NSD-NRC-97-5485, Forwards Responses to FSER Open Items on AP600.Summary of Encl Responses Provided in Table 1

Forwards Responses to FSER Open Items on AP600.Summary of Encl Responses Provided in Table 1

ML20197B481
Person / Time
Site:	05200003
Issue date:	12/12/1997
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-97-5485, NUDOCS 9712230390
Download: ML20197B481 (102)

Text

,

_.-m_______._______.____..

i

,e E

] -

1 g

i Wecingtiouse Energy Systems b 3%

{

PittsburEh Penn v a 12 5

Electric Corporation NSD-NRC-97-5485 I'

Docket No.: 52-003 December 12,1997

. p V

\\

Document Control Desk i

i U.S. Nuclear Regulatory Commission 1

- ll Washington, DC 20555 j

ATFENTION: T.R. QUAY i

SUBJI!CT:

AP600 Rl!SPONSE TO FSER OPEN ITEMS I

Dear Mr. Quay:

. y

~

1 Enclosed with this letter are the Westinghouse responses to FSER open items on the AP600. A l

ummarv of the enclosed responses is provided in Ttiole 1. Included in the table is the FSER open g

s l

J item number, the associated OITS number, and the status to be de.ignated in the Westinghouse status j

q column of OITS.

The NRC should review the enclosure and inform Westinghouse of the status to be designated in the "NRC Status" column of OITS.

u l

Please contact me on (412) 374-4334 if you have any questions concerning this transmittal.

i B ia A. Mc. e, Manager At vanced Plant Safety and Licensing jml Enclosure cc:

W. C. IlulTman, NRC (Enclosuref

[.l J. M. Sebrosky, NRC (Enclosure) f ~.

()g[/

l[l T. J. Kenyon, NRC (Enclosure) 7 J. P. Segala, NRC (Enclosure)

U f

D. C. Scaletti, NRC (Enclosure)

[

N. J. Liparuto, Westinghouse (w/o Enclosure) u.ala R.l.llll.lli.l.Ii.lli.lll r

m==E mh '

A PDR

i DCF/NRCl180 NSD NRC 97 5485 2

December 12,1997 Table 1 List of FSElt Open items included in Letter DCP/NitC l FSElt Open item OITS Number Westinghouse status in OITS 410.306F 6064 Confirm W 410.339F 6194 Confirm W 410.341F 6196 Action N 410.342F 6197 Confirm W 410.345F 6200 Action N 410.352F 6207 Confirm W 410.356F 6211 Confirm W 410.358F 6213 Confirm W 410.362F 62!7 Confirm W 420.126F 6221 Confirm W 470.45F 6222 Confirm W 480.1088F 6223 Confirm W 480.td89F 6224 Confirm W 480.1094F 6229 Confirm W 480.1095F 6230 Confirm W 480.1096F 6231 Confirm W 480.1097F 6232 Confimi W 480.1098F 6233 Cs nfirm W 480.1099F 6234 Confirm W 480.1102F 6237 Confirm W 480.1103F 6238 ConGrm W 720.423F 6135 Confirm W 720.442F 6180 Confirm W l

,w.g C

j

- Table I List of FSER Open items included in Letter DCP/NRC FSER Open item OITS Number Westinghouse status in OITS 720.432F 6161 Confirm W Response to NRC letter of 6316 Confirm W i

October 2,1997 (Related to 720.432F) j 720.443F.

6181 action N 720.454F 6250 Action N 720.455F 6251 Action N

_.=

4 8

Enclosure to Westinghouse Letter DCP/NRC1180 December 12,1997 un..g

1 U.,.

! NRC FSER OPEN ITEM ;

l HEM!

e.

~I

- Question 410.306F (OITS 6064) -

l i

Re:

L Iri Revision 14 of the SSAR, Table 10 4.51 was deleted. The staff cannot complete its review without the.

reference design information that was provided in the table.

Response

Table 10 4 51, Design Parameters for Major Components of the Circulating Water System will be

- reincorporated into the SSAR. The table title will be revised to," Design Parameters for Major Components of

, i the Circulating Water System" This will be cor.sistent with the AP600 standard plant scope as defmed in 2 SSAR I.8, interfaces for Standard Design.

2 JSSAR Revision-

-l Add attached SSAR Table 10.4.51 Also revise SSAR 10.4.51.1, General Description, as shown below:

" Classification of cornponents and equipment in the circulating water system is given in Section 3 2 The circulating water system and cooling tower are subject to site specific modification or optimization. The system described here is applicable to a broad range of sites. The C'.mbined License applicant - will determine the final

- system configuratnon. Table IM,31 provides circulating waer system design data based on a conceptual -

design,'

)

e 4

-[

s 9,.

4 4

t g

410.306F 1' g

d l.

.u,m.

.~.

NRC FSER OPEN ITEM ti!!!

t Table 10 4.5 I DESION PARAMETERS FOR MAJOR CIRCULATING WATER SYSTEM COMPONENTS (Conceptual Design)

Circulating Water Pump Quan Two per unit Flow rate (gal / min) 206.800 llead (ft) 92 Natural Draft Cooling Tower Quality One per unit Approach temperature (*F) 10 Inlet temperature CF) 114.1 t luttet temperature (*F) 87 Temperature range (F) 27 i Flow inte (gal / min) 400,000 lleat transfer (litu/hr) 4,900 x 10' Itasin storage volume (gal) 2.23 x 10'

'Vind velocity design (mph)

I10 Seismic design criteria per Uniform Building Code Predicted performance during limiting site conditions:

Outlet temp 91 wet bulb temp of 80*F (1% exceedance) 90*F 410,306F 2 W Westinghouse i

m -

2

~ iNRC FSER OPEN ITEM' vmamn 3

1 Ouestion 410339F (OITS 6194).

g p

iThe' system desc'ription, design parameters, and P&lDs are provided in Section 9.4.2. Table 9.4.21 through j

9.4.2 7, and Figure 9.4.21, of the SSAR, respectively. Table 3.2 3 of the SSAR provides the classification of"

- the Annex / auxiliary buildings non radioactive heating ventilation and air conditioning system (VXS) system and

=

components? However, Westinghouse needs to provide the following:

~

.l; updated P&!Ds -

2.

a standard safety analysis report (SS/ R) figure reflecting the updated P&lDs ahowing major

- instrumentation, and system interact.ons with other ' systems such as supply of the chilled and hot water as provided from the central chihW water system (VWS) and plant hot water system (VYS).

13? ;a_figura for the ancillary diesel generator room showing air supply to and from the mechanical equipment -

__ troom heating ventilation and air conditioning system (HVAC) subsystem.

4.

SSAR table. listing : major subsystem component parameters for,the VXS inc'm vg air handling units, supply and exhaust fans, electric unit heaters and heating coils, and ancillary du.. generator room exhaust fan

5. __ update SSAR Table 3.2 3 to include classification data for the ancillary diesel generator room exhaust fan !

and correct code data for filters and fans: _

6.

revise SSAR Section 9.4.2.2.2 to include. descriptions for humidifiers, hot' water unit heaters, and isolation

' ?

l dampers, i.

7.

revise SSAR Section 9.4.2.2.3.5 to state that "To replace the filters of an air handling unit, the unit is-stopped and isolated from the duct system by means of isolation dampers."

(

' N.

revise SSAR Section 9.4.2.2.1.3 to designate ' reactor switchgear" as " reactor trip switchgear."

9, revise SSAR Section 9.4.2.2.3.1 to state that the served areas in the annex building are maintained at a positive pressure (state pressure) with respect to the adjacent creas during filter replacement mode.

10..reviw SSAR-Section 9.4.2.1.2 to provide the design temperature conditions for the elevator machme room and boric acid batching room.

Response

't.

-VXS P&lDs (Rev, 5) will be forwarded to the NRC via separate correspondence.

. 2.

' Applicable updated SSAR figures for "XS are attached.-

--3, SSAR figurc 9 4.21 (Sheet 5 of 5) is attached and has beer. revised to show the HVAC servicing the

_. ancillary diesel

~ 4f - SSAR Tables 9.4.21 and 9.412 provide major subsystem component parameters for the VXS defense in-

depth systemsi S%ilar information requested for nondefense-in depth components is considered to be

escessise since has;c information is already provided in the texty Also, there is no reason to control this nondefense in-depth uhmation in Tier 2. For these reasons, no change to the SSAR is necessary.

~ 5.1 The exhaust fan for the ancillary diesel generator room provides ventilation for the' diesel fuel stoied in the aream The fan does not operate during operation of the ancillary diesel generator.. " Ventilation and cooling

_ for the' room when the ancillary diesel generators operate is provided by means of manually operated N

410,339F-1 g

I

$[

,--e-

-- mss y

A m

m-

-.-e-

++

.e ss W

-,.n

.NRC FSER OPEN ITEM -

iu= m ein if

':1 dampers and opening doors to allow radiator discharge air to be exhausted direct td outdoors ", as noted in-SSAR 9A.2.2.1.5 - For this reason, no change to the SSAR is necessary.

- 6? SSAR 9 4.2.2.2 will be updated to include descriptions for humidifiers, hot water unit heaters, and isolation dampers _

7.

SSAR 9 4 2 2.3.5 will be revis(d to state, _"To replace the filters of an air handling t nit, the unit is stopped -

and isolated from the _ duct system by means of isolation dampers ", as requested 8=

SSAR 9.4 2.21,3 will be revised to designate " reactor switchgear" as " reactor trip switchgear.?, as-requested.

9 The general area llVAC subsystem will maintain a slightly positive pressure during filter replacement as

, noted in the SSAR.- A positive pressure exists because the supply air flow rate is maintained greater than the exhaust air flow ate. The reason for maintaining a positive prenure in this subsystem is to provide additional assurance that controlled access areas such as the health physics area are maintained at a negative pressure relative to other areas of the building..This is a nonsafety related function. Since the health physics and hot machine shop itVAC system normally maintains a slightly negac ce pressure by design, even if the general area }{VAC subsystem is not operating, the controlled access areas would still be maintained at a negative pressure. For this reason, no change to the SSAR is necessary.

10 The design temperature conditions for the elevator machine room and boric acid batching room will be added to SSAR 9 4.2 l.2 as requested.

t SSAR Revision:

Ilowever, forward VXS P&lDs (Rev.-5) to the NRC via separate correspondence.

1.

Nre.

2.

Update SSAR figure information based on attached VXS figures 3.

Update. SSAR figure 9.4.21 (Sheet 5 of 5) to show the llVAC servicing the ancillary diesel. (Attached) 4.

None.

5 _ None-6.

IJpdate SSAR 9 4.2 2.2 to include descriptions for humiditiers, hot water unit heaters, and isolation dampers as follows:

' Just before the section entitled Shutoff, Control, Balancing and Backdraft Dampers add these three

-subsections:

"Humidoper

. The humidifier is a packaged electric steam generator type which converts water to steam and distributes it through the supply duct system. The humidifier is performance rated in accordance with ARI 6M -

. (Reference 1.1).

410.339F, 2 (N Westinghouse

NRC FSER OPEN ITEML jMluliiM m-

'Hent Waler Unit Heaters The' hot water unit heaters consist of a fan section and hot water heating coil section factory assembled as a complete and integral unit. The unit heaters are either horizontal discharge or vertical downblast type. The coil rat!ngs are in accordance with ANSI /ARI 410 (Reference 12).

. Ist>lallon Datnpers -

4 isolation dampers are bubble tight, single-orparallel-blade type. The isolation dampers have spring return actuators which fall closed on loss of electricalpower or loss of air pressure. The isolation -

dampers are constructed, qualified and tested in accordance with ANSI /AbfCA $00 (Reference 14).*

- 7.

/\\dd the following to the end of the second paragraph of SSAR 9.4 2.2.3.5: "To replace thefilters of an,

air handling unit, the unit is stopped and isolatedfrom the duct system by means ofisolation dampers.'

R.

Revise SSAR Section 9.42.2.1.3 to designate ' reactor switchgear" as " reactor trip switchgear

• 9.

None.

10: Revise SSAR Section 9 4.21.2 to add the following design temperature conditions at the end of the Normal Operation list:.

Room or Area

. Temperatures

('F)

Normal Operation

• elevator machine room 30 105 boric acid batching room 3010$*

d Z

4 a

1 T Westinghouse

F f

VXS EQUIPMENT ROOM llVAC SUBSYSTEM 4

5 From AlIth (

To AllUs f MS-02A/B qts-02A/B l

VXS 004 Ed.9t. ja VXS 004 Jk VXS UO4 Air Ilandlmg ti uspmera JL t

m Room lishause Pienum 2

^

Central AlarmStation Ik Access Corndor Exhaust Secur Room 2 MA33 9

' To A tat %

_~

I 2-bM, Vux>3 m

g y

l Rest Rooms E

Em4A#.r Access Area

__M 3f Secunty Room 1 D A' 8

}g Fnte-Q a-v/ayj

=

E Battery Rooms m

21 Hattery Charger Rooms V

^

'To iaust F s MA--

A/B, g-

{

{

RCC/Non-1E Penetration Room

/n VXS S itchge com 3 S<

IIV CSul stem E

9 y

ReactorTrip Switchgear Rooms h~

ICC/Non I-E Penetration Room V 003

/

e er

,se 3

mE o,, g me=

3 m

Sh

[

h

+ Adds n lerbe w.4 h ch;tlal waler syslern (MC) l}

hi a

j !!X fIs e

a and hobder trnhg c3.dern (Vys) 4

~

P.

h'-

r

~

VXS SWITCllGEAR' ROOM IIVAC SUBSYSTEM To AllUs F

(

From AllUs #

MS4)SA/B

)

stmos t's.< c. or MS-05A/B VXS VXS 003 to atmos VXS 003 Ik l

Exhaust um 1 b

U k

Electrical Switchgear Rooms 1 & 2 y

Ex l

E Fans hZ MA-9A/B

!i

'N g

y From Battery ooms in VXS

.I

]

. uipment R Subsystem s

- r, G

a

=,

~

a ha R

r 2

h C hi.lk( d (A)

CC SjsdhdYWS}

6

1. g hk ts Cr (=C E wi g

and hok wake bewk<< sgtem (yys) g 1

Ut

y VXS MECllANICAL EQUIPMENT AREA SUBSYSTEM h

B 4

I Upper South Air IIandling Equipment Room E

g

$ 'jI7

) Lower South Air Handling Equipment Room

/ +

Airliandling Unit" 4

4 n

a

,k o G

(

_/

n U

)

I

( mb ]N 4

Air ilandling Unit 4

[

,)

(

ArR W

h E

32 Boric Acid Batching Room gE kk2 i

a 5

5 e

v 22 D

E 4 "E 38 3EP i

Demineralized Water Deoxygenating Room

_o 3gg gj

'r l

T

-,.ti f E"

E y

h v

J

'i - P.

g a

g 3

[ $%

5 B

S 5

- 1

• Aduc, o tuQe ~.4h

~

A m. ilc. < 3 Ch;Oed wder % stem (VWSm{

1 b,b iGen<c.d I

7e

' " ~

12cey 4

> a rmer, be4 w t.c gwh uj% (qyg i

NRC FSER OPEN ITEM ir

's Queston 410342F (OITS 6197)

Re:

The radiologically controlled area v ntilation system (VAS) description, component design parameters, and PAIDs are gisen m Section 9.41, Tables 12-3 and 9 4 31, and Figure 9 4 31 of the SSAR. respectively.

Ilowever, Westmghouse needs to provide the followmg' l.

updated piping and instrumentstion diagrams (P&lDs) and esplain how the annex building exhaust is morutored for detection of high radioactivity and then diverted to VFS 2.

u SSAR figure redecting the updated P& ids for the supply and exhaust of the AABVS and FilAVS showing major instrumentation including system interactions with other systems such as the supply of chilled and hot water as provided from the VWS and VYS.

3 revise SSAR Table 3 2 3 to correct code data for coolers, tilters, and fans.

4.

in SSAR Sections 9 4,3.2.1.1 and 9 4.3 212, provide the capacity of the exhaust fans for AABVS and f il AVS 5

in SSAR Section 9 4.3 2 2, provide the code data for the RNS and CVS pump room unit coolers

& revise SSAR Tab!c 9.431 to list subsystem parameters for air handling umts, supply and exhaust fans, and electric unit heaters and heating coils

Response

1.

VAS P&lDs (Rev. M) will be forwarded to the NRC via separate correspondence. SSAR figure 9 4 31 pheets 2 and 3) presently identifies radiation monitors and that the VAS exhaust Oow is diverted to 'FS when radiation is detected. SSAR I1.5 2 31. describes how the annex buildmg exhaust is monitored for detection of high radioactivity and then diverted to VFS For this reason, no change to the SSAR is necessary.

2 SSAR figure 9.4 31 (sheets 2 and 3) will be revised to reDeet the interactions with other systems such as the supply of chilled and hot water as provided from the VWS and VYS.

1 The code data for the VAS coolers, filters, and fans presently in SSAR Table 32-3 is considered to be correct For this reason, no change to the SSAR is necessary, 4

The mformation requested is considered to be excessive since the equipment does not perform a defense-in-depth function and basic information is already provided in the text. Also, there is no reason to control this mformatwn m Tier 2. For these reasons, no change to the SSAR is necessary.

5.

Code data for the RNS and CVS pump room unit coolers will be provided as shown m item 5 under "SSAR Revision

• Lelow 6

Informahon is provided in tables for defense in-depth IIVAC sy stems. Additional basic mformation is provided m the text for other systems. The informatmn requested is considered to be excessive smee the equipment does not perform a defense-in-depth function and basic information is already provided in the test Alm there is no season to control this inform, tion in Tier 2. For these reasons. no change to the SSAR is necessarv.

WC5tlDgil0tlSe

NRC FSER OPEN ITEM

"!n

"[

r SSAR Revisjort-1.

None. Ilowever, VAS P&lDs (Rev. 8) to be forwarded to the NRC via separate correspondence.

2.

Revise SSAR figure 9.4.31 (sheets 2 and 3) to reflect the interactions with other systems such as the supply of chilled and hot water as provided from the VWS and VYS (Attached) 3.

None, j

4.

Nonc j

Revisc SS AR Section 9.4 3.2.2 by adding the following final sentence under Unit Coolers "The principal construction code is the manufacturer's standard.'

6 None.

t t

410.342F, 2

[ Westlfigh00Se

l$

t S

T E

outs,de A,,

VAS AUX / ANNEX BUILDING HVAC SUBSYSTEM 5

Auxiliary Busidang

.w g

VAS 004 h

o CCS Val re Room. Stairwell. Access Corrxior. Vestabule. VAS g-h Equipment Roosn. Personnel Access Area.VES Air

['.j rATL A 1I ujMt EUCtLu Storage / Operating Deck. Staging As(a. Security Room nab +F'ffd

• s

AIIUs [

Nf

• ~

'l /s

- MS41 A/Bi Corridor El 117*- 6', VFS Penetration Room. VFS/SFS/PSS t

g I

Annex Buildmg Staging and Storage Area.

Corndor El.100* 0". RNS Ileat Exchanger Room. Waste Monitor Contam Air Fahration Exhaust Rooms.

Rooms A & B.Madfle Annulus.Msddle Annulus Access Room.

4 Contamment AccessCorridor Personnel liasch. Mamsenance Floor Stagi.ig Area Comdor El. 82*- 6". Paping/ Valve Roum %1S 1%mp Rooms. SF3 1f Pump Roorns. SFS Ils Rooms. Demsa Faher Access Area. Rad 9

1I Chem Lab. CVS Makeup 1%mp Room. SFS Penetrauon Roorn.

m g

imwer Annulus Southeast. Lower Annulus Southwest. I.mwer g

Annulus Valve Area. Lower Annulus East. Pipe Chase. DegasiGer C.

Column Roosa. Containment isolation Valve Room o

l e.

a Corndor a a - 6. rrimary Sampic Room. RNS iv, Rooms.

5:

Denuneralizer Faher Room. WGS Equipment Room. WLS

.3!

Equipment Room. Degassner Discharge f%mp Room. Effluent

.- h IloidopTank Rooms. Aus Building Sump Room. Waste Momtor a$ %

Tank Room C.Chenucal Waste Tank Room 1f lI

• 1 n

(

E N

RE

- - - - - - + To VFS Fihered Exhaust

• l ga

{ - - - - - - y-To VFS Plant Vent nn ypS.ogg II Qg y

VFS-001 I

G

c. E,

.}.

a Exhaust Fans c

y$

3$$

y MA-02A/B

• Felsered Exhaust used when radiation is detected E

<c3 2

I h.

N,@g II b

o P.

a w $ U; g5s e

'l o.

1

(&

I

" I!!

>g 83 5d

~

nF li;!

? ?

3E

/

VAS FUEL HANDLING AREA HVAC SUBSYSTEM i

a Outside Air VAS 004 To VFS fikered Exhaust

• If

~'""'"^i" sor A rrE 4Ar.4sv5A"

-[ MS-02A/B '

A AllUs r + - -

(Y N)

I I

.I

.'t g

3f Fuel fiandling Area. Rail Car Bay / Filter Storage Area. Resin l

1

> To VFS Plant Vent

(-

g, Transfer PumpfValve Room. Spent Resin Tank Roorn. Waste s

F DisposalCentminer Area.WSS Valve, Piping Area. Elevator E

Machir Room 5.

Exhaust Fans g

MAMA /B 3!q

?9 3!!

Ea

" 3.

n 22 lk R$

?

==

v h

• littered Exhaust used when radiation is detected

[

3 = sr

[=

r S{

<E" E

4 b4 N Y

6 o wj3I 2-E b$

g 3

g n

i d

i NRC FSER OPEN ITEM jer n;j Question 410.345F (OITS 6200)

Re:

The AA11VS exhaust fans are located in the auxiliary budding at Elevation 145' 9" The supply and exhaust duct. Save isolation dampers During normal operation, the subsystem's exhaust is unfiltered and directed to the plant vnt for discharge During high radiation isolation mode, the normal unfiltered ventilation subsystem is isolated from the affected zone when high airborne radioactivity is detected, and the isolated area is exhausted through the VFS to the monitored plant sent. The VFS exhaust fans prevent unfilter6d airborne releases by mamtaining these areas at a slight negative pressure with respect to the outside environment and adjacent clean plant areas llowever, Westinghouse needs to state in specific terms exactly what negative pressure is required to be maintained in these areas with respect to the outside environment and adjacent clean plant areas Rasponse; The VFS system will maintain a slightly negative pressure with respect to the outside environment and adjacent clean plant areas as noted in the SSAR when high radiation is detected A negatise pressure is maintained by,

nr.mtoring the differential pressure between the high radiation area in the auxiliary building and the outside environment and modulating a dampe to the VFS exhaust Al{U. The reason for maintaining a negative pressure is to ensure that airborne radioactivity is filtered rather than released to surroundmg areas in an uncontrolled fashion The value of slightly negative pressure" is not critical as this is a nonsafety-related functian For this reason, no change to the SSAR is necessary The norninal setpoint value for the differential pressure is approximately 0.15 in. II,0.

SSAR Revision:

None-W Westinghouse

~

=,

NRC FSER OPEN ITEM-.

qMstin

- 1 i

- Question 4101'2F (OITS 6207)t Re:

Ench VFS supply and exha'ust air subsystem fans conform to ANS!/AMCA 210-85," Laboratory Method of-iTesting fans for Rating Purposes," ANS!/AMCA 21185

• Certified Ratings Program Air Performancei and ANS!/AMCA 300 85," Reverberant Room Method of Testing Fans for Rating Purposes? The chilled water; (cooling coils and not water heating coils are designed and rated in accordance with ASHRAE 33 78,?Mithod of i

> Testing for Rating Forced Circulation Ait Cooling and Air lleating coils." and ANS!/ARI 410 91," Forced -

Circulation Air Cooling and Air IIcating Coils? The VFS supply air subsystem airnow is measured and balanced in accordance with SMACNA 1983,"l!VAC Systems Testing, Adjusting and Balancing? The low l

efficiency (25 percent) and high efficiency (80 percent) filters are rated to dust spot efficiency based on' L ASHRAli 52 76," Method of Testing Air Cleaning Devices Used in General Ventilation for Removing Particulate Matter," and meet Class I construction criteria of UL 900.86, " Test Performance of Air Filter Units? -

1 The exhaust air subsystem high efficiency particulate air (l! EPA) filters and charcoal adsorbers are constructed and tested to conform with ASME N5091989,

• Nuclear Power Plant Air Cleaning Units and Components," and

- ASME N5101989. " Testing of Nuclear Air Cleaning Systems" and Regulatory Guide (RO) 1.140 1979,_

Revision _1," Design, Testing. and Maintenance Criteria for Normal Ventilation Exhfust Air Filtration and Adsorption Units of Light Water-cooled Nuclear Power Plants? ~ llowever, Westinghouse needs to reference i r

Revision 1 of RO I,1401979 in SSAR Section 9 4,13 and provide this reference to all HVAC Subsections including 9.4.1 and 9.4.7.

