NSD-NRC-98-5524, Provides W Responses to FSER Open Items on AP600.Summary of Encl Responses,Fser Open Item Number,Associated Oits Number & Status to Be Designated in W Status Column of Oits, Included in Table 1.Rev 4 to WCAP-13914 Encl
| ML20198Q924 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/15/1998 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198Q931 | List: |
| References | |
| NSD-NRC-98-5524, NUDOCS 9801230174 | |
| Download: ML20198Q924 (82) | |
Text
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+ 4 Ba 355 Westinghouse Energy Systems Pittsburgh Fennsykanta 15730 0355 Electric Corporation DCP/NRCl214 NSD-NRC-98 5524 Docket No.: 52-003 January 15,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T.R. QUAV SUlUECT: AP600 RESPONSE TO FSER OPEN ITEMS AND WCAP-13914 (REVISION 3)
Dear Mr. Quay:
Enclosure i of this letter provides the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1. Included in the table is the FSER open item number, the associated OITS number, and the status to be designated in the Westinghouse status column of OiTS. The NRC should review the enclosures and inform Westinghouse of the status to be designated in the "NRC Status" column of OITS. linclosure 2 provides Revision 3 of WCAP-13914," Framework for AP600 Severe Accident Management Guidance." l' lease contact me on (412) 374-4334 if you have any questions concerning this transmittal. fhAN/ Ilrian A. Mtlntfre, Manager Advanced Plant Safety and Licensing jml Enclosures f W. C. Iluffman, NRC (Enclosure 1) ,f cc: T, J. Kenyon, NRC (Enclosure 1) \['gf J. M. Sebrosky, NRC (Enclosures 1 & 2) ( D. C. Scaletti, NRC (Enclosure 1) N. J. Liparule, Westinghouse (w/o Enclosures) _ um.g 9001230174 990115 3 ADOM 0520 \{, , , , ug [ ,
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1 i DCP/NRCl214 i NSD-NRC 98 5524 2- January 15,1998 Table 1 List of FSER Opeu Items Included in Letter DCP/NRCl214 FSER Opt'n Item OITS Number Westinghouse status in OITS 250.34F 6493 Confirm W 410.414F 6520 Confirm W 440.756F 6414 Action N 720.422F 6134 Action N 720.434F 6163 Confirm W 720.439F 6177 Confirm W 720.461F 6487 Action N 720.462F 6488 Action N mu .,r
e
-e Enclosure i to Westinghouse Letter DCP/NRCl214 January 15,1998 a
e d kk k Mf
- FSErl Open item m.
w x 2 g. 250.34F (Ol S #6493) In reviewing Revision 17 of the Westinghouse AP600 SSAR, issued on October 31,1997 and. Revision 18, dated December 5,1997, the issues below were identified.
- a. In Revision 18, Westinghouse has identified twelve PlVs in the PXS and RNS systems in a separate table (Table 3.918). However, the IST Program table (Table 3.916) was not revised making these valves Category A and requiring a leak test in accordance with the Code.
A6ditionally, in Revision 18, valves CVS-PL V03 and CVS-PL V80 have been rJded to the IST 4 Program table (Table 3.916). Note 6 has been identified for these valves, however, note 6 was not revised to include these valve numbers. N.>te 32 was added to Revision 18 and addresses operability testing. The third sentence appears incomplete and it is not evident how the note is different from Note 31,
- b. Revision 17 added some valves "not constructed to the ASME Code...nese vahes art relied on in some safety analysis (SSAR 3.9.6)." Are these valves safety related? If not, they should be identified in Table 3.9-16 as augmented, preferably with a note explaining why they are included.
Bcsed on the change to 3.9.6.2, they are not safety related. The turbine bypass valves have no safety function (per SSAR 10.4.4.1.1) and were included in this revision. Also regariing these valves, Note 29 states that iesting these valves will result in an undesirable temperature transient. These valves are used for normal load reductions. Wutinghouse has not demonstrated that it is impractical to exercise these valves, either fully or panially (e.g., due to the potential of equipment damage or plant transient). Per SSAR 10.4.4.5, the " bypass valves may be tested while the unit is in operation." Westinghouse should refer to NUREG 1482, Section 2.4.5 and 3.1.1.
- c. Valve FNS PL V001 was added in Rev.17. His valve is exercised at refueling, however, no note provides the basis for the deferral. Valve MSS PL-V016 was added to the table. Note 29 is indicated for this valve, however, note 29 does not address this valve. Note 21 is indicated for valve PCS PL V439 in the table, however the note refers to V039A.
Response
3 Table 3.916 will be revised to include leakage requirements for the PlVs. Safety seat leakage will be added to Safety Functions; ASME IST Category will be changed to A; and pressure isolation leak test will be added to inservice Testing Type and Frequency. See attached SSAR markup. Note 6 will be revised to include CVS-PL V003 and CVS PL-V080. Note 31 requires that the Combined License applicant evaluate whether opembility testing is required and to determine the test frequency. Note 32 requires operability testing. Note 32 does not permit longer test frequencies for operability testing. Note 32 ',s complete and a period will be added to the final sentence,
- b. The valves in Table 3.916 that are in the main steam rystem (MSS) and the main turbine system (MTS) are not safety related. As noted in subsection 3.6.1.3.3 for those cases in which the mpture of the main steam or feedwater piping inside containment is the postulated initiating event the D" 250.34-1 l
FSER Open item . . _ turbine stop, turbine control, moisture separator reheater stop, and turbine bypass valves are I credited in single failure analysis to mitigate the event. A note will tx added to Table 3.916 to explain why they are included in the table. Opening the turbine bypass valves during normal operation at steady state power would result in an undesirable temperature transient and loads on condenser tubes. These valves are not used for normal load reductions. ney are used for rapid load reductions and at low power levels during startup and shutdown. He reference to testing the valves during operation will be deleted from subsection 10.4.4.5. . 'I.. relevent portion of section 10.3 and 10.4 will be revised as shown below to reflect that these valves are included in the inservice test program. 3
- c. FN" PL-V001 will be corrected to FHS-PL V001. His va've closes one end of the fuel transfer 5
tube. It should remain closed during operation to minimize the potential for leaking water from the fuel transfer canal in to the refueliag cavity. A note to this effect will be added to Table 3.9-16. The note for MSS-PL V016 was corrected to Note 25 in SSAR Revision 18. Note 21 will be conected to correct the reference to valve PCS-PL-V039. In addition to changes to address comments noted above, the turbine stop valves will be added to Table 3.916 to be consistent with Technical Specfication 3.7.2. A note will be added to the table to reflect that on each inlet line to the turbine either the turbine stop or the turbine control valve must be acceptable per the test requirements. SSAR Revision: Revise the second paragraph of subsection 3.6.1.3.3 as follow: For those cases in which the rupture of the main steam or feedwater piping inside I containment is the postulated initiating event the turbine cc,ntrol, turbine stop, moisture separator reheater stop, and turbine bypass valves and to a limited extent, the control sys* ems for the turbine stop and feedwater control valves (which are nonsafety-rel^d equipment) are credited in single failure analysis to mitigate the event. This equis .nt is not protected from pipe ruptures in the turbine building because the postulated pipe mpture for which it provides protection is inside contair"nent. De assumed single active failure far this analysis is the funct'on of the safety-related valve that would normally isolate the piping. This isolation function is addressed in more detail in Chapter 10. See attached markup of Table 3.9-16. Add the following to subsection 10.3.1.1 1
- Re nonsafety-related turbine stop, turbine control, and moisture separator reheater stop i valves are credited in a single failure analysis to mitigate the event for those cases in 250.34-2
. - _ - .. .. . .. . -- . . . ~
FSER Open item , . , , i i which the rupture of the main steam or feedwater piping inside containment , the i postulated initiating event . Revise the second paragraph of subsection 10.3.4.2 as follows: 1 Additional description of in service inspection and in service testing of ASME Code, Section fil, Class 2 and 3 components is contained in Section 6.6 and subsection 3.9.6. I The nonsafety related turbine stop, turbine control, and moisture separator reheater stop i valves are included in the inservice test program discussed in subsection 3.9.6. Revise subsection 10.4.4.1.1 as follows: 10.4.4.1.1 Safety Design Basis The turbine bypass system serves no safety-related function and therefore has no nuclear i safety desigr. basis. The nonsafety-related turbine bypass valves are credited in a single i failure analysis to mitigate the event for those c w .1 which the rupture of the main I steam or feedwater piping inside containment is t.., gstulated initiating event. Revise subsection 10.4.4.5 as follows: 10.4.4.5 Inspection and Testing Requirements Before the system is placed in service, turbine bypass valves are tested to verify they function properly. The steam lines are hydrostatically tested to confirm leaktightness. The by;;rc v:.hcr. ,cy b :n:cd wh!!c the un4 r., h cgre.:ica System piping and valves are I accessible for inspection. No inservice inspection and testing is required escept for the I turbine bypass valves which are included in the inservice program as discussed in I subsection 3.9.6. U 250.34-3
A
- 3. Design of Structures, Componenta, Equipment, and Systems Table 3.9-16 (Sheet VALVE INSERVICE TEST l Valve Tag Safety Related -
Number Description (3) Valve Type Missions Safety CVS-PL V094 Hydrogen Addition IRC Isolation Check Maintain Close - Active Transfer Close Containmei Safety Scai
- Remote Po CVS-PL V100 Makeup Line Containment Isolation Relief Check Maintain Close Active Transfer Close Containmes Transfer Open Safety Seat CVS PL-V136A Demineralized _ Water System Isolation Remote Maintain Close Active-to-F Transfer Close Remote Po-CVS PL V136B Demineralized Water Syrtem Isolation Remote Maintain Close Active-to-F Transfer Close Remote Po:
DWS PL V244 Demineralized Water Supply Containment Isolation Outside Manual Maintain Close Containme Safety Seat. DWS PL V245 Demineralized Water Supply Containment Isolation - Inside Check Maintain Close Containme Safety Seat FPS PL-V050 Fire Water Containment Supply Isolation . Manual Maintain Close Containme Safety Seat FPS-PL-V052 Fire Water Containment Supply Isolation -Inside Check Maintain Close Containmer Safety Seat l FHNS PL-V001 Fuel Transfer Tubes isolation Valve Manual Transfer Close Active Maintain Open MSS-PL V001 Turbine Bypass Control Valve Remote Maintain Close Active-to@ Transfer Close Remote Pot MSS PL-V002 Turbine Bypass Control Valve Remote Maintain Close Active-to@ Transfer C'ose Remote Po; MSS PL V003 Turbine Bypass Control Valve Remote Maintain Close Active-to-Fi Transfer Close Remote Pos MSS-PL-V004 Turbine Bypass Control Valve Remote Maintain C';se Active-to-F: Transfer Close Remote Pos M flgl100$e ww
r.i--- . - , , . - l L __ t of 19) (EQUIREMENTS ASME IST F nctions(2) Category Inservice Testing Type and Frequency IST Notes AC Remote Position Indication, Exercise /2 Years 27 it Isol tion Containment Isolation Leak Test (See Notes) Leakage Check Exercise / Quarterly Operation
,ition AC Containment isolation Leak Test /2 Yests 23,27 it Isol: tion Check Exercise / Refueling Shutdown leakage siled B Remote Position Indication, Exercise /2 Years 30,31 ition Exercise Full Stroke / Quarterly Operability Test /See Notes ailed B Remote Position Indication, Exercise /2 Years 30,31
,ition Exercise Full Stroke / Quarterly Gperability Test /See Notes t Isol: tion A Containment Isolation Leak Test (See Notes) 27 teatage t isolition AC Containmcnt Isolation Leak Test (See Notes) 27 gg3 v Leak:ge APERTURE t Isolation A Containment isolation Leak Test (See Hotes) 27
) CAQg Lcakage t Isol: tion AC Containment Isolation Leak Test (See flotes) 27 Also Ava#eNe LcakaGC %rture Card B Exercise Full Stroke / Refueling Shutdov n 33 tiled B Remote Position Indication, Exercise /2 Years 29,31 ition Exercise Full Stroke / Cold Shutdown Operability Test /See Note
,iled B Remote Position Indication, Exercire/2 Years 29,31 ition Exercise Full Stroke / Cold Shutdown Operability Test /See Note
.iled - B Remote Position Indication, Exercise /2 Years 29,31 ition Exercise Full Stroke / Cold Shutdown Operability Test /See Note iled B Remote Position Indication, Exercise /2 Years 29,31
, tion Exercise Full Stroke / Cold Shutdown Operability TesvSee Note 0050/7 -OI Revision: 20 Draft,1998 l omnit.6 ris wpt.oi 59s 3.9-135 1
__._.-m __._.__--______m_. _ - - . - _ _ _ _ - . . - - _ - . _ _ -. ' '
.- - - _ . _m>muI
- 3. Design of Structures, Components, Equipment, and Systems Table 3.9-16 (Shect' VALVE INSERVICE TEST)
Valve Tag Safety Related Number Descriptien(I) Valve Type M'ssions Safe MSS-PL-V016 Moisture Seprator Reheater Steam Supply Control Valve Remote Main ain Close Active-to-{ Transfer Close Remote 04 l MTS PL V00l A Tu+ine StopValve Remote Maintain Close Active-to-l Transfer Close Remote
- i l
l MTS-PL V001D Turbine StopValve Remote Maintain Close Active-to l Transfer Close Remote is l l MTS-PL V002A Turbine Centrol Valve Remote Maintain Close Active-t t
-l Transfer Close Remote ' .
l MTS PL-V002B Turbine Control Valve Remote - Maintain Close f l Transfer Close Active-Remote P l MTS-PL V003A Turbine StopValve Remote Maintain Close Active-to-f l Transfer Close Remote Po'
-l l MTS-PL V003B Turbine StopValve Remote Maintain Close Active-to F l Transfer Close Remote Po I
i l MTS PL-VO(MA Turbine Control Valve Remote Maintain Close Active-to-F l Transfer Close Remote Po l MTS PL-V004B Turbine Control Valve Remote Maintain Close Active to-F l Transfer Close Remote Po PCS PL V001A PCCWST Isolation Remote . Maintain Open Active-to-F Transfer Open Remote Po: PCS PL V001B PCCWST isolation Remote Maintain Open Active-to-F Transfer Open Remote Po: PCS PL-V002A PCCWST Series Isolation Remote Maintain Open Active Transfer Open Remote Poi [ W85tingh00$8
w --
!!U En E :ii 4 of 19)
REQUIREMENTS ASME IST F:nctions(2) Category laservice Testing Type and Frequency IST Notes tiled B Remote Position Indication, Exercise /2 Years 25, 31 sition Exercise Part Stroke / Operation Exercise Full Stroke / Cold Shutdown Operability TesVSee Note
*iled . B Remote Position Indication, Exercise /2 Years 31,34,35, sition Exercise Full Stroke / Cold Shutdown 36 Operability Test /Sec Note ailed B Remote Position Indication, Exercise /2 Years 31,34,35, sition Exercise Full Stroke / Cold Shutdown 36 Operability Test /See riote a ed B Remote Position Indication, Exercise /2 Years 25, 31, 34. ;ition Exercise Pan Stroke / Operation 36 Exercise Full Stroke / Cold Shutdown Operability TesvSee Note Remote Position Indication, Exercise /2 Years 25,31,34.
g CV ailed B
;ition Exercise Part Stroke / Operation 36 APERTURE Exercise Full Stroke / Cold Shutdown Operability Test /See Note gg Remote Position Indication, Excreise/2 Years ailed B 31,34,35,36 4I40 Ava% g
.ition Exercise Full Stroke / Cold Shutdown Operability Test ice Note ADMtutt Card ailed B Remote Posit;m. Indication, Exercise /2 Years 31,34,35,
.ition Exercise Full Suoke/ Cold Shutdown 36 Operability Test /See Note ailed B Remott Position Indication, Exercise /2 Years 25, 31, 34,
.ition Exercise Part Stroke / Operation 36 Exercise Full Stroke / Cold Shutdown Operability TesvSee Note sited B Remote Position Indication, Exercise /2 Years 25,31,34, ition Exercise Part Stroke / Operation 36 Exercise Full Stroke / Cold Shutdown Operability TesvSee Note iiled B Remote Position Indication, Exerci.c/2 Years 30,31 ition Exercise Full Stroke / Quarterly Operability Test /See Notes iiled B Remote Position Indication, Exercise /2 Years 30,31 ition Exercise Full Stroke / Quarterly Operability Test /Sce Notes B Remote Position Indication, Exercise /2 Years 30,31 ition Exercise Full Stroke / Quarterly Operability Test /See botes j Revision: 20 Draft,1998 i 0.109ntl6.r18 w pf-Ot t $98 3.9-137
- 3. Design of Struct:res, Components, Equipment, cnd Systerns Table 3.9-16 (Sheet VALVE INSERVICE TEST Yahe Tag Safety Related Number Description (3) *. Jve Ty pe Missions Safet.s PXS-PL-V016D Core Makeup Tank B Discharge Check C. n Maintain Open Active Transfer Open Remote Pi Transfer Close PXS.PL V017A Core Makeup Tank A Discharge Check Check Maintain Open Active Transfer Open Remote P<
Transfer Close PXS-PL V017B Core Makeup Tank B Discharge Check Check Maintain Open Active Transfer Open Remote Pt Transfer Close PXS PL V022A Accumulator A Pressure Relief Relief Maintain Close Active Transfer Open Transfer Close PXS-PL V0223 Accumulator B Pressure Relief Relief Maintain Close Active Transfer Oper Transfer Close PXS-PL-V027A Accumulator A Discharge Isolation Remote Maintain Open Remote Po PXS-PL-V027B Accumulator B Discharge Isolation Remore Maintain Open Remote Po l PXS PL-V028A Accumulator A Discharge Check Check Maintain Close Active Transfer Open RCS Presa l Remote Po I Safety Seal l PXS PL-V028B Accumulator B Discharge Check Check Maintain Close Active Transfer Open RCS Pressi I Remote Po l Safety Seat i PXS-PL-V029A Accumulator A Discharge Check Check Maintain Close Active Transfer Open RCS Pressi l Remote Po I Safety Seat l PXS PL-V029B Accumulator B Discharge Check Check Maintain Close Active Transfer Open RCS Presst l Remote Ps l Safety Seat PXS-PL-V042 Nitrogen Supply Containment Isolation ORC Remote Maintain Close Active-to-I-Transfer Close Containmer Safety Seat Remote Po. PXS PL-V043 Nitrogen Supply Containment Isolation IRC Check Maintain Close Active Transfer Close Containmer Safety Seat Remote Po 3 WestingflouSe
i i La 7 of 19) REQUIREMENTS ASME IST Functions (2) Category Inservlee Testing Type and Frequency IST Notes BC Remote Position Indication, Exercise /2 Years 10
- ition Check Exercise / Refueling Shutdown DC Remote Position Indication Exercise /2 Years 10
.:ition Check Exercise / Refueling Shutdown DC Remote Position Indication. Exercise /2 Years 10 sition Check Exercise / Refueling Shutdown DC Class 2/3 Relief Valve Tests /10 Years and 20% in 4 Years BC Class 2/3 Relief Valve Tests /10 Years and 20% in 4 Years ANSTEC Remote Position Indication, Exercise /2 Years ition B Remote Position Indication, Exercise /2 Years APERTURE kition ABC B
Remote Position Indication, Exercise /2 Years 9 CARD are Boundary Check Exercise / Refueling Shutdown
;ition Also Avaliabl*
Pressure Isolation Leak Test / Refueling Shutdown Aperture hg Leakage
.ABC Remote Position Indication Exercise /2 Years 9 Boundary Check Exercise / Refueling Shutdown tion - Pressure Isolation Leak Test / Refueling Shutdown Le:kage ABC Remote Position Indication Exercise /2 Years 9 Check Exercise / Refueling Shutdown pitionBoundary
- Pressure Isolation Leak Test / Refueling 3hutdown fleakige ABC Remote Position Indication Exercise /2 Years 9 re Boundary Check Exercise / Refueling Shutdown
.ition Pressure Isolation Leak Test / Refueling Shutdown Leakage siled A Remote Position Indication, Exercise /2 Years 27,30,31
!t isolation Containment Isolation Leak Test (See Notes)
Leak:ge Exercise Full Stroke / Quarterly lion Operability Test /See Notes AC Remote Position Indication, Exercise /2 Years 27 t Isolation - Containment Isolation Leak Test (See Notes) Leakage Check Exercise / Quarterly % ition - Revision: 20 Drafi,1998 owwa riswprotis9s 3.9-143 i _ _ L -. . . _
O ~' .
