Text

Forwards Responses to Followon Questions Re AP600 Ssar Section 14.3 & AP600 Insps,Tests,Analyses & Acceptance Criteria.Westinghouse Status Column Will Be Changed to Closed in Open Item Tracking Sys,As Result
	ML20199B959
Person / Time
Site:	05200003
Issue date:	11/11/1997
From:	Mcintyre B; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NSD-NRC-97-5426, NUDOCS 9711190161
	Download: ML20199B959 (105)
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Westinghouse. Energy Systems Sx355 Electr!c Corporation

^"85* *8*858 523"355 DCP/NRC1125 -

NSD-NRC 97-5426 Docket No.: 52-003

" November 11,1997 bocument Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 NITENTION: T.R. QUAY SUllJECT: RESPONSES TO FOLLOWON QUESTIONS REGARDING TIIE AP600 SSAR SECTION 14.3 AND TIIE AP600 INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE "RITERIA (ITAAC)

INar Mr. Quay:

l'nclosed are three copies of Westinghouse's revised response to RAI 640.60 related to the Standard Safety Analysis Report Section 14.3, Certified Design Material. This revised response incorporrtes comments from the staff received at the ITAAC Task Group meeting held in White Hint on September 25,1997, as documented in your letter dated October 2,1997. This revised RAI response also includes SSAR markups of changes made in Section 14.3 to be incorporated in the Al'600 SSAR, Revision 17.

This submittal closes, from Westinghouse's perspective, open item 5719. As a result, the Westinghouse status column will be changed to " Closed" in the Open item Tracking Sys;em (OITS).

The NRC should review this response and inform Westinghouse of the status of this open item to be designated in the "NRC Status" colur.m of the OITS.

Please contact Mr. Eugene J. Piplica at (412) 374 5310 if you have any questions concerning this

' transmittal.

/

J~ Yh b 11rian A. McIntyre, Manager ;;gOf Advanced Plant Safety and Licensing jml- {

\

Enclosure i

J. M. Sebrosky, NRC (w! Enclosure) cc:

J. N. Wilson, NRC (w/ Enclosure)

N. J. Liparuto, Westinghouse (w/o Enclosure)

DR 0 m es ; n u m . m1 gg.EHIEjNEN

NRC REQUECT FOR ADDITIONAL INFORNvATION A

l Question 640.60, Revis/on 1 Re:

In AP600 SSAR Revision 14 dated June 17,1997, Westinghouse provided Tables 14.3 8, which identified important design features that were credited in various analyses to their treatment in Tier 1. However, these tables do not contain a disposition column which makes it dif6 cult for a reviewer to determine w hat ITAAC verines the important design features. Therefore, the staff requests that Westinghouse provide a more detailed cross reference that provides a disposition column (i.e., verifying ITAAC) for the important design features. General Electric and ABB-CE provided this type of information to the staff for their ITAAC in a March 31,1994, letter, and a June 10,1994 letter, respectively, l' Response, Rev/s/on 1 :

Tables 640.60-: through 640 60-7 provide a cross reference between Tables 14.3-2 through 14.3-8, respectively, of the AP600 SSAR, revision 14 and the AP600 Cer'ified Design Material, Revision 3. The nrst three columns of the attached tables contain the same SSAR References, Design Features, and Values listed in the corresponding tables of Section 14.3 of the AP600 SSAR. The last three columns, ITAAC, System, and ITA provide the cross reference to the AP600 Certified Design Material. These columns list for each Design Feature the applicable section number (ITAAC), the system acronym (System), and the inspection, Test, Analyses, and Acceptance Criteria item designator (ITA). These p"ameters are further described in Section 1.0, introduction, of the AP600 Certified Design Material.

l Tables 14.3 2 through 14.3 8 of the A P600 SSAR have been reviewedfor conshtency and completeness. Changes l made as a result of this review are attached to this RAI response. These changes include the elimination of two l duplicate entries, typographical corrections and the deletion of an incorrect reference to SSAR Section 18.8.3.2 l In Tabh 14.3 7, Radiological Analysh. In addition, design changes to the .tfain Control Room Emergency l liabitabillo. System (VES) have been incorporated. Changes are marked with a har in the left column. Additions l are shown in bold Italics and deletions are shown by a unkeews.

I l Concurrent with the review of Tables 14.3 2 through 14.3-8 above, a review of Tables 640.60-1 through 640.60-8 l has been performedfor conshtency with the SS4R tables. The same changeformat as above h used.

References:

None SSAR Revisions:

l SSAR Secthm 14.3, change pages to Tabies 14.3-2,14.3-4,14.3-5,14.3 6 and 14.3-7 anached.

640.60-1 T Westinghouse Revision 1 a

. NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 1 ci 24)

ITAAC References for SCAR Table 14.3 Design Basis Accedent Analysis SSAR Reference Design Feature Value ITAAC ' System ITA Secton 5.1.2 Sa'ety valves are installed above and conneded b the pressunzer to 2.1. 2 RCS 1 peovide overpreme protechen for the reactor coolant systerrt 8.a Sechon 5_1. 2 The RCS has two hot legs and four cou legs. 2.1.2 RCS 1 Secton 5.1. 2 The RCS has two steam gene ators and four reactor coolant pumps 2.1. 2 - 'RCS 1-Sechon 5.1.' 2 The RCS contains a pr'asurizer and a surge line conneded to one hot leg 2.1.2 'RCS 1 Sechon 5.1.3.3 Rotating inertia needed for flow coast 4m is provided. 2.1.2 RCS 8.b Table 5.1-3 Thermal 1,ign !!ow rate with 10% tube plugging (gpmSoop) 94.800 2.1.2 RCS 9. a Table 5.1- 3 in tial rated reactor core thermal power (MWt) 1933 INTRO NSSS -

Sechon 5.2.2 Reactor coolant system and steam system overpressure protechon during 2.1.2 RCS 8. a power operation are provded by the pressurizer safety valves and the 2.2.4 SGS 8.a steam generator safety valves, in axijuncbon wih the acbon of the PMS. 2. 5. 2 PMS 6.a i

Wg 640.60-2 Rev on i

NRC REQUEST FOR ADDITIONAL INFORC.1ATION Tatne 640.60-1 (Sheet 2 of 24) -

ITAAC References for SSAR Tatde 14.3 Desegn Basi

Accsdent Analysis SSAR Reference Design Feature Value ITAAC System ITA Sechon 5.22.1 Safety valve capacity ensts to prevent exceedng 110 percent of system 2.1.2 RCS 8.a desgn pressure for the fo! lowing events:

-Loss of elecincal load and,or turtune inp 4)ncontroRed rod withdrawal at power

-Loss of reactor coolant flow

-Loss of normal fredwater l -Lc= c' &c p = to % d2 rdrt =d '&d=

-Loss of offsite power to the station auxdianes Sechon 5. 2. 2.1 Overpressure protechon for the steam system is provxled by steam 2.2.4 SGS' 81a generator safety valves Sechon 5.3.2.3 NorHjestruchve exammahon (NDE) of the rea:: tor vessel and its 2.1.3 RXS 3 aopurtenances is conducted in accordance with ASME Code Sechon 111 requirements,.

Secton 5. 3. 2. 5 The initial Charpy V-notct; minimum upper shelf fracture energy levels for 2.1.3 RXS 7 the reactor vessel be'.ihe base metal traverse drechon and welds are 75 foot-pounds, as reesired by Appendix G of 10 CFR 50.

l Sesteen 5.3.2.5 he W C5py v ~0t2 m-9 _m 62 rdre erg; ' =5 6 t-3 RXS -7 l the read ~ .ed t:5= bre d' te=0 ed :~' _1 : 75 l '^^t p~d, = ragh-d by ?W: G ^'10 CcT'!A ey 640.60-3

== Revisson 1

NRC REQUEOT FOR ADDITIONAL INFOhMATION 4

Table 640.60-1 (Sheet 3 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis Sechon 5.4.1.3.4 11is irnportant to reactor potecbon that the reactor coolant conbnues to 2.1.2 RCS 8.b flow for a bme after reactor inp and loss of electncal power. To powde this flow, each reactor molant pump has a high-inetta rotor.

Sechon 5.4.1,3.4 A safety 4 elated pump trip occurs on hgh beanng water temperature. 2.52 PMS 6.b Sechan 5.4.5.2.3 Power to the pressunzer heaters is blocked when the core makeup tanks 2. 5. 2 PMS 6.b are actuated Secnon 5.4.6 Automatic depressurizabon system stage 1,2 and 3 valves are hw 2.1.2 RCS 1 l to the pressurizer and discharge via the spargers to the in-containment refuehng water storage tank.

Sechon 5. 4. 6 Automat: depressurizabon system stage 4 valves are connected to each 2.1.2 RCS 1 hot leg W M @ se ay

NRC REQUE3T FOR ADDITIONAL INFORMATION A

Table 640.60-1 ISheet 4 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accedent Analysis SSAR Reference Design Feat 7e Value ITAAC Systein ITA Sechon 5 4.9.3 in the analysis of overpressure events, the pressunzer safety va!ves are 2.1.2 RCS 6.b assurned to actuate at 2500 psia The safety valve flowrate assurned is based on ful flow at 2575 psia, assuming 3 percent accumutabon.

Secton 5. 4. 9. 3 The pressunzer safety valves prevent reactor coolant system pressure from 2.1.2 RCS 8.a exceeding 110% of systsrn desgn pressure.

Table 5. 4- 1 FAnanum reactor coolant motor / pump moment of inetta (It>ft'). 2 5.000 2.1.2 RCS 6.b Table 5. 4-11 Reactor Coolant System Desgn Pressure Setbngs: 2.1. 2 RCS 8. a

- Safety valves begin to oper. (psig) 2485 Table 5. 4-17 Pressurizer Safety Vafves - Desgn Parameters: 2.1.2 RCS 8.a

- Number 2 8.b

- Mnimum required releving capaaty per valve (itmhr) 2400.000 l - Set pressure (psig) 24851 25 Section 6.1.2.1.3 The exterior of the containment vessel is coated with the same inorganc 2.2.2 PCS 8 b. i zine as is used inside of the containment.

Figure 6.2.2-1 The passive containment coohng system consists of a water storage tank. 2.2.2 PCS 1 cooling water flow disdurge path to the cc.itainment shel, a wier d:stribution system for the contauunent sheD, and a coohng air flow path.

W-WBStiPgh00S0 Revisson 1

NRC REQUE3 FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 5 of 24)

ITAAC Referen<:es for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Design Feature Value ITAAC System ITA Figure 6.2.2-1 The minimum duraton the PCS mohng water fbw is provded from the 2 72 2.2.2 PCS 8 a .ii PCCWST (hours)

TWe 6.2.2-1 The water coverage of the containment shell exceeds the amount used in 2.2.2 PCS 8. b. i the safety analysis.

Table 6.2.2-1 7.e minimum drain Tow rate capacity of the upper annulus drain {gpm). 2 450 2.2.2 PCS 8.d Table 6.2.2-1 The minimum makeup flow rate capabihty from an extemal source to the 2 62 2.2.2 PCS 8.e PCS water storage tank (gpm).

Table 6.2.2-1 The minimum makeup flow rate capability from the PCS water storage tank 2 50 12.2 PCS 8. f. i to the spent fuel pit (glm).

Table 6. 2. 2- 1 The minimum PCS water storage tank volume for makeup to the spent fuel 2 400.000 2.2.2 PCS 8. f s pit (non-coincident with PCS operaton) (ga!!ons).

Table 6.2.2-1 The mtnimum long term makeup capabhty from the PCCAWST to the 24 2.2.2 PCS 9. a PCCWST (days)

Table 6.2.2-1 The minimum bog term makeup flow capability from the PC ' GT to the 2 62 2.2.2 PCS 9.b PCCWST (gpm)

Table 6.2.2-2 The first or top standpipe's elevaton above the lowest or bottom standpipe 21.710.25 2.2.2 PCS 8.aii (feet).

W==

Westinghouse 84 .88-8 Revision 1

NRC REQUEDT FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 6 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SS AR Reference Design Feature Value ITAAC System ITA Table 62,2-2 The second standpipe's elevation above the lowest or bottom standppe 14.210.25 2.2.2 PCS 8. a iii (feet).

Table 6.2.2-2 The third standpipe's elevation above the lowest or bottom standppe (feet). 6.710.25 2.2.2 PCS 8. a W Figure 6. 2. 2- 3 The minimum passive containment coohng water flow rate with water 2 71.5 2.2.? PCS 8. a i inventory at a height above the lowest standpipe of 13.5510. 25 ft (gpm)

Figure 6. 2. 2- 3 The mirumum passive contanment coohng water Row rate with water 2 442 2.2.2 PCS 8. a i inventory at a height above the lowest standppe of 23.7510.25 ft (gpm)

Figure 6.2.2-3 The minimum passive containment coohng water flow rate with water 2 122 2.2.2 PCS 8. a i inventory at a height above the lowest standpipe of 2 .6510.25 ft (gpm)

Secton 6.3 The paWve core coohng system povides emergency core decay heat 2.2.3 PXS 8.b removal dunng transients, acodents or whenever the normal hea' removal paths are lost Secton 6.3 The passive core cochng system provides makeup and boraton durra 2.2.3 PXS 8.c transients or acodents when the normal reactor coolant system makeup cupply from the chemical and volume control system is unavailable or is insuffcient W Westinghouse .

- Revision 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 7 of 24) -

ITA AC References for SSAR Table 14.3 Design Sasis Accident Analysis SSAR Reference Design Feature Value ITAAC System (TA Sechon 6. 3.' 1.1 The passive core coobng system is desgned to provide emergency core 2.2.3 PXS 8.b cooling dwing events involving increases and decreases in secondary side heat remeval and decreases in reactor coolant system inventory.

Secton 6.3.2.1.1 The heat exchanger consists of a bank of C-tubes, undreded to a 2.1. 3 PXS 1 tubesheet and channel heat arrangement at the top (inlet) and bottom (out!et). The passive exchanger connects to the reactor coolant system l through an inlet hne from one reactor coolant system hot leg and an outlet hne to the assodated steam generator cold leg pienum (reactor coolant pump suction).

Section 6.3.2.1.1 For the passive residual heat removal heat exchanger, the normal water 2.2.3 PXS 8.b temperature in the inlet hne will be hotter than the discharge hne.

Sechon 6.3.2.1.2 The actuation of the mre makeup tanks following a steam kne break 2.2.3 PXS 11.b provdes injechon of borated water via water recirculabon to trutigate the reactivity transient and provide the required shutdown margin.

S cbon 6 3.2.2.3 The in<ontainment refuelmg water storage tank contains one passive 2.2.3 PXS 1 residual heat removal heat exchanger Sechon 6.3 2.2.6 Automabc depressurizabon system stage 1,2 and 3 valves are connectef 2.1.2 RCS 1 l to the pressurizer and discharge via the spargers to the in-containment refueling water storage tank.

W gg

=

640.60-8 Revision 1

NRC REQUEST FOR ADDITIONAL INFORL1ATION Table 640.60-1 (Sheet 8 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accdent Analysis SSAR Reference Design Feature Value ITAAC System ITA Sechon 6.3.3.2.1 For a loss of man feedwater event, the passive resdual heat removal heat 2. 5. 2 PMS 6.b exchanger is actuated. If the core makeup tanks are not initial!y actuated, they actuate later when passive resdual heat exchanger coohng suffoenty reduces pressunzer level Seebon 6.3.3.2.2 For a feedwater system pipe fadure event, the passive residual heat 2. 5. 2 PMS 6.b removal heat exchanger and the core makeup tanks are actuated.

Sechon 6.3.3 3.1 For a steam generator tube rupture event, the nonsafety-related makeup 2. 5. 2 PMS 6.b pumps are automaSca!!y actuated when reactor coolant system inventory decreases and a reactor trip occurs followed by actuation of the startup feedwater pumps. Makeup pumps automatically function to mantain the programmed pressurizer level. The core makeup tanks subsequenny actuate on low pressurizer level, if they are not already actuated. Actuabon of the core makeup tanks automaticaDy actuates the passive resdual teat remova system heat exchanger.

W Westiligh00S8

- Revision 1

s NRC REQUEST FOR ADOfTIONAl. INFO 3MATION r

Table 640.60-1 (Sheet 9 of 24)

ITAAC References for SSAR Table 14.3 Desegn Basis Accident Analysis SSAR Reference Desagn Feature Value ITAAC System ITA Sechen 6.3.6.1 The piping resistances connechng the fo80mng PXS components and the 2.2.3 PXS- 8-RCS are bounded by the resistances assumed in the Chapter 15 safety analysis:

- Core makeup tanks

- Accumulators .

- IrH:entamment refue!ng water storage tank

- Cudeineal recmilabon

- Passive resdual heat removal heat exdanger

- Automabc depressunzabon system vahm Secbon 6.3.6.1.3 The bottom of the core makeup tanks are located above the reactor vessel 27.5 2.2.3 PXS 8. c drect vessel injechon nozzle centerline (ft).

l Sechon 6.3.6.1.3 The bottom of the instamment refueleg water storage tank is located 23.4 2.2.3 PXS- 8. c above the direct vessel injechon nozzle centerline (ft).'

Figure 6. 3- 1 The passive core coorng system has two direct vessel injechon Enes. 2.2.3 PXS 1 Table 6. 3- 4 The passive core coohng system has two core makeup tanks. each wth a - 2 2,000 2.2.3 PXS 1 minimum required volume (ft') 8. c Table 6. 3- 4 The passive core cochng system has two accumulators, each with 'a 2 2,000 2.2.3 PXS 1 minimum required volume (ft) 8.c i Table 6. 3- 4 The passive core coohng system has an inctainment refuehng water 2 557,000 2.2.3 PXS' 1 storage tank eth a muumum required water volume (gabons) 8.c Wg 640.60 Reviseen 1

NRC REQUEST FOR ADDITIONAL. INFORMATION Table 640.60-1 (Sheet 10 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SCAR Reference Design Feature Value ITAAC System ITA Table 6.3-4 The passve core cochng system has two pH adjustment baskets each wJh 2 107 2.2.3 PXS 8. d a minimum required volume (ft ')

Table 6.3-4 The passive residual heat removal heat exchanger mmanum heat transfer 2 106.000,000 2.2.3 PXS 8b rate (BTU!hr)

Secbon 7.1.2.11 Isolabon devices are used to maintam the electncal odependence of 2.5.2 PMS 7.a divisions and to see that no oterachon occurs between nonsafety-related 7. b systems and the safety-related system. Isolation devices serve to prevent credble faults in circuit from propagating to another circuit.