' Response:

'The SSAR will be revised to provide a reference to Revision i of RG l.140,1979 for sections 9.4.1, VBS and

~

9.4.7, VFS Westinghouse does not consider that RG l.140 is applicable to the other SSAR sections.

SSAR Revision:

Add-to the end of C9AR 9 4.13 References the following:

- 30. "Iksign Testing, and Maintenance Criteriafor Normal Ventilation Exhaust Air Filtration and

. Adsorption Units of Light-Water Cooled Nuclear Power Plants," Regulatory Guide (RG) 1.140 1979,

. Revision I, Replace the term " Regulatory Guide 1.140* with the term ! Regulatory Guide 1.140 (reference 30)* in SSAR

. sections 9 4.1.1.1, 9,4.1.2 2, 9.4.1.4, 9 4.7.2.2 (two places), 9.4,7A and 9 4.12.

+

2 5

.p d

4 M

a li I

i NRC FSER OPEN ITEM i

7

- Queston 410.356F (OITS 6211)

[

The system description and components classi0 cation are given m SSAR Section 9.4.8 and Table 3.2 3. As identified in Table 3 2 3 of the SSAR, the VRS components are non-nuclear safety class, non seismic category, and the quality assurance requirements of Appendix il to 10 CFR Part 50 do not apply. Ilowever, Westinghouse needs to (1) revise SSAR Table 3.2 3 to provide code data for the exhaust air system's exhaust fans and high and low ef0ciency filters to reficct the information in the text of SSAR Section 9.4.8.2 2. (2) l provide exhaust fsn rating data in the text of SSAR Section 9.4.8.2.1,(3) provide a SSAR figure showing major instrumentation, and interactions with other systems such as chilled and. hot water' supply as provided from the

.VWS and VYS,'and update the VRS p&lD (i.e. update M6 001,002, and 003), and (4) update SSAR Table -

0.7 2 to list the VRS Ogure requestad in (3) above.

[

1

Response

1.

SsAR Table 3.2,3 p rsently includes the requested data.

2 The information regt ested is considered to be excessive since the equipment does not perform a defense in!

~ depth function and b sic information is already provided in the text < Also, thers i o reason to control this information in Tier 2 For these reasons, no change to the SSAR is necessary.-

3 A figure for VRS is attached and will be added to the SSAR.

4.

SSAR Tabic 1.7 2 will be revised as shown in item 4. of 'SSAR RevisionP below.

SSAR Revision:

1.

None.

2.

None.

3; i a.

Add attached VRS figure to SSAR as Figure 9.4.81.

b.

Add the following statement under SSAR 9,4,8.2 System

Description:

J The rodwatre building HVAC system is shown in Figure 9.4.81.*

4, Revise VRS information in SSAR Table 1.7 2 (Sheet 3 of 3) as noted; i

" Table 1.7 2 (Sheet 3 of 3)

AP600 SYSTEM DESIGNATORS AND SYSTEM DIAGRAMS Dedgns:or:

System SSAR Section SSAR Figure VRS=

• Radwaste Building ilVAC System 9.4.8 N/A 9.4.8-l" 410,356F-1 g

..n u

A

.W'

.VRS RADWASTE BUILDING HVAC SYSTEM

~

- Outside Air.

4 m

gg3 f

~

g

' kU)

MS-01 A/B'_ m (vosa Package Waste Storage Room, Waste Accumulation Room.

RE gwn...

Electrical / Mechanical Equipment Room, HVAC Equipment

' Room, Mobile Systems Facility

.y

> To VFS Plant Vent Unit Unit Unit Unit

. I feater

• Ileater*

Heater

• Heater

• gg p MA-02A/B Ec s h ' Ua:4 He<Aer infeches

[

u A h 4he. bot oder henPg s3ss em (vys3 I'.gure ' ci,+.g._ [

6-

= 6du254e. Mki'ine NVM -

. g_

.sss w hEf).Y NbOO.1, -OOL,. 00'

NctC FSER OPEN ITEM

. Question 410 358F (OITS 6213)

Re.

Durmg a meetmg on December 12 through 14, 1994, Westmghouse deferred the response to the staffs concern regardmg the design parameters for system components and piping, and instrumentation diagram, and classification of the VTS system end components. Westinghouse needs to (1) revise SSAR Table 12 3 to provide data for exhaust sentilators and correct code data for fans and dampers to correspond with the text of SSAR Section 9.4 9.5,(2) provide a SSAR figure referencing and reflecting the updated VTS piping and mstrumentatmn diagram showing interactions with other systems such as the supply of chilled and bt water which is provided from the VWS and VYS, respectively. (3) revise SSAR Table 1.7-2 to reference SSAR Figure 9.4 9 and (4) revise SSAR text and tables as appropriate to show design data for exhaust ventilators for the general area sentilation subsystem and local arer. heating and ventilation subsystem and exhaust fans for local area heating and ventilation subsystem Response-Table 3 2 3 will be revised to mdicate the ifincipal construction code for " fans, ductwork" as I

a "SMACNA" in lieu of " Manufacturer Std."

h inhaust ventilators will not be added to table 3 2 3 as the informatmn requested is considered to be escemve since the equipment does not perform a defense in-depth function and basic information is already provided in the text. Also, there is no reason to control this information in Tier 2. Far these reasons, no change to the SSAR is necessery.

2.

A figure for the VTS general area subsystem is attached and will be added to the SSAR.

3 SsAR Table 1.7 2 will be revised as shown in item 3. of "SSAR Revision:" below.

4 The informatmn requested is considered to be excessive since the equipment does not perform a defense m-depth function and basic informatmn is already provided in the text. Also, there is no reason to control this information m Tier 2. For these reasons, no change to the SSAR is necessary.

SSAR P,avision 1.

a Change the "prmeipal constructwn code" designation for the " Turbine Building Ventilation System (VTS)* " fans, ductwork* from " Manufacturer Std." to "FMACNA" in SSAR Table 3.2 3 (presently sht.

61 of 67 in SSAR Rev.17).

b None.

2 a Add attached VTS figure to SSAR as Figure 9.4.9-1.

h Add the followmg statement under SSAR 9 4.9 2 System

Description:

"The Turbine Building 11l'AC System General Area Subsystem is shown in Figure 949-1,"

m8M W Westinghouse

NRC FSER OPEN ITEM 3.

RcVise VTS information in SSAR Table 1.7 2 (Sheet 3 of 3) as noted:

" Table 1.7 2 (Sheet 3 of 3)

AP600 SYSTEM DESIGNATORS AND SYSTEM DIAGRAMS Designator System

$SAR Section SSAR Figure VTS Turbine Isuilding IIVAC System-9.4.9 N/AP.4.v l*

4 None-d 410.358 F.-2 3 Westilighouse

l VTS TURBINE BUILDING HVAC SYSTEM GENERAL AREA SUBSYSTEM Outside Outside Outside.

Exhaust Exhaust Exhaust -

4k Ak Ak Outside Exhaust j( --

' Roof Roof Roof Exhaust Exhaust Exhaust I

Ventilators Ventilators Ventilators -

Roof Unit Exhaust 11 eaters' Jk Grating Ventilator

- Outside Unit Outside Air d(

lleaters#

Grating 4

./

Air Ak l

Outside Unit Air J\\

lleaters*

Grating 4-.fgtsik.

j(

Air Outside Unit r

Air d(

lleateof Ou ide 1

E

• Eo.cbua.4 Acaler Mter0,ces

, % u a.kr M t aj

%"* C' i S -I o

u.....m.&rrr a

-

y;.

s1s%(y g)

%b.aeiL;g,ng tw ac.gA ry i

' 7i (RCF) VTs 001 F:

e8 2

=m

-a------ - - - - - -.. - - - -

NRC FSER OPEN ITEM Queston 410 362F (OITS 6217)

Re:

As identified in Table 3 2 3 of the SSAli, the VliS components are non nuclear safety class and non seismic category As such, the quality assurance requirements of Appendix B of 10 CFR Part 50 do not apply. The sy stem description is gisen in SSAR Section 9 4.11. Westinghouse needs to revise SSAR Table 3.24 to list the correct code data for the air handling units with filters as

• Manufacturer Standard, UL 900 and ASilRAE $2" and list separately the supply and exhaust fans with their code data as " ANSI /AMCA 210,211, and 300."

Additional,, Westmghouse needs to provide a SSAR figure referencing and retlecting the updated VIIS piping and instrumentation diagram. The figure should show interactions with other systems such as a discharge to plant sent via VFS, and how the supply of the chilled and hot water is prosided from the VWS and VYS, respectisely; SSAR Table 1.7 2 should be revised to reference the SSAR figure; and provide a SSAR table showing major components (such as supply air handling units and supply and exhaust fans) design data.

Response

o I.

a.

SSAR Table 3 24 will be revised to list the correct code data for the air handling units with filters as '

' Manufacturer Standard, UL 900" UL 900 is sufficient for the filters ASIIRAE 52 need not be listed consistent with the way other filters are listed in the table.

h.

The supply and exhaust fans are already covered in Table 3 2-3 under "SMACNA" for fans and ductwork.

2.

A figure for VilS is attached and will be added to the SSAR.

1 SSAR Table 17 2 will be revised as shown in item 3. of "SSAR Revision:" below.

4.

A SSAR table showing major components (such as supply air handling units and srpply and exhaust fans) design data will not be provided. The information requested is considered to be excessive since the equipment does not perfunn a defense in depth function and basic information is already provided in the text. Also, there is no reason to control this mformation in Tier 2. For these reasons, no change to the SSAR is necessary.

SSAR Revision:

1.

a.

Change the "ptmeipal construction code" designation for the "licalth Physics and Hot Machine Shop IIVAC System (VilS)"

• air handling units with filters' from "UL 900" to

• Manufacturer Std., UL 900*

in SSAR Table 3.2-3 (presently sht. 60 ot. 7 in SSAR Rev.17).

b None.

2..

a Add attached VIIS figure to SSAR as Figure 9.4.11 1.

b, Add the following statement under SSAR 9 4.11.2 System

Description:

• The health physics and hot machine shop flVAC system in shown in Figure 9.4.11-1.*

m 362 M W Westinghouse

1 +

a

.~o e

-a-na.

- - - - = -

s n -us

~s--

r

~_

=

NRC FSER OPEN ITEM uit:

'it!

11:

3: Revise VIIS information in SSAR Table 1.7 2 (Sheet 3 of 3) as n'oted:

" Table 1.7 2 (Sheet 3 of 3)

. AP600 SYSTEM DESIGNATORS AND SYSTEM DIAGRAMS i

Designaler System SSAR Section SSAR.I'igure -

VilS.

Ilealth Physics and flot Machine Shop HVAC 9.4.1 ! =

NJA9.4.II l*

System 1

4.

None.

k l

1 s

i-.

N 4

~ 410,362F. 2 W Westinghouse 7'

VHS HEALTH PHYSICS AND HOT MACHINE SHOP HVAC SYSTEM Outside Air W der do.hugjt,o -

w<.bGtled lE Ch;lleg5 (y ys,; -

AIIUs

" 6fW5)

. MS-01A/B L HEALTH PHYSICS Radiation Monitor Calibration Room, HP y

Office, Non-RCA Entry / Exit Area, RCA Entry / Exit Area, Decon Room,llP Counting Room, OfTice I

> To VFS lf

(,.

Plant Vent HOT MACHINE SHOP Jk JL if' >

Pump Seal Shop, i1ot Machine Shop Exhaust Fans MA-02h/B f>.

Ak r=

lli Efliciency MachineTools Filter Exhaust Fan (intermittent

-k operation)

,6 Reyre.4.'t.1, -I g

lbalh Pn siaand tioF

=.

(nac h.ne bcp HVAC Q

C.9 teen (GeF)V tis co l, co 4 co3

't

7 NRC FSER.OPEN ITEM

-j

- )

6 Question 420.126F (OITS. 6221) l LCO 3.3.2, ESFAS "

The NRC Instrumentation and Controls BrancS received a letter from _Westir ghouse dated -

November 5,1997, (NSD NRC 97 5415) which contained mark-ups related to the AP600 I&C =

technical specifications (TS). "These mark upa should be incorporated into the next revision of the i

technical specifications. In addition, the mark-ups did not completely address the incousistent use of-

" channel" versus"channet/ division." (e.g., ESFAS Condition "V" and Required Action "V.1");

. Westinghouse should confirm that all cases of this inconsistency ilave been identified and corrected;

Response

ESFAS Coadition V_ applies to Functions 2.b,9.b,10.b, and ll.d. which only ipecify operability of instrumentation ' channel." Therefore, " division" will be deleted from the Required Action, Basis

- discuuion (no change is required for the LCO Required Ac on V.1).

The NRC identified correction (LCO 3.3.2, ESFAS, Condition V) will be incorporated in the next revision of the AP600 Technical Specifications as shown in the attached markup of the Condition V Bases insert previously transmitted by NSD NRC-97-5415.

j This response, together with the markup in NSD-NRC 97-5415, provides resolution of all known inconsistencies.

SSAR Revision: See attached markup.

420.126F 1

1 REQUIRED ACDON CORETlON DME

/-

CCh0lTIONj'

/

/ esWe N rioperatie channel (s).

/

R 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> V,

Requred Mean and VT assocmIId Completon Time

/-

/

of Tade 3 321 spected

/QB-

/

' Coperton not met -

/

j

/

/

V2.1 Se o MODE 5 180 s

,/

g

/,

j t

/

V22 Inicater ac$on to open me RCS j' 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br />

./

\\

ary estatsch a viatdo level ej

)

pressunzer.

,/

I l

i J

Revised Bases,msort on page B 3.3114 f

~~

r-4G V.1 V 21 and V 2P 760 of if me R od Acton and N assoo Trne of the Arat Conditon listed in Tatdo 3.3 21 is met the requesd channel (s)(divoong not bypassed wein 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, he inoperatie Mannel(s) pyvmon(0 must be restored wen 165 hour0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> The 168 hos Completon Trne is based on the abdity of the two pt)temaming OPERABLE man sqr devig to provide the protectNo Functon even mei a angle fadure.

If the channel (s) arvoo3n is not resWed wthm the 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> Complebon Time, the plant and be placed ti

(^.L a conditon m anicn the hhehhood and consequences of an evoet are minmzod. This is accomplianed by i.g - plaang me plant in MODE 5 wtwi 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> tem next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). Once ti MODE 5, acan snan be citated to open se RCS and estatsen a vmble level m Ine pressunter. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable Ome to reach MODE 5 from MODE 4 wti RCS coohng provided by me RNS (approxrnately 350*F) e an orderty manner

. wthout madengmg piant systems.

/

Openeg the RCS pressure boundary assures eat coolmg water can be miected..*out A

/

' the RCS to proves a cable level in the pressunzer memzes be consequences of a loss of decay heat

/

removal overt Rg04fd PEl offd IT&*n

%20./2G F

/2/17 AP600 Tech Specs DRAFT October 29.1997

Question 470.45F (OITS. 6222)

Adm nistrative Controls. Section 5.7 liigh Radiation Area The timergenc) Preparedness and Radiation Pro' action tiranch has identified two places in Technical Sp;cincation 5.7 (Iligh Radiation Area) where symbols associated with dose rate limits were omitted.

A mark up copy of Section 5.7 with the miss;ng symbols is provided in Enclosure 2. Section 5.7 should be revised to incorporate these correctiot.s.

Responset The identified cortections will be incorporated in the next revision of the AP600 Technical Specifications as shown in the attached mark up of pages 5.0 22 and.23.

SSAR Revision: See attached raaikup.

I 1

i R

470.45F 1 W Westinghouse

High Radia%1on Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20. paragraph 20.1601(c). in lieu of the requirements of 10 CFR 20.1601. each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is

> 100 mrem /hr but < 1000 ares /hr. shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Individuals qualified in radiation protection procedures (e.g.. (Health Physics Technicians)) or personnel continuously escorted by such individuals may be exempt from ths RWP issuance requirement during the performance of their assioned e

duties in high radiation areas with exposure rates (ion protection 1000 mrem /hr.

provided they are otherwise following plant radiat procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more' of tt)e following:

a.

A radiation monitoring device that continuously indicates the radiation dose rate in the area, b.

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring' device may be made after the dose rate levels in the area have been established and personnel are aware of

them, c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the (Radiation ProtectionManager)intheRWP.

(continued) h AP600.m,,o,a mi 5.0 22 08/97 Amendment 0 ooi=-

o

.___________-_l___-______-____ - _

..e

...e..

5.7 5.7 High Radiation Aren (continued) 5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation leve1Q1000 mrea/hr shall be provided with locked or y

continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision.

Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the innediate work areas and the maximum allowable stay times for individuals in those areas.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) contir.uous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.7.3 For individual high radiation areas with radiation levels of

> 1000 aree/hr, accessihie to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the, individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device, fj0,,,_,

5.0 23 08/97 Amendment 0

NRC FSER OPEN ITEM A

Ouottion 410 341F (OITS 6196)

Pa.

Mechamcal liquipment Areas llVAC Subs) stem The subn stem serves the ancillary diesel Fenerator (DO) toorn, the demineralised deoxygenating room, boric acid batching / transfer rooms, and upper and lower south air handhng equipment rooms in the auxihary building. The subsy stem maintains served areas at a slightly positive pressure with respect to the adjacent buildings by a constant volume of outside air. The sut-systern is not credited for plant abnortnal conditions lloweser, Westinghouse needs to state in specific terms euctly what positne pressure is required to be maintained in these areas with respect to the outside environment and adjacent plant areat

Response

The mechanical equipment areas llVAC subsystem is designed to maintain a shghtly positne pressure as noted in the SSAR A positne pressure exists because the supply air flow rate is maintained greater than the exhaust air now rate. The teason for maintaining a positise pressure in this subsystem is to proside additional assurans that controlled access areas such as the health physics area are maintained at a negative pressure relatne to oth' r e

areas of the building. This is a nonsafet) related function Smce the health physics and hot machine shop ilVAC s.vstem normally mamtatns a slightly negatae pressure by design, even if the mechanical equipment areas llVAC subn stem is not operatmp, the controlled access areas would still be maintained at a negaine preuure For this reason, no change to the SSAR is necesu.ry SSAR Reesion None t

99 4

. 41 F 1 W weeneouse 8

y!MW NRC FSER OPEN ffEM r

Questk>n 480.1088F (OITS. 6213) i LCO 3.6.4 Containment Pressure The LCO 3.6.4 liaws provides only a reference to SSAR Section 6.2

• Containment Analysis " and does not provide the results of the contain nent analyses. His is unacceptable and needs to be i

corrected following the staff's review and acceptance of the analyses methods. A place holder with reference to the current SSAR revision number should be provided, for example [44.8 psig. SSAR Rev.13). His is considered to be an open item.

Response

The value for the containment peak pressure will be adoed to the AP600 LCO 3.6.4, Containment Pressure, Bases. Brackets are not considered necessary, as with all information in the Bases, if changes are made which affect the Bases, then revisions will be required and made at the time the changes are made.

The second sentence of the second paragraph in the Applicable Safety Analyses section will be replaced with the following AP6(X). specific version of the STS Bases statements.

"This resulted in a maximum peak pressure from a steam line break of 44.8 psig."

SSAR Revision: See attached markups, i

i i

480,1088F 1 g

PCS Operating 3.6.6 3.6 6 #AINMENT SYSTEMS

9. ;.1 %Ive Contairment Cooling System (PCS)

Operating

[

2 0 3.6.6

'fhe passive containment cooling system shall be OPERABLE.

f APPLICABILITY:

H0 DES 1. 2, 3. and 4.

ACTIONS CON 0! TION REQUIRED ACTION COMPLETION TIME A.

One passive A.1 Restore flow path to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> containment cooling OPERABLE status, water flow path inoperable.

B.

Water storage tank B.1 Restore water storage 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not tank to OPERABLE status, within limit.

Water storage tank volume not within limit.

C.

Required Action and C.1 Be in H00E 3, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Conditions A AND or B.

C.2 Be in H00E 5, 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> LCO not met for reasons other than A or B.

h AP600 3.6 13 08/97 Amen @ent 0 ov. - m m a m i,

3.6.6 SURVEILLMCE REQUIREMENTS SURVEILL.ANCE FP/TENCY SR 3.6.6.1 Verify the water storage tank

.... NOTE....

temperature 3 40 'F and 5 120'F.

Only required when the ambient temperature is I

< 32'F or E100'F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7 days SR 3.6.6.2 Verify the water storage tank volume t 531,000 gallons.

o 1

SR 3.6.6.3 Verify each passive containment cooling 31 days system, power operated, and automatic valve in each flow path that is not locked, sealed, or otherwise secured in position, is in the correct position.

(continued) b AP600 3.6 14 08/97 A-nt 0

3.6.6 l

SURVEILLANCE REQUIREMENTS (continued) j SURVEILLANCE FREQUENCY SR 3.6.6.4 Verify each passive containment cooling 24 months system automatic valve in each flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.6.5 Verify the air flow path from the shield 24 months building annulus inlet to the exit is unobstructed and, that all air baffle sections are in place.

Y10 years SR 3.6.6.6 Verify passive containment cooling

[

system flow and water coverage performance in accordance with the Systes Level Operability Testing Program.

5 fir S'i~

Rerumui N

cr too./of 6F M) 3.6 15 08/97 Amendment 0

,,,M$n.e.,. u,

B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions asstmed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of transients which result in a negative oressure.

Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived from the operating band of conditions used in the containment pressure analyses for the Design Basis Events which result in internal or external pressure loads on the containment vessel.

Should operation occur outside these limits, the initial containment pressure would be outside the range used for contairment pressure analyses.

APPLICABLE Containment internal pressure is an initial condition used SAFETY ANALYSES in the DBA analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered, relative to containment pressure, are the Lt)CA and SLB, which are analyzed using computer pressure transients. The worst case SL5 generates larger mass and energy release than the worst case LOCA. Thus, the SLB event bounds the LOCA event from the containment peak pressure standpoint (Ref. 1).

pt3

,,,,= g,gf,g,

/

The initial pressure contairnen useddn the containment analysis was 15.7 psia (1.0 ps-

.[: his resulted in -the.o-ir.dicated in=

maximum peak pressure from a Q referer.ce 1.

The containment analysis (Ref. 1) shows that the maxista peak calculated contairment pressure. P,. results from the limiting SLB. The maximum containment pressure resulting from the worst case IDEA s not exceed the containment design pressure 45 psig.

SLS c ff h) 17 9 t

+46.16 BBF (continued)

,, g,$ L,,,,,,

B 3.6 24 08/97 Amendment 0

8 3.6.4 i

BASES i

i APPLICABLE The containment was also designed for an external pressure SAFETY ANALYSES load equivalent to 3.0 psig. The limiting negative pressure j

d (continued)-

transient is a loss of all AC power sources coincident with extreme cold weather conditions which cool the external surface of the containment vessel.

The initial pressure l

condition used in this analysis was 0.2 psig. This resulted in a minimum pressure inside containment, as i

illustrated in reference 1, which is less than the design i

load. Other external pressure load events evaluated j

include:

{

Failed fan cooler control l

Ma1 functional of containment purge system L

Inadvertent Incontainment Refueling Water Storage Tank (IRWST) drain Inadvertent Passive Containment Cooling System (PCS) actuation Inadvertent Containment Spray System actuation l

Since the containment external pressure deign limits can be i

met by ensuring compliance with the initial pressure L

condition,NUREG1431LCO3.6.12,VacuumReliefSystemjs not applicable to the AP600 containment.

4 L

Containment pressure satisfies Criterion 2 of the NRC Policy Statement.

LCO Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of i

a DBA, the resultant peak containment accident pressure will i

remain below the containment design pressure.

Maintaining

~

containment pressure at greater than or equal to the LCO lower pressure limit ensures that the containment will not i

L exceed the design negative differential pressure following negative pressure transients.