- 3. Design of Structures, Components, Eq:ipmenta and Systems Table 1.9-16 (Sheet VALVE INSERVICE TEST Valve Tag Safety Related Number Description W Valve Type Missions Safet:
RCS PL-V013A Third Stage Automatic Depressurization System Isolation Remote Maintain Open Active Maintain Close RCS Pres Transfer Open Remote Pi RCS-PL V013B Third Stage Automatic Depressurization System Isolation Remote Maintain Open Active Maintain Close RCS Pres: Transfer Open Remote Pi RCS PL V014A Fourth Stage Automatic Depressuriaition System Isolation Remote Maintain Open Remote Pi RCS PL V014D Fourth Stage Automatic Depressurization System Isolation Remotc Maintain Onen Remote Pi RCS-PL-V014C Fourth Stage Automatic Depressurization System isolation Remote Maintain Open Remote P RCS IL V014D Fourth Stage Automt. tic Depressurization System Isolation Remote Maintain Open Remote P RCS PL V150A Reactor Vessel ifcad Vent Remote Maintain Open Active-tp Maintain Close RCS Press Transfer Open Remote Pd RCS PL VISOB Reactor Vessel Head Vent Remote Maintain Open Active to-Maintain Close RCS Press Transfer Open Remote RCS PL-Vl50C Reactor Vessel licad Vent Remote Maintain Open Active-tod Maintain Close RCS Pres.d Transfer Open Remote Pd RCS-PL-V150D Reactor Vessel licad Vent Remote Maintain Open Active-toq Maintain Close RCS Press Transfer Open Remote Pd KCS-K03 Safety Valve Discharge Chamber Rupture Disk Relief Transfer Open Active RCS KM Safety Valve Discharge Chamber Rupture Disk Relief Transfer Open Active l RNS-PL V00; A RNS liot Leg Suction isolation - Inner Reraote Maintain Close Active Transfer Close RCS P , 1 Safety Sed Remote Pt l RNS PL-V001B RNS110 Leg Suction Isolation inner Remote Maintain Close Active Transfer Close RCS Press, 1 Safety Sea! Remote Pd l RNS-PL-V002A RNS Hot Leg Suction and Containment Isolation - Outer Remote Maintain Close Active Transfer Close RCS Press l Containme l Safety Sec Remote & W Westinghouse M ess *e
1 J t I of 19) REQUIREMENTS ASME IST FunctionsW Category incryice Testing Type and Frequency IST Notes B Remote Position Indication, Exercise /2 Years 3, 31 ute Boundary Exercise Full Stroke / Cold Shutdown
,sition Operability TesvSee Note B Remote Position Indication, Exercise /2 Years 3.31 are Bouniary Exercise Full Strcke/ Cold Shutdown ,sition Operability Test /See Note isition 13 Remote Position Indication, Exercise /2 Years isition B Remote Position Indication, Exercise /2 Years rsition B Remote Position Indication Exercise /2 Years isition B Remote Position Indication, Exercise /2 Years til:d B Remote Position Indication, Exercise /2 Years 4, 31 ute Boundary Exercise Full Stroke / Cold Shutdown isition Operability Test /See Note ANSTEC tiled kemote Position Indication, Exercise /2 Years 4,31 tre Bound:ry B
Exercise Full Stroke / Cold Shutdown APERTURE sition Operability Test /See Note CARD liled B Remote Position Indication, Exercise /2 Years 4, 31 are Boundary Exercise Full Stroke / Cold Shutdowr, Also Available on sition Operability Test /Sce Note Aperture Card
' ailed B Remote Position Indication, Exercise /2 Years 4,31 are Boundary Exercise Full Stroke / Cold Shutdown sitioi Operability Test /See Note BC Inspect and keplace/5 Years BC Inspect and Replace /5 Years AB Remote Position Indicaticn, Exercise /2 Years 15,30,31 ne Boundary Exercise Full Stroke / Cold Shutdown Lc:Lage Pressure isolation Leak Test / Refueling Shutdown sition Operability Test /See Notes AB Remote Position Indication, Exercise /2 Years 15,30,31 tre lloundary Exercise Full Ftrokc/ Cold Shutdown Lc:Lage Pressure Isolation Leak Test / Refueling Shutdown sition Operability Test /See Notes AB Remote Position Indication Exercise /2 Years 15,16,30, tre Boar.dary Exercise Full Stroke / Cold Shutdown 31 it Isol: tion - Pressure isolation Leak Test / Refueling Shutdawn Le:Lage Operability Test /See Notes sition
/ /
/,4 l ~
Revision: 20 Draft,1998 ommici ris wpf.oiisos 3.9-151
ikj
.a Table 3.9-16 'cbeet VALVE INSERVICE TEST l RNS-PL-V002B RNS flot Leg Suction and Containment Isolation - Outer Remote Maintain Close Active Transfer Close RCS Press l Containme l Safety Sea Remote P T Westinghouse
- 3. Design of Struct:res, Components, Eq:1pment, cnd Syst:ms l
of 19) QUIREMENTS AB Remote Position Indication, Exercise /2 Years 15,16,30, Bandary Exercise Full Stroke / Cold Shutdown 31 Isolation Pressure Isolation Leak Test /R-fueling Shutdown
- Lage Operability Test /See Notes tion ANSTEC APERTURE CARD-
. Also Awegable ou 0/250/7 '
Revision: 20 Draft,1998 OMNntl6 tl8 wpf Oll598 3.9-152 l l
ua '. Desig:: of Structures, Components,1:q:ipment, cnd Systems Table 3.916 (Sheet VALVE INSERVICE TFSU Yalie Tag Safety Related Number Description W Vahe Type hfissions Safet RNS PL V003A RCS Pressure Boundary Valve Thern,al Relief Check Maintain Close Active Transfer Open RCS Prec Transfer Close RNS PL-V003B RCS Pressure Boundary Valve Thermal Relief Check Maintain Close Active Transfer Open RCS Pres Trarssfer Close RNS-PL-V0l l RNS Discharge Containment Isolation Valve - ORC Remote Maintain Close Active Transfer Close Containm< Safety Sc. Remote P, RNS PL V013 RNS Discharge Containment Isolation IRC Check Maintain Close Active Transfer Open Containm< Transfer Close Safety See i RNS PL-V015A RNS Discharge RCS Pressure boundary Check Maintain Close Active l Transfer Close RCS Pres - l Safety sea l RNS-PL V0158 RNS Discharge RCS Pressure Boundary Check Maintain Close Active
! Transfer Close RCS Press l
Safety Sea l RNS PL V017A RNS Discharge RCS Pressure Boundary Check Mainnin Close Actise I Transfer Open RCS Presq l Transfer Close Safety Sea l RNS.PL V017B RNS Discharge RCS Pressure Boundary Check Maintain Close Active I Transfer Open RCS Presa l Transfer Close Safety Sch RNS PL V021 RNS liot Leg Suction Pressure Relief Re'ief Maintain Close Active Transfer Open Containnd Transfer Close Safety Sch RNS-PL-V022 RNS Suction IIcader Containment Isolation ORC Remote Maintam Close Active Transfer Close Containme Safety Sec Remote Po RNS PL-V023 RNS Suction from IRWST Containment Isolation Remote Maintain Close Active Transfer Close Containme Safety Seal Remote Po RNS-PL-V045 RNS Pump Discharge Relief Relief Maintain Close Active Transfer Open Transfer Close T Westinghouse
F~ m 2 of 19) SQUIREMENTS AS$1E IST Functions (2) Category Insenice Testing Type and Frequency IST Notes BC Check Esenise/ Refueling Shutdown 23 rre Boundary BC Check Exercise / Refueling Shutdown 23
;te Boundary A Remote Position Indication, Exercise /2 Years 27,30,31 Isolation Containment Isolation Leak Test (See Notes)
Lc:Lage Exercise Full Stroke / Quarterly
.ition Operability Test / Scc Notes AC Containment isolation Leak Test (See Notes) 27 aIsolrtion Check Exercise / Quarterly leak:ge ABC Check Exercise / Refueling Shutdown 24 ANSTEC
- re Bound:ry Pressure Isolation I cak Test / Refueling ShutJown Ir kare APERTURE ABC Check Exercise / Refueling Shutdown 24 CARD m Boundary Pressure Isolation Leak Test / Refueling Shutdown
'le:krge Also Available on ABC Check Exercise / Refueling Shutdown 24
'7e Boundary Pressure Isolation I cak Test / Refueling Shutdov>n Lc:k:ge ABC Check Exercise / Refueling Shutdown 24 33 Bound:ry Pressure Isolation Leak Test / Refueling Shutdown Le9e;e AC Containment Isolation Leak Test /2 Yean, 17,27
- t Isol
- tion Class 2/3 Relief Valve Tests /10 Years and 20% in 4 Years Irk:g A Remote Position Indication, Exercise /2 Years 27,30,31 t isol: tion Containment l<olation Leak Test (See Notes)
Leakage Exercise Full Stroke / Quarterly ition Operability Test /See Notes I A Remote Position Indication, Exercise /2 Years 17,27,30, h Isolation Containment isolation Leak Test (See Notes) 31 Leakuge Exercise Full Stroke. Quarterly .ition Operability Test /Sce Notes BC Class 2/3 Relief Valve Tests /10 Years and 20% in 4 Years f- , Revision: 20 Draft,1998 oxenettris .vai ts9a 3.9-153
- .- . . .. - . ~ - - - -- .- . - _ _ - - .. -- '.-, -- 3, Design of Structrres, Ccmpone:ts, Equipme t,ind Systems
- 2. Valves listed as having an methe or an active.to-failed safety related function provide the safety related. valve transfer .
cambihties identified in the safety related mission column. Valves having an active-to-failed function will transfer to the po ition identined in the safety related ruission column on loss of motive power,
- 3. This note aoplies to the AOS stage 1/2/3 valves (RCS-V001 A/B, V002A/B, V003A/B, V011 A/B, V012A/B, V013A/B), Hese values are normally closed to. maintain the RCS pressure boundary. Dese valves have a safety.related function to open fo' sowing LOCAs to allow safety injection from lower pressure water supplies (accumulators and IRWST). Rese valves ako hoe beyond design basis functions to depressurize the RCS Dese salves have the same design pressure as the RCS and E
. ae AP600 equipment class A. Downstream of the second valve is a lower design piessure and is equipment class C. The-discharge of these valves is open to the containment through the IRWST.
Both ADS valve: in each line are normally closed during normal reactor operation le s ordance with 10CFR50.2 and ANS/ ANSI $1.1. If one of these valves is opened, for example for testing, the RCS ptr e houndary is not maintained in accordance with the criteria contained in these two documents. In addition, the ADS va configuration is similar to the normal residual heat remosal system suction valve configuration. Even though the RNS suction valve conGguration includes a third vahe in the high pressure portion of the line, and the first two RNS salves have safety related functions to transfer closed, they are not stroke tested during normal reactor operation to avoid a plant configuration where the mispositioning of one valve would cause a LOCA. Note 15 describes the justification for testing the RNS valves during cold shutdown. Dese ADS valves rae tested during cold shutdowns when the RCS pressure is reduced to atmospheric pressure so that mispositioning of a single valve during this IST will not cause a LOCA. Testing these vahes every cold shutdown is consistent with the AP600 PRA which assumes more than 2 cold or refueling shutdowns per year.
- 4. His note applies to the reactor vessel head vent solenoid vahes (RCS-V150A/B/C/D). Exercise testing of these valves at power represents a risk of loss of reactor coolant and depressurization of the RCS Lhe proper test sequence is not followed.
Such testing may also result in the vahes developing through seal leaks. Exerck ierkg of these valve 5 will be performed at cold shutdown.
- 5. His note applies to squib valves in the RCS and the PXS. He squib valve charge is removed and test fired outside of valve.
Squib valves are not exercised for inservice testing, neir position indication sensors will be tested by local inspection. l 6. This note applies a the CVS reactor coolant pressure boundary isolation valves (CVS.V001, V002, V003, V080, V081, V082). Closing these vahes at power will result in an undesirable temperature transient on the RCS due to the interruption of purification flow. Derefore, quarterly exercise testing will not be performed. Exercise testing will be performed at cold shutdown. 7, his no; app;ies to the pressuriier safety valves (RCS-V005A/B) and to the main steam safety valves (SGS V030A/B, V031 AA V032A/B). Since these valves are not exercised for inservice testing, their position indication sensors are tested by local inspection without valve exercise.
- 8. His note applies to CVS valve (CVS V081). He safety functions are satisfied by the check valve function of the valve.
- 9. His note applies to the PXS accumulator check valves (PXS V028A/B, V029A/B). To exercise these valves, flow must be -
provided through these valves to the RCS. Rese valves are not exercised during power operations because the ac umulators cannot provide flow to the RCS since they are at a lower pressure, in addition, providing flow to the RCS durMg power operation would cause undesirable thermal transients on the RCS. During cold shutdowns, a full flow stroke test is impractical because of the potential of adding significant water to the RCS, and lifting the RNS relief valve. There is also a risk of
. inject'.ig nitrogen into the RCS.' A partial stroke test is practical during longer cold shu: downs (248 hours in Mode 5). In this test, flow is provided from test connections, through the check valves and into the RCS, Sufficient flow in not available i
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y 'i _ to provide a detectable oL.urator movement. Full stroke exercise testing of these valves is conducted during refueling thutdowns.
- 10. nis note applies to the PXS CMT check vahes (PXS V016A/B, V017A/B). Rese check vahes are biased open + sives and gre sully open during normal operation. These valves will be verified to be open quarterly. In order to exercise these check v:hes, significant reverse flow must be provided from the DVI line to the CMT. Dese valves are not tested during power operations because the test would cause undesirable thermal transients on the portion of the ime at ambient temperatures and ch:nge the CMT boron concentration nese 5abes are not exercised during cold shutdowns because of changes that would result in the CMT boron concentration. Because this parameter is controlled by Technical Specificauons, this testing is impractical. Dese vahes are exercised during refueling when the RCS boron concentration is nearly equal to the CMT concentration and the plant is in a mode where the CMTs are n:t required to be available by the Technical Specifications.
- 11. His note apphes to the PXS containment recirculation check vahes (PXS Vil9A/B). Squib vahes in line with the check vahes present the use of IRWST water to test the vahes. To exercise these check valves an operator must enter the containment, remove a cover frorn the recirculation screens, and insert a test desiee into the recirculation pipe to push ope the check valve. The test device is made to interface with the vahe without causing valve damage. The test device incorporates loads measuring sensors to measure the initial opening and full open force. These vahes are not exercised during power operations because of the need to enter highly radioactive areas and because during this test the recirculation screen is bypassed. Dese vahes are not exercised dunny cold shutdow n operations for the same reasons. Tnese vahes are exercised during refueling conditions when the recirculation lines are not required to be available by Technical Specifications LCOs 3.5.7 and 3.5.8 and the radiation levels are reduced. ,
- 12. His note applies to the PXS IRWST injection check valves (PXS V122A/B, V124 A/B). To exercise these check valves a test cart must be moved into containment and temporary connections made toh t ese check vahes. In addition, the IRWST injection line isolation valves must have power restored and be closed. These valves are not exercised during power operations because closing the IRWST injection s ahe is not permitted by the Technical Specifications and the need to perform si Fnificant wo L inside containment. Testing is not performed during cold shutdown for the same reasons, nese valves are exercised during refueling conditions when the IRWST injection lines are not required to be available by TechnicC Specifications and the radmtion levels are reduced.
- 13. Deleted.
- 14. Component cooling water system containment isolation motor-operated vahes CCS-V200, V207, V208 and check valve CCS-V201 are not exercised during power operation. Exercising these salves would stop cooling water flow to the reactor cwlant pumps and letdown heat exchanger. Loss of cooling water may result in damage to equipment or reactor trip. Rese v:hes are exercised during cold shutdowns when these components do not require cooling water.
- 15. Normal residual heat removal system reactor coolant isolation motor operated valves (RNS V001 A/B, V002A/B) are not exercised during power operation. These sabes isolate the high pressure RCS from the low pressure RNS and passive core cooling system (PXS). Opening during normal operation may result in damage to equipment or reactor trip. Rese salves are exercised during cold shutdowns when the RNS is aligned to remove the core decay heat.
- 16. Normal residual heat removal system containwnt isolation moto:-operated valves (RNS-V002A/B) are not containment isolation leak tested. De basis for the exception is:
He valve is submerged during post accident operaticos which prevents the release of the containment atmg radiogas or acrosol. He RNS is a closed, seismically-desigud safety class 3 system outside containment De valves are closed when the plant is in modes above hot shutdown APERTURE 9 r rd Ohr.tl6 rlB wpf.Oll59s 3.9-169
- --- 3. Desig; of Struet:rcs, Components, Equipment, cnd Systems
- 17. Normal residual heat removal system containment penetration reber valve (RNS V021) and comainment isolation motor operated salve (RNS V0231 are subjected to containment leak testing by pressurizing the lines in the reserse direction to the Dow s.hsch accompanies a containment leak in this path.
- 18. His note apphes to the CAS instrument air containment isolation vahes (CAS V014, V015). It is not practicalIc curcise these vahes during power operation or cold shutdowns. Exercising the valves during these conditions may trault in sone air-operated $ ahes inadscriently opening or closing, resulting in plant or system transients. These sahes are esercised during refueling conditions when system and plant transients would not occur.
- 19. Primary sampling system containment isolation check vahe (PSS.V024)is hicated inside containment and considerable effort is requ red to irntall test equipenent and cap the discharge line. Exerche testing is not performed during cold shutdown operations for the same reasons. nese s ahes are exercised during refuchng conditions w hen the radiation levels are teduced.
- 20. His note applies to the main steam isolation vahes and main feedwater isolation vahes (SGS-V040NB, V057A/B). ne sabes are not full stroke tested quarterly at power since fuh sabe stroking will result in a plant tramient during normal power operation. Therefore, these salves will be partially stroled on a quartes'c, basis and will be full stroke tested on a cold shutdown frequency basis. The full stroke testing will be a full " slow" closure operation. The large size and fast stroking nature of the sabe makes it adsantageous to limit th; number of fast closure operations which the valve experiences. De timed slow closure senfies the vahes oparability status and that the salve is not mechanically bound.
- 21. Post 72 hour check sahrs that require temporary connections for inservice testing are exercised escry refueling outage These sahes require tra.nport and installation of temporary test equipment and pressure /Duid supplies. Since the vahes are normally used very infrequently, constructed of stainless steel, maintained in controlled environments, and of a simple design, there l is little benefit in testing them more frequently. For example, $ahe pCS-V039A is a simple vahe that is opened to provide the sdJition of water to the PCS post-72 hout from a temporary water supply. To exercise the valve, a temporary pump and water supply is connected using temporary pipe and fittings, and the Dow rate is observed using a temporary now measuring device to confirm sabe operation.
- 22. Exerci e testing of the auxiliary spray isolation vahe (CVS V084, V085) will resu't in an undesirable temperature transient on the pressurizer due to the actuation of auxiliary spray flow. Therefore, quarterly exercise testing will not be performed.
Exercise testing will be performed during cold shutdowns.
- 23. Thermal relief check sabes in the normal residual heat remos al suction line (RNS V003A/B) and the Chemical and Volume Control System makeup line (CVS.V100) are located inside containment. To exercise test these vahes, entry to the containment is required and temporary connections made to gas supplies. Because of the radiation exposure and effort required, this test is not conducted di. .ng power operation or during cold shutdowns. Exercise testing is performed during refueling shutdowns.
- 24. Normal residu;n heat removal system reactor coolant isolation check valves (RNS V015A/B, V017A/B) are not exercise tested quarterly. During normal powet operation these vahes isolate the high pressure RCS from the low pressure RNS. Opening during normal operation would require a pressure greater than the RCS normal pressure, which is not as ailable. It would also subject the RCS connection to undesirable trnsients. These valves will be exercised during cold shutdowns.
- 25. His note applies to the main feedwater control .alves (SGS.V250A/B), moisture separator reheater steam control valve (MSS-V020), turbine control vahes (MTS V002A/B, V004 A/B). De valves are not quarterly stroke tested since full stroke testing would result in a plant transient during power operation. Normal feedwater and turbine control operation provides a partial stroke confirmation of valve operability. De valves will be full stroke te :ed during cold shutdowns.
W Westinghouse
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iE iiij; , _ 26: This note appli : to containment compartmen' drain line check vahes (WLS.V071 A/B/C, V072A/B/C). Tnese check valves are located inside containment and require temporary connections for exercise testi..g. Because of the radiation exposure and effort required, these valves are not exercised during pow:r operation or during cold shutdow ns. The vahes will be exercised during refuelings.
- 27. Containn.ent isolation vahes leakage test frequency wi'. be conducted in accordance with the "t rimary Containment Leakage ,
Rate Test Program" in accordance e ith 10 CFR 50 Appendix J. Refer to SSAR subsectio,i 6.2.5.
- 28. This note applies to the chilled water system containment i olation vahes (VWS.V058,V062, V082 and V086). Closing any of these vabes stops the water now to the containment fan coolers. His water now may be necessary to maintain the containment air temperature within Technical SpeciGeation limits. As a result, quarterly exercise testing will be deferred w hen plant operating conditions and site climatic conditions would cause the containment air temperature to exceed this limit during +
testing.
- 29. Exercise testing of the tu:bine bypass control valves (MSS.V001, V002, V003, V004) will result in an undesirable temperature .
transient on the turbine, condenser and other portions of the turbine bypass due to the actuation of bypass flow. Therefore, quarterly exercise testing will not be performed. Exercise testing aill be performed during cold shutdowns.
- 30. nese vahes are required to operate with low differential pressure. The Combined License applicant will provide an evaluation based on test data to scrify that the valves have adequat: margin and operability testing is not required. De test data may include data from type tests. See subsection 3.9.8.4 fnr the Combined License applicant infonnation item.
31 Dese valves may be subject to operabihty testing. See subsection 3.9.6.2.2 for the factors to be considered in the evaluation of operability testing and subsection 3.9.8.4 for the Combined License information item. He specified frequency for opernbility testing is a masimum of once escry 10 years. He test frequency is the longer of every 3 refu-ling cycles or 5 years until suf0cient data exists to determine a longer test frequency is appropriate in accordance with Generic Letter 96-05.
- Some of the valves will be tested the Grst time after a shorter period to provide for trending information.
- 32. Dese valves are subject to operability testing. He operability testing will be conducted using nonintrusive techniques to ossess valve operability under now and differential pressure test conditions. The test frequency is the longer of every 3
) refueling cycles or 5 years.