Sechon 7.1.4.1.6 The protechon and safety monitanng system equipment is seismcally 2 5. 2 PMS 2 qualified to meet design basis earthquake levels.

Sechon 7.1.4.1,6 The abihty of the protecbon and safety monitoring system to initate and 2. 5. 2 PMS 3 accornplish prctective funcbons is maintained despite degraded conditions caused by intemal events such as fire, flooding, explosons, missiles, electrical faults and pipe whip.

Secton 7.1.4.1.6 The design of the protection and safety morutorng system equipment has 2.5.2 PMS 4 l adebonal margin to accommodated a loss of the normal HVAC.

Sechon 7.'.4.2.6 The flexibihty of the protecton and safety monitonng system enables 2.5.2 PMS 5. b physical sepa ation of redundant divisons.

~

W Westinghouse 640;som Revisson 1

NRC REQUEST FOR ADDITIONAL INFORMATION . ,. .,

sin '%

Tatde 640.60-1 (Sheet 11 of 24)

ITAAC References for SSAR Table 14.3 Desegn Basis Accident Analysis SSAR Reference Desegn Feature Value - ITAAC . System . ITA Secbon 7.2.2.2.1 ine potection and safety rnonitoring system intbates a reactor tnp . 2. 5. 2 PMS 6.a whenever a condebon monitored by the system reaches a peset level.

Secbon 7.2.2.2.8 The reactor is tripped by actuabng one of two manual reactor inp mntrois 2. 5. 2 PMS 6. c from the main control room.

Sechon 7.3.1 2.14 The desswdeed water system isolation valves close on a signal from the 2.12 CVS . 8. b protechon and safety monitoring system denved from either a reactor Inp signal, a source range flux doubling signal, low input vo' age to the 1E de and urunterrupbble power supply battery chargers, or a safety injecton signal.

Sechon 7.3.1.2.15 The chemcal and volume control system makeup Ime isolatun valves 2.3.2 CVS 8.c automatcally close on a signal from the protection and monitanng system denved from either a high-2 pressurizer level, high steam generator level signal, or a safeguards sgnal concident with high-1 pressunzer level.

l Secton 7.3.1.2.2 The in-containment refueling water storage tank is aligned for injechon 2. 2. 3 _ PXS 11.b upon actuabon of tne fourth s@ a.lomate depressunzabon system via the protechon and safety monitoring system.

Sechon 7.3.1.2.3 The core makeup tanks are aligned for operabon on a safeguards actuaban 2.2.3 PXS 11.b signal or on a low pressurizer level sgnal via the pumciksi and safety nuatciig system.

p- 340.60-12 Revesson 1

NRC REOU5ST FOR ADDITIONAL 'INFORWIATION

' fi 2

.u Table 640.60-1 (Sheet 12 of 24) -

ITAAC References for SSAR Table 14.3-2 'hsegn Basis Accitient Arwysis SSAR Reference Design Feature Value ITAAC System ITA ' i Sechon 7.3.1.2.4 The fourth stage valves of the automabc depressunzabon sys%m receive a 2.1.2 RCS. 11.b signal to open upon the won.aba.o of a low core makeup tmk water level and low reactor coolant system pressure foRowrig a preset hme delay aber the third stage depressunzabot valves receive a signal to open via lhe protechon and safety monitonng system.

Sechon '7 3.1.2.4 The first stage valves of the automatic depressunzation system open upon 2.1.2 RCS 11.b' receipt of a signal generated from a core makeup tank infechon abgnment signal coincident with core makeup tank wate level less than the low-1 setpoint in either core makeup tank via the ruim.:mm and safety sinsuksig system.

Sechon 7.3.1.2.4 The second and third stage valves open on brne delays folkwnng 2.1.2 RCS 11.b-generation of the first stage actuation signal via the protecton and safety monitanng system.

Sechon 7.3.1.2.5 The reactor coolant pumps are inpped upon generabon of a safeguards 2.1.2 RCS 11.b actuahon signal or upon generabon of a low pressurizer water level signal.

Sectm 7.3.1.2.7 The passive residual heat removal heat exchanger control valves a*e 2. 2. 3 ' PXS 11.b opened on low steam generator water level or on a CMT actuation signal via the protection and safety moi.itoring system.

gg 640.60-13 Revisson 1

NRC REQUEST FOR ADDITIONAL INFORMATION _

Table 640.60-1 -

(Sheet 13 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Desegn Feature Value ITAAC System ITA. .

Section 7.3.1.2.9 The contamment recrculabon isolaban valves are r pened on a safeguards 2.2.3 PXS- 11.b l- actuaban signal in co h with low in-contanment refuehng water storage tank water level via the puieh and safety monitoring systent Secbon 7.3.2.2.1 The protection and rnandonng system automatically generate an actuabon 2. S. 2 PMS 6.b signal for an engmeered safety feature whenever a mondored <xindebon reaches a preset level.

Sechon 7.3.2.2.9 Manual initiation at the system-level exists for the engmeered safety 2.S.2 PMS 6. c features actuabon-Secbon 7.4.3.1 if temporary evacuation of the main control room is requeed because of 3.2 HFE 12 some abnormal main control room conddion, the operators can estabksh and maintain safe shutdown conditions for the plant from outside the main control roorn through the use of controls and mcn'ung located at the remote shutdown workstabon. -

Secbon 7.4.3.1.1 The remote shutdown workstaban equipment is salar to the operator 3.2 HFE 9 workstabons in the mam <xmtrol room and 4 designed to the same 11; standards. One remote shutdown workstaban is provided.

Sechon 7.4.3.1.3 The remote shutdown workstation achieves and maintans safe shutdown 3.2 HFE- 12-cordbons from full power condsbons and maintains safe shutdown conditons Ifeeafter.

W-640.60-14 Revneson 1

NRC REQUEST FOR ADDITIONAL INFORMATION A

Table 640.SO-1 (Sheet 14 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Acadent Analysis SSAR Reference Design Feature Value ITAAC System ITA Secbon 7.5.4 The protecbon and safety nun'uig system provdes sgnal conditioning. 2.5.2 PMS 8.a communcations, and display funcbons for Category 1 vanables and for Category 2 vanables that are energized from the Class 1E uninterruptbie power supply system.

Secbon 7.6.1.1 An intertock is provded for the normally dosed motorgerated normal 2. 5. 2 PMS 9. c residual heat removal system inner and outer suct.on isolation valves.

Each valve is intertocked so that it cannot be opened unless the reactor coolant system pressurr is below a preset pressure.

Sechon 8.3.2.1.2 The non-Class 1E de ar:d UPS system (EDS) mnsists of te ekdr;c power 2. 5. 2 PMS 5. a supply and distnbubon equrprrent that provides de and uninterruptde ac power to nonsafety4 elated loads.

Secbon 9.1.1.2.1 In the unlikely event of a d'apping of an unirrafated fuel assembly. 2.1.1 FHS 6 aadental deformabon of the fuel rack wiR be determined and evaluated in the crith:ahty analysis to demonstrate that it does not cause critcakty caiterion to be vio!ated.

Secbon 9.1.2.2.1 in the unhkely event of a dropping of an irradiated fuel assembly, 2.1.1 FHS 6 accdental deformation of the fuel rack win be teterrruned and evaluated in the criticahty analysis to demonstrate that it does not cause onbcahty criterion to be violated.

Revisson 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 15 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Design Feature ,

Value ITAAC System ITA Sechon 9.1. 3. 5 The spent fuel pod is designed such that a water level is matnta:ned 2.3.7 SFS 8b above the spent fuel assemblies for at least 7 days fonowng a loss of the spent fuel cochng system usag only safety-related rnakeup water sources l (See Table 9.1- 4).

Sechon 9 1.3.5 The spent fuel pool molag system includes safety-related wims b 2.3.7 SFS 8b establish safety-related makeup to te spent fuel pool fonowing a desgn basis event induding a seisme event Secton 9.1.4.1.1 in the event of a safe shutdown earthquake (SSE), handhng equ:pment 2.1.1 FHS S cannot fail in such a manner as to prevent required funcbon of setsme Category 1 equ:pment Section 9.3.6.3.7 The diemcal and volume control system contains two redundant safety- 2.3.2 CVS 8.b related valves to isolate the demineralized water system from the makeup pump suchon Sechon 9.3.6.3.7 The chemmal and volume control system contains two safety-related valves 2.3.2 CVS 8. c to isolate the makeup now to the reactor coolant system.

Secbon 9.3.6.4.5 The chemical and volume control system contains two safety-related valves 2.3.2 CVS 8. c to isciate the makeup flow to the reactor cooldit system.

W gggg 640.60-16 Revision 1

NITC REQUEST FOR ADDITIONAL INFORMATIGN Table 640.60 (Sheet 16 of 24) iTAAC References for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Design I-eature Value ITAAC System ITA Secton 9.3 6.4.5.1 The chemical and volume control sydem contains two redundant safety- 2.3.2 CVS 8. b related valves to isolate the deminerdized water system from the makeup pump suction Sectxm 9.3.6.7 The demmerali7ed water system isolabon valves close on a signal from the 2.3.2 Ci/S 8.b protection and safaty monitority sys'em derived from either a reactor inp signal, a source range flux doubling signal, low input voltaje to the 1E de and uninterruptble power supply battery chargers, or a safety injechon signal. ,

Secton 9.3.6.7 The chemical and volume control system makeup line isolaban valves 2.3.2 CVS 8. c automatically close en a signal from the protectni and safety monitanng system denved from either a high-2 pressurizer level, high steam gererator lew! signal, or a safeguards signal coincident with high-1 pressunzer level.

Sechon 10.1.2 Safety valves are provided on both main steam sines. -

2.2.4 SGS 1 Sechon 10.2.2.4.3 The flow of th: main steam entering the high-pressure turbine is controlled 2.2.4 SGS 9. a by four stop valves and four goveming control valves. The stop valves are dosed by actuabon of the emergency trip syvem devices Secton 10.3.1.1 The main steam supply system is provided with a man steam isolabon 2.2.4 SGS 8. b valve and associated MSIV bypass v4ve on each main steam hne from th, 8. c respective steam generator.

N- W8Sllfigh00S8 .

Revisson 1

NRC REQUEST FOR ADDITIONAL INFORMATION mr Tatde 640.60-1 (Sheet 17 of 24) '-

ITAAC References for SSAR Table 14.3 Desegn Basis Accedent Analysis SSAR Reference Desegn Feature Value ITAAC System ITA Sechon 10.3.1.1 Man steam isolation va!ve (MSIV) prevent the uncontrosed blowdown of ' . 2. 4 i SGS . 8. a -

more than one steam generator and isolate nonsafety-related portons of 2.5.2 PMS 7. a the system.

Secbon 10.3.1.2 Power. operated atmosphenc relef valves are provided to allow controred 2.2.4 SGS 9.. b cooldown of the steam generator and the reactor coolant system when Ile condenser is not available Secton 10.3.2.1 T% main steam supply system includes 2.2.4 SGS' .1

-One main steam isolatx,n valve and one main steam isolabon vahe bypass valve per main steam line.

-Main steam safety valves. .

-Power <perated atmosphenc relief valves and upstream isolabon valves. .. t Secbon 10.3.2.3.2 in the event that a desgn basis accident occurs, which results in a large 2.2.4 SGS 8. b steam line break, the main steam isolabon vates with assousted main .8.c-

~'

steam isolabon bypass valves automabcaRy dose. 2. 5. 2 PMS 6. b .

Figure 10. 3. 2- 1 The steam generator system consists of two main steam, tao main 2. 2. 4 ' SGS ,1 feedwater, and two startup feedwater h Table 10. 3. 2- 1 Desgn data for main steam supply system vtes: 2.2.4 SCS -- 8.a

-Number per main steam kne 3 4

-lVhnnwm relieving capacity per valve at 110% of desqrt pressure (bhr) 1,540.000 W=

640.60-18 Revesson 1 3

NRC REQUEDT FCR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 18 of 24)

ITAAC References for SSAR TatAe 14.3 Design Basis Acadent Analysis SSAR Reference Design Feature Value ITAAC System ITA Table 10. 3. 2- 2 The moimum ibw capacity of the steam panerator safety valves (Itmtr) 2 4,600,000 2.2.4 SGS 8. a i Table '0.3.2-2

. The mannum set pressure of the steam generator safety valves (ps4 s 1.195 2.2.4 SGS 8. a i Sechen 10.3.8.3 The safety-related porbons of toe steam generator blowdown system are 2.2.4 SGS 8. b located in the contanment and auxiliary bulldogs and are desgned to remain funchonal after a safe shutdown earthquake.

Secton 10.4.7.1.1 Double valve main feedwater isolation is provided via the man feedwater 2.2.4 SGE 8. b control valve and rnan feedwater isolation valve. Both vaims dose 8. c automatically on man feedwater isolation sgnals, an appropriate

. engineered safety features isolation signal, within the bme establisted with the Technical W4k-is, Sechon 16.1. The startup feedwater control valve also serves as a containment isolabon valve.

Seebon 10.4_7.1.1 The condensate and feedwater system provdes redundant isolabon valves 2.2.4 SGS 1 for the main fedwater lines routed into conta:nment.

Secton 10.4.7.1.1 For a main feedwater or man steam kne break (MSLB) inside t*e 2.2.4 SGS 8.b cor.tainment, the condensate and feedwater system is designed to kmrt high energy fluid to the broken loop.

W WOSilligh00S8 640.60-19

-- Rev h 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table S40.60-1 (Sheet 19 of 24)

ITAAC References for SSAR Table 14.3 Desegn Beeis Accident Analysis SSAR Reference Desegn Feature Value ITAAC System ITA Secbon - 10.4.7.1.2 The booster / main feedwater pumps are tripped simultaneously wHh the . 2. 2. 4 SGS 9. a

- feedwater isolabon signal to dose the man feedwater isolabon vahes.

Sechon 10.4.7.2.1 The main feedwater pumps and booster pumps are inpped wdh the 2.2.4 SGS 9. a feedwater isolabon signal that closes the main feedwater isolahan vahes.

The same isolabon sign'al closes the isolabon vahe in the cross connect line between the main feedwater pump drscharge header and the startup feedwater pump discharge header.

Sechon 10.4.7.2.2 One MFIV is installed in each of the two main feedwater lires outside the 2.2.4' SGS 8b contsnment and downstream of the feedwater control valve. The MFIVs '

are installed to prevent uncontrolled blowban from tie steam generators '

in the event of a feedwater pipe rupture. Tne main feedwater check valve provides backup isoiabon. !n the event of a secondary side pipe rupture

' inside the containment, the MFIVs bmit the quanbty of high energy fluid that ,

enters the contamment through 7e broken loop and limit cooldown ' The MFCV provides backup isolaban to lurut cooldown and high energy fluid ad$ tion.

Sechon 10.4.7.2.2 In the event of a secondary side pipe rupture inside the contamment, the 2.2.4 SGS. 8.b-main feedwater control valves provide a redundant isolabon to the.MFIVs to limit the quanhty of high energy fluid that enters the contamment livough !

the broken loop b

9Af 640.60-20 Revesson 1

% ,- _a

_---_rm__u_ -- _ _ . _ _ _ _m

NRC REQUEST FOR .iDDITIONAL INFORMATION Table 640.60-1 (Sheet 20 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Acculent Analysis SSAR Reference Design Fea re ,

Value ITAAC System ITA Section 10.4.7.3 For a man feedwater line break inside tre contaernent or a rnan steam 2.2.4 SGS 9.a kne break, the MFIVs and the rnan feedwater control valves aulu mik,asy close upon receipt of a feedwater isolabon signal.

Section 10.4.7.3 For a steam generator tuba rupture event, positive and iedundant isolabon 2.2.4 SGS 9. a is provided for the man feedwater (MFIV and MFCV) with isolabon signals generated by the protechon and safety morutonng system (PMS).

Sectx.m 10.4.8.2.2.7 Blowdown system isolabon is actuated on low steam generator water 2.2.4 SGS 9.a levels The isolabon of steam generator blowdown grovdes for a continued avalabihty of the steam generator as a heat sink for decay teat removal in conjunchon with operabon of the passive resdual heat removal system and the startup feedwater systent Section 10.4.8.3 The safety-related portions of the steam generator blowdown system 2.2.4 SGS 8. c located in the containment and auxiliary buddings ae designed to reman functional after a safe shutdown earthgsake.

Sechon 10.4.9 1.1 Double valve startup feedwater isolabon is provided by the startup 2.2.4 SGS 8. b feedwater control valve and the startup feedwater isolabon valve. Both 8. c valves close on e startup feedwater isolation signal, an appropnate 9. a eng:neered safeguards features signal, within the brne established within the Technical Specifications, Secton 16.1.

gg

==

640.60-21

. Revision 1

NRC REQUEST FOR ADDITIONAL INFORMATION -

Tuble 640.60-1 (Sheet 21 of 24)

ITAAC Referencea for SSAR Table 14.3 Design Basis Accedent Analysis SSAR Reference Design Feature Value ITAAC. System ITA Secbon 10.4.9,1.1 For a stearn generator tube rupture event poseve and redundant isolaban 2.2.4 SGS 9.a is govded for the startup feedwater system (starkp feedwater isolabon signal and startup feedwater contra valve), w'th isofabon signals generated by the potecbon and safety mor.stonng system.

Secton 10.4.9.2.2 in the event of a steam generator tube rupture, the startup feedwater 2.2.4 SGS 8. b isolation valve and startup feedwater control valve Imt overfJ of the steam generator by terminabng startup med Bow.

Sechon 10.4.9.2.2 in the event of a secondary pipe rupture inside containment, the startup 2.2.4 SGS 8. b feedwater isolation valve and startup feedwater control valve provde isolation to limit the quanhty of high energy fluid th3t enters the containment.

Sechon 10.4.9.2.2 The startup feedna'y isolabon valve is provided to gevent the uncontrolled 2.2.4 SGS 9.a blowdown from more than one steam generator in the event of startup feedwater line rupture. The startup feedwater isolabon valve provdes backup isolabon.

Table 15. 0- 1 !n$al core thermal power (MWt) 1933 !NTRO NSSS -

Table 15. 0- 3 Nominal values of perbnent plant parameters used in acckht analysis with 2.1.2 RCS 9.a 10% steam generator tube pluggng

- Reactor coolant flow per loop (gpm) 9.48 E+04 W

Westinghouse 840:60-22.

