Text

Forwards Correspondence Previously Sent Informally Over Period of 970605-970724
	ML20216H677
Person / Time
Site:	05200003
Issue date:	09/10/1997
From:	Mcintyre B; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NSD-NRC-97-5308, NUDOCS 9709170017
	Download: ML20216H677 (211)
	v • d • e

_ -_ - ___ -

s w

'y

t

.t

-Westinghouse Energy Systems b 355 Electric Corporation Pittsburgh Pennsylvarus 15230-0355 DCP/NRC1023 NSD-NRC-97 5308 Docket No.: 52 003 September 10,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY

SUBJECT:

INFORMAL CORRESPONDENCE Dear Mr. Quay; a

Please find enclosed a formal transmittal of correspondence we have previously sent to you informally.

This informal corresi'ondence was sent over the period June 5,1997 through July 24,1997.

Attachment I provides the index of the attached material as you have requested.

Ilrian A. McIntyre, M iager Advanced Plant Safety and Licensing jml Attachment Enclosure cc: N. J. Liparuto, Westinghouse (w/o Attachment, Enclosure)

J. Roe, NRC (w/o Enclosure) i

l

..s }

m 9709170017 970910 PDR ADOCK 05200003-A PM _

usu .er (, n 3

" ' "(. 3 talspBE. LWR

u h

Attachment I to Westinghouse Letter DCP/NRC1023 DATE ADDRESSEE DESCRIPTION 6/5/97 Jackson Status of open items.

6/6/97 Jackson Site related open items.

6/5/97 Jackson SSAR markup of SSAR section 3.2.2.7 It will go into revision 14 of the SSAR unless we hear from you.

6/5/97 Iluffman Draft ESFAS tech specs.

4 6/10/97 Quay Request for confirmation of receipt of reports.

6/10/97 Iluffman Summary of 6/9 meetings with lilCB on tech specs.

l 6/11/97 Sebrosky Level 2 PRA insights.

6/12/97 Iluffman OITS update based on 6/9 tech spec meeting.

6/12/97 Jackson Status of open items.

6/11/97 Sebrosky' Letter DCP/NRC0907 - Responses to questions on the initial test program.

6/13/97 Jackson Markup of SSAR table 5.2-3, 6/16/97 Jackson Markup of SSAR section 3.7.3.5.2.

6/17/97 Jackson Markup of SSAR section 3B.7.

6/18/97 Jackson Markup of SSAR section 5.4.12.

6/17/97 iluffman Advanced response to Chapter 15 discussion item #39.

6/18/97 Huffman Response to items from 3/12/97 meeting.

6/20/97 Jackson Chapter 9 open items.

6/19/97 Huffman OITS update.

6/25/97 iluffman Responses to remaining MAAP4/NOTRUMP benchmarking RAls.

6/26/97 Quay Open item status.

6/30/97 Jackson Structural open items update.

6/30/97 Iluffman OITS update.

6/30/97 Quay / Slosson Wednesday reminder of items needing NRC acknowledgement.

7/2/97 Jackson /Kenyon OITS update.

7/10/97 Jackson Information to resolve open item 18')3. Will be in revision 15 of the SSAR.

3%7a wg

u 4

7/15/97 Sebrosky Telecon items related to MAAP4 and MELCOR.

7/16/97 Kenyon Information to resolve open item 293. Will go into SSAR revision 15 unless we hear from you.

7/14/97 Quay Question accuracy of date referenced in NRC letter of 6/9/97.

7/17/97 Sebrosky OITS update.

7/21/97 Kenyon Correction of SSAR section 9.5 sent 7/18/97. Will be in 4 SSAR revision 15.

7/22/97 Sebrosky Markup of SSAR section 13.3.1 and response to open item 720.391.

7/2/97 liuffman Advance responses to non-LOCA chapter 15 discussion items.

7/21/97 Chu insert for tech spec bases 3.8.1.

i 7/23/97 Kenyon Markup of SSAR section 9.3.1.2.2 to close open item 3968.

l 1

Will be in next revision to the SSAR unless we hear from you.

7/24/97 Quay / Slosson Chapter by chapter breakdown of open items. Request assistance to close gap between what Westinghouse believes is still open and what the NRC thinks is still open.

{

}1674 mpf

. j

C AP600 Open Item Tracking System Project Status Report .

Selection: Full Selection Status as of: l 6/5/97 l l

Resolution Status (W / NRC)

E ' Dropped Confrm-W Confrm-N Audit N Action W Action N Resolved M Total 0 15 / 17 9 / 1 / 0 71 / 157 43 / 317 5 / 492 1106 / 268 1252 / 1252 DSI:R - O! 0 / 1 I5ER - Confmnatory 0 / 0 0 / 0 0 / 0 0 / 0 2 / 9 1 / 7 p / 30 77 / 34 80 / 80 DSTR - COL 0 / 0 0 / 0 0- / 0 0 / 0 2 / 6 1 / 37 ? / 62 162 / 60 165 / 165 DSI'R - 0I50 62 / 62 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 62 / 62

_. . , _ . . _. . o .. . _ , . _ - . . . .

RAI - 01 0 / 1 5 / 0 4 / 2 0 / 0 753 / 865 #8 / 677 0 / 221 1293 / 337 2103 / 2103 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 0 / 0 RAI -Confumatory 0 / 0 0 / 0 Meeting - Of 1 / 1 3 / 6 2 / 1 0 / 0 255 / 308 39 / 215 3 / 1 71 623./ 224 926 / 926 leconfcrence- O! 0 / 0 0 / 0 4 /. 0 0 / 0 5 / 21 6 / 11 0 / 15 41 / 9 56 / 56 Key issue 0 / 0 0 / 0 0 / 0 0 / 0 34 / 43 5 / 1 0 / 1 6 / 0 45 / 45 .

Total: 63 / 64 23 / 23 19 / 4 1 / 0 1122 / 1409 143 / 126% 8 / 992 3310 / 932 4689 f 4689

_ _ 4

Lindgren, Donald A. -

From:' Lindgren, Donald A.

-font:- Friday, June 06,199711:00 AM To: ' Jackson, Diane

Cc: McIntyre, Brian A; Winters, James W.; Orr, Richard S.; Prasad, Narendra

Subject:

Site Related Open items Our recent letters to address RAls 231.35 through 231.42 provided information that address other items. These items are listed below. The Westinghouse status on these will be Action N. I have included the item numbers and

. the most recent NRC question or request. I have also provided more information about DSER open item 2.5.4.41.

OITS# 549, DSER 2.5.4.4-1 The April 25,1997 NRC letter requested information on the effects of using dry soil densities for saturated soils.

Information about the soil densities used for soil structure interaction analyses and the range and effects of variation of soil densities is included in SSAR subsectior,2B.3.6.

OITS #628, DSER 3.7.1.1-1 Respond to RAI 231.40,231.41,231.42 Responses to these RAIS were provided in letter DCP/NRC0858 dated June 5,1997 OITS #768, DSER 3.8.5 10 - Westinghouse should include the loads due to construction sequence in the basemat I- design.

The response te RAI 231.38 in letter DCP/NRC08 dated June 3,1997 addressed the issue of construction loads and provided a markup of SSAR subsection 3.8.5.4.3.

OITS #769 DSER 3.8.5 The SSAR 2.5.4 draft is incomplete: more information is needM on the geotechnical program.

f Letter DCP/NRC0895 dated 6/3/97 provided responses to RAI 231.35,231.36,231.37, and 231.39 that -

addressed NRC questions about the geotechnical program. A markup of SSAR section 2.5 that provided additional information on the requirements for the geotechnical program was also provided.

. OITS #4997, RAI 231.34 - New RAls in NRC letter dated 5/2/97.

Responses to RAIS 231.35 231.42 were provided in letters DCP/NRC0895 dated 6/3/97 and DCP/NRC0858 -

dated June 5,1997 -

I hope this will permit you to clean these up. Please contact me if you need additional information on any of these.