- 33. His note applies to valve FilS V001. His valve closes one end of the fuel transfer tube. It must remain closed during operation to minimite the potential for draining water from the fuel transfer canal in to the refueling cavity.
\ , 34. His note applies to the main feedwater control vahes (SGS-V250A/B), moisture separator reheater steam control valve (MSS-V020), turbine control valves (MTS V002A/B, V004 A/B), main turbine stop valves (MTS.V001 A/B, V003AIB), the turbine bypass control valves (MSS V001, V002, V003, V0GI) . Rese valves are not safety related. Rese valves are relied on in the safety analyses for those cases in which the rupture of the mrin steam or feedwater piping inside containment is the postulated initiating event. These vahes are credited in single failare analysis to mitipte the event.
- 35. Dis note applies to the turbine stop valves (MTS-V001 A/S, V003A/B). De valves are not quarterly stroke tested since full l
stroke testing would result in a plant transient during power opention. The valves will be full stroke tested during cold - shutdowns. , 36 In each of the four turbine inlet lines, there is a turbine stop valve aad turbine control valve. Only one of the valves in each of the four lines is required by Technical Speci0 cation 3.7.2 to be operable. on b() r. -no I/
- - ,. n m m.,
vb APERTURE Resim 20 CARD "" 03mnite ris wpt.oiis9s 3.9-17. Also Aveliable on Aperture Card _ _ . _ _ _ __. ._ _. _.~ .
NRC FSER OPEN ITEM Question 410.414F (OITS . 6520) The staff's comments on the main control room habitability system technical specification are reflected in the mark-ep to TS 3.7.6 provided in enclosure 2 to this letter. Westinghouse is requested to incorporate these chuges. This is an open item.
Response
Items a through g summarize the changes provided in the NRC mark up. A response is provided for each item. a) LCO Condition D (8.CO and Bases) Condition D is marked up to add Required Action D.2: Be in MODE 4 with a Completion Time of 12 hours.
Response
The AP600 Required Actions D.i (Be in A10DE 3 (within 6 hours)) and D.2 (Be in A10DE 3 (within 36 hours)) are consistent with STS NUREG-1431. NUREG 1431 does not include any precedents for spectfication of of0DE 4 in actions which end in Af0DE 3. The most directly applicable NUREG 1431 precedent is LCO 3.7.10. Control Room Emergency Filtration System (CREFS). Condition B, which is identical to the existine AP600 Condition D. No changes to AP600 Technical Specifications are required. b) SR 3.7.6.1 (LCO and Bases) SR 3.7.6.1 is marked up to change the temperature to 5 75'F. The second sentence of the SR 3.7.6.1 Bases is marked as follows: The surveillance limit of 75'F is the initial MCR envelope heat sink temperature used in the VES thermal analysis.
Response
The requested changes will be incorporated in the AP600 Technical Specifications. 410.414F-1
NRC FSER OPEN ITEM c) SR 3.7.6.5 (LCO and Bases) SR 3.7.6.5 h marked up to change dampers to valves. The first sentence of the SR 3.7.6.5 Bases is marked up to change devices to valves. 1 Response - The requested changes will be incorporated in the AP600 Technical Specifications. d) SR 3.7.6.9 SR 3.7.6.9 is marked up to ead as follows: Verify that one VES air delivery train maintains an 1/8 inch water caugg positive pressure in the MCR envelope relative to the adjacent areas at the required air addition flow rate of 60 d scfm using the safety related compressed air emergency air storage tanks. The Bases discussion of SR 3.7.6.9 is marked up to reflect the pressure and flow rate requirements.
Response
The requested additions to SR 3.7.6.9 will be incorporated by reference to the System Level Operability Testing Program specijled in the SSAR asfollows.
- Verify that one VES air delivery train maintains a positive pressure in the MCR relative to the adjacent areas at the required air addition jlow rate in accordance with the System Level Operabiliy Testing Program.
The requested changes wi!. ie incorporated in the AP600 Technical Specifications. e) SR 3.7.6.10 New surveillance, SR 3.7.6.10. has been added as follows: Verify that the air quality of the air storage tanks meets the requirements of Appendix C, Table C 1 of ASHRAE Standsrd 62. Frequency - Quarteily t 410.414F-2 T Westinghouse m e 4
m *E F NRC FSER OPEN ITEM i
Response
SR 3,7.6.10 will be added as requested, except that the Frequency will be specified as 92 days, and the surveillance will be placed in order with the existing surveillances aaording to Frequency as SR 3.7.6.5 (and the subsequent SRs renumbered), consistent with NUREG 1431. The following Bases di.cussion of this new suneillance is included in the mark up: SR 3.7.6.5 Verification that the air quality of the air storage tanks meets the requirements of Appendix C, Table C 1 of ASHRAE Standard 62 is required every 92 days. If air has not been added to the air storage tanks since the previous veripcation, verification may be accomplished by confirmation of the acceptabiliy of the previous surveillance results along with examination of the documented re:ord of air make up. The Purpose of ASHRAE Standara 62 states: "Tnis standard specifies minimum ventilation rates and indoor air quality that will be a:ceptable to human occupa,,.ts and are intended to minimize the potentialfor adverse health effects." Veripcation of the initial air quality (in combination with the other surveill . aces) ensures that breathable air is availablefor 1i AfCR occupantsfor at least 72 hours. f) SR 3.7.6.II A new surveillance, SR 3.7.6.11, is proposed in the NRC mark up. Verify that the maximum unfihered Mr in-leakage (infiltration) into the MCR envelope under accident conditions is 5 scfm when the VES is operadng and 140 scfm when only VBS is operating during "high" gaseous radioactivity signal in accordance with ASTM E741. Respome Considering that the existing AfCR VES pressurization surveillance demonstrates the integrity of the envelope, additional testing for AfCR in leakage is not needed. See response to RAI 410.379F. g) Bases Mark-up In addition to the LCO-related Bases changes discussed above, the NRC Bases mark-up includes changes to the VES system description and to the initi:d MCR temperature (75*F). 410.414F 3
i - 1 (? R NRC FSER OPEN ITEM l l
Response
The NRC proposed Bases changes are in general agreement with the Bases mark-up provided by i
. Westinghouse letter DCP/NRC107), dated October 10,1997. A r.ew Bases mark up has been prepared which combines the NRC and Westinghouse mars ups.
SSAR Revision: See at'xhed markup. 410.414F-4 T Westinghouse
Main Control- Room Habitability System (VL)
- * 'O
( M AVO.4/4 F > 3.7 PLANT SYSTEMS
-3.7.6 Main Control Room Habitability System (VES)
TWE LCO.3.7.6 -Two Main Control Room (MCR)' Habitability System 4pe,wshall be OPERABLE. APPLICABILITY: MODES 1. 2, 3. and 4. During movement of irradiated f Je1 assemblies. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I vna oA com: - A. OneVES4ee4n1 A.1 Restore VES4 4ee4a to 7 days inoperables OPERABLE status.
~8. MCR air temperature ? B.1 Restore NCR air 24 hours not within' limit. ~ temperature to within limit.
C. Loss of integrity of C.1 Restore MCR pressure 24 hours L MCR presst.re boundary to OPERA 8LE E boundary, status D.- Required Action and 0.1 Be in MODE 3. 6 hours
- i. associated Completion - -
t Time of Conditions A, AND B or C not met in-MODE 1. 2, 3, or 4. D.2 Be in MODE 5. 36 hours-E. Required Action and E.1 Suspend CORE Immediately
. associated Completion ALTERATIONS.
Time of Conditions- A, l B, or C not met _ AND , during movement of irradiated fuel. E.2 Suspend movement of_ Imediately l irradiated fuel
- i. assemblies.
L I-l L (continued) k,,.AF60.0,- 3.7 12 08/97 Amendnent 0'
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- . - - - .._ ~ .
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Main Control doom Habi%ab111%y System (VES) R&Z Wo. +/9 F 3.7.6 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TlHE F. JweVES t:;;.,e F.1 Be in H00E 3. 6 hours inoperable in Mu0E 1.
- 2. 3. or 4. AND F.2 Se in MODE 4. 12 hours M
F.3 Restore o m.VES tee 4e to 36 hours OPERABLE status. G. Joe' VES tee 4w G.1 Suspend CORE Innediately inoperable during ALTERATIONS. movement of irndiated fuel. @ G.2 Suspend movement of Inmediately irradiated fuel ' assemblies. 3.7 13 08/97 Amendment 0
@ AP600..
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Main Control Room Habitability System (VES) 3.7.6 l Q+I 4/o . 4/ f SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify Main Control Room air temperature 24 hours i s < 46f F.
~ L- 7s SR 3.7.6.2 Verify that the compressed air storage 24 hours tanks are pressurized to Q 100 psig].
SR 3.7.6.3 Verify that each VES air delivery In accordance isolation valve is OPERABLE. with the Inservice Testing Program SR 3.7.6.4 Verify that each VES air header manual 31 days gg isolation valve is in an open position.
, sR3.7.L.S .
vis u s s SR 3.7.6. Verify th all VBS Main Control Room 24 months isolation dampose 3 are OPERABLE and will close upon receipt of an actual or simulated actuation signal. SR3.7.6.h Verify that each VES pressure relief In accordance isolation valve within the MCR pressure with the boundary is OPERABLE. Inservice Testing Program art 3.7.6. Verify that each VE'S pressure relief 24 months damper is OPERABLE. (continued)
@e,.-
AP600 3.7 14 02/97 Amendment 0 eunw .w am, q ,3 g4 y ,,7
i
- i i
RAI 410.414F LCO 3.7.6 - Main Control Room Habitability System (VES) INSERT SR 3.7.6.5 - PAGE 3.714 SR 3.7.6.5 [ t Verify that the air quality of the air storage tanks meets the requirements of ! Appendix C, Table C 1 of ASHRAE Standard 62.- Frequency 92 days'
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Main Control Room Habitability System (VES) Ar+[ 4/0, 41if F SURVEILLMCE REQUIREMENTS (cortinued) SURVEILLANCE FREQUENCY SR 3.7.6. Verify that the self contained pressure In accordance regulating valve in each YESatretw is with the OPERABLE. Ad oeuvidV Inservice
*" W Testing Program to .
ru.a Nrn SR 3.7.6./ Verify that one VES air deriver 24 months maintains a positive pressure i the MCR, relative to the adjacent areas 46-tM T@iists eir ^II'.ti^" M " 4%e.
/Al Aftalette%E Wo7t/
7HE fYSTEMfs/EL g M / 4 dito r y p s % ,r.n ac e s M. l l l l l l I HAP 600 3.7 15 08/97 Amendient 0
= % = = = ,c ,
4 1 0 .'4 i 4 F 9
Main Control Room Emergency Habt%Abili%y System R&r WO . J/t+ F B 3.7 PLANT SYSTEMS B 3.7.6 Main Control Room Emergency Habitability Sy..em BASES BACKGROUND The Main Control Room Habitability System (VES) provides a protected envirornent from which operators can control the plant following an uncontrolled release of radioactivity. The system is designed to operate following a Design Basis Accident (DBA) which requires protection from the release of radioactivity. In these events, the Nuclear Is1cnd Non. Radioactive Ventilation System (VBS) would continue to function if AC power is available. If AC power is lost or a High 2 main control room (MCR) radiation signal is received, the VES is actuated. The major functions of the VES are:
- 1) to provide forced ventilation to deliser an adequate supply of breathable air for the MCR occupants: 2) to provide forced ventilation to maintain the MCR at a 1/8 inch water gauge positive pressure with respect to the surrounding areas; and 3) to limit the temperature increase of the MCR equipment and facilities that must remain functional during an accident, via the heat absorption of passive heat
- h Lee 4.< osm'Mg7W-The VES consists of M 71.
_..[s. and ; ira eechEech instrumentation. wit L+rcompressed sociated valves,air ofstanks contains enough storage piping, breathable air to supply the requ, ired air flow to the MCR fe. at least 72 hours. The VES system is designed to g maintain C,0, concentration itss than 0.5% for up to 11 MCR ,
' occupants.pntnfooth tr s operst WitWone operati VE5 maintains concent It % 1ess tha
- for up to 5 pants, an( ntains.C0, concentration ss 1 Whn1.0(for _
to 11 MCR occuphnts.i Sufficient thermal mass exists in the surrounding concrete structuce (includi walls, ceiling and floors) to absorb 75 tho heat generated fnside the MCR. which is initially at or be' F. Heat sources inside the MCR include o rator work ations, emergency lighting and occupants. S fficient insulation is provided surrounding the MCR pressure boundary to preserve the minimum required thermal capacity of the heat sink. The insulation also limits the heat gain from the adjoining areas following the loss of VBS cooling. (continued) honAP600 ounne.,oro a., B 3.7 26 08/97 Amendment 0 qi o.4i 4 F -Io
I
- Main Control Room fmergency Habitabiltgy System !
B 3.7.6 R4I WC .4/y F BASES BACKGROUND if tt; O r; ;ine scovetid,1,-felb;ing tt,; 77 iwr perter, (continued) eeelir,g cf tt HCR ir=ii edne.g, by pc-rt:ble air ccg,1;re, i N5M The compressed air storage tanks are initially pressurized to 3400 psig. During operation of the VES, a self contained pressure regulating valve maintains a constant downstream pressure regardless of the upstream pressure. An orifice downstream of the regulating valve is used to control the air flow rate into the MCR. The MCR is maintained at a 1/8 inch water gauge positive pressure to minimize the infiltration of airborne contaminants from the surrounding areas. APPLICABLE
74 s 7se SAFETY ANALYSES suchthatIee,c,eeMlcom>ressedairstoragetanksaresized h set of tants has a combined capacity that provides at least 72 hours of VES operation. ,
1 Operation of the VES is automatically initiated by either of l two safety related signals: 1) undervoltage to Class 1E ' battery charger, or 2) high 2 particulate or iodine radioactivity. . In the event of a loss of all AC power, the VES functions to l provide ventilation, pressurization, and cooling of the MCR ! pressure boundary. In the event of a high level of gaseous radioactivity outside of the MCR. the VBS continues to operate to provide pressurization and filtration functions. The MCR air supply downstream of the filtration units is monitored by a safety , related radiation detector. Upon == M i-; ;
-"W undervoltage to Class 1E battery charger or hi"gh.2 particulate or iodine radioactivity setpoint, a safety related signal is generated to isolate the MCR from the VBS and to initiate air flow from the VES storage tanks.
Isolation of the V85 consists of closing safety related
# 7 4empers in the supply a.id exhaust ducts that penetrate the MCR pressure boundary. VES air flow is initiated by a safety related signal which opens the isolation valves in the VES supply lines.
The VES functions to mitigate a DBA or transient that either assumes the failure of or challenges the integrity of the fission product barrier. (continued)
@ AP600 8 3.7 27 08/97 Ameneent 0 .
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i RAI 410.414F LCO 3.7.0 Main Control Room liabitability System (VES) INSEltT BACKGROUND l' AGE B 3.7 27 in the unlikely event that power to the VBS is unavailable for more than 72 hours. MCR envelope habitability is maintained by operating one of the two MCit ancillary fans to supply outside air to the MCR envelope. 4toA M P- q
~
, Main Control Room Emergency Habitabili2y System Rfr .ho myf <
8ASES APPLICABLE The VES satisfies the requirt.ments of Criterion 3 of the NRC l SAFETY ANALYSES Policy 5thtement. ' (continued) LCO The VES limits the NCR temperature rise and maintains the HCR at a posit %e pressure relative to the surrounding environment. AsR., OFLW141 FLdd hfTW3 Two i:4 .._.... .. . ,~...;. ~C tiei.a are required 1o be OPERABLE to ensure that at least one is available, asstming a single fa11ure,46e:M :: 5 : i t nis. The VES is considered OPERABLE when the individual components necessary to deliver a su to the MCR are OPERABLE.9 _... ;..;pply of breathable ne. This includes air components listed in SR 3.7.6.2 through 3.7.6.8. 'In addition, the MCR pressure boundary must be maintained, i including the integrity of the walls, floors, ceilings., duet- ' W . electrical and mechanical penetrations, and access doors. APPLICABILITY The VES is required to be OPERABLE in H00ES 1, 2, 3. and 4 and during movement of irradiated fuel because of the , potential for a fission product release following a DBA. The VES is not required to be OPERABLE in MODES S and 6 when irradiated fuel is not being moved because accidents resulting in fission product release are not postulated. ACTIONS A.1 p vnug et perYR. cwe When +ne VES trei s inoserable, etion is required to restore theleysteur o OPEMBLE sta us. A Com)1etion Time of 7 days is permitted to restore the before action m.st be taken to redu;peir, to 0)ERABLE ce power. status The Completion 1 Time of 7 days is based on engineering judgmerit, considering i the low probability of an accident that would result in a l significant radiation release from the fuel, the low l probability of not containing the radiation, and that the remaining testa can provide the required capability. L entwMn i ! (continued) l M AP600 8 3.7 28 08/97 Amendment 0 _ _ ** P "'"'"' .___ _ ._.. .-_ Aid d 'I
_ . . _ _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ Main Control Room Emergency Habi%Abil1%y System Q.T W O .'tI + F BASES ACTIONS B.1 ' (continued) When the main control room air temperature is outside the i acceptable range during VBS operation, action is required to restore it to an acceptable range. AC letion Time of 24 hours is permitted based upon the avai ability of temperature indication in the MCR. It is judged to be a i sufficient amount of time allotted to correct the deficiency in the nonsafety ventilation system before shutting down. i C.1 If the MCR pressure boundary is damaged or otherwise degraded, action is required to restore the integrity of the 4 pressure boundary and re: tore it to OPERABLE status within 24 hours. A Completion Time of 24 hours is persitted based upon operating experience. It is judged to be a sufficient amount of time allotted to correct the deficiency in the pressure boundary. 0.1 and 0.2 ' In MODES 1, 2, 3 or 4 if Conditions A, B, or C cannot be restored to OPERABLE status within the required Completion Time, the plant must be placed in a MODE that minimizes accident risk. This is done by entering MODE 3 within 6 hours and MODE 5 within 36 hours. , E.1 and E.2 g {jng y emy t p {M!a Q f ass p ie,3 3 }he p,
- Required Actions A.1, 8.1, or C.1 cannot be completed within
+
- the required Completion Time, the movement of fuel and core alterations must be sus . Performance of Required Action E.1 and E.2 shal not preclude completion of actions to establish a safe condition, F.1, F.2, and F.3 73's M A'T 7% is If 4sth VES Aeainsre/ rs inoperable in MODES 1, 2, 3, or 4.
the VES may 6e capable of performing the intended function, st be brought to MODE 4, where the probability and consequences of an event are minimized, and 7 5 Acum russo mim.VES toe $v munt be restored to OPERABLE status within 36 hours. Tiis 'sMone-by ente *+nt MODE 3 within 6 hours and MODE 4 within hours, L 4 7,y w , (continued) I h AP600 8 3,7 29- 08/97 Amendment 0
- >4u=,
q g p., p.
, Main Control Room Emergency Habitability System B 3.7.6 Mr WQ.4/+f BASES ACTIONS G.) and G.2
~'
(continued) mt During movement of irradiated fuel assemblies withino VES
$pe$ns inoperable, the Required Action is to immediately suspend activities that present a potential for releasing radioactivity that might enter the MCR. This places the plant in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.
SURVEILLANCE SR 3.E6.1 75 REQUIREMENTS The'MCit. air temperature is checked at frequency of 24 hours to verify that the VBS is per orming as required to I maintaintheinitialconditiontemperItureassumedinthe safety analysis, and to ensure that MCR temperature will not exceed the required conditions a r loss of m 1 cooling. The surveillance limit of 'Fisthefno#n "M , g - at g ine ya L1s wius cr ter gik s j
/"" "#I L2'F _ _ L- at ur- ai nt . l 'he 24 hour Fr y is 1
acceptable based on the availability of tempera re "K7F'8NfWlindicationintheMCR. Aswmeo 10 wc I VE5 TH4*stL SR 3.7.6.2 ' \ l da+ eses . i ,;
' Verification every 24 hours that compressed air storage tanks are pressurized to & 3400 psig) is sufficient to ensure that there will be in adequate supply of breathable air to maintain MCR habitability for a period of 72 hours.
The Frequency of 24 hours is based on the availability of pressure indication in the MCR. SR 3.7.6.3 VES air delivery isolation valves are required to be verified as OPERABLE. The Frequency required is in accordance with the Inservice Testing Program. SR 3.7.6.4 VES air header isolation valves are recuired to be verified o:en at 31 day intervals. This SR is cesigned to ensure t1at the pathways for supplying breathable air to the MCR are availabic should loss of VBS occur. These valves should be closed only during required testing or maintenance of downstream components, or to preclude complete depressurization of the system should the VES isolation valves in the air delivery line open inadvertently or begin to leak. (continued) HAP 600 8 3.7 30 08/97 Amendment 0 -
--. 4 io .M F-If ._
Main Control Roore Emergency Habitability System
,, Rhr Wo. WVt= B 3.7 6 BASES ISQ. S '7. 6S f
SURVEILLANCE SR 3. 7. 6.g/. REQUIREMEKTS Wt./55 (continued) Verification that all VBS isolation s-=- are operable and will actuate upon demand is required overy 24 months to ensure that the HCR can be isolated upon loss of VBS operation. SR 3. 7. 6.g7 Verification that each VES pressure relief isolation valve within the HCR pressure boundary is OPERABLE is required in accordance with the Inservice Tertirg Program. The SR is used in combination with SR 3.7.6.7 to ensure that adequate vent area is available to mitigate HCR overpressurization. SR 3. 7. 6.)t $ Verification that the VES pressure relief damper is OPERABLE is required at 24 month intervals. The SR is used in combination with SR 3.7.6.6 to ensure that adequate vent area is available to mitigate ER overpressurization. SR 3.7.6.#9 y p g ,g gy ggas fw; Verification of the operability of the elf contained pressure regulating valve in each VES'teate is required in accordance with the Inservice Testing Program. This is done to ensure that a sufficient supply of 61r is provided as l required and that uncontrolled air flow into the HCR will not occur. , SR 3.7.6.5/4 D*BC This-SR r ires the performance of a system performance
/ r --' test of'tha.VES to verify MCR pressurization ca> abilities. f The system p' e rformance test demonstrates that.tw HCR .
pressurization ass'uned in dose analysis is' maintained.