Revision 1

NRC REQUEOT FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 22 of 24) -

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Design Feature Value ITAAC System ITA Secton 15.1.2.1 Continuous addiban of excessive feedwater is prevented by the steam 2.2.4 SGS 8.b generator high-2 water level sgnal trip, which doses the feedwater isda$on 12.a valves and feedwater control valves axi trips the turtune, main feedwater 2.2.5 PMS 6.b pumps and reactw.

Sects 15.1.4.1 For an inadvertent opening of a steam generator rehef of safety valve, core 2.2.3 PXS 11.b makeup tank actuabon oaurs on a safeguards CS) signal from one of four 2. 5. 2 PMS 6.b sources

- Two ;t of four low pressurizer pressure sgnals

-Two out of four low pressurizer lesal signals

-Two out of four low T , signals in any one loop

-Two out of four low steam line pressura signals in any one loop Section 15.1.4.1 After an inadvertent operung of a steam generator relief of safety valve, 2.2.4 SGS 8.b redundant isotabon of the main feedwater lines doses the feedwater control 12.a valves and feedwater isolation valves, and inps the main feedwater pumps. 2.5.2 PMS 6. b Section 15.1.5.1 Following a steam kne rupture, core makeup tank actuation occurs on a 2.2.3 PXS 11.b safeguards CS) signal from one of five sources: 2.5.2 PMS 6.b

- Two out of four low pressunzer pressure signals

-Two out of four high-1 conta#nment pressure signals

-Two out of four low steam kne pressere signal in an, loop

-Two out of four low T , signals in ar.y one loop

-Two out c4 four low pressurizer level signals

~

W W8Stingfl0t'Se Rev:ision 1

==

k

. NRC REQUEDT FOR ADDITIONAL _20 FORMATION Table 640.60-1 (Sheet 23 of 24)

ITAAC References for SSAR Table 14.3 Design Basis Accident Analysis SSAR Reference Design Feature Value ITAAC System ITA' Secton 15.1.5.1 Aher a steam kne rupture, redundant isolabon of the main feedwater lines 2.2.4 SGS 9.a closes the feedwater control valves and feedwater isolaban valves, and trips the main feedwater purnps

[f' Secbon 15.1.5.2.1 Core makeup tanks and the accumulan;. are the portons of the passne 2.2.3 PXS t core coohng system used in rningabng a steam Ene rupture.

Sechon 15.1.6.1 The heat sok for the PRHR heat exdenger is provided by the IRWST, in 2.2.3 PXS 1.

which the PRHR heat exchanger is submengd Sechon 15.2.6.2.1 FORowing a kx,3 of ac power, the PRHR heat exchanger is actuated by the 2.2.3 PXS 11;b low steam generator water level (wide range). 2. 5. 2 PMS 6.b Secton 15.2.8.2.1 Receipt of a low steam line pressure sig.tal in at least one steam kne 2.2.4 SGS 8.b initiates a steam Ene isolaton signal that closes an main steam be and 12.a feed line isolation valves. This signal also gives an "S" signal int initiates 2.5.2 PMS. 6.b Scw of cold borated water from the core makeup tanks to the reactcr coolant system.

Sechon 15.J.3.2.2 The pressunzer safety valves are fuay open at 2575 psia. Their capacity 2.1.2 RCS' 8a for steam relief is desenbed in Sechon 5.4.

Sechon 15 4 6.2.2 A safety signal from the protechon and safety monitoring system 2.3.2 CVS- 8.b automaticaHy isolates the potenbaGy unborated water from the .

11,b:

l demineralized water transfer and storage system and thereby terminates - 2.5.2 PMS 36. b . r

' the dilution Wg 640.60-24:

Revis6on 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-1 (Sheet 24 of 24)

ITAAC References for SSAR Table 14.3 Desegn Basis Accident Analysis SSAH Reference Design Feature Value ITAAC ' System ITA Seebon 15.5.1.1 Folkmng inadvertent operaten of the mre makeup tanks dumg' power 2.2.3 PXS 11.bl operation, the high-3 pressurizer level signal actuales the PRHR heat 2.5.2 PMS . 6. b exchangar and blocks the pressurizer heaters.

Secton 15.5.2.1 Tne pressunzer heaters are blocked, and the main feedwater toes, steam 2. 5. 2 PMS 6. b lines, and cherncal andvolume control system are isolated 9.a Sechon 18.8.3.2 The main control area indudes two reactor operator workstabons, the o.2 -HFE- 6 supervisor's workstabon, the dedcated safety panel and the wall panel informabon system.

Section 18.8.3.2 The human system interface resources avadable at each workstaban are 3. 2 HFE 8 the plant information system displays, the control drsplays (soft controis),

the alarm system support displays, procedure system, and the screen and component selector.

Re l

NRC REQUEDT FOR ADOITIONAL INFORMATION Table 644.60 (Sheet 1 of 2)

ITAAC References for SSAR Table 14.3 Anticapated Transient Without Scram

4. SSAR Reference Desegn Feature Value ITAAC System ITA Sec5cn 7. 7. 1.11 Tne dverse actuanon system is a nonsafety-related system that provdes a 2. 5.1 DAS 1 dveise backup to the protecbon and safety morutonng system.

Sechon 7. 7. 1.11 The dverse actuabon system inps the reactor control rods and the antnne 2. 5.1 DAS 2.a on low vnde range stearn generator water level and on los pressurizer water level.

Sechon 7. 7. 1.11 The dverse actuaton system irnbates passive resdual heat removal on low 2. 5.1 DAS 2.b wide range steam generator v ater level or high hot leg temperature, actuales core makeup tanks and trips the reactor malant pumps on low pressurizer water level, and isolates selected contanment penetrabons and s: arts passive contanment coohng Secbon 7. 7. 1.11 The manual actuabon functon of the dverse actuabon system is 2. 5.1 -DAS 2. c unplemented by wiring the controls located in the main control room directly 2. d to the foi.i loads in a way that bypasses the norme path through the control room rnuftiplexers, the engmeered safety features actuabon cabnets, and the diverse actuation system logic Section 7. 7. 1.11 The 6 verse actuabon system uses a microprocessor board dfferent from 2.5.1 DAS 3. a those used in the protechon and safety morntonng system.

Sechon 7. 7. 1.11 The diverse actuabon system hardware unplementaban is diferent from that 2. 5.1 DAS 3. b o' tM protechan and safety monitoring system. <

t 1

1 l yf gp 640.so.2s -

'" Revision 1

k NRC REQUEST FOR ADDITIONAL INFORMATION ' a 9

Table 640.60-2 (Sheet 2 of 2)

ITAAC References for SSAR Table 14.3 Anticipated Transient Without Scram SSAR Reference Design Feature Value ITAAC System ITA Section 7. 7.'1.11 The operabng system and pwismag languays of the dverse actuabon 2. 5.1 DAS 3. c system is dfferent from that of the protechon and safety monstonng syttem.

640.60-27 RStingtIOUS8 Revision 1 ;

NRC REQUEST FOR ACDITIONAL INFORMATION Table 640.60-3 (Sheet 1 of 41 -

i. ITAAC References for SSAR Tame 14.3 Fire Protection W2 Reference Design Feature Value ITAAC Systone ITA Sechon 3.4.1.1.2 The boundanes behseen rm1.& c4 equpment rooms aid me electrcal 3.3 N1 4. b and instnanentaban and enntrol equpmerd rooms d the arminary bunchng are desgned b prevent cong of rooms hat c!ntan sale shut $own equpment up to the rnannum flood level for esda roont Seebon 3.4.1.1.2 Separahan is manta.wt between Cass 1E dneans and between Qass 3.3 NI 6. c 1E dnoons and non-Cass 1E cables in accordance wei me fire areas.

Su: bon 3.4.1.1.2 The AP600. ..yowoe prowdes physcal separabon d redundant safety- 3. ? NI 6.c related components and systems from each ome and from nonsafety-related w .w A ;

3.4.1.2.2 Sechon The boundares between mechanca' equipment rooms insde contanment 3. 3 NI 4b and the elecincal ar.d instrumentaban and control equpment rooms d be i aumliary buildng are designed to prevent floodng d rooms that contan sale shutdown equipment up to the maximum flood level for eadi roont i

Sechon 3.4.1.2.2 Boundanes erst to prevent floodng between the foRowng rooms whch 3.3 NI 4.b catan safety 4 elated equpment PXS vafve! accumulator room A, PXS valve /accumulattr room B, and chemcal and volume control room W wesunpouse **o.60 2s

= Revenion 1

NRC REQUEST FOR ADDITIONA' INFORMATION Table 640.60-3 l (Sheet 2 of 4)

ITAAC References for SSAR Table 14.34 - Fwe Protection Desegn Feature Value ITAAC System ITA SSAR Reference Sechan 3.8.4.1.1 The conical roof supports the passrae contanment cochng system tank. 3.3 NI 9 which is constructed wrth a star %ss steel liner on ic.&M concrete waEs.

Secbon 7.1.4.1.6 The atAty of the rviccu. and safety susus system to r6f.e and 2. 5 2 PMS 3 ouuividi rwiedie $mcbons is mantaned desprie degradeo ci4;u6 caused by internal events such as fire and Soodng

3. 2 HFE 12 Secbon 7.4.3.1 If temporary evacuabon of the man control room is regured because of some abnormal man control room condbon. the operators can estabhsh and mantan safe shutdown con 6 bons for the plant from outside the man control room through the use of controls and nuang located at the remote shutdown workstaban.

?

Sechon 7.4.3 1.1 The remote shutdown workstaten equipment is smlar to the operator 3.2 HFE 9 workstabans in the man control room and is desagned to the sarte 11 standard One remote shutdown workstaban is prtmded Secbon 7.4.3.1.3 The remote shutdown workstaban adwhw and mantans safe shutdown 32 HFD 12 con 6 tons from fut power uidius and matans safe shutdown aidiu6 thereafter.

640.60-29 W- g. Revision 1

NRC REQUEST FOR ADDITIONAL. INFORMATION Table 640.60-3 (Sheet 3 of 4)

ITAAC RMerences for SSAR Table 14.3 Fire Protection SSAR Reference Design Feature Value ITAAC System ITA Secton 8.3.2.2 The four Oass 1E battery chargers Qass 1E vdtage regulahng 3.3 N1 6.b trao u.ses are independent. located in separate rooms, cannot be 6c intertonnected, and their cretats are routed in de6cated, physca5y separated raceways.

Secbon 8.3.2.3 Each safety 4 elated orcud and raceway is gnren a umque identficaban 3.3 N1 6. a number e disbnguish between crcuits and raceways of ddferent voltage level or separatm groups Secbon 8.3.2.4.2 Cabies of one separabon grc4 are run in separate raceway ara physcally 3. 3 NI 6. d separated from cables of other sep.du. groups. Group N raceways are separated from safety 4 elated groups A. B. C, and D. Noncass 1E orcuits are elecincally isolated by isolabon devi s, shieking and unng techruques, physcal separabon, or an appropnate arnbnabon thereof.

Secbon 9.5 1.2.1.1 Separabon is maintaned between % 1E dnnssons and between Cass 3. 3 Nt 6. c 1E dvisens and noncass 1E cables in accordance mth fre areas Sechan 9.5.1.2.1.5 The standpipe system is supphed mth water from the safety 4etated passive 2.3.4 FPS 1 contanment cochng system storage tank and normaty operates andependency of the rest of the fee prote: bon systerrt The supply. tine draws water from a dedwied porton of the storage tank. using water allocated for fre yviadut W g-gg 640.60-30 devassort 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-3 (Sheet 4 of 4)

ITAAC References for SSAR Table 14.3 F'we Protection SSAR Reference Desegn Feature Va>Je ITAAC System ITA Secton 9.5.1.2.1.5 The :4 4g system sennng areas ud--r; equprnent requred for safe 2.3.4 FPS 2 shutdown foBowog a safe shutbn earthquake is desgned and supported so that it can wtstand the efects of a safe shutdown earthquake and remam funcbonal Secbon 9.5.1.2 1.5 The volume of the water in the PCS tank is suf5 cent to supply two hose 2 18.000 2.3.4 FPS 4. i streams, each we a flow of 75 ganons per trunute, for two hours (gal)

Sechon 18.8. 1 2 The human system interface resources ava:lable at eads workstaban are 3.2 HFE 8 the olant informaban system chspiays, the control dW (soft controls),

the alarm system support displays, proceoure system, and the sceen and ui@st selector Sectoi 18.8.3.4 The trusson of the remote shutdown workstabon is to provde the resour s 3.2 HFE 9 to bnng the pimt to a safe simidown cond; bon after an evacuaban of the 11 main control room. 12 Sechon 18 12. 3 The controis, deplays, and alarms Ested in Table 18 12. 2- 1 are retnewable 2. 5. 4 DOS 2 from the remote shutdown workstaban gg 640.60-31

- Revision i

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-4 (Sheet 1 of 3)

ITAAC References f at SSAR Table 14.3 Flood Protectxm SSAR Reference Design Feature Value ITAAC System ITA Secton 1. 1.43 The lowest level of the awaiery %3 ding. elevaban 66~ 6*, contans the 33 N1 5

/ppendix 1-A -psits of the radwaste system w@n a common flood zone we watertght !bors and waEs, Thss volume of this enM flood zone is suffcent to contan the contents of the radwaste systent Table 2- 1 Plant elevaton for maxrnum !!ood level (ft) s100 5 1/F Secdon 3.4.1.1.1 The seismc category I sinxtures tHow grade are protected aganst 3.3 NI 4. a finoding by waterstops and a wateiruving system.

Sechar 3.4.1.1.2 The boundanes between mechancal eqmpment rooms and the electncal 33 N1 4b and instrumentaton and control equipment rooms of the aux &ary buddog are desgned to prevent flooding of rooms that contan safe shutdown equipment up to the maxrnum flood level for each room.

9ff g 640.60-32 Revision 1

=;

NRC REQUEST FOR ADDITIONAL INFORASATION . , . - ,

n ~ . . . .

Tatde 640.60-4 (Sheet 2 of 3)

ITAAC References for SSAR Tatde 14.3 Flood Protecton SSAR Reference Doesgn Feature Value ITAAC System ITA Sechan 3.4.1.2.2 The boundanes between mechancal equpment rooms rede amtanment 3.3 !a 4b ar.d me eledncal and instumentaton and contal equgment rooms of he auxikary buileng are desgned b prevent Soodng ci rooms that contan safe shutdown equgment up b the mammum Sood level for each room.

Secton 3.4 t.2.2 Boundanes exist b prevent Soodng between the lolowng rooms which 3.3 NI 4. b antan safety-related eqtW PXS valve / accumulator room A PXS _.

valwlaccumulator room B. and chemcal and volume antrol roont Sechan 3.4.1.2.2 The AP600 arra.vec=4 prowdes physical + A, of redundant safety- 3.3 N1 4. b

, related wiyurets and systems from each omer and from nonsafety-related m. wets.

Secbon 3 4.1.2.2 The safety-related wiyu.ada available for safety hedown are looped in 13 N1 4.b 3

the awahary buildng and inside contanment. No credit is taken for operaton or sump pumps e megate me consequences or nooding I

640.60-33 j

@ Revieson 1 c

NRC REQUEST FOR ADDITIONAt. INFORMATION Table 640.60-4 (Sheet 3 of 3)

ITAAC References for SSAR Table 14.3 Flood Protection SSAR Reference Design Feature Value ITAAC System ITA Sechon 3.4.1.2.2.1 The PXS-A wiwL:oa. PXS-B wiwkien and the cherncal and 3.3 N! 4. c volume comrol system wip.L.od are physcaBy separated and isolated frorn each other by structural waEs such that floodog n any one d these wiwiseits is m the reactor coolant system wip.Lioa canrot cause floodng in any of the other wnpsLich Secton 3. 6 In the event of a hgh- or moderateenergy ppe fadure zh the plant. 3.3 N1 ,

adequate pdmJa . is prowded so that essental strudures, systems, or wivysits are tot vrpacted by the adverse effects of %W ppe falure.

Sechon 7.1.4.1.6 The ab6ty of the protedal and safety uvuiving systen to inchate and

2. 5. 2 PMS 3 accompr as h protectve funcbons is mantaned despie degraded 0 -4k s caused by intemal events such as fire and flooding 9ff g- 640.60-34

= Revision 1

NRC REQUEST FOR ADDITIONAL miFORMATION Table 640.60-5 (Sheet 1 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Table 2. 3 - 3 The u,;p.ests identfied under Reactor Systems in Table 3. 2- 3. as 2.1.3 RXS 2 ASME Code Secbon III are desgned and constracted in mu,J.u: with ASME Code Secton 111 Regurements.

Secton 3.2.1.3 The Nudear Island stPJctures indude the contanment and the sheid and 3.3 NI 1. a auxAary buddings These structures are seeme Category I.

Table 3. 2- 3 The Nuclear Island structure = include the containment and the Shield and 3.3 NI 1.a Auxmary Buindogs. These structures are seeme Category 1.

Sedon 3.4.1.1.2 The boundanes between mechancal equrpment rooms and the elecincal 3.3 Ni 4b and instrumentahca and control equpnent rooms of the auxhay building are desrped to prevent flooding of rooms that contan safe shutdown equipment up to the maximum Sood level for each roont Secbon 3.4.1.1.2 The AP600 arraiyen=d pnmdes physical separaban of redundant safety- 3.3 N1 4b related uaiwsits and systems from each other and from nonsafety-related u,.ir, wits.

gg 640.60-35

- Revision 1

NRC REQUEST FOR ADDITIONAL INFORh1AT10N Teble 640.60-5 (Sheet 2 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Sechon 3.4.1.1.2 Separation is mantaned be+ ween Class 1E dvsons and between Qass 33 N1 6. c 1E dvsons and non-Cass 1E cables in accordance wrlh the fire areas.

Sechon 3 4.1.1.2 The AP600 i.npiod prtmdes physcal separaton of redundant safety- 3.3 N1 6. c related u,.wmits and systems from eadi other and from nonsafety-related components.

l Section 3.4.1.2.2 Boundaries exist to prevent 11oodng between the foRowing 3. 3 M 4. b l rooms which contain safety-related equqpment: PXS l valve / accumulator room A. PXS valve / accumulator room B, l and chemical and volume controlroom.

l Section 3.4.1.2.2 The boundaries between mechanical equrpment rooms inside 3. 3 NT 4. b l containment and the electrical andinstrumentation and l control equipment rooms of the auxEary buidmg are l designed to prevent floodmg of rooms that contain safe l shutdown equipment up to the maximurn floodlevelfor each l room.

l Section 3.4.1.2.2 The safety-related components avaRable for safety shutdown 3. 3 M 4. b l are locatedin the auxEary buBding andinside containment.

l No credt is taken for operation of sump pumps to mitigate l the consequences of floodng.

