterion 2.
Response
The spent fuel facility is classified as seismic Category I. The building structures associated with the spent fuel storage facility, including the spent fuel pool, fuel tranfer canal, and cask loading pit are seismic Category I, as indicated in SS.AR Table 3.2-2. The classification of components is given in SSAR Table 3.2-3. The spent fuel racks, the cask handling crane, and the fuel handling machine are also seismic Category I.
SSAR Revision: NONE 41 .232-1 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
= mu IN Question 410.239 Provide all of the site-specific interface requirements and combined operating license (COL) applicant information, as appropriate, for the HVAC systems in the corresponding sections of the SSAR.
Response
There are no site-specific interface requirements or Combined Operating license applicant information beyond that specified in SSAR Section 1.8 and Chapter 2.
SSAR Revision: NONE i
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NRC REQUEST FOR ADDITIONAL INFORMATION mg s
Question 410.252 Paragraph 11.2 of Section 10.3 of the SRP states that the design of the MSSS is acceptable if the integrated design of the system meets GDC 4, with respect to the safety-related portions of the system being capable of withstanding the effects of external missiles, internally generated missiles, pipe w hip, and jet impingement forces associated with pipe breaks, and Position C.1 of RG 1.115 as related to the protection of structures, systems, and components important to safety from the effects of turbine missiles. How does the AP600 MSSS design meet this guidance?
Response
Protection of safety-related portions of the main steam supply system are described in the following sections of Chapter 3:
External missiles - Section 3.5 Internally generated missiles - Section 3.5 Pipe whip, and jet impingement forces associated with pipe breaks - Section 3.6 Turbine missiles - Section 3.5 SSAR Revision: NONE l
l T Westinghouse 410.252-1
I NRC REQUEST FOR ADDITIONAL INFORMATION E- as
.,7 Question 410.256 I l
Section 10.4.2.2.1 of the SSAR states that no hydrogen buildup is anticipated in the main condenser as described in Subsection 10.4.1. 2.1. " The staff believes that the referenced subsection number should be 10.4.1.3 instead of !
10.4.1.2.1 because the subject of hydrogen is not mentioned in Section 10.4.1.2.1. Section 10.4.1.3 indicates that no hydrogen buildup in the main condenser is anticipated. Is this correct?
Response
No hydrogen buildup in the main condenser is anticipated. SSAR Subsection 10.4.2.2.1 will be revised to reference SSAR Subsection 10.4.1.3.
SSAR Revision:
Revise the first paragraph of Subsection 10.4.2.2.1 as follows: .
The condenser air removal system, as shown in Figure 10.4.2-1, consists of three liquid ring vacuum pumps that remove air and noncondensable gases from the two condenser shells during normal operation and provide condenser hogging during startup. The noncondensable gases, together with a quantity of vapor, are drawn from ,
the condenser shell through the air cooler section to the suction of the air removal equipment. These j noncondensables consist mainly of air, nitrogen, and ammonia. Since no hydrogen buildup is anticipated in the main '
condenser (as described in Subsection 10.4.1.143), no hydrogen buildup is anticipated in the system. Dissolved oxygen is present in the condensate and. condenser hotwell inventory. Only trace amounts of this oxygen are ,
released in the condenser, and the amounts are considered negligible compared to the amount of gas and vapor being !
evacuated by the system. Therefore, the potential for explosive mixtures within the condenser air removal system ;
does not exist.
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NRC REQUEST FOR ADDITIONAL INFORMATION
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b $7 Question 440.67
. Section 9.2 of Chapter 3 of the EPRI ALWR Requirements Document identifies requirements regarding the RHR l system during mid-loop operation.- For example, paragraph 9.2.1.3 states that a single failure in the RHR system ;
with reactor vessel head removal should not cause the water in the RCS or the reactor cavity to boil; i paragraph 9.2.2.1.1 states that analysis should be performed for all potential RHR conditions that properly account :
for sources of error during mid-loop operation; paragraph 9.3.1.2 lists various featums to prevent or mitigate the 'i effects oflosing suction to the RHR pumps when the RCS level is lower. Table B.1-2 of Appendix B to Chapter 1 of the EPRI Requirements Document indicates that designers meeting the guidance of this document should comply with Generic 1.etter 88-17 regarding loss of decay heat removal during mid-loop operation. ;
A Address compliance of the AP600 RHR system during mid-loop operation with the EPRI Requirements Document, identify any deviations, and provide justification for each of the deviations identified (see also Q440.72). >
t
Response
While the ALWR Utility Requirements Document is not a regulatory requirement, the AP600 normal residual heat removal system design conforms to the EPRI ALWR Utility Requirements Document. A summary of the AP600 compliance with Generic Letter 88-17 is presented in SSAR Subsection 1.9.5.1 Advanced Light Water Reactor ,
Certification Issues, SECY-90-016 under the subheading of "Mid-loop Operations." ,
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SSAR Subsections 1.9.5.1, 5.4.6.1, 5.4.7.2.1, and 6.3 discuss the design features and system performance capabilities incorporated into the design of the AP600 reactor coolant system, normal residual heat removal system, and passive core cooling system to address the issues identified by the Industry and the NRC related to reactor coolant system mid-loop operations !
j SSAR Revision: NONE .
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NRC REQUEST FOR ADDITIONAL INFORMATiON n:e ze; Lp %
Ouestion 440.72 Generic Letter 88-17 identified action items to reduce shutdown risk. The 14 action items as summarized in Table 5.2 of NUREG-1449 address issues ranging from operations, events, hardware design procedures, analyses and instrumentation. Address the proposed resolution to each of these action items for the AP6(X) design during shutdown and mid-loop operations (see also Q440.53,0440.55, Q440.56, Q440.58, Q440.67, and Q440.71).
Response
A summary of the AP600's compli:uice with Generic Letter 88-17 is presented in SSAR Section 1.9.5.1-Advanced Light Water Reactor Certification Issues. SECY-90-016--under the subheading of "Mid-loop Operations." See also the responses to RAls 440.53,440.55,440.56,440.58,440.67, and 440.71 for related discussions of this issue.
SSAR Revision: NONT i
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NRC REQUEST FOR ADDITIONAL INFORMATION E fiil 2 iEl mi Question 440.97 Section 6.3.2.2.3 of the SSAR states that flow out of the IRWST during the injection mode includes conservative allowances for spill flow during a direct vessel injection line break.
- a. Describe these conservative allowances.
- b. Has the bypass phenomena of the safety injection through the DVI been properly modeled in the analysis codes, and verified and validated by appropriate testings?
Response
The conservatisms considered in a direct vessel injection (DVl) line break LOCA include assuming minimum line resistances in the faulted injection line to maximize the blowdown of the reactor and the spill from the in-containment refueling water storage tank. In addition, maximum line resistances are assumed in the intact injection line to minimize the safety injection flow that reaches the reactor.
Both of the APU)0 integral test facilities (SPES-2 and OSU) will perform double ended direct vessel injection line break tests. When these tests are completed, the test results will be used to verify the AP600 safety analysis models.
SSAR Revision: NONE W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION F
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Question 480.53 Table 6.2.31 of the SSAR references Regulatory Guide (RG) 1.11 for four containment pressure instrument line penetrations. Since the isolation valves in these lines are not automatic or remote-operated, the penetrations do not conform to this regulatory guide. The instrument lines are also described in Section 6.2.3.1.1 of the SSAR. The penetrations should conform to RG 1.11 unless an other acceptable basis is defined. Address this concern.
Response
The design of the four containment pressure instrument lines is consistent with Regulatory Guide 1.141
" Containment Isolation Provisions for Fluid Systems", and ANS standards N271-1976 " Containment isolation Provisions for Fluid Systems" and ANSI /ANS - 56.2 " Containment Isolation Provisions for Fluid Systems after a LOCA" These lines sense pressure within containment and are connected to pressure transmitters located outside containment. Signals from the transmitters are utilized to generate a safety injection and containment isolation signal and the channels are an integral part of the post accident monitoring function. In view of these functions, it is essential that the lines remain open and not be isolated following an accident. To provide these safety related functions and also to meet the containment isolation needs, a sealed sensing line is utilized.
Each of the four channels has a separate containment penetration. The design configuration includes a sealed bellows inside containment connected to a pressure transmitter located outside containment in close proximity to the i penetration. The pressure transmission between the sealed bellows and transmitter is via a sealed, fluid filled, tube.
The tubing. bellows and transmitter are designed to maintain integrity during containment transients. The configuration therefor provides a double barrier (a sealed bellows inside and a pressure transmitter outside containment) to prevent the release of radioactivity from tne cora.inment atmosphere and consistent with the intent of the General Design Criteria.
SSAR Revision:
A revised version of Table 6.2-3 is provided in response to RAI 480.61.
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NRC REQUEST FOR ADDITIONAL INFORMATION
= l 9 j Ouestion 480.54 Figure 5.4-7 of the SSAR does not depict TV&D connections necessary for Type C testing of the residual heat removal (RHR) suction isolation valves. However, Table 6.2.3-1 indicates that Type C testing is to be performed.
Are the necessary testing provisions provided? Clarify the discrepancy bernen Figure 3.4-7 and Table 6.2.3-1 as to w hich valve closes on a radiation signal.
Response
Table 6.2-3 is correct in specifying type C testing will be perfornvxl on nonnal residual heat removal system suction line containment penetration in accordance with Appendix J requirements. Although the current P&ID does not specify the necessary test connections to perform the testing, Revision 2 of the AP600 SSAR willinclude a revised version of Figure 5.4-7 with necessary connections added to perform the required testing.
SSAR Revision:
Figure 5.4-7 will be revised with the connections necessary to perform the required testing added.
W WB5tingh0US8
NRC REQUEST FOR ADDITIONAL INFORMATION
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Question 480.65 This question pertains to Westinghouse's statement of conformance to paragraph 6.1.1 of the Standard Review Plan,
" Engineered Safety Features Materials," Criteria B.4, Regulatory Guide (RG) 1.54, and paragraph 6.1.2 of the Standard Review Plan. " Protective Coating Systems (Paints)," Criteria B.4. Regulatory Guide (RG) 1.54, that are identified on page 6-5 of Revision i to WCAP-13054, "AP600 Compliance with SRP Acceptance Criteria."
The WCAP indicates that coatings are non-safety-related.
Does the non-safety-related aspect of the coatings mean that debris is not considered a safety issue for the AP600 design? If debris could be a hazard, explain this position. Could this debris clog sump strainers, condensate gutters, or the drain of the in-containment refueling water storage tank?
Response
The coatings inside the containment are not safety-related as explained in the response to RAI 281.4 and RAI 252.122. Even though they are not safety related, the coating materials will still be qualified, applied, and inspected in accordance with stringent specification requirements. SSAR subsection 6.1.2.1.5 addresses each type of coating utilized within containment and why potential debris is not considered a safety issue. The subsection proceeds to identify steps that will be taken so that the non safety-related coatings perform as intended. Additional details on the coatings are also provided in response to RAI 252.28.
SSAR Revision: NONE i
480.654 T Westinghouse
l NRO REQUEST FOR ADDITIONAL INFORMATION
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Question 480.73 This question pertains to Westinghouse's statement of conformance to paragraph 6.2.4 of the Standard Review Plan.
" Containment isolation System," Item 6.n, that is identified on page 6-16 of Revision I to WCAP-13054, "AIWX)
Compliance with SRP Acceptance Criteria."
The WCAP indicates that a closure of 60 seconds or less has been established with the new source term being used to justify the time. It is the staff's position that closure time should also be as quickly as practical. The staff believes that the containment integrity will be in question until the valves are actually closed. Therefore, closure time will be viewed from the ability to close as well as the source term. Address this concern.
Response
The selection of closure time for containment isolation is based on an integrated design approach focussing on reliability and safety rather than speed. Consideration included optimizing the component design consistent with the physically based source term timing, selection of valve operating requirements based on maximizing reliability, and identification of critical valves to retain rapid closure times based o ,ih:hing releases or other overriding safety-related factors.
- Industry standard closure times were selected for penetzations when closure within that time frame would not represent any decrease in public safety due o delayed closure compared to the typical 10 second closure time of previous pressurized water reactors. The standard ;adustry closure time permits a more reliable design focussed on opening and closing performance without sacrificing for speed.
- Containment isolation valve closure times are providd in SLAR Table 6.3.2-1. The critical valves identified are containment air filtration system and the main ste. , ace', feedwater isolation valves. The containment air filtration system valves are regarded as critical since unlire other penetration, releases to the containment atmosphere are immediately available for release through the coutainment air filtration system. He main steam and feedwater isolation valves are determined to requir rapid closure based on limiting mass and energy to the containment following a main steam line break.
De result of this integrated approach is not a containment design that responds to an event as quickly as practical or that raises additional questions of the ability to establish integrity. Rather, the containment isolation design is based on a methodical design cancept that has been established based on the variety of design and scenario considerations and enhances overall containment integrity potential.
SSAR Revision: NONE 480.73-1 g_ Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 2 Ouestion 720.158 Provide a detailed desenption of the methodology used in performing the PRA-based seismic margins analysis. This description is particularly important because there are limited examples of the practical implementation of this methodology.
Response: tRevision 2)
Attachment I to this RAI response provides the methodology for perfonning the AP600 risk-based seismic margins analysis.
Attachment 2 to this RAI response provides the AP600 seismic margin assessment. The snethodology presented in Attachment I was followed with the exception of the scismic containment performance analysis. The containment performance was evaluated by considering the availability of the air baffle between the vessel and shell and by closure of containment isolation valves to maintain containment integrity. These events were included on the seismic event trees.
PRA Revision: See Attachment 2.
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NRC REQUEST FOR ADDITIONAL INFORMATION N e.
Response Revision 2 ATTACllMENT 1 to RAI 720.158(R2)
METilODOLOGY FOR PERFORMING Tile AP600 RISK-HASED SEISMIC MARGINS ANALYSIS
1.0 INTRODUCTION
In accordance with Section ll.N Site-Specific Probabilistic Risk Assessments and Analysis of External Events. of SECY-93-087 (Reference 1), the Nuclear Regulatory Commission approved the following staff recommendation:
"PRA insights will be used to support a margins-type assessment of seismic events. A PRA-based seismic margins analysis will consider sequence-level High Confidence, Low Probability of Failures (HCLPFs) and fragilities for all sequences leading to core damage or containment failures up to approximately one and two-thirds the ground motion acceleration of the Design Basis SSE."
The seismic margins analysis presented in AP600 Probabilistic Risk Assessment (PRA) Appendix H.3 and request for additional information (RAI) response 720.15 was performed prior to the issuance of the SECY report. The Appendix H.3 seisuse margins analysis is based on a cutset-level rather than on a sequence-level evaluation.
The most efficient and complete way to answer RAls 720.158, 720.159, 720.162 and 720.166 is to perform additional seismic margins analysis that supports and supplements the present analysis.
The risk-based seismic margins analysis methodology discussed herein and the analysis to be performed in accordance with this methodology satisfies the intent of SECY-93-087 and answers the NRC RAls.
Table 1.0-1 provides brief definitions of PRA terminology.
Since the AP600 nonsafety related components are not seismic category I,it is conservatively assumed for the risk-based seismic margins analysis that no credit is taken for the mitigation functions of the nonsafety-related components and systems. A PRA sensitivity analysis has been performed that does not credit the mitigation functions of the nonsafety-related systems. The analysis is referred to as the focused PRA (Reference 2) and was submitted to the NRC as a part of resolving the Regulatory Treatment of Nonsafety Systems (RTNSS) issue for the AP600. The focused PRA is based on the AP600 baseline PRA analysis (Reference 3). The focused PRA event tree and fault tree models are modified as necessary for the seismic margins analysis. Note the mission time for the baseline and focused PRAs and thus for the seismic margins analysis is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time an accident occurs.
720.158(R2)-2 3 Westinghouse
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1he safety-related systems and functions which are included in the focused PRA evaluation, and thus in the risk-based seismic margins analysis, are listed in Table 1.0-2. i For this risk-based seismic margins analysis, HCLPFs are calculated and reported for systems at the sequence level. This is accomplished by calculating HCLPFs for each seismic event tree top event that represents a safety-related system or function. Once HCLPFs for the necessary systems are calculated, HCLPF values are calculated for each event tree core damage sequence. In addition, insights related to random and/or human failures are reported, as deemed appropriate, for each sequence.
'E 720.158(R2)-3 1 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table 1.0-1 (Sheet 1 of 2)
DEFINITION OF PRA TERMINOLOGY RTNSS Regulatory Treatment of Nonsafety Systems. He AP600 uses passive safety systems that rely on natural forces such as density differences and gravity to provide water for core and containment cooling. Rese passive safety systems do not include active equipment such as pumps. The AP600 active systems are designated as nonsafety-related systems except for limited portions of the systems that provide safety-related isolation functions, such as containment isolation.
Credit is not taken for these active systems in the Chapter 15 licensing design basis accident analyses unless their operation makes the consequences of an accident more limiting. The nonsafety-related systems in the AP600 supplement the capability of the safety-related passive systems. The NRC and industry have defined a process to evaluate the importance of the nonsafety-related systems and for maintaining appropriate regulatory oversight referred to as RTNSS.
Baseline PRA Re baseline PRA credits mitigation functions for those safety-related and nonsafety-related systems and components modeled in the PRA. The PRA evaluation, which is documented in the AP600 PRA report (Reference 3), was submitted to the NRC in June 1992.
Focused PRA The focused PRA credits mitigation functions for only those safety-related systems and components modeled in the baseline PRA. The focused PRA was submitted to the NRC in September 1993.
Initiating event A failure caused by the seismic event that induces a reactor trip.
Event tree Inductive logic models for identifying the possible outcomes of a given initiating event. An event tree is developed for each type of initiating event. De event tree models the subsequent events (called top events) that are required to mitigate the accident.
Top event System or function success / failure criteria for the mitigation of an accident, as defined by the event tree.
720.158(R2)-4 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION 1EE iss Response Revision 2 Table 1.0-1 (Sheet 2 of 2)
DEFINITION OF PRA TEIB11NOLOGY Fault tree A logic model that represents the possible ways a system or function may fail based upon the failure criteria defined by the event tree top event. The logic model indicates how the combination of failure of a component or other events such as operator errors cause the system to fail.
Basic event The equipment faults or other faults (such as operator errors) modeled in the fault tree. No further development of the fault is evaluated from this point.
Cutset Combination of basic event failures that cause the top event to occur.
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Response Revision 2 Table 1.0-2 SAFETY-RELATED SYSTEMS AND FUNCTIONS CREDITED IN THE RISK-IIASED SEISMIC MARGINS ANALYSIS
- Passive core cooling
- In-containment refueling water storage tank injection and containment recirculation
- Core makeup tank
- Passive residual heat removal (PRHR)
- Automatic depressurization
- Containment systems
- Passive containment cooling
- Containment isolation
+ Class IE de and uninterruptible power suppiy (UPS) power
+ Protection and safety monitoring
- Engineered safeguards actuation
- Safety-related monitoring
+ Reactor coolant pump trip
- Steam generator isolation -
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l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 2.0 METIIODOLOGY Steps involved in performing a risk-based seismic margins analysis include: seismic initiating event evaluation, seismic event tree development, systems analysis, accident sequence quantification, and containment isolation capabilities. Evaluation of the seismic initiating events and development of the seismic event trees is discussed in Section 2.1. HCLPFs are calculated for the seismic category I safety-related systems which are called upon via the seismic event trees to mitigate an accident caused by the initiating seismic event. De systems evaluation is discussed in Section 2.2. De core damage sequences defined by the seismic event trees are " quantified." he event trees are not quantified to generate core damage sequence frequencies, but rather the result of the quantification is a list of core damage sequences with the IICLPF for each sequence and, where appropriate, the random and human error failure probabilities for the sequence. The quantification process is discussed in Section 2.3. The final step of the risk-based seismic margins analysis is to perform a seismic containment performance analysis. We process is discussed in Section 2.4.
2.1 SEISMIC INITIATING EVENT AND EVENT TREE ANALYSIS The first step to performing a seismic analysis is to evaluate which initiating events could occur as a result of a scismic event. He risk-based seismic margins analysis does not consider seismic hazard curves; therefore, initiating event frequencies are not calculated for each seismically-generated initiating event category. Although seismically-generated initiating event frequencies are not calculated, it is important to evaluate the seismic vulnerability of the components and systems that contribute to the initiating event categories. This is done by estimating a HCLPF for each seismic initiating event category. The categories, which are based on the RTNSS focused PRA initiating events, include:
General transient (includes loss of offsite power)
Loss of coolant accidents (LOCAs)
Steam generator tube rupture (SGTR)
Steam line breaks Loss of feedwater without scram (ATWS) (seismically-induced failure to scram).
In the RTNSS focused PRA, seven of the baseline PRA transient initiating events are grouped with the loss of offsite power event creating a " general transient" category. These seven events include: turbine or reactor trip, loss of feedwater to steam generator, secondary to primary side power mismatch, spurious "S" signal, loss of component cooling water system, loss of service water system, and loss of compressed air system. For the seismic margins analysis, the general transient event is essentially a loss of offsite power event.
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- =1 Responso Revision 2 m __
It is not necessary to estimate a HCLPF for the general transient initiating event, since offsite ac power is a nonsafety-related system for AP600. Mitigation of the general transient initiating event is evaluated in the risk-based seismic margins analysis. ;
A HCLPF is estimated for the remaining four initiating event categories.- If the seismic initiating event HCLPF is projected to be less than 0.5g, which represents one and two-thirds the ground motion ;
acceleration of the AP600 design basis SSE, then the initiating event is further evaluated in the risk-based l seismic margins analysis. If the seismic initiating event HCLPF is greater than 0.5g, then no further l seismic margins evaluation is required for that event.