'Althouah the likelihood that system performance would degradellmerf time is-loi.' it is considered prudent to
) periodicajl pverify system performance;'The System Level ( . { OperabkMty Testing Program provides specific test i requirements and acceptance criteria. p (continued) h AP600 B 3.7 31 08/97 Amendnent 0 d#n9W P-(L
1 RAI 410.414F LCO 3.7.6 Main Control Room Habitability System (VES) INSERT B8SES SR 3.7.6.5 PAGE 3.7 31 , i SR 3.7.6.5 Verification that the cir quality of the air storage tanks meets the requirements of Appendix C, Table C 1 of ASHRAE Standard 62 is required every 92 days. If air has not been added to the air storage tanks since the previous verification, verification may be accomplished by confirmation of the acceptability of the , previous surveillance results along with examination of the documented record of air make up. The Purpose of ASHRAE Standard 62 states:"This standard , specifies minimum ventilation rates and indoor air quality that will be acceptable to human occupants and are intended to minimize the potential for - adverse health effects." Verification of the initial air quality (in combination with the other surveillances) ensures that breathable air is available for 11 . MCR occupants for at least 72 hours. INSERT BASES SR 3.7.6.10 PAGE 3.7 31 SR 3.7.6.10 Per reference 1, a functional test is required to establish that one VES air delivery flow path, using the safety related compressed air storage tanks, , pressurizes the MCR envelope to at least a positive 1/8 inch water gauge pressure relative to the surrounding spaces at the required air addition flow ; rate of 65
- 5 scfm. The test need not last 72 hours, only long enough to demonstrate the ability to achieve the required differential pressure. The MCR envelope leakage rate must be within the design capacity of the VES to pressurize the MCR for 72 hours. One air delivery flow path is tested on an alternating basis. The system performance test demonstrates that the MCR '
pressurization assumed in dose analysis is maintained. E g io Wi+F -
Main Control Room Emergency Habitability System B 3.7.6 BASE 3 (continued) REFERENCES 1. AP600 SSAR. Section 6.4,
- Main Control Room Habitability Systems.' '
- 2. AP600 SSAR, Section 9.4.1, ' Nuclear Island Non Radicactive Ventilation System.'
- 3. SECY.95 132, ' Policy and Technical Issues Associated .
With The Regulatory Treatment of Non Safety Systems ' (RTNSS) In Passive Plant Designs (SECY.94 084),* May 22, 1995. l-e 4lO. 4 H F - I
.@ AP600 8 3.7 32 08/97 Amendment 0
1 I pammm ) NRC F3ER OPEN ITEM - L._ e Question 440.756F (OITS . 6413 l Calculation Note SSAR OSC 377 l I In an NRC inspection of Westinghouse AP600 design control activities from November 17 through 21, 1997, the stalf reviewed calculation note SSAR GSC 377 related to WCOBRAfrRAC long term ! cooling calculations and has the following questions:
]
I (a) A large mass discharge from ADS l 3 was noted in two scenarios analyzed in this calculation note. Ifowever, it was noted that the calculating was initialized with the pressurizer empty and, in this . time window, the IRWST should also be empty. If these initial conditions for the window are correct, I how could there be such a substantial ADS l 3 mass now? Because the ADS 13 How is used to calculate the for mass balance, what impact does this have on the validity of the case analyzed (if the ADS l 3 dow is not real)7 (b) In the same calculation it was noted that there was substantial negative now through the DVI lines ; back into the sump. Please explain if this is physically possible. What does this mean for the validity of the case analyzed? Responses (a) De ADS Stage 13 mass discharge referenced occurs during the sump injection phr.se of two two-inch cold leg break AP600 long term cooling scenarios. ECOBRAfrRAC is initialized for these cases with the pressurizer empty, which is consistent with the long term cooling test simulations of Reference 440.756F-l. During the initial, pre 4teady state portion of the analyses, ECOBRA/ TRAC predicts liquid to enter the pressurizer, which is consistent with the fact that "the ECOBRA/ TRAC calculanon is initially a transient until the mass redistributions occur," as noted in subsection 3.7 of Reference 440.756F l. The code overrides whatever initial conditions are input to start a window mode calculation to define the solution for the boundary conditions specified. The insensitivity to input initial conditions was established in subsection 3.10 of Reference 440.756F l. The ECOBRA/ TRAC mass balances for the scenarios referenced show that these cases are valid. : Considering the two-inch cold leg break window of SSAR.GSC 377 reported in SSAR subsection 15.6.5.4C.3.5 (Reference 440.756F 2), the average total injection now rate iito the vessel through the DVI lines is 76 lbm/second. The average total now through the ADS Stage 4 flow paths from the hot legs is equal to the injection flow. Therefore, the venting capability necessary for adequate core l cooling is provided by ADS Stage 4 operation alone. The mass balance between the DV' and ADS How rates is the same for the other SSAR GSC 377 scenario referenced. . A mass balance on the pressurizer dudng the %ection 15.6.5.4C.3.5 sump window analysis is also instructise. When mass _is present in the pre' ~rer, the discharge of same through ADS Stages 13 is to be expected. Mass is discharged through ALS Stages 13 at an average rate of about 4 lbm/second
'during the subsection 15.6.5.4C.3.5 cue. The pressuriter behavior is effectively independent of the 440.756F-1 r
l l iiF in NRC FSER OPEN ITEM reacto senel during this discharge because the preuuriier recches no inlet Acw through the surge line; the now through ADS Stages 13 depletes the preuurirer man insentory. The predicted flows through the ADS Stage 4 How paths and through ADS Stages 13 are understood and are rational in the referenced SSAR GSC 377 long-term cooling scenanos. (b) The ma;nitude and direction of How between the IRWST and/or the sump and the reactor venel through the DVI lines during long term cooling is determined by the pressure balance that esists. If the preuure in the reactor vessel upper plenum temporarily increases to a large enough estent above containment prenure, injection now from the IRTSThump will reverse direction, until the vessel prenure has been reduced by the venting of steam through the ADS Stage 4 How paths. As described in SS AR sub',ections 15.6.5.4C.3.4 and 5 (Reference 440.756F 2), the upper plenum pressure rises when the liquid level is high in the upper plenum and in the hot legs because more liquid (and less steam)is vented through the ADS Stage 4 How paths. Once the liquid level falls slightly within the hot leg perime;c and the quality of the misture being vented through ADS Stage 4 increases, the upper plenum preuure decreases and positive injection through the DVI lines is reestablished. As discussed in Reference 440.756F 2, the AlWX) design provides the necessary venting capacity in the ADS Stage 4 How paths to anure that injection from IRWST and/or sump occurs during long term cooling under conservatise Appendia K auumptions. The AP600 SSAR long term cooling analysis results make physical sense and demonstrate that the performance of the AP6(X) complies with the 10CFR$0.46 requirements for long-term core cooling. Referenees: 440.756F l: Garner, D. C. et.al., "ECollRAfrRAC OSU Long Ter n Cooling Final Validation Report," WCAP 14776, Revision 2, May,1997. 440.756F 2: A1600 Standard Safety Analysis Report, Revision 13, subsections 15.6.5.4C.3.4 and 15.6.5.4C.3.5, May 30,1997. SSAR Resiston: None 440,756F 2 3 Westingh00S6
i NRC FSER OPEN ITEM
= - _ l ' Ouestion: 720.422F (OITS # 6134) [
WCAP 13914. Revision 2, does not address the need for the COL applicant to deselop guidance and procedures for: l (1) powering the hydrogen ignitiers from batteries,(2) containment venting in the event of core concrete interactions. [ and (3) post <72 hour actions, The report should be revised to include guidance in this regard. This in Open item : 19.2.5 2. Response! WCAP 13914 has been revised to address the issues raised by this open item. Revision 3 of WCAP 13914 is provided as Enclosure 2 of Westinghouse letter DCP/NRCl214. dated January 15, 1998. SSAR Revision: None. PRA Revision: None. , r
+
t t e 720.422F 1 '
=
, w... - - - - , ,
NRC FSER OPEN ITEM y:n A!!n 1 Ouestion: 720.434F (OITS #6163) As documented in a May 13,1997, meeting summary of an April 15,1997, meeting, Westinghouse proposed that the staffs set of insights rewlting from their review of the PRA be shared with Westinghouse. Unless Westinghouse determined that there was technically inccrrect information in the staffs list there would be no new meetings or information transfer, and the staffs insight. would be added to the Westinghouse insights. To that end Enclosure 2 [of NRC letter to Westinghouse dated Nos ember 7,1997] contains the staffs insights as a result of the review of the level i PRA. Enclosure 2 contains add; ional insights from those contained in Chapter 59 of Westinghouse's PRA. Incorporation of the additional insights that esist in Enclosure 2 to the Westinghouse insights is an open item.
Response
De NRC May 13,1997 letter summariting the April 15,1997 PRA insights rnecting states "Unless Westinghouse determined that there was technically incorrect information in the staffs (insights] list, there would be no new meetings or information transfer. De staffs insights would be added to the AP600 Design Control Document." Westinghouse did not agree during the April 15 meeting to include the staffs insights into the Westinghouse AP600 PRA report, nor does the NRC May 13,1997 letter state the commitment. De NRC's PRA insights provided as Enclosure 2 of their November 7,1997 letter, will be includeo in the AP600 Design Control Document per the agreement on April 15, and not duplicated in Chapter 59 of the AP600 PRA. Westinghouse did agree at the Aoril 15, 1997 meeting to review the staffs insights for technical accuracy and appropriateness. In that vem, attached is the Westinghouse feedback on the staff insights. Note that much of what is provided it, the staffs insights is already provided in Table 59 29 of the AP600 PRA. Westinghouse does agree to provide some additional insights into PRA Table 59 29 based on the staff insights. Rese are shown below. The wording of the staffs insights versus what n included within the AP600 Design Control Document will be worked out with the staff prior to issuance of the Design Control Document. PRA Revision: The RAP section in the SSAR has been moved to section 17.4, per request of NRC. As a result, when "SSAR 16.2" is named in the disposition column of PRA Tab!c 59 29, it will be changed to "SSAR 17.4". De following changes will be made to PRA Table 59 29:
- Under item ib (ADS);
Stage 1,2, and 3 valves are stroke. tested escry 6 -S cold shutdown. If RNS is loss during reduced inventory conditions with the reactor coolant system open, a rent path through the ADS 4th stage is required to preclude the occurrence t.f surge lineflooding and thereby not I affect gravity injection. [ disposition = PRA Attachment 54B] l l ADS 4th etage squib valves receive a signalto open during shutdown conditions using PMb tow hot leg i Irrellogic. (disposition = SSAR 6.31 [ W6Stifigh00$8
l NRC FEER OPEN ITEM e Under item le (PRHR): Capabihty exists for the control room operator to identify a leak in the PRilR llX before it can degrade to a tube rupture, dwmg : ab=;=-: &;ip b= =!&,HDBAh
! The PRHR HX, in conjunction with the PCS, can provide core coolingfor an indefinite period of time.
l Aper the IRWST water reaches its saturation temperature, the process of straming to the containment l initiates. Condensation orcars on the containment ressel, and the condensate is collected in a sgfety. l related gutter arrangement which returns the condensate to the IRWST. The gutter normally drains to the containment sump, but when the PRHR HX actuates, sqfety related isolation ralres in the gutter drain line shut and the gutter overpow returns directly to the IRHST. (disposition = SSAR 6.3.2.1.1)
- Under item 6 (RNS):
VIanned maintenance of the RNS and its support systems (CCS and SWS) is perforard e-power in hfodes I,2.J. (disposition = SSAR 16.3)
+ Under item 10:
The operathn of RNS and its support systems (CCS, SWS, main ac power and onsite power)is RTNSS-l important for shutdown decay heat removal during reduced RCS inventory operations. [SSAR 16.3 disposition will be added) Operation of RNS during at power conditions provides margir.for long term cooling T&H uncertainty. Short term availability controls of the RNS are prcrided. (disposition = SSAR 16.3)
+ Under item 13, add:
l Topreventfooding in a radiologically controlled area fRCA)in the a':xiliary buildingfrom propagating i to non radlulogically controlled areas, the non RCAs are separatedfrom the RCAs by 2 and 3. foot walls i and poor slabs. In addition, riectrical penetrations between RCAs and non RCAs in the auxiliary building are located above the maximum pood Irrel. (disposition = SSAR 3.4.1.2.2.2)
- New item (#42):
l No safety related equipment is located outside the Nuclear Island, (disposhion = SSAR 3.4.1) 720.434F 2 g
i NRC FSER OPEN ITEM I t WP"P.t.. 1 NRC Staff Insights of the AP609 Le' eel 1 PRA and Westinghouse Feedback ; i OnmLA. plant wide reouirements
- 1. WEC will maintain a list of risk important systems, structures and components (SSCs) in the D R8 W Response: The risk important SSCs within the scope of D RAP are provided in SSnR Table 17.41. There is no additional action required by Westinghouse to maintain this list pfter Final Design Approval.
Westinghouse does not agree that this item is an insighl of the Isvel 1 PRA. rather the PRA results are usedfor identyying the risk important SSCs in D RAP . 2, De COL Applicant should perform a scismic walkdown to ensure that the as built plant conforms to the assumptions in the AP600 PRA based seismic matgins analysis and to assure that seismic spatial systems interactions do not exist. Details of the seismic walkdown will be developed by the COL epplicant. W Response: As provided in the response to FSER open items 720.43f f through 720.433F, the seismic margin Combined Ucense applicant action item will be changed in AP600 PRA Revision 11, subsection 39.10.6 to read as follows: The Combined Ucense applicant referencing the AP600 cert $ed design should perform a seismic walkdown to cor$rm that the as built plan; corforms to the design used as the basis for the seismic margin evaluation and that seismic spatial systems interactions do not nist. Details of the seismic walLJown will be developed by the Combined Ucense applicant.
- 3. WEC will maintain a list of the SSC HCLPF values used in the AP600 Seismic Margins Assessment in the D.
RAP. De COL Applicant should compare the as built SSC HCLPFs to those assumed in the AP600 seismic margins analysis (SMA). Deviations from the llCLPF values or assumptions in the SMA should be esaluated by the COL Applicant to determine if any vulnerabilities have been introduced. W Response: The llCLPF values usedfor the AP600 seismic margin analysis are provided in AP600 PRA Table
$S.I. The SSCs captured by the D RAP process using the results of the seismic margin analysis as the rationale for inclusion, are provided in SSAR Table 17.41. There is no additional action requi.ed by Westinghouse to maintain this list qfter Final Design Approval. Westinhouse does not agree that "WEC will maintain a list of the SSC HCLPF values"is an insight a the Level i PRA.
As provided in the response to FSER open items 720.451F through 720.433F, the following Combined Ucense applicant action item will be included in AP600 PRA Revision 11, subsection 39.10.6: The Combined Ucense applicant referencing the /P600 :en$ed design should compare the as built SSC HCLPFs to those assumed in the AP600 seismic margin evaluation. Deviationsfrom the HCLPF values or assumptions in the seismic margin evaluation should be evaluated by the Combined Ucense ! applicmt to determine y unacceptable vulnerabilities have been introduced. 720.434F 3
l NRC FSER OPEN ITEM g a!
- 4. 'Ihe COL Apphcant will maintain an operation reliability assurance process based on the system reliabihty infonnation derised from the PRA and other sources The COL Applicant should incorporate the hst of risk.
importar.t SSCs, as presented in the SSAR section on D RAP,in its D RAP and operation rehabihty assurance prncess. W despcmse: nere is a Combined License applicant 0 RAP action within SSAR subsection 17.4.8 that reads the " Combined License applicant is responsiblefor performing the tasks necessary to maintain the reliability of risk signWeant SSCs." in addition SSAR subsection 17.4.7.3 states the " COL appiscant well need to establish PRA importance measures, the espert panel process, and other Jeterministic methods to determine the site specific list of SSCs under the scope of RAP." These two COL action arms address the stafs insight statements. d Fr.e COL qplicant shot.ld consider the information on risk important operator actions from the PRA, as 6 vend .n Chapter 18 of the SSAR on burntn factors engineering,in developing and implementing procedures, traNny mi other hu nan reliability re:ated programs W Response: In the AP600 PRA. credit is taken for various tasks to be performed in the control room by the team of trained operators. These tasLi are rule. based and procedurali rd. ne tasks refer to the completion of a well defined mission by a team of trained operators following procedures. As stated in SSAR section 18.10 operator training is the responsibility of the COL Westinghouse input to the COL is srmided in WCAP 14635. PRA Table 59 29, item it, also reflects what is wntren in SSAR che.t ter 18. Westinghouse believes what is provided in SSAR section 18.10, and how it is captured in PRA Table 59 29, addresses the stafs insight statement.
- 6. During detailed dr i phase, the COL Applicant should update the PRA using the final design information and site. specific infermation. As deemed necessary, the COL Applicant should update the PRA, including the fire and flood analyses for both at power and shutdown operation. Based on site specific information, the COL Applicant shouM also re evaluate the qualitative screening of external events. If any site specific susceptibilities are found, the applicable etternal event should be included in the updated PRA.
E Resp <msr. There is a COL item provided in PRA subsection 39.10.6 that reads the " Combined License applicant referencing the AP600 cert @ed design will vorfy the as built plant is consistent with the design used as the basisfor the base.'ine AP600 PRA." It is the COL's responsibility to describe how this will be done and whether any portions of the baseline PRA need to be updated.
- 7. No safety related equipment is located outside the Nuclear Island.
W Response: his is an accurate statement. T westinghouse
NRC FSEM OPEN ITEM piii amig
- 8. De AlW10 low pressure systems which interface with the RCS are protected against interfacing s) stems LOCA (ISLOCA) by a combination of multiple isolation valves, valve interlocking, increase in the piping pressure limits and pressure relief capability. l r
W Response: This is an accurate statement. PRA Table 59 29, item 6, specyically discusses the elements which prevent interfacing system LOCA between the RNS and the RCS.
- 9. Solid state switching devices and electro mechanical relays resistant to relay chatter will be used in the Ap600 l&C systems. Use of these devices and relays either eliminates or minimites the mechanical discontinuities associated with similar devices at operating teactors.
W Response: It is not understood why the stafs statement is an insightfrom the AP600 PRA. The stq[f would : need to explain why this is an important insight of the PRA tojustsfy its placement in the DCD. ' The stafs statement is accurate, but h not explicitly stated in the SSAR or PRA.
- 10. Dere are no watertight doors used for Good protection in the AP600 design.
W Response: This is an accurate statement per SSAR subsection 3.4.1.1.2.
- 11. De AP600 design minimites potential Gooding sources in safety related equipment areas, to the extent possible.
De design also minimises the number of penetrations through enclosure or barrier walls below the probable maximum Good level. All Good barriers (e g., walls. Doors and penetrations) are designed to withstand the , maximum anticipated hydrodynamic loads as well as water pressures generated by Goods in adjoining areas. W Response: E2cluding the ending phrase "as wellas water pressures generated byfloods in adjoining areas," the stq0 statement is supported by SSAR subsection 3.4.1.1.2. This is essentially item 23 of PRA Table 39 29.
- 12. Drains are capable to remove now from an assumed break in a line up to 4" in diameter and include features, such as check valves and siphon breaks, that prevent backDow.
, W Response: The irformation as worded in the sinfs statement is not supported by text in the AP600 SSAR. SSAR subsection 9.3.5.1.2 does read
- Plugging of the drain headers is minimi:ed by designing them large enough to accommodate more than the design flow and by making the flow path as straight as possible. Drain headers are at least 4 inches in diameter." Regarding the portion of the stqfs statement on backflow prevention, see the last bulletfrom item l$ below.
- 13. Dere is no cable spreading room .in the AP600 design. ,
W Response: This is an accurate statement, e
. , . - . .- . , - _ . _ _ ---,y __, ,_.,______,-mm. ,
NRC FSER OPEN ITEM i i
- 14. %e separation of equipment and cabhng auociated with different daisions of safety related equipment as ucli as the separation of safety related from nonsafety related equipment, minimites the likehhood that a Gre or Good would affect more than one safety related system or train.
W Response: This is an accurate statement. PRA Table 39 29, item 13, provides the same information.
- 15. The following minimise the probabihty for fire or Good propagation from one area to another and helps limit risk from internal Orcs and Goods:
- Fire barriers are scaled and Good barriers are watertight.
W Resporue: This statement isfrom PRA Table 39 29, item 14, but is missing the words "to the entent possible" after the word sealed.
- Each fire door is alarmed in the control room.
W Resp <mse: PRA Table 59 29, item 14 provides the same statement.