W g-gg 640.60-36 Renseon 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 3 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk f asessment SSAR Reference Design Feature Value ITAAC System ITA I

l Section 3.4.1.2.2. The PXS-A compartment, PXS-B corrpartment, and the 3. 3 NT 4. c l 1 chemicaland volume controlsystem compartment are l ' physicaWy separated andisolated from each other by l structural waEs such that Annning in any one of these l compartments or its the reactor coolant system c&wrin i, ,,n l cannot cause flooding in any of the other compartments.

l l Section 3D.6 RXS equement in Appendx 3D is sersmeceWy quag 6ed. 2. 1. 3 RXS 5 l

l Section S. 1. 3 ADS has four stages. Each stage is arrangedinto two 2. 1. 2 RCS 1 l separate groups of valves and nines.

l -Stages 1, 2, and 3 dscharge from the top of the pressurizer l to the IRWST.

l -Esch stage 4 dscharges from a hc! Jeg to the RCSloop l compartment.

l l Section 5. 3. 1. 1 The reactor vesselprovides a high integritypressure 2.1.3 RXS 4 l boundary to contain the reactor coolant, heat generating l reactor sre, and fuel fissian products. The reactor vesselis l the primary boundary for the reactor coolant and the l secondary barrier against the release of radoactive Hssion l products.

gg

==

640.60-37 Reviason 1

NRC REQUEST FOR ADDITIONAI. INFORMATION Table 640.60-5 ISheet 4 of 18) -

ITAAC References for SSAR Table 14.3 Probabilistic Risk A.wssment SSAR Reference Design Feature Valve ITAAC System ITA 1

l Section 5. 4. 6 ADS has four stages. Each stageis arrangedinto two 2. 1. 2 RCS 1 l separate groups of valves and Enes.

l -Stages 1. 2. and 3 dscharge from the top of the pressunzer l to the IRWST.

l -Each stage 4 dscharges from a hot leg to the RCS toop l compartment.

l l Section 5. 4. 6. 2 Each ADS stage 1. 2. and 3 Ene contains two normady 2. 1. 1 RCS 1 l closed motor-operated valves (MOVs).

I l Section 5. 4. 6. 2 Each ADS stage 4 Gne contains a normaRy open MOV valve 2. 1. 1 RCS 1 l and a normady closed squib valve.

Se: bon 5.4.7 The RNS removes heat from the core and reactor oaotant system at 2.3 6 RNS 1 reduced RCS pressure and temperature condtons after shutdown Secton 5.4.7 The normal resdual heat removal system (RNS) provdes a safetym 2.3.6 RNS 8. a mean; of perfomwg the Iceowing funcbons

- Containment isolaton for the RNS Enes that penetrate the contanment

- Long-term, post-acadent makeup water to the RCS Wg 640.60-38

- Revision i

. NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 5 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Sechon 5.4.7.1.1 The RNS wdameit isolaban and pressure boundary valves are safety- 2.3.6 RNS 7.b related. The motorvated valws are powered by Cass 1E de power.

Sechon 5.4.7.1.2.1 The component cochng water system (CCS) provdes cooiing to the RNS 2.3.1 CCS 3 heat exdianger. 2.3.6 RNS 1 Sechon 6.2.4 The containment hydrogen control system prowdes nonsafety-related 2.3.9 VLS S hydrogen igraters for control of the contaoment hydrogen concentraban for beyond desxp basis acodents.

Sechan 6.2.4.2.3 .u leas' 64 hydrogen ign:ters are provded 2.3.9 VLS 3 Table 6. 2. 4- 2 The muumum passive autocatalybe recomtaner depiebon rate at 120* F and 21 2.3.9 HCS 4. b. ii atrmspheiic pressure (scfm)

Sechon 6. 3 The automate depressunzaton system provdes a safety-re!ated means of 2.1.2 RCS 8. c depressunzing the RCS.

SN MW

"" 8 - 8 Revemon 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 6 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Secton 6.3 The exontamment refuehng water stora9e tartk subsystem provdes a 2.2.3 PXS 8. c safety-related rneans of pe/,un c3 the followog funchons.

- Low-pressure safety irgecton

- Core decay heat sink dunng desgn basis events

- Flooding of the lower containment, the reactor cavdy and the loop wiv unent by drarung the IRWST sito the contanment

- Barated water Sechon 6.3.1 The core makeup tanks prowde safety-related means of safety irgecnon of 2.2.3 FXS 8c barated water to the RCS.

Secton 6.3.1 Passive resdual heat removal (PRHR) provdes a safety-related means of 2.2.3 PXS 8.c removrig core decay heat dunng desgn basis events Secbon 6.3.2 The ADS valves are powered from Oass 1E de power. 2.1.2 RCS 7.b W WeStingh0US0 0.60-40 Revisson 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 7 of 181 ITAAC Refere .ces for SSAR Table 14.3 Probabehstic Risk Assessment SSAR Reference Design Feature value ITAAC System ITA Seccon 6.3.2 There are tvo CMTs each wth an irgectan Ine to the reacer vesseuDVI 113 PXS 1 nozzie.

- Eact. CMT has a pressure balanw ine from an RCS cold leg

- Each injecse: ine is isolated mth a paraBei set of arcerated valves (AOVs)

- Thesa AOVs open on loss of ar.

- The irgecbon kne for each CMT also has two check vanes in sera Seebon 6.3,2 The IRWST subsystem has the fotbwng Rowpaths 123 PXS 1

- Two (redundant) irgecbon Enes from the IRWST to the reactor vesseLDVI nozzle Each kne is isolated w:th a paraRei set of vanes; each set wth a check vane in senes wth a squb vahe.

- Two (redundant) recrculaban hnes from the undo.osd to the IRWST injechon kne. Each rearculatnn kne has two paths one path contans a squb valve anti an MOV, the eter path contans a squb valve and a check valve.

- The two MOV/squb valve Enes also provide the capabkty 10 flood the reactor cavity.

p- 640.GO-41 Revesson 1

NRC IEQUEST FOR ADDITBONA1. INFORMATION Tatde 640.64 's (Sheet 8 of 18)

ITAAC References for SS4R Tatde 14.3 N. Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Sechen 6.3.2 There are screens for ead IRWST anjecnon kne and rearculaban Ire 1.2.3 PXS 1 Sechon 6.3.2 PRHR is actuated by operung redundant, paras argerated vahes. 2.2.3 PXS 12 These argerated vahes open on toss of ar.

Secton 6.3.2.2 The passive core cooing system (PXS) is composed of the followruJ 2.2.3 PXS 1

- Aa:unrAator subsystem

- Core makeup tank (CMT) subsystem -

I

- Inwri inent refuehng water storage tank (IRWST) subsystem -

- Passive resdal heat removal (PRHR) subsystem.

- The automabc depressunzabon system (ADS), wNch is a subsystem of the reador coolant system (RCS), also supports passne core coohng funcbons.

Sechon 6.3.2.2.2 There are No accumulators, each wth an insecton hne to the reactor 2.2.3 PXS 1 vesseLWect vessel ingechon (DVI) nozzle. Eam irvechan bne has too check valves in seres. '

Secbon 6.3 2.2.2 The accunuMors prowde a safety 4etaled means of safety insechan of 2.2.3 PXS 8,c barated water to the RCS.

i W-640.50-42 Rowseon 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 9 of 181 -

ITAAC References for SSAR Table 14.3 ProbabsTestic Risk Assessenent SSAR Reference Design Feature Value ITAAC System ITA Sechon 6 3.3 IRWST sque vahres and MOVs are poweed by Cass 1E de power. I13 PXS 7. b Sechon 6.3 3 The CMT AOVs are automatcaEy and manuaRy actua:ed from PMS and 2.51 DAS 2 UAS. 152 PMS 6 Cechon 6 3.3 The PRHR ar-operated valves are auhsw.4y actuated and rnanuaffy 2.51 DAS 2 actaated from the control room by ecer PMS or DAS. 2.5 2 PMS 6 Secton 6.3.3 The squb vahres and MOVs for ryectan and recrculaban are automa0caDy 15.1 DAS 2. c and manuaRy actuated ma PMS. and manuary aduated via DAS 2. 5. 2 PMS A Sechan 6.3.3 The 5 pb valves and MOVs for lower contanrnent and reactor cavity 2.51 DAS 2. c floodmg are manuany actuated ma PMS and DAS from tha control room 15.2 PMS 6. c Secbon 6.3 7 The posums of the containment recrculaban csotaban MOVs ::re recated 22.3 PXS 8a m the control room.

Secbon 6. 3 7 The posfoon of the niet PRHR valve is recated in the control room. 12.3 PXS 8. a Secbon 6.3.7.6.1 The ADS frst , second . and thad -stage valve positxos are in$cated n 2.23 PXS 8a the antrol room 640 60-43 Revision 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sheet 10 of 181 ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Desegn Feature Value ITAAC System ITA Sechon 7.1.1 The dese actuaton system provdes a nonsafety-related means of 2.51 DAS 2b Weg.g the foSowng funct:ans- 1d

- Irdates automabc and rnanual reactor tnp

- Automabc and manual actuabon of selected engmeered safety features

- Man arttrol room dsplay of selected plant parameters Secton 7.1.1 The pv;edu. and safety usab.9 system powdes a safety-related 15.2 PMS 6.b means of gun-g the followng furamons:

- Automabc and manual reador inp

- Automate and manual actuaban of ergneered safety features (ESF).

Sechon 7.1.1 PMS provdes for the momum inventory of fixed postban controls and 15? PMS 8a dsplays in the control room.

Secton 7.1.2 Each PMS dvson is pows 1 from its respecbve Oass 1E de dvision. 15.2 PMS 5.a Sechon 7.1.2 PMS has four cnnsons of reactor tnp and ESF actuaban 15.2 PMS 9b Secton 7.1.2 10 PMS automatcally produces a safety-related reacts t,ip or ESF rubaban 25.2 PMS 6.b upon an attempt to bypass more than two channels of a funchon that uses 8.b 2o.it-of-4 logic.

Wg 640.60-44 Rewsson 1

NRC REQUEST FOR ADDITIONAt. INFORMATION Table 640.60-5 (Sheet 11 of 18)

ITAAC References for SSAR Table 14.3 Probabelistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Secton 7.1.2.15 Tne PMS hardware and software are developed usng a planned desgn 2. 5. 2 PMS 11 process whct providas for speche desgn documentaban and revews dunng the desgn requirement, system de6nibon, development test and inMaban phases Sechon 7.1.2.6 PMS has redundant dmsons of safety-reta:ed post-acodent parameter 25.2 PMS 8. a 6 splay.

Sechon 7.1.4.1.6 The abihty of the protcGon and safety rnonstonng system to inabate and 2. 5. 2 PMS 3 accomphsh protechve funcbons is mantaned despite degraded condibons causcd by intemal events such as 6re and flooding Sechon 7.1.4 1.6 The desgn of the protecton and safety monitonng system equpnent has 2. 5. 2 PMS 4 l margri to is.wi.idted a loss of the formal HVAC.

Sechon 7.1.4.2.6 The fexblity of the protechon and safety monstonng system enables 2. 5. 2 PMS 5.b physcal separabon of redundant dvisions.

Figure 7.1- 8 PMS has redundant dmsons of safety-related post-acadent parameter 2. 5. 2 PMS 1 display.

== Reviseon 1 i

l l

NRC REQUEST FOR f.DOfTIONAL INFORMATION Table 640.60-5 (Shest 12 of 18)

ITAAC References for SSAR Table 14.3 Proh=hn=tsc Bish Assessment SSAR Reference Desegn Feature Value ITAAC System ITA Sechon 7,2.1 2.1 The protechan and safety Huh.9 system reales a reactor tnp 2. 5. 2 PMS 6.a whenever a condsbon rnandored by the syste7 reaches a preset level.

Secton 7.3 The PMS allows for the transfer of control capability from the man amtros 2. 5. 2 PMS 8.b room to the remote shutdown room. The rmnenum imrentory of drsplays and controis in the remote shutdown room is prowded.

Secton 7.3.1 The ADS valves are powered from Cass 1E de power. 2.1.2 RCS 7.b Sechon 7.3 1 The ADS valves are automabcaly and manuaBy actuated wa the p.viedui 2. 5.1 DAS 2. c and safety nu ib.9 system (PMS). and manuafy actuated via me diverse 2. 5. 2 PMS 6b actuabon system (DAS). ;

I Sechan 7.3.1 The CMT AOVs are aubnata.4y and mamally arinahx! from PMS and 2. 5.1 LAS 1c DAS. 2. 5. 2 PMS 6.b Sechon 7. 3.1 The sque valves and MOVs for injecton and rearculaeon are aub1My Z 5.1 DAS 2. c and manuaBy actuated ma PMS. and manually actuated via DAS. 152 PMS 6. b Sechon 7.3.1 The sque valves and MOVs for reactor cavity floo@g are manuaBy 2. 5.1 DAS 2. c actuated via PMS and DAS from the control room. 15.2 PMS 6. b ey 640.60-46 Revenson 1

NRC REQUEST FOR ADDITIONAL INFORMATION A

lable 640.60-5 (Cheet 13 o; 18)

ITAAC References for SSAR Table t' . PrdW15stic Risk Assessment SOAR Reference Desigr costure VJoo ITAAC System ITA Section 7.3.1 The PRHR ar-or,erM valves r- automatcacy emW mi ma%.aty 25,1 DAS 2. c aduated frcm the control room by W PMS or DAS. 2. 5. 2 PMS 6b Sechon 7. 3.1 1he RNS cutamment isolation MOVs are actuated via F4S. 15.2 PMS 6. b Sechon 7.6.1.I An intariock is provded for the normally closed rnotorgerated normal Z 5. 2 PMS 9c resdual heat removal system oner and outer sucbon esolabon va!ves.

Each ve8ve is interlocked so that it cannot be opened unless the reacity coolant system pressure is below a preset pressure Sechon 7. 7. 1.11 The drverse actuation system is a nonsafety-related system 54 prowdes a 2. 5.1 DAS 1 diverse backup to the protecton and safety mornin=9 system.

Sechon 7. 7. 1.11 The diverse ataban systein trps the reactor control rods and the %ne 2. 5.1 DAS 2.a on low wde range steam g=rooin water level and on bw pressur zer water le.t Secbon T . 1.11 DC manual matrn funcbons are rnplemented in a manner that 15.1 DAS 1c bypasses the siyual prh9 equp-mt of the DAS.

W WB5tillgil00Se

":*?.

Revissor.

NRC REQUEST FOR ADDITIONAL INFORASATION Table 640.60-5 ISheet 14 of 18) .

ITAAC References for SSAR Table 14.3 Probabiksenc Risk Assosoment SSAR Reference Design Feature Value ITAAC System ITA Secton 7.'7. 1.11 The DAS automate actuaban say.als are generated in a functonaBy 15.1 GAS 3.a dverse manner from the PMS segnals Drversdy between DAS and PMS 3. b is actueved by the use of deerent archdecture, dSerent hardware 3. c rnpiementabons, and dSerent solbware.

Secton 8.3.1.1.1 On loss of power b a 4160V desel-backed bus, the assooated desel 2.6.4 ZOS 2. a generator auinn.&,4y s: arts and produces ac power. The source crcud 1c breakers and bus load crcud tweakers are opened, ad me generator is conneded b the bus. Eadt generator has an automate load sequencer to enable controlled loadng on the assocated buses.

Secton 8.3 1.1.1 1 Two onsde standby desel @s.k urats prowde power b the selected 164 ZOS 2.b ir.nbaicip-related ac loads. P Secton 8.3 1.1.3 The man ac power systern estnbutes non-Oass 1E power from ensde 2.6.1 ECS 4. a '

sources b selected nonsa rety-rehded loads.

i Secton 8.3.2.1 The Cass 1E de and unintenuptde power supply (UPS) system (IDS) 2.6.3 OS 1 provides de and uninterruphble ac power for me salety-related equenent.

t b

j 640.60-4a

[ Mtstilighouse Reviesen 1

- NRC REQUEST FOR ADDITIOrdAL INFORMATION A

Table 640.60-5 (Sheet 15 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC. System ITA Sechan 8 3.2.1.1.1 There are four niependent. Cass 1E 125 Vdc dvsors Desons A and 2.6.3 IDS 1 D are each convosed of one ba:tery bank one setsboard, and one bacary charger Dvsons B and C are ead composed of two banery banks, two swrdh u6. and two banery chargers The frst banery bank in the four dvisions is desgnated as the 244mur banery bank. The second bacery bank in Dvsons B and C is desgnated as the 724 our bahery banit Secbon 8.3.2.1.1.1 Banery chargers are connected b & 5 746,isd buses. The snput ac 2.6 3 IDS 1 power for the Cass 1E de battery chargers is supphed from onsste dasel-generator-baded hvottage ac power supphes Sechon 8.3 2.1.1.1 The 24-hour battery banks provde power to the loads for a perod of 24 2.6 3 IDS 4. c l hours w:thout rechargas The 72-hour banery banks suppfyes a oc 4d smidha d bus load for a penod of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months without isday.4 Secbon 8.3.2.1.2 The non-Cass 1E de and UPS system (EDS) consists of the electnc 2.5.2 PMS 5.a power *py and dstrbution equipment that prowdes de and tanterruptble ac power to nonsafety-related loads.

Secbon 8 3.2.1.2 The non-Cass 1E de and UPS system (EDS) conststs of the electnc 2.6.2 EDS 1 rower supply and dstrbubon equpment that provdes de and uninterruphble ac power to nonsafety-related loads.

W

==

westinghouse ** :8 #8 Revis:on 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-5 (Sivret 16 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Dessgn Feature Value ITAAC System ITA Secson 8 3 1 1.2 EDS bad gcups 1, 2. and 3 provde 125 Vdc power b the associated 2.6.2 EDE 2 riverter units that supply the ac power b the rmClass 1E unotenupttle power suppiy sc system.

Secbon 8.3.I 1.2 Battery chargers are connected to de s-6.Lv -d buses. The input ac 2.6.3 IDS 1 power for the Class 1E de battery chargers is supphed from onsite desel-germob-backed bw-voltage ac power suppies Secbon 8.3.2.1.2 The onsite standby $esel<jeneratcr-ba-ked low-voltage ac power supply 16.3 IDS 1 provides the normal ac power b the battery chargers Secbon 8.3.2.1.3 Separabon is provded between Cass 1E divmons, and between Cass 1E 2.6 3 IDS 3 divoons and non-Class 1E cables Sechon 9. I 1 Tin se ce water system is a nonsafety4 elated system that transfers heat 2.3.8 SWS 3

' rom the wwod coolog water heat enW to the atmosphere Sedon 9.2.1.2.1 The SWS is arranged ind two trains Each tran rdudes one pump and 2.3.8 SWS 1 one cooling tower cet.