Once the seismic initiating events with a HCLPF of less than 0.5g are determined, a seismic event tree f is developed. The event trees used in the AP600 focused PRA provide the starting point for developing {
the necessary seismic event trees.
An example of a loss of offsite power event tree is shown in Figure 2.1-1. The top events on the event l tree represent the safety-related systems that are required to operate in order to mitigate the accident from !
progressing to a core damage event. Since ac power is nonsafety-related, the probability of recovering the grid within I hour or within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as questioned in the event tree, is set to 1.0, meaning offsite power is unavailable following the seismic event. The example loss of offsite power event tree can be redrawn as shown in Figure 2.1-2. j in order to evaluate the focused PRA event trees for the risk-based seismic margins analysis, a HCLPF '
and, as appropriate, a random failure or human error probability is assigned to each event tree top event. l The seismic event trees are created by assigning a HCLPF and random failure or human error probability 3 to each top event.
The methodology for assigning HCLPFs and failure probabilities for each event tree top event is discussed l in Section 2.2. !
i 2.2 SYSTEM SEISMIC ANALYSIS l
i in order to quantify the seismic event trees, each event tree top event is assigned a HCLPF value. In addition, random failures and human error events whose probabilities are greater than or equal to IE-3 l are recorded for each top event. The probability cutoff of IE-3 is based on the NRC request in RAI 720.159, part b. This section discusses the calculation process used to define the system or top event HCLPF value. An example is provided.
For each top event, a HCLPF value is estimated. For example, the top event would be the failure of a !
safety-related system which is requested to operate to help mitigate an accident caused by a seismic event. .
The component and structural failmes of the system and its supports are considered. The components to l l
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 he evaluated could include: pipes, valves, tanks, heat exchangers, and any support system components such as instrumentation and control equipment. The buildings containing the equipment are also included.
A conservative assumption to be used in the PRA-based seismic evaluation for creating the seismic fault trees is that if one component fails due to the seismic event, then all components of that same type for that system fail as well; thus creating a common cause failure event. For example,if a system contains two parallel motor-operated valves, and one fails due to the seismic event,it is assumed that the second valve fails as well. This assumption is conservative because the components may experience different
~
excitation due to their location, orientation or support condition.
A seismic fault tree for each event tree top event is created. Included in the fault tree are the structural and component seismic failures that could cause the system to fail. A HCLPF value is assigned for each of the fault tree basic events that represent components and structures. HCLPF values are reported in Table H-1 of the AP600 PRA report for components and stmetures.
The seismic fault tree is " quantified" using the min-max approach to determine the system HCLPF. The min-max approach is defined as follows:
l
+
For events under an AND gate, all basic events must occur in order to progress up a level in the fault tree. Thus, the highest HCLPF value for the events under an AND gate represents the cutset or top event HCLPF.
+
For events under an OR gate, any one of the basic events occurring cause progression up the fault tree. Rus, the lowest HCLPF value represents the top event HCLPF.
Figure 2.2-1 represents a seismic fault tree. He HCLPF values for each fault tree basic event are recorded under the event (event X has a HCLPF of 0.62g, event Y has a HCLPF of 0.91g). The HCLPF value that represents the event "A and B and C occurs" is calculated as follows:
HCLPF for event A 0.53g HCLPF for event B 0.62g HCLPF for event C 0.91g Since these three events are connected under an AND gate, the maximum (highest) HCLPF value for the events is selected to represent the HCLPF for event "A and B and C occurs." Thus, the HCLPF for this l cvent is 0.91g. Now the HCLPF for system Q is calculated as follows:
i l
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NRC REQUEST FOR ADDITIONAL INFORMATION 1HE Hrd ,
Response Revision 2 HCLPF for event X 0.62g HCLPF for event Y 0.91g HCLPF for event Z 0.86g HCLPF for event A&B&C 0.91g Rese four events are combined under an OR gate. Using the min-max approach, the minimum (lowest)
HCLPF value under the OR gate is selected. Thus, the HCLPF for top event " System Q" is 0.62g.
To further illustrate the system seismic analysis the loss of offsite power example discussed in Section 2.1 is continued here.
As shown in Figure 2.1-2, there are seven top events:
- No battery common cause Passive RHR (PRHR)
. Core makeup tank
- Automatic depressurization Accumulators
- Injection
+ Recirculation A HCLPF is calculated for each top event. In order to calculate the HCLPF, a seismic fault tree is I developed for each top event. The seismic fault tree illustrates the ways the system may seismically fail.
For example, the passive RHR may seismically fail due to failure of any of the following structures or components:
. PRHR heat exchangers
. PRHR valves
+ PRHR piping
- Containment building interior structures (stmeture housing the PRHR system) {
+ Support system failures (dc power to open valves) J Since any one of these structure or component groups failing cause system failure, the events are combined under an OR gate. Figure 2.2-2 illustrates the passive RHR seismic fault tree. Based on the ,
HCLPF values for components and structures presented in AP600 PRA Table H-1, a HCLPF value is i assigned to each fault tree basic event. The HCLPF values are shown on the fault tree under each event.
The fault tree is then quantified using the min-max approach.
Since all events in the fault tree are under an OR gate, the minimum HCLPF value represents the passive RHR system HCLPF. Rus, the passive RHR system HCLPF is evaluated to be 0.5g.
720.158(R2)-10 W westinghouse i
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 The second item that is evaluated for the passive RHR top event is whether any random failures or human errors with a cutset probability of IE-3 or greater exist for the system. This is determined by examining the passive RHR system dominant cutsets from the focused PRA. Any cutset whose probability is greater than IE-3 is recorded. De cutset may represent only random failures, only human errors, or a combination of both random and human failures.
According to the example loss of offsite power event tree shown in Figure 2.1-2, two passive RHR fault trees are called upon: RHR1 and RHR2. De cutset files for both of these trees is reviewed to determine if any cutsets exist whose probability is greater than IE-3. In neither case was a cutset found to be greater than IE-3. Bus, there are no random or human failures to report for the passive RHR top event in the example loss of offsite power seismic event tree.
Once a HCLPF value is determined for each seismic event tree top esent, the event trees are quantified as discussed in Section 2.3.
2.3 SEISMIC EVENT CORE DAMAGE SEQUENCE EVALUATION Once the seismic event trees are created and a HCLPF is estimated for each event tree top event, the event tree core damage sequences can be evaluated.
De seismic event tree is quantified using the min-max approach that was discussed in Section 2.2. Each event tree sequence leading to a core damage end state is evaluated.
An example of a transient seismic event tree with fictitious top event HCLPFs is presented in Figure 2.3-1.
He quantification of the seismic event sequences leading to a core damage end state is presented in Table 2.3-1.
2.4 CONTAINMENT PERFORMANCE SEISMIC EVALUATION The next step in the risk-based scismic margins analysis is to proceed to the containment evaluation by generating a list of seismic event sequences leading to containment failure.
The Level 2 focused PRA containment event trees (CET) are the starting point for developing the seismic CETs. De process steps for evaluating the Level 2 PRA CETs are essentially the same as described in Sections 2.1 through 2.3 for the scismic core damage sequences. First, the focused PRA CETs are modified as necessary to create seismic CETs. Second, the systems and functions representing the seismic CET top events are evaluated by calculating a HCLPF for each top event. This is done in the same i manner as described in Section 2.2. The third step is to quantify the seismic CETs. Before the seismic l CETs can be quantified, the entry HCLPF from the core damage sequences leading into the seismic CET !
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 must be determined. Note that the core damage end states (l A, IB, etc.) define widch containment event tree to proceed to. The minimum sequence HCLPF from the core damage evaluation that leads to the specific CET is conservatively used. For example, suppose three core damage seismic sequences lead to core damage end state I A, and the three sequence HCLPFs are:
Seismic Event Tree Seauence No. HCLPF Small LOCA 5 0.62g Steam generator tube rupture 2 0.86g Loss cf offsite power 35 0.59g The HCLPF chosen to represent the entry into the 1A seismic CET is 0.59g. Any one of these three sequences lead to end state I A, which is representative of an OR gate. Using the min-max approach, the minimum HCLPF value is chosen because of the "OR gate" This is a conservative approach to evaluating the seismic CET and is used for the AP600 risk-based seismic margins analysis in order to simplify the evaluation.
Once the seismic CET top event HCLPFs and the entry HCLPF into the seismic CET are determined, the last step of the containment performance seismic evaluation is performed. The seisnde CET quantification follows the same process as described in Section 2.3 for the core damage seismic evaluation.
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NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 2 Table 2.3-1 EXAMPLE SEISMIC ACCIDENT SEQUENCE QUANTIFICATION OF EVENT TREE SIIOWN IN FIGURE 2.3-1 Secuence No. Accident Sequence Ouantification Seouence HCLPF 3 RH
- R = 0.70g
- 0.76g 0.76g 4 RH
- I = 0.70g * (0.76g + SE-3) 0.76g + (0.70g
- SE-3) 5 RH
- AD = 0.70g
- 0.68g 0.70g 7 RH
- R = 0.70g
- 0.52g
- 0.76g 0.76g 8 RH
- I = 0.70g
- 0.52g * (0.76g + 5E-3) 0.76g + (0.70g
- SE-3) 9 RH
- AC = 0.70g
- 0.52g
- 0.75g 0.75g 10 RH
- AD = 0.70g
- 0.52g
- 0.68g 0.70g 12 B
- RH = 0.62g
- 0.70g 0.70g I
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NRC REQUEST FOR ADDITIONAL INFORMATION 1.g..
I Response Revision 2 l
- -- --- w. ..,te tic ,,, _ ,,,,,
Rettery ete wie
- * *
- D*w *** w- weeterm
'"l***'** weetion OH o l t e , , , , , , ,,,,, n.wa name co m A me seww. To'm *notten IE B R1 A24 4H CM AD AC l A i
een 4
eos 1 #
41 Cl24 2 OK "C ' " 3 3BL i 4 3EE "3 5 1B a oc RHR1 Decent y gg
- 8 3EE AC M 6 3ER f
A% 10 in BATT 12 1C Figure 2.1-1 Example Loss of Offsite Power Event Tree 720.158(R2)-14 W
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Lo.. of p ..sv. ca. Autom.t,e Accum. m.cere.
~" :
.... . .>.~.- ......-
IE B AH CM AD AC i A 1 OK 2 OK AECIAC 3 3BL I 4 3EE AOS $ Sg 6 OK AHA1 AECIAC 7 3DL S 3EE ACCW 9 3EA Cur AD5 10 is 11 OK BATT AHR2 qp qc Figure 2.1-2 Example Loss of Offsite Power Seismic Event Tree 720.158(R2)-15 ;
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 siSTEM Q SEIEWIC FAULT TREE
/N n
i I I I Y occurm Z occurs A and B and C I occure occurs 0.62g 0.91g 0.86g
(~ 3 l 1 A occurs 5 occure C occurs O Mg O 62g O.913 Figure 2.2-1 Example Seismic Fault Tree 720.158(R2)-16 W-Westinghouse I
NRC REQUEST FOR ADDITIONAL INFORMATION n:t :
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Responso Revision 2 .[
PAHA System Fella SeismlCatfy
/\
f%
i l I I
PAHA Heat PAHA Piping Containment Loss 1E pAHA Valves Exchangers and Supports Buildino DC Power
" FaiI Fa1I Faile to Open AOva O.65g O.95g 0.55g
/\ \
,, n I l Containment I nter i or- of DC SwGA Os tt er- l es /
Shell Cont 31 rynent Faile Racks Falls Falle Fal1 l 0.95g 0.7c 0.5g 0.7g Figure 2.2-2 Example Passive RHR Seismic Fault Tree 720.158(R2)-17 W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION tr ~a
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~i Response Revision 2 i
HCLPF n E2g D 70g C.52g o.eng D 7So D.769 0 780 N OD
- SE- 2
'* 4 3EE A05 5 1B 6 OK AHR1 AECIAC 7 3BL
'C 9 3EE ACCT.N 9 3EA CMT ADS 10 18 11 OK BATT nwet2 1C 12 Figure 2.3-1 Example Seismic Event Tree 720.158(R2)-18 W-Westingh00Se
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Response Revision 2
3.0 CONCLUSION
Re risk-based seismic margins analysis methodology presented in this RAI response is specifically for use on AP600. The methodology meets the PRA-based seismic margins requirement as specified in SECY-93-087 and addresses the NRC seismic margins questions posed in several RAls.
The result of the risk-based seismic margins analysis will be a listing of the seismic sequences leading to a core damage end state for each seismic event tree. The results provide a list of the seismic sequences leading to a containment failure (release category) end state for the seismic containment event tree.
4.0 REFERENCES
- 1. SECY-93-087, Policy. Technical and Licensine Issues Pertaining to Evolutionary and Advanced Licht-Water Reactor (ALWR) Desiens, July 1993.
- 2. Attachment to "AP600 Implementation Report for Regulatory Treatment of Nonsafety-Related Systems (WCAP-13856)," Westinghouse letter number ET-NRC-93-3974/NSRA-APSL-93-0356, dated September 24,1993.
- 3. AP600 Probabilistic Risk Assessment, Westinghouse Electric Corporation and ENEL, DE-ACO3-90SF18495, June 26,1992.
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l NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 ATTACllMENT 2 to RAI 720.158(R2)
APPENDIX H are envekiped by the site interface requirements defined in Section 2.5 of the AP600 Standard Safety Analysis Report (Reference 3). If characteristics are outside the SEISMIC MARGIN ASSESSMENT range discussed in Section 2.5, then a site-specific risk-based scismic analysis may be necessary. It is anocipated that u evaluauon win be necessary fm site H.1 METHODOLOGY SELECTION liquefaction potential as long as the site foundations have qm e w static hg capaciys The scismic margin plant examination is made based . ,
. . .. considering the safe shutdown carthquake. This will on established criteria, design specifications, existing ensure a reserve margin that exceeds a 0.5g seismic qualification test repons, established basic design gg charactenstics and configurations, and public domain A qualification seismic review of the design will be generic data. Vendor-specific evaluations are not performed with the purpose ofidentifying vulnerabilities E" ""#
and confirming assumptions made in this generic seismic Scismic margins methodology is employed to g pg, , g g9 identify potential vulnerabilities and demonstrate seismic walkdown will be performed with the purpose of margin beyond the design level Safe Shutdown identifying differences of the as-built from design and Earthquake (SSE). The caprity of those components ensuring vulnerabilities were not created.
required to bring the plant to a safe, stable condition is A seismic margin evaluation is used to identify assessed. The structures, systems, and components areas where special design considerations are appro-identified as important to seismic risk are addressed.
priate For example, additional design and mounting A review level earthquake equal to 0.5g has been requirements may be needed to provide adequate reserve established for the seismic margin assessment, and is scismic capacities for components.
used to demonstrate margin over the safe shutdown Two approaches are used in this assessment for t.arthquake of 0.3g. This review level earthquake was computing liCLPF (high confidence of low probability chosen to be consistent with the upper (0.5g) bin level of failure). One is fragility analysis and the other is the established by the USNRC (References 1 and 2). conservative deterministic failure margin (CDFM). He The seismic margin canhquake that is used is based liCLPF is defined using the log-normal statistical on the NUREG/CR4K)98 median shape spectrum distribetion and associated median capacities and anchored to 0.50g peak ground acceleration (pga). His standard deviations when fragility analysis is used. De spectmm is related to that used for seismic structural HCLPF is also conservatively defined using seismic analysis. Since the evaluation is performed for a genene acceleration design levels given in AP600 design site, both rock and soil foundation condiuons are h uons fm @y componem dich is considered. This median shape spectrum ts used since Reference I states that "the scismic margin evaluations CDFM edolm should utilize the NUREG/CR-0098 median rock or soil spectrum anchored at 0.5g ". There will be no necessity to perform a new site-specific risk-based seismic analysis by an applicant for a combined operating license as long as the site-specific parameters l
l 720.158(R2)-20 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION g.... 7 Response Revision 2 1 H.2 SEISMIC MARGINS METHOD Table 3.2-3 (Reference 3) , AP600 Classification of Components, Equipment and Systems, provides the seismic categories of the non-piping items in the plant H.2.1 Review of Plant Information (valves, pumps, etc.). The Piping and Instrumentation Diagratn Legend GW-M6-001 provides the design codes The assessment uses the following plant for the plant piping. See the Line Numbering Example in nnadon:
and the Line and Valve / Damper Specifications Table in GW-M6-001. The design code shown in the table
+ Structural design and seismic design criteria and nelu s A ec. ass, ng w RC procedures Quality Group. Paragraphs 3.2.2.3, 3.2.2.4 and 3.2.2.5
+ Structural design calculations e A tan ty AnMysis k pon
+ Structural and equipment design methodology (Reference 3) correlate the NRC Quality Groups, AP600
+ Design and equipment specifications Equipment Classes, seismic categones and ASME
+ Generic fragility data m class with one amthec
. AP600 plant response spectra.
H.2.3 Analysis of Structure Response H.2.2 System Analysis The purpose of a fragility analysis is to define the Section 7.4 of the AP600 Standard Safety Analysis maximum limit f functi nal capability or operability Report (Reference 3) provides a discussion of'the w a a nanam y orp compmen s and systems required for safe shutdown. The structures and structures that could have an effect on safe shutdown of components associated with these systems are considered the plant following an extemal event. Fragility is the in the seismic margin assessment el celuadon msulthg b fme ne ped pu Paragraph 3.2.1.2 of the AP600 Standard Safety "" **
Analysis Report (Reference 3) explains the in most
"'"
- 8 *" *
- E " *"I I 'Y' r oc m. aUur cmld acur de to loss of a cases, safety-related items are also seismic Category I pr ssure boundary, significant inelastic deformation, items. The paragraph also states that when portions of panial c llapse, or a combination of failure modes.
systems are identified as seismic Cavgory I, the boundaries of seismic Category I ponions of the system are shown on the piping and instrumentation drawing of that system.
AP600 Equipment Class ASME Sect. NRC Quality Seismic Category ill, Class Group A 1 A 1
(
B 1 B 1 ,
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NRC REQUEST FOR ADDITIONAL INFORMATION
.@ 9E i-Response Revision 2 A fragility evaluahon is made for the key structures Ah = HCLPF capacity in terms of ground and components (Section H.2.2) using information as acceleration level discussed in Section H.2.1. The high con 0dence of low Am' = Median capacity probability of failure for the equipment and structures r = Standard deviation associated with addressed in this submittal are established based on one randomncss of the following: u = Standard deviation associated with uncertainty
+ Fragility analysis
+ Conservative deterministic failure margin based on The conservatisms and vanability identined and design specification seismic levels considered in this assessment are associated with stress
. Generic fragility data (References 4 and 5) and strength margin factors. These factors are used to derme mean peak seismic ground capacities. The mean Fragility analysis is the primary method and is capacity factors are converted to median values having discussed below. The stress and seismic enteria are log-normal statistical relationships.
used, along with the design configuration, to define The AP600 design is based on Reg. Guide 1.60.
stress allowables, areas of conservatism and margin Seismic margin methodology as dermed in Reference 1 factors in the design. This information is used to uses the ground response spectra given in perform the fragility analysis to derme the HCLPF levels NUREG/CR-0098 (Reference 8).
that employ standard deviations, median capacity, and The mean peak seismic margin factor is related to the log-normal distribution. the stress and strength design margin factors by the There are many sources of conservatism and following expression:
variability in the estimation of seismic ground acceleration capacity. A probabilistic approach is Am = (Q X, )Ao appropriate for evaluation purposes. Instead of using probabilities directly to assess the capacity of structures, where, lower bound estimates of the seismic capacity are denved from median capacity using formulas based on Am = Mean peak seismic ground capacity log-normal distnbution for the seismic margin approach. X, = ith design margin factor Seismic capacity levels are dermed by the value R = Product notation associated with the probabilistic value identined by the Ao = Nominal peak seismic ground capacity HCLPF. This value reflects a 95 percent confidence (probability) of not exceeding a five percent probability Stress and strength margin factors that are of failure (Reference 6). appropriate for the component or structure are defined.
An expression for the HCLPF capacity (Ah) is The variability of these factors are represented in the defined (References 6 and 7) using the log-normal evaluation using standard deviations representing distribution: randomness ( r) or uncertainty ( u). These standard deviations are related to a composite uncertainty, c.
Ah = Am'
- e t -i op*M pg = p,2 + u 2
1 l
where, j l
I i
I 4
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NRC REQUEST FOR ADDITIONAL INFORMATION i
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' Response Revision 2 The various design margin factors considered are is buckling, the material strength factor is unity. ,
discussed below. The failure mode must be defined to establish the correct material factor.