APPLICABILITY In MODES 1. 2, 3, and 4, a 08A could cause a release of radioactive material to containment.

Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analyses are

- maintained, the LCO is applicable in MODES 1, 2, 3. and 4, (continued) k AP600 8 3.6 25~

08/97 Amendment 0

_. m.o = u n w. m.m a t t

_=

j BASES APPLICABILITY In MODES 5 and 6 the probability and consequences of these (continued) events are reduced due to the 3ressure and temperature i

limitations of these MODES. Tierefore, maintaining containment pressure within the limits of the LCO is not required in MODES 5 or 6.

ACTIONS AJ When containment pressure is not within the limits of the LCO, it must be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the time required to restore the containment to within limits, the conservative assumption of the containment analysis and mirar pressure deviations expected during normal operation.

B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be l

broughttoMODE5wheretheprobabilityandconsequencesion an event are minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are l

reasonable, based on operating experience, to reach the recuired plant conditions from full power conditions in an orcerly manner and without challenging plant systems.

l SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was develo*! based on operating experience related to trending of bot 1 containment pressure variations during the applicable MODES.

Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications available in the main control room, including alarms, to alert the operator to an abnormal containment pressure condition.

REFERENCES 1,

AP600 SSAR, Section 6.2, " Containment Analysis."

8 N m,or.

8 3.6 26 08/97 Amendment 0

B 3.6.5 8 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperature BASES BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core i

following a Design Basis Accident (DBA). The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB).

The containment average air temperature limit 's derived from the input conditions used in the containment functional analyses and the containment structure external pressure analyses.

This LCO ensures that initial conditions assumed in the analysis of containment response to a DBA cre not violated during plant operations.

The total amount of energy to be removed from containment by the passive containment cooling system during post accident conditions is dependent upon the energy released to the containment dpe to the event, as well as the initial containment temxrature and pressure. The higher the initial temperature, tw more energy that must be removed, resulting in higher peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis.

Operation with containment temperature in excess of the LCO limit violates an initial condition assumed in the accident analysis.

APPLICABLE.

Containment average air temperature is an initial condition SAFETY ANALYSES used in the DBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. The limit for containment average air temperature ensures that operation is maintained within the assumptions used in the DBA analyses for containment (Ref. 1).

The limiting DBAs considered relative to containment OPERABILITY are the LOCA and SLB.

The DBA LOCA and SLB are analyzed using computer codes designed to predict the

+

resultant containment pressure transients.

No t m DBAs are (continued)

,g6g B 3.6 27 08/97 Amendment 0

B 3.6.5 g

% w - P r *

M M

/d

, : c.

BASES assumed to occur skmultaneoudy or consecuti ely.

The APPLICABLE 4

SAFETY ANALYSES postulated D8As are analyzed with regard to ngineered (continued)

Safety Feature (ESF) systems. assuming the loss of one tF bus which'is thy worst case single active failure.

~ ~ ~ ~

N le 1E % t

1. resu,lting in one tcai,a --d ;f ti. Ce,t;icct S ey syst-.

\\

~

f> 4 L - k b n, W b,

g.., u..s a --...beinfren5 red 1nokabieI ~

~ ~ " 7 e.

.a w.~ e s a. e....

7

• LO ' A '

i Q_- (iWfd hp The limitinh;c,.r the maximum peak containment air.

i GbM fo l

temperature r vi e 6." ed OLO. The initial containment average air temperature assumed in the design basis analyses (Ref. 1) is 120*F.

This resulted in a maximum containment air temperature ;; Lot,etG in d N,E !

-i

, ;;.;; 1. %

i :gM

.l

^

T temperatu limit u used t establish the enyironmental qua fication rating enulope r containment. NThe

. g f\\

provi, peak co gnment air tempereture variation with tim maxi the basis (or environmental qu(lification envelope ltJL -

to ensur the perto e of safety related equipment ins (de N

containee (Ref. 2).

The temperature limit is also used in the depressurization analyses to ensure that the minimum pressure limit is maintained following an inadvertent actuation of the Passive Containment Cooling System (Ref.1).

The containment. pressure transient is sensitive to the initial air mass in containment and, therefore, to the initial containment air-temperature. The limiting DBA for establishin_ g the maximum peak containment internal pressure is a SL8.

The temperature limit is used in this analysis to ensure that in the event of an accident the maximum containment internal pressure will not be exceeded.

Centainment average air temperature satisfies Criterion 2 of

.the NRC Policy Statement.

The containment pressure transient is sensitive to the initial containment air temperature. The temperature limit is used in this analysis to ensure that in the event of an

~

accident the maximum containment internal pressure will not be exceeded.

i Containment-average air temperature satisfies Criterion 2 of the NRC Policy Statement.

(continued) h AP600 B 3.6 28 08/97 Amendment 0 m.oso.nmmn.ror.ums.

a-

_,u._..2__._,

u-_.__,_

l INSERT A Note the following insert attempts to follow the STS to the degree gxmible and consistent uith design process for safety related equipment.

l 1he DB A temperature uansients are used to establish the environmental quahfication operating envelope for containment. The basis of the containment environmental qualification temperature envelope is to ensure the Ivrformance of safety related equipment inside contairunent (Ref.2). The contairunent vessel design teroperature is 280 *F The contairunent vessel temperatare remains below 280 T for DB A's.

Therefore it is (oncluded that the calculated transient containment uit temperature is acceptable for the Dil A's.

4 e

BASES (continued)

LCO During a DBA, with an initial cor+.ainment average air temperature less than or equal to the LCO temperature limit, the resultant peak accident temperature is computed to remain within acceptable limits.

As a result, the ability of containment to perfors its design function is ensured.

APPLICABILITY In H0 DES 1, 2, 3 and 4 a DBA could cause a release of radioactive material to containment.

In H0 DES 5 and 6, the p-obability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

i Therefore, maintaining containment average air temperature within the limit is not required in MODE 5 or 6.

ACTIONS Aj When containment average air temperature is not within the limit of the LCO, it must be restored to within its limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

This Required Action is necessary to return operation to within the bounds of the containment analysis.

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is acceptable considering thg sensitivity of the conservative analysis to variations in this parameter, and provides sufficient time to correct minor problems.

B.1 and B.2 If the containment average air temperature cannot be restored to within its limit within the required Completion Time, the plant must be brought to MODE S where the probability and consequences on an event are minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an order'.y manner and without challenging plant systems.

(continued)

$.361.....m.,

s 3.6 29 08/97 Amendment 0

.,a-,,,

~-

= _ - _ _ _

BASES (continued)

SURVEILLANCE SR 3.6.5.1 REQUIREMENTS Verifying that the contairment average air temperature is within the LCO limit ensures that containment operation remains within the limits assumed for the containment analyses.

In order to determine the containment average air temperature, a weighted average is calculated using 1

measurements taken at locations within the containment selected to provide a representative semple of the associated containment atmosphere.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this Surveillance Requirement is considered acceptable based on observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of containments).

Furthermore the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is considered adequate in view of other indications available in the main control room, including alarms, to alert the operator to an abnormal containment temperature condition.

REFERENCES

1. AP600 SSAR. Section 6.2. ' Containment Systems.*

2.

10 CFR 50.49 ' Environmental Qualification of Electrid Equipment Important to Sefety for Nuclear Power Plants.*

.. [ b a m. m. m..,

B 3.6 30 08/97 Amendment 0

B 3.6.6 O

8 3.6 CONTAINMENT SYSTEMS 8 3.6.6 Passive Containment Cooling System (PCS)

Operating to cM se M/k*H W5

}

ThePCSprovidescontainmentcoolingtolimitIbstaccident BACKGROUND pressure and t@Myre in containment to lesu than the

-design values.

Re9n, ion of containment pressore reduces l

the release of fin ion product radioactivity face 4

containment to the environment, in the event of a Design Basis Accident (08A). The Passive Containment Cooling C 38 System is designed to meet the requirements o @f

' Containment Heat Removal

• and @C 40 'Testi o

I Containment Heat Removal Systems' (Ref.-1).

The PCS consists of a 531.000 gal cooling water tank, four-headored tank discharge lines with flow restricting i

orifices, and-two separate full capacity discharge flow paths to the containment vessel with isolation valves, each cepable of meeting the design bases. Algae growth is not-expected within the PCCWST: however, to assure water clarity l

is maintained, a prevailing concentration of hydrogen e

l Deroxide is maintained at 50 ppe.j.The PCS valve room gg.

temperature must not be below Tree:ing for an extended M)8&d period to assure the water flow path to the containment shell is available. The isolation valves on each flow path are powered froni a separate Division.

Upon actuation of the isolation valves, gravity flow of L

water from the cooling water tank (contained in the shield i

building structure, above Q containment) onto the upper portion of the containment shell reduces the containment pressure and temperature following a 08A.

The flow of water 1

to the containment shell surface is initially established to assure that the required short term containment cooling requirements following the postulated worst case LOCA are achieved. As the decay heat from the core becomes less with time, the water flow to the containment shell is reduced in two steps.- The change in flow rate is attained without active componentr. in the system and is dependent only on the i

decreasing water level in the elevated storage tank.

In order to ensure the containment surface is adequately and-effectively wetted, the water is introduced at the center of the containment done and flows outward.

Weirs are placed on l

l (continued) 1 L

f HAP 600 8 3.6 31 08/97 Amendment 0

,o

.m

,mn

,.am,

-i i-

i 1

I

- 480.1005F (iii) & (ivi i

INSERT BASES B 3.6.0 PAGE B 3.6 31 '

BACKGROUND 1

The recirculation pumps and heater provide freeze protection for the passive.

}

containment cooling water storage t.ank. However, OPERABILITY of the tank is assured by compliance with the temperature limits specified in SR 3.6,0.2 i

and not by the recirculation pumps and heater, in addition to the recirculation 1

pumps and heater, the PCS water storage tank temperature can be maintained 1

within limits by the ambient temperature, the large thermalinertia of the j

tank. H

,, by heat from other sources.

er l

e Y

4 9

i k

l i

i A

i

?

t I

1 i

l

rw vpereung l

8 3.6.6 1

BASES 8ACKWl0VW the done surface to distribute the w.1ter and ensure (continued)-

effective wetting of the done and vertical sides of the containment shell. The monitoring of the containment surface through the Reliability Assurance Program (RAP) and the Inservice Testing Program assures surface containment i

does not unacceptably degrade containment heat removal i

performance. Contamination can be removed by PCS actuation or by using coating vendor cleening procedures, The >ath for the natural circulation of air is from the air i

intates in the-shield building, down the outside of the.

baffle, up along the contairment shell to the to exit-in tto shield building and is always open. p, center The d.ains i

in the upwr annulus region must be clear to prevent water from blocdng the air flow path. Heat is removed from l

within the containment utilizing the steel containment shell l

as the heat transfer surface combining conductive heat transfer to the water film, convective heat transfer from the water film to the air, radiative heat transfer from the film to the air baffle, and mass transfer (evaporation) of i

the water film into the air. As the air heats up and water evaporates into the air, it becomes less dense then-the i

cooler air in the air inlet annulus.

This differential causes an increase in the natural circulation of the air upward along the containment surface, with heated air / water i

vapor exiting the top /conter of the shield building.

Additional system design details are provided in 31, referenceg,*

The PCS is actuated either automatically, by a containment High 2 >ressure signal, or manually. Automatic actuation opens tis cooling water tank discharge valves, allowing i

gravity flow of the cooling water onto the containment shell. The manual containment cooling actuation consists of t

four momentary controls, if two associated controls are i

operated simultaneously actuation will occur in all divisions. The discharge continues for at least three days, The PCS is designed to limit post. accident pressure and temperature in containment to less than the design values, Reduction of containment pressure reduces the release of j

fission product radioactivity from containment to the i

environment, in the event of a 08A, (continued) l f

go_ _,

8 3,6 32 08/97 Amendment 0 4

l

~-...-... m

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

B 3.6.6 BASES BACKGROUND The PCS is an ESF system and is designed to ensure that the (continued) beat removal capability required during the post accident period can be attained.

F.

APPLICABLE The Passive Containment Cooling System limits the SAFETf ANALYSES temerature and pressure that could be experienced followin a DBA.

the limiting DBAs considered are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No DBAs are asstmed to occur simultaneously or consecutively. The postulated DBAs are analyzed with regard to containment ESF system, assuming the loss of one Class 1E Engineered Safety Features Actuation Cabinet (ESFAC)

Division, which is the worst case single active failure and rdults in one PCS flow path being inoperable.

T*' unalysis and evaluation show that under the worstpe sc.t io, the highest peak contairment pressure = ' miceted n,44 I,p s tM E in re nee 4 occurs during a Sl8 and is less n

contai

.t design pressure. Tnt analysis s that the l

P gg peak contai t temperature is as indiga inreferenceA'd i

also occurs du a SLB.

Both res s meet the intent of the design basis.

See the Bas or LCO 3.6.4,

" Containment Pressur

• and 3.6.5 for a detailed discussion.) The analy and evaluations assume a unit specific power leve 1

Wt. one nassive containment conditions of 1 and 1.0 p The analyses also assume

.. response delayed initiati o provide conservative peak cal ated containment pressure d temperature res s.

The total response time inc s actuation time p1 the time required to achieve full flow the tainment shell.

The modeled Passive Containment Cooling System actuation iesponse time from the containment analysis is based upon a resxnse time associated with exceeding tire contairnent Hig12 pressure setpoint to opening of isolation valves.

The Passive Containment Cooling System satisfies Criterion 3 l

of the NRC Policy Stbtement.

i

{

~

(continued) l N

. 3n,,

B 3.6 33 0B/97 Amenchent 0 L

Con y t Spray S g (Atmos BASES j

4 APPLICA8LE t

ict the resultant containment pressure an SAFETY ANALYSES tempera tents.

No DBAs ar2 assu occur (continued) simultaneously or ively.

T ulated DBAs are analyzed with regard to con ESF systems, assuming h h gF'N the loss of one ESF ich is t case single pg active failur results in one train of t nment 1

l

%ND gb Spray and Containment Cooling Systes being ren b d yp 1

able.

The analysis and evaluation show that under the worst case 7

W w ario, the highest peak cont i ment pressure is

/

MB TM.',f psig (experienced during

'JGA).

The analysis shows that the peak containment temperature is $3H4. {'-f 390,3 'F (experienced during an SLB).

Both results meet UFe inten) of the design basis.

(See the Bases for LCO 3.6.4%,

i

" Containment Pressure,' and LCO 3.6.58 for a detailed ty,3MOlpecifJc-power _leyeLofS0$is, oneAconuinmenta N,a discussion.) Theanalysesandevalggjoosassune s

train d ;r.; centeinment-osl4ng-trak operating, nd initial (pre accident) containment conditions of((~12 F and s i

l0 Ti $ psig.

The analyses also assume a resp me delayed initiation to provide conservative peak calculated containment pressure and temperature responses.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative.

In particular, the effectiveness of the Ph hs..., Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment I

backpressure.

For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with10CFR50,AppendixK(Ref.2).

TH> tif{t Q of an inadvertent containment spray on has been anain.. 4 Qadvertent s r n results in a (2.0)psigcontainmenc

... a d is associated with the sudden cooling. r n the interior of4.ded tioht contal dditional discussion is providedYtM3tm.

0 3.6.'4A.

^

Thr : d ! d cant 31nment Spray System rom the containment analysts n t:^f w lesponse time associated

2' ' a n ainment High J prn.wi.

with exceedlag.- W

(

ar.hintnffull flow through the containment spray nozzlW levt4anad)

(APPLICABLE nt Spray System total responia 4 #

loss of offsite powerkbtR.gemeretF DG) startup (for l60) seconds n A diattel SAFETY ANALYSES

'itrenag-equipment, (continued) containmeDLsPr0 pump startup, and spray 4111g (W$

sv h Containment cooling tyv4a performance for post accident conJitionsisgiveninReferenceK.3Theresultofthe analysis is that each train can provide 100% of the required peak cooling capacity durin the post accident condition.

e train pos ccident coo ng capacity un varying co ineent aab condition,

equired to pe rs the accide

analyses, o shown Reference 5.

h oodeled Containment Cooling 5 tem actuation from the conthmentanalysisisbasedupon esponse time associated with exceeding the contain t High 3 pressure setpoint tt achieving full Containment C ing System air and safety gradescpoling water flow.

The tainment C oling System total response time of (60] se

,ds, includes s

al delay, DG startub({or loss of offsite po r),and serv water pump startup t (Ref.6).

The Conta nt Spray System.and the tainment Coolin System satis erion 3 of the NRC Po Statement.

O n'g a DBA, a minimum of one containment cooli dnand one containment spray. rain are required to m3tfifgd LCO ain the containmen(peak pressure and temperature,below the design limits (Ref.% Additionally, one c3ntfinment spray train is also requited tAremove iodine from the containment oncentrations below those assumed atmosphere and maintiiq1q en'sure that these requirements in the safety analysis.

are met, two containment 4 pts trains and two containment coolingtrainsmustJe'OPERABL Therefore, in the event of an accident, at leatt one train 1 sch system operates, assuming the wo t case single activeM ilure occurs.

Each Conta nt Spray System typically inc es a spray pump, sp y headers, nozzles, valves, piping,

truments, and co rols to ensure an OPERABLE flow path capa of tak suction from the RWST upon an ESF actuation s nal automatically transferring suction to the contain t

ump.

(continued)

WOG STS B 3.6 68 Rev 1, 04/07/95 v

~-

B 3.6.6 V

BASES (continued)

/

LCO During a DBA, one passive containment cooling ater flow path is required to maintain the containment ak pressure and temperature below the design limits (Ref. 41. To ensure that this requirement is met, two passive containment cooling water flow paths must be OPERABLE.

Therefore, in the event of an accident, at least one flow path operates, assuming the worst case single active failure occurs.

The PCS includes a cooling water tank, valves, piping.

instruments and controls to ensure an OPERABLE flow path capable of delivering water from the cooling water tank upon an actuation signal. An OPERABLE flow path consists of either the normally closed air operated valve capable of H0.n f M automatically opening or the air operated valve open and the motor operated valve closed and capable of automatically 0P'"i"9'

/Mssty 3

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the PCS.

During shutdown the PCS may be required to remove heat from containment. The requirements in MODES 5 and 6 :re specified in LCO 3.6.7, Passive Containment Cooling System (PCS)

Shutdown.

ACTIONS M

With one passive containment cooling water flow path ir. operable, the affected flow path must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In this degraded condition, the remaining flow path is capable of providing greater than 100% of the heat removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was chosen in light of the remaining heat removal capability and the low probability of DBA occurrin, during this period.

(continued)

.M

,.a.,

B 3.6 34 08/97 Amenchent 0 y

w c.-,,

-,-,,..,w.,

i 480,1097F A

l INSERT BASES B 3.G.G PAGE B 3.G 34 LCO i

The PCS cooling water storage tank ensures that an adequate supply of water i

is as ailable to cool and depressurize the containment in the event of a Design Basis Accident (DBA). To be considered OPERABLE, the PCS cooling water.

i storage tank must meet the water volume and temperature limita established in the SRs. To be consiiered OPERABLE, the air Dow path from the shield building annulus inlet to the exit must be unobstructed, with unobstructed upper annulusprains providing a path for containment cooling water runoff to preclude blockige of the air flow psth.

/

4-cau i

r

?

f 1

i f

1 5

l-

-......f..

_ ~, __

..a

8 3.6.6 BASES-ACTIONS 8.1 (continued)

If the cooling water tank is inoperable it must be restored to OPERA 8LE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The tank may be declared i

Inoperable due to low water level or temperature out of l

limits. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is reasonable based on i

the remaining heat removal capability of the system and the availability of cooling water from alternate scurees.

j U

If any of the Required Actions and associated Completion Times for Condition A or 8 are not met, or if the LCO is not met for reasons other than Condition A or 8. the plant must l

be brought to MODE 5 where the probability and consequences i

on an event are minimized. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

The extended interval to reach MODE 5 allows additional timefand i

i 4

is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant l

System is reduced in MODE 3.

i l

l SURVEILLANCE SR 3.6.6.1 REQUIREMENTS i

This surveillance requires verification that the cooling water tauperature is within the limits assumed in the accident analyses. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is adequate to i

identify a temperature change that would approach the temperature limit and has been shown to be acceptable in similar applications.

gN SR 3.6.6.2 t

Verification that the cooling water volume is above the required minimum ensures that a sufficient supply is available for containment cooling.

Since the cooling water volume is normally stable and low-level is indicated by a main control room alarm, a 7 day Frequency is appropriate -

and has been shown to be acceptable 0; x 7 :;:nt W i

fd>.NfdF

- (22)

%. ;Ee in similar applicatio'is.

(continued) 3 0 _,,,

83.635 08/97 Amendment 0 l

i i

I 480.1098F (i)

I INSERT BASES B 3.6.6 PAGE B 3.6 35 SR 3.6.6.1 i

I The SR is modined by a Note that el.ninates the' requirement to perform this S'irveillance when ambient air temperature are within the operating limits of the PO9 water storage tar)., With ambient temperatures within the band, the

[

PCS waer storage tank temperature should not exceed the limits.

j i

r t

i 1

1

. f i

}

I t

e L

I i

4

?

4 W

k' 9

Vrd'7'?-

R eP Tre te9-1Mrg-+'WeemvekrT% p3 rT w zfe s was h ef m*=

yev w ' ~. F

'18-p-'>-r-e' d_

u

-7w-'-

g re a g y 91'Y af-kWW bT*- P M F W

• g -

'F 3 W '4 4 7" W

7._

B 3.6.6 BASES SURVEILLANCE SR 3.6.6.3 REQUIREMENTS Verifying the correct alignent of power operated, and automatic valves, excluding check valves, in the Passive Containment Cooling System provides assurance that the proper flow paths exist for system operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct positions prior to being secured, This SR does not

1. 0 /N require any testing or valve manipulation.

Rather, it involves verification, through control room instrumentation or a system walkdown, that valves capable of potentially jpsygy % being mispositioned are in the correct 'msition f bY SR 3.6.6.4 This SR requires verification that each automatic isolation valve actuates to its correct position upon receipt of an actual or simulated actuation signal.

This Su,*veillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. Yhe 24 month Frequency is based on the need to perfors these Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at r= ar a Operating experience has shown that

/y these connonents usually pass the Surve111ances when Od

>erformed at the 24 month Frequency. Therefore, the requency was concluded to be acceptable from a reliability standpoint.

SR 3.6.6.5 Ne %equires-ver44tceti^= that tM eir 'b p:th 'rc-tM iD44miiding anish 1rikt t-tW$ fit is-f n/

n+ t..-* 2 w i g ).3y ; g 7 _yj,.g.j..i'ication ggpLAca u #

____m

" *R+l : 7:2h r;dilE7.M1though there are no b')

N!.

!bdk chNr b i 3 d b,E ENb'4 n

' 5 5 N E 3 M biiity ic ~ M b i densidered--pe"d-a+ t fy edyf4 montifr

~

~

(continued) b AP600 B 3.6 36 08/97 Amendment 0 mm o.a ma:..o

480.1098F (iii)

INSERT BASES B 3.6.0 PAGE B 3.0 30 SR 3 0.6.3 The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a smgle flow path. This Frequency has been shown to be acceptable through operating experience.

480.1098F (iv)

INSERT BASES B 3.6.0 PAGE B 3.0 30 SP, 3.6.6.4 W

The JtCmonth Frequency is also acceptable based on consideration of the design, rehability (and confirmed by operating experience) of the equipment.

480.1098F (v)

INSERT BASES B 3.6.6 PAGE B 3.0 30 SR 3.6.G.5

,&y.edM Periodic inspections of the PCS air [ow path from the shield building annulu inlet to the exit ensure that it is ynobstructed ' the bafile plates are properly insta!!ed and the upper annuluskirains are unobstructed. Although there are no anticipatet' unhanisms which would cause air flow path or annulus drain obstruction and the effect of a missing air baffle section is small, it is considered prudent to verify this capability every 24 months. Additionally, the 24 month Frequency la based on the desire to perform this Surveillance under mditions that apply during a plant outage, on the need to have access to the

. ations, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation in similar situations.

m.

m

B 3.6.6 BASES SURVEILLANCE SR 3.6.6.6 REQUIREMENTS (continued)

This SR requires performance of a Passive Containment Cnoling System test to verify system flow and water coverage capabilities. The system perfonunce test demonstrates that the containment cooling capability assumed in accident analyses is maintained by verifying the flow rates via each standpipe and measurement of containment wetting coverage.