Don Lindgren (412) 374 4b56 m

Page 1

(

- - _ -[W) Westinghouse------- - - - TAX COVER SHEET -

g RECIPIENT INFORMATION SENDER INFORMATION OATE: Juu cr f,/997 NAfIE: % py,yg TO: ,

- LOCATION:

/ eM6 J Atterco ENERGY CENTER -

EAST PHONE: FACSIMILE: PHONE:

OmC8:V/2- 5 7V-s 2 9o COMPANY: Facsimito; win: 204 4M7 U 7 ^/40-outside: (412)374 4087 LOCATION:

_ -- L Covor + Pages 1+/

The follow!ng pages are being sent from the Westinghouse Energy Center. East Tower, Monroeville. PA. If any problems occur during this transmission, please cati:

WIN: 284 5125 (Janice) or Outside: (412)S74 5125.

COMMENTS: ~ ~ " '

l DMe-1 hl5 Av%Id a d S Hou.LD )? E sm e v e~ (HA at. ') con c ew cy Au, c p3 .-

e h f [

vr a=v w e uru. m ,/$a .

(e, bwCGlN pf

(.

W r%Tf/c~ .

L u tN W*)

f& r%4 7 &fM N

0 e

om.

l

3. Desisiof Structures, Compone:ts, Eq:lpment end Systems

Prevent interaction that could result in preventing Class A, B or C structures, systems, and components from performing required safety related functions 3.2.2.7 Other Equipment Classes 4

Equipment classes E, F, L P, R, and W are nonsafety related. They apply to structures, systems, and :omponents not covered in the above classes. They have no safety related function to perform. They do not contain sufficient radioactive m-terial that a release could exceed applicable limits, l Structures, systems, and components that do not normally contain radioactive fluids, gases, j or solids but have the potential to become radioactively contaminated are classified as one of j -

these nonsafety.related classes if *he following criteria are satis 0ed; Gil The system is only potentially radioactive and does not normally contain radioactive materi-

!- y The system has shown in plant operations that the operation with the system containing radioactive wL material meets or can meet unrestricted area release limitQ

+

An evaluation of the system confirms that the system contains features and components

- that keep the consequences of a system failure as low as reasonably achievabg ud

+

I The system has no other regulatory guidance requiring its inclusion in Classes A, .

I B, C or D.

This review of the system features and components ;ncludes the following as a minimum:

Features and components that control and limit the radioactive contamination in the system >

Features that facilitate an expeditious cleanup should the system become contaminated a

Featiires and components that limit and control the radiological consequences of a potential system failure -

+

1he means by which the system prevents propagation to an event greater consequence.

There are no special quality assurance requirements for Class E, F, L. P, R, and W structures, systems, and components. Unless specifically specined,10 CFR Part 21 and Part 50, Appendix B do not apply. The systems and components are normally not designed for seismic loading. However, there may be special cases where some seismic design is required.

See subsection 3.2 I for more details.

Revision: 13 3 Westingh00$8 3.2-11 May 30,1997

- TX CONFIRMATION REPORT ** AS OF IUN 6 '97 9:30 PAGE.01 AP600 DESIGH CERT DATE TIME T0/FROM MODE MIN /SEC PGS 3TATUS 01 6/ 6 09:36 223:NRC G3--S 01'02 02 OK 9

O e

_1_--

49 -oD oJ? kN ggg MO 13oU#$52(

g- M" c

L6 1 ' F *j

/ pp MN 1:s 4**

yw nre t t z u sa/.-- )'9 ( '

~

d E'E-hb 9!l- NP 4

)CP ppQ b P F" S(

e KO TLUSER 63 .SGP 50'21 T EGASU 23:61 50/80 FMIT .TS

DI NOITCENNOC '

SSERDDABUS 200251410318 LET NOITCENNOC 7714 ON XR/IT KO ADISSIMSNART sesessetssessassesasa ses TROPER IT ees ssssssssssssssssssses T73 214 IAF 44 61 llil iH'EW/WW

' ~'

Ti. -

Oc Orc -I(ru Esi:As gcs _e gg. 4

9. '

A b~

y a,OFman'/

1 7/ (\os+ p y b d bd 14 is g .

, Le 3,3.2 0 52 F- (g) ),

.. P 3.3-44 I

9108 0 DN 17dS3h 0 'SDd 00.00 1 3DYS3 21:91 S0/90 3 nil '1S 01 NOI133NNO3 SS3HOQVB3S l 200Z919IOCIS 131 NOID3NNO3 ELit ON YH/Il 0313~1dR0310N SVu NOIDNad Y1

s s*: 1H0d3H II H0HH3 s*:

sessasssssssssssssssssssnes

. _.q1.E S klf 7ik YV4 PT,QT lit i 7 R / c n z3%_____

L lp' ,arD oJD kN w

) p aaL

} \09 kE is

,m t r .z u>#f-)*(('

d t'{-hh l

210# 51 GN TLUSSK '

51 .SGP 13'50 T EGASU 41:51 50/60 EMIT .TS DI NOITCENNOC SSERDDABUS 200251410318 LET NOITCENNOC 1714 ON XR/XT DETELPMOC TON SA# NOITCNUF XT ses TROPER XT RORRE ::

sessssssssssssssssssssssssa m ar ,4.ai tna es au uu

E5FAS Instrumentation

,) e 8 3.3.2 3.3 INSTRUMENTATION 3.3.2EngineeredSafetyFeatureActuationSystem(ESFAS) Instrumentation LCO 3.3.2 The shall beESFAS instrumentation for each function in Table 3.3.2 1 OPERABLE.

APPLIC'<BILITY: According to Table 3.3.2 1.

ACTIONS

....................................N0TE......................................

Separate condition entry is allowed for each Function.

CONDil10N REQUIRED ACTION COMPLETION TlHE A. One or more Functions A.1 with one or more Enter the Condition immediately required channels or referenced in divisfohs inoperable, Table 3.3.2-1 for the channel (s)or ,

division (s).

B. 458 s75 One manual initiation B.1 Restore manual 8 device inoperable, J(hours initiation device to OPERABLE status.

t e.: '!;rify :;n;;l 72 5::r:

'-iti:ti:n d:;i::;,

iielent te th.

ntrei ree; devises, in th; n;;ete Shutd;;; ";rk;t

tien er. OPERA"L: end &

d;di :t:d ;p;r;ter .

ahe is in centinu;ue

i :ti:r with th; ;;ntrei r;;; i;-

teti;n;d et the -

d:vic L QB (continued)

(!)

' m, AP600 3.3-17 i maim 08/96 Amendment 0

.J

t$tAS instrumentation

.. .. 3.3.2 ACTIONS (continued)

CONDITION ' REQUIRED ACTION COMPLETION TIME 2,

.st' s B. (continued) B./.1 Be in MODE 3. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 8 MA 2-72 '

B /.2 Be in MODE 4,W W M jlthours emA; Mauso at W Rus.

4 5 73 C. All Engineered Safety C.1 Restore 1 Actuation ifhours '

Features Actuation Subsystem in the Cabinets (ESFACs) inoperable division battery backed logic to OPERABLE status, groups in one l

division iaoperable. OR 12-l C.2.1 Be in MODE 3. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months l

AND S +2 C.2.2 BeinMODEf. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 575 D. One required D.1 Verify the interlocks interlock inoperable. I hour L are in the required state for existin plant conditions.g M

7 D.2/ Place any functions IVhours associated with ino)erable interlocks in )ypass.

-0.2.2 P.0:ter:-interl::h t: 15? 5:r;

{PEPfSLE :t:te:. .