- Stress Interaction
- Inelastic Response This factor reflects the margin between the acci ,
stress state and the code-mandated allowable stress A large amount of energy is absorbed by inelastic for the safe shutdown earthquake level. structural response. The structum or system is capable of performing its function even though it is
+ Criteria responding in an inelastic range. The following statements are made in Reference 10, page 34 Cc41e-mandated allowable limits include factors of concerning this phenomenon:
safety. An example is shear walls. In Reference 9, acceptable methods are given for justifying " Numerous observations of the actual performance increased capacity that represent the conservatism in of structures subjected to seismic motions have the design codes. The procedures given in demonstrated the capacity of structures to absorb Reference 9 are used to define median ultimate and dissipate much energy when strained in inelastic shear strength of concrete shear walls. Standard response. The energy absorption obtained from a deviations associated with the variability are given. linear elastic analysis performed to the design or yield level is only a fraction of the total energy
. Material absorption capability of a structure. Unless corrected for inelastic-response capability, a linear This factor represents the ratio of the average elastic-response analysis can not account for the strengths to the minimum material strengths. inelastic energy absorption capacity of a structure." i Minimum values of material properties (yield stress) are used to specify lower bound code allowables. Relationships between ductility factors (p) and Actual values of material yield stress and ultimate median margin factor (F,) are given in Reference 1I stress are higher than those used to define the which are dependent on the natural frequency range design allowable stmss for the steel used in of the structure. For the frequency range from 2 to structures and the reinforcement steel in concrete. 8 bz, the equation is:
Further, the dynamic yield stress, not considered in the calculation of the allowable stress,is higher than F, = (2p - 1)"
static yield stress used to establish the allowable.
Large variability in concrete compressive strength When the dynamic response is in the rigid range, can be expected. To achieve the mininuun specified the equation ts:
compressive strength, the concrete m'.x is specified such that the test cylinders after 2S days will have F, = p" ". .
a strength in excess of the minimum compressive ,
strength used in design. The compressive strength In Reference 12, page 4-39, hysteretic loop associated with concrete increases with age. " pinching" effects on the actual inelastic energy Further, the material strength factor is dependent on dissipation are discussed. Actual inelastic energy the failure mode. If the failure mode of a member dissipation can be less with severe pinching behavior. !
720.158(R2)-23 W-Westinghouse l
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11RC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Tlus potential effect that can reduce energy loss is provide a lower scismic response; therefore inherent considered m the seismic tnargin evaluadon. Tbc conservatisms exist in the seismic loads that have adjusted ductility factor (F,) in terms of the mean been developed using the lower damping values for ductility factor (F" ) is given below: design.
F,= 1 + CgF", -1) H.2.4 Evaluation of Seismic Capacities of Components and Plant where Table H-1 provides the seismic fragility data for the Cn = 0.6 equipment, structures and systems considered in the seismic margin evaluadon. Also shown in this table is
. Redundancy the basis used for the assessment of each item (AP600-specific or genenc). All of the liCLPF values are above Addinonal design margin exists in indetenninate the review level earthquake (0.50g).
structures due to load redistnbution. As portions of in the design of the AP600, careful consideration is the structure reach their yield and plastic moment given to those areas that are recognized as important to lunits, loads are redistributed through other ked plant seismic risk. In addition to paying special paths. Failure of a part of a system does not attention to tbose entical components that have 11CLPF necessarily mean the loss of a system or structure to values close to the review level earthquake, the design perform its intended function. This represents a process considers potential interaction with both safety-significant reserve strength for many systems and related and nonsafety-related systems or structures, as structures allowing scismic excitations that are much well as adequate anchorage load transfer and structural higher than the estabhshed safe shutdown ductility. The seismic margin evaluanon provides a earthquake. means of identifying specific equipment or structures which are vulneraNe to beyond design basis seismic
. Analysis events.
A conservadve approach is used to deline the Conservausms exist in the methods employed to llCLPF for the equipment. Qualified testing and analyze structures and equipment. Design margin analysis levels associated with the same or similar exists in the seismic qualification that is associated equipment are used with small variability to defirv the with the analysis itself. Conservative analydcal llCLPF level. The minimum margin factors above test techniques are employed to avoid costly and long levels are used to conservatively account for both analysis approaches in the finite element analyses as structural integrity and operability requirements. There well as boundary .;onditions. The manner in which is additional reserve strength existing for these items.
the loading is defined or applied to a structure can Acuve valves are grouped according to size and introduce conservatism. type. Grouping consists of gate / globe or butterfly with the operator being motor, air, pneumatic or solenoid.
. Damping The valve capacides are evaluated assuming that the valves are line-tnounted. Given that the valves in each Damping values used to design and qualify group are subject to various seismic levels, the capacity ;
structures are lower than actual damping that is of each valve group is calculated assuming the highest )
observed within a structure or component durtng seismic level is experienced by all of the valves in that I dynamic excitation. The higher actual damping wi)1 group. If a limited number of valves or specialized 720.158(R2)-24 3 Westingflouse
1 NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 components are subject to unusually high scismic levels, determined using the guidance in EPRI NP-6041 their capacities are determined on a case by case basis. and are compared to the review level carthquake From the assessment performed, the HCLPF seismic requirements.
acceleration levels are within acceptable limits defined by the review level earthquake (0.5g). . As pan of the walkdown, the existence of equipment in the plant whose mounting details H.2.5 Verification of Equipment Fragility could result in the equipment becoming a missile or Data seismically interacting with nearby safety related equipment dunng a seismic event will be noted.
The AP600 safety-related equipment is designed to meet the safe shutdown earthquake requirements defined
= Document the results of the evaluation and the in Chapter 3 of the AP600 Standard Safety Analysis corrective actions.
Report (Reference 3). An as-constmeted assessment confirms that the as-built safety-related equipment and H.3 SEISMIC MARGIN ANALYSIS construction details are consistent with the seismic margin study discussed in Secdon H.l. This section presents the seismic margin analysis.
The scismic margin evaluation is a two-phase An event tree (Figure H-1, discussed in detail in process: (1) performance of a generic seismic margin Section H.3.2) models the seismic-induced initiahng assessment based on the as-designed condition and events. The paths of this event tree define the accident (2) verification of the seismic margin assessment of the scenarios. The progressions for these specific scenarios as-installed conditions. are further developed in subsequent event trees Phase I focuses on demonstrating that the design of (described in Section H.3.3). Small fault trees model the nuclear island structure, safety-related equipment, the HCLPFs determined for each of the top events for and equipment supports can carry the loads induced by the event trees. These fault tree models are discussed in ;
the review level carthquake discussed in this Appendix. Section H.3.4. The HCLPFs for components are This Phase I evaluation incorporates as-specified reponed in Table H-1. The min-max approach is used equipment data and will be performed by the combined to quantify the fault tree and event tree models. The operahng license applicant. min-max approach is described in Section H.3.5. Table Phase 2 focuses on the review of the any field H-2 reports the HCLPF values determined for each event change notices frotn design. The deviant as-installed tree top event. Quantification of the event tree equipment configuration is examined for seismic sequences are shown in Tables H-3 thmugh H-7.
structural interacdon between adjacent equipment. The The analysis demonstrates that all components steps used in the Phase 2 Seismic Margin Assessment of required to maintain the plant in a safe stable state are deviant features include the following: expected to function following a seismic event of 0.5g acceleration.
. Visually examine the as-built mounting details for installed components using the guidance provided in H.3.1 Assumptions Applied to the Seismic EPRI NP-6041 (Reference 9). Margin Analysis
. Where significant differences between the Phase 1 The following assumptions are used to develop the f evaluation and the as-installed conditions are found, seismic margin logic models:
the capacity of the as-installed mounting details are 720.158(R2)-25 g
NRC REQUEST FOR ADDITIONAL INFORMATION E Response Revision 2
. The seismic event is assumed to occur while the e identical pipe segments are modeled as failing at the plant is operaung at full power. same acceleration level and at the same time.
. A review level carthquake equal to 0.5g is used for . If the safeguards signals should fail, safety-related the seismic margin analysis, air-operated valves will still perform their function on a loss of compressed air. The main steam
. The offsite insulators on the feed lines from the isolation valves and steam generator power-ope ated offsite power grid fail such that a loss of offsite valves close, air-operated valves in the passive heat power occurs. removal system and core makeup tank system open, and air-operated containment isolation valves close.
. The control rod motor generator sets are powered by ac load centers that are de-energized on loss of H.3.2 initiating Events offsite power sources. When the control rod motor generator sets are de-energized, current to the The following initiating events from the AP6(X) magnetic ja<;k mechanisms stops and the gripper Level 1 PRA are reviewed:
coils open, allowing the rods to drop into the core.
Therefore, failure of the reactor trip signal is not . Steam generator tube rupture (SGTR) modeled. . Loss of coolant accidents (LOCAs)
. Main steam line or main feed line breaks
. The nonsafety-related systems are assumed to be . Transients non-functional. . Loss of offsite power
. Loss of support systems (component cooling water,
. It is assumed that piping will fait prior to failure of service water, compressed air) associated pressure boundary valves. Therefore, . Anticipated transient without scram ( ATWS) valves that are not required to change posinon are not included. Orifices and flow elements are also The transients events are bounded by the loss of not included since it is assumed that the piping will offsite power event (reactor trip, turbine trip, loss of fail before these elements. main feedwater, loss of RCS flow). The nonsafety-related systems in the turbine building are assumed to be
. Failure of the normal residual heat removal isolation non-functional. The steam lines downstream of the main valves is not included in the analysis, since the pipe steam isolation valves and the feed lines upstream of the ,
sections are expected to fail before the valves could main feedwater isolation valves are assumed to be !
fail. These valves are closed in normal operation. ruptured.
Offsite power is lost and the nonsafety-related
. The entena established for containment integrity for support systems (diesel generators, component cooling I the AP6(X) level 1 PRA is applied to the seismic water, service water and compressed air) are assumed to l margin analysis. be unavailable. Therefore, the loss of ac power sources ,
bounds these special initiating events. l
. If the containment air path remains functional. The remaining initiating events are evaluated as passive containment air cooling is sufficient. seismically-induced initiating events. An event tree fFigure H-1) is used to show the possible initiating i
. Redundant components are modeled as failing at the events that could be induced by a seismic event. The I same acceleration level and at the same time. top events for the event tree shown on Figure H-1 are 720.158(R2)-26 W-Wes!!f!gt100Se l
i i
I I
NRC REQUEST FOR ADDITIONAL INFORMATION e m.e.m Response Revision 2 desenhed in Sections H.3.2.1 through H.3.2.10. These PRHR INTACT are not evaluated; if the next three top top ever r c: events do not fail (BUILDINGS INTACT, IRWST INTACT, and RCS COMPONENTS INTACT), then SEISMIC EVENT -- The scismic event that starts the path 34 is continued on the event tree ADVS accident progression. (anticipated transient without scram).
If the top event PIPES LNTACT fails, two failure CORE ASSEMBLY INTACT -- The core assembly branches are shown. De middle branch is to identify remains intact following a seismic event. that the pipes in the reactor coolant system have failed and the bottom branch is to identify that the main steam PIPES INTACT - The pipes inside containment remain line or feed line pipes inside containment have failed.
intact. If the next three top events do not fail (BUILDINGS INTACT, IRWST INTACT, and RCS COMPONTNTS PRHR 11X INTACT - De passive residual heat INTACT), then path 18 is continued on the event tree removal heat exchanger remains intact following the LL (large LOCA) and path 26 is continued as a main seismic event. steam line break inside containment.
If the top event PRHR HX LNTACT fails, and the BUILDINGS INTACT -- ne buildings housing safety- next three top events do not fail (BUILDINGS INTACT, related equipment remain intact following the seismic IRWST INTACT, and RCS COMPONENTS LNTACT),
event. then path 10 is continued on the event tree LIJPRHR, large LOCA due to failure of the passive residual heat IRWST INTACT -- ne in-containment refueling water exchanger, storage tank remains intact following the seismic event. The next three top events, BUILDINGS INTALT, IRWST INTACT, and RCS COMPONENTS INTALT RCS COMPONENTS INTACT -- Large equipment such are addressed on the success and failure paths of the nrst as the' steam generators and reactor vessel remain intact three top events. If top event BUILDINGS INTACT, following the seismic event. fails, core damage with containment failure is assumed (CD/CF). If either of the top events IRWST LNTACT NO SMALL LOCA - A consequential small LOCA or RCS COMPONENTS INTACT fails, core damage is does not occur following the seismic event. assumed, and the next two top events (AIR BAITLE INTACT and ISOLATION VALVES CLOSE) are AIR BAITLE INTACT -- The contamment air haffle addressed. If neither of these top event fails, the path is remams intact following the seismic event, assigned as CD (core damage). If the top event AIR BAFFLE INTACT fails, core danuge with containment ISOLATION VALVES CLOSE - Successful closure of failure is assigned (CD/CF), if the top event containment isolation valves. ISOLATION VALVES CLOSE fails, then core damage with failure of containment isolation is assigned The first three top events (CORE ASSEMBLY (CD/CI). If no seismic induced failures occur, then a INTACT, PIPES INTACT, PRHR INTACT) consequential small LOCA (NO SMALL LOCA) is determine the outcome of plant response to seismic addressed, if a small LOCA does not oaur (shown on events, since the tree structure for all subsequent top path 1), then the path is defined as MSBO (main steam events is identical. If the first top event, CORE break outside of containment) and this path is continued ASSEMBLY INTACT, fails, then PIPES INTACT and 720.158(R2)-27
NRC REQUEST FOR ADDITIONAL INFORMATION HE! !EE Response Ret ision 2 on the event tree MSBO. If a small LOCA occurs, to be a large LOCA. A fault tree model is not required path 2 is continued on event tree SL (small LOCA). to determine this HCLPF.
The high confidence of low probabihty of failure (HCLPF) values are shown on the event tree for each of H.3.2.5 Buildings Intact the top events. The HCLPFs for the individual components are shown on Table H-1 and the HCLPFs This top event is to show that the following for event tree top events are shown on Table H-2. De buildings remain intact following a seismic event:
top events for the event tree shown on Figure H-1 are desenbed in the following subsections. . Shield bmiding
+ Containment vessel .
H.3.2.1 Seismic Event . Intenor containment
+ Auxiliary building.
This top event is to identify that a seismic event occurs. Figure H-3 is the fault tree to evaluate the HCLPF for this top event. Failure of any one of these buildings H.3.2.2 Core Assembly intact is not evaluated funher and core damage with containment vessel failure is assigned.
The reactor is tripped. The core assembly (and guide tubest and control rod drive systems remain H.3.2.6 in-Containment Refueling Water functional so that the rods can fall into the core. Figure Storage Tank (IRWST Intact)
H-2 is the fault tree to evaluate the HCLPF for this top evenL Failure of this path transfers to event tree ATWS This top event is to show that the in-containment to continue the accident sequence model. refueling storage tank is intact. Sparger failure would not prevent the flow of water from an intact H.3.2.3 Pipes intact in-containment refueling storage tank. Therefore, the spargers are not included and a fault tree model is not This top event is split into three paths. The top required. Core damage is assigned if the in-containment branch is success, the second branch is to show that the refueling water sto'ar,e tank fails.
pipe breaks could be in the reactor coolant system and the third branch is to show that the pipe breaks could be H.3.2.7 RCS Components intact in the main steam line or feed line pipes. As more than one pipe break could occur, all pipe breaks in the reactor if the buildings are intact, then this top event shows coolant system are assumed to be large pipe breaks and that the following primary system components (and their this path transfers to address a large LOCA (LL). supports) maintain pressure boundary integrity:
The HCLPF applies to all pipes inside containment.
Therefore, a fault tree model for this top event is not . Reactor vessel !
required.
- Pressurizer
- Steam generators H.3.2.4 Passive Residual Heat Removal - Reactor coolant pump Heat Exchanger (PRHR HX Intact)
- Polar crane This top event is the passive residual heat removal Although the polar crane is not a reactor coolant heat exchangers are intact. The failure path is assumed system component, it is included in this top event as it 720.158(R2)-28 W~
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 is postulated to cause damage to other components if it event tree paths. However, the intact air baffle and fails. containment vessel integrity (containment isolation) are Figure H-4 is the fault tn e to evaluate the HCLPF included as additional end states on the seismic margin for this top event. If any of tl ese components fail, core event trees. If containment isolation fails on the path, damage is assigned. CI is included as part of the path identifier. CF is used to identify that the air baffle failed and containment H.3.2.8 No Small LOCA cooling is not availabic. '
This top event is that a consequential small LOCA H.3.3.1 Path MSBO does not occur. A consequential small LOCA could occur because the pressunzer safety valves open and do Path 1 of the event tree shown on Figure H-1 not close, small pipe breaks occur, or for any other transfers to the event tree MSBO, mam steam and main reason. The failure branch transfers to a small LOCA feed breaks outside containment. The buildings and event tree (SL) to evaluate the small LOCA following a reactor coolant components maintain their function.
seismic event. From tbc Level 1 PRA the probability of There are no breaks in the prunary system piping or the a consequential small LOCA is 1.0E-2. steam hne and feed line systems inside containment.
The reactor is tripped. There are no ac power sources in H.3.2.9 Air Baffle intact the plant and the steam lines and/or feed lines downstream of the isolation valves have failed. Figure This top event is to show that the air bafile between H-5 shows the event tree that continues the accident the contaimnent vessel and the containtnent shell remains sequence model of MSBO, defined on path 1. The top functional. A fault tree model is not required. Failure events for tree MSBO are:
of this top event is not modeled further and containment vessel failure and core damage are assigned to this path. IE -- This is simply the entry into the tree from path I of Figure H-1.
H.3.2.10 Isolation Valves Close DC -- DC power system intact.
This top event is closure of the isolation valves to maintain containment vesselintegrity. Evaluation of the PMS - Protection and monitoring system intact.
HCLPF for this top event are discussed in Subsection H.3.4.
VALVES CLOSE - Intact main steam lines, feed lines and steam generator blowdown lines such that the valves can close and isolate the reactor coolant system.
H.3.3 Event Tree Models Continued The accident progression defined on the paths of the NO SGTR -- This event checks that a steam generator event tree shown on Figure H.1 can transfer to event tube rupture does not occur as a consequence of the trees as defined by the path identifiers. He transfer steam line break.
trees and their top events are described in the following sections. The descriptions of the systems and fault tree PRHR -- Passive residual heat removal system intact.
logic models that define the top events are described in Subsection H.3.4. The end states shown on the event CMT - Core makeup tank system intact.
trees are the same as those defined for the Level 1 PRA 720.158(R2)-29 l
w NRC REQUEST FOR ADDITIONAL INFORMATION e .-
i ]
Response Revision 2 ADS - Automatic depressurization system intact. operators must activate the automatic depressurization system and the accumulators ( ACC) must inject to allow ACC - Accumulator intact. gravity injection and recirculauon.
If either de power (DC) or the protection and GRAVITY INJECT - Intact gravity injection from the monitoring system (PMS) is not available, then systems in-containment refueling water storage tank. carmot be automatically activated. However, the valves in the passive residual heat removal system and valves RECIRC -- Valves required for recirculation are intact. from the core makeup tank fail open so that these systems are available. The isolation valves fail closed AIR B AFFLE - Air baffle intact. on loss of power or instrument air. If the valves close, then the plant can be maintained in a safe, stable C1 -- Contmnment isolation valves intact, condition. If the valves do not close but a steam generator tube rupture does not occur, then the plant can Availabihty of de power (DC) and the protection be maintained in a safe, stable condition.
and monitoring system (PMS) are addressed first on Figure H-5. If both de power and PMS are available H.3.3.2 Small LOCA Path and the main steam lines, feed hnes and steam generator l blowdown lines are isolated (VALVES CLOSE) and the Path 2, identified as SL, on the initiating event tree passive residual heat removal system (PRHR) is (Figure H-1) is to show the evaluation of a small LOCA.
activated, then safe shutdown of the plant can be This path is continued on Figure H-6. The top events achieved and maintained even if a steam generator tube for the small LOCA event tree are:
rupture occurs. If the isolation function fails, then the top event steam generator tube rupture (NO SGTR) is SL -- This is simply the entry into the tree from path 2 addressed. If there is not a consequential steam of Figure H-1.
generator tube rupture and passive residual heat removal (PRHR) is available, then core damage does not occur. DC -- DC power system intact.
If the passive residual heat removal system is not available, then injection from the core makeup tanks PMS - Protection and monitoring system intact.
(CMT), operahon of the automatic depressurization system (ADS), injection from the in-containment PRHR -- Passive residual heat removal system intact.
refuelmg water storage tank (GRAVITY INJECT) with recirculation (RECIRC) are required. If injection from CMT - Core makeup tank system intact.
the core makeup tanks is not available, then the operators must activate the automatic depressurization ADS -- Automatic depressurization system intact.
system and the accumulators (ACC) must inject to allow gravity injection and recirculation. ACC -- Accumulator intact.
If the isolation function fails and a steam generator tube rupture occurs, then injection from the core makeup GRAVITY INJECT - Intact gravity injection from the tanks (CMT), operation of the automatic depressurization in-contamment refueling water storage tank.
system (ADS), and injection from the in-containment refueling water storage tank (GRAVITY INJECT) with RECIRC - Intact valves required for recirculation.
rectreulation (RECIRC) are required. If injection from the core makeup tanks is not available, then the AIR B AFFLE -- Air baffle intact.
720.158(R2)-30 3 WeStirigh00Se i
NRC REQUEST FOR ADDITIONAL INFORMATION w
Response Revision 2 ;
l CI -- Containment isolation valves intact. GRAVITY INJECT -- Intact gravity injection from the j in-containment refueling water storage tank.
The failure of de power and the protection and j monitoring system would be the same as shown on RECIRC -- Valves required for recirculation intact. l Figure H-5. For a consequential steam generator tube '
rupture combined with a small LOCA, the sequences AIR BAFFLE - Intact air baffle.
would be the same, except that injection must continue l with and without isolation of the steam generators. C1 - Containment isolation valves intact. l Thus, only the small LOCA event is shown on j Figure H-6. Decay heat removal and injection and H.3.3.4 Anticipated Transients Without l recirculation are required for mitigation of the small Scram LOCA.