Sechan 9.2.2 ar cooling water system is a nonsafety4 elated system that The wiv ent 2.3 1 CCS 3 removes heat from various wupu ents and transfers the heat to the sennte water system (SWS) g g.

640.60-50 Reviseon 1

NRC REQUEST FOR ADDl!!ONAL INFORMATION A

Table 640.60-5 (Sheet 17 of 18)

ITAAO References for SSAR Table 14.3 Probabslistic Risk Asswssment ITA SSAR Reference Design Feature Value ITAAC System Sec'xn 9.2.2.2 The CCS is arra94 into two trarts. Each tran rdudes one pump and 2.3.1 CLS 1 oce heat exhanger Secbon 9.3.6 TN CVS provdes a nonsafety-related means b pe brm the baowog 23.2 CVS 9 funcbons-

- Makeup water to the RCS dunng normal plant operaban

- Boraban foDowog a falure of reactor inp

- Coolant to the pressunzer auxiEary spray kne.

Sechon 9.3.6.1 The chemes! and volume control system (CVS) provdes a safety-related 2.3.2 C7S 8h means to temw. ate inadvertent RCS baron ddutKn.

Secbon 9.4.1 The man control room has ds own ventdabcn system and is pressunzed 2.2.5 VES 1 The ventdaban system for the remote shuidown room is independent of the ventdaban system for the man control room.

Sechon 9.5.1.2.1.1 The PMS allows for the transfer of control capabhty from the man control 2.5 2 PMS 8. b room to the remote shutdown workstaban. The trunimum inventory of dtsplays and controls in the remote shutdown room is prowded.

Secbon 9.5.1.2.1.1 Class 1E cables are routed in their respechve dartsional raceways. 33 Ni 6b g- . 640.60-51 Reidsion 1

NRC REQUEST FOR ADDITIONAL INFORMA70N ,

Table 640.60-5 (Sheet 18 of 18)

ITAAC References for SSAR Table 14.3 Probabilistic Risk Assessment SSAR Reference Design Feature Value ITAAC System ITA Sechon 9.5.1.2.1.1 Separabon is e-6=1 between Cass 1E dmsons and between 3. 3 NI 6c Oass 1E dmsons and ro>Cass 1E cabees ri accordance wdh the Ere weas.

Secton 16.2.1 important rehabary assumptons made as part of the AP600 pie'M-:- 3_7 DRAP nsk m.ad (PRA) wiB reman vahd thmughout plant kfe.

Secton 18.8.3.2 The mari control area indudes two rex *w operator w%stabans, the 3. 2 HFE 6 superv:sor's wcrkstabon, the dedcated safety panel and b.* was panel informabon system.

Sechon 18.12.2 The mrumun inventory of instrumentabon adudes those displays. 2. 5. 2 PMS 8a controls, and alarms that are used to morntor the status of the enbcal safety funcbons and to manuaDy actuate the safety 4 elated systems that acheve the enbcal safety funcbons The mrumum swentary resufbng from the implementaban of the selecton entena is provded in Table 18.12. 2- 1.

Wg 640.60-52 Reviseon 1

NRC REQUEST FOR ADDITION.AL INFORMATION Table 640.60-6 (Sheet 1 of 6) -

ITAAC References for SSAR Table 14.3 Radiologecal Analysis SSAR Reference Design Feature Value ITAAC System ITA Tabie 2- 1 Plant eleva*xn for max: mum Sood level (ft) s 100 C t/F Secto 2.3 4 AL,.ug,erc %w factors - X/O (sec/m') 5 I!F

- Sde Boundary X/O l (0-2 hour bme intervag s t o x 10'

- Low Populaban Zone Bou.dary 7/O O - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s 1.35 x 10' 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s 1.0 x 10' 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months s 5 4 x 10'

% - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months s 2.2 x 10' Table 6. 2. 3- 1 Contarment penetrabon isolaban features are configured as in 2.2.1 CNS 1 Tatde 6. 2. 3- 1 Table 6.23-1 Maxrnum closure bme for remotely operated contanment purge valves s5 2. 2.1 CNS 8i (seconds)

Table 6. 2. 3- 1 Maximum dosure brne fa a1 other remotely operated contamment isolaban s 50 2.2.1 CNS 8. E valves (seconds)

Sechan 6.4.2.3 The moimum storage capacity of each set of storage tanks n the VES s 122.021 2.2.5 VES 8. a. i (scQ ga- 640.GO-53 Revisson 1

NRC REQUEST FOR t.ODITICNAL IfeFORMATION Table 640.60-6 (Sheet 2 of 6)

ITAa0 References for SSAR Table 14.3 RadologicM Analysis SSAR Reference Design Feature Value ITAAC System ITA Secbon 6.4.3.2 The maxrrun temperature nse in the rnan control room pressure s 15 2_2.5 VES 8.c boundary follow:ng a loss on the nuclear estand iu.ihve vent!abon sysWn over a 724our penod (* F)

Secton 6.4.3.2 The maxrun ternperature in the instnuTudoiui and control rooms and s 125 2. 2. 5 VES 8. c dc equpment rooms following a loss of the nudear island nonradioachve venblaban system remacs over a 72-hour penod (*F).

l Sechon 6.4.4 The man ccntrol emergency habtatsfy system nommally provdes 2565 254-2 2. 2. 5 VES 8. a. i l scfm of ventiabon at to the man control room from the evessed at 65+5 l storage tanks d m tr 6 MW. ~ r^ W ! M t9: 20 l 6' =d~; (scfm).

l Sechon 6. 4. - Tey FeSixty-five scfm of ventiabon flow is su!5cient to pressonze 1:3* 2. 2. 5 VES 8.b the control room to 1/8* n:h water gauge dferental pressure (W1C).

l F9.ne 6.4-2 The man control room criopsy t&Ety system consists of twoa 2. 2. 5 VES 1 sets of cHapsy ar storage tanks and an ar de!ivery system to the l

man control room.

gg 640.60-54 Revision i

NRC REQUEST F_. ADDITIONAL INFORMATION Tatde 640.60-6 ISheet 3 of 6)

ITAAC References for SSAR Table 14.3 Radiologecal Analysis SSAR Reference Design Feature Value ITAAC System ITA Secbon 6. 5. 3 The passive heat rernoval process and the Imded leakage from the 2.2.1 CNS 8. i contamment result rt offsste doses less than the regulatory gudehne Imts.

Secton 8.3.1.1.6 FW~a' penetrabons through the mntamment can withstand the 2.2.1 CNS 9 maxrnum shortount currents 1va:!able either conhnuously without exceedmg ther lhermal Imt, or at least longer than tha field cables of the cucu:ts so that the fault or overload currents are interrupted by te protecbve devres pnor to a potenbal falure of a pewtra$on.

l Secbon 9.4.1.1.1 The VBS isolates the HVAC destwodpipeg that penetrates the man 17.1 VBS 4 control room boundary on hgh partcAate or iodine concer,trabons in the main control room supply air or on extended loss of ac power to support operabon of the mam control room errapKy habstabsty system.

l Sesten 18 8 3 2 Bemr Med = re:M e M mr~ p_t d dte., 4-2 HFE -6 l thnuperese( ddtm, N W'~J @ pc-d _W N d papei l % dm g!r W

Westinghouse 6"U:60-55 Rem.on i i

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-6 (Sheet 4 of 6)

ITAAC References for ESAR Table 14.3 Radiological Analysis SSAR Reference Design Feature Value . ITAAC System ITA Secbon 12 3 2.2.1 Dunng reactor operaton, the shieu budchng protects personnel ocnipyng 3.3 NI 2 arigarant plant structures and yard areas from raiaton ongmahng in the reactor vessel and pnmary toop umpicts. The concrete sheid buinna wag and the reactor vessel and steam generator umystist shield wars reduce rariation levels outside the shield budding to less than 0 25 mremlbr from sources inside contamment. The shield tuidmg completely surrounds the reactor um,r s =&

Sechon 12.3.2.2.2 The reactor vessel is shielded by the concrete pnmary sheid and by the 3.3 NI 2' concrete secondary shield which also surrounds other pnmary loop components. The secondary sheid is a structural module filled with concrete surrourxhng the reactor coolant system equipment, indudog ;

pipeg, pumps and steam generators. Extensne sheldog is provided for areas surroundeng the rebehng cav:ty and the fuel transfer canal to Imt the radsabon levels.

l g== 640.60-55.

Revesson 1 l

l l

t

NRC REQUEST FOR ADDITIONAL INFORMATION g. .

S ..

Table 640.60-6 (Sheet 5 of 6)

ITAAC References for SSAR Table 14.3 Radiolog6 cal Analyses Desegn Feature Value - ITA4C System ' ITA SSAR Reference Shieldmg is provded for :'wa liquid radwaste, gaseous radwaste and spent 33 Ni 2 Sechon 12.3.2.2.3 resin handkng systems consWent with the maxrnum padW acInnty Comdors are generai!y shieldeo to allow Zone 11 acx:ess, and operator areas far valve modules are generet, Zone 11 or hl for access Sheeldog is provded to attenuate radiabon from nomial resdual heat removal equipment durirg shutdown (nohng operations to levels consistent wih radiahon zonog requirements of ad}acent areas.

l lhe concrete shield wars surroundog the spent fuel cask loadog and 3. 3 NI 2 l Sechon 12.3.2 2.4 decontamination areas, and the shield walls surroundog the fuel transfer and storage are as-are4ufficiently thick to limit radebon levels outsde the l l shield walls in accessbie areas to Zone 11. The buildog waNs are sufficient to shield external plant areas which are not controlled to Zone 11.

Secbon 12.3.2.2.5 Sheldeg is provded as necessary for the waste storage areas in the 3.3 NI- 3 radwaste building to meet the radiation zone and ao:ess requirements.

. = = _

=_ .

NRC REQUEST FOR ADDITIONAL INFORMATION 1r d

Tatde 640.60-6 (Sheet 6 of 6)

ITAAC References for SSAR Table 14.3 Rodolopcal Analyses SSAR Reference Design Featt.re Value ITAAC System ' ITA '

Section 12.3.2.2.7 Shelding combmed eth other ergneered safety features is provxled to 3. 3 N1 3.

permit access and occupancy of the cxmtrol roorn fotomng a postulated lossda:oolant acodent, so that radiabon doses are Imted to Ne tem whole body from contnbuhng modes of exposure for the durabon of the acadent, in &wd w with Generat Design Cntena 19.

Sectxm 12.3.2.2.9 The spent fuel transfer tube is shielded to wthas adjacent area radebon 3. 3 NI - 2 l limits, is completely enclosed in concrete, and there is no unsheided porbon of the spent fuel transfer tube dunng the refuehng operabon.

b W Wesunghouse 640.60-sa Revesson 1 l

k

NRC REQUEJT FOR ADDITIONAL INFORMATION Table 640.60-7 (Sheet 1 of 2)

ITAAC References for SSAR Table 14.3 Severe Accsdent Analysis SSAR Reference Desegn Feature Value . ITAAC - System ITA l Ses6en 12.3.2.2.9 5 p'Fr8r d W k- W. te t 9 - t re:4adelen 3 M -3 l i--1, k ~ -,_'i e-e=~' b. =-- W r.n k - ; .2:;5ed l W. c'ihe m' Fm' ed W t- g N d6; pma-.

Sechon 5. 3.1. 2 There are no penetrabons in the reactor vessel below the core. 2.1.3 RXS 1 Sechon 6.2.4 2.1 _ The hydrogen concentrabon iru A .g subsystem cmsists of two grotps 2.3.'9 HCS 1 of eight hydrogen sensors eadt Sechon 6.2.4.2.2 The hydrogen recomt*, suosystem consist of two passive 2.3.9 HCS 1 autocatalybc recombiners irtstaNed inside the contamment above te operahng decit Secbon 6.2.4,2.3 The hydrogen igniton subsystem consists of 64 hydrogen igniters 2.3 9 HCS 1 strategically distnbuted throughout the contamment.

Table 6. 2. 4- 3 The minimum surface temperature of the hydrogen ignitors (*F). 2 1,700 2.3.9 HCS 5. s Sechon 63 The ADS provides a safety 4 elated means of depressuriang the RCS. 2.1. 2 'RCS 1 1

i W

==

640.60-59 Revesson 1

NRC REQUEST FOR ADDITIONAL INFORMATION Table 640.60-7 (Sheet 2 of 2)

ITAAC References for SSAR Table 14.3 Severe Accident Analysis

~

SSAR Reference Design Feature Value ITAAC . System ' ITA Section 6.3 The PXS provides a safety-related mes.ns of Soodug the reactor cavity by 2.2.3 PXS 1 draming the IRWST into the containment.

Section 7.3.1.2.9 Signals to align the IRWST containment reorculation isolabon valves are ' 2. S. 2 ; PMS 6. c generated by manual initiaton.

Sechon 7. 7. 1.11 initiation of containment recirculabon is a diverse manual funcbon 2. 5.1 DAS 2.c p== 640.60 Revesson 1

ATTACliMENT TO RAI 640.60, Revision 1 Changes to SSAR Tables 14.3-2,14.3 4,14.3 5,14.3-6, and 14.3 7 4

1

14. I:ltial Test Progran Table 14.3 2 (Sheet I c. z!) .

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 51.2 Safety valves are installed above and connected to the pressurizer to provide overpressure protection for the reactor coolant system.

Section 5.1. 2 The RCS has two hot legs and four cold legs.

Section 5.1. 2 The RCS has two steam generators anf four reactor coolant pumps.

Section 5.1. 2 The RCS contains a pressurizer and a surge line connected to one hot leg.

Section 5.1. 3. 3 Rotating inertia needed for flow coast-down, is provided.

Table S.1-3 Thermal design flow rate with 10% tube plugging 94,800 (gpm!!oop)

Table 5.1-3 initial rated react. core thermal power (MWt) 1933 Section 5. 2. 2 Reactor coolant system and steam system overpressure protection during power operation are provided by the pressurizer safety valves and the steam generator safety valves, in conjunction with the action of the PMS.

Section 5. 2. 2.1 Safety valve capacity exists to prevent exceeding 110 percent of system ciesign pressure for the following events:

-Loss of electrical load and/or turbine trip Uncontrolled rod withdrawal at power

-Loss of reactor coolant flow

-Loss of normal feedwater Loss of offsite power to the station auxiliaries Section 5. 2. 2.1 Overpressure protection for the steam system is provided by steam generator safety valves Draft for Rev.17 October 31.1997 14,3 1 g [ W85tl!1gt10USS

14. I:ltial Test Program Table 14.3 2 (Sheet 2 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 5.3.2.3- Non-destructive examination (NDE) of the reactor vessel and its appurtenances is conducted in accordance with ASME Code Section 111 requirements.

Section 5. 3. 2. 5 The initial Charpy V-notch minimum .spper shelf fracture energy levels for the reactor vessel beltline base metal traverse direction and welds are 75 foot pounds, as required by Appendix G of 10 CFR 50.

l E+edon 5. ? 2, 5 The ':!:! Chr:" " c :h rinN . uppr 'h:!' # ::::::

I .._:gy !:=!: re - 'h: :_=:^ =:! 5:!'"n: Ex: ::!

I - . :.: d::::::c xd :!6 r: 75 fue: peen 6, ::

I - ::q :::d by '.ppxdb O d '^ CFR 5^

Section 5.4.I.2.1 Resistance temperature detectors (RTDs) monitor motor cooling circuit water temperature. These detectors provide indication of anomalous bearing or motor operation. They also provide a system for automatic shutdown in the event of a prolonged loss of component cooling water.

Section 5. 4.1. 3. 4 It is important to reactor protection that the reactor coolant continues to flow for a time after reactor trip and loss of electrical power. To provide this flow, each reactor coolant pump has a high-inertia rotor.

Section 5.4.1.3.4 A safety related pump trip occurs on high bearing water temperature.

Section 5.4.5,2.3 Power to the pressurizer heaters is blocked when the core makeup tanks are actuated Draft for Rev.17 M 0gh0038 14,3 19 October 31,1997

\.

14. I:itial Test Program Table 14.3 2 (Sheet 3 of 21)

DESIGN HASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 5. 4. 6 Automatic depressurization system stage 1,2 and 3 valves are connected to the pressurizer and discharge via the spargers to the in-containment refueling water storage tank.

Section 5. 4. 6 Automatic depressurization system stage 4 valves are connected to each hot leg.

Section 5. 4. 9. 3 in the analysis of overpressure events, the pressurizer

- safety valves are assumed to actuate at 2500 psia. The safety valve flowrate assumed is based on full flow at 2575 psia, assuming 3 percent accumulation.

Section 5. 4. 9. 3 The pressurizer safety valves prevent reactor coolant sptem pressure from exceeding 110% of system design pressure.

Table 5.41 Minimum reactor coolant motor / pump mcment of inertia 2 5,000 2

(Ib-ft ).

Table 5. 4-11 Reactor Coolant System Design Pressure Settings:

- Safety valves begin to open (psig) 2485 Table 5.4-17 Pressurizer Safety Valves - Design Parameters:

- Number 2

- Minimum required relieving capacity per valve (Ibm /hr) 2400,000 1 - Set pressure (psig) 2485* 25 Section 6. 1. 2. 1. 3 The exterior of the containment vessel is coated with the same inorganic zine as is used inside of the containment.

Draft for Rev.17 October 31,1997 14.3 20 W85tillgt10Use

Z" P I4. I:ltlal Test Progr:m

~

't Table 14.3 2 (Sheet 4 of 21)-

DESIGN BASIS ACCIDENT ANALYSIS-SSAR Reference Design Feature Value Figure 6.2.21 The passive containment cooling system consists of a water storage tank, cooling water How discharge path to the containment shell, a water distribution system for the containment shell, and a cooling air now path.

Figure 6.2.21. The minimum duration the PCS cooling water How is 2 72 provided from the PCCWST (hours)

Table 6.2.21 The water coverage of the containment shell exceeds the

, amount used in the safety analysis.

Table 6.2.21 The minimum drain now rate capacity of the upper 2450 annulus drain (gpm).

Table 6, 2. 2-1 The minimum makeup Dow rate capability from an 2 62 external source to the PCS water storage tank (gpm).

Table 6.2,21 The minimum makeup flow rate capability from the PCS 2 50 water storage tank to the spent fuel pit (gpm).