The System Level Operability Testing Progrn provides specific test requirements and acceptance criteria.

Although the likelihood that system performance would degrade with time is low, it is considered prudent to sr4 - /

periodically verify system wrformance. TheGTyear R

Frequency is based on the Mility of the mork frequent Ma surveillances to verify the OPERABILITY of the active components and features which could degrade with time.

+9aM9ff AAI REFERENCES 1.

10 CFR 50, Appendix A. " General Design Criteria for Nuclear Power Plants."

S, t W 18. 10 C"! S0, @rdh Y,, "ECC Evelvetur, "dh.".

Ho.Janf O.

""500 SSA", Chapi.er 15, "Avvi A. Ar.alysis. '

3 K /. AP600 5SAR, Chapter 6.2, " Containment Systems."

b AP600 B 3.6 37 08/97 Amendment 0

B 3.6.7 I

B 3,6 CONTAINMENT SYSTDtS B'3.6,i Passive Containment Cooling System (PCS)

Shutdown BASES BACKGROUND A description of the OCS is provided in the Bases for LCO 3.6.6 " Passive Contairment Cooling System Operating."

APPLICABLE The PCS limits the temperature and pressure that could be SAFETY ANALYSES experienced following a Design Basis Accident (DBA). The limiting DBAs considered during shutdown are the loss of decay heat removal and loss of shutdown margin events.

For shutdown events, the Reactor Coolant System (RCS) sensible and decay heat removal requirements are reduced as compared to heat removal requirements for MODE 1, 2, 3, or 4 events. Therefore, the shutdown containment heat removal requirements are bounded by analyses of MODES 1, 2, 3, and 4 events. A discussion of MODES 1, 2, 3 'and 4 OBAs is provided in the Bases for LCO 3.6.6 " Passive Containment Cooling System (PCS)

Operating."

The PCS satisfies Criterion 3 of the NRC Policy Statement.

LCO For postulated shutdown events, one passive containment cooling water flow path is required to provide the required containment heat removal capability (Ref. 1), To ensure that this requirement is met, two passive containment cooling water flow paths must be OPERABLE. Therefore, in the event of an accident, at least one flow path operates, assuming the worst case single active failure occurs.

The PCS includes a cooling water tank, valves, piping, instruments and controls to ensure an OPERABLE flow path capable of delivering water from the cooling water tank upon an actuation signal.

I A)3EG.->

WBo.Ilo2 P

~

(continued) 300_,,,_

B 3.6 38 08/97 Amendment 0

d 480.1102F INSERT BASES 3.6.7 PAGE B 3.0 38 LCO The PCS cooling water storage tank ensures that an adequate supply of water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA). To be co nidered OPERABLE, the PCS cooling water storage tank must meet the water volume and temperature limits established in the SRs. To be considered OPERABLE, the air flow path from the shield building annulus inlet to the exit must be unobstructed, with unobstructed upper annulus rains providing a path for containment cooling water runoff to preclude block go of the air flow path.

Sakl - rd4h{

y i

7-. -.

+

BASES (continued)

~ APPLICABILITY-OPERABILITY of the PCS is required in H0 DES 5 and 6 with the reactor shutdown and the calculated reactor decay heat greater than 6 mit for heat removal in the event of a loss cf nonsafety decay heat removal capabilities.

With the decay heat less than 6 mit, the decay and sensible heat can be easily removed from containment with air cooling alone.

The PCS requirements in MODES 1, 2. 3. and 4 are specified in LCO 3.6.6, Passive Containment Cooling System (PCS).

Operating.

ACTIONS AJ With one passive containment cooling water flow path inoperable, the affected flow path must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In this degraded condition, the remaining flow path is capable of providing greater than 100% of the heat removal needs after an accident. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was chosen in light of the remaining heat removal capability and the low e

probability of DBA occurring during this period.

B.1 If the cooling water tank is inoperable, it must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The tank may be declared inoperable due to low water level or temperature out of limits. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is reasonable based on the remaining heat removal capability of the system and the availability of cooling water from alternate sources.

(continued) 1 3 00 _,,,

B 3.6 39 08/97 Amendnent 0 9

_e--

,...-_.e-.

,,----F a-w1-n- - - -

9-

B 3.6.7 l

BASES ACTIONS C.1 (continued)

Action must be initiated if any of the Required Actions and associated Completion Times for Condition A or B are not met, or if the LCO is not met-for reasons other than Condition A or B.

If in MODE 5 with the RCS open and/or pressurizer level not visible, action must be initiated.

innediately, to increase the RCS level to a visible pressurizer level and to close the RCS so that the PRHR HX operation is available.

If in MODE 6 with the upper internals in place and/or the refueling cavity less than full, action must be initiated, inmediately, to increase the refueling cavity level to full with the upper internals removed.

In both cases, the time to RCS boiling is maximized by maximizing the RCS inventory and maintaining RCS temperature as low as practical. Additionally, action to suspend positive reactivity additions is required to ensure that the shutdown margin is maintained.

Sources of positive reactivity addition include boron dilution, withdrawal of reactivity control assemblies, and excessive l

cooling of the RCS.

e These Actions place the plant in a condition which maximi'ze the time to actuation of the Passive Containment Cooling System, thus providing time for repairs or appl:.ation of alternative cooling capabilities.

b SURVEILLANCE SR 3.6.7.1 REQUIREMENTS The LC0

.6 Surveillance Requirements (SR 3.6.6.1 through 3.6.6.

are applicable.

The Frequencies associated with each specified SR are applicable.

Refer to the corresponding Bases for LCO 3.6.6 for a discussion of each SR.

REFERENCES 1.

AP600 SSAR. Section 6.2. " Containment Systems."

3 00 _,,,

B 3.6 40 08/97 Amendnent 0 m

~.

4 Q

NRC FSER OPEN ITEM Question '480.1089F (OITS. 6224)

' LCO 3.6.5 Containment Air Temperature ne following open item relates to LCO 3.6.5 Bases. The Applicable Safety Analyses Bases discuulon deviates from the STS. The response to SCSB comment 32 (Ref. Westinghouse letter NSD NRC-97 5263, Response to NRC SCSB Comments on Containment Systems Technical Specification," from Brian A. McIntyre (W) to T. R. Quay (NRC), dated August 19,1997) is only partially acceptable.

(i) De identification of the limiting accident is not provided, it would be the DEO Hot Leg LOCA based on the staff's current understanding of the supporting analyses, SSAR Rev.13. This is considered to be an open item.

-(ii) De 9sults of the analyses are not provided, only a reference to SSAR Section 6.2, " Containment

' nalyses," is provided. This is unacceptable and needs to be corrected following the staff's review and acceptance of the analyses methods. A place holder with reference to the current SSAR revision number should be provided, for example [390.3*F for a DEG Hot Leg LOCA, SSAR Rev.

13]. In addition, the containment design temperature,280*F, needs to be included. This is considered to be an open item.

(iii)

The discussion with respect to the influence of the short durations when the calculated containment air temperature exceeds the design temperature must be included. De analyses to demonstrate that these short deviations are acceptable with respect to the containment design temperature need to be performed and referenced. This is considered to be an open item.

Response

The AP600 Bases 3.6.5, Containment Air Temperature, Applicable Safety Analysis section, will be revised as follows to identify the limiting accident, to provide analysis temperature results, and to address peak temperature greater than design temperature.

. SSAR Resision: See markup provided with FSER Open item 480.1088F.

480.1089-1

dill NRC FSER OPEN ITEM Question 480,1094F (OITS. 6229)

LOC 3.6.6 Passive Containment Cooling System (PCS) - Operating The LCO 3.6.6 Surveillance Frequency of verifying PCS flow and water coverage, SR 3.6.6.6,is unacceptable. De Frequency needs to be consistent with STS Containment Spray and Cooling Systems (Credit not taken for iodine removal by the Containment Spray System) LCO SR 3.6.6B.8.

The mechanical alignment of the PCS water distribution system (becket and weirs) needs to be maintained and it needs to be demonstrated that the PCS flow and water coverage are maintained consistent with the Applicable Safety Analysis. A Surveillance test at the first refueling outage [24 months] needs to be added to the 10 year interval. This is considered to be an open item.

Response

The AP600 system test, SR 3.6.6.6, Frequency will be revised to be "At first refueling and 10 years" consistent with all five of the STS containment spray nozzle surveillances (SR 3.6.6A.8, SR 3.6.68.8, SR 3.6.6C.5, SR 3.6.6D.5, and SR 3.6.6E.7). This revision, which matches the STS, is a slight variation to that provided in response to FSER 01480.1084.

e SSAR Revision: See markup provided with FSER 01480.1088F.

W westingnouse

NRC FSER OPEN ITEM Question 480.1095F (OITS 6230)

LCO 3.6.6 Passive Containment Cooling System (PCS)- Operating (a)

De LCO 3.6.6 Bases Background needs dditional descriptive information based on a review of STS Containment Spray and CoolinF (Credit not taken for iodine removal by the Containment Spray System) LCO 3.6.6B.

(i)

Augment the discussion to include, as an example, "He PCS is designed to ensure that the heat removal capability required during the post accident period can be attained. The PCS limits and maintains post accident conditions to less than the containment design values." This it considered to be an open item.

(ii)

Augment the GDCs with "10 CFR 50 Appendir. A."

This is considered to be an open item.

(iii) Augment the design details to include the recirculation pumps. This is considered to be an open item.

(iv) Augment ti.;

• sign details to include the recirculation heater. His is considered to be an, open item.

(v)

Augment the design details to include the ancillary water storage tank and additional support systems. His is considered to be an open item.

(vi) Correct the typographical error in " Insert Background, fifth paragraph."

This is considered to be an open item.

Response

(i)

The requested information is already in last two paragraphs of Background on pages B 3.6-32 and -33.

(ii)

"10 CFR 50 Appendix A" will be added to the last sentence of the first Background paragraph (page B 3.6.31).

(iii & iv)

De following discussion of the recirculation pumps and heater will be added to the second Background paragraph:

"The recirculation pumps and heater provide freeze protection for the passive containment cooling water storage tank during normal operations. However, OPERABILITY of the tank is

[ W85tingh00S8

NRC FSER OPEN ITEM assured by compliance with the temperature limits specified in SR 3.6.6.2 and not by the recirculation pumps and heater. In addition to the recirculation pumps and heater, the PCS water storage tank temperature can be maintained within limits by the ambient temperature, the large thermal inenia of the tank, or potentially by heat from other sources."

(v)

The ancillary water storage tank is not a suppon system for the Technical Specification operability of the PCS (i.e., mit!gation of DBAs). Requirements applicable to the ancillary water storage tank are specified in SSAR subsection 16.3.1, Investment Protection Shon Term Availability Controls.

(vi) This correction was incorporated in the SSAR Technical Specification revision, dated 8/97.

SSAR Revision: See markups provided with FSER Open item 480.1088F.

f 480.1095F-2 W Westinghouse

iii tais

+

e NRC FSER OPEN ITEM Question 480.1096F (OITS 6231)

LCO 3.6.6 Passive Containment Cooling System (PCS) Opera'ing (a) ne LCO 3.6.6 Bases description of the Applicable Safety Analyses lacks specinc details and contains errors.

(i) The peak containment air temperature occurs during a DEG llot Leg LOCA (Ref. SSAR Rev.

13, Table 6.2.1.1-1), not the SLB as indicated. This is considered to be an open item.

(ii) As discussed in LCO 3.6.4 Bases 4(b)(ii) and in LCO 3.6.5 Bases 4(b)(i),4(b)(ii), and 4(b)(iii), the analyses results need more specific details. This is considered to be an open item.

Response

(i) Tne event which causes the peak containment air temperature is DEHLG LOCA, LCO 3.6.6 Bases description of the Applicable Safety Analysis has been revised to correct the event identification.

(ii) The current value for the containment peak pressure will be added to the AP600 LCO 3.6.6, PCS -

Operating, to be consistent with the SSAR and the Open Item 480.1088F change to Bases 3.6.4, Containment Pressure, Bases.

De AP600 Bases 3.6.6, PCS - Operating Applicable Safety Analysis section, will be revised to be consistent with the SSAR and the Open item 480.1089F change to Bases 3.6.5, Containment Air Temperature, to identify the limiting accident, to provide analysis temperature results, and to address peak temperature greater than design temperature.

The requested PCS Bases revisions are provided in the attached markup.

SSAR Revision: See markup provided with FSER Open item 480.1088F..

480,1096F1 W westingh00Se

p:-

nih NRC FSER OPEN ITEM

^

Question 480.1097F (OITS 6232)

LCO 3.6.6 Passive Containment Cooling System (PCS) - Operating ne LCO 3.6.6 Bases description of the LCO is unacceptable. Both sections only discuss an OPERABLE flow path. There is no discussion of an OPERABLE PCCWST. This is considered to be an open item. Add, for example, based on LCO 3.5.4:

(i)

"The PCCWST ensures that an adequate supply of water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA)."

(ii)

"To be considered OPERABLE, the PCCWST must meet the water volume and temperature limits established in the SRs."

(iii). "To be considered OPERABLE, the upper annulus drains must be unobstructed and provide a path for PCCWST water runoff to preclude blockage of the air flow path." 'This is considered to be an open item.

Response

(i)

Item (i) will be added to the Bases LCO section as shown in the attached markup.

(ii)

Item (ii) will be added to the Bases LCO section as shown in the attached markup.

(iii) Consistent with item (iii), the following will be added to the Bases LCO section:

To be considered OPERABLE, the air flow path from the shield building annulus inlet to the exit must be unobstructed, with unobstructed upper annulus drains providing a path for containment cooling water runoff to preclude blockage of the air flow path.

SSAR Resision: See markup provided with FSER Open item 480.1088F.

[ WeSililgt10USS

NRC FSER OPEN ITEM Question 480.1098F (OITS 6233)

LCO 3.6.6 Passive Containment Cooling System (PCS) Operating The LCO 3.6.6 Dases description of the Surveillance Requirements need to be modified.

(i)

For SR 3.6.6.1, a reference to the Surveillance note is requires, for example, "The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperature are within the operating limits of the PCCWST, With ambient temperatures within the band, the PCCWST temperature should not exceed the limits." His is considered to be an open item.

(ii)

For SR 3.6.6.2, the text "through operating experience" needs to be removed.

This is considered to be an open item.

(iii)

SR 3.6.6.3 is partially consistent with 3.5.2 ECCS Operating SR 3.5.2.2. The Surveillance frequency and its justi6 cation need to t>e included in the Bases. This is considered to be an open item.

(iv)

SR 3.6.6.4 is partially consistent with 3.5.2 ECCS - Operating SR 15.2.5 and SR 3.5.2.6.

The Surveillance frequency and its justification need to be included

Bases. This is considered to be an open item.

(v)

SR 3.6.6.5 is partially consistent with 3.5.2 ECCS - Operating SR 3.5.2.8 While the SR is indeed prudent, the format is unacceptable. To be consistent with the STS the test shoald be revised to " Periodic inspections of the PCS air flow path from the shield building annulus inlet to the exit ensures that it is unrestricted that it stays in proper operating condition by confirming proper placement of the baffle plates. De 24 month Frequency is ba.;ed on the desire [nced] to perform this Surveillance under conditions that apply during a plant outage, on the need to have access to the locations, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. His Frequency has been found to be sufficient to detect abnormal degradation in similar situations." This is considered to be an open item.

("i)

SR 3.6.6.6 is partially consistent with STS Containment Spray and Cooling (Credit not taken for iodine removal by the Containment Spray System) LCO SR 3.6.68.8. However the passive nature of the PCS which relies on the mechanical alignment of the PCS water distribution system (bucket and weirs) needs to be maintained and it needs to be demonstrated that the PCS flow and water coverage are maintained consistent with the Applicable Safety Analysis. A Surveillance test at the 6rst refueling outage [24 months] needs to be added to tlie 10 year interval. This is considered to be an open item.

(vil)

SR 3.6.6.7 needs to be added for the upper annulus drains, consistent with the attemative used to address I (iii). His is considered to be an open item.

T Westinghouse

(

NRC FSER OPEN ITEM

Response

(i)

The recommended statements will be tJded.

(ii)

The specified text will be deleted from the Bases discussion of SR 3.6.6.2.

(iii)

None of the STS containment cooling specincations 16.6 Bases (specs A through E) include the requested statements for the valve position verification surveillance. However, the following STS SR 3.5.2.2 Bases statements (with one minor, plant-specific modi 0 cation) will be added to the AP600 SR 3.6.6.3 Bases discussion:

ne 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single P.ow path. This Frequency has been shown to be acceptable through operating experience.

(iv)

The AP600 Bases discussion for SR 3.6.6.4 (automatic valve actuation) is based on the STS containment cooling specifications 3.6.6 Bases (specs A through E) and includes the STS Surveillance Frequency and justification. The STS SR 3.5.2.5/6 Bases Frequency discussion is the same as the STS containment cooling specification Bases, with the addition of the following statement:

The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirmed by operating experience) of the equipment.

This statement will be added to the AP600 SR 3.6.6.4 Bases discussion with the plant specific fuel cycle frequency.

(v)

The requested text will be incorporated into the AP600 SR 3.6.6.5 Bases discussion as follows:

Periodic inspections of the PCS air Oow path from the shield building annulus inlet to the exit ensure that it is unobstructed, the baffle plates are properly installed and the upper annulus safety related drains are unobstructed. Although there are no anticipated mechanisms which would cause air now path or annulus drain obstruction and the effect of a missing air baf0e section is small, it is considered prudent to verify this capability every 24 months.

Additiona'.'j, the 24 month Frequency is based on the desire to perform this Surveillance under conditions that apply during a plant outage, on the need to have access to the locations, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power, his Frequency has been found to be sufficient to detect abnormal degradation in similar situations.

(vi)

The Surveillance Frequency has been revised to specify "First refaeling outage."

480.1098F-2 T Westinghouse

NRC FSER OPEN ITEM (vii)

PRA RESOLUTION As discussed in the.csponse to SCSB Open item 480.1091F, the SR 3.6.6.5, air flow path inspection includes inspection of the upper annulus drains.

SSAR Resision: See markup provided with FSER Open item 480.1088F.

I 48.1 98F 3 w wesingnouse

NRC FSER OPEN ITEM Question 480.1099F (OITS. 6234)

LCO 3,6.6 Passive Containment Cooling System (PCS) Operating The LCO 3.6.6 Bases References should be corrected. References 2 and 3 are not used and need to be deleted. Reference 4 becomes reference 2 and the attendant text needs to be corrected. This is considered to be an open item.

Response

The references will be revised to eliminate superfluous entries and to correctly reference sources.

SSAR Revision: See markup provided with FSER Open item 480.1088F.

r i

l 480.1099F-1 W851111gh00S8

NRC FSER OPEN ITEM Question 480.1102F (OITS 6237)

LCO 3.6.7 Passive Containment Cooling System (PCS) Shutdown LCO 3.6.7 Bases discussion on the LCO is unacceptable. Both sections only discuss an OPERABLE flow path, ~1nere is no discussion of an OPERABLE PCCWST. This is considered to be an open itent Add, for example, based on LCO 3.5,4:

(i)

"The PCCWST ensures that an adequate supply of water is available to cool and depressurir.e the containment in the event of a Design Basis Accident (DBA)."

(ii)

"To be considered OPERABLE, the WCWST must meet the water volume and temperature limits established in the SRs."

In addition, in consideration of Open item 1097F above, also add:

(iii)

"To be considered OPERABLE, the upper annulus drali,5 must be unobstructed and provide a.

path for PCCWST water runoff to preclude blockage of the air flow path." This is considered to be an open item, e

- Response:

(i) item (i) will be added to the Bases LCO section as shown in the attached markup, (ii)

Item (ii) will be added to the Bases LCO section as shown in the attached markup (iii) Consistent with item (iii), the following will be added to the Bases LCO section:

To be considered OPERABLE, the air flow path from the shield building annulus inlet to the exit must be unobstructed, with unobstructed upper safety related annulus drains providing a path for containment cooling water runoff to preclude blockage of the air flow path.

SSAR Revision: See markup provided with FSER Open Item 480.1088F.

48.u 2F.i W wesuouse

9:r qs NRC FSER OPEN ITEM Question 480.1103F (OITS 6238)

LCO 3.6.7 Passi,c Containment Cooling System (PCS) Shdtdown The LCO 3.6.7 Bases discussions on Applicability. Actions, and Surveillance Requirements are referenced to LCO 3.6.6 (PCS - Operating) SRs 3.6.6.1 through 3.6.6.8. Under the current version of LCO 3.6.6, there are only six SRs. The referenced number of SRs needs to be corrected. His is considered to be an open item.

Response

The Bases discussion of SR 3.6.7.1 will be corrected to reference SRs 3.6.6.1 through 3.6.6.6.

SSAR Revision: See markup provided with FSER Open Item 480.1088F.

480.1103F-1 W westinghouse

t t

' T re i

NRC FSER OPEN ITEM -

i

Question' 720.423F (OITS 6135).

. In meeting the RTNSS criteria, credit was taken for external reactor vessel cooling (ERVC) as a.

l strategy for retaining molten core debris in vessel. This results in the majority of core melt accidents

(-90 percent) being arrested m vessel, thereby avoiding RPV failure and associated containment '

challenges from ex vessel phenomena. Successful RCS d: pressurization and reactor cavity Gooding are prerequisites for ERVC, and credit for these aspects of ERVC in the focussed PRA is appropriate 1 since both functions are fulfilled by safety related systems. Howeser, the nonsafety related RPV thcrmal insulation system is also required for successful ERVC, The thermal insulation system limits

~

thermal losses during normal operations, but provides an engineered pathway for supplying water cooling to the sessel and venting steam from the reactor cavity during severe accidents. Attributes of-the system include specific RPV/ insulation clearances and water / steam now areas based on scaled tests, integral ball and cage check valves and buoyant steam vent dampers which change position during Good up of the reactor cavity, and insulation panel and support members designed to withstand the hydrodynamic loads associated with ERVC.

- If credit for ERVC is reduced, large release frequency and CCFP would increase proportionally since -

-all RPV breaches are assumed to lead to early containment failure in the PRA. Under the most-limiting assumption of no credit for ERVC, the large release frequency would approach the core melt i

frequency and CCFP would approach 1.0. In view of the reliance on ERVC to meet the Commission's large release frequency goals, the staff will require an appropriate level of regulatory oversight of the !

RPV thermal insulation system This oversight should provide reasonable assurance that the as built insulation system conforms with design specifications contained in Chapter 39 of the PRA, and that the operability of the system is confirtned through periodic surveillance.

The RPV insulation design description and functional requirements are not currently included in the SSAR, ITAAC, or reliability assurance program. The design description and functional requirements for the RPV iasulation should be added to the SSAR, and important criteria associated with the insulation design should be incorporated into the ITAAC, including information related to the j-necessary clearances /Dow areas, and the check valves and steam vent dampers. The system should be included as a risk signi6 cant SSC in the reliability assurance program, and reliability / availability controlc and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained.

c Westinghuuse Response:

Functional requirements for the reactor sessel insulation was incorporated in section 5 3.5 of Revision

-14 of the AP600 SSAR. Based on discussions with the NRC staff, more information was requested to be ~ included in the SSAR. SSAR Section 3.3.5 is modined to include the design bases and design

- description for the reactor vessel insulation and is attached. In addition, the reactor vessel insulatica is included as a risk-significant_SSC in the reliability assurance program as shown in the proposed revision to SSAR section 17.4; The AP600 Certined Design Materialis~also revised to include appropriate ITAACs for the reactor vessel insulation per the response io RAI 720 442F.

T W85tingh0050-

~.. _....... _......_. _._.._._._..

.y i

T NRC FSER OPEN ITEML 7

SSAR Revision:-

i Revised SSAR Sections 5.3.5 and 17,4 aitsched. See the response to.RAl~720.442F for changes to the :

j

. Cenified Design Maieriato

.[

?

?s 1

1 b

r I

e l

5 1

- g

,i J

I.

i r

!r t, _ ' g ';

i,

}

4 I

d 4

' 720,'423F 2 Y WOStingh0080 l.

h.

.-ra.

u

,n-,e n,,

+

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

~,

Prn=u'

$[ Rhcirr Ccalant Systini ecd Cemected Syst mis.