Path 34 of Figure H-1 is continued on Figure H-8 to H.3.3.3 Large LOCA from Passive model the accident sequence of an anticipated transient Residual Heat Removal Break without scram. Because offsite power is postulated to
- have been lost, the control rod motor generator sets ,
Failure of the passive residual heat removal heat would be de-energized even if the reactor trip function exchanger is postulated to occur because of a large failed. If the core assembly or the control rod system break. Path 10 of Figure H-1, identified as LilPRHR, failed, the rods are postulated to fail to insert into the j is continued on Figure H-7. Note that the initiation core. 'Ilus is not a realistic scenario, as the HCLPF for event HCLPF of 0.65g is greater than the HCLPFs for the initiating event is 0.97g, while the HCLPFs for ;
buildings or fuel in the vessel. Rus..this is not a failure of the passive residual heat exchanger (0.65g), !
realistic accident sequence but is retained for failure of buildings (0.62g), and failure of the fuel in the completeness. Figure H-7 is the event tree that shows vessel (0.54g) are all less than 0.97g. However, this the accident sequences of a large LOCA. De passive accident sequence is continued on Figure H-8 for residual heat removal system is not required. The reactor completeness and the top events are-coolant system is depressurized so that the automatic !
depressurization system is not required. The top events ATWS - This is simply the entry into the tree from path ;
for the large LOCA event tree are: 34 of Figure H-1.
LL-PRHR -- This is simply the entry into the tree from DC - DC power system intact.
path 10 of Figure H 1. i I
PMS - Protection and monitoring system intact.
DC -- DC power system intact. ,
PR -- Ris top event is pressure relief with the i PMS -- Protection and monitoring system intact. pressurizer relief valves. A probability of 0.24 is ,
assigned for unfavorable exposure time when sufficient j CMT - Core makeup tank system intact reactivity feedback is not present. I ACC - Accumulator intact. PRHR -- Passive residual heat removal system intact.
CMT - Core makeup tank system intact.
720.158(R2)-31 W-Westinghouse ,
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-j NRC REQUEST FOR ADDITIONAL INFORMATION I asarmi ,
Response Revision 2 l W-ADS - Automahc depressurization system intact. H.3.4.1 DC Power System (DC)
ACC - Accumulator intact. The switchgear for the four safety-related 125 vde switchgear buses (in the auxiliary building) supply the GRAVITY INJECT - Intact gravity injecnon from the integrated protection and control system and plant in-containment refueling water storage tank. protection equipment. De 125 vdc class IE switchboards supply power for the 120 vac distribution ;
RECIRC -- Valves required for recirculation intact. system,125 de main distribution system (for reactor trip switchgear, air-operated solenoid valves and reactor AIR B AFFLE -- Intact air baffle. coolant pu'ap trip switchgear), and the motor control centers which supply the de motor-operated valve loads.
C1 -- Containment isolation valves intact. For the seismic evaluation, it is assumed that ac power is not avaliable. Herefore, the battery chargers are The consequential steam generator tube rupture and modeled as being unavailable so that de power is only closure of the isolation valves for the main steam line supplied by the batteries.125 vde requires:
break are not included on this event tree since the sequences would not change significantly. . 125v 60 cell batteries
- 125 vde switchboards H.3.3.5 Other Pipe Break initiating Events - 125 vdc distribution panels
- Motor control centers The pipe breaks addressed by the top event, Pipes . Inverters (120 vac)
Intact, are mainly large breaks (large LOCAs). He
- 120 vac distribution panels large LOCA event tree model shown on Figure H.7 would apply, if the break is postulated in the The de power system is modeled as also requiring:
accumulator system, the top event ACC would be remov'ed. If the break is postulated in the core makeup
- Cables (and cable trays) tank system, the top event CMT would be removed. . Transfer switches If the steam line or feed line break is inside . Battery racks ,
containment, the event tree model would address a '
consequential steam generator tub- rupture with the Figure H-9 show the fault tree model to evaluate the isolation valves closed as well as with the isolation HCLPF for this top event.
valves failed. This increases the number of event tree sequences but does not change the HCLPFs on the H.3.4.2 Protection and Monitoring accident sequence paths. System (PMS) .
H.3.4 Accident Sequence Top Events The protection and monitoring system requires the following cabinets:
The descriptions of the top events used for the event tree models described in Subsection H.3.2 are provided . ESFAC cabinets here. . Protection logic cabinets
= Integrated protection cabinets
- Multiplexer cabinets t
720.158(R2)-32 W Westinghouse i
NRC REQUEST FOR ADDITIONAL INFORMATION w m:
Response Revision 2 q The actuation circuitry, boards, cards and software The main steam line pressure transmitter is included are not evaluated separately for the seismic margins to model actuation of the valves. Figure H-11 is the analysis. These components are modeled as being part fault tree to evaluate the HCLPF for this top event.
of the cabinets. The field signals are wired directly from the sensor, transmitter, switch, to the input / output H.3.4.4 No Steam Generator Tube termination boards. The reactor trip switchgear and Rupture (No SGTR) reactor trip breakers are not required for the seismic margin analysis bemuse failure of ac power is assumed. The top event, No SGTR, is to check that a steam Figure H-10 is the fault tree model for this top event, generator tube mpture does not occur following a main steam line break. The failure path denotes the H.3.4.3 Valves Close probability of a steam generator tube rupture following a main steam line break in which blowdown of both This top event is successful closure of isolation steam generators occur. The probability assigned to this valves following a steam line break or steam generator top event the AP600 Level 1 PRA is 1.6E-2.
tube rupture.
Valves that must close for a steam line break are: H.3.4.5 Passive Residual Heat Removal
. Main steam iselation valves SGS V(M0A and V(MOB During normal operation the passive residual beat Main feedwater isclation valve 50S VG57A and removal heat exchangers are full of cold water and are V057B at reactor coolant system pressme. The heat exchangers Steam generator power-operated relief valve block are submerged in the in-containment refueling water valves SGS V027A and V027B storage tank. The line from the passive residual heat removal heat exchangers to the cold leg is blocked by Valves that must close for a steam genemtor tube two normally closed, fail open, parallel air-operated rupture are: valves. The line from the reactor coolant system hot leg to the top of the passive residual heat removal heat Steam generator pov cc-operated relief valve block exchangers is normally open. If the steam generators are valves SGS V'J27A and V027B unavailable for heat removal or injection from the core Main steam isolation valves SGS V(M0A and makeup tanks is activated, the passive residual heat V040B removal heat exchangers are actuated. He actuation
. Main feedwater isolation valves SGS V057A and signals automatically open the normally closed valves in V057B the line from the passive residual heat removal heat Steam generator blowdown isolation valves exchangers to the reactor coolant system cold leg. The SGS V074A and V074B heat exchangers transfer heat from the reactor coolant Steam generator blowdown isolation valves system to the water contained in the in-contaimnent SGS V075A and V075B refueling water storage tank by circulation of reactor Main feedwater control valves SGS V250A and coolant systern fluid from the hot leg to the cold leg side V250B. of the steam generator channel head. He water in the in-containment refueling water storage tank provides a heat sink, if a LOCA should occur, the reactor coolant system pressure decreases and initiates a reactor trip and 720.158(R2)-33 l
NRC REQUEST FOR ADDITIONAL lNFORMATION Response Revision 2 safeguard actuation. If the loss of water inventory from . PXS-V015A and V015B the reactor coolant system is sufficient to reduce the . PXS-005A and 005B pressuruer level to the core makeup tank actuation . CV PXS-016A and 016B setpoint, core makeup tank water is injected into the . CV PXS-017A and 17B reactor vessel. When core makeup tank level decreases to a low level setpoint, the automatic depressurization PXS-V002, PXS-V003, PXS-V014 and PXS-V015 A system is activated. Reactor coolant system and B each fail open upon loss of de power or depressurization is assisted by heat transfer in the compressed air. Motor-operated valves PXS-005A and passive residual heat removal heat exchangers. The PXS-005B are normally open. A pressurizer pressure components required for passive residual heat removal sensor and pressurizer level transmitter are also included.
are: Figure H-13 is the fault tree to evaluate the HCLPF for this top event.
. PXS-V108A This fault tree was evaluated with the design
. PXS-V108B changes documented in References 13 and 14. These
. PXS-V101 changes do not have an impact on the HCLPF obtained from the fault tree.
PXS-V108A and PXS-V108B fail open on loss of either compressed air or de power. Motor-operated H.3.4.7 Automatic Depressurization valve PXS-V101 is open, but must remain open dunng System (ADS) the seismic event. Although there are various signals that will actuate passive residual heat removal. The automatic depressurization system (ADS) is representative transmitters are included in the fault tree used to fully depressurize the reactor coolant system.
model. These transmitters are the steam generator (SG) Signals to actuate automatic depressurization system narrow range transmitter, the steam generator wide range occur when low core makeup tank water level is sensed level transmitter, and the passive residual heat removal by temperature transmitters. H ere are four heat exchanger flow tmnsmitter. Figure H-12 is the fault depressurization stages in the automatic depressurization tree to evaluate the HCLPF for this top event. system.125 vde and 120 vde are required to actuate the valves. Air is provided by backup air bottles sufficient H.3.4.6 Core Makeup Tank (CMT) for at least a full opening and re-closing of the air-operated valves.
Injection from the core makeup tanks is actuated upon receipt of a low-2 pressurizer level signal, a The valves in the ADS system are:
safeguard actuation signal, or a low steam generator water level coincident with high reactor coolant system . RCS-PL-V001 A, RCS-PL-V001 B, RCS-PL-V001C, hot leg temperature. The air-operated valves open to RCS-PL-V001D align the core makeup tanks to the reactor coolant . RCS-PL-V002A, RCS-PL-V002B, RCS-PL-V002C, system. The check valves (CV)in the system must also RCS-PL-V002D open. He required components are: . RCS-PL-V003A, RCS-PL-V003B, RCS-PL-V003C, RCS-PL-V003D
. Tanks PXS-02A and 02B . RCS-PL-VONA, RCS-PL-VONB, RCS-PL-VOMC, a PXS-V002A and V002B RCS-PL-V004D
- PXS-V003A and V003B
+ PXS-V014A and V014B 720.158(R2)-34 3 Westinghouse
i NRC REQUEST FOR ADDITIONAL INFORMATION 2._
Response Revision 2 3_1 The representative success enterion used for the depressurized before the accumulators will inject. De seismic margins fault tree is three out of four lines of components are:
stages two and three must open or one out of two stage four and one out of six stage one, two and three lines . Tanks PXS-MT-01 A and PXS-MT-01B must open. . CV PXS-028Aand PXS-028B Figure H-14 is the fault tree to evaluate the HCLPF . CV PXS-029A and PXS-029B for this top event. . PXS-027A and PXS-027B If an operator acdon is required to initiate the system, the following components are included in the Figure H-16 is the fault tree to evaluate the HCLPF fault tree: for this top event.
. Qualified data processing system cabinet H.3.4.9 Gravity injection (Gravity inject)
. Quali0ed data processing system and main control room display If the reactor coolant system is depressunzed by
. Main control room supervisory operator station either a LOCA or the automatic depressurization system,
. Main control room switch stadon gravity injection from the in-containment refueling water
. Operator action storage tank is initiated. He components required are:
Figure H-15 is the fault tree to evaluate the HCLPF . CV PXS-122A and PXS-122B for this top event if an operator action is included (such . CV PXS-123A and PXS-123B as for event tree paths where the core makeup tanks . PXS-121 A and PXS-121B fail).
The fault trees shown in Figures H-14 and H-15 PXS-121 A and PXS-121B are normally open, were evaluated with the design changes documented in Figur'e H-17 is the fault tree to evaluate the HCLPF for References 13 and 14. Rese changes do not have an this top event.
impact on the HCLPFs obtained from these fault trees.
H.3.4.10 Water Recirculation to Reactor H.3.4.8 Accumulators (ACC) Vessel (Recirc)
During normal operation, the accumulators are in the long-term cooling mode, the core is covered isolated from the reactor coolant system by two isolation and steam is released to the containment via the check valves, in series. Rese valves mamtain the depressurization valves. The water in the containment reactor coolant system pressure boundary. When the sump provides recirculation flow to the core through the reactor cochmt system falls below the accumulator direct vessel injection lines. Some of the steam that pressure, the check valves open and borated water is condenses on the containment steel shell, which is forced into the reactor coolant system by gas pressure. cooled by the passive containment cooling system. is The motor-operated isolation valve in the discharge line returned. Actuation signals to open the motor-operated from each accumulator to the reactor vessel is normally valves are: low in-containment refueling water storage open (PXS-027A and PXS-027B). He accumulator tank water level in coincidence with an S-signal.
provides sufficient inventory for short-term core cooling. Components required are:
ne reactor coolant system must be partially
. CV PXS-Il9A and PXS-Il9B i
1 720.158(R2)-35 3 Westinghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION l
\
Responso Revision 2 ,
h
. CV PXS-120A and PXS-120B . WLS-VON ,
. PXS-Il7A and PXS-Il7B - WLS-V006
- PXS-Il8A and PXS-118B . WLS-V0li [
+ WLS-V012 !
The in-containment refueling water storage tank - WLS-V055 ;
(IRWST) level transmitter is included. Figure H-18 is
- WLS-V057 !
the fault tree to evaluate the HCLPF for this top event. + VFS-V0ll ;
. VFS-V012 :
H.3.4.11 Isolation Valves Close (Cl) ,
A containment pressure sensor and transmitter, main This top event is closure of the isolation valves to steam ime pressure transmitter and pressurizer sensor maintain containment integrity. Automatic actuation of and transmitter are included. Figure H-19 is the fault !
the containment isolation signal is generated by: tree to evaluate the HCLPF for this top event. f I
+ Hi-1 containment pressure H.3.5 Evaluation of Event Tree End States I
- Low pressurizer pressure
- Low-3 reactor coolant system T,y To quantify the seismic event trees, each event tree l
- Low steam line pressure. top event is assigned a HCLPF value. The random failures and human error events used in the AP600 i
Containment penetrations were screened out of the Level 1 PRA were reviewed- There were no single i
AP600 Level 1 PRA analysis if they met one of the cutsets for the safety-related system with a failure following screening criteria' probability of 1.0E-3 or greater. One human enor ,
probability is shown on the event trees. This operator
- The penetration is connected to a closed system action is to start ADS given the CMT injection failed. .
inside containment. The human error probability of 2.2E-3 is from the ,
AP6001.evel 1 PRA. ,
+ The penetration has isoladon valves plus closed The initiating events defined for the seismic margin system outside of containment. analysis are specified as end states on the event tree l shown on Figure H-1. These initiating events are :
l
- The penetration has additional system valves inside continued on additional event trees to evaluate the containment normally closed or automatically accident sequences.
closed. Small seismic fault trees that involve multiple components are developed in Subsection H.3.4 These )
+ The penetration has at least one blind flange. fault tree models are used to evaluated the HCLPFs for i the top events of the event trees. The basic events for The following valves did not meet the above each seismic fault tree are assigned a HCLPF value from j screening criteria and must be closed for containment Table H-1. The seismic fault tree is then quantified ;
isolation: using the " min-max" approach to determine the system j (or event tree top event) HCLPF. The min-max
. CAS VNO approach is defined as:
- CAS-VN1
+ CAS-V084 . For basic events under an AND gate, all basic
+ CAS-V085 events must occur to progress up a level in the fault 720.158(R2)-36 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
? % '2 Response Revision 2 tree. Thus the highest HCLPF value is assigned for accident sequences evaluated as a pipe break (path 18 on the AND gate value. Figure H-1) defined the large pipe break event. The HCLPFs on the paths given in Table H-6 would be
. For basic events under an OR gate, any one of the similar, with Q(pipe breaks) of 0.78g replacing basic events cause progression up the fault tree. Q(PRHR-HX) of 0.65g.
Thus the lowest HCLPF value is assigned for the Evaluation of the HCLPFs for the ATWS sequences OR gate value. for the event tree shown on Figure H-13 is given in Table H-7. The HCLPF for core assembly intact is Values for event tree top events not modeled using 0.97g. This HCLPF is greater than the HCLPF for fault trees are taken from Table H-1 for the particular buildings, the IRWST, or the fuel in the vessel.
component. The event tree for path 26, MSBI (main steam line Once the event tree top event HCLPFs are breaks inside containment) would be similar to the event evaluated, the scismic event trees are quantified using tree for path 1, MSBO. This is not a realisuc sequence, the min-max approach. Combinadons of HCLPFs on a as the HCLPFs for the buildmgs, fuel in the vessel, and single accident sequence path are combined as "AND" passive residual heat exchanger are less than the HCLPF '
events. Table H-2 summarizes the HCLPF values for this imnating event (0.78g).
determined for each of the event tree top events.
Evaluadons of the HCLPFs for the paths of the H.3.6 Level 2 Analysis event tree shown on Figure H-1 are given on Table H-3.
Success paths are not included in the evaluation. The The end states are partiuoned by intact containment, failure paths of the top events are identified as containment fails (CF), or containment isolation fails Q (top eventh (CD. The conditional probability that the containment Evaluation of the HCLPFs for the paths of the event fails randomly as evaluated in the Level 2 PRA is tree shown on Figure H-5 (MSBO) are given on approximately 1.0E-2.
Table' H-4. The operator action is identified as OA. The end state defined as CF/CD with HCLPF The consequential steam generator tube rupture is of 1.28g is containment failure (because of failure of the identified as R(SGTR) and the steam generator tube air baffle) with core damage occurring at some later rupture conditional probability of 1.6E-2 is included in time. The HCLPFs for the remaining plant damage the sequences addressing a steam generator tube rupture. states are combined and the HCLPF for each end state Evaluation of the HCLPFs for the paths of the event is provided below. All end states identified as CF/CD tree shown on Figure H.6 for the consequential small have a HCLPF of 1.28g.
LOCA event are given on Table H-5. The consequential The following states have an associated random small LOCA is defined as R(Small LOCA) and the containment failure of 1.0E-2:
consequential probability of 1.0E-2 is included in these sequences. . 3BL = 1.07g, or 0.67g, or 0.78g, or 0.97g = 0.67g Evaluation of the HCLPFs for the paths of tbc event tree shown on Figure H-7 for the large LOCA resulting . 3BE = 1.07g, or 0.67g, or 1.48g, or 0.65g, or 0.78g, from rupture of the PRHR heat exchanger are given in or 0.97g = 0.65g Table H-6. The initiation event HCLPF of 0.65g is greater than the HCLPF for intact buildings or fuel in . 3BR = 1.07g, or 0.67g, or 0.78g, or 0.97g = 0.67g the vessel. Thus, this is not a realistic accident sequence but is retained for completeness. The HCLPFs for the . I A = 1.42g, or 1.07g, or 1.99g = 1.07g 720.158(R2)-37
NRC REQUEST FOR ADDITIONAL INFORMATION v.n =m
' G Response Revision 2
- 3A = 1.42g, or 1.07g. or 0.92g, or 1.48g = 0.92g . 6E/CF =1.28g, or 1.42g, or 1.48g, or 1.99g = 1.28g
. 6L = 0.67g or 1.07g = 0.67g H,4 CONCLUSIONS
. 6E = 0.67g, or 1.07g, or 1.42g, or 1.48g, or 1.99g The AP600 seismic margin analysis has
= 0.67g demonstrated that for all structures, systems, and components required for safe shutdown se high The following end states are associated with failure confidence of low probability of failure (HCLPF) of containment isolation: magnitudes are equal to or greater than 0.60g except for the fuelin the reactor vessel. A HCLPF of 0.54g was
. 3BL/Cl = 1.07g, or 0.67g, or 0.78g, or 0.97g = detennined for the fuel in the vessel.
0.67g Thus, the AP600 design exceeds the requirement set forth in Reference 2 that the plant meet or exceed a a
3BE/Cl = 1.07g, or 0.67g, or 1.48g, or 0.65g, or review level earthquake of 0.5g.
0.78g, or 0.97g = 0.65g
+ 3BR/Cl = 1.07g, or 0.67g, or 0.78g, or 0.97g =
0.67g " Procedural and Submittal Guidance for the 1.
Individual Plant Examination of Extemal Evetits
. 1 A/Cl = 1.42g, or 1.07g, or 1.99g = 1.07g (IPEEE) for Severe Accident Vulnerabilities," Final Repon, NUREG-1407, U.S. Nuclear Regulatory
+ 3 A/CI = 1.42g, or 1.07g, or 0.92g, or 1.48g = 0.92g Commission, June 1991.
- 6L/CI = 0.67g or 1.07g = 0.67g 2. "SECY-93-087 - Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced
. 6E/CI = 0.67g, or 1.07g, or 1.42g, or 1.48g, or Light-Water Reactor (ALWR) Designs," USNRC 1.99g = 0.67g Memorandum, July 21,1993, Chilk to Taylor.
The f ollowing end state s are associated with seismic 3. AP600 Standard Safety Arnlysis Report, Revision 1, failure of the air baffle or containment fadure: January,1994.
- 3BL/CF = 1.28g 4. " Handbook of Nuclear Power Plant Seismic Fragilities," NUREG/CR-3558, June,1985.
- 5. " Shutdown Decay Heat Removal Analysis of a
+ 3BE/CF = 1.28g or 1.48g = 1.28g Westinghouse 3-Loop Pressunzed Water Reactor"
- I A/CF = 1.42g, or 1.28g, or 1.99g = 1.28g
- 6. Budaitz, R. J., et al, "An Approach to the
- 3A/CF = 1.42g. or 1.28g, or 1.48g = 1.28g Quantification of Seismic Margins in Nuclear Power Plants," NUREG/CR-4334 UCID-20444,
= 6L/CF = 1.28g August 1985.