Table 6.2.21 The minimum PCS water storage tank volume for 2 400.000 makeup to the spent fuel pit (non coincident with PCS operation) (gallons).

Table 6, 2. 2- 1 The minimum long term makeup capability from the 24 PCCAWST to the PCCWST (days)

Table 6. 2. 2- 1 The minimum long term makeup flow capability from the 2 62 PCCAWST to the PCCWST (gpm)

Table 6. 2. 2-2 The first or top standpipe's elevation above the lowest or 21.7 0.25 bottom standpipe (feet).

Draft for Rev.17 UM 14.3 21 October 31,1997

14. I:itial Test Prograu Table 14.3 2 (Sheet 5 of 21)

DESIGN HASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Table 6.2.22 The second standpipe's elevation above the lowest or 14. 2

0.25 bottom standpipe (feet).

Table 6.2.22 The third standpipe's elevation above the lowest or 6. 7

0.25 bottom standpipe (feet).

Figure 6. 2. 2-3 The minimum passive containment cooling water flow 2 71.5 rate with water inventory at a height cbove the lowest standpipe of 13.55

0.025 ft. (gpm)

Figure 6.2.23 The minimum passive containment cooling water flow 2 442 rate with water inventory et a height above the lowest standpipe of 23.75

0.25 ft. (gpm)

Figure 6.2.23 The minimum passive containment cooling water flow 2 122 rate with water inventory at a height above the lowest standpipe of 20.65

0.25 ft. (gpm)

Section 6. 3 The passive core cooling system provides emergency core decay heat removal during transients, accidents or whenever the normal heat removal paths are lost.

I Section 6. 3 The passive core cooling system provides makeup and boration during transients or accidents when the normal reactor coolant system makeup supply from the chemical and volume control system is unavailable or is insufficient.

Section 6.3.1.1 The passive core cooling system is designed to provide emergency core cooling during events involving increases and decreases in secondary side heat removal and decreases in reactor coolant system inventory.

Draft for Rev.17 October 31,1997 143 3, W Westiligt10US8

m. u=

14.' ' 1:itial Test Prograa . 5 Tabic 14.3-2 (Sheet 6 of 21)

DESIGN BASIS ACCIDENT ANAINSIS SSAR Reference . Design Feature Value Section - 6.3.2.1.I The heat ex: hanger consists of a bank of C-tubes, connected to a tubesheet and channel heat anangement at the top (inlet) and bottom (outlet). The passive exchanger ,

connects to the reactor coolant system through an inlet line from one reactor coolant system hot leg and an outlet line to the associated steam generator cold leg plenum (reactor coolant pump suction).

Section 6.3.2.1.I For the passive residual heat removal heat exchanger. the normal water temperature in the intet line will 6 hotter than the discharge line.

Section 6.3.2.1.2 The actuation of the core makeup tanks following a steam line break provides injection of borated water via water recirculation to mitigate the reactivity transient and provide the required shutdown margin.

Section 6.3.2.2.3 The in containment refueling water storage tank contaias one passive residual heat removal heat exchanger.

Section 6.3.2.2.6 Automatic depressurization system stage 1,2 and 3 valves are connected to the pressurizer and discharge via the spargers to the in containment refueling water storage tank.

Section 6.3.3.2.I For a loss of main feedwater event, the passive residual heat removal heat exchanger is actuated. If the core makeup tanks are not initially actuated, they actuate later when passive residual heat exchanger cooling sufficiently reduces pressurizer level.

Section 6.3.3.2.2 For a feedwater system pipe failure event, the passive residual heat removal heat exchanger and the core makeup tanks are actuated.

Draft for Rev.17 3 W86tingt10038 14.3-23 October 31,1997

L

14. I:iti:1 Test Progran Table 14.3 2 (Sheet 7 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 6.3.3.3.I For a steam generator tube rupture event, the nonsafety-related makeup pumps are automatically actuated when reactor coolant system inventory decreases and a reactor trip occurs, followed by actuation of the stanup feedwater pumps. Makeup pumps automatically function to maintain the programmed pressurizer level. The core makeup tanks subsequently actuate on low pressurizer level, if they are not already actuated. Actuation of the core makeup tanks automatically actuates the passive residual heat removal system heat exchanger.

Section 6.3.6.1 The piping resistances connecting the following PXS components and the RCS are bounded by the resistances assumed in the Chapter 15 safety analysis:

- Core makeup tanks

- Accumulators

- In-containment refueling water storage tank

- Containment recirculation

- Passive residual heat removal heat exchanger

- Automatic depressurization system valves Section 6.3.6.1.3 The bottom of the core makeup tanks are located above 27.5 the reactor vessel direct vessel injection nozzle centerline (11).

Section 6.3.6.1.3 The bottom of the in-containment refueling water storage 23.4 tank is located above the direct vessel injection nozzle centerline (ft).

Figure 6.3-1 The passive core cooling system has two direct vessel injection lines.

Table 6. 3-4 The passive core cooling system has two core makeup 2 2,000 tanks, each with a minimum required volume (ft')

h Draft for Rev 17 October 31,1997 14.3 24 Westinghouse

4. I:itti Test Program -

Table 14.3 2 (Sheet 3 of 21) ._

.1 DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value ,

Table 6. 3-4 The passive core cooling system has two accumulators, 2 2.000 each with a minimum required volume (ft')

Table 6.34 The passive core cooling system has an in containment 2 557,000 refueling water storage tank with a minimum required water volume (gallons)

Tabt: 6.34 - The passive core cooling system has two pli adjustment 2 107 baskets each with a minimum required volume (ft ').

+

Table 6. 3-4 The passive residual heat removal heat exchanger 2 106,000,000 minimum heat transfer rate (BTU /hr)

Section 7.1.2.11 Isolation devices are used to maintain the electrical independence of divisions and to see that no interaction occurs between nonsafety related systems and the safety-related system. Isolation devices serve to prevent credible faults in circuit from propagating to another circuit.

Section 7.1. 4.1. 6 The protection and safety monitoring system equipment is scismically qualified to meet design basis earthquake levels.

Section 7.1. 4.1, 6 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire, flooding, explosions, missiles, electrical faults and pipe whip.

l : Section ' 7.1. 4.1. 6 The design of the protection and safety monitoring-system equipment has margin to accommodate a loss of the normal HVAC.

Draft for Rev.17

[ M ingtIOUSS I4.3 25 October 31, t997

14. I:itial Test Program Table 143 2 (Sheet 9 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 7.1.4.2.6 The flexibility of the protection and safety monitoring system enables physical separation of redundant divisions.

Section 7.2.2.2.1 The protection and safety monitoring system initiates a reactor trip whenever a condition monitored by the system reaches a preset level.

Section 7.2.2.2.8 The reactor is tripped by actuating one of two manual reactor trip controls from the main control room.

Section 7. 3. 1. 2. 14 The d: mineralized water system isolation valves close on a signal from the protection and safety monitoring system derived from either a reactor trip signal, a source range flux doubling signal, low input voltage to the IE de and uninterruptible power supply battery chargers, or a safety injection signal.

Section 7.3.1.2.15 The chemical and volume control system makeup line isolation valv: automatically close on a signal from the protection and moni'oring system derived from either a high-2 pressurizer level, high steam generator level signal, or a safeguards signal coincident with high-l pressurizer level.

Section 7.3.1.2.2 The in-containment refueling water storage tank is aligned for injection upon actuation of the fourth stage automatic depressurization system via the protection and safety monitoring system.

Section 7.3.1.2.3 The core makeup tanks are aligned for operation on a safeguards actuation signal or on a low pressurizer level signal via the protection and safety monitoring system.

Draft for Rev.17 October 31,1997 14,3 26 Westinghouse

I 14.1:itial Test Program I

'l Table 14.3 2 (Sheet 10 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR. Reference Design Feature Value -

Section 7.3.1.2.4 The fourth stage valves of the automatic depressurization i system receive a signal to open upon the coincidence of a low core makeup tank water level and low reactor coolant system pressure following a preset time delay after the l third stage depressurization valves receive a signal to

.open via the protection and safety monitoring system.

Section 7.3.1.2.4 The Grst s' age valves of the automatic depressurization system open upon receipt of a signal generated from a 3 I

core makeup tank injection alignment signal coincident with core makeup tank water level less than the Low-l setpoint in either core makeup tank via the protection and safety monitoring system.

Section 7.3.1.2.4 The second and third stage valves open on time delays following generation of the Orst stage actuation signal via the protection and safety monitoring system.

Section -7.3.1.2.5 The reactor coolant pumps are tripped upon generatio'n of a safeguards actuation signal or upon generation of a low pressurizer water level signal.

Section 7.3.1.2.7 The passive residual heat removal heat exchanger control valves are opened on low steam generator water level or on a CMT actuation signal via the protection and safety monitoring systsm.

Section 7.3.1.2.9 The containment recirculation isolation valves are opened on a safeguards actuation signal in coincidence with low in-containment refueling water storage tank water level via the protection and safety monitoring system.

Draft for Rev.17 M fgh0088 14.3 27 October 31,1997

C',_

~

14. Initial Tee Progr:m

. Table 14.3 2 (Sheet 11 of 21)

DESIGN H4 SIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 7.3.2.2.1 The protection ara menitoring system automatically generate an actuation signal % . r an engineered safety feature whenever a monitored condition reaches a preset level.

Section 7.3.2.2.9 Manual initiation at the system level exists for the engineered safety featutes actuation.

Section 7.4.3.1 If temporary evacuation of the main control room is required because of some abnormal main control room condition, the operators can establish r.nd maintain safe shutdown conditions for the plant from outside l.e main control room through the use of controls and monitoring located at the remote shutdown workstation.

Section 7.4.3.1.1 The remote shutdown workstation equipment is similar to the operator workstations i.! the main control room and is designed to the same standards. One remote shutdown workstation is provided.

Section 7.4.3.1.3 The remote shutdown workstatien achieves and maintains safe shutdown conditions from full power conditions and maintains safe shutdown conditions thereafter.

Section 7.5.4 The protection and safety monitoring system provides signal conditioning, communications, and display functions for Category i variables and for Category 2 variables that are energized from the Class IE uninterruptible power supply system.

Draft for Rev.17 October 31,1997 14.3 28 Westlngt10t!Se

= 4 E

~

14. I;hial Test Program Table 14.3 2 (Sheet 12 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value

- Section 7. 6. L 1 An interlock is provided for the normally closed motor-operated normal residual heat removal system inner and outer suction isolation valves. Each valve is interlocked-so that it cannot be opened unless the reactor coolant system pressure is below a preset pressure.

Section 8.3.2.1.2 The non-Class lE de and UPS system (EDS) consists of the electric power supply and distribution equipment that provides de and uninterruptible ac power to nonsafety-related loads.

Section 9.1.1.2.1 -In the unlikely evem of a dropping of an unirradiated fuel assembly, accidental deformation of the fuel rack will be determined and evaluated in the criticality analysis to demonstrate that it does not cause criticality criterion to be violated.

Section 9.1. 2. 2. I - in the unlikely event of a dropping of an irradiated fuel assembly, accidental deformation of the fuel rack will be determined and evaluated in the criticality analysis to.

demonstrate that it does not cause criticality criterion to be violated.

Section - 9.1. 3. 5 The spent fuel pool is designed such that a water level is -

maintained above the spent fuel assemblies for at least 7 days following a loss of the spent fuel cooling system using only safety related makeup water sources (See Table 9.14).

Section 9.1. 3. 5 The spent fuel pool cooling system includes safety-related connections to establish safety-related makeup to the spent fuel pool following a design basis event including a seismic event.

4 Draft for Rev.17 Y WO5tillM [a,3 29 October 31,1997 )

e

I l! 14. I itial Test Program l l

1

.-Table 14.3 2 (Sheet 13 of 21) - .l DFSIGN BASIS ACCIDENT ANALYSIS I

SSAR Reference Design Feature Value Section 9.1, 4.1. I in the event of a safe shutdown earthquake (SSE),

handling equipment cannot fail in such a manner as to prevent required function of seismic Category I equipment.

- Section 9.3.6.3.7 The chemical and volume control system contains two redundant safety-related valves to isolate the demineralized water system from the makeup pump suction.

Section 9.3,6.3.7 The chemical and volume control system contains two safety related valves to isolate the makeup Dow to the reactor coolant system.

Section 9.3.6.4.5 The chemical and volume control system contains two safety-related valves to isolate the makeup Dow to the reactor coolant system.

Section 9.3.6.4.S.1 The chemical and volume control system contains two redundant safety related valves to isolate the demineralized water system from the makeup pump suction.

Section 9.3.6.7 The demineralized water system isolation valves close on a signal from the protection and safety monitoring system derived from either a reactor trip signal, a source range Dux doubling signal, low input voltage to the IE de and uninterruptible power supply battery chargers, or a safety injection signal, d .

Draft' for Rev.17 October 31,1997 }4,3 30 MDll88

. . . - . . . . ~ - . . . - - .- -. . . . . ,. -

"l4.1:ltial Test Progra :a -

Table 14.3-2 (Sheet 14 of 21)

DESIGN BASIS ACCIDENT ANALVSIS SSAR Reference Design Feature Value Section _9.3.6.7 . The chemical and volume control system makcup line isolation valves automatically close on e sisnal from the protection and safety monitoring system derived from either a high-2 pressudzer level, high steam generator levei signal, or a safeguards signal coincident with high-1 pressurizer level.

Section 10. 1. 2 Safety valves are provided on both main steam lines.

Section - 10. 2. 2. 4.'3

. The flow of the main steam entering the high-pressure turbine is controlled by four stop valves and four governing control valves. Th: stop valves are closed by actuation of the emergency trip system devices.

Section 10.3.1.1 The main steam supply system is provided with a main steam isolation valve and associated MSIV bypass valve on each main steam line from its respective steam generator.

Section 10.3.1.I Main steam isolation valve (MSIV) prevent the uncontrolled blowdown of more than one steam generator and isolate nonsafety-related portions of the system.

Section 10.3.1.2 Power operated atmospheric relief valves are provided to allow controlled cooldown of the steam generator and the reactor coolant system when the condenser is not available.

' Section .10. 3. 2.1 The main steam supply system includes:

-One main steam isolation valve and one main steam isolation valve bypass valve per main steam line.

-Main steam safety valves.

-Power-operated atmospheric relief valves and upstream isolation valves.

I Draft for Rev.17 l 3 W8StiflgtIOUS8 14,3 31 October 31,1997

i=i 14.1:1t1:1 Test Prograra

' Table 14.3 2 (Sheet 15 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference - Design Feature Value Section - 10.3.2.3,2 In the event that a design basis accident occurs, which results in a large steam line break, the main steam isolation valves with associated main steam isolation bypas; valves automatically close.

Figure 10. 3. 2-1 The steam generator system consists of two main steam, two main feedwater, and two startup feedwater lines.

Table 10, 3. 2-1 . Design data for main steam supply system valves:

Number per main steam line 3 Minimum relieving capacity per valve at 110% of design 1,540,000 pressure (Ib/hr)

Table 10. 3. 2-2 The minimum flow capacity of the steam generator safety 2 4,600,000 valves (Ibm /hr)

Table .10,3. 2 2 The maximtm set pressure of the steam generator safety s 1,195 valves (psig)

Section 10.3.8.3 The safety related portions of the am generator blowdown system are located in the containment and auxiliary buildings and are designed to remain functional after a safe shutdown earthquake.

Section 10,4.7.1.I Double valve main feedwater isolation is provided via the main feedwater control valve and main feedwater isolation valve. Both valves close automatically on main feedwater isolation signals, an appropr; ate engineered safety features isolation signal, within the time established with the Technical Specifications, Section 16.1. The startup feedwater control valve also serves as a containment isolation valve.

. Draft for Rev 17 October 31,1997 14.3-3' M@US8

^

~ l4. liitial Test Progr:m

, Table'14.3 2 (Sheet 16'of 21)

- DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference : Design Feature Value Section 10._4. 7. 1. 1 The condensate and feedwater system provides redundant isolation valves for the main feedwater lines routed into containment.

Section 10. 4. 7.-1. I For a main feedwater or main steam line break (MSLB) inside the containment, the condensate and feedwater system is oesigned to limit high energy fluid to the broken loop.

Section - 10.4.7.1.2 The booster / main feedwater pumps are tripped simultaneously with the feedwater isolation signal to close the main feedwater isolation valves.-

Section 10.4.7.2.I The main feedwater pumps and booster pumps are tripped with the feedwater isolation signal that closes the main feedwater isolation valves. The same isolation signal-closes the isolation valve in the cross connect line between the main feedwater pump discharge header and the startup feedwater pump discharge header.

Section 10.4,7.2.2 One MFIV is installed in each of the two main feedwater lines outside the containment and downstream of the feedwater control valve. The MFIVs are installed to prevent uncontrolled olowdown from the steam generators in the event of a feedwater pipe rupture. The main feedwater check valve provides backup isolation. In the event of a secondary side pipe rupture inside the containment, the MFIVs limit the quantity of high energy fluid that enters the containment through the broken loop and limit cooldown. The MFCV provides backup isolation to limit cooldown and high energy fluid addition.

Draft for Rev.17

[ M llgtl0088 14.3-33 October 31,1997 P

l sh

14. I:iti:1 Test Progran 3

w-

- Table 14.3 2 (Sheet 17 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value

-Section- - 10.4,7.2.2 In the event of a secondary side pipe rupture inside the containment, the main feedwater control valves provide a-redundant isolation to the MFIVs to limit the quantity of-high energy fluid that enters the containment through the 4

broken loop '

Section 10.4.7.3 For a main feedwater line break inside the containment or a main steam line break, the MFIVs and the main feedwater control valves automatically close upon receipt of a feedwater isolation signal.

Section 10.4,7.3 For a steam generator tube rupture event, positive and redundant isolation is provided for the main feedwater (MFIV and MFCV) with isolation signals generated by the protection and safety monitoring system (PMS).

- Section - 10,4.8,2.2,7 Blowdown system isolation is actuated on low steam generator water levels. The isolation of steam generator blowdown provides for a continued availability of the steam generator as a heat sink for decay heat removal in conjunction with operation of the passive residual heat removal system and the startup feedwater system.

- Section 10.4.8.3 The safety-related portions of the steam generator blowdown system located in the containment and auxiliary buildings are designed to remain functional after a safe shutdown earthquake.

Section 10.4.9.1.I Double valve startup feedwater isolation is provided by the startup feedwater control valve and the startup feedwater isolation valve. Both valves close on a startup feedwater isolation signal, an appropriate engineered -

safeguards features signal, within the time established within tne Technical Specifications, Section 16.1.