~

e-i Reactor. vessel studs, nuts, and washers can be removed to dry storage during refueling.L Access is provided to the reactor vessel nozzle safe enda. The insulation coverinF the.

- nozzle to pipe welds may be removed, l

j, Because tr.diation levels and remote underwater accessibility limits access to the reactor' vessel, several stept have been incorporated into the doi n end manufactunng procedures in b

preparanon for the periodic nondestructive tests shich ec required by the ASME Codo insmice inspection requirements.. These are as fo!!ows:

Shop ultrasonic examinations are performed on in*ernally clad surfaces to r.

acecptance and repair standard to provide an adequate cladding bond to allow later ultrasonic testing of the base fuetal from the inside surface. The size of claddir.g bond defect allowed is 0.25 inch by 0.75 inch with the greater direction parallal to the weld in the region bounded by 2T'(T = wall thickness) on 'ooth sides of each full-penetration pressure boundary weld, Ur. bounded areas exceeding 0.442 square inches (0.75-inch diameter) in other regions are rejected.

The design of the reactor vessel she!! is an uncluttered cylindrical surface to permit future positioning of the test equipment without obstruction.

The weld deposited clad surface on both' sides of the welds to be inspected is specifically prepared to ensure meaningful ultrasonic examinations.

During fabrication, full penetration ferritic pressure boundary welds are ultrasonically examined in addition to code examinations.

Aner the shop hydrostatic testing, full penetration ferritic pressure boundary weids (with the exception of the closure head welds), as well as the nozzies to safe end welds, are uitrasonically examined frc.m both the inside and outside diameters in addition to ASME Code, Section !!! requirements. The closure head ferritic pressure boundary welds are examined from the outside diameter only.

The vessel design and construction enables inspection in accordance with the ASME Code,Section XI. The reactor vessel inservice inspection program is detailed in the technical speciGeations.

5.3.5 Reactor Vessel Insulation

~

in the "-lihe'y :.: t of: b:ycad 3:::;n b:s aceid:::,ih:::=::: v==! := lation de:: ct interfere-witM eet4emeva! f c: ect: d:b-i: !: beiling - th: ::::id: :d : ef 'h: ::=:::

. = ! The f=cden:! ::; :eemen:: :'npc-:::: :: : :=n! ::::::!

.:: = fc!!c. ::

Draft Revision: 19 Y WOStiflgh0USSI 53 19 December 12,1997 i

,e

,--ww

-~

w

5. Rncezr Ce l=t spt;m c:d can::ct:d spiems 4-water 4nte445-proWded464he4ottom-of-the4nsulation-The-water 4alet4 Hired 4*eh a

uaHhe-pressure-drop-througlHhe4 ale 64e-negligible-during-theireulation ef water accesued e.ithahe-inwessel-retention-phenomena, b.

The4nsulatin-preWde,-e-steam-vent-aHhe :cp cf the4ielogier4-shiehiwall--The steam 4ent-erea-it-greater 4han-et-equal-to-the-minimum-Gow-area 4ishutruetures form in g4he-eircu l at ionloop-( sol-ineluding-t he4nsu lat ion 4 t sel fb-4he-m inimum-flow area 49-744',

I 5.3.5.1 Res.ctt,r Vessel insulation Desittn Bases l

Reactor sessel insul.ition is provided to minimize heat losses from the primary system.

I Nonsafety re:ated reDective inculation simitsr to that in use in surrent ;tessurized water I

reactors is utilized. The AP600 reactor <essel insulation contains design features to promote I

in vessci retention following severe accidents. In the unlikely esent of a beyond design basis I

accident, the scactor cavity is Gooded with water, and the reactor vessel insulation allows heat I

removal from core debris via boiling en the outside surface of the reactor vessel. The reactor I

vessel insulation permits a water layer next to the reac;or vessel to promote heat transfer from I

the reactor sessel This is accomplished by providing:

I A means of allowing water free access to the region between the reactor vessel and 1

I insulation.

l l

A means to allow rem generated by water contact with :he re.mtor vessel to escape I

from the region surroundir.g the reactor vessel.

I I

A support frame to prevent the insulation panels from breaking fre: and blocking I

water from cooling the reactor vesset exterior surface.

I I

The reactor sessel insulation and its supports are designed to withstand bounding pressure i

differentials across the reactor sesselinsulation panels durmg the period thu the reactor vessel I

is externally Gooded with water and the core retained in the reactor vessel through heat i

removal from the vessel wall accomplished by the water. This is accomplished by providing 2

i a minimum Dow area of 7.5 A in the portions of the now paths required to vent steam. The i

Dow path from the reactor loop compartment to the reactor cavity provides an open new path I

for water to Good the reactor cavity. The reactor vessel insulation inlet assembles are I

designed to minimize the pressure drop during ex vessel cooling to permit water to cool the I

sessel-l S.3.5.2 Description of Insulatinn l

A schematic of the reactor sessel, the vessel insulation and the reactor cavity is shown m i

Figure 53 7. The insulation is mounted on a structural frame that is supported from the wall I

of the reactor cavity. The vertical insulation panels are designed to have a minimum gap i

between the insulation and reactor vessel not less than 2 inches undei static load conoitions Dcaft Resision: 19 December 12,1997 53 20 W Westingfl00S8

5. R act:r Coolzt S) stem cd Cunct:d Systems I

l associated with containment Goodup. A nc nital gap (with no dcDection) of more than twice I

the minimum gap is provided.

l I

The conical design of tht; bottom portion of the sessel insulation is constructed of Gat panels.

l This provides a single point of contact with the spherical portion of the sessel in the event I

that an insulation panel becomes dislodged. This prevents hot spots from developing on the I

reactor vessel and permits suf0cient flow to maintain in vessel retention. The nominal gap i

between the conical portion of the insulation and the spherical portion of the reactor vessel I

is not less than 9 inches I

The structural frame supporting the insulation is designed to withstand the bounding severe I

accident loads without exceeding denection criteria. The fasteners holding the insulation I

panels to the frame are also designed for these loads.

I l

At the bottom of the insulation are water inlet assemblies. Each water inlet assembly is

.I normally closed to prevent an air circulation path through the vessel insulation. The inlet I

assemblies are self actuating passise devices. The inlet assemblies open when the cavity is Alled with water. This permits ingress of water during a severe accident, while presenting i

excessive heat loss during normal operation.

I The total now area of the water inlet assembhes have suf0cient margin to preclude significant i

pressure drop during ex-vessel cooling during a severe accident. The minimum total Dow I

area for the water inlets assemblies is 6 ft. Due to the relatisely low approach velocities in 2

I the flow paths leading to the reactor casity, and due to the relatively large minimum now area I

through each water inlet assembly (> 7 in ), the water inlet assemblies are not susceptible to I

clogging from debris inside containment.

I i

Near the top of the lower insulation segment are steam vent dampers. These dampers are I

normally closed to prevent reactor sessel heat loss, and a small buildup of steam pressure i

under the insulation will.cause them to open to the vent position. The steam sent dampers I

are passive, self-actuated devices ond will operate when steam is generated under the I

insulation with the cavity filled with water.

I I

in-vessel retention requires a flow path from the reactor coolant loop compartments to che I

reactor cavity. The path from the loop compartments to the reactor cavity is open, and free l

from obstructions that could block water from flooding the cavity during an accident. Doors I

in this now path that could preclude a minimum Cow area of 6 ft' are required to open to I

permit water to flood the reactor cavity compartment. This includes the door between the I

reactor coolant drain tank room and the reactor cavity. This door is not capable of being I

latched closed, and opens with minimal force from the water as it floods the reactor cavity.

l l

Extensive maintenance of the vesselinsulation is not normally required. Periodic verincation I

that the vessel insulation moving parts can be performed during refueling outages.

'l i

Draft Resision: 19

[ Westiflgl10USe 5.3-21 December 12,1997

(

5. React 1r C :la:t s,ut:m c:d cc :ect:d syntimi l

The expected forces that may be expected in the reactor cavity region of the AP600 plant I

during a core damage accident in which the core has relocated to the lower head and the I

reactor cavity is renooded have been consen atively established based on data from the ULPU i

test program (Reference 5). The particular configuration (Configuration 111) reviewed closely I

models the full scale AP600 geometry of water in the region near the reactor vessel, between i

the reactor sessel and the reactor vessel insulation. The ULPU tests provide data on the i

pressure generated in the region between the reactor vessel and reactor sessel insulation.

l These data, along with observations and conclusions from heat transfer studies, are used to

)

I develop the functional requirements with respect to in vessel retention for the reactor vessel I

insulation and support system. Interpretation of data collected from ULPU Configuration 111 l

experiments in conjunction with the static head of water that would be present in the AP600 l

is used to estimate forces acting on the rigid sections ofinsulation. Further evaluation of the I

forces on the reactor vessel insulation and supports is provided in the AP600 Probabilistic l

Risk Assessment.

I l

5.3.5.4 Design Es aluation l

l A structural analysis of the AP600 reactor cavity insulation system demonstrates that it meets I

the functional requirements discussed above. The analysis encompassed the insulation and I

support system and included a determinanon of the stresses in support members, bolts, I

insulation panels and welds, as well as dcHection of support members and insulation panels' l

I The results of the analyses show that the insulation is able to meet its functional requirements.

l The reactor vessel insulation provides an engineered pathway for water cooling the vessel and I

for venting steam from the reactor cavity.

I I

The reactor vessel insulation is purchased equipment. The purchase specitication for the i

reactor vessel insulation will require confirmatory static load analyses.

5.3.6 Combined License Information 5.3.6.1 Pressure-Temperature Limit Curves The pressure-temp. curves shown in Figures 53 2 and 53 3 are generic curves for AP600 reactor vessel design, and they are the limiting cunes based on copper and nickel material composition. liowever, for a speciGe AP600, these curves will be plotted based on material composition of copper and nickel. Use of plant-specific curves will be addressed by the Combined License applicant during procurement of the reactor vessel.

5.3.6.2 Reactor Vessel Materials Surveillance Program The Combined License applicant will address a reactor vessel reactor material surveillance program based on subsection 53.2.6.

5.3.6.3 Reactor Vessel Materials Properties Verification Draft Revision: 19 December 12.1997 53 22

[ Westingt10US8

~-

. @ semi!!E F

' $ Mrctir Cult.at Systim cad Ce:n:cted Systims EAM -

j The Combined License applicant will address verificuion of plant specide belt line material

~

properties consistent with the requirements in subsection.5.3.3.1 and Tables 5.31 and 5.3 3.

'l-5.3.6,4-Reactor Vessel Insulation -

-I

.4 s.

1 The Con.bined License applicant will address verification that the reactor vercel insulation

l' is consistent with.the design bases astablished for in vessel retention.-

l 5.3.7 References 1.

- ASTM E 185 82, Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels."

-2.

Soltesz, R. G., et al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation. Volume 5 Two Dimensional Discrete Ordinates Techniques,"-

WANL PR-(LL) 034, August 1970.

L 3.

S AILOR RSIC Data Library Collection DLC 76," Coupled,Self Shielded,47 Neutron, 20 Gamma Ray, P3, Cross Section Library for Light Water Reactors."

'4.

NRC Policy issue," Pressurized Thermal Shock," SECY 82 465, November 23,1982.

g 1

Theofanous, T.G., et al., "In Vessel Coolability. and Retention of a Core Melt,"

I DOE /ID 10460, July 1995.

l-i I

7 Draft Revisioni 15 Y W85tingh0080L

-5.3 23' December 12,1997

in iu

$. Rrct:r Cr:la:t System cnd Cz::ect:d Syst;ms reQCtor T

vessel r

steam vent (0 UjfI Dg p__

__q shield wall l

l I

l l

Core l

l l

1 l

L__

__J insutation (4 I

% ':: % water intet a%dlin (3I (1) Minimum steam vent flow area provided in section 5.3.5.1 (2)- Minimum gap between insulation and vessel insi.!ation provided in section 5.3.5.2 (3) Minimum flow area provided in section 5.3.5.2 Figure 5.3-7 Schematie of Reactor Vessel Insulatica Draft Resision: 19 December 12,1997 5.3 36 3 W65tiflgh00S8

ga M

17. Qu:lity Ass 2r ca Table 17 41 (Sheet 7 of 9)

RISK SIGNIFICANT SSCs WITillN Tile SCOPE OF D. RAP Splem, Structure, or Rationale

• Insights and Assumptions Component (SSC)'"

system: Reactor Coolant System (RCS)

ADS Stages 1/23 EP,L2 The ADS provides a controlled depressuritation of the RCS Motor-Operated Valves following LOCAs to allow core cooling from the accumulator. IRWST injection, and containment recirculation. The isDS provides

• bleed" capability for feed / bleed cooling of the core. The ADS also provides depressurization of the RCS to prevent a high-pressure core melt sequence ADS 4th Stage Squib RAW /CCF The ADS provides a controlled depressurization of the RCS Vahes following LOCAs to allow core cooling from the accumulator, IRWST injection, and containment recirculation. The ADS provides

• bleed" capability for feed / bleed cooling of the core. The ADS also provides depressurization of the RCS to prevent a high-pressure core melt sequence.

Pressurizer Safet) Vahes EP Th(se valves provide overpressure protection of the RCS I

Reactor Vessel Insulation EP These devices provide an engineered flow patt o promote 1

Water inlet and Steam Vent in. vessel retention of the core in a severe accident.

I Deuces System: Normal Residual Heat Removal System (RNS)

RNS Pumps FP These pumps provide shutdown cooling of the RCS They also provide an alternate RCS lower pressure injection capabihty followmg actuation of the ADS.

The operatien of these pumps is RTNSS-important durmg shutdown reduced inventory conditions RNS valve ruhgnment is not required for reduced inventory conditions.

RNS Motor operated RRW/FVW These MOVs align a tTowpath for nonsafety-related makeup Valv's to the RCS following ADS operation.

System: Spent Fuel Coolmg System (SFS}

SFS Pumps EP These pumps provije flow to the heat exchangers for removal of the design basis heat load.

Draft Revision: 19 W Westinghouse 17 17 December 12,1997

l

(

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION ut te

/

1 Short Term Availability Control Questions rat 720.432F (OITS #G161)

Westinghouw has proposed availability controls on Normal Residual 11 eat Removal system (RNS) and its support systems (Service Water System (SWS). Component Cooling Water System (CCS), and AC power) when Reactor Coolant System (RCS) level is not visible in the pressurizer until the refueling cavity is h:uf full and the upper internals are removed. The staffi review found that this additional regulatory oversight for RNS and its support systems (CCW, SSW and AC power) must be extended to Mode 5 operation when the RCS is open.

Westinghouse needs to modify Section 16.3 of the SSAR to require additional regulatory oversight for RNS and its support systems (CCW. SSW and onsite AC powe.) for the whole period of Mode 5 when the RCS open.

This is an open item.

Response

We agree with the NRC comment. The short term availability controls have been revised to eliminate the mention of reduced RCS inventory and to use RCS open to denne the MODE nf applicability. Words have been added to the BASES to clairify what conditions constitute RCS open, e

i SSAR Changes: See attachment.

SSAR Section item Change Table 16.31 Lfst of Short Term Avail Controls Change Description of items 2.2, 2.3, 2.4, 3.2 Avail Control 2,2 RNS RCS Open Change " reduced inventory" to "RCS Open" Avail Control 2.3 CCS RCS Open Cha:,e " reduced inventory" to "RCS Open" A vail Control 2.4 SWS RCS Open Change " reduced inventory" to "RCS Open" Avail Control 3.2 AC Power Supplies RCS Open Change " reduced inventory" to "RCS Open" ITAAC Changes:

None W Westinghouse l

l l

y

)

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION m

n 5

)

l, NRC Letter Dated 10/2/97 Question (1)

In a letter to Westinghouse dated June Y.1997, the NRC sta'ed that the RMSS administrative controls should include a commitment to satisfy the hiaintenance Rule and to establish the.recilability and reliability perfonnance goals which would be applicable for those SSCs under the COL applicantiimplemenuuion of the Maintenance Rule. De proposed procedures do not provide any link to the hiaintenan.cc Rule. De discussion in the introductory sections of the short term availability controls, as well as the bases section of each system control procedure, should clarify the relationship between the hiaintenance Rule and availability / reliability for the subject SSCs.

Response

De first paragraph of'SS AR subsection 163.1 has been revised to satisfy this comment. See attachment.

Ouestion (2) f The BASES sections for most of the admimstrative availability control procedures cite a minimum availability '

of 75 percent of the function, it is not clear what is meant by minimum availability and if this value relates to maintenance rule perfonnance goals. It is aho not clear if the cited availability applies at a system, train, channel (where applicable), or component ic$el. Although a 75% availability value represents a significant increase in margin for the focused PRA results (since no availability was previously assumed),it does not com-pletely address the staffi uncertainties in the data used in the focused PRA (e.g., software and check valve reliability) %e staffi teuew finds that an annigc fuel cycle availability of 90 percent or better is a more realistic performance goal for maintenM.cc rule implementation on these systems. Bis should be easily achievable. Based on staff audits and inspections at operating plants,it is.are for any industrial grade SSC to t e unavailable more than 15 percent of the time (8$ percent availability), and most risk significant SSCs have availabilities greater than 90 percent (over the course of a fuel cycle). The staff would expect the AP600 defense in depth SSCs to perform even better. Westinghouse should explain how the availability values were detennined tmd why availability / reliability goals more in line with operating plant experience and the projected baseline pRA values are not proposed.

Response

The availability goals were developed based on expert opinion with consideration given to the capabilities of the SSCh as well as to the demands for planned maintenance. We agree that 90% availability should be reasonably achievable for important nonsafety-related functions that are included in the hiaintenance Rule and are covered by these investment protection short term availability controls. The availability goal applies to the function described in the short term availability controls BASES. Examples include. DAS ESFA actuation (1.2) includes the sensors and functions list?d in table 1,2-1 and the RNS (2.1) includes one train of RNS injection. The availinility goal includes both maintenance unavailability and failures to operate during the MODES of ap? cability of the short-term availability control. The availability should be averaged over severrd years in l!

o' der to provide meaningful statistics. De number of years used for averaging will be determined during the i

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION l

?

impicmentation of the Maintenance Rule by the COL applicant. The DASES sections of the availab!!ity controls has been revised based on this response. See attachment.

Question (3)

"Ihe proposed administrative controls cite availability values but do not provide nny reliaHlity values, floth reliability and availability should be siecified for establishing maintenance rule performance goals. 'The

=

availability values should be specined in conjunction with a period of time e.g.. (90 percent per e. h fuel cycle),

to ensure that there is no ambiguity in allowable unavailability times. Similarly, reliability values should include number of f ailures and demands over a period of time. Some of the SSCs in Table 163.2 (e.g. the i'assive Containment System Water Storage Tank Malcup. Mam Control Room, Instrumentation and Control Room Cooling AC Power Supplies, and the DC Power Supplies) do not contain any infonnation on reliability and availability auumptions. Appropriate availability and reliability goals should be provided for these systems.

Response

As discuned in our response to question 2, the availability goals provided in the short tenn availability controls emer both availability and reliability (maintenance unavailability and failures to operate) during the MODES of applicability of the short term availability control. As a result, separate reliability goals are not denaed. The availability should be averaged over several years in order to provide meaningful statistics. The number of years used for averaging will be determined during the implementation of the Maintenance Rule by the COL applicant, lhe short term availability controls that cover SSC's that provide long term shutdown functions did not hase availability goals because long tenu shutdown operations are not modeled in the PRA. Availability goal of 90%

has been added to these f unctions. See attachment.

Question (4)

For SSCs which cannot be restored to operable status within the speciDed compt t% n time. Westinghouse states that the COL shall " Document the justification for the actions taken and input prosided to 0 PA P in plant records" The meaning and intent of this action statement is unclear. It appears that Westinghouse might be recommending tlyat a repon be prcpared and placed in the plant records that documents the temedial measuses implemented to improve long tenn reliability and availability t.tilliing established operational reliabihiy awurance pnicesses such as the maintenance rule or the quality assurance program. Westinghouse %ds to clanty thk action.

Response

Two activities were intended. One activity was to document in tiie plant records the actions taken 10 restore the f unction in OPERA 14LE status, since the completion time to restore the function was exceeded. The second activity was to provl.c input to the Maintenance Rule so that the event can be reviewed and appropriate NRC 10797 2 3 Westinghouse

i RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION i,n '

'ilE 1

measures takn. Since the Maintenance Rule will already review this infonnation,it is unnecessary to include this second actiuty in the short term asailabilily controls. To simplify the short.tenn availabilit) comrols and to asoid duplication el admmistrahve controls these words are revised as follows:

Cursent:

l>ocwncnt the jusufication for the actions taken and input providrJ to 0 RAP in plant records Yevised.

Doc ument in plant records the just(fication fer the at tions tulen to restore the function to OPERAltLE Ouestion (5)

De bases for Availability Control 2.8," Hydrogen Ignitors " states that the hydrogen ignitor system should be auntable during severe accidents because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety related SSCs to mitigate al power and shutdown events. This is not an adequate basis for the ignitoss. A inore appropriate basis would be to meel 10 CFR 5034(f)(2)(ix). De bases should reflect that this system h being provided to safely accommodate hydrogen generated by the equivalent of a 100 percent fuel < lad metal water reachon to address the lessons learned from the accident at Three Mile Island. Tbc bases should also cover the more important aspects of the design such as the ability to promote hydrogen burning soon alter the lower llammability limit k reached by providing at least one ignitor from each group for every e

inne,

Response

%e il ASES for the hy Irogen ignitor (2.8) has been revised to address these comments. See attachment.

Question (6)

It is not clear how the minimum availability value relates to the ignitor system operability, For instance, is the ignitor system considered unavailable if one or more ignitors are inoperable or can multiple ignitors be ineperable as long as 754 of them are available? To avoid confucion in this area, Westinghouse should provide availability goals for the power supplies each group of ignitors, and each coverage tone, The sudi expects availability goah for the ignitor system to be consistent with ignitor systems in operating plants. Although ignitors in operating plants were installed to meet dif ferent regulations, both ignitor systems, AprMs and operating plants..tre designed to promote hydrogen burning soon after the lower llammability limit is reached.

The stalt estimates the availability of ignitor sysvems at nperating plants to be greater than 909.

Response

Lach of the arcat of the containment listed in Table 2.81 contain redundant sets of ignitors. This short term availability control requires that in each area one of these sets of ignitors be available. Words have been added to note 2 of thh table to clarify the ignitor availability, See attachment. As an example, the Refueling Cavity area has lynitors 55 & 58 (Group 1) and ignitors 56 & $7 (Group 2), if ignitors 55 & 58 (Group 1) are available then the hydrogen ignitor function in that area is available even if one or both of ionitoi % & $7 W Westinghouse

l HESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION (Group 2) are unavailable. At the same time,it would be acceptable for the IRWST Inlet area ignitor 15 (Group

2) to be available and ignitor 16 (Group 1) unavailable.

The ignators have redundant power supplies including offsite power, onsite diesel generators ar.d IE hatteries /

inserters. The diesel generators have short term availability controls and the IE batteries / inverters have Technical Specifications. No changes are necessary for Ignitor power supplies because of the redundancy of power supplies and because one whole group of ignitors may be unavailable.

The availability goal has been clarilled and thn value has been increased to 90%. Refer to the response to questton 2.

Question (7).

Westinghouse states that the ignitor function should be available during MODE I and MODE 2 when core decay heat in high. There are no availability controls in MODES 3 and 4 when decay heat levels can be comparable to MODE 2. Westinghouse should provide a basis for not providing availab lity controls for the hydrogen ignitors in Modes 3 and 4 f

Response

The li ASES for the hydrogen ignitors explains that the ignitors should be available in MODES 5 and 6 when the RCS is open. The importance of hydrogen ignitors is less risk important in MODES 3 & 4 because of reduced decay heat and because the probability of core damage is less as compared to MODE 1. MODE 2 is more like MODES 3 & 4. however it was included with MODE I to be conservative. No change is necessary.

Question (8)

The DAS sorveillance requirements on pages 16.3 7 and 16.310 recommend that a CHANNEL CllECK be per-tonned at a licquency of once escry 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The stalf finds this Irequency excessively long since the channel check can he performed by a computer on a per shilt basis and does not represent a burdensome task to the control room opeia!or. The stall considers a channel check of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be a more reasonable frequency.