- ne COL Applicant will ensure the reliable performance of fire barriers through appropriate inspection and maintenance of doors, dampers, and penetration seals. Also, all water tight penetrations will be maintained with high rehability durmg power operation to prevent the propagation of water from one area to the next.
W Response: The staff's statement appears to be concentrating on a COL item for inspection and maintenance of fire barriers and maintenance of reliable water tight penetrations. Westinghouse is not spectfying the COL items to this level because it is the COL's responsibility to describe how this nill be done. Rather, Westinghouse includes a COL item provided in SSAR subsection 9.5.1.8 that reads the
- Combined License applicant will address qualfication requirements for individuals respcmssble for development of the fire protection program, training offire fightmg personnel, administrative procedures and controls governing the fire protection program during plant operation, andfire protection system maintenance." In addition, as stated in SSAR Table 9.3.1 1, items 29, it is the COL's responsibility for " establishing administrative controls to maintain the performance of thefire protection system and personnel."
- De COL Apphcant will ensure the availability of proper Gre Oghting equipment in all plant areas, and especially in the most risk signincant Ore areas.
E Response: SSAR Table 9.5.1 1, items 4,30, and 32, cover this staf statement. Note that it is not appropriate to add the phrase "and especially in the most risk significantfire areas" because Table 9.511 cosers allfire areas. There is no need to limit this to the most risk sigmficantfire areas. t 720m W westinghouse
l l 1
~ NRC FSER OPEN ITEM
+ The COL Applicant will maintain an adequately staffed, well trairied, and well prepared fire brigade. ,
W Response: SSAR Table 9.$.11, items 4 and 30 through 34, cover this stagstatement, i
. When a fire door, fire barrier penetration, or flood barrier penetration must be open to allow specific i maintenance (e.g., during plant shutdcwn), appropriate compensatory measures will be taken to mini-mise risk, Risk during shutdown is minimized by appropriate outage ruanagement, administrative controls, procedures, and operator knowledge of plant configuration. In particular, this will require configuration control of fire / flood barriers to ensare the integrity of fire and flood barricts between areas containing equipment performing redundant safe shutdown functions.
E Response:. The intentions of what is described in the stafs statement is covered by good plant operating practices. It is covered in a higher level by SSAR Table 9.5.1 1, items 4 and 29.
- Drains include features, such as check valves and siphon breaks, that prevent backflow.
W Response: Assumption m, as written in PRA Chapter $6, reads *forpoor drains, appropriate precautions such as check valves, back pow preventors, and siphon breals are assumed to prevent back pow and any potentialfooding."
- 16. Fire detection and suppression capability as well as flooding control features and sump level indication are provided in the AP600 design. Appropriate compensatory measures will be taken by the COL Applicant to rnaintain adequate detection and suppression capability during maintenance activities. ,
E Response: Per SSAR section 13.$. the Combined License applicant is responsible for developing the plant procedures. The stafs statement is part of good plant practices, and should be addressed by the applicable procedures which the COL will develop.
- 17. In addition to the MCR which has its own dedicated ventilation system, separate ventilation systems are provided for each of the two pairs of safety related equipment divisions supporting redundant functions (i.e.,
divisions AAC and B&D). Furthermo:c,the plant ventilation systems include features to prevent propagation of smoke from a non safety related area to a safety related area or between safety related areas supported by two different divisions. The COL holder tr~t ensure the reliable performance of such smoke propagation prevention features. W Response: Excluding the COL statement, the stafs statement is covered by item 20 of PRA Table 39 29. Regarding the COL statement, this level of detailis not included within Westinghouse COL items of SSAR 9.3.
+
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- 18. He COL applicant should implement the maintenance guidelines as described in the Shutdown Evaluation Report (WCAP 14837).
K Respumse: SSAR section 13.3.1 fas revisedper the respcmse to FSER open item 440.763F, Westinghouse letter , UCP/NRC1198, dated December 22, 19971 includes the following statement' WCAP 14837 provbles input to the Combined Ucense applicantfor ihr development of plant spec @c refueling , plans." This means the maintenance guidelines, as urli as other guidelines spec @ed within the WCAP, should be considered by the Combined Ucense applicant shen they develop the plant procedu r This SSAR COL item covers as a higher level the stpfs statement.
- 19. De COL appna .suld control transient combustibles during shutdown operations.
W Response: The intentions of what is described in the staff's statement is covered in a higher level by SSAR Table 9.3.1 1, item 4d. Main Conttpl Room (MCR) and Remote Shuldgwn Workstation (RSW)
- 1. The automatic function of the AP600 actuation systems (i.e., PMS and CAS)is not affected by a fire in either the MCR or the RSW. His ensures an independent, automatic means, to reach safe shutdown even when a fire occurs in the MCR or the RSW (manual actuation is not needed unless the automatic actuation fails). Also, even though a fire in the MCR may defeat manual actuation of equipment from the MCR, it will not affect the manual operation from the RSW. His is because the I&C cabinets are located in fire areas outside the MCR and the RSW.
W Response: The stafs statement is covered by item 19 of PRA Table 39 29,
- 2. Redundancy in MCR operations,in terms of both monitoring and manual control of safe shutdown equipment, is provided within the MCR itself. His provides an alternative mer -
mitigating certain MCR fires before deciding to evacuate the MCR and use the RSW. E Response: The stafs statement is covered by item 17 of PRA Table $9 29. , s
- 3. If MCR evacuation is necessary, the RSW provides complete redundancy in terms of control for all safe shutdown functions.
W Response: This statement is paraphrasedfrom SSAR section 7.4.3.1.1. The stafs statement is covered by item i2 of PRA Table 59 29.
- 4. The MCR has its own dedicated ventilation system and is pressurized. His climinates the possibility of .
smoke, hot gases, and fire suppressants, originated in areas outside the MCR, to migrate via the ventilation
- 5) stem to the controi room. ,
W Response: The stafs stassment is covered by item 20 of PRA Table 59 29. Note it is recommended that the stpfs wording of " eliminates' be changed to ' prevents'. , 720.434F 8
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i 1 NRC FSER OPEN ITEM
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$. He MCR and the RSW are in separate fire and Good areas, ney have s,;parate and independent ventilation systems. ,
W Response: The stafs statement is covered by items 18 and 20 of PRA Table 59 29,
- 6. Al%00 MCR fire ignition frequency is limited as a result of the use of low voltage, low-current equipment ,
and fiber optic cables. H Response: ne stqff's staremort is covered by item 16 of PRA Table 59 29. Containment / Shield Building
- 1. Containment isolation functions are protected from the impact of internal fires and Goods by redundant containment isolation valves in each line which are located in separate fire and Good areas and, if powered, are served by different power and control divisions. Always, one isolation component in a given line is located inside containment, while the other is located outside containment, and the containment wall is a fire /Dood barrier.
r W Response: ne staff's statement is covered by item 22 of PRA Table 59 29.
- 2. Although the containment is a single fire area, redundant divisions are generally separated by continuous structural or Gre barriers without penetrations and by labyrinth passageways. In a few situations, the divisions ,
are separated by large open spaces without intervening combustibles. , W Response: Westinghouse recommends the stafs wording of this insight repect what is written in SSAR subsection 9A.3.1.1, spec (fically: The containment / shield building comprises onepre area which is separated intopre :ones. "These :ones are based on the establishment of boundaries (structures or distance) that inhibit pre propagation from :one to :one. Complete pre barrier separation cannot be provided inside containment because of the need to maintain thefree e.nchange of gases for purposes such as passive containment cooling."
- 3. Dere are only two compartments inside containment (PXS A and PXS B) containing safe shutdown equipment other than containment isolation valves that are Goodable (i.e., below the maximum Dood height),
Each of these tuo compartments contains redundant and essentially identical equipment (one accumulator with associated isolation valves as well as isolation valves for one CMT, one IRWST injection line and one containment recirculation line). Dese two compartments are physically separated by 2 or 3 foot walls and Door slabs to ensure that a Good in one compartment does not propagate to the other. Drain lines fmm the PXS A and PXS B compartments to the reactor vessei cavity and steam generator compartment are protected from backflow by redundant backDow preventers. 4 no.434F 9 , (
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l i NRC FSER OPEN ITEM l W Response: Westinghouse recommends the staff remose the word "only"in the prst sentence. The canty alw has source range detectors. it is correct that the PXS A and PXS B compartments are physically such that u flood in one compartment does not propagate to the other; honeser, Westinghouse recommends the staf remove the specifics that the ;ompartments are separated "by 2 or 3 foot walls andfloor slabs.* It appears the staginadvertently used the words regarding 2 and J foot walls andfloor slabs that appear in SSAR subsection 3.4.1.2.2.2 which pertains to the asailiary building separation of RCA and nonRCA areas. Once these recommendations are implemented. the staff's statement is fully supported by SSAR subsection 3.4.1.2.2.l.
- 4. Containment isolation valves located below the mar %um Good height inside containment or in the Auxiliary Huilding are normally closed and are designed to fail closed when submerged.
W Response: The staffs statement is not technically accurate. The valves are not designed tofall closed when submerged. Westinghouse recommends the staff change the wording of t'selv statement to read consistently with SSAR subsection 3.4.L2.2.1, Specifically, Ihr SSAR reads "There are four automatically actua:ed containment isolation valves inside containment subject toflooding. These four normally closed containment isolation valves would not fail open as a result of the compartmentflooding. Also, there is a redundant, normally closed containment isolation valve located outside containment in series with each of these valves."
- 5. The fragility of salve rooms, labeled 11206/11207, where the passise core cooling astem valves are concentrated is an important factor in the AP600 capability to withstand earthquakes. P capacit) of the as-built SSCs to meci the HCLPF values assumed in the AP600 PRA will be checked by a seismic walkdown.
W Response: It is not understood what the stag means by thefragility of valve rooms 11206 and 11207 is an import 2ntfactor in the capability of AP600 to withstand earthquakes. The HCLPF valuefor these rooms is 0.96g (per PRA T 'e $$.1) The HCLPF valuefor that valve rooms is not the limtting HCLPF elementfor the nuctnr island. Westinghouse recommenas thefirst sentence of the stafs statement be removed. The stafs statement regarding a seismic walkdown is already addressed under item 2 of " general
& plant wide requirements."
- 6. The passive containment cooling system (PCS) cooling water not evaporated froru the vessel wall Dows Jow n to the bottom of the inner containment annulus into floor drains. De redundant Door drains rcute the excess water to storm drains. De drain lines are alwsys open (withcut isolation valves) and each is sired to accept maximum PCS How. The interface with the storm drain system is an open connection such that any blockage in the storm drains would result in the annulus drains overdowing the connection, draining the annulus independently of the storm drain system.
720mo W westinghouse
NRC FSER OPEN ITEM i i W Response: Westinghouse recommends the staf revise this statement to read "The passis e containr-ent cooling system (PCS) wolin;' water not evaporatedfrom the vessel nallflows dann to the bottom of the inner containment annulas. Two 100 percent drain openirigs, located in the slJe wall of the shield building, are alway s open with screens provided to prevent entry ofsmall animals into the drains." Note that the specifle dram corfguration has changed since what was modeled in PRA Revision
- 8. when the drains n ere located on thefloor of the annulus (see also response to FSER open item 720.440F). Thus the stafs statement should be revised.
- 7. De annulus floor drains, w hich are essentially pipes embedded into the wall of the Shield Building, will have the same (or higher)IICLPF value as the Shield Building. his ensures that the drain system will not fail at lower acceleration levels causing water bkicking of the PCS air bafue.
W Response: Refer to trem 6 above regarding placement of the annulus drains.
- 8. The COL applicant should develop and .mplement policies, procedures, and training to c.ose containment penetrations during Modes 5 and 6 in accordance with TS 3.6.8.
W Response: A COL item in SSAR subsection 13.3.1 states the Combined License applicant nill address plant procedures. A COL item in SS AR subsection 13.2.1 states the applicant will develop and implement training programs for plant personnel. These items inherently include following the Technical Spec $ cations. The COL items in SSAR chapter 13 cover the staffs statement. Anihartilmldins I. Separate ventilation systems are provided for each of the two pairs of safety related equipment divisions supporting redundant functions (i.e., divisions A&C ar.d D&D). His prevents smoke, hot gases, and Gre suppressants originating in divisiotis A or C from propagaung to divisions B and D. h Response: The stafs statement is covered by item 20 of PRA Table 39 29. Note this is essentially a duplicate ofitem 17 of " general & plant. wide requirements."
- 2. De majer rooms housing divisional cabling and equipment (the battery rooms, DC equipment rooms, I&C rooms, and penetration rooms) are separated by 3 hour rated Orc walls without openings. There are no doors, dampers, or seals in these walls, he rooms are served by separate ventilation subsystems. In order for a Gre to propagate from one divisional room to another, it must move past a 3 hour barrier (e.g., a door) into a common corridor and enter the other room through another 3 hour barrier (e g., another door).
W Response: This is an accurate statement. It is essentially what is descrsbed in SSAR subsection 9A.3.1. 720MF 11 W westinghouse
o NRC FSER OPEN ITEM
- 3. A two foot concrete floor (barrier) protects important safety-related !&C equipment as well as the main control room and the ternote shutdown panel. located in N north end of the Auxiliary Building, from potent'al debris produced by a postulated seismically induced structural collapse of the adjacent Turbine Building and propagated through the accen bay separating the two buildings.
K Resp <mse: To be an accurate statement, the stafs wording should be changed asfollows: fI) change "a two. foot conc?rtefloor fbarrier)~ to "An access bay *; and f 2) delete the ending words 'and propagated through the access bay separating the two buildings." By changing these words, the statement is now consistent with PRA subsection 33.3 8.
- 4. There are no connections to sources of "unhmited" quantity of water m the Auxiliary Building.
W Response: It is not understood what is the definition of ~ unlimited quantity of water" or the purpose of this statement. Uponfurther understanding of this statement, it may be accurate to state there are no normally open connections ... .
- 5. To ensure that a flooding ir. a radiologically controlled area (RCA) in the Auxiliary Building does not propagate to non RCAs (where all tafety related equipment except for some containment isolation valves is located). the non RCAs are separated from the RCAs by 2 and 3 foot walls and Door slabs. In addit.or.,
electrical penetrations between RCAs and non RCAs in the Auxiliary Building are located above the maximum flood level. E Response: As it is not appropriate to use the word " ensure" since its interpretation is subjective. Westinghouse recommends the stafs statement be rewosded to read "To preventflooding in a RCA in the auxiliary buildingfrom pro;n. gating to . ." and to remove the statement in parentheses. The statement will then be conhte at uith SSAR subsection 3.4.1.2.2.2.
- 6. 'lle two 72 hour rated Class IE division B and C batteries are located above the maximum flood height in the Auxiliary Building considering all p>assible flooding sources (including propagation from sources located outside the Auxiliary Building).
W Resp <mse: It is not clear why the stqgincludes thss statement as an important insightfrom the PRA. The 24 hour Class iE batteries are usedfor safe shutdon n operation; the 72. hour batteries are usedfor functsons such as post-accident sampling.
- 7. Flood water propagated frorn the Turbine B;.ilding to the Auxiliary Building vahe/ piping penetration room at grade level (the only Auxiliary Building area that interfaces with the Turbine Building) is directed to drains and to outside through access doors. This. combined with the presence of water tight walls and Door of the valve / penetration room. limits the maximum Good height in the valve / piping penetration room (to about 36 inches) and ensures that the flooding does not propaga.e beyond this area.
E Resp <mse: Change the words
- ensures that the flooding does not propagate . . ~ to " prevents ficoding from propagatir.g beyond this area." The statement is then accurate per SSAR subsection 3.4.1.2.2.2.
Antiliary Building Isvel J. wn RCA discussion. 720.434F 12 g8 Westinghouse
O NPC FSER OPEN ITEM
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- 8. The mechanical and electrical equipment in the Auxiliary Building are separated to p? vent propagation of leaks from the p!ning and mechanical areas to the Class IE electrical and riass lE l&C equipment rooms.
W Response: By revising the wording to read " the piping and mecisanical cguipment areas", the staff's statement becomes consistent with SSAR subsection 3.4.1.2.2.2. Turbine Building
- 1. No safety related equipment is beated in the turbine building. There is a 3 hour fire barrier wall between the turbine bcilding and the safety related areas of the Nuclear Island.
E Response: This is an accurate statement, per SSAR subsection 3.4.1.2.2.3. [ Note - there was not item 2 or 3 in Attachment 2 of NRC's November 7.1997 letter.)
- 4. Connections to sources of"large" quantity of water are located in the Turbine Building. They are the service water system (SWS) which interfaces with the corponent cooling water system (CCS) and the circulating water system (CWS) which interfaces with the turbine building closed cooling system (TCS) and the condenser. Features that minimize nood propagation to other buildings are:
- Flow from any postulated ruptures above grade level (elevation 100' 0")in the Turbine Building Dows down to grade level via Coor grating and stairwells. This grating in the Doors also prevents any significant propagation of water to the Auxiliary or Annes Buildings via now under the doors.
. A relief panel in the Turbine Building west wall at grade level directs the water outside the building to the yard and limits the maximum Good level in the Tuibine Building to less than 6 inches. Flooding propagation to areas of the adjacent Auxiliary and Annex Buildings, via now under doors or backDow through the drains,is possible but is bounded by a postulated break in those areas.
E Response: Information in SSAR subsection 3.4.1.2.2.3 supports the staffs statement once the word " Annex" is removedfrom the two sub buttets. Annet Buildine
- 1. There is no safety related equipment located in the Annex Building.
W Response: This is an accurate statement, per SSAR subsection 3.4.1.2.2.3.
- 2. Flood water in the Annen Building grade level is directed by the sloped floor to drains and to the yard area through the front door of the Annex Building.
2 Response: Remove the word " front"from the statement, and then it becomes an accurate statement, per SSAR subsection 3.4.1.2.2.3. W Westinghouse
l I NRC FSER OPEN ITEM siHM=e t a
- 3. Flow from any postulated ruptures above grade levelin the Annex Building is directed by floor drains to the Annex Buildmg sump whict discharges to the Turbine Building drain tank. Alternate paths include Dows to the Tutbine Building via flow under access doors and down to ;;rade lesel via stairwells and elevator shaft.
W Response: Remose the word "any"from tne statement, and then it becomes consistent with SSAR subsection 3.4.1.2.2.3.
- 4. The Doors of the Annex Building are sloped away from the access doors to the Auxiliary Building in the vicinity of the access doors to prevent migration of flood water to the non-radiologically controlled areas of the Nucleat Island where all safety-related equipment, except for some containment isolation valves, is located. [lTAAC].
. E Response: This is an accurate statement per SSAR subsection 3.4.1.2.2.3.
- 5. There are no connections to sources of " unlimited" quantity of water in the Annex Building.
W Response: It is not understood what is the definition of " unlimited quantity of water" or the purpose of this statement. RcasicLcoolant system
- 1. To prevent overdraining, the RCS hot and cold legs are vertically offset which permits draining of the steam generators for nonle dam insertion with a hot leg tevel much higher than traditional designs. This level is nominally 80 percent level in the hot leg.
E Response: This is an accurate statement per SSAR subsection 3.4.6.2. Although the second sentence may be an insight of the Shutdown haluation Report, it is not understood why this is an important insight from the PRA.
- 2. To lower the level in that hot leg the vortexing can occur, a step nozzle connection between the RCS hot leg and the RHR suction line is used. The step nozzle is a 20 inch schedule 140 pipe, approximately 2 feet long.
E Respome: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this detail ofinformation is an important insight of the PRA. For example, the schedule of the piping is not important in calculating the failure probability. However, if the stag explains why this is important as a PRA insight, then please revise the opening of thefirst sentence since it appears to be missing some words. To be consistent with SSAR subsection 3.4.7.2.1, Westinghouse recommends the sentence read "To lower the RCS hot leg level at which a vortex occurs in the
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RNS suction line, a step no.ule . . 72-u W w.anc s.
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NRC FSER OPEN ITEM
- 3. Should vortering occur, the maximum air entrainment into the pump suction was shown experimentally to be no greater than 5 percent.
E Respnse: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this information is an important insight from the PRA. However, {f the stag explains why this is - important as a PRA insight. then please revise the sentence to read " . RNS pump suction . "
- 4. There are two safety-related RCS hot leg level channels, one located in each hot leg. These level instruments are independent and do not share instrument lines, nese level indicators are provided primarily to monitor RCS level during midloop operations. One level tap is at the bottom of the hot leg, and the other tap is on the top of the hot leg as close to the steam generator as possible.
E Response: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this information is an important insightfrom the PRA.
- 5. Wide range pressurizer level indication (cold calibrated)is provided that can measure RCS level to the bottom of the hot legs. The upper level tap is connected to an ADS valve inlet header above the top of the pressurizer. The lower level tap is connected to the bottom of the hot leg. His non safety related pressurizer level indication can be used as an alternative way of monitoring level and can be used to identify inconsistencies in the safety related hot leg level instrumentation.
. E Response: Although this may be stated within the Shutdown Evaluation Report, it is not understood why this information is an important insightfrom the PRA.
- 6. The RNS pump suction line is sloped continuously upward from the pump to the reactor coolant system hot leg with no local high points. This design eliminates potential problems in refilling the pump suction line if a RNS pump is stopped when cavitating due to excessive air entrainment. His self-venting suction line allows the RNS pumps to be immediately restarted once an adequate level in the hot leg is re established.
This is an accurate statement per SSAR subsection 5.4.7.2,1. ' W Response: i-
- 7. De COL applicant should have procedures and policies to maximize the availability of the non-safety related wide range pressurizer level indication (cold calibrated) during RCS draining operations during cold shutdown. De operators shall be trained to use this indication to identify inconsistencies in the safety related hot leg level instrumentation to prevent RCS overdraining.