B (continued)

\

. AP600 3.3-18 mi- .no=== i m 08/96 Amendment 0

ESFAS Instrumentation

.. 3.3.2

. ACTIONS (continued)

CONDIT10N' REQUIRED ACTION COMPLETION TIME D. (continued) 0.3.1 Be in MODE 3. IX hours '

6Np 13 0.3.2 Be in MODE 4. PM hours s1s E. One channel E.1 Place channel in #ghours D inoperable, trip.

F. One required channel g .57s F.1 Place channel in f inoperable. bypass.

J hours f.2.1 " g;' vip d er.rel tv 10o ;im >

v muu ...s ..

B 12.

F.2./.1 Be in MODE 3. Ptihours Aja

/4 F.2./.2 Be in MODE 4. Prfhours G. All battery backed G.1 6 575 Restore 1 Functional "Pt hours C logic groups in one Logic Subsystem in Protection Lo the inoperable Cabinet (PLC)gic '

inoperable cabinet to OPERABLE status.

E '

IL G.2.1 Be in MODE 3. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

AND 5 +1 G.2.2 BeinMODE/.

9tIhours (continued)

, HAP 600 3.3-19 mi-i = muse 08/96 Amendment 0

E5fAS Instrumentation

. ,. 3.3.2 AC110NS CON 0lT10N

REQUIRED ACTION COMPLET10N TlHE ,

H.

ct2fS One channel H.1 Restore channel to inoperable. OPERABLE status, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months "[f g

g , o,rs) 0B H.2 Suspend movement of irradiated fuel 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months assemblies, One required channel g 57S

1. 1.1g Place channel in 7' hours #

inoper:ble, bypass.

.:.: neatere d.ii..wi to 1:0 :,...

-0PEPfalE ;tetu;.

- 93 12.

l.2.[ Be in MODE 3. Ptf hours ;

J.

sn One manual initiation J.1 Restore manual O O

( channel inoperable, initiation channel to 7tlhours OPERABLE status.

98 J.2 Initiate action to be M in MODE 5 with RCS Thours open and visible level in pressurizer.

(continued)

AP600 APg1 w tMBSB02J06401W 3.3-20 08/96 Amendment 0 ,

.. . . - 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME 375 K. All ESFAC battery K.1 Restore 1 Actuation ~ ##

backed logic groups 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Subsystem in the in one division inoperable division inoperable, to OPERABLE status.

_OB K.2.1 If in MODE 5 with the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Reactor Coolant System (RCS) open and level not visible in pressurizer, initiate action to be in MODE 5 with RCS open and visible level in pressurizer.

l M i K.2.2 If in MODE 5, isolate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

' the flow path from the demineralized water storage tank to the RCS by use of at least one closed and de activated automatic valve or closed manual valve.

AND K.2.3 If in MODE 6 with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months upoer internals in place and cavity level less than full, initiate action to be in MODE 6 with the upper internals removed and the -

cavity full.

M K.2.4 Suspend positive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months reactivity additions.

(continued)

, h AP600 3.3-21 a.mo.m soimosu e 08/96 Amendment 0 a

I (SFAS Instrumentation

.. .. 3.3.2 ACTIONS-CONC,lTION '

REQUIRED ACTION COMPLETION _ TIME L. All battery backed S77 L.) Restore 1 Functional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ~##

logic groups in one Logic Subsystem in PLC inoperable. the inoperable cabinet to OPERABLE 4

status.

01 L.2.1 If in MODE 5 with Rr5 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months open and level not visible in pressurizer, initiate action to be in MODE 5 with RCS open and visible level in pressurizer.

AND l.2.2 If in MODE 5, isolate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months -

the flow path from th$ demineralized water storage tank to the RCS by use of at least one closed and deactivated automatic valve or closed manual valve.

AND L.2.3 If in MODE 6 with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months upper internals in place and cavity level less than full, initiate actior to be in MODE 6 with the upper internals removed and the '

cavity full.

AN,Q (continued)

, h AP600 mi-ieu nm euse 3.3-22 08/96 Amendment 0

($iA$ lastrumentation 3.3.2 4

AC110NS CCN0lil0N -

DEQUIRED AC110N COMPLET10H TlHE L. (contlitued) L.2.4 Suspend positive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months reactivity additions, M. One required channel g, 675 inoperable. H.l.% Place channel in yhours N 64-bypass.

M

" 2.1 ne; tere ;h;nn;l t; 100 h;;. ;-

OPCPfa C ;t:tue.

~

. . . u nE - - - ~ '

10 oa peg (s)me's f

F p j,94.Ars4147Ef*F ?wy H.2 1 01a the flow path lV

, IMrhours u40Ftt 40*W 6T N O o wT4dAS. deminerilized Water

- " ~ " ~ ~

/

.) storage tank to the

'* RCS by use of at least one closed and de activated automatic valve or closed manual valve.

(continued)

, h AP600 i - ieu u m min. 3.3 23 08/ 96 Amendment 0

[5fAS lastrumentation

.. 3.3.2 AC110NS (continued)

CONDITION REQUIRED ACTION COMPLE110N TlHE N. One required channel N.1 inoperable, Place channel in - E-bypass.

A3 N.2.1 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months Restore channel to OPERABLE status.

_0B N.2.2.1 If in MODE 5 with RCS open and level not i

visible in pressurizer, initiate action to be in MODE 5 with RCS open and visible level in pressurizer.

.. E 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months N.2.2.2 If in MODE 6 with upper internals in place and cavity level less than full, initiate action to be in MODE 6 with the upper internals removed and the cavity full.

E 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months N.2.2.3 Suspend positive reactivity additions.

(continued) t u h AP600 3.3-24 mi-i momameian 08/96 Amendment 0

thiAS instrumentation

,. . 3.3.2

.. ACTIONS (continued)

CONDIT10N REQUIRED ACT10N COMPLET10N TIME 575

0. One manual initiation 0.1 Restore' manual * * **T device inoperable, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months initiation device to OPERABLE status.

9.8 0.2.1 If in H0DE 5 with RCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months open and level not visible in pressurizer, initiate af. tion to be in MODE 5 with RCS open and visible level in pressurizer, bhQ 0.2.2 If in H0DE 6 with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months upper internals in place and cavity 1evel less than full, initiate action to be in MODE 6 with the

, upper internals i

removed and the

cavity full.

AND 0.2.3 Suspend positive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months reactivity additions.

1 (continued) s h AP600 3.3 25 08/96 Amendment 0 m i - o.n e w ines

L5fA5 Instrumentation

.. .. 3.3.2 ACTIONS (continued)

CONDITION '

REQUIRED ACTION COMPLE110N TlHE l P. One manual initiation P.1 Restore manual device inoperable. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i initiation device to ,

i OPERABLE status. '

i 98 P.2.) If in H0DE 5 initiate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action to be in MODE 5 with the RCS intact and visible level in ,

pressurizer.

2 S

P.2.2 If in MODE 6 with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

~

upperinternalsin i piace and cavity level less than full, '

i initiate action to be 1 - in MODE 6 with the

' upper internals removed and the cavity full. '

2 E ,

P.2.3 Suspend positive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months reactivity additions.

(continued) i 1

1 i

5 (l)AP600 3.3-26 m i - i u m m isse 08/96 Amendment 0

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

I,5fA5 instrumentat1on

.. .* 3.3.2 ACTIONS (continued)

CONDITION '

REQUIRED ACT,10N COMPLEil0N 11HE Q. One required channel (. STS Q.1 Place channel in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months -#WO inoperable, trip.

8EQ Q.2.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

1 0

Q.2.2.1 If in H0DE 5 initiate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action to be in H0DE 5 with the RCS intact and visible level in ~

pressurizer.

MO 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Q.2.2.2 If in H0DE 6 with upper internals in place and cavity :

level less than full, initiate action to be in MODE 6 with the upper internals removed and the cavity full, bbQ

'72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Q.2.2.3 Suspend positive reactivity additions.