Response

The DAS is not a safety related protection system and these short term availability controls are not Technical Specifications. It is rmt appropriate to require the s:une surveillances on DAS as on the PMS. In addition, the DAS instrument readouts will most likely be dedicated readouts that are not input to a computer. In order to resolve this concern, the channel check frcquency has been reduced to once per 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. See attachment.

NRC 10'2'97 4 3 Westinghouse

I RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION e

r Question (9)

The definition of CilANNEL CALF BRATION on page 1631 should tw revised to delete the 3rd and 4th l

wntences on RTD cahbration. RTD calibiation should follow the guidance provided in Chapter 7 of the Standard Review Plan and tiranch Technical Position lilCB.13.

Response

"the definition of CllANNEL CAllBRATION contained in SSAR subsection 163 is identical with that contained in the Al%00 Tect.nical Specifications in SSAR subsection 16.1. It is also identical with the definition contained in NUREO 1431, revision 1. No change is necessary.

Question (10) llawd on precedent set in the rulemaking process for the evolutionary reactor desigr) certifications, the staff requests that the Design Reliability Assurance Prognun (D RAP) be moved from Chapter 16 of the Al%00 SS AR to Chapter 17. This is consistent with where the D RAP information is hicated in the evolutionary design I

SSARs.

i

Response

The D RAl'section (16.2) will be moved to SS AR chapter 17 in the next SSAR revision.

Question (11)

Editorial Conunents:

The information under the Corsbined License infonnation section on page 163-4 should be revised to read ".. referencing the AlWW) will develop procedures to control the operability./

On pages 163 7 and 16310, the setpoint for the PRilR llX Actuallon signal on 50 Wide Range Level i=

should be in tenns of 4 of span rather than pounds mass.

Response

The first comment will be incorporated. See attachment.

The 50 wide level setpoints are expressed in tenus of pounds because that is the value used in safety analysis.

This same approach is used in the Al%00 Technical Specification. No change is necessary.

NRC 10/2/97 5 a

i

..., +

l f

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION 16J Intestment Protection 16J.1 Insentment Protection Short. Term Availability Controls i

The importance of nonsafety-related systems, structules and components in the ANel has been evaluated.'

The evaluation uses PRA insights to identify systems, structures and components that are important in l

protecting the utilities investment and for preventing and mitigating severe accidents. To provide reasonable assurance that these systems, structures and components are operable during anticipated events i

Short term availability controls are provided. These investrnent protection systems, structures and components are also included in the D RAP /O RAP (refer to Table 16.21), which provides confidence that

~

l availability and reliability are designed into the plant and that availability and reliability are maintained I

throughout plant life through the maintenance rule. Technical Specifications are r.ot required for these systems structures and components because they do not meet the relection criteria applied to the AP600 (refer to subsection 16.1.1),

t Table 1631 lists nonsafety related systems. structures and components that have investment protection i

short term availability controls. This table also lists the number of trains that should be operable and the plant operating MODES when they should be operable. Table 163 2 contains the investment protection short term availability controls. Thesc short. term availability controls define:

e l

- Equipment that should be operable Operational MODES when the equipment should be operable.

- Testing and inspections that should be used to demonstrate the equipment's operability.

- Operational MODES that should be used for planned maintenance operations Remedial nctions that should be taken if the equipment is not operable Tables 1631 and 163 2 contain defined terms that appear in capitalized type. These terms are defined

[

below.

AC7 IONS-shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

CHANNEL CAlJllRATION-shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input.1he CIIANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an

. in place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel, Whenever a sensing element is replaced, the next required CHANNEL CAllBRATION shall include an in place cross calibration that compares the other sensing elements with

'the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

NRC 10'2/97 6 3 W95tiflgt10088.

+

~

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

{

1 CHANNEL CHECN-shall be the qualitati.e assessment, by observation, of channel behavior during operation. 'Ihis determination shall include, where possible, comparison of the channel indication and status to other indications or status derived (tom independent instrument channels measuring the same parameter, CHANNEL OPERAT/DNAL TEST (COT)-shall be the injection of a simulated or TEST (COT) actual signal into the ch:mnel as close to the sensor as practicable to verify the OPERABILITY of required alann, interkick, display, and trip functions. 'Ihe COT shall include adjustment as necessary, of the required alarm, inteth>ck, and trip setinints so that the setpoints are within the required range and accuracy, MO/E-shall conespond to any one inclusive combination of core reactivity condition, power level, nverage reactor coolant temperature, and reactor vessel head closure bolt tensioning specified below with fuel in the reactor vessel, OPERAlfLL' O/'ERAll/L/IT-system, subsystem, train, component, or device is OPERABLE or has OPERABILITY when it is capable of perfonning its specified safety function (s) and when all necessary attendant instrumentation, controls, nonnal or emergency electrical power, cooling and seal water, lubrication, and oll er auxillary equipment that are required for the system, subsystem, train, component, or 3

device to perfonn its specified sidcty function (s) are also capable of perfonning their related suppon function (s),

16J,2 Combined 1.icense Information I

Combined License applicants referencing the Ap600 will develop a procedure to control the operability of investment protection systems, structures and components in accordance with Table 163 2.

NRC 10'2/97 7

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION s

hiODES

% RATED AVERAGE REACTIVITY Ti(ERhtAL REACTOR C(X)LANT h1 ODES TITLE CONDITION IV)WERUd.

TEh1PERATURE

( K,,)

('F) i 1

. Power Operation

' 2 0.99-NA 2

Startup 2 0.99 s$

NA 3

llot Standby

< 0.99 NA

> 420 4

Safe Shutdown (h)

< 0.99 NA 420 2 T.n > 200 f

5 Cold Shutdown (b)

< 0.99 NA s 200 6

- Refueling (C)

NA NA.

NA i

(a) Excludirig decay heat, (b) All reactos vessel head closure bolts fully tensioned.

[

(c) One or more reactor sessel head closure bolts less than fully tensioned.

B 4

NRC 10/2/97 8

[ W8Silngh0088 3

5

-1...d,

.,b.,

.~.--...#-

RESPONSES TO NRC REOUEST FOR ADDITIONAL INFO.MATION j

id t

l Table 1631 LIST Ol' INVESTMENT PROTECTION SHORT. TERM AVAILAlllLITY CONTROLS Spirms, Structures. Components Number MODES Trains (a)

Operation (b) 1.0 instrumenuition Systerns 1.1. DAS ATWS Mitigation 2

1 1.2 DAS ESF Actuation 2

1,2,3,4,5,6 (3)

_i 2.0 Plarit Systems 2.1 RNS 1

1,2,3 1

2.2 RNS MCS Open 2

5,6 (2,3) l 23 CCS HCS Open 2

5,6 (2.3) l 2.4 SWS RCS Open 2

5.6 (2J) 2.5 PCS Water Makeup Long Tenn Shutdown 1

1,23,4,5,6(4) 2.6 MCR Cmling. Long Tenn Shutdown i

1,2,3,4,5,6 2.7 I&C Room Cooling. Long Term Shutdown i

1.2.3,4,5,6 2.8 Ilydrogen Ignitors i

1,2,5,6 (2.3) 3.0 Electrical Power Systems e

3,1 AC Power Supplies 1

1,2,3,4,5 l

3.2 AC Power Supplies MCS Open (1) 5,6 (2,3) 33 AC Power Supplies. Long Tenn Shutdown

'l 1,2,3,4,5,6 3.4 DC Power Supplies. DAS 2

1,2J,4,5,6 (3)

Aloha Notes:

(a) Refers to the number of trains covered by the availability controls.

(b) Refess to the MODES of plant operation where the availability controls apply, Notes:

(1) 2 of 3 AC power supplies (2 standby diesel generators and I of fsite power supply),

I (2) MODE 5 with RCS open.

(3) MODE 6 with upper internals in place and cavity level less than full, (4) MODES 5 and 6 with the calculated core decay heat greater than 6 Mwt.

NRC 10/2/97 9

.. _,,. _ _ _ _ _ ~.. _. _

l RESPONSES TO NRC REQUECT FOR ADDITIONAL INFORMATION Tattle 163 2 l

INVESTMENT PROTI:CTION SilORT. TERM AVAILAllti.lTY CONTROLS 1.0 flottumentation Systems 1.1 Diverse Actuation Systern (DAS) ATWS hiitigation OPERAlllLITY:

DAS ATWS mitigation function listed in Table 1.1 1 should be operable APPLICABILITY:

htODB i ACTIONS CONDITION REQUIRED ACTION COhfPLETION Tih18 I

A.

DAS ATWS Function A.I Notify [ chief nuclear offlect) or 72 houts with one or more required

[on call allemate),

chantwls inogerable.

AND A.2 Restore required channels to 14 days operable status, 11 Required Action and 11.1 Submit repon to lchief nuclear iday associated Completion offleer) or (on call attemate)

Time of Condition A not detailing interim compensatory

met, measures, cause for inoperability, an. schedule for restoration to

~

OPERABLE.

AND I

11.2 Document in plant records the-

-I month justillcation for the actions taken I-to restore the function to 1

OPERAHl.E.

NRC 10/2/9710 3 W85tingh0086

I ti.

Icth:le:1 Specificatiorn Table 16.3 2 (Cont.)

l INVESTMENT PROJECTION SilORT.TEltM AVAll.AlllLITY CONTROL.S SURVEll. LANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 1.1.1 l'erfonn Cil ANNEL CllECK on each required channel.

30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> SR I.l.2 Pert'onn CilANNEL OPERATIONAL TEST on each required 92 days thannel.

SR I.1.3 Perfonn CilANNEL CAllBRATION on each required 24 months thannel.

Table 1.1 1. IIAS ATWS Functions DAS Initiating Number Channels Setpoint Function Signal Installed Required Rai Drive MG SG Wide 2 per SO 1 per SG

> [25 (XX) Ib]

Set Trip. Turbine Range Level Trip iual PRIIR llX Actuation NRC 10/2/9711

16. Tcch ic:1 Specific tions Table 16.3 2 (Cont.)

i INVESTMIENT PROTECTION SilORT.TERS1 AVAll.Allit.ITY CONTROI.S 1.0 Instrumentation Systems 1.1 DAS ATWS hiitigation llASES:

The DAS ATWS mitigation function of reactor trip, turbine trip and passive residual heat removal heat enhanger (PRl!R llX) actuation should be available to provide ATWS mitigation capability.

'this function is imponant based on 10 CFR $0.62 (ATWS Rule) and because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety related SSCs to miti ate at power and F

I shutdown events. The margin provided in the PRA study assumes a minimum availability of 907c for 1 this function during the MODES of appilcability, considering both maintenance unavailability

,I and failures to actuate.

The DAS uses a 2 out of 2 logic to actuate autornatic functions. When a required channel is unavailable the automatic DAS function is unavailable. SSAR section 7.7.1.11 provides additional infonnation. The DAS charmels listed in Table 1,1 1 should be available.

Automated operator aids may be used to facilitate perfonnance of the CilANNEL CllECK. An automated tester may be used to facilitate perfonnance of the CilANNEL OPERATIONAL TEST.

The DAS ATWS initlyation function shouhl be available during MODE I when ATWS is a Ilmiting event. Planned maintenance affecting this DAS function should be perfonned MODES 3,4,5. 6; these MODliS are selected because the reactor is tripped in these MODES and ATWS can not occur.

NRC 10/2/9712 W Westinghouse

j ij;;

t!!![

16. lattkt:I Spnific9kms 1

(

Table 163 2 (Cont.)

l l

1 INVESThtENT l'ROTECTION SilORT.TER$1 AVAll.Allit.lTY CONTROL.S l

1.0 hntnlutenfallon Systenis l

1.2 DAS Engineering Safeguants Features Actuation (ESFA) 1 OPERADILITY:

DAS ESFA functions listed in Table 1.21 should be operable APPLICAlllLITY:

h10DE 1, 2, 3, 4, 5, h10DE 6 with upper intemals in place vid cavity level less than full ACTIONS CONDITION REQUIRED ACTION COhtPLETION Tih1E A.

9AS ESFA Functions A.I Notify [ chief nuclear ollicerj or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with one or inore required lon-call altentatel.

thannels inoperable.

AND A.2 Restore required chaluicis to 14 days operable status.

IL Required Action and B.I Subinit report to [ chief nuclear 1 day associated Completion of ficer] or [on-call attemate]

Time of Condition A not detailing interim compensatory

met, measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND i

B.2 Document in plant records the i month justification for the actions taken i

in restore the function to l

OPERAllLE.

NRC 10T9713

16. Tecicleel Specificztions j

TaNe 16.3 2 (Cont.)

i I

INVESTMENT l'ROIECTION Sil0RT. TERM AVAll.AHil.ITY CONTROLS l

1 l

SURVEILLANCE REQUIREMENTS SURVEll LANCE FREQUENCY i

SR I.2.1 Perform CilANNEL CliECK on each mquired 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> CllANNEL.

l 4

t SR 1.2.2 Perf0nn CilANNEL OPER ATIONAL TEST on each 92 days required CilANNEL.

SR 1.2.3 Perforn CllANNEL CALIBRATION on each required 24 Inonths CilANNEL f

i Table 1,21, ilAS ESFA Functions IIAS Initiating Number Channels Setpoint l' unction Signal Installed Required PRilR llX SO Wide 2 per 50 I per SO

> [2$.000 lb)

Actuation Level or llo Temp iperliL 1 per liL

< [620lF r

CMT Actuation Pir Level 2

2

> [7]%

and RCI' trip 1%sive Cont.

Cont. Temp 2

2

< [200lF Croling and Selected Cont.

Isolation Actuation b

NRC 10'2/9714 N

. +. -

,,m-.--.

..- - -.m

m

16. Tetknical Spectrications Table 16.3 2 (Cont.)

l INVENTMINT PHOTITTION SHOWT.TI:RM AVAILAllII.ITY CONTROL S i

i 1.0 Imtmmentation Sptems 1.2 DAS ESFA i

I IIASES:

l The DAS ESFA functions listed in Table 1.21 should be available to pnivkle accident midgation I

capability. This function is imponant because it provides margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at power.md shutdown events.

The I margin provided in the PRA study assumes a minimum availability of 90% for this function during

' I the MODES of applicability, considering both maintenance unavailability and failures to actuate.

t The DAS uses a 2 out of 2 logic to actuate automatic functions. When a required charmel is unavailable the automatic DAS function is unavailable. SSAR section 7.7,1,11 provides additional j

infonnation. The DAS channels listed in Table 1.21 should be available.

i 1

Automated operator aids may be used to facilitate perfonnance of the CllANNEL CllECK. An automated tester may le used to facilitate perfonnance of the CilANNEL OPERATIONAL TEST.

7 I

1he DAS ESFA mitigation functions should be available during MODES 1,2,3,4,5.6 when accident mitigation is beneficial to the PRA results. The DAS ESFA should be available in MODE 6 with upper intemals in place and cavity level less than full. Planned raalntenance af fecting these DAS functions should be perfonned in MODE 6 when the refueling cavity is fullt this MODE is selected t

because requiring DAS ESFA pre not anticipated in this MODE.

b i

1 4

r NRC 10/2/9715 W hiinghouse 4

s m -

~

.v.,ie.,,,w,,,,-

www~

- -,',.,'.,.r.

a e,,-

w, 7

.e_... _...,,,, -.

I

16. Tech:letl Specincetions i

Table 16.3 2 (Cont.)

' l INVESTMENT PROTECTION SHORT.1ERM AVAll.AHILITY CON 1ROLS 2.0 l'lant Systerns-2.1 Nonnal Residual liest Removal System (RNS)

Ol'EHAlilL11Y:

One train of RNS injection should be operable Al'PLICAlllLITY:

MODE 1, 2, 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required train not A.I Notify [ chief nuclear officer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operable.

[on-call attemate).

i AND A.2 Restore one train to operable status 14 days

11. Required Action arul 11.1 Submit report to [ chief nuclear i day associated Completion ollicerj or (on-call attemate]

Time not met.

detailing interim compensatory measures, cause for inoperability, arnt schedale for intorr.!on to Ol'ER AllLE.

AND 1

B.2 Document in plant ncords the i month justification for the actions taken ll to restore the function to I

I OPERAHLE.

t NRC 10/2/9716 3 WOStingh0084 j

-c-++

1. 4--

w

,.-,,y-=

y y-i-,,L g-,,*-+vy-,-e-r--..

,r, y,-e-,w

---

w=-r,g-3,, - - '

'T F'

'"~f

• 9"

T"F^

T

'"P"7

t

16. Tubale:1 Spninciskens t

?

Table 16.3 2 (Cont.)

l f

INVESTMENT PHOTECTION SHORT. TERM AVAILAlllLITY CONTROLS I

L SURVE11 LANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY t

SR 2.1.1 Verify that one RNS pump develops a differential head of 92 days

[330) feet on recirculation flow t

i SR 2.1.2 Verify that the following valves stroke open 92 days l

RNS V0ll RNS Discha.ge Cont. Isolation i

RNS V022 RNS Suction lleader Cont, Isolation a

RNS V023 RNS Suction from IRWST isolation f

b i

9 i

P g -

NRC 10'2/9717 7

g

.w..

, '. 4%

r-,,,

,ri,.-

n--,

,...,vn.,,

. ~ -, - -

,,-v,-

i f

16. Tschiltr.1 Specificolons a.

~

Table 16.3 2 (Cont.)

i I

INVESTMENT PROTECTION SHORT. TERM AVAll.Afill.lTY CONTROt.S.

l 2.0 Plant Systerns i

2.1 RNS l

h lIASES:

'1he RNS injection function provides a nonsafety related means of injecting IRWST water into the f

RCS following ADS actuations. The RNS injection function is imponant because it provides margin l

in the PRA sensitivity perfonned assuming no credit for nonsafety related SSCs to mitigate at power

~

and shutdown events. The margin provided in the PRA study assumes a minimum availability of

1. Wi% for this function during the MODES of applicability, etmsidering both maintenance I unavailability and Iallures to operate.

One _ train of RNS injection includes one RNS pump and the line from the IRWST to the RCS Threc of the valves in the line between the IRWST and the RCS are nonnally closed and need to be opened i

to allow injectf an. 'this equipment does not nonnally operate during h10 DES 1,2,3. Refer to SSAR section 5.4,7 for additional infonnation on the RNS.

i The RNS injection function should be available during MODES 1. 2. 3 because decay heat is higher and the need for ADS is greater.

Phuuied maintemuice on redundant RNS SSCs should be perfonned during MODES 1,2. 3. Such maintenance shouhl be perfonned on an RNS SSC not required to be available, The bases for this recommendation is that the RNS is more risk important during shutdown MODES when it is nonnally operating than during other MODES when it only provides a backup to PXS injection.

i Planned maintenance on non redundant RNS valves (such as Voll, V022 V023) should be perfonned to minimite the impact on their RNS injection and their contahunent isolation capability. Non-pressure boundary maintenance shouhl be perfomied during MODE 5 with a visible pressurizer level or MODE 6 with the refueling cavity full, in these MODES, these valves need to be open but they do not need to be able to close. Contakunent closure which is required in these MODES can be satisfied by one nonnally open operable valve, Pressure boundary maintenance can not be perfonned during MODES when the RNS is used to cool the core, therefore such maintenance should be perfonned during MODES 1,2,3. Since these valves are also containment isolation valves, maintenance that renders tu valves inoperable requires that the containment isolation valve located in series with the inoperable valve has to closed and de activated..The bases for this recommendation is that the RNS is more risk imponant during shutdown MODES when it is nonnally operating tlum during other MODliS when it only provides a backup to PXS injection. In addition, it is not possible to perfonn pressure boundary maintenance of these valves during RNS operation.

NRC 10/2/9718 T watinghoun j

rw--

w y

y.~,.

.w,,

y y.y

,_,__,,mg

.r.,,_,,e,.,,,,,

yy.,_,7,

,_mm,,,,,,.,,

,_m, 7

i 16.1edakal Specifkatic.ms j

TaNe 16.3 2 (Cont.)

{

i I

INVESTMENT PROTECT!ON SHORT. TERM AVAll.Allit.ITY CON 1NOI.S

-_2.0 Plant Systems

}

l 2,2 Nonnal Residual lleat Removal System (RNS). RCS Open OPER AlllLITY;-

lloth RNS pumps siiould be operable for RCS cailing l

l _ APPLICAlllLITY:

MODE,9 with RCS pressure imundary open.

MODE 6 with upper intemals in place and cavity level less dwi full f

i ACTIONS -

3 CONDITION REQUIRED AcrinN COMPLETION TIME A.

One pump not operable.

A,1 Remove plant Inim applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i

MODES e

i

!),

Required Action and il 1 Submit report to [ chief nuclear iday assalated Completion of flect] or [on-call alternate) t Time not met, detailing interim compensatory l

measures, cause for inoperability.

and schedule for restoration to OPERAllLE, AND I

it.2 Document in plant records the I month justification for the actions taken

-l to restore the function to

- 1 OPERAHl.E.

a q

4 k

4 ':. S ensue-NRC !0/2/97.19

.--.a

.-,..,,e.~.n-

,,.,e

,.,m-,_$,,,_me s

,a,,

.,m,,,wwaw, w.w

,,m_.,

n.-s w

-n--

r g

-r

16. Tech:Irel Sperinc:lle Table 163 2 (Cont.)

i INVESTMENT PHOTECTION SHORT 1ERM AVAILANil.ITY CONTHOI.S l

SURVEll. LANCE REQUIREMENTS i

SURVEILLANCE IHEQUENCY-l SR 2.2.1 Verify that one RNS pump is..i operation and that each Within i day prior to RNS pump operating individually circulates reactor entering the MODES I

coolant at a flow > 19(0] gpm of applicability OR l

?

Verify that tvth RNS pumps are in operation and circulating reactor conhuit at a ibw > [1300) ppm 4

t i

i i

I NRC 10"L'97;20~

N8IM I

,.,c.

,,. -..--4

....,-....._.,,,#y.

g.,-m

,,....__,,e.m_..-

.,c.

. _ _,,,..... ~.,.

.m

?i Iri, f ecicital Specifkclkms

]

l Table 16.3 2 (Cont.)

i INVESih1ENI pHOfEC110N SilORT.1LRhl Aull.Allit.lTT CONTROL.S 2D Plant Systems i 2.2 RNS - RCS Open llASES:

The RNS cooling function prmides a nonsafety related means to nonnally cool the RCS dunng I shutdown operations (h10 DES 4,5,6). This RNS cooling function is important during conditions I when the RCS pressure boundary is open and the refueling cavity is not flooded because it reduces the pnibabilky of an initiating event due to loss of RNS cooling and because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety related SS('s to mitigate at.

I power and shutdown events. The RCS is considered open when its pre.aure boundary is not I capable of being re established from the control rman. The HCS is also considered open if there i is n,i 6whle level in the pressurlier, The margin provided in the PRA study assmnes a minimum I avai',abilliy if W1 for this function during the $10 DES of applicabillly, considering hath I maintenance unavailability and failures to operate.

lhe RNS cooling of the RCS involves the RNS suction line from the RCS l'!., the two RNS pumps and the RNS discharge line retuming to the RCS through the DVI lines.1hc valves located in these I lines should be open prior to the plant entering these conditions. One of the RNS pumps has to be operating; the other pump may be operating or may be in standby. Standby includes the capability of being able to tic placed into operation from the main contnd room. Refer to SSAR section $A.7 for additional infonnation on the RNS.

I Iloth RNS pumps should be available during the $10 DES of applicability when the loss of RNS I cooling is risk important. If both RNS pumps are not available, the plant should tot enter these I conditions. If the plant has entered these conditions, then the plant should take ac son to restore i system operation or lease the 510 DES of applicability, planned maintenance allecting this RNS cooling function should be perfonned in h10 DES 1,2,3 when the RNS is not nonnally operating. The bases for this mcommendation is that the RNS is mow I risk imponant during shutdown h10 DES, especially during the 510 DES of applicability conditions than during other h10 DES when it only provides a backup to PXS injection.

NRC 10'2/97 21

16. Techtle:1 Specincations Tatile 16,3 2 (Cont.)

l INVESTMENT PROTECTION SHORT 'IERM AVAILAtill.lTY CONTNol.S 2.0 Plant Systerns l 2.3 Component Cooling Water System (CCS). RCS Open i

f OI-ERAlllLITY:

llo'n CCS pamps should te operable for RNS cooling l

1 APPLICAlllLITY:

$.10DE 5 with RCS pressure boundary open, h10DE 6 with upper intemals in place and cavity level less than full ACTIONS CONDITION REQUIRED ACTION COh1PLET10N TIME i

A.One pump not open'ble.