E Response: . SSAR section 13.5 provides the committment that the Combmed License applicant is responsible for developing procedures. The COL items reponed in secrwn 13.5 provide the committment at a higher level than described in the stafs statement above. I 720.434F-15 L l-
NRC FSER OPEN ITEM idR Passive Cate Cooling Systems (PXS) The passive core cooling system (PXS) is composed of (1) the accumulator subsystem, (2) the core makeup tanks (CMTs) subsystern, (3) the in containment refueling water storage tank (IRWST) subsystem, and (4) the passive residual heat removal (PRHR) subsystem. In addition, the automatic depressurization system (ADS). which is part of the reactor coolant system (RCS), also supports passive core cooling functions. W Response: The stafs statement is covered by item 1 of PRA Table 59 29. A.cnnullnati ne accumulators provide a safety related means of safety injeuion of borated water to th RCS. He following are some important aspects of the accumulator subsystem as represented in the PRA:
- There are two accumulators, each with an injection line to the reactor vessel / direct vessel injection (DVI) nozzle. Each injection line has two check valves in series.
- The reliability of the accumulator subsystem is important. He COL will maintain the reliability of the accumulator subsystem.
- Diversity between the accumulator check valves and the CMT check valves minimizes the potential for common cause failures.
E Response: The stafs statement on accumulators is covered by item la of PRA Table 59 29. Core Makeun Tanks (CMTs) The CMTs provide safety-related means of high-pressure safety injection of borated water to the RCS. He following are some important aspects of CMT subsystem as represented in the PRA:
- There are two CM l's, each with an injection line to the reactor vessel /DVI nozzle. Each CMT has a normally open pressure balance line from an RCS cold leg. Each injection line is isolated with a parallel set of air.
operated valves (AOVs) which open on loss of Class IE de power, loss of air, or loss of the signal from the PMS. De injection line for each CMT also has two normally open -heck valves in series.
- The CMT AOVs are automatically and manually actuated from PMS and DAS and their positions are indicated and alarmed in the control room.
- CMTlevelinstrumentation provides an actuation signal to initiate automatic ADS and provides the actuation signal for the IRWST squib valves to open.
- Be CMTs are risk important for power conditions because the level indicators in the CMTs provide an open signal to ADS and to the IRWST squib valves as the CMTs empty. ne COL will maintain the reliability of the CMT subsystem. Rese AOVs are stroke tested quarterly.
W Westinghouse
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- Ch1T is required by the Technical Speci0 cations to be available from power conditions down through cold shutdown with RCS pressure boundary intact.
E' Response: The staf's statement on CofTs is covered by item Ic of PRA Table 59 29. In Containment Refueline Water Storare Tank (IRWST) The IRWST subsystem provides a safety re! ted means of performing (1) low pressure safety injection following ADS actuation, (2) long4erm core cooling via containment recirculation, and (3) reactor vessel cooling through the flooding of the reactor cavity by draining the IRWST into the containment. He following are some important aspects of the IRWST subsystem as represented in the PRA:
- IRWST subsystem has the following flowpaths:
- Two (redundant) injection lines from IRWST to reactor vessel DVI nozzle. Each line is isolated with a parallel set of valves: each set with a check valve in series with a squib valve.
-- Two (redundant) recirculation lines from the containment to the IRWST injection line. Each recirculation line has two paths: one path contains a squib valve and a N10V, the other path contains a squib valve and a check valve.
- He two htOV/ squib valve lines also provide the capability to Good II.c reactor cavity.
- There are screens for each IRWST injection line and recirculation line which ensure that they are not clogged by debris or other materials generated in the IRWST or containment sump. De COL Applicant will maintain the reliability of such screens.
7
- Explosive (squib) valves provide the pressure boui.dary and protect the check valves from any potential
- i. adverse impact of high differential pressures.
~
- The Squib valves and h10Vs are powered by Class IE de power and their positions are indicated and alarmed in the control room.
i
- The squib valves and htOVs for injection and recirculation are auiomatically and manually actuated via PhtS, l and manually actuated via DAS.
1
- Re squib valves and h10Vs for reactor cavity flooding are manually actuated via PhtS and DAS from the control room.-
* - Diversity of the squib valves in the injection lines and recirculation lines minimizes the potential for common cause faih;re between injection and recirculation / reactor cavity Gooding.
- Automatic IRWST injection at shutdown conditions is provided using PhtS low hot leg level logic.
l 720.434F-17 l-
NRC FSER OPEN ITEM ym 1 t
- IRWST injection and recirculation check valves are exercised at each refueling. IRWST injection and recirculation squib valve actuators are tested every 2 years for 20 percent of the valves. IRWST recirculation MOVs are stroke tested quarterly.
- Re reliability of the IRWST subsystem is important. De COL will maintain the reliability of the IRWST subsystem.
- IRWST injection and recirculation are required by Technical Specifications to be available from power conditions to refueling without the cavity flooded.-
E Response: The stafs statements above on IRWST is covered by item Id of PRA Table 59-29, except Westinghouse wishes to note thefollowing change should be made to what is written above:
. Second bullet . remove the work " ensure"for reasons provided earlier in this document.
An accurate statement would read *, . recirculation line which prevents clogging by debris
.. ." Also now the COL item is covered by a higher level action of the COL will maintain the reliability of the IRWST subsystem (SSAR Section 17.41 The IRWST provides a safety related long term source of water during shutdown conditions. He following are some additional important aspects of the IRWST subsystem as represented in the shutdown PRA.
- The COL applicant should provide administrative controls to control trash generated during shutdown operations from entering the RCS and the IRWST which could possibly plug the screens.
W Response: As stated in SSAR section 13.5, the Combined License applicant is responsible for developing administrative controls. The COL item in SSAR chapter 13 covers the staff's statement at a higher level. 'r
- On low hot leg level, the PMS actuates the squib valves to open allowing gravity injection from the IRWST.
W Response: This statement is a duplicate of the 8th bullet on IRWST(see above). Passive Residual Heat Removal (PRHR) System The PRHR provides a safety-related means of performing the following functions: (1) removes core decay heat during accidents. (2) allows adequate plant performance during transient (non-LOCA and non ATWS) accidents without ADS,(3) allows automatic termination of RCS leak during a SGTR accident without ADS, and (4) provides core cooling and pressure control during the early phase of an ATWS accident. E Response: For item (2), recommend changing the word " allows" to "provides." item (4) is ambiguous by using the words early phase of an ATWS. The phrase should read, " allows plant to ride out an ATWS event without rod insertion." o 720.434F-18 g
NRC FSER OPEN ITEM De following important aspects of the PRHR design and operation features are incorporated in the PRA models:
- PRHR is actuated by opening redundant parallel air operated valves (AOVs). Dese AOVs are de jned to fail open on loss of Class lE power, loss of air, or loss of signal from the protection and safety monitoring system (PMS).
- The PRHR AOVs are automatically actuated by two redundant and diverse I&C systems: (1) the safety-telated protection and safety monitoring system (PMS) and (2) the nonsafety-related diverse actuation system (DAS). De PRHR can also be actuated manually from the control room using eithet PMS or DAS.
- Diversity of the PRHR AOVs from the AOVs in the core makeup tanks (CMTs) minimizes the probability for common cause failure of both PRHR and CMT AOVs.
- The positions of the inlet and outlet PRHR valves are indicated and alarmed in the MCR, W Response: The staffs abose statements on PRilR are covered by item le of PRA Table 59 29.
- Re PRHR AOVs and isolation MOV are tested quarterly. The PRHR HX is How tested at shutdown.
W Response: It is true the PRHR A0Vs are tested quarterly, per IST(SSAR subsection 3.9.6). As stated in the PRA and SSAR, the AIOV is closed to test the AOVs so indirectly, the Af0V is also tested; however, the AIOVis not specofied as such per ISTand the PRA. The words "and isolation A10V" should be removedfrom the stafs statement to be technically accurate. It is accurate to say the PRHR HX ispow tested (as is stated by item le in PRA Table 59-29), but it is misleading to say it is tested at shutdown. The HX is pow tested at shutdown, but not every time the plant is shutdown, Per Technical Specipcation, the PRHR HX isflow tested every JO years. It is not an insightfrom the PRA to include this level ofdetail(thepow testfrequency). The recommendation is the stafs bullet above be changed to what is provided by itern le in PRA Table 59-29.
- Use of the PRHR heat exchanger (HX) for long term cooling causes the IRWST water to heat up, resulting in inventory loss through cvaporation. To ensure successful long-term cooling by the PRHR HX, the evaporated IRWST inventory must return to the IRWST after condensed on the containment liner and collected in the IRWST gutter system. The IRWST gutter system, which directs the water to the containment sump during normal plant operation, is automatically re aligned to direct the water back to the IRWST during an accident. The following design features ensure proper re-alignment of the gutter system valves to direct water to the IRWST dunng accidents:
- the IRWST gutter and its isolation valves are safety grade
- the valves that re-direct the Dow are designed to fail-safe on loss of compressed air, loss of Class IE DC power, or loss of the PMS signal.
- the isolation valves are actuated automatically by PMS and DAS.
T m Sunewuse
. . - -_ . - - - . _ _ - - ~-~ - - _ = -- ~ - - - . . - , .
o-NRC FSER OPEN ITEM
)V Response: The staff's statement should be reworded asfollows, to be technically accurate. Note the statement below is consistent with SSAR subsection 6.3.2.1.1.
"The PRilR HX, in corqunction with the PCS can provide core coolirgfor an indefinite period of time. After the IRWST water reaches _its saturation temperature, the procers of steaming to the containment initiates. Condensation occurs on the steel containment vessel, and the condensate is collected in a safety-related gutter arrangement which returns the condensate to the IRWST. The gutter normally drains to the containment sump, but when the PRHR HX actuates, sqfety related Isolation valves in the gutter drain line shut and the gutter overflow returns directly to the IRWST. The followin; design fra.vtes provide proper re alignmentfu the gutter system valves to direct water to the IR W 9Tt" The staff's three sub bullets above are accurate, except change the word "spfety-grade" to " safety.
related" and " fall safe" to " fall closed."
* - Use of the PRIIR HX for long term cooling will tesult in steaming to the containment. The steam will normally condense on the containment shell and return to the IRWST via the gutter system. If the condensate does not return to the IRWST, the IRWST volume is sufficient for at least 72 hours of PRHR operation.
Connections to the IRWST are provided from the spent fuel system (SFS) and chemical and volume control system (CVS) to extend PRHR operation. A safety telated makeup connection is also provided from outside , the containment through the normal residual heat removal system (RNS) to the IRWST. 4 W Response: This is an accurate statemeat.
- Capability exists in the control room u identify a leak in the PRHR HX which could degrade to a tube
! rupture under the stmss conditions, such as RCS pressure increase and temperature gradients inside the HX tube walls, likely to occur during a postulated accident requiring PRHR operation. W Response: Recommend the stafs statement stop after the words " tube rupture". By continuing with the ? specifics of tying this to a transient, it deminishes the leak tightness capability. Note the statement will be consistent with PRA Table 59-29, item le, by ending the sentence as recommended. Also note the operator guidance is provided via Technical Specylcation 3.4.8.
- sTechnical Specifications requ ire t eh PRHR_to be available, with RCS boundary intact, from power conditions down through cold shutdown. Guidance is provided for operator action when a leak is detected m the PRHR -
HX which could degrade to a tube tupture during normal' power operation conditas or under stress conditions, such as RCS pressure increase and temperature gradients inside the HX tube walls,likely to occur during a postuhted accident requiring PRHR operation. W Response: The first sentence is an accurate statement. The second sentence is essentially a repeat of the
' previous bullet. Recommend the second sentente be deleted.
L L L l
. __. _ _ - ._ _ _ ___ _ _ _ __ ____ __~ - _ _ _ __ _
i NRC FSER OPEN ITEM
. tpaanm(
- The PRHR systems provides a safety related means of removing decay heat following loss of shutdown cooling during safe / cold shutdown with the RCS intact.
}Y Response: Change the words " shutdown cooling" to "RNS cooling" This is an accurate statement and is covered by Technical Specification bases 3.33.
Automatic Deoressurization System (ADS) ADS provides a safety-related means of depressurizing the RCS. The following are some important aspects of ADS as represented in the PRA:
* - ADS has four stages. Each stage is arranged into two separate groups of valves and lines. Stages 1,2, and 3 discharge from the top of the pressurizer to the IRWST. Stage 4 discharges from the hot leg to the RCS loop compartment.
- Each stage 1,2 and 3 line contains two MOVs in series. Each stage 4 line contains an MOV valve and a squib valve in series.
- The valve arrangement and positioning for each stage is designed to reduce spurious actuation of ADS.
a Stage 1,2, and 3 MOVs are normally closed and have separate controls.
- Each stage 4 squib valve has redundant, series controllers.
- Stage 4 is blocked from opening at high RCS pressures.
- De ADS valves are automatically and manually actuated via the protection and safety monitonng system (PMS), and manually actuated via the diverse actuation system (DAS).
- Re ADS valves are powered from Class IE de power and theii positions are indicated and alarmed in the control room.
* ' Stage 1,2 and 3 valves are stroke tested every 6 months. Note: Westinghouse has indicated that this requirement may change as a result of an NRC review. Stage 4 squib valve actuators are tested every 2 years for 20 percent of the valves.
- The reliability of the ADS is important. The COL will maintain the reliability of the ADS.
- ADS is required by the Technical Specifications to be available from power conditions down through refuelmg without the cavity flooded.
- Depre'ssurization of the RCS through ADS minimizes the potential for high pressure melt ejection events.
Procedures will be provided for use of the ADS for depressurization of the RCS during a severe accident.
- 720.434F-21 2
y
d
. 1 1
i NRC FSER OPEN ITEM r ... E Response: The stafs above statements on ADS are covered by item Ib of PRA Table $9 29. Notefor the 6th bullet, as a result of NRC review, the stage 1, 2, and 3 valves are now stroke-tested every cold shutdown. With the number of cold shutdowns and refuelings assumed in the shutdown PRA, the test frequency is equivalent to being tested every 6 months. PRA Table 59 29 will be revised appropriately. Notefor the 9th bullet, the wording "during a severe accident"should be changed to "qfter core uncovery."
- Fire induced hot shorts, especially in I&C copper cables from the protection logic cabinets to the squib valve operators, could cause detonation of a squib valve. His risk important concern should be addressed by appropriate power and control cable separation and routing and by the incorporation of features and requirements in the detailed design of ADS cabling.
E Response: Westingh,use recommends the words of the stafs statement be changed to read as described in SSAR subsection 9A.2.7.1, specsfically, " Spurious actuation of squib valves is prevented by the use of a squib valse controller circuit which requires multiple hot shorts for actuation, physical separation of potential hot short locations, and provisions for operator action to remove power from thefire zone ~ Note as stated in the internalfire PRA analysis, it is conservatively modeled in the PRA analysis that or hot short can cause spurious ADS squib valve actuation, whereas, per design, multiple hot shorts are required.
- The first, second, and third stage valves, connected to the top of the pressurizer, provide a vent path to ptolude pressurization of the RCS during shutdown conditions if decay heat removal is lost. One fourth stage ADS valve is required to open if gravity injection is actuated during cold shutdown and refueling with the RCS is open to preclude surge line flooding. On low low hot leg level (empty hot leg), the PMS signals the ADS 4th stage squibs to open to preclude surge line flooding.
E Response: This is an accurate statement. A statement will be added to PRA Table $9 29. Normal Residual Heat Removal System (RNS) ne normal residual heat removal system (RNS) provides the following nonsafety-related means of core cooling during accidents: (1) RCS recirculation at shutdown conditions,(2) low pressure pumped injection from the IRWST, and (3)long term pumped recirculation from the containment sump Such RNS functions provide defense in depth in mitigating accidents, in addition to that provided by the passive safety related systems.
- W Response: This is an accurate statement. The statement is covered by item 6 of PRA Table 59 29.
De following are some important aspects of RNS as represented in the PRA:
- Re RNS has redundant pumps, powered by separate non Class IE buses with backup connections from the diesel generators, and redundant heat exchangers. .
F Response: This is an accurate statement and is covered by item 6 of PRA Table 59-29.
- 720.434F-22 i
NRC FSER OPEN ITEM i
- De RNS provides safety related means for (1) containment isolation at the penetration of the RNS lines,(2)
RCS isolation at the RNS suction and discharge lines, and (3) 1RWST and containment sump inventory makeup. W Response: Exceptfor point (3), the above is an accurate statement and covered by item 6 of PRA Table 59-
- 29. Item (3) is incorrect. RNS does u.ot provide a safety related means, but rather a defense-in.
depth function ofIRWST and containment sump inventory makeup.
- The RNS is manually aligned from the control room to perform its core cooling functions [SSARJ.
Emergency Response Guidelines (ERGS) are provided for aligning the RNS from the control room for RCS injection and recirculation. W Response: This is an accurate statement.
- Recirculation from the containment sump is actuated automatically by a low IRWST level signal or manually from the control room, if automatic actuation fails.
W Response: This statement is misleading as worded. It should read "PXS recirculation valves are automatically actuated. ~ lt is believed the stag was intending to mean the IRWST recirculation valves rather than an RNS recirculation (i.e., pumps stop, start) as could be interpreted by the statement. Note that if RNS is operating, the RNS pumps will continue to operate and provide containment recirculation.
- For long-term recirculation operation, the RNS pumps take suction from only one of the two sump recirculation lines. Unrestricted flow through both parallel paths (one containing an MOV and a squib valve in series, the other containing a check valve and a squib valve in series) is required for success of the sump recirculation function when both RNS pumps are running. If one of the two parallel paths fails to open, operator action (in the control room through PMS) is required to manually throttle the RNS discharge MOV (VO!!) to prevent pump cavitation. [ ERGS).
W Response: This is an accurate statement per the PRA.
- With the NRHR pumps aligned either to the IRWST or the containment sump, the pumps' net positive suction head (NPSH) is adequate to prevent pump cavitation and failure even when the IRWST or sump insentory is saturated.
W Response: Change NRHR to RNS. This above is an accurate statement.
- De RNS containment isolation and RCS pressure boundary val es are safety related. The MOVs are powered by Class IE de power.
W Response: This is consistent with item 6 of PRA Table 59 29. 72o w 23 T westinghouse
l .a I t NRC FSER OPEN ITEM mmann
- The containment isolation valves in the RNS piping close automatically via PMS with a high radiation signal.
Westinghouse analyses indicate that under all accident conditions but large LOCAs, the containment radiation level is well below the point that would cause the RNS MOVs to automatically close. E' Response: Thefirst sentence is consistent with item 6 of PRA Table 39 29. The second sentence tenis to lead beyond an insightfrom the PRA. However, if the staf explains why it considers this an insight. then Westinghouse recommends the second sentence be reworded to read: The actuation serpoint was established consistent with a DBA non mechanistic source term associated with a large LOCA. "
+ ne following AP600 design features contribute to the low likelihood of interfacing system LOCAs through the NRHR system:
- The portion of the RNS outside containment is capable of withstanding the operating pressure of the RCS.
- A relief valve located in the common RNS discharge line outside containment provides protection against excess pressure.
- Each RNS line is isolated by at least three valves.
- De pressure in the RN5 pump suction line is continuously indicated and alarmed in the main control room.
- The pump suction isolation valves connecting the RNS pumps to the RCS hot leg are interlocked with RCS pressure so that they cannot be opened until the RCS pressure is less than 450 psig. His prevents overpressurization of the RCS when the RNS is aligned for shutdown cooling.
- The two remotely operated MOVs connecting the suction and discharge headers, respectively, to the IRWST are interlocked with the isolation valves connecting the RNS pumps to the hot leg. This prevents inadvertent opening of any of these two MOVs when the RNS is aligned for shutdown cooling and potential diversion and draining of reactor coolant system.
- De power to the four isolation MOVs connecting the RNS pumps to the RCS hot leg is administratively blocked at their motor control centers during normal power operation. [ COL).
. He operability of the RNS is tested, via connections to the IRWST,immediately before its alignment to the RCS hot leg. for shutdown cooling, to ensure that there are no any open manual valves in the drain lines. [SSAR, COL, Procedures].
720.434F-24 g
c. NRC FSER OPEN ITEM P. 2 4 E Response: ' Westinghouse has the following commentsfor the sta.[f's above statement:
- Change "NRilR system" to "RNS".
- Second sub-bullet is a true statement, but notfactored into the PRA and is not a key to providing a lo s likelihood ofinterfacing systems LOCA. Thus, Westinghouse does not see this as an important statement to include as an insight.
- Last sub-bullet: It is true that the system is tested; howes er, it is done to test operability of the system, not solely to minimize potentialfor interfacing syetems LOCA or to detect an open valve in the drain lines. lionever, the testing does have this end result efect.
The words should be revised appropriately.
- The IRWST suction isolation valve (V023) and the RCS pressure boundary isolation valves (V001 A, V001 B, V002A and V002B) are qualified for DBA conditions, W Response: It is not understood why the stqf]'s statement is an insightfrom the PRA.
- The reliability of the IRWST suction isolation valve (V023) to open on demand (for RNS injection during power operation and for IRWST gravity injection via the RNS hot leg connection during shutdown operation) is important, De COL will ensure high reliability. [ COL, D RAP].
W Response: This item is acceptable and is covered by SSAR section 17.4 (RAP).