720.158(R2)-38 3 Westinghouse i
t I
NRC REQUEST FOR ADDITIONAL INFORMATION ,
'I -
Response Revision 2 I
7 Kennedy, R. P., et al, " Assessment of Seismic 11. Ravindra, M. K., R. D. Campbell, " Probability of !
Margin Calculation Methods", NUREG/CR-5270, Pipe Failure in the Reactor Coolant Loop of UCID-21572, March,1989. Westinghouse PWR Plants," Volume 3, ;
NUREG/CR-3660, February,1985. i
- 8. " Development of Criteria for Geismic Review of .
Selected Nuclear Power Plants " NUREG/CR-0098, 12. Kipp, T. R., T. A. Wesley, and K. K. Nakaki, f May,1978. " Seismic Fragilities of Civil Structures and i Equipment Components at the Diablo Canyon l
- 9. "A Methodology for Assesstnent of Nuclear Power Power Plant," Repon No. 1643.02, NTS ;
Plant Seismic Margin," Electric Power Research Engineering, September,1988.
Institute, EPRI NP-6041, Revision 1. August 1991. l
- 13. Attachment to Letter NTD-NRC-94-4065, "AP600 l
- 10. Coats, D. W., " Recommended Revisions to Nuclear Design Change Description Repon," February 1994.
Regulatory Commission Seismic Design Criteria,"
NUREG/CR-1161, May,1980, 14. Attachment to Letter NE-NRC-94-4175, "AP600 ;
Design Change Description," June 1994. !
[
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~: i NRC REQUEST FOR ADDITIONAL INFORMATION r"
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= Response Revision 2 -
Table H-1 (Sheet I of 6)
SEISMIC MARGIN HCLPF VALUES __
Description Median lieta r Beta u Beta e IICLPF Basis Value Buildings / structures Shield Building Wall 2.03g 0.17 0.21 0.27 1.09g (1)
Roof 1.07g 0.17 0.16 0.23 0.62g (1)
Containment Vessel Overturning 5.75g 0.07 0.44 0.44 2.49g (1)
Buckling 2.53g 0.05 0.15 0.16 1.82g (1)
Interior Containment 2.23g 0.17 0.22 0.27 1.18g (1)
Auxiliary Building Intenor wall 1.64g 0.18 0.24 0.30 0.82g (1)
Extenor wall 3.77g 0.17 0.24 0.29 1.92g (1)
Floor 1.71g 0.17 0.13 0.21 1.02g (1)
Corr.amment baffle 2.19g 0.16 0.16 0.23 1.28g (1) support failure IRWST 2.23g 0.17 0.22 0.27 1.18g (1)
Primary Components Reactor Pressure vessel - - - - 1.25g (2)
Support 1.44g 0.16 0.13 0.20 0.89g (1)
Integrated head package - - - - 1.25g (2)
Fuel 1.08g 0.23 0.19 0.30 0.54g (1) i 720.158(R2)-40 T Westingflouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-1 (Continued)
(Sheet 2 of 6)
SEISMIC MARGIN HCLPF VALUES Description Median Beta r Beta u Beta e IICLPF Basis Value Core assembly - - - - 1.25g (2)
(not fuel)
Control nxi drive mechanism - - - -
0.97g (2)
Pressunter support controlled Upper lateral support 1.14g 0.16 0.13 0.20 0.71g (1)
Lower suppon 1.12g 0.16 0.13 0.20 0.70g (1)
Steam generator 0.79g (2)
Column 1.18g 0.16 0.13 0.20 0.74g (1)
Upper lateml suppon 1.16g 0.16 0.19 0.24 0.66g (1)
Lower lateral support 1.48g 0.16 0.17 0.23 0.87g (1)
Reactor coolant pump - - . - 0.94g (2)
Mechanical Equipment Polar crane 1.14g 0.16 0.17 0.24 0.66g (1)
Piping 1.24g 0.11 0.17 0.21 0.78g (1) support controlled Cable trays 1.24g 0.11 0.17 0.21 0.78g (1) support controlled Heat exchanger (PRHR) 0.77g 0.103 0.0 0.103 0.65g (1)
Tank PXS-MT 1 A/B 1.35g 0.103 0.0 0.103 1.14g (1)
Tank PXS 2A/B 1.35g 0.103 0.0 0.103 1.14g (1)
(Core Makeup Tank) 720.158(R2)-41
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NRC REQUEST FOR ADDITIONAL INFORMATION !
Response Revision 2 KN -- i i
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Table H-1 (Continued) l (Sheet 3 of 6) ;
SEISMIC MARGIN HCLPF VALUES Description Median Beta r Beta u Beta e HCLPF Basis i Value Air bottles for ADS (supports) 1.24g 0.11 0.17 0.21 0.78g (1)
Vahes CVS 1/2 236g 0.103 0.0 0.103 1.99g (1)
SGS 233A/B 1.58g 0.103 0.0 0.103 133g (1) i t
SGS 40A/B 2.15g 0.103 0.0 0.103 1.81g (4) :
SGS V27A/B 0.80g 0.103 0.0 0.103 0.67g (1)
SGS 57A/B 236g 0.103 0.0 0.103 1.99g (4)
SGS 74A/B 236g 0.103 0.0 0.103 1.99g (4)
{
SGS 75A/B 236g 0.103 0.0 0.103 1.99g (4)
SGS 250A/B 236g 0.103 0.0 0.103 1.99g (4) ;
PXS 108A/B 236g 0.103 0.0 0.103 1.99g (4) ,
PXS 101 236g 0.103 0.0 0.103 1.99g (4) i PXS 2A/B 236g 0.103 0.0 0.103 1.99g (4) f PXS 3A/B 2.36g 0.103 0.0 0.103 1.99g (4) i PXS SA/B 2.36g 0.103 0.0 0.103 1.99g (4)
PXS 14/B 236g 0.103 0.0 0.103 1.99g (4) l PXS ISA/B 2.36g 0.103 0.0 0.103 1.99g (4) i PXS 16A/B 0.80g 0.103 0.0 0.103 0.67g (1) !
PXS 17A/B 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 28A/B 0.80g 0.103 0.0 0.103 0.67g (1) !
I PXS 29A/B 0.80g 0.103 0.0 0.103 0.67g (1) j 720.158(R2)-42 3 Westingflouse 1
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-1 (Continued)
(Sheet 4 of 6)
SEISMIC MARGIN HCLPF VALUES Description Median Beta r Beta u Beta e IICLPF Basis Value PXS 27A/B 236g 0.103 0.0 0.103 1.99g (4)
PXS 122A/B 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 123A/B 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 12] A/B 236g 0.103 0.0 0.103 1.99g (4)
PXS Il9A/B 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 120A/B 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 117A/B 2.36g 0.103 0.0 0.103 1.99g (4)
PXS Il8A/B 236g 0.103 0.0 0.103 1.99g (4)
RCS V001 A/B/C/D 238g 0.103 0.0 0.103 2.85g (4)
RCS V002A/B/C/D 238g 0.103 0.0 0.103 2.85g (4)
RCS V003A/B/C/D 238g 0.103 0.0 0.103 2.85g (4)
RCS VONA/B/C/D 0.80g 0.103 0.0 0.103 0.67g (4)
CAS 40/41 0.80g 0.103 0.0 0.103 0.67g (1)
CAS 84/85 0.80g 0.103 0.0 0.103 0.67g (1)
WLS 4/6 2.36g 0.103 0.0 0.103 1.99g (4)
WLS 11/12 0.80g 0.103 0.0 0.103 0.67g (1)
WLS 55/57 0.80g 0.103 0.0 0.103 0.67g (1)
VFS 11/12 0.80g 0.103 0.0 0.103 0.67g (1)
PXS 120A/B 0.80g 0.103 0.0 0.103 0.67g (1) 720.158(R2)-43
NRC REQUEST FOR ADDITIONAL INFORMATION v --@
Response Revision 2 w-Table H-1 (Continued) l (Sheet 5 of 6) {
l SEISMIC MARGIN HCLPF VALUES j Description Median Beta r Beta u Beta e HCLPF Basis i Value ,
i Electrical Equiprnent Battery 1.Mg 0.048 0.0 0.048 1.51g (4)
' 0.78g Battery rack 0.92g 0.103 0.0 0.103 (5) 125 vde distnbution panel 4.86g U.48 0.74 0.88 0.65g (3) 120 vac distnbution panel 4.86g 0.48 0.74 0.88 0.65g (3) ,
125 vde switch 5oard 4.86g 0.48 0.74 0.88 0.65g (3) 125 vde MCC 1.73g 0.N8 0.0 0.048 1.60g (3)
Inverter 6.59g 0.26 0.35 0.44 2.41g (3)
Transfer switch 0.80g 0.048 0.0 0.N8 0.74g (1)
ESFAC cabinet 1.89g 0.048 0.10 0.11 1.48g (4)
Protection logic cabinet 1.89g 0.048 0.10 0.11 1.48g (4)
Integrated protection cabinet 1.89g 0.N8 0.10 0.11 1.48g (4) switchgear Multiplexer cabinet 1.89g 0.048 0.10 0.11 1.48g (4)
ODPS cabinet 2.52g 0.048 0.10 0.11 1.98g (4) ,
MCR supervisory operator 1.00g 0.M8 0.0 0.048 0.92g (1) '
station ,
MCR switch station 1.00g 0.N8 0.0 0.N8 0.92g (1) >
QDPS and MCR display 4.97g 0.N8 0.10 0.048 3.89g (4)
CMT level switch 1.54g 0.N8 0.0 0.N8 1.42g (4)
SG narrow range transmitter 1.15g 0.048 0.0 0.048 1.07g (4) l
)
i SG wide range level 1.15g 0.048 0.0 0.048 1.07g (4) transmitter I
720.158(R2)-44 i W65tiflgh0USS l
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NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 2 Table H-1 (Continued)
(Sheet 6 of 6)
SEISMIC MARGIN HCLPF VALUES Description Median Beta r Beta u Beta e HCLPF Basis Value PRZ pressure sensor 1.54g 0.048 0.0 OM8 1.42g (4)
PZR level transmitter 1.54g 0.048 0.0 0.N8 1.42g (4)
Containment pressure sensor 2.06g 0.048 0.0 0.N8 1.90g (4) and transmitter IRWST level transmitter 1.54g 0.N8 0.0 0.048 1.42g (4)
PRHR HX flow transmitter 2.64g 0.048 0.0 OM8 2.44g (4)
MSL pressure transmitter 1.75g 0.N8 0.0 ON8 1.61g (4)
Notes: ,
(1) Design-specific basis from Westingbouse design informadon.
(2) Lower bound HCLPF values based on use of design-specific Westinghouse specification seismic design les el.
(3) Genenc data that is representa6ve of supplied equipment for Westinghouse plants. The fragility data used are conservative since they are representative of early Westinghouse plants.
(4) Fragility data based on generic seismic qualification tests performed on similar equipment.
(5) Fragility data based on quantification analyses on similar equipment.
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720.158(R2)-45 i l
VJ Westinghouse l
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NRC REQUEST FOR ADDITIONAL INFORMATION l t
- t
- .mx Response Revision 2 i M- $
i Table H '
SEISMIC MARGIN HCLPF VALUES FOR EVENT TREE TOP EVENTS Top Esent identifier Top Event HCLPF Core Assembly Intact 0.97g Pipes intact 0.78g [
PRHR HX Intact 0.65g Buildings Intact 0.62g f IWRST Intact 1.18g l RCS Components intact Oncluding fuel) 0.54g No Small LOCA Consequential Event i Air Baffle Intact 1.28g [
Isolation Valves Close 0.67g 0.65g !
de power (DC)
Protection and monitormg system (PMS) 1.48g Main steam / feed and SG isolation valves close (Valves Close) 0.67g No SGTR 1.6E-2 (consequentiaD Passive residual heat removal system (PRHR) 1.07g ;
Passive residual heat removal with de power or PMS not available 1.99g ,
Core makeup tanks (CMT) 0.67g !
Accumulators (ACC) 0.67g Automatic depressurization system, ( ADS) 1.42g i Operator action to start ADS (OA) 0.92g. 2.2E-3 (OA)
Gravity injection (Gravity Inject) 0.67g .
Recirculation (Recirc) 0.67g 1 i
Air Baffle intact 1.28g Isolation Valves Close (CI) 0.67g l
720.158(R2)-46 Westingtiouse
i NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 i
l i
Table H-3 (Sheet 1 of 3)
EVALUATION OF HCLPFS ON INITIATING EVENT TREE Path Quantification End State 1 All success paths Transfer to MSBO 2 All success paths, consequential small LOCA assumed Transfer to Small LOCA (SL) 3 Q(RCS Components Intact) = 0.54g CD 4 Q(RCS Components intaco
- Q(Isolation Valves Close) =
0.54g
- 0.67g = 0.67g CD/Cl 5 Q(RCS Components Intac0
- Q(Air Baffle Intact) =
0.54g
- 1.28g = 1.28g CD/CF 6 Q(IWRST Intact) = 1.18g CD 7 Q(IWRST Intact)
- Q(Isolation Valves Close) =
1.18g
- 0.67g = 1.18g CD/CI 8 Q(IWRST Intac0
- Q(Air Baffle intact) =
1.18g
- 1.28g = 1.28g CD/CF 9 Q(Buildings Intact) = 0.62g CD/CF 10 Q(PRHR HX) = 0.65g Transfer to Large LOCA Event Tree (LIJPRHR) 11 Q(PRHR HX)
- Q(RCS Component Intact) = 0.65g
- Q(RCS Components intact)
- Q(Isolation Valves Close)
= 0.65g
- 0.54g
- Q(RCS Components Intact)
- Q(Air Baffle Intact)
= 0.65g
- 0.54g
- 1.28g = 1.28g CD/CF l
72a m 2N7 i W Westinghouse I
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NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H 3 (Continued)
(Sheet 2 of 3)
EVALUATION OF HCLPFS ON INITIATING EVENT TREE Path Quantification End State 14 Q(PRHR HX)
- Q(IWRST Intaco = 0.65g
- Q(IWRST Intaco
- Q(Isolation Valves Close) = l 0.65g
- 1.18g
- 0.67g = 1.18g CD/CI l l
- Q(IWRST Intac0
- Q(Air Baffle Intac0 =
0.65g
- 1.18g
- Q(Buildings Intact) = 0.65g
- 0.62g = 0.65g CF/CD l
18 Q(Pipes Intact) = 0.78g Transfer to Large LOCA Event Tree (LL) 19 Q(Pipes Intact)
- Q(RCS Component intaco =
0.78g
- 0.54g = 0.78g CD 20 Q(Pipes Intac0
- Q(RCS Components Intac0
- Q(Isolation Valves Close)
= 0.78g
- 0.54g
- 0.67g = 0.78g CD/Cl 21 Q(Pipes Intac0
- Q(RCS Components intaco
- Q(Air Baffle Intac0
= 0.78g
- 0.54g
- 1.28g = 1.28g CD/CF 22 Q: Pipes intact)
- Q(IWRST Intact) = 0.78g
- 1.18g = 1.18g CD 23 Q(Pipes Intaco
- Q(IWRST Intac0
- Q(Isolation Valves Close) =
0.78g
- 1.18g
- 0.67g = 1.18g CD/CI 24 Q(Pipes Intaco
- Q(IWRST Intaco
- Q(Air Baffle intact) =
0.78g
- 1.18g
- 1.28g = 1.28g CD/CF 25 Q(Pipes Intac0
- Q(Buildmgs Intact) = 0.78g
- 0.62g = 0.78g CD/CF 26 Q(Pipes Intaco = 0.78g MSBI 27 Q(Pipes Intac0
- Q(RCS Component Intaco =
0.78g
- 0.54g = 0.78g CD 720.158(R2)-48 W Westinghouse
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NRC REQUEST FOR ADDITIONAL INFORMATION l
- =-u Response Revision 2 n-Table H-3 (Continued)
(Sbeet 3 of 3) ;
EVALUATION OF HCLPFS ON INITIATING EVENT TREE i Path Quantification End State I 28 Q(Pipes Intaco
- Q(RCS Components intac0
- Q(Isolation Valves Close)
= 0.78g
- 0.54g
- 0.67g = 0.78g CD/Cl 29 Q(Pipes Intac0
- Q(RCS Components intac0
- Q(Air Baffle Intact)
= 0.78g
- 0.54g
- 1.28g = 1.28g CD/CF 30 Q(Pipes intact)
- Q(lWRST Intac0 = 0.78g
- 1.18g = 1.18g CD 31 Q(Pipes Intac0
- Q(IWRST Intaco
- Q(Isolation Valves Close) =
0.78g
- 1.18g
- 0.67g = 1.18g CD/CI 32 Q(Pipes Intact)
- Q(IWRST Intact)
- Q(Air Bafne Intact) =
0.78g
- 1.18g
- 1.28g = 1.28g CD/CF >
33 Q(Pipes intact)
- Q(Buildings intact) = 0.78g
- 0.62g = 0.78g CD/CF 34 Q(Core Assembly Intact) = 0.97g Transfer to ATWS Event Tree ,
35 Q(Core Assembly Intact)
- Q(RCS Component Intact) =
0.97 g
- 0.54g = 0.97g CD 36 Q(Core Assembly intac0
- Q(RCS Components Intact)
- Q(Isolation Valves Close)
= 0.97g
- 0.54g
- 0.67g = 0.97g CD/CI ,
37 Q(Core Assembly Intaco
- Q(RCS Components Intac0
- Q(Air Baffle Intact) .
i
= 0.97g
- 0.54g
- 1.28g = 1.28g CD/CF l
)
38 Q(Core Assembly intact)
- Q(IWRST Intact) =
0.97 g
- 1.18g = 1.18g CD h
39 Q(Core Assembly intact)
- Q(IWRST Intac0
- Q(Isolation Valves Close) =
0.97 g
- 1.18g
- 0.67g = 1.18g CD/CI 40 Q(Core Assembly Intac0
- Q(IWRST Intac0
- Q(Air Baf0e intact) =
O.97 g
- 1.18g
- 1.28g = 1.28g CD/CF i
41 Q(Core Assembly Intact)
- Q(Buildings intac0 = .
0.97 g
- 0.62g = 0.97g CD/CF I
720.158(R2)-49
NRC RE8UEST FOR ADDITIONAL INFORMATION l
-.- g
- 1 Response Revision 2 l Table H-4 (Sheet i of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 2 Q(Air Baffle) = 1.28g CF/CD 4 Q(PRHR)
- Q(Air Baffle) = 1.07g
- 1.28g = 1.28g CF/CD 5 Q(PRHR)
- Q(Recric) = 1.07g
- 0.67g = 1.07g 3BL 6 Q(PRHR)
- Q(Recirc)
- Q(Cl) = 1.07g
- 0.67g
- 0.67g = 1.07g 3BUCI 7 Q(PRHR)
- Q(Recire)
- Q(Air Bafue) = 1.07g
- 0.67g
- 1.28g = 1.28g 3BUCF 8 QlPRHR)
- Q(Gravity inject) = 1.07g
- 0.67g = 1.07g 3BE 9 Q(PRHR)
- Q(Gravity Inject)
- Q(Cl) = 1.07g
- 0.67g
- 0.67g = 1.07g 3BE/Cl 10 Q(PRHR)
- Q(Gravity inject)
- Q(Air Baffle)
= 1.07g
- 0.67g
- 1.28g = 1.28g 3BE/CF i1 Q(PRHR)
- Q(ADS) = 1.07g
- 1.42g = 1.42g 1A ,
12 Q(PRHR)
- Q(ADS)
- Q(Cl) = 1.07g
- 1.42g
- 0.67g.= 1.42g lA/CI 13 Q(PRHR)
- Q(ADS)
- Q(Air Bafue)
= 1.07g
- 1.42g
- 1.28g = 1.42g lA/CF ,
15 Q(PRHR)
- Q(CMT)
- Q(Air Baffle) = 1.07g
- 0.67g
- 1.28g = 1.28g CF/CD 16 Q(PRHR)
- Q(CMT)
- Q(Recric) = 1.07g
- 0.67g
- 0.67g = 1.07g 3BL 17 Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(Cl)
= 1.07g *0.67 g
- 0.67g
- 0.67g = 1.07g 3BUCI 18 Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BUCF 19 Q(PRHR)
- Q(CMT)
- Q(Gravity inject) = 1.07g
- 0.67g
- 0.67g = 1.07g 3BE 20 Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
- Q(CI)
= 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.07g 3BE/CI 21 Q(PRHR)
- Q(CMT)
- Q(Gravity Inject)
- Q(Air Baffle)
= 1.07g
- 0.67g
- 1.28g = 1.28g 3BE/CF 720.158(R2)-50 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-4 (Continued)
(Sheet 2 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 22 Q(PRHR)
- Q(CMT)
- Q(ACC) = 1.07g
- 0.67g
- 0.67g = 1.07g 3BR 23 Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(CI)
= 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.07g 3BR/CI 24 Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle)
= 1.07g
- 0.67g
- 1.28g = 1.28g 3BR/CF 25 Q(PRHR)
- Q(CMT)
- Q(OAADS)
= 1.07g
- 0.67g
- 0.92g + OA(2.2E-3) = 1.07g + 0A(2.2E-3) 1A 26 Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(CD
= 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 0.67g = 1.07g + OA(2.2E 3) 1A/CI 27 Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 1.28g = 1.28g + OA(2.2E-3) 1A/CF 29 Q(Valves Close)
- Q(Air Baffle) = 0.67g
- 1.28g = 1.28g CF/CD 31 Q(Valves Close)
- Q(PRHR)
- Q(Air Baffle)
= 0.67g
- 1.07g
- 1.28g = 1.28g CF/CD 32 Q(Valves Close)
- Q(PRHR)
- Q(Rectic)
= 0.67 g
- 1.07 g
- 0.67g = 1.07g 3BL 33 Q(Valves Close)
- Q(PRHR)
- Q(Recirc)
- Q(Cl)
= 0.67g
- 1.07 g
- 0.67g
- 0.67g = 1.07g 3BUCI 34 Q(Valves Close)
- Q(PRHR)
- Q(Recirc)
- Q(Air Baffle)
= 0.67g
- 1.07 g
- 0.67g
- 1.28g = 1.28g 3BUCF 35 Q(Valves Close)
- Q(PRHR)
- Q(Gravity Inject)
= 0.67g
- 1.07 g
- 0.67g = 1.07g 3BE 36 Q(Valves Close)
- Q(PRHR)
- Q(Gravity inject)
- Q(CI)
= 0.67g
- 1.07 g
- 0.67g
- 0.67g = 1.07g 3BE/CI 37 Q(Valves Close)
- Q(PRHR)
- Q(Gravity inject)
- Q(Air Baffle) ;
= 0.67g
- 1.07g
- 0.67g
- 1.28g = 1.28g 3BE/CF 720.158(R2)-51 1
1
NRC RES.UEST FOR ADDITIONAL INFORMATION 4-M Response Revision 2 Table H-1 (Continued)
(Sheet 3 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 38 Q(Valves Close)
- Q(PRHR)
- Q(ADS)
= 0.67g
- 1.07 g
- 1.42g = 1.42g 1A 39 Q(Valves Close)
- Q(PRHR)
- Q(ADS)
- Q(CI)
= 0.67g
- 1.07 g
- 1.42g
- 0.67g = 1.42g 1A/CI 40 Q(Valves Close)
- Q(PRHR)
- Q(ADS)
- Q(Air Baffle)
= 0.67g
- 1.07g
- 1.42g
- 1.28g = 1.42g lA/CF 42 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(A tifle)
= 0.67g
- 1.07 g
- 0.67g
- 1.28g = 1.28g CF/CD 43 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recric)
= 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g 3BL 44 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(Cl)
= 0.67g
- 1.07g *0.67 g
- 0.67g
- 0.67g = 1.07g 3BUCI 45 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 0.67g
- 01.07g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BUCF 46 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- QI sravity Inject)
= 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g 3BE 47 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
- Q(CI)
= 0.67g
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.07g 3BFJCI 48 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Gravity Injec0
- Q(Air Baffle)
= 0.67g
- 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BE/CF 49 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
= 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g 3BR SO Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(ACCO
- Q(Cl)
= 0.67g
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.07g 3BR,CI 51 Q(Valves Chwe)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(Air Baffic)
= 0.67g
- 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BR/CF
$2 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
= 0.67g
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3) = 1.07g +0A(2.2E-3) 1A 720.158(R2)-52 3 Westingh0!tse
NRC REQUEST FOR ADDITIONAL INFORMATION 159 nin.