A.1 Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i

MODBS i

t Required Action and 11.1 Submit report to (chief nuclear iday associated Comf etion officer) or (on call alternate]

l Time not met, detal',ing interim compensatory measures, cease for inoperability, and schedule for restoratbn to OPERAllLE, AND I

11.2 Document in plant records the i mor.th justification for the actions taken i

l to rcsture the function to I

OPERAllt.E.

NRC 10T97.'t T woninghouse I

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

..,.,,,,, ~ _. _...

.,.,.,.,_..,4._,-,,.-,,,,

16. Technicci Sperinc.stlorn Table 16.3 2 (Cont.)

l INVENTMENT Pl;OTECTION ~;llORT.1ERM AVAll.Allli.ITY CONT ROI.S SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 2.3.1 Verify that one CCS purnp is in operation aid each CCS Within ! day prior to i

putnp operating individually provides a CCS flow through entering the MOI)ES I

one RNS heat exchanger > [2520] pprn of applicability OR Verify that both CCS pumps are in operation and the CCS 110w through each RNS heat exchanger is > [2520] pprn r

NRO 10t97 23

16. Technical Specincctions Table 16.3 2 (Cont.)

l INYF.STAIENT PROTECTION SilORT TEtth! AVAILAllit.lTY CONTROLS 2.0 Plant Splems I

. 3 CCS RCS Open 11ASES:

1he CCS cooling of the RNS IlXs provides a nonsafety related means to nonnally cool the RCS during shutdown operations (MODES 4,5,6). This RNS cooling function is impertant because it reduces the probability of an initiating event due to loss of RNS cooling and bec"Jse it provides margm in the l'R A sensitivity perfonned assuming no credit for nonsafety related SSCs to mitigate at-I pmer and shutdown events. The RCS is considered open when its p. essure boundary is not I capable of hting re established from the control room. The RCS is also considered open if there I is no sisable lesel in the pressurlier. The margin provided Iri the PRA study assumes a minimum i availability of 90% for this function during the hf 0 DES of applicability, considering both I maintenance unanilability and failures to operate.

The CCS cooling of the RNS involves two CCS pinnps and ilXs and the CCS line to the RNS IlXs '

1hc valves around the CCS pumps arul llXs and in the lines to the RNS IlXs should be open prior to I the plant entering these conditions. One of the CCS pumps and its llX has to be operating. One of the lines to a RNS IlX also haa to be open. The other CCS pump and ilX may be operating or may be in standby. Sta1xiby includes the capability of being able to be placed into operation from the main control room. Refer to SSAR section 9.2.2 for additional infomiation on the CCS.

i lloth CCS pumps should be available during the MODES of applicability when the loss of RNS 1 cooling is risk importuit, if both CCS pumps are not available, the plant should not enter these I conditions. If the plant has entered these conditions, then the plant should take action to restore both I CCS pumps or to leave these conditions.

Planned maintenance af fecting this CCS cooling function should be perfonned in MODES 1,2,3 when the CCS is not supponing RNS operation.1he bases for this recommendation is that the CCS is I more risk imponant during shutdown MODES, especially during the MODFS of applicability conditier.s than during other MODES.

NRC 10'2/97 24 W Westinghouse

EIN

16. Tethuic 1 Spetiric tions Table 16.3 2 (Cont.)

1 INVESTMENT PROTECTION SHORT T'tRM AVAILAllil.ITY CONTROL.S 2.0 Plant Systems 1 2.4 Service Water System (SWS) RCS Open OPERAlllLITY:

lloth SWS pumps and cooling tower fans should be operaNe for CCS cmling l APPLICADILITY:

MODE $ with RCS pressure boundary open.

MODE 6 with upper internals in place and cavity level less than full ACTIONS CONDITION REO' iED ACTION COMPLETION TIME '

A.

One pump or fan not A.1 Rer

. plant fium applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operable.

M' dS e

'i. Required Action and 11.1 Submit irport to [ chief nuclear -

1 day awwiated Completion -

ofilcer) or [on call alternate)

Time not met, detailing interim compensatory measums, cause for inoperability.

and schedule for restoration to OPERAllLE.

AND I

11.2 Document in plant records the I month justillcation for the actions taken I

to restore the function to I

OPERAllt.E.

a w

NRC 10/2/97 25 4 +

w w-e-w s

e

.+t i

--ye-,-,+

w r 2 r w - w-e-

w

---r-,r,:---or

t

15. Tech:le:1 Specification Table 16.3 2 (Cont.)

f 1

INVESTMENT PROIECTION SHORT.TF*dt AVAll.AHil.lTY CONTROL.S i

SURVEILLANCE REQUIREMENTS SURVEILLANCli FREQUENCY SR 2,4,1 Verify that one SWS pump is operating and that each Within I day prior to I

SWS pump operating individually pmvides a SWS Ilow entering the MOI)ES i

l

> [6200) ppin of applicability i

l SR 2,4,2 Operate each cooling tower fan for > 15 min Within i day prior to I

entering the MOI)ES of applicability l

i t

r 1

NRC 10'2/97 26 T Westinghouse t

-n--

r n-.,,

__n.-.m,-

a s

s

p

16. Teticid Specific tionn h

Table 163 2 (Cont.).

l INVESTMENT pHOTECllON SHORT. TERM AVAll.Altll.ITY CONTNot.S 2.0 Plant Systems l

l 2.4 SWS RCS Open l

i 11ASES:

r lhe SWS cooling of the CCS IlXs provides a nonsafety related means to.onnally cool the RNS IlX which cool the RCS during shutdown operationa (MODES 4,5,6). This RNS cwling function is imponant because it reduces the probability of an initiating event due to loss of RNS cooling and i

because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafetr related l SSCs to mitigate at power and shutdown events. The RCS 15 cimsidered open when its pressure houndary is not capable of being re established from the control room. The RCS is also

- l considered open if there is no visable lesel in the pressuriter. The margin provided in the PRA 1

I study assumes a minimum availability of 90% for this function during the MODES of applicability, I considering both maintenance unavailabliity and failures to operate.

?

r The SWS uxiling of the CCS IlXs involves two SWS pumps and cooling tower fans and the SWS l. line to the RNS IlXs The valves in the SWS lines should be open pdor to the plant entering these 1 etmditiont. 0:n of the SWS pumps and its cooling tower fan has to be operating. The other SWS pump and cooling tower fan may be operating or may be in standby. Standby includes the capability of being able to be placed into operation fnim the main control room. Refer to SSAR section 9.2.1 for additional infonnation on the CCS.

l Hoth SWS pumps and cooling tower fans should be available during the MODES of applicability i

when the loss of RNS cooling is risk imponant. If tuth SWS pumps and cooling tower fans are not I available, the plant shouhl not enter these conditions. If the plant has en'ered there conditions, then I the plant shouki take action to restore both SWS pumps / fans or to leave these conditions.

pituuted maintena: ice affecting this SWS cooling function should be perfonned in MODES when the SWS is not supponing RNS operation,le during MODES 1,2,3 lhe bases for this recommendation I is that the SWS is more risk imponant during shutdown MODES, especially during the MODES of I applicability conditions than during other MODES, 4

I t

l NHC 10297 27 i

we--,

v.r-..-

.-v

.c.-

-.,.y

,,.r.,

.,,.e,_.

,_.~,e y

e

-w.

,-e-e.

e

- r

. - = -

.<.nv

<iv-v

-ww-w-

16. Teth: loc.l Specincations Table 163 2 (Cont.)

l INVI SThfENT l'ROTI:CIION Siluftl.11'Rht AVAILAlllLITY CONTROL.S 2.0 Plant Systems 2.5 l'assise Contailunent System Water Storage Tsuik (PCCWST) hiakeup Long Tenn Shutdown OPERAillt.lTY:

Long tenn makeup to the PCCWST should be operable Al' PLICA 13ILIIT:

h10 DES 1, 2, 3,4, h10 DES $ and 6 with calculated co decay heat > 6 htWI.

ACTIONS CONDITI' N REQUIRED ACTION COhtPLETION TlhtE O

A.

W:.ter volume in PCS A,1 Notify khlef nuclear officer} or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ancillary tank less than

[on call attemate).

limil.

AND A.2 Restore volume to within limits 14 days II.

One required I'CS II,1 Notify [ chief nuclear ollicer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> recirculation pump not

[on call attemate].

operable.

AND II.2 Restore pump to operaNe status 14 stays NRC 10'2/97 28 T Westinghouse

16. Tuhnitt Spninttions TaNe 16.3 2 (Cont.)

l INVEN1 MENT Pit 01ECTION Silultt.'IEltM AVAll.AllilIlY CONTitol.S C.

Required Action and C.l Submit repon to [ chief nuclear i day anociated Completion ofilcerl or [on-ca:1 alternate]

Titne of Condition A, il detailing intetim cornpensatory not met.

measures, cause for inoperability, and schedule for restoration to OPERABl.E.

AND I

C.2 Document in plant records the I month justification for the actions taken I

to restiire the function to 1

OPERAlli E.

r SURVEll. lab'CE REQlllREMENTS SURVEll.l.ANCE FREQUENCY SR 2.5.1 Verily water volume in the PCT ancillary tank is 31 days

> 1362JX01 gal.

SR 2.5 2 Record that the required PCS recirculation pomp pnnides 92 days recirculation of the PCCWST at > 162) ppm.

SR 2.5.3 Verify that each PCS recirculation pump transfers 10 years

> [62] gpm fnim the PCS ancilla:y tank to the PCCWST.

During this test, each PCS retircu. tlon pump will be powered from a ancillary diesel.

NRC 10'2/97 29

y uu i

!i g

16. Technical Specifications Table 163 2 (Cont.)

I

. INVESTMENT PROTECTION SilORT. TERM AVAILAlllLITY CONTROLS 2.0 Plant Systems 2.5 PCCWST Makeup Long Term Shutdown i

BASES; Vie PCS recirculation pumps prt.. de long tenn shutdown suppon by transferring water from the PCS ancillary tank to the PCCWST. This water is used to maintain PCS coo'i.ig du.ng the 3 to 7 day time period following an accident. After 7 days water brought in from offsite allows the PCCWST to continue to provide PCS cooling and also to provide makeup to the spent fuel pit. This PCCWST I makeup function is important because it supports long-tenn shutdown operation. A minimum I aiailability of 90% is assumed for this function during the MODES of applicability, considering i both maintenance unavailability and failures to operate.

The PCCWST makeup function involves the use of one PCS recirculation pump, the PCS ancillary tiuik and the line connecting the PCS ancillary tank with the PCCWST. One PCS Teirculation pump nonnally operates to recirculate the PCCWST. Refer to SSAR section 6.2.2 for additional infonnatior/

on the PCCWST makeup function.

I The PCCWST makeup fur; tion should be available during MODES of operation when PCS cooling is required; one PCS recirculation pump and PCS ancillary tank should be available during MODES l 2,3A, and MODES 5 and 6 with calculated core decay heat > 6 MWt. 'lhe PCS is required to be available during these MODES; it is not required to be available in MODE 5 or 6 with calculated core decay heat < 6 MWt.

Planned maintenance affecting the to pred PCCWST recirculation pump should not be perfonned during required MODES; planned maimenance should be perfomied on the redundant pump (ie the pump not required. N available). Planned maintenance affecting the PCS ancillary tank that requires less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to perfonn can be perfonned in any MODE of operation. Planned maintenance requiring nr.. m 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> should be perfonned in MODES 5 or 6 when the calculated core decay heat is < 6 M'bt e c ~es for this recommendation is that the long-tenn PCS makeup is not required in this condition.

NRC 10/2197 30 W Westinghouse

nr u

n

16. Technic:l Specifications l

i Table 163 2 (Cont.)

l INVESTSIENT PROTECTION SilORT TERM AVAll.AlllLITY CONTROI 2.0 Plant Systems 2.6 hiain Control Room (h1CR) Cooling Long Term Shutdown OPERABILITY:

Long tenn cooling of the MCR should be operable APPLICABILITY:

MODES 1, 2, 3, 4, 5, 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required MCR A.1 Notify [ chief nuclear officer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ancillary fans not

[on-call attematel.

opeiable, e

AND A.2 Restore one fan to operable status 14 days 11.

Required Action and D 1 Submit n port to [ chief nuclear i day associated Completion officerl or [on-call attemate]

1;me not met, detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND 1

B.2 Document in plant records the I month justification for the actions taken I

to restore the function to 1

OPERAllLE.

NRC 10/2/97 31

E

~ {

1

16. Technicti Specifications j

L__u"i Table 16.3 2 (Cont.)

i INVESTh1ENT l'ROTECTION SilORT.TERSI AVAILAllILITY CONTROLS SURVEILLANCE REQUIREh1ENTS SURVEILLANCE FREQUENCY SR 2.6.1 Operate required h1CR ancillary fan for > 15 min 92 days SR 2.6.2 Verify that each h1CR ancillary fan can provide a flow of 10 years air into the $1CR for >l5 min. During this test, the htCR ancillary fans will be powered from the ancillaq diesels.

NRC 10.2!97 32 3 Westh ghouse

I

??

iti

16. Tcch:ic:l Specine tions j

Table 16.3-2 (Cont.)

i INVESTMENT l'ROTECTION SilORT TERM AVAILAlllLITY CONTROLS 2.0 Plant Systems 2.6 MCR Cooling - Long Tenn Shutdown liASES:

The MCR ancillary fans provide long tenn shutdown support by cooling the main control room. For the first three days alter an accident the ernergency ilVAC system (VES) together with the passive heat sinks in the MCR provide cooling of the MCR. After 3 days, the MCR ancillary fans can be used to circulate ambient air through the MCR to provide cooling. De long term MCR cooling function shouhl be available during all MODES of operation. This long tenn MCR cooling function I is important because it suppons long tenn shutdown operation.

A minimum availability of 90% is I assumed for this function during the MODES of applicability, considering both maintenance I unasailability and failures to operate.

The long tenn MCR cooling function involves the use of a MCR ancillary fan. During SR 2.6.1 the fan will be run to verify that it operates without providing flow to the MCR. During SR 2.6.2 cach fan will be connected to the MCR and operated such that thev provide flow to the MCR. Refer to SS AR section 9.4.1 for additional infonnation on the long tenu MCR cooling function.

One MCR ancillag fan should be available during all MOJES of plant operatica. Planned maintenance should not be performed on the required MCR ancillary fan dunng a required MODE of operation; planned maintenance should be perfonned on the redundant MCR ancillary fan (ie the fan not required to be available) during MODES 3 or 4, MODE 5 with a visible pressurizer level or MODE 6 with the refueling cavity full; these MODES are selected becat3e the reactor is tripped in these MODES and the risk of core damage / hydrogen generation is low.

NRC 10."2/97 33

16. Technical Specifications nii Table 16.3-2 (Cont.)

I INVEST 51ENT PROTECTION SilORT TERh1 AVAILAllILITY CONTROLS 2.0 Plant Systems 2.7 I&C Room Cooling - Long Term Shutdown OPERABILITY:

Long term cooling of I&C rooms B & C should be operable APPLICABILITY:

h10 DES 1, 2, 3, 4, 5, 6 ACTIONS CONDITION REQUIRED ACTION CON 1PLETION TIh1E A.

One required I&C room A.1 Notify [ chief nuclear 6fficer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ancillary fan not operable.

[on-call attemate).

e AND A.2 Restore one fan to operable status 14 days B.

Required Action and B.1 Submit report to [ chief nuclear i day associated Completion officerl or (on call altemate)

Time not met, detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND I

B.2 Document in plant records the 1 month justification for the actions taken I

to restore the function to 1

OPERABLE.

NRC 10/2/97 34 W Westinghouse l

l

. ~

. ~ ~ -. - -.....~

16. Tschildt Specificatints:

)

-Table 16.3 2 (Cont.).

.l

-l-INVESTMENT PROTECTION SHORT. TERM AVAILAllILITY CONTROLS i

SURVEILLANCE REQUIREMENTS-SURVEILLANCE FREQUENCY i

92 days S R 2.7.1 Operate required I&C room ancillary fan for > 15 min-i SR 2.7.2 Verify that each I&C room ancillary fan can provide a 10 years flow of air into an I&C room for >l5 min. Dwing this test, the I&C room ancillary fans will be powered from an ancillary diescl; i

l 1

b d

f I

A NRC 10/2/97 35 gggy

- o e.s w--Ap

- e

,n-n

16. Tech:ic:1 Specification Table 16.3-2 (Cont.)

l INVESTMENT PROTECTION SilORT TERM AVAILAlllLITY CONTROLS 2.0 Plant Systems 2.7 I&C Rmm Cooling Long Tenn Shutdown BASES:

he I&C rmm ancillary fans provide long tenn shutdown support by cooling I&C rooms B & C which contain post ar cident instrument processing equipment. For die first three days after an accident the passive L aat sinks in the I&C rooms provide cooling. After 3 days, the I&C room ancillary fans can b., used to circulate ambient air through the I&C room to provide cooling. De long tenn I&C room cooling function should be available during all MODES of operation. This long I tenu l&C room cooling function is important because it supports long-tenn shutdown operation.

A 1 minimum nvailability of 90% is assumed for this function during the MODES of applicability, I considering both maintenance unavailability and failures to operate.

The long tenn I&C room cooling function involves the use of two I&C mom ancillary fans; each fan is associated with one I&C room (B or C). During SR 2.6.1 the required fan will be run to verify that!

it operates without providing How to the l&C room During SR 2.6.2 each fan will be connected to its associated I&C room and operated such that flow is provided to the IkC room. Refer to SSAR section 9.4.1 for additional infonnation on the long tenn I&C room cooling function.

One l&C room ancillary fan should be available during all MODES of plant operation. Planned maintenance should not be perfonned on the required I&C room ancillary fan during a required MODE of operation; phmned maintenance should be performed on the redundant 1&C room ancillary lan.

NRC 10/2/97 36 W Westinghouse

!!!E Hiii

(

16. Technical Specificcthms Table 15.3 2 (Cont.)

1 INVESTNIENT PROTECTION SilORT TERN 1 AVAILAlllLITY CONTROLS 2.0 Plant Systems 2.8 Ilydrogen Ignitors OPERABILITY:

' Die hydrogen ignitors listed in Table 2.81 should be operable APPLICABILITY:

h10DE I,2, I

h10DE 5 with RCS pressure boundary open, h10DE 6 with ut.per intemals in place and cavity level less than full ACI'lONS CONDITION REQUIRED ACTION COh1PLETION Tih1E A.

One or more required A,1 Notify (chief nuclear officer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> e

hydmgen ignitor lon-call altemate).

inoperable.

AND A.2 Restore required ignitors to 14 days operable status.

B.

Required Action and B.1 Submit report to [ chiel nuclear I day associated Completion officer) or [on-call attemate)

Time of Condition A not detailing interim compensatory m et, measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND l

B.2 Document in plant records the I month justification for the actions taken I

to restore the function to I

OPERAllLE, NRC 10,"2/97-37 l

l

p;I T.

16. Technical Specifications Table 16.3 2 (Cont.)

l INVESTMENT I'ROTECTION SilORT TEltM AVAILA111LITY CONTROLS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 2.8.1 Energize cach required hydrogen ignitor and verify the surface Each refueling outage temperature is > [1700) F.

f NRC 10297 38 W Westinghouse

jpH E5!

16. Tech:ic:1 Specifie tions 7

Table 16.3 2 (Cont.)

l INVESThtENT PROTECTION SiiORT.TERhl AVAILAlllLITY CONTROLS Table 2.&l, II drogen Ignitors (1) 3 Location flydrogen Ignitors Group i Group 2 Reactor Cavity note 2 note 2 Loop Companment #1 12,13 11,14 Loop Cornpanment #2 5,8 6,7 Pressuriter Companment 49,60 50,59 Tunnel connecting Loop Companments 1,3,31 2,4,30 Southeast Valve Room & Southeast 21 20 Accumulator Room East Valve Room, Northeast Accumulator 18 17,19 Room, & Nonheast Valve Room Nonh CVS Equipment Room 34 33 Lower Companment Area 22.27,28,29,31,32 23,24,25,26,30 (ChlT and Valve Area)

IRWST 9,35,37 10,36,38 1RWST inlet 16 15 Ret ~ueling Cavity 55,58 56,57 Upper Companment Lower Region 39,42,43,44,47 40,41,45,46,4"'

hlid Region 51,54 52,53

- Upper Region 61,63 62.64 Notes:

1) The table lists the hydrogen ignitors, in each location, all of the ignitors in Group 1 or Group 2 I

should be available. It is not necessary for all of the available ignitors to be in one group.

2) Ignitors in this location are shared with other locations.

W Westinghouse

)-

[

_16. TechIical Specific tion!

i Table 16.3-2 (Cont.)

I INVESTMENT PROTECTION SHORT TERM AVAILAllILITY CONTROLS 2.0 Plant Systems 2.8 flydrogen Ignitors BASES:

The hydrogen ignitors should be available to provide the capability of buming hydrogen generated I : during severe accidents in order to prevent failure of the containment due to hydrogen I detonation. These hydrogen ignitors are required by 10 CFR 50.34 to limit the buildup of

-I hydrogen to less than 10% assuming that 100% of the active zircaloy fuel cladding is oxidized.

' I - This function is also imponant because it provides margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at-power and shutdown events. The margin provided I in the PRA study assumes a minimum availability of 90% for this function during the MODES of I applicability, considering both maintenance unavailability and failures to operate.

1 - The ignitors are distributed in the containment to limit the buildup of h drogen in hical areas. !

3 I Two groups of ignitors are prmided in each areat one of which is sufficient to limit the buildup i of hydrogen. When an ignitor is energized, the ignitor surface heats up to 211700] F. This I temperature is sufficient to ignite hydrogen in the vicinity of the ignitor when the lower i flammability limit is reached. SSAR section 6.2.4 provides additional infonnation.

The hydrogen ignitor function should be available during MODES I and 2 when core decay heat is I high and during MODES 5 when the RCS pressure boundary is open and in MODE 6 when the

.I refueling cavity is not full. Planned maintenance should be performed on hydrogen ignitors I when they are not required to meet this availability etmtrol.

NRC 10/2/97 40

(# Westinghouse.

=-

!jin ::dj

16. Tech:ic:l Specific:tions Table 16.3 2 (Cont.,

1 INVESTMENT PROTECTION SilORT TERM AVAILABILITY CONTROL.S 3.0 Electrical Power Systems 3.1 AC Power Supplies OPERABILITY:

One standby diesel generator should be operable APPLICABILITY:

MODES 1, 2, 3, 4, 5 ACTIONS CONDITION REQUIRED ACTION COhiPLETION TihtE A.

Fuel volume in one A.! Notify [ chief nuclear officerj or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> required standby diesel

[on call altemate].

fuel tank less than limit.

e AND A.2 Restore volume to within limits 14 days B.

One required fuel trar,sfer B.1 Notify [ chief nuclear offleer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pump or standby diesel

[on-call altemate).

generator not operable.

AND B.2 Restore pump a.nd diesel generator 14 days to operable status NF C 10/2/97 41

16. Tech:ic;l Specine:tions Table 16.3 2 (Cont.)

i INVESTMENT PROTECTION SilORT TERM AVAILAlllLITY CONTROLS C.

Required Action and C.I Submit report to [ chief nuclear i day associated Completiot; officer] or [on-call attemate]

Time not met, detailing interim compensatory measures, cause for inoperability and schedub for restoration to OPERABLE.

AND I

C.2 Document in plant records the 1 month justification for the actions taken i

to restore the function to

'I OPERAllLE.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1 Verify that the fuel oil volume in the required standby 31 days diesel generator fuel tank is > [50JKO) gal.

SR 3.1.2 Reconi that the required fuel oil transfer pump provides a 92 days recirculation flow of > [8] gpm.

SR 3.1.3 Verify that the required standby diesel generator starts and 92 days operates at > [3800] kw for > I hour. This test may utillie diesel engine prelube prior to starting and a wannup period prior to loading.

SR 3.l.4 Verify that each standby diesel generator starts and 10 years operates at > [3800] kw for > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This test may utilize diesel engine prelube prior to starting and a wammp period prior to loading. Both diesel generators will be operated at the same time during this test.

NRC 10/2/97 42

^

f Y A'stinghouse

g.