- An alternative gravity injection path is provided through RNS V-023 during cold shutdown and refueling cot ditions with the RCS open. De COL applicant should have policies that maximize the availability of this valve and procedures to open this valve during cold shutdown and refueling operations when the RCS is open.
W Response: - The ERGS cover the operation of the valve. In addition, as stated in SSAR section 13.5, it is the responsibility of the Combined License applicant to develop procedures.
- Re COL applicant will maintain RNS and its support systems (CCS and SWS) during power operation, E Response: To be accurate and consistent with SSAR section 16.3 (Table 16.3 2, item 2.2), change the statement to read: " Planned maintenance afecting the RNS cooling function and its support systems should be performed in Modes I,2, 3 when the RNS is ect normally operating."
- The COL applicant will have administrative controls to maximize the likelihood that RNS valve V-023 will be able to op:n if needed during Mode 5 when the RCS is open, and PRHR cannot be used far core cooling.
W Response: As stated in SSAR section 13.5, it is the responsibility of the Combined License applicant to develop administrative procedures. 720.434F 25
9 b NRC FSER OPEN UEM
. ..= .
t 3++* ,
- _ Since inadvertent opening of RNS valve V024 results in a draindown of RCS inventory to the IRWST and requires gravity injection from the IRWST, the COL applicant will have administrative controls to ensure that -
inadvertent opening of this valve is unlikely, in addition, the COL applicant should evaluate this errot in the human reliability analysis / human factors engineering integration implementation plan. E Response: As stated in SSAR section 13.5. It is the responsibility cr] the Combined License applicant to develop administrative procedures.
- He RNS is an important " defense in depth" system for accidents initiated while the plant is at power or at mid loop during shutdown. The availability control of the RNS and its support systems (CCW, SWS and diesel generators) is covered in SSAR Section 16.3. (RTNSS).
E Response: The reason RNS is important while the plant is at power is not because it is imponant per the PRA results or importance listings, but rather because it provides margin for long term cooling T&H uncertainty. Otherwise, the stafs statement is accurate. Startuo Feedwater System (SFW) He SFW system provides a nonsafety-related means of delivering feedwater to the steam generators (SGs) when the main feedwater pumps are unavailable during an transient. His capability provides an alternate core cooling mechanism to the PRHR heat exchanger for non LOCA and SGTR accidents which minimizes the PRHR challenge rate. De reliability of the SFW system will t.e maintained by the COL Applicant [D-RAP). E Response: The stafs statement is essent: ally taken directlyfrom the SSAR Table 17.4 fRAP). To be accurate, note the words should read startupfeedwater system gumes. The rationale provided in this table for why the startup feedwater pumps are included is based on the Expert Panel, not PRA. Therefore, it is not clear why the stafs statement is considered an insightfrom the PRA. Instrumentation and Control (l&C) ne following three I&C systems are credited in the PRA for providing monitoring and control functions during accidents: (1) the safety related Protection and Safety Monitoring System (PMS),(2) the nonsafety related Diverse Actuation System (DAS), and (3) the nonsafety related Plant Control System (PLS). De PMS provides a safety related means of performing the following functions: i
- Automatic end manual reactor trip.
*~ Automatic and manual actuation of engineered safety features (ESF).
- Monitor the safety related functions during and following an accident as required by Regulatory Guide 1.97.
E Response: The stafs statements on PMS are covered by item 2 of PRA Table 59 29, l 720.43M 26 o
i . NRC FSER OPEN ITEM The DAS provides a nonsafety related means of performing the following functions:
- Automatic and manual reactor trip.
- Automatic and manual actuation of selected engineered safety features.
- Provides control room indication for monitoring of selected safety related functions.
1Y Response: The stafs statements on DAS are covered by item 3 of PRA Table $9 29. The PLS provides a nonsafety related means of performing the following functions: Automatic and manual control of nonsafety-related systems, including " defense-in-depth' systems (e.g., RNS).
- Provides control room indication for monitoring overall plant and nonsafety-related system performance.
}Y Response: SSAR subsection 7.1.1 support the stafs statements on PLS: however, on thefirst bullet. the word
" systems" should be changed to
- functions."
The following are some important aspects of PMS as represented in the PRA:
- The PMS has four (redundant) divisions of reactor trip and ESF actuation and automatically produces a reactor trip or ESF initiation upon an attempt to bypass more than two channels of a function that ut.es 2-out of 4 logic.
The PMS has redundant divisions of safety related post accident paremeter display.
- Each PMS division is powered from its respective Class IE de division.
- De PMS provides fixed position controls in the control room.
- The rehability of the PMS is ensured by redundancy and functional diversity within each division:
The reactor trip functions are divided into two functionally diverse subsystems. De ESF fun,;tions are processed by two microprocessor based subsystems that are functionally identical in both hardware and software. Separate input channels are provided for the reactor trip and the ESF actuation functions, with the exceptian of sensors which may be shared.
- Sensor redundancy and diversity contribute to the reliability of PMS. Four sensors normally monitor variables used for an ESF actuation. Different type sensors, or same typ: sensors in different environment, minimize common cause failures.
72 - 27 T w.sunes.
NRC FSER OPEN ITEM RE a Contmuous automatic PMS system monitoring and failure detection / alarm is provided.
- PMS equipment is designed to accommodate a loss of the normal heating, sentilation, and air conditioning (HVAC). PMS equipment is protected by the passive heat sinks upon failure or degradation of the active IIVAC,
- ne reliability of the PMS is important. De COL will maintain the reliability of the PMS.
- The PMS software is designed, tested, at ; maintained to be reliable under a controlled verificatmn and validation program written in accordance with IEEE 7-4.3.2 (1993) that has been endorsed by Regulatory Guide 1.152. Elements that contribute to a reliable software design include:
- A formalized development, modification, and acceptance nrocess in accordance with an approved software QA plan (paraphrased from IEEE standard, Sec .i.3, " Quality")
- A verification and validation program prepared to confirm the design implemented will function as required (IEEE standard, Section 5.3.4, " Veri 6 cation and Validation")
- Equipment qualification testing performed to demonstrate that the system will function as required in the environment it is intended to be installed in (IEEE standard. Section 5.4, " Equipment Qualification")
- Design for system integrity (performing its intended safety function) when subjected to all conditions, external or internal, that have significant potential for defeating the safety function (abnormal conditions and events) (IEEE standard. Section 5.5, " System Integrity")
- Software configuration management process (IEEE standaro, Section 5.3.5, "Sof tware Configuration Management").
W Response: The stafs above statements on PMS are covered by item 2 of PRA Table 59-29, exceptfor the 7th bullet. Westinghouse does not claim specifically what is written as the third sentence of the stafs 7th bullet. Rather, functional diversity minimi:es the common cause failure among sensors. The following are some important aspects of DAS as represented in the PRA:
- Diversity is assumed in the PRA that eliminates the potential for common cause failures between PMS and DAS. Le DAS automatic actuation signals are generated in a functionally diverse manner from the PMS signals. Diversity between the DAS and PMS is achieved by the use of different architecture, different hardware implementations, and different software.
- DAS provides control room displays and fixed position controls to allow the operators to take manual actions.
- DAS actuates using 2 out of 2 logic. Actuation signals are output to the loads in the form of normally de-energized, energire to-actuate signals. The normally de energized output state, along with the dual 2-out-of 2 redundancy, reduces the probability of inadvertent actuation.
T Westinghouse
1 l NRC FSER OPEN ITEM
- De actuation devices of DAS and PMS are capable of independent operation that is not affected by the operation of the other. The DAS is designed to actuate components only in a manner that initiates the safety function.
- Capabilit, is provided for on-line testing and calibration of the DAS channels, including sensors.
- The DAS manual initiation functions are implemented in a manner that bypasses the signal processing equipment of the DAS automatic logic. This eliminates the potential for common cause failures between automatic and manual DAS functions.
e ne DAS reactor trip function is implemented through a trip of the control rods via the motor-generator (M-G) set which is separate and diverse from the reactor trip breakers. De COL will maintain the reliability of the M G set breakers [D RAP).
- DAS is an important " defense in depth" system. De availability o DAS, with respect to both its reactor trip and ESF actuation functions, will be controlled. [RTNSS). He COL will maintain its reliability [D-RAP).
W Response: The stafs above statements on DAS are covered by item 3 of PRA Table 59 29, exceptfor the 5th bullet, which is supponed by SSAR subsection 7.7.1.11. The following are some important aspects of PLS as represented in the PRA:
- PLS has redundancy to minimize plant transients.
- PLS provides capability for both automatic control and manual control.
- Redundant signal selectors provide PLS with the ability to obtain inputs from the integrated protect on cabinets in the PMS. De signal selector function maintains the independence of the PLS and PMS. The signal selectors select those protection system signals that represent the actual status of the plant and reject erroneous signals.
- PLS control functions are distributed across multiple distributed controllers so that single failures within a controller do not degrade the performance of control functions performed by other controllers.
E Response: The stafs statements on PLS are covered by item 4 of PRA Table 59 29. Onnte Power The onsite power system consists of the main ac power system and the de power system. The main ac power system is a non Class lE system. He de power system consists of two independent systems: the Class IE de system and the non Civ. IE de system.
? Response: The stafs statement is covered by item Sa of PRA Table 59 29.
T westinghouse
NRC FSER OPEN ITEM w MY 1 , e .. The main ac power system is a non Class IE system comprised of a normal, preferred, and standby power system. It distnbutes power to the reactor, turbine, and balance of plant auxiliary electricalloads for startup, normal operation, and normal / emergency shutdown. 2 Response: The stafs statement is covered by item Sa of PRA Table 59 29. The Class IE de and uninterrupdble power supply (UPS) system (IDS) provides reliable power for the safety-related equipment required for the plant instrumentation, control, monitoring, and other vital functions needed for shutdown of the plant. E Response: The stafs statement is covered by item $b of PRA Table 59 29. The non Class IE de and UPS system (EDS) consists of the electric power supply and distribution equipment that provide de and uninterruptible ac power to nonsafety-related loads. W Response: The stafs statement is covered by item Se of PRA Table 59 29. ne following are some important aspects of the main AC power system as represented in the PRA:
- The arrangement of the buses permits feeding functionally redundant pumps or groups of loads from separate buses and enhances the plant operational reliability.
- During power generation mode, the turbine generator normally supplies electric power to the plant auxiliary loads through the unit auxiliary Isansformers. During plant startup, shatdown, and maintenance, the main ac power is provided by the preferred power supply from the high voltage switchyard. The onsite standby power system powered by the two onsite standby diesel generators supplies power to selected loads in the event of loss of normal and preferred ac power supplies.
- Two onsite standby diesel generator units, each furnished with its own support subsystems, provide power to the selected plant nonsafety-related ac loads.
- On loss of power to a 4160 V diesel backed bus, the associated diesel generator automatically starts and produces ac power. De normal source circuit breaker and bus load circuit breakers are opened, and the generator is connected to the bus. Each generator has an automatic load sequencer to enable controlled loading on the associated buses.
W Response: The stafs statements on main ac power are covered by item 5a of PRA Table 59-29. 720.4MF-30 W Westinghouse
t NMC FSER OPEN ITEM The following are some important aspects of the Class IE de and UPS system (IDS) as represented in the PRA:
- There are four indepmdent, Class IE 125 V de divisions. Divisions A and D cach consists of one battery ,
bank, one switchboard, and one battery charger. Divisions B and C are each composed of two battery banks. two switchboards, and two battery chargers. The first battery bank in the four divisions is designated as the 24-hour battery banh. The second battery bank in Divisions B and C is designated as the 72 hour battery bank.
- He 24 hour baticry banks provide power to the loads required for the first 24 hours following an event of loss of all ac power sources concurrent with a design basis accident. The 72 hour battery banks provide power to those loads requiring power for 72 hours following the same event.
...' Battery chargers are connected to de switchboard buses. He input ac power for the Class IE de battery chargers is cupplied from non Class IE 480 V ac diesel generator backed motor control centers.
- The 24 hout and 72 hour battery banks are housed in ventilated rooms apre from chargers and distribution .
equipment.
- Each of the four divisions of de systems are electrically isolated and physically separated to prevent an event from causing the loss of mon than one division.
- Reliability of the Class IE botteries is important. The COL will maintain the reliability of the equipment.
W Resp <mse: The stafs statements on Class IE de power are covered by item $b of PRA Table 59 29. The following are some important aspects of the non-Class IE de and UPS system as represented in the PRA:
- The non Class IE de and UPS system consists of two subsystems representing two separate power supply trains.
- EDS load groups I,2. and 3 provide 125 V de power to the associated inverter units that supply the ac power to the non Class IE uninterruptible power supply ac system.
t
- The onsite standby diesel generator backed 480 V ac distribution system provides the normal ac power to the battery chargers.
- The batteries are sized to supply the system loads fer a period of at least two hours after loss of all ac power i sources.
W Response: The stafs statements on non Class IE de power are covered by item $c of PRA Table 59 29. em
- M0.43#-31 I
L
l NRC FSER Of>EN ITEM I Comnonent Cooline Water System !CCS) Ae component cooling water system (CCS)is a nonsafety-relatW system that removes heat froni various compenents rd transfers the heat to the service water system he fei!c Airg are some important aspects of the CCS as repre-sented in the PRA:
- De CCS is arrangW into two trains. Each train includes one pump and one heat eu hanger.
- During normal operation. (me CCS pump is operating. De standby pump is aliped to automatically start in case of a failure of the operating CCS pump.
- He CCS pumps are automatiet.!!y loaded on the standby diesel generator in the event of a loss of normal ne power. The CCS, therefore, <.ontinues to provide cooling of required components if normal ac power is lost.
W Response: The stafs statements on CCS are covered by item 7 of PRA Table 59 29. Service Water System (SWS) De service water system (SWS)is a nonsafety-related system that transfers heat from the component cooling water heat exchangers to the atmosphere, ne following are some important aspects of the SWS as represented in the PRA:
- The SWS is arranged into two trains. Each train includes one pump, one strainer, and ore cooling tower cell.
- During normal operation, ue SWS train of equipment is operating. The standby train is aligned to automatically start in case of a failure of the operating SWS pump.
- The SWS pumps and cooling tower fans are automatically loaded onto their associated div.cl bus in the event of a loss of normal ac power. Both pumps and cooling tower fans automatically start after power from the diesel generator is available.
}&' Response: The stafs statements on SWS are covered by item E of PRA Table 59-29.
720.434F-32 1
i NRC FSFR OPEN ITEM IP 9!! k Chemical and Volume Cnntro' System (CVS) 1 The chemical and volume control system (CVS) provides a safety-related means to terminate inadvertent RCS b'ron dilution, in addition, the CVS provides a nonsafety related means to (1) provide makeup water to the RCS during normal plant operation,(2) provide boration following a failure of reactor trip,(3) provide coolant to the pressurizer ausiliary spray line, (4) safety related partions of the CVS provide inadvenent boron dilution protection, and (5) safety related portions of the CVS provide isolation of normal CVS letdown during shutdown operation on low hot leg level. E Response: The stafs above statement on CVS is covered by item 9 of PRA Table 59 29 with support from SSAR subsection 9.3.6. Note the second sentence begins by discussing nonsafety-related means, but items (4) and (3) state safety related portinns. It could be a confusing sentence. Also note, item (4) is a repeat of the first sentence. The following are some iraportant aspects of CVS as represented in the PRA:
. De CVS has two makeup pumps and each pump is capable of providing normal makeup.
E Response: This statement is covered by item 9 of PRA Table 59 29
+ One CVS pump is configured to operate on demand while the other CVS pump is in standby. The operation of these pumps will alternate periodically (monthly).
E Response: The stafs statement is accurate per PRA assumptions. The first sentence is true. The second sentence's monthly statement is an assumption of the PRA: however, good operating practices would callfor the COL to periodically alternate the pumps.
- On low hot leg level, the safety related PMS signals three safety related CVS AOVs to close automatically to isolate letdown during Mode 4 (when RNS is in operation), Mode 5, and Mode 6 (with the upper internals in place and the refueling cavity less than half full) as requited by AP600 TS-E Pesponse: Only two of the AOVs are safety-related, the third is nonsafety-related. Exceptfor this error, the above statement is true per the ESF Technical Specification.
- The safety related PMS boron dilution signal automatically re aligns CVS pump suction to the boric acid tank.
His same signal also closes the two safety-related CVS demineralized water supply valves. This signal actuates on any reactor trip signal, source range flux multiplication signal, low input voltage to the Class IE DC power system battery chargers, or a safety injection signal. E Response: This is an accurate statement.
. He COL applicant will maintain procedures to respond to low hot leg level alarms.
E Response: The shutdown ERGS cover the procedure to respond to low hot leg level alarms. 720.434F 33 W - Westinghouse _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.._m
o NRC FSER OPEN ITEM g... l
. Question: - 720.439F (OITS #6177)
Enclosure 2 [of NRC letter dated November 19,1997J contains the staffs insights as a result of the review of the Level 2 PRA. Enclosure 2 contains additional insights from those contained in Chapter 59 of Westinghouse's PRA. Incorporation of the additional insights that exist in Enclosure 2 to the Westinghouse insights is an open item, in addition, the staff believes that certain insights are so important tha they need to be incorporated into the Technical Specifications or into the inspections, tests, analyses, and acceptance criteria (ITAAC). In the cases where the staff believes the disposition of the insight should be Technical Specifications or ITAAC a separate FSER open item number has been assigned.
Response
Westinghouse has reviewed the staffs insights presented in Enclosure 2 of NRC's November 19,1997 letter for technica~ accuracy and appropriateness. Attached is the Westinghouse feedback on the staff insights. As noted in
- the sta. 's question above, several items were provided with a separate FSER open item number. Those items are not being addressed by this FSER open item response, because they are addressed by their assigned FSER open item number.
Changes to the PRA report insights table provided in Chapter 59 are described below. PRA Revision: The R.AP section in the SSAR has been moved to section 17.4, per request of NRC. As a result. when "SSAR 16.2" is named in the disposition column of PRA Table 59 29, it will be changed to "SSAR 17.4". The following changes will be made to PRA Tabte 59 29:
'
- Under item Id (IRWST):
l The operator action toflood the reactor cavity is determined in Emergency Response Guideline FR.C.I, l which instructs the operator to flood the reactor cavity (finjection to the RCS cannot be recovered or l containment radiation reaches a level that indicates fission product releases as determined by a core
! damage assessment guideline. [ disposition = Emergency Response Guidelines]
- Under item 26:
! The reflective insulation panels and support members can withstand pressure differential loading due to the IVR boiling phenomena.
- RPi? insulation panel clearances, water entrance and steam exit flow areas, and loss coefficients are
' based on scale test datafrom the ULPUfacility.-
i L. Water inlets and sseam vents are provided at the optrar. e and exit of the insulation boundary. t Reactor vesselinsulation is an important SSC. . The COL will maintain the reliability of the insulation.
-[ disposition = 5SAR 17.4]
720.439F4 J
NRC FSER OPEN ITEM srmmu n=
- Under item 30:
Operabuity of the hydrogen igniters is addressed by shers term availability controls during modes 1,2, 5 (with RCS pressure boundary open), and 6 (with upper internals in place and cavity levelless than l fu#). - [ disposition = SSAR 16.3) a Under item 36: The Col wul develop and implement severe accident management guidance for operation of the nonsafety related containment spray system using the suggestedframework provided in WCAP 13914. (disposition = PRA Chapter 59 (subsection 59.10.6)]
- Under item 39:
Operabuity of DASfor selected containment isolation actuations is addressed by short term availability controls. [ disposition = SSAR 16.3)
- Under item 41:
The reactor cavity design incorporates features that extend the time to basemat melt-throusk in the event of RPYfailure. The cavity design includes:
. a minimum floor sea of 48 m' availablefor spreading of the molten core debris,
. a minimum thickness of concrete above the embedded containmer.1 liner of 0.85 m, l . there are no interconnecting pipelines embedded in the concrete floor andler sump curb, thereby I preventing debris from passing into the sump l - a sump curb of ssWicient height and width to prevent molten core debris from overflowing or l ablating through the curb. [ disposition = PRA Appendix B)
- New item (#43):
l Capability exists to vent the containment via the RdS suction lines to the spentfuelpool, with the RCS i depressurized and open to the containment atmtsphere via either the ADS or the vessel failure.
-(disposition = PRA Appendix D]
I The COL wi# develop and implement severe accident management guidancefor venting containment using the suggestedframework provided in WCAP 13914. [ disposition = PRA Chapter 59 (subsection 59.10.6)] I 720.439F-2 l.
4 NRC FSER OPEN ITEM i I NRC Staff Insights of the AP600 Level 2 & 3 PRA and Westinghouse Feedback Passive Containment Cooline Systern , The first item of NRC's November 19,1997 Enclosure 2 was FSER open item 720.440F, which pertains to the passive containment cooling system. The response to 720,440F was provided in Westinghouse letter DCP/NRCl l79, dated December 11,1997. There is no change to the PRA insights table presented in PRA Chapter 59 as a result of this PSER open item. The staffs wording is covered by item 37 of PRA Table 59 29, Reactor Cavity Floodine System A safety-related reactor cavity flooding system is included in the AP600 design to prevent reactor vessel breach and ex vessel phenomena in the event of a severe accident. The system is comprised of the following design features:
. two 6-inch diameter recirculation lines that provide a path for gravity draining the IRWST to the reactor cavity. - a squib valve and a motor operated valve in each recirculation line, each powered from the Class IE de power supply, and actuated from the control room, and - a reactor vessel thermal insulation system designed specifically to enhance RPV cooling, as described in FSER Section 19.2.3.3.1.