Draft for Rey,17 October 31,1997 14.3 34 Y DI@M

14. I;itial Test Progr:0 Table 14.3-2 (Sheet 18 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 10.4.9.1.I For a steam generator tube rupture event, positive and redundant isolation is provided for the startup feedwater system (startup feedwater isolation signal and startup feedwater control valve), with isolation signals generated by the protection and safety monitoring system.

Section 10.4.9.2.2 in the event of a steam generator tube rupture, the startup feedwater isolation valve and startup feedwater control valve limit over0ll of the steam generator by terminating

. startup feed Dow.

Section 10.4.9.2.2 in the event of a secondary pipe rupture inside containment, the startup feedwater isolation valve and stanup feedwater control valve provide isolation to limit the quantity of high energy Guid that enters the containment.

Section 10.4.9.2.2 The startup feedwater isolation valve is provided to prevent the uncontrolled blowdown from more than one steam generator in the event of startup feedwater line rupture. The startup feedwater isolation valve provides backup isolation.

Table 15.0 1 initial core thermal power (MWt) 1933 Table 15.0 3 Nominal values of peninent plant parameters used in accident analysis with 10% steam generator tube plugging 1 - Reactor coolant flow per loop (gpm) 9.48 E+04 Section 15.1.2.I Continuous addition of excessive feedwater is prevented by the steam generator high-2 water level signal trip, which closes the feedwater isolation valves and feedwater control valves and trips the turbine, main feedwater pumps and reactor.

Draft for Rev.17

[ WC5tiligh00$8 14.3-35 October 31,1997

14. Iri:ial Test Program Table 14.3 2 (Sheet 19 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 15.l.4.1 For an inadvertent opening of a steam generator relief of safety valve, core makeup tank actuation occurs on a safeguards ("S) signal from one of four sources:

Two out of four low pressurizer pressure signals

-Two out of four low pressurizer level signals Two out of four low T g, signals in any one loop Two out of fou-low steam line pressure signals in any one loop Section 15.l.4.I, After an inadvertent opening of a steam generator relief of safety valve, redundant isolation of the main feedwater lines closes the feedwater control valves and feedwater isolation valves, and trips the main feedwater pumps.

Section 15. 1. 5. 1 Following a steam line rupture, core makeup tank actuation occurs on a safeguards ("S) signal from one of five sources:

- Two out of four low pressurizer pressure signals

-Two out of four high-1 contairiment pressure signals Two out of four low steam line pressure signals in any loop

-Two out of four low T u, signals in any one loop

-Two out of four low pressurizer level signals l

Section 15.1.5.1 After a steam line rupture, redundant isol.aion of the main feedwater lines closes the feedwater control valves and feedwater isolation valves, and trips the main feedwater pumps.

Section 15.1.5.2.I Core makeup tanks and the accumulators are the portions of the passive core cooling system used in mitigating a stam line rupture.

Draft for Rev.17 October 31.1997 14,3 36 3 Westirighouse

_ . . . - - -= - -- -. - -

' 14.1:ltist Test Progra:3 I

' Table 14.3 2 (Sheet 20 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 15.II6.1 The heat sink for the PRHR heat exchanger is provided by the IRWST, in which the PRHR heat exchanger is submerged.

Section 15,2.6.2.I Following a loss of ac power, the PRHR heat exchanger is actuated by the low steam generator water level (wide range).

Section !!.2.8.2.I Receipt of a low steam line pressure signal in at least one steam line initiates a steam line isolation signal that closes all main steam line and feed line isolation valves.

This signal also gives an "S" signal that initiates flow of cold borated water from the core makeup tanks to the reactor coolant system.

' Section 15.3.3.2.2 The pressurizer safety valves are fully open at 2575 psia.

Their capacity for steam relief is described in Section 5.4.

Section 15,4,6.2.2 A safety signal from the protection and safety monitoring i system automatically isolates the potentially unborated water from the demineralized water transfer and storage system and thereby terminates the dilution Section 15. 5. 1. I Following inadvertent operation of the core makeup tanks

' during power operation, the high 3 pressurizer level signal actuates the PRHR heat exchanger and blocks the pressurizer heaters.

Section 15.5.2.I The pressurizer heaters are blocked, and the main feedwater lin:s, steam lines, and chemical and volume control system tre isolated.

i

-s.

Draft for Rev.17 Y W 8Silh M i4,3 37 OctoS:r 31,1997

14. I:iti::t ht Progr::2

. Table 14.3 2 (Sheet 21 of 21)

DESIGN BASIS ACCIDENT ANALYSIS SSAR Reference Design Feature Value Section 18.8.3.2 The main control area includes two reactor operator workstatinas, the supervisor's workstation, the dedicated safety panel and the wall panel information system.

Section 18.8.3.2 The human system interface resources available at each workstation are the plant information system displays, the control displays (soft controls), the alarm system support displays, procedure system, and the screen and component selector.

Draft for Rev.17 October 31,1997 14,3 3g [ W95tingh0USS

y 3 14."I:itial Test Prograa 3 Table 14.3 3 ANTICIPATED TRANSIENT WITHOUT SCRAM SSAR Reference Design Feature - Value Section -7.7.1.11- The diverse actuation system is a nonsafety-related -

system that provides a diverse backup to the protection and safety monitoring system.

Section - 7.7.1.11 - The diverse actuation system trips the reactor control rods and the turbine on low wide range steam generator water level and on low pressurizer water level.

Section 7. 7.~ 1.1 I The diverse actuation system initiates passive residual heat removal on low wide range steam generator water level or high hot leg temperature; actuates core makeup tanks and trips the reactor coolan'. pumps on low pressurizer water level; and isolates selected containment penetrations and starts passive containment cooling.

Section 7.7.1.11 The manual actuation function of the diverse actuation system is implemented by wiring the controls located in the main control room directly to the final loads in a way that byr,sses the normal path through the control room multiplexers, the englneered safety features actuation cabinets, and the diserse actuation system logic Section 7.7.1.Il The diverse actuation system uses a microprocessor board different from those used in the protection and safety monitoring system.

Section 7.7.1.11 The dive se ratuation system hardware implementation is different f om that of the protection and safety monitoring system.

Section 7.7.l.11.

The operating system a-A programming language of the diverse actuation system is different from that of the protection and safety monitoring systerr..

Draft for Rev.17 T WBS!!M 14.3-39 October 31,1997 4

14. I:lti:1 Test Program Table 14.3 4 (Sheet I of 3)

FIRE PROTECTION SSAR Reference Design Feature Value Section 3.4.1.1.2 The boundaries between mechanical equipment rooms and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent Hooding of rooms that contain safe shutdown equipment up to the maximum Dood level for each room.

Section 3.4.1.1.2 Separation is maintained between Class lE divisions and between Class IE divisions and non-Class IE cables in accordance with the Gre areas.

Section 3.4.l.l.2 The Ap600 arrangement provides physical separation of redundant safety related components and systems from each other and from nonsafety related components.

Section 3.4.l.2.2 The boundarias between mechanical equipment rooms inside containment and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent Gooding of rooms that conta:n safe shutdown equipment up to the maximum Dood level for each room.

Section 3.4.1.2.2 lloundaries exist to prevent Hooding between the following rooms which contain safety related equipment:

PXS valve / accumulator room A, PXS vahe/ accumulator room B, and chemical and volume contro! room.

Section 3.8.4.1.1 The conical roof supports the passive containment cooling systein tank, which is constructed with a stainlen steel tirwr on reinforced concrete walls.

I Section 7.1, 4.1. 6 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as Gre and Gooding.

l l

l Draft for Rev.17 October 31,1997 14,3 40 [ Westinghouse

14. I::,::1 Test Program Table 14.3 4 (Sheet 2 of 3)

FIRE PROTECTION SSAR Reference Design Feature Value Section 7.4.3.I If temporary evacuation of the main control room is required because of some abnormal main control room condition, the operators can establish and maintain safe shutdown conditions for the plant from outside the main control room through the use of controls and monitoring located at the remote shutdown workstation.

Section 7.4.3.1.1 The remote shutdown workstation equipment is similar to the operator workstations in the main control room and is designed to the same standards. One remote shutdown workstation is provided.

Section 7.4.3.l.3 The remote shutdown workstation achieses and maintains safe shutdown conditions from full power conditions and maintains safe shutdown conditions thereaner.

Section 8.3.2.2 The four Class IE battery chargers Class lE voltage regulating transformers are independent, located in separate rooms, cannot be interconnected, and their i circuits are routed in dedicated, physical'y separated raceways.

Section 8.3.2,3 Each safety related circuit and raceway is given a unique identincation number to distinguish between circuits and raceways of different voltage lesel or separation groups.

Section 8.3.2.4,2 Cables of one separation group are run in separate raceway and physically separated from cables of other separation groups. Group N raceways are separated from safety related groups A, B, C, and D. Non class IE circuits are electrically isolated by isolation Jevices, shielding and wiring techniques, physical separation, or an appropriate combination thereof.

Section 9. S.1. 2. i . I Separation is maintained between Class IE disisions and between Class IE divisions and non. Class lE cables in accordance with the Gre areas.

Draft for Rev.17

[ W85tingh00$6 14,3 4i October 31,1997 i

14. I:itial Test Program Table 14.3 4 (Sheet 3 of 3)

FIRE PROTECTION SSAR Heference Design Feature Value Section 9. 5.1. 2.1. 5 The standpipe system is supplied with water from the safety related passive containment cooling system storage tank and normally operates independently of the rest of the Hre protection system. The supply line draws water from u dedicated pation of the storage tank, using water allocated for Orc protection.

Section 9.S.1.2.;.5 The standpipe system serving areas containing equipment required for safe shutdown fcilowing a safe shutdown earthquake is designed and rapported so that it can withstand the effects of a sr.fe shutdown earthquake and remain functional.

Section 9.5.1.2.l.5 The solume of the water in the PCS tank is suf0cient to 2 18,000 supply two hose streams, each with a now of 75 gallons per minute, for two iiours (gal).

Section 18.8.3,2 1he human system interface resources available at each workstation are the plant information system displays, the control displays (soft controls), the alarm system support displays, procedure system, and the screen and component selector.

Section 18.8.3.4 The mission of the remote shutdown workstation is to provide the resources to bring the plant to a safe shutdown condition after an evacuation of the main control room.

Section 18.12.3 The controls, displays, and alarms listed in Table 18.12.21 are retrievable from the remote shmdown workstation.

Draft for Rev.17 October 31,1997 14,3 4% W W65tingh0USO

14. In:ial Test Prograa

. Table 14.3 5 (Sheet 1 of 2)

FLOOD PROTECTION SSAR Reference Design Feature Va!ue i Section 1. 1.43 The lowest level of the auxiliary building, elevation l Appendix 1 A 66' 6", contains the components of the radwaste system within a common Hood rone with watertight Doors and walls. This volume of this enclosed Good zone is suf0cient to contain the contents of the radwaste system.

Table 21 Plant elevation for maximum Good level (ft) s; 100 Section 3. 4.1.1.1 The seismic cm;ory I structures below grade are protected against Hooding by waterstops and a waterproonng system.

Section 3. 4.1. I . 2 The boundaries between mechanical equipment rooms and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent Gooding of rooms that contain safe shutdown equipment up to the maximum Dood level for each room.

Section 3.4.l.2.2 The boundaries between mechanical equipment rooms inside containment and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent Hooding of rooms that contain safe shutdown equipment up to the maximun. Good level for each room.

Section 3.4.1.2.2 Houndaries exist to prevent Gooding between the following rooms which contain safety-related equipment:

PXS valve / accumulator room A. PXS valve / accumulator room B. and chemical and volume control room.

Section 3.4.1.2.2 The AP600 arrangement provides physical separation of redundant safety related components and systems from each other and from nonsafety related components.

Section 3.4.1.2.2 The safety related components available for safety shutdown are located in the auxiliary building and inside containment. No credit is taken for operation of sump pumps to mitigate the consequences of Gooding.

Draft for Rev.17

[ W85tingh00$8 14.3 43 October 31,1997

14. I:iti:1 Test Progr:a Table 14.3 5 (Sheet 2 of 2)

FLOOD PROTECTION SSAR Reference Design Feature Value Section 3.4.1.2.2.1 The PXS A compartment, PXS B compartment and the chemical and solume control system compartment are physically separated and isolated from each other by structural walls such that flooding in any one of these compartments is in the reactor coolant sy stem compartment cannot cause flooding in any of the other compartments.

Section 3. 6 in the event of a high- or moderate enerpy pipe failure within the plant, adequate protection is provided so that essential structures, systems, or components are not impacted by the adverse effects of postulated pipe failure.

Section 7.1.4.1.6 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire and flooding.

Draft for Rev.17 October 31,1997 14,3 44 . T Westinghouse

14. I:6thi Test Progr:m f Table 14.3 6 (Sheet I of 12) I PROHABILISTIC RISK ASSESSMENT SSAR Reference Design Feature Value !

Table 2.33 The components identined under Reactor Systems in Table 3.2 3, as ASME Code Section ill are designed and constructed in accordance with ASME Code Section 111 Requirements.

Section 3. 2. l. 3 The Nuclear Island structures include the containment and the shield and auxiliary buildings. These structures are scismic Category I.

. Table 3.23 The Nuclear Island structures include the containment and the Shield and Auxiliary 13uildings. These structures are seismic Category 1.

Section 3. 4.1. l . 2 The boundaries between mechanical equipment rooms and the electrical and instrumentation and control equipment ,

rooms of the auxiliary building are designed to prevent (looding of rooms that contain safe shutdown equipment up to the maximum Dood level for each room.

Section 3.4.1.l.2 The AP600 arrangement provides physical separation of redundant safety related components and systems from each other and from nonsafety related components, Section 3. 4.1.1. 2 Separation is maintained between Class IE divisions and between Class IE divisions and non Class IE cables in accordance with the Arc areas.

Section 3.4.1.. 2 The Ap600 arrangement provides physical separation of redundant safety related components and systems from each other and from nonsafety related components.

Section 3.4.l.2.2 lloundaries exist to present Gooding between the following rooms which contain safety related equipment:

PXS valve / accumulator room A, PXS valve / accumulator room 13, and chemical and volume control room.

Draft for Rev.17 Westingh0088 14.3-45 October 31,1997

14. I:iti:1 Test l'rogr:a Table 14.3 6 (Sheet 2 of 12)

I'ROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Yalue Section 3. 4. 1.

The boundaries between mechanical equipment rooms inside containment and the electrical and instrumentation and control equipment rooms of the auxiliary building are designed to prevent nooding of rooms that contain safe shutdown equipment up to the maximum Dood level for each room.

Section 3.4.1.2.2 The rafety related components available for safety shutdown are located in the auxiliary building and inside containment. No credit is taken for operation of sump pumps to mitigate the consequences of Gooding.

Section 3.4.1.2.2.I The PXS A compartment, PXS B compartment and the chemical and volume control sys'em compartment are physically separaied and isolated from each other by structural walls such that Gooding in any one of these compartments or in the reactor coolant system compartment cannot cause Dooding in any of the other compartments.

Section 3D.6 RXS equipment in Appendix 3D is seismically qualified.

Section 5.1. 3 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.

Each stage 4 discharges from a hot leg to the RCS loop compartment.

Section 5. 3.1. I The reactor vessel prosides a high integrity pressure boundary to contain the reactor coolant, heat generating reactor core, and fuel fission products. The reactor vessel is the primary boundary for the reactor coolant and the secondary barrier against the release of radioactive fission products.

Draft for Rev.17 October 31,1997 14.3-46 3 We @ 00$0

H. I:itial Test Prograa Table 14.3 6 (5heet 3 of 12)

I'ROllABILISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 5,4.6 ADS has four stages. Each stage is arranged into two separate groups of vahes and lines.

Stages 1,2, and 3 discharge from the top of the pressuriier to the IRWST.

Each stage 4 discharges from a hot leg to the RCS loop compartrnent.

Section 5. 4. 6. 2 Each ADS stage I,2, and 3 line contains two normally closed motor operated valves (MOVs).

Section 5. 4. 6. 2 Each ADS stage 4 line contains a normally open MOV valve and a normally closed squib valve.

Section 5.4.7 The RNS removes heat from the core and reactor coolant system at reduced RCS pressure and temperature conditions after shutdown.

Section 5. 4. 7 1he normal residual heat removal system (RNS) provides a safety related means of performing the following functions:

Containment isolation for the RNS lines that penetrate the containment Long term, post accident makeup water to the RCS Section 5. 4. 7. l . I The RNS containment isolation and pressure boundary vahes are safety related. The motor operated vahes are powered by Class IE de power.

Section 5.4.7.1.2.I The component cooling water system (CCS) provides cooling to the RNS heat exchanger.

Section 6.2,4 The containment hydrogen control system provides nonsafety related hydrogen igniters for control of the containment hydrogen concentration for beyond design basis accidents.

Section 6,2,4.2.3 At least 64 hydrogen igniters are provided.

Draft for Rev.17 3 W65tlflgh00S8 14,3 47 October 31,1997

14. I:itial Test Prograa

. Table 14.3 6 (Sheet 4 of 12)

PROBAltlLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Table 6.2.42 The minimum passive autocatalpic rccombiner depletion 21 rate at 120' F and atmospheric pressure (scfm)

Section 6. 3 The automatic deptessurization system provides a safety-related means of depressurizing the RCS.

Section 6. 3 The in containment refueling water storage tank subsystem provides a safety related means of performing the following functions:

Low pressure safety injection Core decay heat sink during design basis events Flooding of the lower containment, the reactor cavity and the loop compartment by draining the IRWST into the containment.

Dorated water Section 6. 3. I The core makeup tanks provide safety related means of safety injection of borated water to the RCS.

Section 6. 3. I Passhe residual heat removal (PRilR) provides a safety-related means of removing core decay heat during design basis events.

Section 6.3.2 The ADS valves are powered from Class IE de power, Section 6.3.2 There are two CMTs, each with an injection line to the reactor vessel /DVI nozzle.

- Each CMT has a pressure balance line from an RCS cold leg.

- Each injection line is isolated with a parallel set of air.

operated valves (AOVs).

These AOVs open on loss of air.

The injection line for each CMT also has two check valves in series.

Draft for Rev.17 October 31,1997 14.3 4g [ Westinghouse

14.1:itial Test Prograa Table 14.3 6 (Sheet 5 of 12)

PRollAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 6.3.2 The IRWST subsystem has the following flowpaths:

.Two (redundant) injection lines from the IRWST to the reactor vessel /DVI nozzle. Each line is isolated with a parallel set of vahes; each set with a ; heck valve in series with a squib salve.