~
Responso Revision 2 Table H-4 (Continued)
(Sheet 4 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 53 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Cl)
= 0.67 g
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 0.67g
= 1.07g + OA(2.2E-3) lA/Cl 54 Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 0.67g
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 1.28g
= 1.28g + OA(2.2E-3) IA/CF 56 R(SGTR)
- Q(Valves Close)
- Q(Air Baffle)
= 1.6E-2
- 0.6 7g
- 1.28g = 1.28g + 1.6E-2 CF/CD 57 R(SGTR)
- Q(Valves Close)
- Q(Recric)
= 1.6E-2
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6L 58 R(SGTR)
- Q(Valves Close)
- Q(Recirc)
- Q(CD
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6UCI 59 R(SGTR)
- Q(Valves Close)
- Q(Recirc)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6UCF 60 R(SGTR)
- Q(Valves Close)
- Q(Gravity inject)
= 1.6E-2
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E 61 R(SGTR)
- Q(Valves Close)
- Q(Gravity inject)
- Q(CD
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E/CI 62 R(SGTR)
- Q(Valves Close)
- Q(Gravity inject)
- Q( Air Baffle)
= 1.6E-2
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 63 R(SGTR)
- Q(Valves Close)
- Q(ADS)
= 1.6E-2
- 0.67g
- 1.42g = 1.42g + 1.6E-2 6E 64 R(SGTR)
- Q(Valves Close)
- Q(ADS)
- Q(CI)
= 1.6E-2
- 0.67g
- 1.42g
- 0.67g = 1.42g + 1.6E-2 6E/Cl 65 R(SGTR)
- Q(Valves Close)
- Q(ADS)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.42g
- 1.28g = 1.42g + 1.6E-2 6E/CF 67 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.28g = 1.28g + 1.6E-2 CF/CD 720.158(R2)-53 l
l
NRC REEUEST FOR ADDITIONAL INFORMATION
=
Response Revision 2
- R.u Table H4 (Continued)
(Sheet 5 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE l'ath Quantification End State 68 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Recric)
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6L 69 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Recirc)
- Q(CD -
= 1.6E-2
- 0.67 g
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6UCI 70 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6UCF 71 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Gravity inject)
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E 72 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Gravity Inject)
- Q(CD
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E/CI 73 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(Gravity Inject)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 74 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(ACC)
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E 75 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(ACC)
- Q(Cl)
= 1.6E-2
- 0.67g
- 0.67g
- 0.67g
- 0.67g = 0.67g + 1.6E-2 6E/Cl 76 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle) '
= 1.6E-2
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 77 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(OAADS)
= 1.6E-2
- 0.67 g
- 0.67g
- 0.92g + OA(2.2E-3)
= 0.92g + 1.6E-2 + 2.2E-3 (OA) 6E 78 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- QlOAADS)
- Q(CD
= 1.6E-2
- 0.67 g
- 0.67g
- 0.67 g
- 0.92g + OA(2.2E-3)
= 0.92g + 1.6E-2 + 2.2E-3 (OA) 6E/CI 79 R(SGTR)
- Q(Valves Close)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 0.67g
- 1.28g
- 0.92g + OA(2.2E-3)
= 1.28g + 1.6E-2 + 2.2E-3 (OA) 6E/CF 81 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Air Baffle)
= 1.6E-2
- 0.6 7g
- 1.07g
- 1.28g = 1.28g + 1.6E-2 CF/CD l
720.158(R2)-54 Westinghouse i
NRC REQUEST FOR ADDITIONAL INFORMATION
- =amme= ~l Response Revision 2 p i
Table H-4 (Continued) l (Sheet 6 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 82 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Recric)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g = 1.07g + 1.6E-2 6L 83 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Recirc)
- Q(Cl)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6UCI 84 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Recire)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6UCF 85 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Gravity inject)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g = 1.07g + 1.6E-2 6E 86 R(SGTR)
- Q(Valves Close)
- Q(PRHK)
- Q(Gravity inject)
- Q(CD
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6EICI 87 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(Gravity inject)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 88 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(ADS)
= 1.6E-2
- 0.67g
- 1.07 g
- 1.42g = 1.42g + 1.6E-2 6E 89 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(ADS)
- Q(CD
= 1.6E-2
- 0.67g
- 1.07g
- 1.42g
- 0.67g = 1.42g + 1.6E-2 6E/CI 90 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(ADS)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.07g
- 1.42g
- 1.28g = 1.42g + 1.6E-2 6E/CF 92 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Air Baille)
= 1.6E-2
- 0.67 g
- 1.07g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 CF/CD 93 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recric)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6E 94 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(CD
= 1.6E-2
- 0.67g
- 1.07g *0.67g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6E/Cl 95 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Recire)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 01.07g
- 0.67g
- 0.67g
- 1.28g = 1.28g + 1.62E-2 6E/CF i
96 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject) l
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6E ,
l 720.158(R2)-55
I NRC RE2UEST FCR ADDITIONAL INFORMATION Response Revision 2 I i
Table H-4 (Continued)
(Sheet 7 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 97 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
- Q(Cl)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6E/Cl 98 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
- Q(Air Baffle)
= 1.6E-2
- 0.67g
- 1.07g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 99 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
= 1.6E-2
- 0.67g
- 1.07 g
- 0.67 g
- 0.67g = 1.07g + 1.6E-2 6E 100 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(Cl)
= 1.6E.2 *0.67g
- 1.07 g
- 0.67 g
- 0.67g
- 0.67g = 1.07g + 1.6E-2 6E/CI 101 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle)
= 1.6E-2 *0.67g
- 1.07g
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 102 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
= 1.6E-2 *0.67g
- 1.07 g
- 0.67g
- 0.92g + 2.2E-3
= 1.07g + 1.6E-2+ 2.2E-3 6E 103 R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(CD
= 1.6E-2 *0.67g
- 1.07g
- 0.67 g
- 0.92g + OA(2.2E-3)
- 0.67g
= 1.07g + 1.6E-2 + 0A(2.2E-3) 6E/CI 1(M R(SGTR)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 1.6E-2 *0.67g
- 1.07g
- 0.67g
- 1.28g
- 0.92g + OA(2.2E-3)
= 1.28g + 1.6E-2 + OA(2.2E-3) 6E/CF 106 Q(PMS)
- Q(Air Baffic) = 1.48g
- 1.28g = 1.48g CF/CD 107 Q(PMS)
- Q(PRHR) = 1.48g
- 1.99g = 1.99g 1A 108 Q(PMS)
- Q(PRHR)
- Q(CD = 1.48g
- 1.99g
- 0.67g = 1.99g 1A/CI e
109 Q(PMS)
- Q(PRHR)
- Q(Air Baffle) = 1.48g
- 1.99g
- 1.28g = 1.99g 1A/CF i10 Q(PMS)
- Q(PRHR)
- Q(CMT) = 1.48g
- 1.99g
- 0.67g = 1.99g 1A i
l11 Q(PMS)
- Q(PRHR)
- Q(CMT)
- Q(CI) ,
= 1.48g
- 1.99 g
- 0.67g
- 0.67g = 1.99g 1A/Cl I 112 Q(PMS)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffle)
= 1.48g
- 1.99g
- 0.67g
- 1.28g = 1.99g 1A/CF 720.158(R2)-56 W~
i 1
l NRC REQUEST FOR ADDITIONAL INFORMATION ;
i Response Revision 2 i
i Table H-4 (Continued)
(Sheet 8 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 114 Q(PMS)
- Q(Valves Close)
- Q(Air Baffle)
= 1.48g
- 0.67g
- 1.28g = 1.48g CF/CD 115 Q(PMS)
- Q(Valves Close)
- Q(PRHR)= 1.48g
- 0.67g
- 1.99g = 1.99g lA 116 Q(PMS)
- Q(Valves Close)
- Q(PRHR)
- Q(Cl)
= 1.48g
- 0.67g
- 1.99g
- 0.67g = 1.99g IA/Ci 117 Q(PMS)
- Q(Valves Close)
- Q(PRHR)
- Q(Air Baffic)
= 1.48g
- 0.67g
- 1.99g
- 1.28g = 1.99g 1A/CF 118 Q(PMS)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
= 1.48g
- 0.67g
- 1.99g
- 0.67g = 1.99g lA 119 Q(PMS)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Cl)
= 1.48g
- 0.67g
- 1.99g
- 0.67g
- 0.67g = 1.99g lA/Cl 120 Q(PMS)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffic)
= 1.48g
- 0.67g
- 1.99g *0.67g
- 1.28g = 1.99g lA/CF 121 Q(PMS)
- Q(Valves Close)*R(SGTR)
= 1.48g
- 0.67g
- 1.6E-2 * = 1.48g + 1.6E-2 6E 122 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(Cl)
= 1.48g
- 0.67g
- 1.6E-2
- 0.67g = 1.48g + 1.6E-2 6E/Cl 123 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(Air Baffle)
= 1.48g
- 0.67g
- 1.28g = 1.48g + 1.6E-2 6E/CF 124 Q(PMS)
- Q(Valves Close)*R(SGTR) 'Q(CMT)
= 1.48g
- 0.67g
- 1.6E-2
- 0.67g = 1.48g + 1.6E-2 6E 125 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(CMT)
- Q(CD
= 1.48g
- 0.67g
- 1.6E-2
- 0.67g
- 0.67g = 1.48g + 1.6E-2 6F CI 126 Q(PMS)
- Q(Valves Close)*R(SG7R)
- Q(CMT)
- Q(Air Baffle)
= 1.48g
- 0.67g
- 1.6E-2 *0.67g
- 1.28g s 1.J.'!g + 1.6E-2 6E/CF 127. Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRH16
= 1.48g
- 0.67g
- 1.6E-2* 1.99g = 1.99g + 1.6E-2 6E 720.158(R2)-57
NRC RE8UEST FOR ADDITIONAL INFORMATION !
If i!
Response Revision 2
}
l Table H-4 (Continued) i (Sheet 9 of 11) l l
EVALUATION OF HCLPFS ON MSBO EVENT TREE !
Path Quantification End State 128 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRHR)
- Q(Cl)
= 1.48g
- 0.67g
- 1.6E-2* 1.99g
- 0.67g = 1.99g + 1.6E-2 6E/CI 129 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRHR)
- Q(Air Baffle)
= 1.48g
- 0.67g
- 1.99g
- 1.28g = 1.99g + 1.6E-2 6E/CF 130 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRHR)
- Q(CMT)
= 1.48g
- 0.67g
- 1.6E-2
- 1.99g
- 0.67g = 1.99g + 1.6E-2 6E 131 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRHR)
- Q(CMT)
- Q(Cl)
= 1.48g
- 0.67g
- 1.6E-2
- 1.99g
- 0.67g
- 0.67g = 1.99g + 1.6E-2 6E/CI 132 Q(PMS)
- Q(Valves Close)*R(SGTR)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffle)
= 1.48g ' O.67g
- 1.6E-2
- 1.99g *0.67g
- 1.28g = 1.99g + 1.6E-2 6E/CF 134 Q(DC)
- Q(Air Baffle) = 0.65g
- 1.28g = 1.28g CF/CD 135 Q(DC)
- Qf PRHR) = 0.65g
- 1.99g = 1.99g lA 136 Q(DC)
- Q(PRHR)
- Q(CI) = 0.65g
- 1.99g
- 0.67g = 1.99g 1A/CI 137 Q(DC)
- Q(PRHR)
- Q(Air Baffle) = 0.65g
- 1.99g
- 1.28g = 1.99g 1A/CF 138 Q(DC)
- Q(PRHR)
- Q(CMT) = 0.65g
- 1.99g
- 0.67g = 1.99g 1A 139 Q(DC)
- Q(PRHR)
- Q(CMT)
- Q(CI)
= 0.65g
- 1.99g
- 0.67g
- 0.67g = 1.99g iA/CI 140 Q(DC)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffle)
= 0.65g
- 1.99g
- 0.67g
- 1.28g = 1.99g 1 A/CF 141 Q(DC)
- Q(Valves Close)
- Q(Air Baffle) = 0.65g
- 0.67g
- 1.28g = 1.28g CF/CD 143 Q(DC)
- Q(Valves Close)
- Q(PRHR)= 0.65g
- 0.67g
- 1.99g = 1.99g 1A 144 Q(DC)
- Q(Valves Close)
- Q(PRHR)
- Q(CI)
= 0.65g
- 0.67g
- 1.99g
- 0.67g = 1.99g lA/Cl 145 Q(DC)
- Q(Valves Close)
- Q(PRHR)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 1.99g
- 1.28g = 1.99g lA/CF 720.158(R2)-58 3 Westirighouse l
l
l l
NRC REQUEST FOR ADDITIONAL INFORMATION j Responso Revision 2 l
l Table H-4 (Continued)
(Sheet 10 of 11)
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 146 Q(DC)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
= 0.65g
- 0.67g
- 1.99g
- 0.67g = 1.99g 1A 147 Q(DC)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(CI)
= 0.65g
- 0.67g
- 1.99g
- 0.67g
- 0.67g = 1.99g 1A/CI 148 Q(DC)
- Q(Valves Close)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 1.99g
- 0.67g
- 1.28g = 1.99g lA/CF 149 Q(DC)
- Q(Valves Close)
- R(SGTR)
= 0.65g
- 0.67g
- 1.6E-2 * = 0.67g + 1.6E-2 6E 150 QCDC)
- Q(Valves Close)
- R(SGTR)
- Q(Cl)
= 0.65g
- 0.67g
- 1.6E-2
- 0.67g = 0.67g + 1.6E-2 6E/CI 151 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q( Air Baffle)
= 0.65
- 0.67g
- 1.6E-2
- 1.28g = 1.28g + 1.6E-2 6E/CF 152 Q(DC)
- Q(Valves Close)
- R(SGTR) *Q(CMT)
= 0.65g
- 0.67g
- 1.6E-2
- 0.67g = 0.67g + 1.6E-2 6E 153 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(CMT)
- Q(CI)
= 0.65g
- 0.67g
- 1.6E-2
- 0.67g
- 0.67g = 0.67g + 1.6E 2 6E/CI 154 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(CMT)
- Q(Air Baille)
= 0.65g
- 0.67g
- 1.6E-2
- 0.67g
- 1.28g = 1.28g + 1.6E-2 6E/CF 155 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
= 0.65g
- 0.67g
- 1.6E-2* 1.99g = 1.99g + 1.6E-2 6E 156 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
- Q(CI)
= 0.65g
- 0.67g
- 1.6E-2* 1.99g
- 0.67g = 1.99g + 1.6E-2 6E/Cl 157 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 1.99g
- 1.28g = 1.99g + 1.6E-2 6E/CF 158 QCDC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
- Q(CMT)
= 0.65g
- 0.67g
- 1.6E-2
- 1.99g
- 0.67g = 1.99g + 1.6E-2 6E 720.158(R2)-59
NRC REOUEST FOR ADDITIONAL INFORMATION E Response Revision 2
~
I Table H-4 (Continued) l (Sheet 11 of 11) !
EVALUATION OF HCLPFS ON MSBO EVENT TREE Path Quantification End State 159 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
- Q(CMT)
- Q(Cl)
= 0.65g
- 0.67g
- 1.6E-2
- 1.99g
- 0.67g
- 0.67g = 1.99g + 1.6E-2 6E/CI q
160 Q(DC)
- Q(Valves Close)
- R(SGTR)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffic)
= 0.65g
- 0.67g
- 1.6E-2
- 1.99g *0.67g
- 1.28g = 1.99g + 1.6E-2 6E/CF i
720.158(R2)-60 W Westinghouse 1
1
NRC REQUEST FOR ADDITIONAL INFORMATION j Responso Revision 2 l
Table H-5 (Sheet 1 of 4)
EVALUATION OF HCLPFS ON SMALL LOCA EVENT TREE l'ath Quantification End State 2 R(Small LOCA)
- Q(Air Baffle) = 1.0E-2
- 1.28g CF/CD 3 R(Small LOCA)
- Q(Recire) = 1.0E-2
- 0.67g 3BL 4 R(Small LOCA)
- Q(Recirc)
- Q(Cl)
= 1.0E-2
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BUCI 5 R(Small LOCA)
- Q(Recire)
- Q(Air Baffic)
= 1.0E-2
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BUCF 6 R(Small LOCA)
- Q(Gravity Inject) = 1.0E-2
- 0.67g 3BE 7 R(Small LOCA)
- Q(Gravity inject)
- Q(CI)
= 1.0E-2
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BE/CI 8 R(Small LOCA)
- Q(Gravity inject)
- Q(Air Baffle)
= 1.0E-2
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BE/CF 9 R(Small LOCA)
- Q(ADS) = 1.0E-2
- 1.42g 1A 10 R(Small LOCA)
- Q(ADS)
- Q(CI)
= 1.0E-2
- 1.42g
- 0.67g = 1.0E-2
- 1.42g 1A/Cl 11 R(Small LOCA)
- Q(ADS)
- Q(Air Baffle)
= 1.0E-2
- 1.42g
- 1.28g = 1.0E-2
- 1.42g lA/CF 13 R(Small LOCA)
- Q(CMT)
- Q(Air Baffle)
= 1.0E-2
- 0.67g
- 1.28g = 1.0E-2
- 1.28g CF/CD 14 R(Small LOCA)
- Q(CMT)
- Q(Recire)
= 1.0E-2
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BL 15 R(Small LOCA)
- Q(CMT)
- Q(Recire)
- Q(Cl)
= 1.0E-2
- 0.67g
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BUCI 16 R(Smal! LOCA)
- Q(CMT)
- Q(Recire)
- Q(Air Baffle) l
= 1.0E-2
- 0.67g
- 0.67g
- 1.28g = 1,0E-2
- 1.28g 3BUCF 17 R(Small LOCA)
- Q(CMT)
- Q(Gravity Inject) l
= 1.0E-2
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BE 720.158(R2)-61
NRC RE80EST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-5 (Continued)
(Sheet 2 of 4)
EVALUATION OF HCLPFS ON SMALL LOCA EVENT TREE Path Quantification End State 18 R(Small LOCA)
- Q(CMT)
- Q(Gravity inject)
- Q(Cl)
= 1.0E-2
- 0.67g
- 0.67g
- 0.07g = 1.0E-2
- 0.67g 3BE/CI 19 R(Small LOCA)
- Q(CMT)
- Q(Gravity Inject)
- Q(Air Baffle)
= 1.0E-2
- 0.67g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BE/CF 20 R(Small LOCA)
- Q(CMT)
- Q(ACC)
= 1.0E-2
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BR 21 R(Small LOCA)
- Q(CMT)
- Q(ACC)
- Q(CI)
= 1.0E-2
- 0.67g
- 0.67g
- 0.67g = 1.0E-2
- 0.67g 3BR/Cl 22 R(Small LOCA)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle)
= 1.0E-2
- 0.67g
- 0.67g
- 1.28g = :.0E-2
- 1.28g 3BR/CF 23 R(Small LOCA)
- Q(CMT)
- Q(OAADS)
= 1.0E-2
- 0.67g
- 0.92g + OA(2.2E-3)
= 1.0E-2
- 0.92g + OA(2.2E-3) 1A 24 R(Small LOCA)
- Q(CMT)
- Q(OAADS)
- Q(Cl)
= 1.0E-2
- 0.67g
- 0.92g + OA(2.2E-3)* 0.67g
= 1.0E-2
- 0.92g + OA(2.2E-3) 1A/CI 25 R(Small LOCA)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 1.0E-2
- 0.67g
- 0.92g + OA(2.2E-3)
- 1.28g
= 1.0E-2
- 1.28g + OA(2.2E-3) lA/CF 27 R(Small LOCA)
- Q(PRHR)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 1.28g = 1.0E-2
- 1.28g CF/CD 28 R(Small LOCA)
- Q(PRHR)
- Q(Recirc)
= 1.0E-2
- 1.07g
- 0.67g = 1.0E-2
- 1.07g 3BL 29 R(Small LOCA)
- Q(PRHR)
- Q(Recirc)
- Q(Cl)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BUCI 30 R(Small LOCA)
- Q(PRHR)
- Q(Recirc)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BUCF 31 R(Small LOCA)
- Q(PRHR)
- Q(Gravity inject) i
= 1.0E-2
- 1.07g
- 0.67g = 1.0E-2
- 1.07g 3BE l
l l
720.158(R2)-62 3 Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-5 (Contmued)
(Sheet 3 of 4)
EVALUATION OF HCLPFS ON SMALL LOCA EVENT TREE Path Quantification End State 32 R(Small LOCA)
- Q(PRHR)
- Q(Gravity Inject)
- Q(Cl)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g = 1.0E-2
- 0.67g
- 1.07g 3BE/CI 33 R(Small LOCA)
- Q(PRHR)
- Q(Gravity inject)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BE/CF 34 R(Small LOCA)
- QlPRHR)
- Q(ADS)
= 1.0E-2
- 1.07g
- 1.42g = 1.0E-2
- 1.42g lA 35 R(Small LOCA)
- Q(PRHR)
- Q(ADS)
- Q(CD
= 1.0E-2
- 1.07g
- 1.42g
- 0.67g = 1.0E-2
- 1.42g 1A/CI 36 R(Small LOCA)
- Q(PRHR)
- Q(ADS)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 1.42g
- 1.28g = 1.0E-2
- 1.42g lA/CF 38 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g CF/CD 39 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Recirc)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BL 40 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Recire)
- Q(CI)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BUCI 41 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BUCF 42 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BE 43 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Gravity inject)
- Q(CD
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BE/CI 44 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(Gravity Inject)
- Q(Air Baffle) !