(continued)

~

h AP600 3 3 27 5 mi - i w 08/96 Amendment 0

! $h5$b b' tarA5 instruu ntation

. 3.3.2

\

'fl0NS (continued) i CONDITION REQUIRED ACTION COMPLET10N TlHE 4 :s n pe channel in / hours ~NML e

ll$

+2

-0.2 ! Dest +re channel to ' ~ >

'""U"_

-OPC PAOL 3tetg3z M

R.2./ Initiate action to be 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months in H00E S with RCS open and visible level in pressurizer.

S. One required channel S.I.t Place channel in inoperable, bypass.

hours 575

-NAVE

~8fff!-

021 -Ice h: rs E!!ig,.;b.M..ne w nua ....

t:

M 12.

S.211 Be in H00E b LNIhours AND 56 S.2./.2 Be in H00E 4 with the ly[ hours RCS cooling provided by the RNS. .

(continued) h AP600 3.3-28

mi--sm= = 08/96 Amendment 0

($fAS instrumentstion

.. .. 3.3.2

,, AC110NS (continued)

CONDITION REQU! RED ACTION COMPLETION 11HE cffr5 T. One channel T.1 8'S M*#

inoperable. Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months t OPERABLE status. A#C OnW l 98 l T.2.1 Verify atlernate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l radiation monitors ;

are OPEPABLE.

&NQ T.2.2 Verify control room 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months isolation and air supply initiation manual controls are operable.

OB 78 T.3.1 Be in H00E 3. JKIhours 8!!Q 5 44 T.3.2 BeinH0DE/. 9(hours V. One ;;ne:1 Iriiti ti 0

>d; ic; in;p;e.h.

U.! 9ett^re -M" t-ttt:tter de'ete t 1

II'^" a Or" RA 0i.: i;etu.-

1 _O3 ygsgem/ U.2 Yerify :re:1 72 $::r:

g g gy. g na,q u_ '

8-!!!:tter devices,

q;ini:rt i - th-

air;l r::: de;i :;,

,. .t. _a.

."...n I

..l."";.'.1

.. us. , i,.

v. ,

are OPEPf!'E ::d :

dedic ted :pereter '

d: i : ! :::t! ::::

s ee -i ir>+<aa muk t'a --tid 7::: i;

- tett- :d at t' .

-de"!:en (continued) 6 h AP600 3.3 29 08/96 Amendment 0 mi-amme, em n .m-- w - n p- .-,.-r-gm

~

A9600 LCO 3.3.2 INSERT Action U ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME U. One manual initiation U.1 Restore manual initiation 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months device Inoperable. device to OPERABLE status.

l U.2.1 Be in MDDE 34 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months AJ U.2.2 Be in MODE 4 with the RCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months cooling provided by the RNS.

J A

.. U.2,3 .

.... ... NOTE . - '

Flow ath(s) may be uniso ated intermittently under administrative controls.

Isolate the affected flow 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months path (s).

MLD U.2.4 Verify the affected flow Once per 7 days path is isolated.

U.3.1 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months AJ U.3.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months b

+4 s 8 e

A W s j

l A 1 To, n . w 3\n NY C h h n'. -

b c O f( S n) f5FA5 $cs (Lco,ne;-bd 4 0 .

h-(

fi f

,7 < a (sos + p y's '

A b4 L* W g 9 to 3.u 4 -

inyr" M )

p 3,3 H bih - 3rc! k -

(9nd cW4ag d4 -no ycgs)

[5fAS Instrumentation

,, .. 3.3.2 CONDITION .

REQUIRED ACTION COMPLETION TlHE

'U. (ceetia.ed) g

^

i

-U.0.1 0; in "00 : aa VW t_.._.

I4ywy y-

{

t ANO U.3.2 Oe in "00: ? with the ^0 h;;r; RCS : :M~g p :M W t,, the M;.

I" W NSul f9c.7/SN 5 : V LQ X

Y E

Ah 66 C.C

, OO 66 PF GG Wd II -

h AP600 3.3 30

' mi-immum.im 08/96 Amendment 0

AP600 LCO 3.3.2 INSERT Action V

-ACT10N3 CONDITION REQUIRED ACTION COMPLETION TIME V. One manual initiation V.1 Restore manual initiation 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months device inoperable, device to OPERABLE status.-

OR V.2.1 Be in MODE 3, 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months A$

V.2.2 Be in MODE 4 with the RCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months cooling provided by the RNS.

g h e

4 4

b 11

AP600 LCO 3.3.2 INSERT Action W ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l W. One manual initiation W.1 Restore manual initiation 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months device inoperable, device to OPERABLE

. status.

1 OB W.2.1 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months l W.2.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months 6 8 w

B5 h

AP600 LCO 3.3.2 INSERT Action X ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME X. One required interlock X.1 Verify the interlocks are 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. In the required state for existing plant-con 01tions.

93 X.2.1 Place any functions 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months associated with the inoperable interlock in bypass.

93 X.3.1 Be in MODE 3. 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months

.. gg X 3.2 Be in MODE 4 with the RCS 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months cooling provided by the RNS.

e 05 I b

4 .

AP600 LCO 3 3.2 INSERT Action V ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME Y. One required channel Y.1.1 Place required inoperable inoperable. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months channel in bypass.

l Y.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

inoperable channel 15 placed in bypass.

OR f

Y 2.1 Bt in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months J

A Y.2.2 Be in H0DE 4 with the RCS 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months a

cooling provided by the '

RNS, 4

4

~~_;

AP600 LCO 3.3.2 INSERT Action 2 '

ACTIONS CONDITION REQUIRED ACTION CCPPLETION TIME l Z. One required channel Z.1.1 Place required inoperable inoperable. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months channel in bypass.

ELO Z.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable channel is placed in bypass.

1 0

Z.2.1 Be in MODE 3, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND

~

I Z.2.2 NOTE -

Flow ath(s) may be uniso ated intermittently under administrative controls.

Isolate the affected flow 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months path (s).

AND Z.2.3 Verify the affected flow Once per 7 days path 15 isolated.

01 Z.3.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

,A,,N,0 2.3.2 Be in MODE 4 with the RCS 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

cooling proviced by the RNS.

m,

. #600 .

LCO 3.3.2 INSERT Act?on AA I ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME AA. One required channel AA.1.1 Place required inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable. channel in bypass.

AND

. AA.1.2 Ensure that unrecaired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable channel is placed in bypass.

93 AA.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ANQ AA.2.2 Be in MODE 4 IB hours 4

b d

AP600

LCO 3'3.2 INSERT Action BB ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME BB. One required channel BB.1.1 Place required inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable channel in bypass.

A,,Np B8.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable channel is placed in bypass.

1 I 0.R BB.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l A,NQ BB.2.2 Be in H30E 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months OR BB.3.3 -

NOTE -

Flow path (s) may be unisolated intermittently under administrative controls.

Isolate the affected flow 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months path (s) by use of at least one closed manual or closed and de-activated automatic

. valve.

6 b

AP600 1.C0 3 3.2 INSERT Action CC -

ACTIONS CONDITION REQUIRED ACil0N COMPLETION 1]ME CC. One required chanr.el CC.1.1 Place required inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperiJble. channel in bypass-CC.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Inoperable channel is placed in bypass.

93 CC.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months CC.2.2 Be in MODE 5. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months 4

w 4

9 t

',AP600 LCO 3.3.2 INSERT Action 00 l

ACTIONS ,_ _ _

CONDITION REQUlRED ACTION COMPLET10N TIME

00. One required channel 00 1.1 Place required inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

inoperable. channel in bypass. I A.30 1 00.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable channel is placed in bypass.

00.21 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months M

00.2.2 Be in H00E 5. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months

.. B 00.3.1 ...........N0TE..........

Flow path (s) may be unisolated intermittently under administrative controls.

Isolate the affected flow 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months pathis).

5 00.3.2 Verify the affected flow Once per 7 days path is isolated.