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 an MOV, the other path contains a squib valve and a check valve.

The two MOV/ squib valve lines also provide the capability to flood the reactor cavity.

Section 6.3,2 There are screens for each IRWST injection line and recirculation line.

Section 6.3.2 PRilR is actuated by opening redandant, parallel air-operated valves. These air operated valves open on loss of air.

Section 6.3.2.2 The passive core cooling system (PXS) is composed of the following:

I Accumulator subsystem 1 Core makeup tank (CMT) subsystem I In containment refueling water storage tank (IRWST) subsy stem 1 Passise residual heat removal (PRilR) subsystem.

1 - The automatic depressurization system (ADS), which is a subsystem of the reactor coolant system (RCS), also supports passive core cooling functions.

Section 6.3.2.2.2 There are two accumulators, each with an injection line to the reactor sessel' direct vessel injection (DVI) nonle.

Each injection line has two check valves in series.

Section 6.3,2,2.2 The accumulators provide a safety rclated means of safety injection of borated water to the RCS.

Draft for Rev.17

[ W8SillighouS8 14,3 49 October 31.1997 1

14.1:itial Test Progra Table 14.3 6 (Sheet 6 of 12)

PROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Sec' ion 6.3.3 1RWST squib valves and MOVs are powered by Class IE de power.

Section 6.3.3 The CMT AOVs are automatically and manually actuated from PhiS and DAS.

Section 6.3.3 The PRHR air operated valves are automatically actuated and manually actuated from the control room by either PMS or DAS.

Section 6.3.3 The squib valves and MOVs for injection and recirculation are automatically and manually actuated via PMS, and manually actuated via DAS.

Section 6.3.3 The squib valves and MOVs for lower containment and reactor cavity Gooding are manually actuated via PMS and DAS from the control room.

Section 6.3.7 The positions of the containment recirculation isolation MOVs are indicated in the control room.

Section 6.3.7 The position of the inlet PRilR valve is indicated in the control room.

Section 6.3.7.6.I The ADS first , second , and third stage valve positions are indicated in the control room.

Section 7.l.1 The diverse actuation system provides a nonsafety-related means of performing the following functions:

Initiates automatic and manual reactor trip Automatic and manual actuation of selected engineered safety features Main control room display of selected plant parameters.

4 Draft for Rev.17 October 31,1997 14.3 50 3 Westilighouse

14. Initial Test Program i Table 14.3 6 (Sheet 7 of 12)

PROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 7.1.I The protection and safety monitoring system provides a safety related means of performing the following functions:

Autornatic and manual reactor trip Automatic and manual actuation of engineered safety featurca (ESF).

Section 7. 1. 1 PMS provides for the minimum inventory of 0xed position controls and displays in the control room.

Section 7.1.2 Each PMS division is powered from its respective Class IE de division.

Section 7.1. 2 PMS has four divisions of reactor trip and ESF actuation.

Section 7.1.2.10 PMS automatically produces a safety related reactor trip or ESF initiation upon an attempt to bypass more than two channels of a function that uses 2-out of 4 logic.

Section 7.1.2.15 The PMS hardware and software are developed using a planned design process which provides for specific design documentation and reviews during the design requirement, system definition, development, test and installation phases.

Section 7. l. 2. 6 PMS has redundant divisions of safety related post-accident parameter display.

Section 7.1. 4.1. 6 The ability of the protection and safety monitoring system to initiate and accomplish protective functions is maintained despite degraded conditions caused by internal events such as fire and flooding.

Section 7.1. 4.1. 6 The design of the protection and safety monitoring system I equipment has margin to accommodated a loss of the normal HVAC.

Draft for Rev.17

[ W65tlligh00S8 14.3 51 October 31,1997

l

14. I:itial Test Program j t

_ Table 14.3 6 (Sheet 8 of 12) l t

PROHAHILISTIC RISK ASSESSMENT SSAR Peference Design Feature Value ,

Section 7.1.4.2.6 The Dexibility of the protection and safety monitoring system enables phys! cal separation of redundant divisions.

Figure 7.18 PMS has redundant divisions of safety related post- ,

accident parameter display.

Section 7.2.2.2.l The protection and safety monitoring system initiates a reactor trip whenever a condition monitored by the i system reaches a preset level. ,

Section 7. 3 The PMS allows for the transfer of control capability from the main control room to the remote shutdown room. The minimum inventory of displays and controls in the temote shutdown room is provided.

Section 7. 3.1 Tbc ADS valves are powered from Class IE de power.

Section 7, 3.1 The ADS valves are automatically and manually actuated :

via the piotection and safety monitoring system (PMS).

md n anually actuated via the diverse actuation system (DAS)

Section 7.3.1 The CMT AOVs are automatically and manually actuated from PMS and DAS.

Section 7. 3.1 The squib valvos and MOVs for injection and recirculation are automatically and manually actuated via PMS, and manually actuated via DAS.

Section 7. 3.1 The squib valves and MOVs for reactor cavity Gooding are manually actuated via PMS and DAS from the control room.

Section 7, 3. 1 The PRiik air operated valves are automatically actuated find manually actuated from the control room by either PMS or DAS.

Draft for Rev.17

, October 31,1997 14,3 9 3 Westingh0088 r- -m-- --~ , 4 ,- , , > , - - e , ,

14. Irri Test Program

. Table 14.3 6 (Sheet 9 of 12)

PROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 7.3.1 The RNS containment isolation MOVs are actuated via PMS.

Section 7.6.1.I An interlock is provided for the normally closed motor-operated normal residual heat removal sy tem inner and outer suction isolation valves. Each valve is interlocked so that it cannot W opened unless the reactor coolant system pressure la below a preset pressure.

Section 7.7.1.11 The diverse actuation system is a nonsafety related system that provides a diverse backup to the protection and safety monitoring system.

Section 7.7.1.11 The diverse actuation system trips the reactor control rods and the turbine on low wide range steam generator water level and on low pressurizer water lesel.

Section 7.7.1.11 DAS manual initiation functions are implemented in a manner that bypasses the signal processing equipment of the DAS.

Section 7.7.1.11 The DAS automatic actuation signals are generated in a functionally diserse manner frem the PMS signals.

Diversity betw en DAS and PMS is achieved by the use of different architecture, different hardware implementations, and different software.

Section 8. 3. l .1. I On loss of power to a 4160V diesel backed bus, the associated diesel generator automatically starts and produces ac [xmer. The source circuit breakers 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.

Section 8.3.1.I.2.I Two onsite standby diesel generator units provide power to the selected nonsafety related ac loads.

Draft for Rev.17 3 W85tingh0US8 14.3 53 October 31,1997

14.1:itial Test Program

. Table 14.3 6 (Sheet 10 of 12)

PROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 8. 3.1.1. 3 The main ac power system distributes non Class IE power from onsite sources to selected nonsafety related loads.

Section 8.3.2.1 The Class lE de and uninterruptible power supply (UPS) system (IDS) provides de and unintei.uptible ac power for the safety related equipment.

Section 8.3.2.1.l.1 There are four independent, Class lE 125 Vdc divisions.

Divisions A and D are cach composed of one battery bank, one switchboard, and one battery charger.

Divisions B and C are e, h composed of two battery banks two switchboards, and two battery chargers. The first battery bank in the four divisions is designated as the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months battery bank. The second battery bank in Divisions !! and C is designated as the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months battery bank.

Section 8.3.2.1.I,i Battery chargers are connected to de switchboard buses.

The mput ac power for the Class lE de battery chargers is supplied from onsite diesel generator backed low-voltage ac power supplies.

Section 8.3.2.1.1.1 The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months battery banks provide power to the loads for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without recharging. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I battery banks supplys a de switchboard bus load for a period of'/2 hours without recharging.

Section 8.3.2.1.2 The non Class IE de and UPS system (EDS) consists of the electric power supply and distribution equipment that provides de and uninterruptible ac power to nonsafety.

related loads.

Section 8.3.2.1.2 The non-Class IE de and UPS system (EDS) consists of the electric power supply and distribution equipment that provides de and uninterruptible ac power to nonsafety-related loads.

Draft for Rev,17 October 31,1997 l4,3 54 Westilighouse

11. I:stial Test Program Table 14.3 6 (Sheet 11 of 12)

PROllAlllLISTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 8.3.2.I.2 EDS load groups 1,2, and 3 provide 125 Vdc power to the associated inverter units that supply the ac power to the non Class IE uninterruptible power supply ac system.

Section 8.3.2.1.2 Battery chargers are connected to de switchboard buses.

1he input ac power for the Class lE de battery chargers is supplied from onsite dicsci g-nerator backed low-voltage ac power supplies.

Section 8.3.2.1.2 The onsite standby diesel generator backed low voltage ac power supply provides the normal ac power to the ', tery chargers.

I Section 8.3.2.1.3 Separation is provided between Class IE divisions, and between Class IE divisions and non Class IE cables.

Section 9. 2. I The service water system is a nonsafety related system that transfers heat from the component cooling water heat ,

exchangers to the atmosphere. l Section 9.2.1.2.1 The SWS :s arranged into two trains. Each train includes one punip and ore cooling tower cell.

I Section 9.2.2 The component cooling water system is a nonsafety-rela'ed system that removes heat from various components and transfers the heat to the service water sy stem (SWS).

Section 9,2.2.2 The CCS is arranged into two trains. Each train includes one pump and one heat exchanger.

Section 9.3.6 The CVS provides a nonsafety related means to perform the following functions:

Maicup water to the RCS during r.ormal plant operation

- Boration following a failure of reactor trip Coolant to the pressurizer auxiliary spray line.

Draft for Rei.17 W Westinghouse 9,3 55 october 31,1997

14. I:itial Test Progran Table 14.3 6 (Sheet 12 of 12)

PROIIAllit.lSTIC RISK ASSESSMENT SSAR Reference Design Feature Value Section 9.3.6.1 The chemical and volume control system (CVS) provides a safety related means to terminate inadvertent RCS boron dilution.

Section 9. 4.1 The main control room has its own ventilation system and is pressurized. The ventilation system for the remote shutdown room is independent of the ventilation system for the main control room.

Section 9.5.1.2..l.1 The PMS allows for the transfer of control capability from the main control room to the remote shutdown workstation. The minimum inventory of displays and controls in the remote shutdown room is provided.

Section 9.5.1.2.1.1 Class IE cables are routed in their respective divisional raceways.

Section 9.5.1.2.l.I Separation is maintained between Class IE divisions and between Class IE divisions and non-Class lE cables in accordance with the fire areas.

Section 16.2.I li.iponant reliability assumptions made as pan of the AP600 probabilistic risk assessment (PRA) will remain valid throughout plant life.

Section 18.8.3.2 The main control area includes two reactor operator workstations, the supervisor's workstation, the dedicated safety panel and the wall panel information system.

Section 18.12.2 The minimum inventory of instrumentation includes those displays, controls, and alarms that are used to monitor the status of the critical safety functions and to manually actuate the safety related systems that achieve the critical safety functions. The minimum inventory resulting from the implementation of the selection criteria is provided in Table 18.12.21.

Draft for Rev.17 October 31,1997 14,3 56 3 Westingh0USB t

14. I:itial Test Progr:m Table 14.3 7 (Sheet 1 of 4)

RADIOLOGICAL ANAINSIS SSAR Reference Design Feature Value Table 2l Plant elevation for maximum flood level (ft) s100 Section 2.3.4 Atmospheric dispersion factors X/O (sec/m')

- Site Boundary X/Q 0 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time interval s 1.0 x 10

Low Population Zone Boundary X/Q 0- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s 1.35 x 10 d 8- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s 1.0 x 10d 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months s 5.4 x 10 4 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months s 2.2 x 10 4 Table 6.2.31 Containment penetration isolation features are configured as in Table 6.2.31 Table 6.2.31 Maximum closure time for remotely operated containment s5 purge valves (seconds)

Table 6.2.31 Maximum closuie time for all other remotely operated s 60 containtnent isolation valves (seconds) l Section 6. 4. 2. 3 The minimum storage capacity of each set of stoiage s 43034314,ln tanks in the VES (sc0 Section 6 4.3.2 The maximum temperature rise in the main control room s 15 pressure boundary following a loss on the nuclear island nonradioactive sentilation system over a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period

(*F)

Section 6.J.3.2 'lhe maximum temperature in the instrumentation and s 125 control rooms and de equipment rooms following a loss of the nuclear island nonradioactise ventilation system remains os er a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period ('F).

I Section 6.4.4 The main control emergency habitability system EMS

25 l nominally provides 245 scfm of sentilation air to the I main contrel room from the compressed air storage tanks I h m W M % ring. ." M n % A 4 m +- m I deP. es; (sefm).

Draft for Rev.17 T W85tingh0US8 14,3 57 October 31,1997

14. I:iti: Test Program

. Table 14.3 7 (Sheet 2 of 4)

RADIOLOGICAL ANALYSIS SSAR Hererence Design Feature Value 1 Section 6.4.4 Tr.:: : %my tj" r scfm of sentilation Cow is 1/8*

suf0cient to pressurize the control room to 1/8* inch water gauge differential pressure (WlC).

l'igure 6,42 The main control room emergency habitability system I consists of twea sets of emergency air storage tanks and an air delivery system to the main control room.

Section 6. 5. 3 The passive heat removal process and the limited leakage from the containment result in offsite doses less than the regulatory guideline limits.

Section 8. 3.1.1. 6 Electrical penetrations through the containment can withstand the maximum short circuit currents available either continuously without exceeding their thermal limit, or at least long(r than the Geld cables of the circuits so that the fault or overload currents are interrupted by the protectise devices prior to a potential failure of a penetration.

I Section 9. 4.1.1. I The VBS isolates the HVAC dewswkp/ ping that penetrates the main control room boundery on high particulate or iodine concentrations in the main control room supply air or on extended loss of ac power to support operation of the main control room emergency habitability system.

I 46% 48 * ' 2 TWwn: e-"^' : :w!ud: '"

r ep r' -

^^

I wwk twh ;, 'h: impen W -4 ::,:!w de: :::ed I afwy ;u :! . 9 'h: " ^'! p: :' ' erm.e ' ,

w Draft for Rev.17 October 31,1997 14,3 3g 3 W85tingh0US8 L

14.1:i:21 Test Prograa Table 14.3 7 (Sheet 3 of 4)

RADIOLOGICAL ANALYSIS SSAR Reference Design Feature Value Section 12.3.2.2.I During reactor operation, the shield building protects personnel occupying adjacent plant structures and yard areas from radiation originating in the reactc,r vessel and primary loop components. The eencrete shield building wall and the reactor vessel and steam generator compartment shield walls reduce radiation levels outside the shield building to less than 0.25 mrem /hr from sources inside containment. The shield building completely surrounds the reactor components.

Section 12.3.2.2.2 i lhe reactor vessel is shielded by the concrete primary shield and by the concrete secondary shield which also surrounds other primary loop components. The secondary shield is a structural module filled with concrete surrounding the reactor coolant system equipment, including piping, pumps and steam generators. Extensive shielding is provided for areas surrounding the refueling cavity and the fuel transfer canal to limit the radiation levels.

Section 12.3.2.2.3 Shielding is provided for the liquid radwaste, gaseous radwaste and spent resin handling systems consistent with the maximum postulated activity. Corridors are generally shielded to allow Zone 11 access, and operator areas for valve modules are generally Zone 11 or 111 for access.

Shielding is provided to attenuate radiation from nonnal residual heat removal equipment during shutdown cooling c,perations to levels consistent with radiation zoning requiremerits of adjacent areas.

Draft for Rev.17 3 W65tingh00$8 14.3 59 October 3 6,1997

14. luttal Test Progran Table 143 7 (Sheet 4 of 4)

RADIOLOGICAL ANALYSIS SSAR Reference Design Feature Value Sec' ion 12,3.2.2,4 The ennerete shield walls surrounding the spent fuel cask loaamg and decontamination areas, and the shield walls i surrounding the fuel transfer and storage are sufficiently thick to limit radiation levels outside the shield walls in accessib!c areas to Zone 11. The building walls are sufficient to shield external plant areas which are not controlled to Zone 11.

Section 12,3,2.2.$ f.hielding is provided as necessary for the waste storage

. areas in :he radwaste building to meet the radiation zone and access requirernents.

Section 12. 3. 2. 2. 7 Shielding combined with other engineered safety features is provided to f.ermit necess and occupancy of the control room following a postulated loss-of coolant accident, so that radiation doses are limited to five rem whole body from contributing modes of exposure for the duration of the accident, in accorda. ice with General Design Criteria 19.

Section 12.3.2.2.9 TSe spent fuel transfer tube is shielded to within adjacent I area radiati'n limits, is completely enclosed in concrete, ai .here is no unshicided portion of the spent fuci transfer tube during the refueling operation.

Draft for Rev.17 October 31,1997 ;4,3 60 [ Westinghouse

14.1:itial Test Progr:a Table 14.3 8 SEVERE ACCIDENT ANALYSIS SSAR Reference Design Features Value Section 5. 3. l. 2 There are no penetrations in the reactor vessel below the Cote.

Section 6.2.4.2.l The hydrogen concentration monitoring subsystem consists of two groups of eight hydrogen sensors each.

Section 6.2.4.2.2 The hydrogen recombination subsystem consists of two passive autocatalytic recombiners installed inside the containment above the operating deck.

Section 6.2.4,2.3 The hydrogen ignition subsystem consists of 64 hydrogen igniters strategically distributed throughout the containment.

Table 6.2.43 The minimum surface temperature of the hydrogen 2 1,700 ignitors ('F).

Section 6. 3 The ADS provides a safety related means of depressurizing the RCS.

Section 6. 3 The PXS provides a safety related means of lbodin<j the reactor ctvity by draining the IRWST into the containment.

Section 7.3.1.2.9 Signals to align the IRWST containment recirculation isolation valves are generated by manual initiation.

Section 7. 7. L 11 initiation of containment recirculation is a diserse manual function.

Draft for Rev.17 W Westinghouse 14.3 61 october 31,1997

NSD-NRC-97-5426, Forwards Responses to Followon Questions Re AP600 Ssar Section 14.3 & AP600 Insps,Tests,Analyses & Acceptance Criteria.Westinghouse Status Column Will Be Changed to Closed in Open Item Tracking Sys,As Result

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NSD-NRC-97-5426, Forwards Responses to Followon Questions Re AP600 Ssar Section 14.3 & AP600 Insps,Tests,Analyses & Acceptance Criteria.Westinghouse Status Column Will Be Changed to Closed in Open Item Tracking Sys,As Result

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