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BE/CF 45 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BR 46 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(CD
= 1 OE-2
- 1.07g
- 0.67g
- 0.67g
- 0.67g = 1.0E-2
- 1.07g 3BR/CI 720.158(R2)-63
NRC REEUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-5 (Contmued)
(Sheet 4 of 4)
EVALUATION OF HCLPFS ON SMALL LOCA EVENT TREE Path Quantification End State 47 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 0.67g
- 1.28g = 1.0E-2
- 1.28g 3BR/CF 48 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
= 1.0E-2
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
= 1.0E-2
- 1.07g + OA(2.2E-3) 1A 49 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Cl)
= 1.0E-2
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 0.67g
= 1.0E-2
- 1.07g + OA(2.2E-3) 1A/Cl 50 R(Small LOCA)
- Q(PRHR)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 1.0E-2
- 1.07g
- 0.67g
- 0.92g + OA(2.2E-3)
- 1.28g
= 1.0E-2
- 1.28g + OA(2.2E-3) 1A/CF 51 R(Small LOCA)
- Q(PMS) = 1.0E-2
- 1.48g 3BE 52 R(Small LOCA)
- Q(PMS)
- Q(Cl)
= 1.0E-2
- 1.48g
- 0.67g = 1.0E-2
- 1.48g 3BE/Cl 53 R(Small LOCA)
- Q(PMS)
- Q(Air Baffle)
= 1.0E-2
- 1.48g
- 0.67g = 1.0E-2
- 1.48g 3BE/CF 54 R(Small LOCA)
- Q(DC) = 1.0E-2
- 0.65g 3BE 55 R(Small LOCA)
- Q(DC)
- Q(CD
= 1.0E-2
- 0.65g
- 0.67g = 1.0E-2
- 0.67g 3BE/Cl 56 R(Small LOCA)
- Q(DC)
- Q(Air Baffle)
= 1.0E-2
- 0.65g
- 1.28g = 1.0E-2
- 1.28g 3BE/CF l
I 720.158(R2)-64 W WBStinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table H-6 (Sheet I of 2)
EVALUATION OF HCLPFS ON LARGE LOCA EVENT TREE Path Quantification End State 2 Q(PRHR-HX)
- Q(Air Baffle) = 0.65g
- 1.28g = 1.28g CF/CD 3 Q(PRHR-HX)
- Q(Recirc) = 0.65g
- 0.67g = 0.67g 3BL 4 Q(PRHR-HX)
- Q(Recire)
- Q(Cl)
= 0.65g
- 0.67g
- 0.67g = 0.67g 3BUCI 5 Q(PRHR-HX)
- Q(Recirc)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 1.28g = 1.28g 3BUCF 6 Q(PRHR-HX)
- Q(Gravity inject) = 0.65g
- 0.67g = 0.67g 3BE 8 Q(PRHR-HX)
- Q(Gravity inject)
- Q(CD
= 0.65g
- 0.67g
- 0.67g = 0.67g 3BE/CI 8 Q(PRHR-HX)
- Q(Gravity inject)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 1.28g = 1.28g 3BE/CF 10 Q(PRHR-HX)
- Q(CMT)
- Q(^ir Baffle)
= 0.65g
- 0.67g
- 1.28g = 1.28g CF/CD 11 Q(PR.HR-HX)
- Q(CMT)
- Q(Recire)
= 0.65g
- 0.67g
- 0.67g = 0.67g 3BL 12 Q(PRHR-HX)
- Q(CMT)
- Q(Recirc)
- QlCI)
= 0.65g
- 0.67g
- 0.67g
- 0.67g = 0.67g 3BUCI 13 Q(PRHR-HX)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BUCF 14 Q(PRHR-HX)
- Q(CMT)
- Q(Gravity inject)
= 0.65g
- 0.67g
- 0.67g = 0.67g 3BE 15 Q(PRHR-HX)
- Q(CMT)
- Q(Gravity inject)
- Q(CI)
= 0.65g
- 0.67g
- 0.67g
- 0.67g = 0.67g 3BE/Cl 16 Q(PRHR-HX)
- Q(CMT)
- Q(Gravity inject)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BE/CF i
17 Q(PRHR-HX)
- Q(CMT)
- Q(ACC)
= 0.65g
- 0.67g
- 0.67g = 0.67g 3BR 720.158(R2)-65
l i
NRC RE2UEST FOR ADDITIONAL INFORMATION i
Response Revision 2 Table H-6 (Continued)
(Sheet 2 of 2)
EVALUATION OF HCLPFS ON LARGE LOCA EVENT TREE Path Quantification End State 18 Q(PRHR-HX)
- Q(CMT)
- Q(ACC)
- Q(Cl)
= 0.65g
- 0.67g
- 0.67g
- 0.67g = 0.67g 3BR/CI 19 Q(PRHR-HX)
- Q(CMT)
- Q(ACC)
- Q(Air Baffle)
= 0.65g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BR/CF i 20 Q(PRHR-HX)
- Q(PMS) = 0.65g
- 1.48g = 1.48g 3BE 21 Q(PRHR-HX)
- Q(PMS)
- Q(CD
= 0.65g
- 1.48g
- 0.67g = 1.48g 3BE/CI 22 Q(PRRR-HX)
- Q(PMS)
- Q(Air Baffle)
= 0.65g
- 1.48g
- 0.67g = 1.48g 3BE/CF 23 Q(PRHR-HX)
- Q(DC) = 0.65g
- 0.65g = 0.65g 3BE 24 Q(PRHR-HX)
- Q(DC)
- Q(Cl)
= 0.65g
- 0.65g
- QCDC)
- Q(Air Baffle)
= 0.65g
- 0.65g
- 1.28g = 1.28g 3BE/CF NOTE: These are not logical sequences because the HCLPFs of the buildings and fuel in the vessel are less than the HCLPFs of either the passive residual heat removal heat exchanger or the reactor coolant system pipes.
720.158(R2)-66 W Westinghouse l
l
NRC REQUEST FOR ADDITIONAL INFORMATION 8t3 t"dt:
Response Revision 2 f Table H-7 (Sheet 1 of 3) l EVALUATION OF HCLPFS ON THE ATWS EVENT TREE Path Quantification End State 2 Q(Core Assembly)
- Q(Air Baffle) = 0.97g
- 1.28g = 1.28g CF/CD 3 Q(Core Assembly)
- Q(Recirc) = 0.97g
- 0.67g = 0.97g 3BL 4 Q(Core Assembly)
- Q(Recirc)
- Q(Cl)
= 0.97g
- 0.67g
- 0.67g = 0.97g 3BUCI 5 Q(Core Assembly)
- Q(Recirc)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 1.28g = 1.28g 3BUCF 6 Q(Core Assembly)
- Q(Gravity inject) = 0.97g
- 0.67g = 0.97g 3BE 7 Q(Core Assembly)
- Q(Gravity inject)
- Q(CD
= 0.97g
- 0.67g
- 0.67g = 0.97g 3BE/Cl 8 Q(Core Assembly)
- Q(Gravity Inject)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 1.28g = 1.28g 3BE/CF 9 Q(Core Assembly)
- Q(ADS) = 0.97g
- 1.42g 3A 10 Q(Core Assembly)
- Q(ADS)
- Q(CI)
= 0.97g
- 1.42g
- 0.67g = 0.97g
- 1.42g 3A/Cl 11 Q(Core Assembly)
- Q(ADS)
- Q(Air Baffle)
= 0.97g
- 1.42g
- 1.28g = 0.97g
- 1.42g 3A/CF 13 Q(Core Assembly)
- Q(CMT)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 1.28g = 0.97g
- 1.28g CF/CD 14 Q(Core Assembly)
- Q(CMT)
- Q(Recirc)
= 0.97g
- 0.67g
- 0.67g = 0.97g 3BL 15 Q(Core Assembly)
- Q(CMT)
- Q(Recirc)
- Q(Cl)
= 0.97g
- 0.67g
- 0.67g
- 0.67g = 0.97g 3BUCI 16 Q(Core Assembly)
- Q(CMT)
- Q(Recirc)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 0.67g
- 1.28g = 1.282 3BUCF 17 Q(Core Assembly)
- Q(CMT) * %(Hi 9 ,ect)
= 0.97g
- 0.67g
- 0.67g = 0.97g 3BE 720.158(R2)-67
NRC RE*.UEST FOR ADDITIONAL. INFORMATION Response Revision 2 Table H-7 (Continued)
(Sheet 2 of 3)
EVALUATION OF HCLPFS ON THE ATV/S EVENT TREE Path Quantification End State 18 Q(Core Assembly)
- Q(CMT)
- Q(Gmvity inject)
- Q(CI)
= 0.97g
- 0.67g
- 0.67g
- 0.67g = 0.97g 3BE/CI 19 Q(Core Assembly)
- Q(CMT)
- Q(Gravity Inject)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 0.67g
- 1.28g = 1.28g 3BE/CF (
20 Q(Core Assembly)
- Q(CMT)
- Q(ACC)
= 0.97g
- 0.67g
- 0.67g = 0.97g 3BR 21 Q(Core Assembly)
- QlCMT)
- Q(ACC)
- Q(Cl)
= 0.97g
- 0.67g
- 0.67g
- 0.67g = 0.97g 3BR/Cl 22 Q(Core Assembly)
- Q(CMT)
- Q(ACC)
- Q(Air Bame)
= 0.97g
- 0.67g
- 0.67g
- 1.28g = 1.2Sg 3BREF 23 Q(Core Assembly)
- Q(CMT)
- Q(OAADS)
= 0.97g
- 0.67g
- 0.92g + OA(2.2E-3) = 0.97g + OA(2.2E-3) 3A 24 Q(Core Assembly)
- Q(CMT)
- Q(OAADS)
- Q(CI) l
= 0.97g
- 0.67g
- 0.92g + OA(2.2E-3)
- 0.67g = 0.97g + OA(2.2E-3) 3A/Cl 25 Q(Core Assembly)
- Q(CMT)
- Q(OAADS)
- Q(Air Baffle)
= 0.97g
- 0.67g
- 0.92g + OA(2.2E-3)
- 1.28g = 1.28g + OA(2.2E-3) 3A/CF l
l 26 Q(Core Assembly)
- Q(PRHR)
= 0.97g
- 1.07g = 1.07g 3A 27 Q(Core Assembly)
- Q(PRHR)
- Q(CI)
= 0.97g
- 1.07g
- 0.67g = 1.07g 3A/Cl 28 Q(Core Assembly)
- Q(PRHR)
- Q(Air Baffle)
= 0.97g
- 1.07g
- 1.28g = 1.28g 3A/CF l 29 Q(Core Assembly)
- Q(PMS) = 1.48g 3A 30 Q(Core Assembly)
- Q(PMS)
- Q(CI)
= 0.97g
- 1.48g
- 0.67g = 1.48g 3A/CI 31 Q(Core Assembly)
- Q(PMS)
- Q(Air Baffle)
= 0.97g
- 1.48g
- 0.67g = 1.48g 3A/CF 720.158(R2)-68 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 ,
Table H-7 (Continued)
(Sheet 3 of 3)
EVALUATION OF HCLPFS ON THE ATWS EVENT TREE Path Quantification End State 32 Q(Core Assembly)
- Q(DC) = 0.97g 3A 33 Q(Core Assembly)
- Q(DC)
- Q(CI)
= 0.97g
- 0.65g
- 0.67g = 0.97g 3A/Cl 34 Q(Core Assembly)
- Q(DC)
- Q(Air Baffle)
= 0.97g
- 0.65g
- 1.28g = 1.28g 3A/CF 35 Q(Core Assembly)
- Q(PR) = 0.97g
- 0.24(PR) 3C 36 Q(Core Assembly)
- Q(PR)
- Q(CI)
= 0.97g
- 0.24(PR)
- 67g = 0.97g
- 0.24(PR) 3C/CI 37 Q(Core Assembly)
- Q(PR)
- Q(Air Baffic)
= 0.97g
- 0.24(PR)
- 1.28g = 1.28g
- 0.24(PR) 3C/CF NOTE: These are not logical sequences because the HCLPFs of the buildings and fuel in the vessel are less than the HCLPFs of the core assembly, i
s f
720.158(R2)-69
4
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. RCS ISOLATION O_ g C CORE M
g SEISMIC ASSEMBLY PIPES PRER NE BUILDINGS 1RWST COMPONENTS NO 8% AIR BAFFLE VALVES PATH END NCLPF N
3 EVENT INTACT INTACT INTACT INTACT INTACT INTACT LOCA INTACT CLOSE NUMBER STATE 26 MSBI 0.789 U g
0 27 CD 0.789 9 H
0.54g l1.28, I .67 9 28 CD/CI 0.78g 9.78g 29 CD/CF 1,28g -,
i 30 CD 1.189 >
1.16g ;0.679 I CD/C1 1.289 Il 1.189 0.62 32 CD/CF 1.289 4 33 CD/CF 0.789 h 34 AtwS 0.979 o
y 0.54g
, 35 CD 0.e79 g g
l.28g {0.67 9 36 CD/CI 0.97q g
9.979 37 CD/CF 1.289 -
38 CD 1.18, O 1.18g yo.67 9 Z g g I 1.28g 39 CD/CI 3.189 0.62g 48 CD/CF 1.28g 41 CD/CF 0.979 KEY:
MSBO - sain steam line and imod line break outside containment SL - small LOCA ATws - enticipated transient without scram mss 1 - main steam line and feed line break inside containment LL/FRER - large LOCA (passive residual heat exchanger piping fails)
LL - large 14CA (core damage can be prevented)
CD - core damage with containment isolation CD/CI - core damage, containent isolation fails CD/CF - core damage and containment failure N
M o
L. Figure H-1 E '
9 w Seismic-induced initiatin9 Events I y (Sheet 2 of 2)
" 1,.
S
h P
~ ll..
$ . i!I h
w SEISMIC-(NDUCED FAILUAES OF THE 4
M CONTROL ROD 5
% }
1
$E15MtC-lNDUCED 3 $El5MIC-INDUCED \
FAILURES OF THE FAILURE OF CORE CONTROL ROD ASSEMBLY (NOT DRIVE SYSTEM FUEL HCLPF = 0.97G HCLPF = 1.25G L J ( )
2 3 z
n O
E D
5 4
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- i
= 6 O
h P
h 54 Figure 11-2 h n O Er o g Seismic Fault Tree - Failure of Control Rods g "g a
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-4 SEISMIC-INDUCED 6 Z
FAILURE OF #
STEAM GENERATOAS 7 m
( )
4 >
6 Z
$ TEAM GENEAATOA STEAM GENEAATOR STEAM GENERATOR STEAM GENEAATOA FAILS COLUMN FAILS UPPEA LATERAL LOWEA INTEAAL SUPPORT FAILS SUPPOAT FAILS HCLPF = 0.79G HCLPF = 0.74G HCLPF = 0.66G HCLPF = 0.87G L J % ) % ) % )
14 15 16 17 U
O Figure 11-4 E _
Sil Seismic Fault Tree - Failure of Primary Components y (Sheet 2 of 2) 5 ,i
I N t:
PO n o
ill h
m
.iiitti M
b k
om CMT ADS ACC GRAVITY RECIRC AIR PATH END RCLPFS IE DC PMS VALVES NO PRNR CLOSE SCTR INJECT , BAFFLE CI NUMBER STATE 1 OK -
l1.28g 2 CF/CD 1.28g 3 OK -
l1.28g 4 CF/CD 1.28g g
5 3BL 1.079 0.67 9y g o.67g
_ ,1.28g 6 3BL/C1 1.079 7 3BL/CF 1.28g 8 3BE 1.07g I
_ 0.67g j g o.679
- 1.28g 9 3BE/CI 1.07q 1.07g 10 3BE/CF 1.28g 11 1A 1.42g 1.42g 0.679 1.2Bg 12 1A/CI 1.42g 13 1A/CF 1.42g 14 OK -
I I.289 I
15 CF/CD 1.28g 2 16 3BL 1.07g M O
0.67gfl.2Sg l0.67g 17 3BL/C1 1.079 m m
3BL/CF 1.28g 18 0.67g 19 3BE 1.07g O 0.67g y 0.67g 20 3BE/C1 1.079 m
I 1.28g 21 3BE/CF 1.28g H 22 3BR 1.079 m 0.679 IO O I I.28g I .679 23 3BR/CI 1.079 M I 24 3 Bit /CF 1.289 >
g 25 1A 1.07g, 2.2E-3 0 0.929, OA = 2.2E-3 0.679 0 I
1.28g 26 1A/C1 1.07g, 2.2E-3 3 27 1A/CF 1.28g 2.2E-3 m Q p e z E t 6o >
" E e
- E Figure 11-5 5 m
$4 m O
?? e "
og. Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree 1 E c >
8 a
(Sheet I of 6) o 3
=f O
- h3 Z t
3 Z D D D
3 m
- E E D co C n g m m
E
- o c IE DC FMS VALVES NO PRER CMT ADS ACC GRAVITY RECIRC AIR PATH END HCLPFS g 6' O g CIDSB SGTR INJEC"T
- BAFFLE CI NUMBER STATE g 2 3>
g 28 CK -
1.2Sg -
I 29 CP/CD 1.28g d g
30 OK -
0 0.67g g l.289 Z 31 CF/CD 1.28g )>
g 32 3BL 1.C7g F 0.67g I I 0.679 -
I.28g 33 3BL/CI 1.079 Z I
34 3BL/CF 1.28g M I
35 3BE 1.07g O 0.67g 0.67g 2 I
l1.289 36 3BE/CI 1.07g E 1.07g 37 3BE/CF 1.28g )>
38 1A 1.42g -4 1.429 y
,1.289 0.679 39 1A/CI 1.423 6
g 40 1A/CF 1.42g g
41 OK -
g 1.28g 42 CP/CD 1.28g r-- 43 3BL 1.079 0.679 l0.67g
,l.289 g
44 3BL/CI 1.079 45 3BL/CF 1.28g 0.67g y 46 3BS 1.079 0.679 , ;0.673
,1.289 47 3BE/CI 1.07g 48 3BE/CF 1.28g y
49 3BR 1.079 0.67g g g o.673 g
l.28g 50 3RR/CI 1.079 51 3BR/CF 1.28g y
52 1A 1.07g, 2.28-3 0.92g, OA = 2.2E-3 0.679
,l.28g g
g 53 1A/CI 1.07g, 2.2E-3 54 1A/CF 1.28g, 2.2E-3 M
M o
. Figure H-5 9 Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree i O (Sheet 2 of 6) 4 -_..s 4 . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ .
l N p:
M p '
l.l.ll
~~ .