O 6

h

Af(Ci LCO J 3 ?

l PiSERT Action EE ACTIOAS CON 0! TION REQUIRED ACTION COMPLET10N TIME l

l Fr One reautred chennel EE.1.1 Place required inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i 1roperable. channel in bypass.

ED t

EE 1.? Ensure that unreautred 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable Channel is placed in bypass.

98 EE.2.1 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months MD, EE.2.2 Be in MODE 5. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months OR EE.3.3 ... . .....N0TE .. ......

Flow ath(s) may be uniso ated intermittently under administrative controls.

Isolate the affected flow 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months path (s) by use of at least one closed manual or closed and de.

activated automatic valve.

O 4

4 b

', AF600-LCO 3.3.2 lNSERT Action FF l

t ACTIONS CON 0! TION REQUIRED ACTION COMPLET10N TIME FF. One reautred channel Fl'.l.1 Place reautred inoperable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable. channel in bypass.

AND 3 , FF.1.2 Ens'Jre that unreQuired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i Inoperable channel is placed in bypass.

FF.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

Ag l

FF.2.2 Be in MODE 4 with the RCS 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months cooling provided by the RNS. -

AND FF.2.3 NOTE -

Flow ath(s) may be uniso ated intermittently under administrative

, controls.

' Isolate the affected flow 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months path (s).

A,,N,p FF.2.4 Verify the affected flow Once per 7 days path is isolated. ,

s 2

FF.3.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months -

FF.3.2 Be in MODE 5 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months

.;600

' ' LCO 3 3. 2 INSERT Action GG ACTIONS CONDIT10N REQUIRED ACT10N COMPLET10N T!ME GG. One required channel GG.1.1 Place required inoperable inoperable. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months channel in bypass.

. A3 GG.1.2 Ensure that unrequired 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable Channel is placed in bypass.

I

~

, GG.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A,Np GG.2.2 ...........N0TE.........

Flow path (s) may be

- unisolated intermittently under administrative controls.

Isolate the affected flow 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months path (s).

UE GG.2.3 verify the affected flow Once per 7 days path 15 1solated.

Og GG.3.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND GG.3.2 Be in MODE 4 with the RCS 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months cooling provided by the RNS. *

,. AP600 LCO 3.3.2 INSERT Action HH i

ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME HH. One required HH.1 Verify the interlock.s are 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months interlock inoperable. In the required state for existing conditions.

HH.2 Place any functions 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months associated with the inoperable interlocks in bypass.

6 HH,3 Be in MODE 3. 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months s

4 D3 O

b

86.00

LCO 3.3.2 INSERT Action !! -

ACTIONS CONDITION RE0V! RED ACTION COMPLETION TIME

11. One channel 11.1 Place inoperable channel 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable. in bypass.

l d 11.2.1 ...... NOTE -

Flow path (s) may be unisolated intermittently under administrat1ve controls.

Isolate the affected floc 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months path (s).

M 11.2.2.1 Isolate the affecter. 7 days flow path (s) by use of with at least one closed and deactivated automatic valve.

closed manual valve, blind flange, or check valve with flow through the valve secured.

SE

!!.2.2.2 Verify the affected Once per 7 days flow path is isolated.

e 0

150

. . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _.)

LSIAS instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS

....................................N0TE........,............................

Refer to Table 3.3.21 to determine which SRs apply for er:h Engineered Safety features (ESF) Function.

SURVElLLANCE FREQUENCY s75 SR 3.3.2.1 Perform CHANNEL CHECK. # 3'T7' l J( hours 3/ Mvs

SR 3.3.2.2 Perform ESF ACTUATION LOGIC TEST =2t ::r,ths M 33 2.2 (ESFALT). W .4.2.1 SL 3.3.2.3 . . . - . . . .

. . N0 T E . . . . . . . . . . - . . . .

Verification of Setpoint not required for manual initiation functions.

Perform TRIP ACTUATING DEVICE OPERATIONAL 24 months TEST (TAD 01).

SR 3.3.2.4 - -- -

- NO T E - - - - - . .

1.

This test shall include verification that the time constants are adjusted to the prescribed values. '

2. Resistance Temperature Detectors may be excluded from the CHANNEL Call 3 RAT!0N.

Perform CHANNEL CAllBRAT10N. 24 months SR 3.3.2.5 Perform ESF CHANNEL OPERATIONAL TEST i

24 months (ESFCOT).

SR 3.3.2.6 Verify ESFAS RESPONSE TIMES are within limit. 24 months on a STAGGERED <

TEST Basis h AP60 mi - .0 ..uw= 3.3 31 08/96 Amendment 0

LSFAS laltrumentation 3.3.2 Toble 3.3.21 (pape 1 of iSt Ingineoted Sofopwatae Attweben System instrumentation Rf 0VIR!D APPLICAtLt CHANNilli NOW4AL FUNCTION M00tl SVRVilLLANCI TRIP OtVitt0NS CON 0illC'NS R(QutREMENTS SCTP0 INT

1. Sciepverde Attwebon

e. eionwelIrvtieben 1.2.3.4 2 setthee A/ 4 SR 3 3.2.3 N/A l 2 setches P SR 3 3.2 3 N/A

b. Conteetnt Pteeswre - Rgh 1 1.2.3.4 3 CC/ SR 3 3 2.1 is 0 0 po'01 l SR 3 3.2.4

SR 3 3 2.5 SR 3 3.2 4

c. Peessuriter Pressure - Low 1. 2. 3I 83 3 f SR 31.2.1 la 1685 po'01 SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.4

d. Steam One Presswee - Low 1.2.38) I 3 per steem F

' SR 3 3.2.1 Ia eOS or line SR 3 3.2.4 l $3 tibi po'0 SR 3.3.2.5 See Note il l SR 3.3.2.4

e. RCS Cold leg Tempeteture 1. 2.3 I81 3 per loop F SR $ 3.2.1 (7,,y) - Love la $10 or SR 3.3.2.4 470'F SR 3.3.2.5 See Note 2' SR 3.3.2.5

2. Cote Mekeup Tonk (CMT) Actuebon

e. ManuelIrvtsetion 1.2.3.49 2 e etches Y/ SR 3.3.2.3 N/A 4W.E'l I 2 ewitchee J SR 3.3.2.3 N/A

b. Presounaer Water Level - Low 2 1.2.3.4@ 3 7/ SR 3.3.2.1 la 7.0%'l SR 3.3.2.4 la 1.0%)

SR 3.3.2.6 SR 3.3.2.4 4W . l I '# 3 R 54 3.3.2.1 lt 7.0%'l .

54 3.3.2.4 la 1.0%)

SR 3.3.2.5 SR 3.2.2.4

.. Seitwe,4e Actue noter to runction 1 (Sefe,werde Aetvenon) toe inrosen, tune. r,s and towiremente. -

(continued!

IU5MT A.

, h AP600 3.3 32 ApeIwaarusentsessetisto400fto 08/96 Amendment 0

_j

APG00 LCO 3.3.2 INSERT PG 3 3 -3E MP,](AB((

W C5 OR SF ! [D tt0V! RED Sutytluas;t talp TVNCTION CON 0!!!ONS CVNNtt$ CONDit!ONS RE M at tht$ sitp01stl8) 2 Cerg Maiesp feet (CMT)

Actuation d AD$ $ttgen 1, 2. & 8efer to fu nct106 9 (AOS $tegel 1. 2 & 3 Actuation' ) for all tettietton functions 3 Actuation and recatremeets.

t we h

4 4

6 b

05fAS Instrumentation

., .. 3.3.2 7668 3.3.21 (e ge 2 0 12 eng.a. ores sete,were. A twei.on sveiem instrumentebon Rf 0 VIRID HOMINAL APPLICA9Lt CHANNfL$f FUNCTION $VRV!!LLANCE T RIP M00t$ OlVISIONS CONDITIONS R t 0V'RE M(NTS St7PO:NT 3, Conte.ntnent isolation

e. Manuellrutistion 1,2,3,4 2 switches W/ $R 3 3 2.3 N/A 5,6 2 switches P SR 3.3.2.3 N'A

b. Manwel Irvt.etion of Passive

! Refst to Functies 12.s (Passive Conte nment Coohng Actuationi lot wtiebng Containment Coohng funebens and requirements.

e, Safegwards Attweben Refer to Function 1 (Safeguares Actwetioni for abobag fwncbons and re quiremente.