?

y

16. Tech'le;l Specifiestions Table 163 2 (Cont.)

l INVESTN1ENT PROTECTION SilORT TERN 1 AVAILAlllLITY CONTROLS 3.0 Electrical Power Systems 3.1 AC Power Supplies

!!ASES:

AC power is required to power the RNS and to provide a nonsafety related means of supplying power to the safety-related PhtS for actuation and post accident monitoring. The RNS provides a nonsafety-related means to inject water into the RCS following ADS actuations in h10 DES 1,2,3,4 (when steam generators cool the RCS). This AC power supply function is imponant because it adds margin to the PRA sensitivity perfonned assuming no credit for nonsafety-related SSCs to mitigate at-power and I shutdown events. Tbc margin provided in the PRA study assumes a minimum availability of 90% for 1 this function during the h10 DES of applicability, cimsidering both niaintenance unavailability I and failures to operate.

Two standby diesel generators are provided. Each standby diesel generator has its own fuel oil transfer pump and fuel oil tank, The volume of fuel oil required is Inat volume that is above the connection to the fuel oil transfer pump. Refer to SS AR section 8.3.1 for additional information.

This AC power supply function should be available dering h10 DES 1,2 3,4,5 when RNS injection and PhtS actuation are more risk important. Planned maintenance should not be performed on required AC power supply SSCs during a required h10DE of operation; planned maintenance should be perfomied on redundant AC power supply SSCs during A10 DES I,2,3 when the RNS is not nonnally in operation. The bases for this recommendation is that the AC power is more risk I important during shutdown h10 DES, especially when the RCS is open as defined in availability I control 2.2, than during other A10 DES.

NRC 10/2/97-43

11in isti!

16. Tech:ical Specificctiecs Table 16.3 2 (Cont.)

l INVESTMENT PROTECTION SilORT. TERM AVAll.AllILITY CONTROLS 3.0 Electrical Power Systems i 3.2 AC Power Supplies RCS Open OPERABILITY:

Two AC power supplies should be operable to support RNS operation I APPLICA131LITY:

MODE 5 with RCS pressure boundary open.

MODE 6 with upper intemals in place and cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required AC power A.1 Remove plant from applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> supply not operable.

MODES 8

H.

Required Action and B.1 Submit repon to [ chief nuclear i day associated Completion officerj or [on-call attemate]

Time not met.

detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND l

B.2 Document in plant records the i month justification for the actions taken I

to restore the function to I

OPERAHLE.

NRC 10/2/97-44 W Westinghouse

m..

i

]

M. Technic;l SPecificall""5 -'

Table 16.3 2 (Cont.)

4 INVESTMENT I'ROTECTION SHORT. TERM AVAILABILITY CONTROLS -

-l' SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY t

SR 3.2.1 Verify that the required number of AC power supplies are Within i day prior to

-l operable entering the MODES 1

of applicability I

e i

5 i

4 e

NRC 10/2/97 45

=

<,r.-

,-,,.--.,_,,+%

y

-n,,.

16. Tech:icil Specific:tions Table 16.3 2 (Cont.)

l INVE5iTMENT PROTECTION SilORT TERM AVAll.Allil.ITY CON'IROI.S 3.0 Electucal Power Systems I 3.2 AC Power Supplies RCS Open IIASES:

AC power is required to power the RNS and its required support systems (CCS & SWS); the RNS provides a nonsafety-related means to nonnally cool the RCS during shutdown operations. This RNS I cooling function is important when the RCS pressure boundary is open and the refueling cavity is i not ihmded because it reduces the probability of an initiating event due to loss of RNS cooling during these conditions and because it provides margin in the PRA sensitivity perfonned assuming no credit i for nonsafety-related SSCs to mitigate at power and shutdown events. The RCS is etmsidered open I oben its pressure boundary is not capable of being re-established from the control room. The i RCS is also considered open if there is no visable level in the pressurizer. The margin provided in I the PRA stu(ly assumes a minimum availability of 90% for this function during the MODES of I applicability, considering both maintenance unasailability and failures to operate.

t Two AC power supplies, one offsite and one onsite supply, should be available as follows:

a)

Offsite power through the transmisdan switchyard and either the main step-up transfonner

/ unit auxillary transfomier or the reserve auxiliary inmsfomier supply from the transmission switchyard, and b)

Onsite power from one of the two standby diesel go ators.

Refer to SS AR section 8.3.1 for additional infonnation on the standay utesel generators. Refer to SSAR section 8.2 for infonnation on the offsite AC power supply, 1 One offsite and one onsite AC power supply should be available during the MODES of applicability when the loss of RNS cooling is imponant. If both of these AC power supplies are not available, the i plant shouhi not enter these conditions. If the plant has already emered these conditions, then the i plant should take action to restore this AC power supply function or to leave these conditions.

Planned maintenance should not be perfonned on required AC power supply SSCs, Planned maintenance affecting the standby diesel generators should be performed in MODES 1,2,3 when the RNS is not nonnally in operation. Phmned maintenance of the other AC power supply should be performed in MODES 2,3 or MODE 6 with the refueling cavity full. The bases for this recommendation is that the AC power is more risk imponant during shutdown MODES, especially I during the MODES of applicability conditions than during other MODES.

NRC 10'2/97 46 3 Westinghouse

16. Technic:1 Specincetions -

Tab!c 16.3 2 (Cont.)

1

- INVESTMENT PROTECTION SHORT TERM AVAILABit.lTY CONTROLS

-3.0 Electrical Power Systems 3.3 = AC Power Supplies Long Tenn Slautdown OPERABILITY:.

One ancillary diesel generator should be operable APPLICAlllLITY:

MODES 1, 2, 3. 4, 5, 6

- ACTIONS '

CONDITION REQUIRED ACTION' COMPLETION TIME A.

Fuel volume in ancillary A.I _ Notify [ chief nuclear officerl or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> diesel fuel tank less than -

[on-call attemate).

limit.

e i

AND A.2 Restore volume to within limits 14 days

11. One required ancillary B.I Notify [ chief nuclear officer) or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> diesel generator not

[on-call attemare).

operable.

AND B.2 Restore one diesel generator to 14 days operable status NRC 10/2/97-47

IE

!j

16. Teciticil Specific tions TaNe 16.3 2 (Cont.)

l INVES1 MENT l'HOTECTION SilORT. TERM AVAILAllILITY CONTROLS C.

Requ red Action and C.1 Submit report to [ chief nuclear i day associated Completion officer] or [on-call attemate)

Time not met.

detaillag interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE.

AND l

C.2 Document in plant records the 1 month justification for the actions taken I

to restore the function to I

OPERAllLE, f

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1 Verify fuel volume in the ancillary fuel tank is >[350] gal 31 days SR 3.3.2 Verify that the required diesel generator starts and 92 days operates for >l Mur connected to a te load > [14.51 kw.

This test may utilize diesel engine warmup perial prior to loading.

NRC 10.'2/97 48 W Westinghouse

i!W Eiii

16. Tech:iccl Specific:tio:s

}

Table 16.3 2 (Cont.)

l INVESTMENT PHOTECTION SilORT. TERM AVAILAlllLITY CONTROI,S SR 3.3.3 Verify that each diesel generator starts and operates for 4 10 yeau hours while providing power to the regulating transformer, an ancillary control room fan, an ancillary I&C room fan and a passive contairunent cooling water storage tank recirculation pump that it will power in a long tenu post accident condition. Test loads will be applied to the output of the regulating transformers that represent the loads required for post-accirlent monitoring and control rmm lighting. This test may utilize diesel engine wannup prior to loading. Both diesel generators will be operated at the same tirne during this test.

f W Westinghouse

iig

16. Tech iccl Specific tion Table 16.3 2 (Cont.)

INVESTMENT PROTECTION SilORT TERM AVAll AlllLITY CONTROL.S 3.0 Electrical Power Systems 3.3 AC Power Supplies - Long Tenn Shutdown BASES:

The ancillary diesel pnerators provide long term power supplia for post accident monitoring, MCR and l&C nom cc.oling, PCS and spent fuel water makeup. For the first three days after an accident the IE batteries provide power for post accident monitoring. Passive heat sinks provide coohng of the MCR and the li rooms. The initial water supply in the PCCWST provides for at least 3 days of PCS cooling. The initial water volume in the spent fuel pi normally provides for 7 days of spent fuel cooling; in some shutdon events the PCCWST is used to supplement the spent ftal pit.

A i

l minimum availabili y of W7c is assumcJ for this function during the STODES of applicability, I considering both maintenance unavailability and failures to operate.

After 3 days, ancillary diesel generators can t e used to power the MCR and I&C nom ancillary fans, the PCS recirculation pumps and MCR lighting. In this time frame, the PCCWST provides water makeup to both the PCS amt the spent fuel pit. An ancillary generator should be available during all MODES of operation. This long tenn AC power supply function is important because it supports long teun shutdown operation.

iae long tenn AC power supply function involves the use of two ancillary diesel generators and rn ancillary diesel generator fuel oil storage tank. Refer to SSAR section 8.3.1 for additional infonnation on the long tenn At power supply function.

One anchhtry diesel generator and the ancilhtry diesel generator fuel oil storage tank should be available dunng all MODES of plant operation. Planned maintenance should not be perfonned on the required aacillary diesel generator during a required MODE of operation; planned maintenance should be perfonned or the redundant ancillary diesel generator. Phumed.naintenance affecting the ancillary diesel fuel tank that requires less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to perfonn can be perfonned in any MODE of operation. Planned maintenance requiring more than 72 homs should be perfonned in MODE 6 with the refueling cavity full. The basis for this recommendation is that core decay heat is low and the risk of core damage is low in these MODES, the inventory of the refueling cavity results in slow response of the plant to accidents.

NRC 10/2/97 50 3 Westinghouse

jf :lij

16. Tcchrie:1 Specifktth s l

Table 163 2 (Cont) l INVESTMENT PROTECTION SilORT TERM AVAILAHILITY CONTROLS 3.0 Electrial Power Systems 3.4 DC Power Supplies OPERABILITY:

Power for DAS automatic actuation functions listed in 1.1 and 1.2 should be operable APPLICABILI~IY:

MODES 1, 2, 3, 4, 5.

MODE 6 with upper intemals in place and cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Power to DAS Function A.1 Notify [ chief nuclear officer] or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> e

inoperable.

[o.1-call altemate).

AND A.2 Restore power supply to DAS to 14 days operable status B,

Required Action and B,1 Submit report to [ chief nuclear i day associated Completion officer] or [on-call altemate]

Time of Condition A not detailing interim compensatory m et.

measures, cause for inoperability, and schedule for restoration to OPERABLE AND 1

B.2 Document in plant records the i month I

justification for ti,e actions taken I

to restore the function to 1

OPERAllt.E.

NRC 10/2/97 51

16. Tech:Ical Specille:tio:s Table 163 2 (Cont.)

l INVESTMENT l'HOTECTION SilORT. TERM AVAILAllit.ITY CONTROL.S SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1 Verify power supply voltage at each DAS cabinet is 92 days 120 volts 5%

f NRC 10.'2/97 52 C'

(W-s) WC5tlnE!10U30

?

16, Technic:1 Specific: lions Table 16.3-2 (Cont.)

INVESTMENT PROTECTION SilORT TERM AVAll Allit.lTY CONTROLS 3.0 Electrical Power Systems 3.4 Non Class IF DC md UPS System (EDS)

BASES:

The EDS function of pic iding power to DAS to support ATWS mitigation is !mportant based on 10 CFR,50.62 (ATWS Rule) und to support ESFA is important based on providing margin in the PRA sensitivity perfonned assuming no credit for nonsafety-related SSCs to mitigate at-power and i shutdown events. 'Ihe margin provided in the PRA study assumes a minimum availability of 90% for I this function during the MODES of applicability, considering both maintenance unavailability I and failures to operate.

The DAS uses a 2 out of 2 logic to actuate automatic functions. EDS power must be available to the DAS sensors DAS actuation, and the devices which control the actuated components. Power may be provided by EDS to DAS by non lE batteries through non lE inverters. Other means of providing power to DAS include the spare battery through a non lE inverter or non lE regulating trartformers. '

The EDS support of the DAS ATWS mitigation function is required during MODE 1 when ATWS is a limiting event and during MODES 1,2,3,4,5,6 when ESFA is important. The DAS ESFA is required in MODE 6 with upper intemals in place and cavity level less than full. Planned maintenance should not be perfonned on a required EDS SSC during a required MODE of c9eration; phumed maintenance should be perfonned on redundant supplies of EDS power 4

}

l l

NRC 10.7./97 53

- ~.

. =.. - -

- ~, -.

I gBMNg NRC FSER OPEN ITEM a-t i

Qdestion 720.442F (OITS 6180)

Key aspects'or the reactor cavity Gooding system and the containment layout 'need to be confirmed by-ITAAC to assure that the reactor cavity will Good and the RPV will redood as modelled in the PRA '

(by grasity draining and by manual actuation of the casity Gooding system). Tne ITAAC should

- include confirmation of internal volumes, levations, and inter compartment vent and drain paths of the subcompartments containing RCS piping components and impacting reactor cavity Goodi.ig and RPV re0ooding. WEC needs to provide this ITAAC.

- Westinghouse Response:

The key elements to succ:ssful ex sessel cooling are:

- Water Goodmg the lower portion of the containment containing the reactor vessel IRWST drain arrangement (ITAAC Table 2.2.3-4, item I and 9.i)

A Dow path to allow water to remove heat from the outside of the reactor vessel RV insulation now area (new ITAAC Table 2.2.3 4, item 9.ii and iii)

Flow path from loop compartment back to under reactor vessel (new ITAAC, Table 22.3 4, item 9,iv)

The containment water lesel required to support ex vessel cooling is less than the provided by the

- IRWST to support post LOCA long term cooling. As a result, there is no need to :MAC the containment solume below the level required to support ex vessel cooling The PXS has an ITAAC (Table 2.2.3 4 item 9) that contains some of these requirements. The associated design description and the items contained in Table 2.2.3 4 are expanded to include the new ITAAC requirements listed above in 2.1 and 2.2.

SSAR Revision:

See the response to RAI 720.423 F.

Certified Design Material Revisionst Section 2.3.4 is revised as attached.

4 720.442F-1 g..

4

.m m

l Certified Design Materid PASSIVE CORE COOLING SYSTEM

{

y Draft Revision: 4 Effeev ',: 12/12/97 b) The Class IE components identified in Table 2 2.31 are powered from their respective Class IE division.

c) Separation is provided between PXS Class IE divisions, and between Class IE disisions and non Class IE cable.

ft The PXS prosides the following safety telated functions:

a) The PXS provides containment isolation of the PXS lines penetrating the containment b) The PXS prosides core deca) heat removal dunng design basis events.

c) The PXS provides reactor coolant system (RCS) makeup, boration, and safety injection during design basis events.

d) The PXS provides pli adjustment of water flooding the containment following design basis accidents l

9-The PXS prosides a function to ecol the outside of the reactor sessel during a severe accident,

drain-thean-containmentaefueling-wateMiof age-tanleilPrWMWnto-the-cont aimr ent.

10 Safety-related displays of the parameters identified in Table 2.2.31 can be retrieved in the main control room (NICR).

I 1.

a) Controls exist in the htCR to cause the remotely operated valves identified in Table 2 2.31 to perform their active function (s)-

b) The sahes identified in Table 2.2.31 as having protection and safety monitoring e stem (PMS) control perform their active function after receiving a signal from the PhtS.

c) The salves identi6cd in Table 2.2.31 as hasing diverse actuation system (DAS) control perform their active function after receiving a signal from the DAS.

12.

a) The motor operated and check salves identified ic Table 2.2.31 perform an actise safety-related function to change position as indicated in the table.

b) After loss of motise power, the remotely operated sahes identified in Table 2.2.31 assume the indicated loss of motive power position.

13.

Displays of the parameters identi0ed in Table 2.2.3 3 can be retrieved in the htCR.

Inspectinn, Tests. Anal ses, and Acceptance Criteria 3

Table 2.2 3 4 specifies the inspections, tests, analyses, and associated acceptance enteria for the PXS.

0 MTMCSvev340.0203 wpf 1b

Certified Design Material 1

i PASSIVE CORE COOLING SYSTEM lLi.

Draft Revision: 4 7

Effective: 12/12/97 Table 2.2.3 4 (cont.)

Inspectioni, Tests, Anal)ses,,and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 9 The PXS provides a function i) A Dow test and analysis for i) The calculated now resistance I

to cool the outside of the tractor each IRWST drain line to the for each IRWST drain line I

sessel during a sesere accident contamment will be conducted between the IRWST and the

, hem 4;m-4 R WAT-mto-4w The test is initiated by opening contamment is s 1.38 x 10" 2

• • ,.tumwe t the isolation vahes m each line ft/gpm.

I ii) Inspections will be conducted ii) The combmed total now area of I

of the reactor vessel steam the steam outlet (s) are not less i

l outlets and water inlets than 7 5 fl. The combined total

}

Cow area of the water inlets is not 8

l less than 6 ft I

I l

ni) Insisctions of the as built iii) A report exists and concludes

}

reactor sessel msulation will be that the minimum now area l

performed between the sessel insulatmn and I

reactor sessel for the now path l

that sents steam is not less than a

3 j

7 5 ft considermg the maumum I

denection of the sessel msulation l

with a static pressure of 12 95 ft of l

w ater.

I iv) Inspections will be iv) A Dow path with a now area 2

)

conducted of the now path (s) iot less than 6 ft exists from the l

from the loop compartments to loop compartment to the reactor i

the reactor sessel castry.

vessel cavity 10 Safety-related display s of the inspection will be performed for Safety-related display s idemified m parameters identified in the retriesability of the safety.

Table 2 2 3-1 can be retnered m

'able 2 2 3-1 can be retriesed in related displays m the MCR the MCR the MCR 11u) Controls cust m the MCR i) Testing will be performed on i) Controls in the MCR operate to to cause the remotely operated the squib valves identified m cause a signal at the squib sake sahes identified in Table 2 2 31 Table 2 2 3-1 using controls in electrical leads that is capable of to perform their actne the MCR, without strokmg the actuatmg the squib sabe function (sl valve.

ii) Stroke testing will be ii) Controls in the MCR operate performed on remotely operated to cause remotely operated valves vahes other than squit, valves other than squib valves to perform identified in Table 2 2.31 usmg their actne functions.

the controls in the MCR.

2.2.3 17 g W85tingh00SO c arTAAcsvev34020:03 wpr 1b-1:1:97

innallirr NRC FSER OPEN ITEM

~

m Qv-stion 720.443F (OITS. 61st)'

The AP600 design includes a reflectise reactor vessel insulation system that provides an engineered Gow path to allow the ingression of water and venting of steam for externally cooling the vessel in the esent of a sescre accident involving core relocation to the lower plenum. Key attnbutes of the insulation systen' are:

RPV/ insulation panel clearances, water entrance and steam exit flow areas, and loss

- coefacients based on scale tests in the ULPU facility, ball and cage checs valves and steam vent dampers at the entrance and exit of the insulation boundary that open due to buoyant forces during cavity Good up, and insulation panels and support members designed to withstand the pressure differential loading due to the IVR boiling phenomena.

- No coatings are applied to the outside surface of the reactor vessel which will inhibit the wettability of the surface. The reactor vessel insulation system should be included as a risk significant SSC in the reliability assurance program, and reliability / availability controls and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained. IT ~ AC and availability controls are also needed to assure thet the RPV insulation system will perform as designed. WEC needs to provide these commitments and ITAAC.

Westinghouse Response:

Per the response to RAI 720 42.3F, the reactor sessel insulation system is included as a risk signiGeant SSC in the reliaHlity assurance program, and appropriate features of the reactor vessel insulation hase been included in the ITA ACs.

SSAR Resision:

See the responses to RAI 720.423F and 720.442F for SSAR and CDM P.evisions.

720.443F-1

NRC FSER OPEN ITEM Question 720.4541' (OITS 6250)

WEC did not submit an esaluation of the potential for debns blockage of the ERVC flow path, or specify a functional requnement to assure that the cages and other parts of the coolant Oow path are not schject to debns blockage Based on information prosided by WEC on related topics, the stali conuders blockcge of the ERVC Dow path to be a potential concern Speci0cally,in response to RAI 480 1079 WEC indicated that the RCS blowdown during 's I.OCA will tend to cany debns created by the accident into the reactor casity, and that during the Dood up time, matenals will settle to the Doors of the loop compartments or the reactor easity. In response to RAI 480.148, WEC noted that the entrance pathway into the CV n.Jation is elesated and the water level at the time of debris irlocation is sescrr.1 meters abose the bottom of the insulation so donting or submerged debris cannot be engested into the insulation t1oupath. Iloweser, the high now rates tl, ugh the sy stem (possibly as high as 10.000 gpm according to estimates presided by Westinghouse during an August 17,19s7, mecting), in conjunction with tba relatisely small RPV insulation entrance and its prosimity to the re;,ctor casily Hoor, could result in re surpension and transport of debris to the entrance of the insulanon sy stem, Also,in response to RAI 480148, WEC indicated that there are no screens to plug and the w ater flow paths in the insulation are large This response is not consistent with the conceptual insulata.r dcign information prosided in Chapter 39 of the pRA and in Appendis K of th-ROAAM report, which indicates that the insulation sy stem includes an unspec Ded number of buo> ant stainic.s sicci balls caelosed within perforated screen like cages Details regarding the now area and,

mesh sire o 'the cages base not been prosided Thus, the staliis unable to conclude that the ERVC Onw path is not subject to plurging. The statT concludes that the potential for debns blockage of the I;RVC Hew path needs to be csaluated by WEC, and that a functional requirement needs to be prouded to assure that the cages and ether parts of the coolant flow path are not subject to debns Idockage Westinghnuse Regunse:

The responu to RAI 48014H discussed the timing sequence of sessel Gooding with respect to es-sessel cooling. As discumsed, the reactor casily is Gooded befoie the core debns relocates to the buct head The entrancewey pathway into the insulation is clevated and the water lesel at the time ot'lVR is ses trai mcters abose the bottom of the insulation so Ocating or submerged debns cannot be injested into the insulation Ocupath The Ocw path from the loup compartments are large, and the velocities in the Dow path (s) are scry low dunng e nessel cooling. l'or example, based on the estimate of 10,000 gpm through the insulat;,,n, the corresponding masimum approach selocity through the entranceway between the reactor coolant drain tank room and the reactor ca;ty is lesa than I ft/sec. Other now paths along the encuit base larger now areas and tonesponding lower selocities Such low approach selocities pres ent entrainment of large debris into the solume below the reactor sessel in addition, the total now area of the water inlet assemblies hase sumcient margin to preclude sigm0 cant pressure drop dunng esoessel cooling during a sescre accident Due to the low approach 2tm 3 Westinghouse I

=

NRC FSER OPEN ITEM selbeities to the reactor casity, and the relatisely large minimum flow area through each water inlet assembly (3 7 in'), the water inlet assemblies are not susceptible to clogging from debris inside containment SSAR Resision:

p See the response to RAI 720 423F for SSAR resisions associated with in seuel retention.

e 720.454F 2 W Westinghouse E f

_.. g NRC FSER OPEN ITEM Question 720.455F (OITS 6251)

WEC indicated that a structural anal) sis uas perfonned for the conceptual design of the RPV insulation s> stem and that the results of the esaluation show that the design was able to meet each of the defined functional requirements. Thus, a design that meets the functional requirements is feasible.

Iloweser, the RPV insulation design description or a complete set of functional requirements are not currently included in the SSAR or ITAAC. In siew of the reliance on ERVC to meet the Commission's large release frequency and containment perfonnance goals, the design description and all funcimnal requirements for the RPV insulation should be added to the SSAR, and important critena associated with the insulation design should be incorporated into the ITAAC, including information related to the necessary clearances / flow areas, and the check sahes and steam sent dam pers Alternatisely WEC can adopt a Design Acceptance Criteria (DAC) approach, in which case, the complete set of design criteria or functional design requirements, including pressure loading and deflection enteria, needs to be incorporated in the SSAR and in system ITAAC. In either case, the sy stem should 9e included as a risk significant SSC in the reliability assurance program, and reliability /availabilit) controls and goals should be presided to assure that operability of the system and mosing parts is maintained Writinghouse I(ciponse:

See the response to RAI 720 423 and associated SSAR updates.

SSAlt 1(esision:

None 2MSSN W Westinghouse