m y
N OD IE DC PMS VALVES 28 0 PRNR CPfr ADS ACC GRAVITY RECIRC AIR PATH END HCLPFS CICSE SGTR INJECT BAFF E CI NUMBER STATE
. 55 OK -
l1.28g 56 CF/CD 1.28g, 1.6E-2 g
57 6L 0.67g, 1. 6 E -2 0.67g 0.679 l1.28g - 58 6L/C1 0.67g, 1.6E-2 59 6L/CF 1.283, 1.6E-2 g
60 6E 0.67g, 1.6E-2 8.67g g ;0.679 y 1.289 61 6E/CI 0.67g, 1.6E-2 1.6E-2 62 6E/CF 1.28g, 1.6E-2 63 GE 1.429, 1.6E-2 1.42g y
,o.679 g
,1.28g 64 6E/C1 1.42g, 1.6E-2 65 6E/CF 1.42g. 1.6E-2 66 CK -
l1.28g 67 CF/CD 1.289. 1.6E-2 e6 6L 0.673, 1.6E-2
,o.67g 7 0.67g ,1.28g g
69 EL/CI 0.679, 1.6E 2 g 70 6L/CF 1.289, 1.6E-2 g 0.679 71 6E 0 67g, 1.6E-2 0.679 y
o.679 M
,1.28g g m 72 73 6E/CI 6E/CF 0.679, 1.289, 1.6E-2 1.6E-2 g
g 74 6E 0.679, 1.6E-2 m 0.67g ,0.679 C#D I I
- I I.289 75 6E/CI 0.679, 1.6E-2 76 6E/CF 1.289, 1.6E-2 77 4 6E 0.92g, 1,6E-2 'O O.92g, OA = 2.2E-3 I N I C.673 l1.28g 78* GE/CI 0.92g, 1.6E-2 >
79 a 6E/CF 1.28g, 1.6E-2 O 80 CK -
0 I
I I
I.28g A 81
, , CF/CD 1.289, 1.6E-2 =
- O
=
it M
- Sequences also includes operator action of 2.2E-3 f
- E E.
o 3' ,
4 Figure 11-5 = 0 e =
< K g _
Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree $ $
$ (Sheet 3 of 6) 3 O u z k_=___.-___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- _ _ . - --_ - - - - . . -
T Z A
- 33 14 eO o 3
= m E E o 2 C S
- m ET un
- g. H Q O h
(4 II DC PMS VALVM CICSE 10 0 SGTR PRMR C1fr ADS ACC GRAVITY RECIRC INJECT AIR BAFFLE CI PATH END IFJMBER STATE HC1.FFS 3 O E
m m
,0.67 9 82 6L 1.079, 1.6E-2 >
0.679 O
- l.28g y y
81 6L/C1 1.073, 1.6E-2 -O 84 6L/CF 1.289, 1.6E-2 85 6E 1.079. 1.6E-2 d 0.679 I
0.679 O
'1.289 I 86 6E/CI 1.079, 1.6E-2 Z 1.07g I
87 GE/CF 1.28g, 1.6E-2 >
t 88 6E 1.42g, 1.6E 2 I 1.42g ,0.679 ==
I 1.28g 89 6E/C1 1.42g, 1.6E-2 90 6E/CF 1.429 1.65-2 O
91 OK
'1.289 I E 92 CF/CD 1.28g, 1.6E-2 91 6E 1.079. 1.6E 2 M 0.67g y 0.67g d g
l.28g 94 6E/CI 1.079, 1.6E-2 O 95 6E/CF 1.28g, 1.6E-2 Z 0.67g 96 6E 1.07g, 1.6E-2 0.67g ,0.679 I
g l.289 97 6E/CI 1.079, 1.6E-2 98 6E/CF 1.289, 1.68-2 p 39 6E 1.079, 1.68-2 0.67g g G.679
,1.29g 100 6E/CI 1.079, 1.6E-2 101 6E/CF 1.289, 1.6E-2 y
102e 6E 1.07g, 1.68-2 0.92g, CA = 2.2E 3 0.67g I
1.28g 103* 6E/CI 1.079, 1.65-2 104* 6B/CF 1.28g, 1.6E-2 105 OK -
1.28g 1.48g 106 CF/CD 1.48g g
107 1A 1.99g y
0.67g y
g l.28g 108 1A/CI 1.999 l 109 1A/CF 1.999 u
M o Figure 11-5 L
lE 1B Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree :
u y (Sheet 4 of 6) w I
@ ,m t
%d g "i PJ 5 i o
a A
h . . .
N w
OP O
IB DC fHS VALVES NO FFJIP CPfr ADS ACC GRAVITT RECIRC AIR FA7H END MCLFFS CLOSE SGTR INJECT BAFFLE CI NUMBER STATE l
' 1.999 110 1A 1.993 0.679 ,o.67 9 y g y
1.28g ill 1A/C1 1.999 112 -1A/CF 1.99g 113 OK l1.28g 114 CP/CD 1.48g 0.679 119 1A 1.999
,0.679
- 1.28g g 116 1A/CI 1.99g 1.999 117 1A/CF 1.99g g
118 1A 1.999 0.67g
,1.28g ,0.679 y
119 1A/C1 1.99q 120 1A/CF 1.99g 121 6R 1.48g, 1.6E-2 y
,0.67
, 9 g l.283 122 GE/C1 1.489, 1.6E-2 123 6E/CF 1.483, 1.68-2 y
124 6E 1.48g, 1.6E-2 g 0.679 g o.67g m 1.6E-2 ,1.28g 125 GE/C1 1.48g, 1.68 2 Q l- 126 6E/CF 1.48g, 1.6E-2 0.67g 127 EE 1.999, 1.6E-2 h g
pl.28g 128 6E/CI 1.999, 1.6E-2 g C 1.999 129 EE/CF 1.99g, 1.6E-2 rn 0.67g ,o.67g 130 6E 1.999, 1.6B-2 $
g
,1,28g 131 GE/CI 1.99g, 1.65-2 132 6E/CF 1.999, 1.6B-2 O g
133 OK -
I I.28g M I
134 CF/CD 1.28g U
. 52
=!
n
- O 2
1 6o >
'~
" En
- E co hgure 11-5 5 4 m O 5' e =
9 Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree i K E (Sheet 5 of 6i E $
un a o h3 Z k . _ _ _ _ _ _
2 2
- 21
- O B =
= m
$ O
) 5 G. ?
=4 h IE DC PMS VA1NES 40 PRNR CMP ADS ACC GRAVITT RRCIRC AIR PATH BND HCLPFS O 3 [
g CLOSE STI'R It67ECT BAFFLE CI NUlfiER STATE M
135 1A 1.999 )>
g l0.679 O 3
1.28g 136 1A/CI 1.999 0 0.65g 1.99g 137 1A/CF 1.999 138 1A 1.999 0.67g 3
0.67g O I 1.28g I Z 139 1A/CI 7.999 I 140 1A/CF 1.999 b 141 OK - I I -
I I.289 2 l 142 CF/CD 1.28g 7
143 1A 1.999 0.679 ;O.67g O !
II.28g 'I 144 1A/C1 1.999 1.99g I 145 1A/CF 1.999 E 146 1A 1.999 > i 0.679 l0.679 d
- 1.28g 3
147 1A/CI 1.999 O 148 1A/CF 1.99g Z 149 6E 0.679, 1.65-2
- o.67g
- l.28 9 g
g 150 GE/CI 0.679. f.6E-2 151 GE/CF 1.28 9, 1.6E-2 152 68 0.679, 1.6E-2 0.673 ;0.679 1.6B-2 ;1.28g ,
y 153 6E/CI 0.67 9, 1.6E-2 154 6E/CF 1.28g, 1.6E-2 y
155 6E 1.99g, 1.6B-2
- g o.6?g
,1.28g 156 68/C1 1.999, 1,68 2 1.999 157 GB/CF 1.993, 1.6E-2 158 68 1.999, 1.6B-2 0.679 y
- 0.67g g
,1.28g 159 6B/CI 1.999. 1.6E-2 160 6E/CF 1.993, 1.6E-2 w
L Figure 11-5 S
10 Main Steam Line \ Feed Line Break Outside of Containment Seismic Event Tree I y (Sheet 6 of 6)
I l.l.ll.
cm.
.w.,. - . . - - ,---, . - . - - - . - . .-.,r. . , - . - . . . . . , ,- .,--w . - - ~ . -. .--e--. . - . ~ - - - . - - - - - - , , - - - + , . . . - . . . . - . ~ . . - ~ , - . - - . , - - . . . . . . ~ , - , . - - - - - .
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S F 9g 9g9 9g9 gg 9g ggg g9g ggg gg g P 87 787 782 22 87 787 787 782 28 8 L
C
- 26 626 624 44 - 26 626 626 629 92 2 A -
M 10 010 011 11 10 01D 010 010 O1 1 C .
D IF CC 1F CC IF D IF CC 1F CC 1F CC IF D O -
E C // // CC C // // // CC C L T /L LLE EE // /L LLE EER RR // /
DA K Fa BBB BBA AAK FB BBB BBB BBA AAK F l l
NT O C3 333 331 11O C3 333 333 331 11O C a ES -
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7 7 7 7 7 7 - 7 .
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h en RANDCM DC PMS PRHR SL CNT ADS ACC GRAVITY RECIRC IRJECT AIR BAFFLE CI PATH END NUMBER STATE MCLPFS O 3 h N
(D y 28 3BL 1.079 >
0.673 ,0.679 O 1.289 5
29 3BL/CI 1.07g Q 30 3BL/CF 1.28g 1.07g g
31 3BE 1.07g 0.67g I O.679 -O I I.28g I 32 3BE/CI 1.079 Z I
33 3BE/CF 1.28g >
34 1A 1.42g F j --
1.42g I 0.673 Z I
I.28g 35 1A/CI 1.42g l 36 1A/CF 1.42g 37 OK -
I 1.289 E 38 CF/CD 1.28g 39 2BL 1.07g >
0.67g 0.679 d l1.28g 40 3BL/CI 1.079 O 41 3BL/CF 1.28g g 0.679 42 3BE 1.079 0.679 o.679
,l.28g g
g 43 3BE/CI 1.07g 44 3BE/CF 1.28g y
45 3BR 1.079 0.67g y y o.67g
,1.28g 46 3BR/CI 1.07g 47 3BR/CF 1.26g 48 1A 1.07g, 2.2E-3 0.92g. OA - 2.2E-3 ,0.67 9 y ,
g l.28g 49 1A/CI 1.07g, 2.2E-3 50 1A/CF 1.28g, 2.2E-3 51 3BE 1.48g 1.48g ,0.679
,1.289 52 3BS/CI 1.48g 53 3BS/CF 1.48g 54 3BE 0.65g 0.659 ,0.679
,1.28g 55 3BE/CI 0.679 5
56 3BE/CF 1.289 N
P0 0
- Rgure H-6 9 Small LOCA Seismic Event Tree e!
ye (Sheet 2 of 2) ii u ... :
7 b
L____. __ _ ______. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
b N
g 1..
.1.
^
LL -
PRHR DC PMS CMT ACC GRAVITY RECIRC INJECT AIR BAFFLE CI PATH NUMBER STATE END HCLPFS b
y 1 OK -
g 1.28g 2 CP/CD 1.28g i 3 3BL 0.679 0.679 0.679 1.28g 4 3BL/CI 0.67g ,
5 3BL/CF 1.28g
- 6 3BE 0.679 0.65g 0.67g 0.679 1.28g 7 3BE/CI 0.67g 8 3BE/CF 1.28g 9 OK -
1.28g 10 CF/CD 1.28g 11 3BL 0.67g 0.67g 0.679 0.67g 1.28g 12 3BL/CI 0.679 13 3BL/CF 1.28g 14 3BE 0.67g 0.679 0.67g .
1.28g 15 3BE/CI 0.67g 16 3BE/CF 1.28g z 0.67g 0.67g 17 3BR 0.67g 8 3BE/CI 0.679
- 1.289 18 m 19 3BE/CF 1.28g O 20 3BE 1.48g $
1.48g 0.67g g 1.28g 21 3BE/CI 1.48g .,,
22 3BE/CF 1.48g Q 23 3BE 0.65g N 0.659 0.67g >
1.28g 24 3BE/CI 0.679 @
25 3BE/CF 1.28g q 6
m
- z 14 C
h Figure 11-7 m
M O
S.
d
-o Large LOCA Seismic Event Tree I.
l-f w
g = a w z t
m e%+- iey+me.-egw-w,w+,.,,y,. 3 ,w,. -rr-1,, ,, , , ,, ,m.-,~w iyg., +=, - -rwr, *- yewy 3 we-g-+gy-t m- *g g-+ g -~mp+ ---g-ery-W+Se ew=~ww,e4em-4www- p+e - - -
4-g,-s--,4*, e, w-. .v cgw --r-rerr-g-e.w-%+g-- eerwr -*e'-
3 2 8 23 PR DC PMS FRHR CMT ADS ACC GRAVITY RECIRC AIR PATH END HCLPFS
$O O m
ATWS TNJECT BAFFLEe CI NUMBER STATE 3 m 4 $ O G 1 OK - C N_ -l 1. 2 8g [ m
- 3 2 CF/CD 1.2Sg < M F d ho 0.67g y
g 0.67q 3 3BL 0.979 g m C
g g
l.28g 4 3BL/CI 0.979 3 O g 5 3BL/CF 1.28g 3 m 6 3BE 0.97g g
y 0.67q ;0 g
- l.28g g .67g g
7 JBE/CE 0.979 g
- B 9
3BE/CF 1.28g 3A 1.42g q
0.979 1.42g I D.679 O I I 10 3A/CE 1.42g Z I I.28g 11 3A/CF 1.42g M 12 OK -
I
{1.28g ~g 13 CF/CD 1.28g m y
14 3BL 0.97q 0 0.67g o.67g 3
- ;l.28g g
g 15 3BL/C1 0.97g g 16 3BL/CF 1.28g y 0.679 17 3BE 0.97g -4 0.679 g l0.679 18 3BE/C1 0.97g 3
- 1.289 g 19 3BE/CF 1.28g 20 3BR 0.979 0.679 ;0.67 9 g
,1.289 21 3BR/C1 0.97g 22 3BR/CF 1.28g g
23 3A 0.979 + 2.2E-3 0.92g, OA = 2.2E-3 0.67g
- 1.289 24 3A/CI 0.97g + 2.2E-3 25 3A/CF 1.28g + 2.2E-3 y
26 3A 1.079 1.079 0.67g
- l.28g g
27 3A/C1 1.07g 28 3A/CF 1.28g 29 3A 1.48g 1.489 y
,0.67g
- 1.28g 30 3A/C1 1.48g 31 3A/CF 1.489
%8 M
o
. Figure 11-8
^
g Anticipated Transient Without Scram Seismic Event Tree $ I%
~ (ShcCt 1 of 2) e h S lillklI L _ _ _ _ _ _ - - - _ .
%J . . ..:
M '
P
~ "
DC PMS PRBR CMT ADS ACC GRAVITY RECIRC AIR PATH END BCLPFS
% ATWS PR INJECT BAJFLE CI NUMBER STATE N
Y 32 3A 0.97g 00 O 0.65g y l0.679 g
l.2Bg 33 3A/CI 0.979 34 3A/CF 1.289 y
35 3C PR=0.24+0.97g PR = 0.24 g o.67g y
g l.289 36 3C/CI PR*0.24+0.979 37 3C/CF PR=0.24+1.289 2
m O
E o
Ei a
O m
0 0
m 3
O l I Z uo >
3 4m Evi Figure 11-8 5 E. m O
- 2 e =
9 Anticipated Transient Without Scram Seismic Event Tree 1 E E (Sheet 2 of 2) 8' $
$ 3 O u z
( ___
33 Z
- x O
l E =
g
$E I SW I C - l@UCE')
FatLURE OF OC PDrER SYSTEM y g,- e m # m
\ _.) -4 Ep m 5 (3 -h
= o
= ,
@ [) t@
7 b 8LE TRAYS Fall 3 dnANSFER kEStulC-I M Dl $20 WCC STsTEu 7 FAtLS its WDC Of5TRtBUTION b emm3R CENT 810L' CENTERS F 4 GL l$w t TCHE S F A IL -
F a tLURE OF gaTTERIES PAEL S F AIL 4 d
MCLPF e 3.63G HCLPF e 1.60G >CLPF = 0.79G HCLPF = 0. 7 E3 O
( ) ( ) (- J L J ( ) Z
[ 4 7 9 10 14 h-3 t
y 0
l ka'TERY RACKS k23 WAC 7 $10 WAC l (25 VOC 3 533v saTTERIES F Att FatL 4NYfRTERS F All DISTRIBUTION $W ITCH 90AO g PAWL 3 FAtt falls HCLPF = 1,$ 1G HCLPF = 0.79G HCL PF e 2.41G HCLPF = 0.63G HCLPf = 0.fl5G H L
4 L J L J L J V J f Z
e, 52 s a is u
PJ o Figure 11-9
$ I
- r 2 Seismic Fault Tree - Failure of DC Power '
Ug lll N 4;;
N
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Y $EISMIC-1NDUCED i
FAILUAE OF PAHR
( )
HALVES PXS 1 kALVE PXS 101 TRANSMITTERS 108Af8 FAILS FAILS Fall HCL PF = 1.99G HCL PF = 1.99G k J ( ) ( )
2 3 4 SG NAA AANGE SG wtDE ANGE hAHA HX FLOW D g TRANSMITTER LEVEL TAANSMITTER FAILS TRANSMITTER FAILS FAILS L
HCL PF = 1.07G J k HCL PF = 1.07G
) L HCLPF= 2.44G J
g)
Ei s a 7 M O
m 0
U Ei
= 0
- z 1 >
r-g 3 e Figurt 11-12 5 M l f4 = 0 E o
- 9 Seismic Fault Tree - Failure of PRHR g j 8
o 3
=!
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- w z
3 2 e .D 1
14 g SE 4 SMiC. I h0UCED 3 F AILURE OF CMT 3 g
(
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> e c h & J h M l
g (3 1
s.
".i T
l
. . . . . n IANKS PXS 2AfD ' SEISMIC + INDUCED 3 # SEISMIC.INDUCEO 3 SEISulC. INDUCED SEISMIC INDUCED SEISMtC-INDUCEO ' M FAIL FAILURE OF F AILUAE OF AIR F AILUAE OF e4]VS FAILURE OF PAZ F AILUAE OF PZR O l CHECK VALVES OPERATED VALVES PXS v005 afb PRESURE SENSOR LEVEL U TRANSulTTER WLPF = 1.14G HCLPF = 1.99G K'LPF e 1.42G HCL PF = 1.42G d" u 1 L > e J V 2 L J L ) O
()
{D 6 7 I"
2 3 4 5 l
~Z
, . . . m CV PAS D15 afb 3 b prs 017AtB 3 (ALVES PXS ' kALVES PXS I HALVES PRS HALVES PXS O Fall Fall VOO2 afb FA1L V003 afb Fall V014 afb Fall 015AlB Fall 3 l
E ,
HCLPF = 0.67G HCL F F = 0.87G HCLPF e 1.99G HCLPF e 1.99G K1PF e 1.99G HCLPF = 1 99G >
L J L J V J L J L. J L J -
O z
10 11 12 13 14 15
%4 PJ o
Figure 11-13
$ i lin Seismic Fault Tree - Failure of CMT E S e I
ll..l
.II.
= +
L
M M t.
- t 7
h me kE I SW IC - t 4.1UCED FAILUME OF ADS i
( )
w N ,
SEISu!C I43UCED ' SEISWIC-tNDUCED 3 sAILUME OF CMi F AiLURE OF ADS LEVEL SWITC>t HCLPF e 9.42G L J L >
STAM Two DR 3 $7Am 1,2,3 AND 3 Siam THREE 4 VALVES FAIL VALVES F AIL L J L J 19 12 k S 002fAfBfCfD k S V003Af9fCfD $TAE 1 V AL VE S ST AE TWD WAL VEh ITAGE 3 VALVES iTAGE 4 VALVES (ST AGE 2 VALVES] (STAE THREE ) FAIL (RCS FAlt (RCS FAIL (RCS FAtt FAIL VALVES F AIL V001 rat OfCfD) AUD2fAfBfCfD) A003fAf5FCfD)
HCL PF = 2 O3G >CLPF = 2.53G HCL PF = 2.95G HCLPF = 2 SSG >CLPF = 2.33G M L } L ) V J k ) L ) L ) Q 13
( 14 15 15 17 lh te
=
m m
itR BDTTLES Fall' iTAM 4 VALVES FAet (ACS A004AfBfCfD) M
>CLPF = 0.79G FCLPF e 0.67G O
g L J L J 0
to 20 o m
3 O
l 3 z n >
5 r-5 " En m Figure H-14 $
E. m O
- s e =
- g. Seismic Fault Tree - Failure of ADS 1 E o r >
c= 0 :!
=
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- Figure H-15 8
9 Seismic Fault Tree - Operator Fails to Start ADS Y M
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Question 920.2 l l
Section 13.6.5.2 of the AP600 SSAR states that a listing of vital equipment is provided in the AP600 Security l Design Report. The Security Design Report, dated June 30,1992, did not contain a list of vital equipment. Provide )
a listing of vital equipment for the AP600 design.
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Response (Revision 1): >
Table 2 of the Security Design Report has been updated to provide a listing of AP600 vital equipment and will be provided with the next revision of AP600 Standard Security Design Report. The revised Table 2 is provided as an attachment to letter NTD-NRC-94-4195 i
SSAR Revision: NONE !
Security Design Report Revision: Table 2 of the AP600 Standard Security Design Report will be revised to j include the AP600 vital equipment. ;
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