4. Steam Une lootebon

e. Menval trwbation 1.2,3,40) 2 switches V SR 3.3.2.3 N/A

b. Containment Pressure - High 1 1,2,3,40I 3 $ $R 3.3,2,1 lg a_0 poig}

SR 3.3.2.4

$4 3.3.2.5 SR 3.3.2.4

c. Steam Une Pressure til Steem U'ne Pressure - Low 1,2,3I *I 3 per steem F 7 SR 3.3.2.1 la 405 or l

line SR 3.3.2.4 525M P0 SR 3.3.2.6 See Note 1l SR 3.3.2.6 (2) Steem Une ald) 3 per steem Pressure Nogetive F SR 3.3.2.1 (s 100 Eno $4 3.3.2.4 poi with time Rete - High SR 3.3.2.5 constant k 60 SR 3.3.2.6 secondel

d. T,,y - Low 1,2,3 I81 3 per loop F SR 3.3.2. '. Ik 110 of SR 3.3.2.4 470'F SR 3.3.2.s seeNoie21 SR 3.3.2.4

. iconbnw.ai s @AP600 3.3 13 m i - in ea m ase m 08/96 Amendment 0

_ --- - -- - - - - ~ ~

bfA5 instrumentation 3.3.2 Table 3.3.21 (pege 3 of 11)

Ingineered Sofoguardo Actuation System Instrumentetson REQUIRED NOM)NAL APPL: CABLE CHANktLSI FUNCTION $URVtiLLANCE TRIP MODES DIVISIONS CONDITIONS REQUIREMENTS l SETPolNT

, 5. Turb no Top

, s. Menval Main Feedwater leolabon 1, Refer to Funcoon 6.e (Manuel Mein Feedwater Isolation) for se guiremente.

b. SC Narrow Range Water Level - 1.2 3 3 per SG l SR 3.3.2.1 is 95%)

Rgh 2 SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.8 s, Safeguards Actuadon L Refer to Funcnon 1 (Selegverde Actuetson) for erwtiating functions and re quir eme nts.

d. Reactor Trip <

Refer to Functon 18.e (LSFAS Interiocks. Reactor Trip, P di for requiremente.

4. Close Mein Feedwater Control Velves

a. Menval Main Feedwater toolation 1,2,3,4(m) 2 ewitches U SR 3.3.2.3 N/A

b. SG Narrow Range Water Level - 1,2,3,44m) 3 per 50 Hgh 2 G4 / SR 3.3.2,1 1595%)

SR 3.3.2.4 ER 3.3.2.5 SR 3.3.2.8

c. Safeguards Actuation Refer to Function 1 (Safeguards Actuebon) for alliratieung functione and re quiremente.

> d. Reactor Coolant Average 1,2 3 i

SR 3.3.2.1 (* 542*F'l Temeereture (T,yg) - Low 1 SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.4 Coincident with Reactor Trip Refer to Function 13.s (ESFAS Interiocke, Reactor Tne, P.4) for requiremente.

7. Tng Maen Feedwater Pumpe and Closure of leolation and Crossover Velves

a. Menval Mein Feedweter teolation 1,2,3,4(m) 2 ewitches U SR 3.3.2.3 N!A

b. eG Horrow Renge Water Level - 1,2,3,44m) 3 per SG 4(, 5 SR 3.3.2.1 Mgh 2 (s95%)

SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.4

c. Safeguarde Actuation Refer to Funcuon 1 isefegwde Actuation) for initiating functione and requiremente. '

d. Reactor Coolant Average 1,2 3 1 GR 3.3.2.1 la 542'F'l Temperature T,yg - Low 2 SR 3.3.2.4 SR 3.3.2.5 SR 3.3J.6 Coincident with Reactet Trip Refer to Funcuen 18.e (ESFAS Interlocks, Reactor Tne, P 4) for requiremente.

(centrased) h AP600 3.3-34 08/96 Amendment 0

% i oemesso e.o =

ESFAS Instrumentation 3.3.2 Table 3.3.21 toege 4 of 12)

Engineered Sefegverde Actueben System instrumentet on REQUIRED APPLICABLE NOM NAL CHANNELSI FUNCTION SURVEILLANCE TRIP MODES OlVISIONS CONDITIONS REQUIREMENTS SETPOINT

4. Startup Feedwatet teolabon

e. SG Netrow Range Water Level - 1,2, .4l 'I 3 per 50 ff g SR 3.3.2.1 High 2 ls 95%l SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

b. T,,g - L o w 1,'.

~

3 per loop F SR 3.3.2.1 la $10 or SR 3.3.2.4 470'F SR 3.3.2.5 See Note 21 SR 3.3.2.6

9. ADS Stages 1,2 & 3 Actuebon

e. Manuelliwtiation 1,2,3,4 2 sets of v/ 4 SR 3.3.2.3 N/A 2 smtches 5,6(9I 2 sets of 0 SR 3.3.2.3 N/A 2 outches

b. Core Makeup Tank (CMT) Level 1,2,3,4 3 per tenk

- Low 1 CL/ SR 3.3.2.1 (a 67.5%)

SR 3.3.2.4 volume SR 3.3,2.5

.. SR 3.3.2.6 S'I I

3 per ter* R SR 3.3.2.1 3 SR 3.3.2.4 (a 67.5%1 i SR 3.3.2.5 volume SR 3.3.2.6 j Coincident with CMT Actuation Refer to Function 2 (CMT Actuation) for en irvtiating functione and requiremente, (contnued) e,. h AP6C0 no, m smoosos m eno m 3.3-35 08/96 Amendment 0 o

($fA5 Instrumentation 3.3.2 Table 3.3.21 (pege $ of 12)

Enginotted $sfeguards Attuation System instrumentation

-Rt0VtRt0 NOMINAL APPLICABLE CHANNELS / SURVilLLANCE FUNCTION TRIP MODES DIVISIONS CONDITIO NS REQVIREMENTS SETPOINT

10. ADS Stage 4 Actuation

e. Manuelin tiation Coinc4ent with 1.2.3.4 2 sets of N/ SR 3 3.2.3 N!A 2 switches 5,6(91 2 sets of 0 SR 3.3.2.3 N/A 2 twitches 1

Rcs Wde Range Pressure - 1.2.3,4 (ow 3 CC./ SR 3.3.2.1 (t 1200 psegl SR 3.3.2.4 SR 3.1.2.5 SR 3.3.2.6 5,6(91 3 N SR 3.3.2.1 (t 1200 psigl SR 3.3.2.4 SR 3.3.2,5 SR 3.3.2.6

b. CMT Level - Low 2 1,2.3.4 3 Po' tank CC/ SR 3.3.2.1 la 20% volume SR 3.3.2.4 levetspan)

SR 3.3.2.5 1 SR 3.3.2.6 i -

S ICI 3 per tank R SR 3.3.2.1 Ia 20% volume SR 3.3.2.4 level span)

SR 3.3.2.5 SR 3.3.2.4 Coincident with RCS Wde Range 1,2,3,4 Pressure - Low, eM 3 C8-/ SR 3.3.2.1 In 1200 poig)

SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.4 5(*) 3 R SR 3.3.2.1 lk 1200 poigl SR 3.3.2.4 SR 3.3.2,5 SR 3.3.2.4 Coincdont with ADS Refer to Functon 9 (ADS Stages 1,2 & 3 Actuation) for irutietng functore end Stages 1,2 & 3 Actuanon requiremente (conunued) e