E Response: The information provided in the stafs two bullets (above) are covered by item Id of PRA Table
$ 9-29.
The reactor vessel insulation is described by item 26 of PRA Table $9 29. The IRWST injection squib valves are diverse from the containment recirculation squib valves. [ Diversity between these vahes is specified in SSAR Section 6.3.2.2.8.9, but the criteria for confirming that diversity has been achieved is not provided. This needs to be addressed by ITAAC. This is Open item 720.44tF.]
- K Response: The stafsJIrst sentence is covered by item Id of PRA Table 59-29. The stafs statement which has been placed in brackets is addressed :n the response to FSER open item 720.441F, The response was provided by Westinghouse lensr DCP/NRCI168. dated December 4.1997.
The containment recirculation squib valves and isolation MO'/s, and containment recirculation screens are included as risk significant SSCs within D-RAP. W Response: The information provided in the stafs statement is covered by item Id of PRA Table 59 29. Note the PRA insights table states the reliability of the IRWST subsystem is important, and then directs the reader to SSAR 16.2 (now 17.4) which is the section on RAP. 720.439F-3
-- - - ~ - .
k NRC FSER OPEN ITEM - , E L Surveillance and maintenance requirements on the related piping and valves are provided in the 'n Service Inspection and Testing Programs. E Response: The information on valves is specyled in item Id of PRA Table 39 29, and directs the reader via the disposition column to SSAR subsection 3.9.6, where ISTprogram is discussed. The operator action to Good the reactor cavity is provided in Emergency Response Guideline FR.C l which instructs - the operator to flood the reactor cavity if injection to the RCS cannot be recovered or containment radiation reaches ) levels that indicate fission product releases as determined by a core damage assessment guideline. E Response: This is a correct statement. An insight will be added to item Id of PRA Table 59 29.
-[ Key aspects of the reactor cavity Ocoding system and the containment layout need to be confirmed by ITAAC to assure that the reactor cavity will Dood snd the RPV will renood as modelled in the PRA (by gravity draining and by manual actuation of the cavity flooaing system). The ITAAC should include confirmation of internal volumes, elevations, and inter compartment vent and drain paths of the subcompartments containing RCS piping components and impacting reactor cavity Gooding and RPV reflooding. WEC needs to provide this ITAAC. 'lhis is open item 720.442F. The response to FSER open item 7?O.442F was provided by Westinghouse letter DCP/NRCl180, dated December 12, 1997.)
RPV Thermal Insulation System The AP600 oesign includes a reacctive reactor vessel insulation system that provides an engineered now path to allow the ingression of water and venting of steam for externally cooling the vesselin the event of a seve:e accident involving core relocation to the lower plenum. H Response: Tiis is a correct statement and is consirtent with item 26 of PRA Table 59 29. Key attributes of the insulation system are:
, RPVlinsulation panel clearances, water entrance and steam exit now areas, and loss coefficients based on scale tests in the ULPU facility.
- ball and cage check valves and steam vent dampers at the entrance and exit of the insulation boundary that open due to buoyant forces during cavity flood.up, and -
- insulation panels and support members designed to withstand the pressure differential loading due to the IVR
' boiling phenomenal E Response: Thefirst and therJ bullets above are accurate statements. The second bullet should be reworded asfollows to be consistent with design information provided in the SSAR: " water inlets and steam .
vents are provided at the entrance and exit of the insulation boundary that open due to buoyant. forces during cavity flood-up ~ The information provided in the stafs statements will be added to item 26 of PRA Table 59 29 with the changes noted. g 1
']
NRC FSER OPEN ITEM EF =is No coatings are apphed to the outside surface of the reactor vessel which willinhibit the wettability of the surface. W kesponse: This is a correct statement and is consistent with item 26 of PRA Table 59 29. [%c reactor vessel insulation system should be included as a risk-significant SSC in the reliability assurance program, and teliability/ availability controls and goals should be provided, consistent with maintenance rule Fuidelines, to assure that operat'ility of the system and moving parts is maintained. ITAAC and availability controls are also needed to assuse that the RPV insulation system will perform as designed. WEC ceeds to provide these commitsnents and ITAAC. his is Open item 720.443F. The response to FSER open item 720.443F was provided by Westinghouse letter DCP/NRCll80, dated December 12, 1997. Note per the response to FSER open item 720.443F, the reactor vessel insulation is included as a risk significant SSC in the RAP. The tie to RAP will be provided as an insight with item 26 of PRA Table $9-29.] hetection of Containment From Diffusion Flames The containment layout prevents the formation of diffusion flames that can challenge the integrity of the containment shell Specifically:
. the openings from the accumulator rooms and CVS compartments that can vent hydrogen to the CMT room are either located away from the containment wall and electrical penetration junction boxes, or are covered by a sccure hatch, and . IRWST vents near the containment wall are oriented to direct releases away from the containment shell.
W Response: The above information is covered by items 31 and 38 of PRA Table 39 29. (Rese provisions need to be confirmed by ITAAC. WEC has not provided this ITA AC. This is Open item 720.444F. The response to FSER open item 720.444F was provided by Westinghouse letter DCP/NRCl209. dated January 9,1998.} Operation of ADS stage 4 provides a vent path for the seere accident hydrogen to the steam generator compartments, bypassing the IRWST, and mitigating the condition required to produce a diffusion flame near the , containment wall. W Response: The above information is covered by item 38 of PRA Table (9 29. W Westinghouse
I NRC FSER OPEN ITEM
@ M Hiq Containment Isolation svutm Containment isolation valves in lines that represent risk-significant release paths are controlled by DAS in addition to PMS to further hmit offsite releases following core melt accidents. These lines are: containment air filter supply and exhaust RCDT out, and normal containment sump. He containment isolation vahes controlled by DAS are
' included as risk significant SSCs within D RAP. The operability of DAS actuation of these isolation valves is addressed by short term availability controls for DAS.
W Response: The above information is covered by item 39 of PRA Table 59 29. A sentence will be added to item 39 to address the availability controls on this DASfunction.
& actor Cavity Desien for Direct Containment Heating He reactor cavity and RPV arrangement provides no direct Dow path for the transport of particulated molten debris from the reactor cavity to the upper containment regions.
W Response: The above information is covered by item 29 of PRA Table 59 29. Mactor Cavity Desien for Ex Vessel Fuel Coolant Interactions ne design can withstand a best estimate ex vessel steam explosion without loss of containment integrity. W Response: The above information is cowred by item 28 of PRA Table 59 29. Peactor Caviiv Desien for Core Concrete Interactions De reactor cavity design incorporates features that protect against basemat melt through in the event of RPV failure. The cavity design includes:
- a minimum floor area of 48 m 2available for spreading of the molten core debris,
- layout, efevations, and flow areas of the reactor cavity and RCDT subcompartments and interconnecting ventilation duct consistent with Figure B 3 of Appendix B of the PRA and the supporting ANL analysis,
- a minimum thickness of concrete above the embedded containment liner cf 0.85 m.
- provisions to prevent core debris from passing into the sump via interconnecting pipelines embedded in the concrete floor and/or sump curb, and
- a sump curb of sufficient height and width to prevent molten core debris from overflowing or ablating through the curb.
[WEC still needs to confirm these items as pr of Open item 19.2.3.3.31 (Open item 720.418F). Note the response to F5ER open item 720.418F was provided by Westayhouse letter DCP/NRCil71, dated December 9.1997.) T Westingnouse l l
t e NRC FSER OPEN ITEM E Response: It is not technically accurate to say the reactor cavity design incorporates features that grotect against basemat melt through, but rather the plant is designedfor innessel retention of monten core debris and the reactor cavity incorperutes features that extend the time to basemat melt. tlsroush in the event of vesselfailure. Thefirst, third, andfifth sub bullet above are accurate. The founh sub-bullet is accurats y written as "there are no interconnecting pipelines embedded in the concrets floor and/or sump curb thereby preventing debrisfrom passing into the sump." Thefirst. third,founh, andfifth sub bullets ident(fy characteristics of the reactor cavity design, whereas the second sub bullet contains secondary irformation and is written too broadly, thus Westinghouse does not consider that it is an insight. Item 42 of PRA Table 39 29 willinclude the appropriate ir; formation per this Westinghoun response. A specific type of cor. crete is not specified for use in the basemat. n' Response: The above information is covered Iy item 41 of PRA Table 39-29. [hdronen leniter System The AP600 design includes a hydrogen ipniter system to limit the concentration of hydrogen in the containment during severe accidents. The features of the system are: 66 glow plug igniters distributed throughout the containment
- powered from 'he ' an safety related onsite ac power system, but also capable of being powered by offsite ac power, onsite non essential diesel generators, or non Class IE batteries via de-to-ac inverters
- manually actuated from the control room when core exit temperature exceeds 1200F, as the first step in ERG FR.C 1 to ensure that the igniter activation occurs prior to rapid cladding oxidation.
The igniter system is non safety related but is subject to investment prctection short term availability controls. The AP600 design also includes four passive autocatalytic recombiners (PARS) strategically located within the containment. The PARS are provided primarily to cope with hydrogen production during design basis accidents, but are expected to function to reduce combustible gas concentrations during seve:e accid-nts.
' W Response: Much of the above ir\ formation is covered by item 30 of PRA Table 39 29. As part ofitem 30, the igniters and PARS are introduced, and then the reader is directed to the Certified Design klarerial forfunher design information (i.e., number ofignters, etc). The operation of the igniters via the ERGS is also presented in item 30. A statement will be added to item 30 to cover the igniter shon-term availability control which is covered by SSAR section 16.3, 720.439F 7 '
g
O NRC FSER OPEN ITEM d2MAfgtv CQntammtDL$ pray A non. safety grade containment spray system is included in the APNX) design with the capability to supply water to the containment spray hende, from an esternal source in the event of a sesere accident. Loss of ac power does not contribute sigm0cantly to the core damage frequency, therefore, non safety telated containment spray does not need to be ac independent. De spray system comprises the following design features:
- two contamment spray ring headers equipped with a total of 66 spray nozzles and providing approsimately 80
> percent containment coverage.
- a 6 inch diameter supply pipe connecting the spray ring headers to the Gre protection system header inside containment, contaimng one normally closed, ait operated valve with remote actuation from the control room (V701), and one normally open, manual valve (V700),
- 6 inch diameter pipmg connecting the Gre neader inside containment to the Gre main header outside containment, and capable of being supplied from both the diesel driven 6te pump and the motor driven Gre pump.
1 De detailed design and location of all associated valves and connections will take into account expested radiation
~
levels and shielding requirements for any required local operator actions. De COL applicant willdevelop and implement guidar.cc and procedures for use of the non safety containment spray system as part of the COL Action item regarding accident management program.
}\' Response: As stated in item 36 of PRA Table 39 29, " Containment spray is not credited in the PRA. Failure of the nonsafety related containment spray does not prevent the plant achieving the safety goals."
Since the nonsafety related containment spray rystem is not credited in the PRA, it is not appropriate to provide the stqffs abose Julgn ccrtfipration .itatements in the PRA insights. - llowever, a staten snt will be added to its : 36 regarding the COL development of SAbfG usir.g the suggested framewors provided in WCAP 13914. (Nots WCAP 13914 is being revised to l mcorporate this aspect.) CDalainment Vent he following will be completed after W submits design description. His is idenuned as an Open item in FSER Secuon 19.2.5 (Open item 720 421F). In the event of a severe accident that results in gradual containment pressurization, the AP600 containment can be vented via the __, ime to prevent over pressure failure. Fmion praluct releases from the ,_ line are routed to the stack. i Valses in the __. line are qualified to operate at containment pressures corresponding to i.ervice Level C. The .,,,_. line is capable of withstanding the pressures associated with vent actuation at Service C. 720,439F 8
i l NRC FSER OPEN ITEM iii Detailed procedures for use of the containment vent system will be developed by the COL applies.nt, as part of the COL Action item tegarding accident management.
'ih : section contingent on WEC's response to staff request concerning providing a vent pursuant to 10 CFR 50.34(fX3Xiv).
)y Response: The response to FSER open item 720.421F was provided by Westinghouse letter DCP/NRC1194 dated December 18,1997. liased on the response to FSER open item 720.421F, an statement will be added to Table 39 29 to address the containment vent issue. The statement will be asfollows:
Capability esists to vent the containment via the RNS suction lines to the spentfuelpool, with the RCS depressurized and open to the containment atmosphere via either the ADS or the vesselfailure. (disposition = Appendis 01 In addition, a statement will also be added to Table 59 29 to address the COL development of SAAfGfor venting containment using the suggestedframework provided in WCAP.13914. AtgJdent Managemem
*lhe COL will develop and implement severe accident management geidance and procedures using the framework provided in WCAP.13914. Revision 2.
}X Resp <mse: The above irtformation is covered by a COL item described in PRA subsection 39.10.6.
120.439F 9
4 NRC FSER OPEN ITEM i ! Question: 720.461F (OITS #6487) In the response to RAI 720.306, the probability of failure to achieve containment closure is shown to be small and have insignificant impact on the estimated failure probability for containment isolation, flowever, the underlying flEP analysis includes averal ass.mptions that are either inconsistent with or not assured by the AP600 TSs. Specifically:
- a. it is assumed that c:.ch opch penetration will be manned by 2 persons having specific resp (msibility for closing the penetration. The means by which this as6umption will be met (e g.,TS, administrative controls) should be provided
- b. it is assumed that detailed written procedures for closing the openings will be developed and used. This should be identified as a COL action item
- c. in quantifying the llEP for failure to achieve containment closure, it is assumed that the hatches / penetrations are open at 28 hours (when the time to coolant boiling is about 17 minutes). Consequently, actions to achieve containment closure are tequired (and assumed) to take place under harsh conditions following steam release to containment. TS 3.6.8 does not permit the subject penetrations to be open unless they are capable of being closed prior to steaming into the containment. Since the minimum closure time in the scenario (45 minutes, based on 15 minutes for decision making and 30 minutes for implementation)is longer than the time to steam release to the containment, the situation which forms the basis for the llEP assessment would not be permitted.
The llEp analysis should be updated to be consistent with coneraints of the TSs. The time between loss of RNS and the receipt of the alarm inside containment, and the time required for the crew responsible for closing the equipment hatch to esit containment needs to be accounted for in the revised liEP analysis
- d. the alarm which provides the cue to the staff manning the penetrations should be identified, and TSs for the alarm should be provided Response;
- a. "Ihe assumptions regarding personnel responsibility for manning and closing the containment penetrations is disused below in item c. An administrative program, for which the COL applicant is responsible, is used to show compliance with the Technical Specifications.
- b. SSAR section 13.5 includes the COL information item (s) associated with plant procedures.
- c. According to TS 3.6.8, the COL applicant is required to ensure (through analysis) that containment penetrations can be closed prior to steaming into the containment following any event at shutdown.
The humu reliabi;ity analysis (llRA) of the capabihty to close seven containment openings during mid loop operation (doeurented in the response to RAI 720.306) was performed before TS 3.6.8 was written. The HRA made conservative assumptions; primarily, assuming the containment is open 28 hours after shutdown which could result in coolant boiling 17 minutes after the loss of RNS. It is also assumed in the lira that the contamment is habitabk 33 minutes after boiling / steaming occurs, and, therefoie, closing of the containment can 72a m W wwingrase
e NRC FSER OPEN ITEM 4 til contmue through this time frame. Westinghouse agrees with the NRC that, according to TS 3 6,8. this situation wculd not be permitted. Iloweser, it is shown in the following paragraphs that, based on the modeling guidelines of TilERP used in the llRA, the human error probability (llEP) of 2.2F'h03 for closing the contamment opening is reasonable and does not change, According to TS 3.6 8, if coolant tuling is espected 45 minutes af ter the loss of RNS, the COL applicant must demonstrate that the containment openings can indeed be closed within 45 minutes for the scenario. Since the time to boihng varies with decay heat levels and the time after shutdown when RNS failure occurs, the COL applicant would be expected to demonstrate containment closure capability for the most limiting cases; the hmiting case defined as conditions such that the time from loss of RNS to steam release into the containment approsimates the time to manually close the containment penetrations, in other words, for many scenarios involving the loss of RNS, the time to coolant boiling and steam release into the containment would be longer than the limiting cases, and, therefore, the crew could have longer time windows to close the containment penetrations. In that regard, for the current HRA,it is reasonable to assume the crew would have a time window of at least 45 minutes to close the containment penetrations, if the actual time to close those penetrations is anumed to be 45 mmutes. In the response to RAI 720.306, the !!RA assumes a time window of 50 minutes for closing the containment openings, and an actual time of about 45 minutes for the actions. The actual time is broken down into 15 mmutes for diagnosis of the event and 30 minutes for action execution. The actual time of 45 minutes is based on engineering judgement and is believed to be conservatise. Shortly after the loss of RNS, the control room crew would be prosided indications such as loss of RNS flow and high hot leg temperature. The control room crer is expected to diagnose the event, assess the urgency for closing the containment, and communicate to the local personnel the need to initiate containment closure. There is no alarm annunciaton in the containment for the loss of RNS. The control room crew is expected to use phones and/or loud speakers to communicate the plant status and the need for containment closure to the personnel working locally. According to the assumptions in the response to RAI 720.306, the control room crew is expected to diagnose the loss of RNS event and request action from the local crew within 15 minutes from the loss of RNS. Given that there are fewer operating /available systems during reduced inventory conditions than at pressurtied conditions, there would be fewer nuisance alarms in the control room; therefore, the cues for the loss of RNS at mid loop would be clear to the control room operators and the event would be quickly diagnosed.1herefore, the assumed actual time of 15 minutes for diagnosing the event and decision making is reasonable. The ilRA also assumes that each penetration would be manned by 2 persons. This assumption was based on the most limiting scenario; specifically, where it was assumed in the llRA that steam would be released into the containment about 17 minutes after the loss of RNS. Given the need for the COL appheant to ensure containment closure capability, it is believed that each penetration could be manned by two persons for plant status with the more limiting time to boiling, and by one person for scenarios with longer times to taling. On the other hand, the COL applicant may decide that the penetrations do not have to be manned where very long time window s exist. The basic assumption is that one person would close each penetration and a second person would ass st in serifying the action. The COL applicant may determine that one verifier can he responsible for seseral penetrations. 72 m 2 W wesunghouse 4
e NRC FSER OPEN ITEM A in current operating plant practices, local activities are checked by an independent crew member; the same practice is capected for actnities on the AP600. In applying the 111ERP ruethodology, it could be assumed that each penetration is manned by two persons or by one person with another person being available to serify the action. The curren: HRA assigns a basic human error probabihty (DHEP) of 5.0E 02 for the second crew member manning the penetration. If the second crew member is taken to be a checker, then a IlllEP of 1.6E 02 can te assigned; this would reduce the overall llEP of 2.28E 03. The llRA estimated that failure to diagnose the event has a probabihty less than 1.0E 05. It was assumed tbt the event would be diagnosed in the control room and a local alarm of the esent would be provided. As stated atxne, the event would indeed be diagnosed in the control room, but local crew awareness of the event would occur through communication and direction from the control room. The current diagnosis error probabihty calculation is still valid even thoupi no local alarm annunciation is provided.
'the IIRA assumes a high stress environment for closing the containment penetrations, shown by a multiplier of 5 to the BilEP. A high stress factor is still applicable for the more limitmg cases, as defined previously.
Dased on the discussion abose. Westinghouse believes the HEP of 2.28E 03 for closing the containment penetratwn during the loss of RNS at mid hop to te a reasonable estimate, even though it uses a scenario not permitted by Technical Specification 3.6.8, and the HRA for this activity need not be revised. d) As discussed in item c, there is no local alarm annunciation which provides cues to the staff manning the contamment penetrations. The staff would be directed by communications from the main control room. PRA Revision: None. 720 e 3 W wesunews.
O NRC PSER OPEN ITEM p__9r Ouestion: 720.462F (OITS #6488) Opiation of stage 4 of ADS would significantly impact the time to steam release to the containment, and the likelihood of achieving containment closure. Please explain how operation (or failure) of 4th stage ADS should be accounted for in determining the time to steam release to containment under TS 3.6.8. Response:- Technical Specification 3.6.8 requires containment closure capability during shutdown operations in Modes $ and
- 6. In these modes, the containment may be open, but potential leak paths must be capable of being closed prior to the time steam covi be released into containment following a loss of RNS cooling event. Figure 3.6.81 provides a series of plots v,'.. h shows the time permitted for containment closure for the various RCS conditions applicable to Modes 5 and 6. Tha following RCS conditions are identified la the Figure:
Case 1. Mode 5, RCS intact Case 2 Mode 5, Mid loop Case 3. Mode 6 Cavity not Hooded Case 4. Me:le 6 Cavity flooded For these four cases. ADS stage 4 operatior is assumed for Case 2 and Case 3 curves. For these configurations, operation of ADS stage 4 would be assumed to occur early in an event, and steam release to containment could occur fairly soon (less than I hour). For Cae 1, the RCS intact, and the PRHR heat exchanger is available. In such a condition, ADS stage 4 operation would *.ot occur early (if at all), and steaming to containment would occur as a result of the PRHR boiling the water in the IRWST. For Case 4, the reactor cavity is flooded, and the reactor vessel head and upper internals are removed. In this condition, operation of the 4th stage ADS is not required, and its effect on steam release into containment is not relevant. PRA Revision: None, f 1 720 4 2F.1 Y WDElingheute
Enclosure 2 to Westinghouse 14tter DCP/NRCl214 January 15,1998 n u s .pt}}