~

._ h AP600 3.3-36 08/96 Amendment 0

_m,m,eamesmaam

LSFAS instrumentation

. 3.3.2 Table 3 3.21 (page 6 of 12)

Ingineered Sefeguards Actwabon System Instrumentebon REQUIRED NOM;NAL APPLICABLE CHANNELS /

FUNCTION SURVEl(LANCE TRIP MODis DIVISIONS CONDITIONS R(QUIREMENTS SETPOINT

11. Reactor Cootent Pump Tnp

e. ADS Stages 1,2 & 3 Actuation Refet to Funcbon 9 (ADS Stages 1,2 & 3 Actwebon) for erutet.ng functions and requiremente.

b. Reactor Coolant Pump Beanng Water Temperature - High 1,2 3 per RCP AA / SR 3.3.2.1 (s 320*F']

SR 3.3.2.4 gMb A SR 3.3.2.5 SR 3.3.2.6

c. gMT Actwebon Refer to Function

((MT Actuation) for c.L.o ~. .. . .. 4 reavirements.

WL 12, Passive Containment Cooling Actuation -

e. Manual Irubation 1,2,3,4 2 sets of 8 SR 3.3.2.3 N/A 2 setchee 5,6('l 2 sets of P SR 3.3.2.3 N/A 2 smtches

b. Containment Pressure - High 2 1,2,3,4 3 F SR 3.3.2,1 [s 3.0 psig)

SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6 I

I (concaved)

L p)sEKT A d e u h AP600 3.3-37

,,,$ -i eowno n-= 08/96 Amendment 0 l

APdoo LCO 3.3.2 INSERT PCr JI.3 3 7 APPLICABtt M00E5 OR CTHER SPECIFIED RE0VIRED Set.E!Lustt FLNCTION

's!P CON 0!i10NS CnANNELS CON 0l?!ONS RE1:AE=E%TS SEip0;st(8) 11 Reactor Coolant P ep Trip

d. Pressurizer Water 1.2.3.4(J) 3 y SR 3321 (a7 Cla)

Lh'I LO" 2 SR 3 3 2.4 ( 1.01)

SR 3.3 2.5 SR 3.3.2.6 4 ") 5(C J3

. 3 R SR 3 3.2.1 r,7.ct.)

SR 3.3 2.4 [al.ct)

SR 3.3 2.5 SR 3.3 2.6

e. Safeguards Refer to Fur.ction 1 (Safeguards Actuation) for initiating functions and Actuation requirements.

l l

e 4

t b

ESFAS Instrumentation 3.3.2 Table 3.3.21 (page 7 of 12)

Engineered Sefeguards Attuabon System instrwmentabon REQUIRIO APPLICABLE NOM 1NAL CHANNELS 1 FUNCTION M00ts SURV(ILLANCE TRIP Olvist0NS CONDITIONS REOVIREMENTS SETP0 TNT

13. Passive Residwel He at Removal Hest f achanger Actuation

e. Manual Initisbon 1,2,3.49 2 switches Vg SR 3.3.2,3 N/A 4th) 503 2 ewitches J SR 3.3.2.3 N/A

b. SG Narrow Range Water Level - 1,2.3.4W Low 3 per 50 Y/ SR 3.3.2.1 Ik 45.000 lbml SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6 Coincident eth Startup 1,2,3.4G 2 per 50 Feodwater Flow - Low E SR 3.3.2.1 la 200 gpm per SR 3.3.2.4 SO'l SR 3.3.2.5 SR 3.3.2.6 c, SG Wde Range Water Level - 1,2,3,4@ 3 per 50 Low Y/ SR 3.3.2.1 la 25.000 lbml SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

d. ADS Stages 1,2 & 3 Actuation Refer to Functio 9 (ADS Stages 1,2 & 3 Actuabon) for initating funcbons and requiremente,

e. CMT Actuat[on -1.2.3.4@ Refee to Functon 2 (CMT Actuabon) for inicanng functons and requiremente.

4(Al, SUI Refer to Funcuene 2.e end 2.b (CMT Actuacon) for irweaung funcbons and requiremente.

Ih~5f.4T B, f (contnued) t

.- h AP600 3.3-38 mtw esemon- 08/96 Amendment 0

. 4 APG00 LCO 3.3.2 INSERT 7C 7,3*38 APPtiCABLE MODES OR OTHER SPECIFIED RIOUIRED StavEttLANCE TRIP FUNCT:0h CCN01T10NS CHANNEL 5 CONDIT!;NS Iil RE'UlsEMENT5

. SETFOINT 13 Passive Resicsal Feat Removal beat En:tanger Actuation

f. Pressu*1rer water 1.2.3.4(3) 3 trains v SR 3 3 2.1 (92t*)

Level, High '3 SR 3 3.2.4 SR 3325 SR 3.3.2.6 O

4 h

a . - _ _- _ - - _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - - . _ _ .

f.5fA5 Instrumentation

', 3.3.2 Tebte 3.3.21 faspe 8 of 12)

Engineered Safeguarde Actweben System instewmentation REQViRED APPLICABLE NOM:NAL CHANNELS / SURy[ftAANCE TRtP FUNCTION MODES OlVist0NS CONDITJONS REQUIREMENTS SETPOINT

14. SC Blowdown isotebon lj l) J

4. Passive Residwal Heat Removal 3 Heat (schenger Actuation Refer to Function 13 (Passive Residwet Heat Removal Heat ( changer Actwation) for ellinitiebon functions and requirements.

b. SC Narrow Range Water Level - 1,2.3.4 9 N 3 per 30 Low J/ SR 3.3.2.1 12 45.000 lbml set 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

15. Boron Oilwuon Block

e. Source Range Neutron nur 210,3, Mwitplic ebon 3 DD / SR 3.3.2.1 Is Source SR 3.3.2.4 Range Awu X SR 3.3.2.5 1.6 in 50 SR 3.3.2.6 minutes) 5 3 M SR 3.3.2.1 SR 3.3.2.4 is Source SR 3.3.2.5 Range Aux X SR 3.3.2.6 1.6 in 50 rn nutee

b. Reactor Top ,

Refer to Function 18.e (ESFt.S Interiocks. Reactor Tnp, P 4) for all requirements.

c. Battery Charger input Voltage - 1,2,3 low 2 per OD / SR 3.3.2.3 1= 343 v'n charger m 3 SR 3.3.2.4 5 M SR 3.3.2.3 (a 343 V']

2 per SR 3.3.2.4 charger in 3 eMoione (contnued) s AP600 3.3 39 mi-teeseenemesom 08/96 Amendment 0

3.3.2.

Table 3.3.21 (page 9 of 12)

- Engineered Safeguards Actuatson System Instrumentation RE QUIRE D APPLICABLE NOMINAL CHANNELS / SURVtiLLANCE TRIP FUNCTION M00t$ OlVISIONS CONDITIONS REQUIREMENTS FITPotNT

16. Chemical volume end Controi System Makeup leolation

e. SG Narrow Reage Water 1,2,3,4JI level - High 2 3 per SG @/ SR 3,3,2,g gsg$g)

SR 3.3.2.4 SR 3.3.2.5

b. Pres sunzer Water Level -

H;gh 1 1.2. 3 83 [ SR 3.3.2.1

[s 30%')

SR 3.3.2.4 SR 3.3.2.5

/gj j SR 3.3.2.6 Coincident wth Safeguardo i

Refer to Function 1 (Safeguards Actuation) for init.eting functions and requaements.

Actuation

c. ,,es.u,i,er w.i.t teve -

i2.3.4(e(.N

, -' 3 por SR 3.3.2.1 is 67% o, High 2 SR 3.3.2.4 74%. See SR 3.3.2.5 Note 3.1 SR 3.3.2.6

d. Containtnent RadioactMey -

High 2 1.2. 3 8h/ SR 3.3.2.1 (s 100 R/hrl SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

17. Normal Residual Heat Removal S yetem isolanon

s. Containment Radioactvity -

High 2 1,2, 3 6b [ SR 3.3.2.1 (5100 R/hr)

SR 3.3.2.4 SR 3.3.2.5

/fM h. SR 3.3.2.6

18. ESFAS Intertocks

a. Reactor Top. P 4 1.2.3 3 dMsions. D SR 3.3.2.3 N/A 1 RTB per dMsion

b. Pressunzer Pressure, P.11 1,2,3 3 D SR 3.3.2.1 (s 1970 poig)

GR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

c. Intermediate Range Neutron Flux. P 8 2 3 ## # SR 3.3.2.1 (a it.10 SR 3.3.2.4 ampel SR 3.3.2.5 SR 3.3.2.6

d. Pressunzer Level. P.12 1,2,3 3 0 SR 3.3.2.1 (Abow SR 3.3.2.4 Pressunzer SR 3.3.2.5 Water Level -

SR 3.3.2.6 Low 1

/WRT e. . eetpoint of 20si (cononued) h AP600 3.3-40 08/96 Amendment 0 mi-ie.com-m

AP000 '

LCO 3,3.2 INSERT f4 --

APPtlCABLE H00($ OR OTH[R SPECIFitD RIOulRED !.RvEIttANCE TRIP FLNCTION CONDITIONS CHANNELS CONDIT:0NS S E]uMEwiNTS 5tTFCINT(8)

17. Normal Residual Heat Removal System Isolation

b. Automatic or 1.2,3,4(*) Refer to Function 1 (Safeguares Actuation) for all initiation manual safeguards functions and requirements actuation signal

18. ESFAS Interibcis

e. RCS Pressure,P 19 1.2.3.4(3) 3 I SR 3 3.2.1 (a700psig)

SR 3.3.2,4 SR 3.3.2.5 SR 3.3.2 6 o

G 8

l b

l

ESFAS Intrumentation

.

3.3.2 Teble 3.3.21 trage 10 of 12)

Eng neered Safegverde Actvet.on System Instrumentation REQUIRED NOM;NAL APPLIC A BLE CHANNELS /

FUNCTION SURv(ll' ANCE TRIP MODE S DIVISIONS CONDITIONS REQUIREMENTS SETP0 INT

19. Conta nment Air filtration System isolation

s. Containment Radioactmty - 1,2,3 High 1 3 $/ SR 3.3.2.1 15 2 R!ht)

SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

b. Containment isof auon Refer to Function 3 (Containment isolation) for irubating twnctions and requirements.

20. Main Control Room isolation end Air Supply Irvt:4 bon

's. Control Room Air Supply 1.2.3,4 2 T SR 3.3.2.1 Radiabon - High 2 Is 2:10*

SR 3.3.2.4 cunestm' Dose SR 3.3.2.5 Equiva!ent SR 3.3.2.6 61311 Note (hl 2 H SR 3.3.2.1 1* 2x104 SR 3.3.2.4 curie s/m' Dose SR 3.3.2.5 Equivalent SR 3.3.2.6 l1311

b. Battery Charger input Voltage 1,2,3,4

- Low 2 per charger CC./ SR 3.3.2.3 [k 343 V'l

, in 3 divisione SR 3.3.2.4 Note (h) 2 per charger H SR 3.3.2.3 m 3 drvisione la 343 V'l SR 3.3.2.4

, 21, Punf'icaton Une Isolanon

e. Pressuriset Water Level - 1,2 Low 1 3 hk / SR 3.3.2.1 12 0.0 % 'l SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6 22, Open IrwContainment Refuehog Water Storage Tonk (IRWST) iniecuon une VaNoe e, Manual inicaton 1,2,3,4@ 2 sets of VW SR 3.3.2.3 NIA 2 smtches 4I "I,5,6 I9I 2 sets of P SR 3.3.2.3 N/A 2estdes

b. ADS 4th Stege Actuenon Refer to Function 10 (ADS 4th Stage Actuation) foriritionne functione and requiremente,

c. Co ncident RCS Loop 1 and 2 4IAI,5,6 I9) 1 perloop Q Hot Leg Level - Low SR 3.3.2.1 la 3 in. above SR 3.3.2.4 bottom inside SR 3.3.2.5 surface of the SR 3.3.2.6 hdt level (conunued) u ao, h AP600 - ,esseeo m m 3.3 41 08/96 Amendment 0

ESFAS Instrumentation 3.3.2 Toble 3.3.21 (pege 11 of 121

-Engineered Sefeguards Actwebon System lastrumenteben i RE QUIRE D APPLICABLE - NOMINAL CHANNELSI SVRV!RANCE FUNCTION MODES TRIP ONISIONS CONDITIONS REQUIREMENTS S(TPOINT

23. Open Ap IRWST Containment Recirculebon Velves

e. Man elinstistion 1

1.2.3.40I 2 sets of Y# SR 3.3.2.3 N!A 2 switches 1

4(al.S.6 91 2 sets el P SR 3.3.2.3 N/A 2 switches

b. Safeguards Actueuon Refer to Funecon 1 (Setegue'de Actuebon! for ellitsbebng funcuens and re quir eme nts .

Coincident with 1RWST Level - 1.2.3.4UI Low 3 3 T# SR 3.3.2.1 ie Contsin-

- SR 3.3.2.4 ment Eleveuen

' SR 3.3.2.5 @ 107'2*l 4(n). S.6 I93 3 0 1

SR 3.3.2.1 Ik Contain-SR 3.3.2.4 ment Elevoton SR 3.3.2.5 9 107'2*l-SR 3.3.2.6 4

1 (conbnved) 4 4

4 i

4 4

- h AP600 3 3-42 08/96 Amendment 0 mt-i : .

,

ESFAS Instrumeatation a .

3.3.2 Table 3.3.21 (page 12 of 12)

. Inpinested Sefe0uerds Actushen System instrumentebon R10VIRED APPLICA8(t NOMINAL CHANN(LS1 SURVilLLANCE TRIP FUNCTION M00t$ OtVISIONS CONDITIONS Rf0VIREMENTS SETP0 INT

24. Spent Fuel Poolisolation

e. Spent Feel Pool Level - Low 6 2 37 g SR 3.3.2.1 137.S ft.)

SR 3.3.2.4 SR 3.3.2.5 SR 3.3.2.6

25. (SFACs Logic

. a. Actuabon Subsystems 1.2,3,4 4 divsions, C SR 3.3.2.2 N/A 1 bettery-backed subsystem perdMeion 5.6,I9I 4 drvisione, K SR 3.3.2,2 N/A 1 bettery-backed wbs ystem pet dMaion

26. PLCe

e. Funcbonel Lo0ic Subsystem 1,2,3.4 4 omsiofw. O SR 3.3.2.2 NA

.+

1 bettery

backed subsystem per cabinet 5,6(91 4 diesions. L SR 3.3.2.2 NA 1 bettery.

backed subsvtem per cabinet (conbnued)

/A158/2 T 21228 h AP600 3.3-43 08/96 Amendment 0 Ape 1wassummelee30DeeJM400M

e APG^0 LCO 3.3.2 INSERT ,

P 4 J, 3 -f'3 APPL! CABLE MODES OR OTHER SPECIFitD RE0V! RED SURvC! LANCE TRIP FUNCTION CONOITIONS C w NELS CONDITIONS GE0VIRE"INTS SETFO!NT(a)

27. Pressur12er Heater Tetp

4. Core Makevo tant Actuation Refer to Function 2 (Core Matevo Tan' Actuation) for al' initiation functiers and requirements.

l 28 Chemical and Volume

' Control System Letcown

! solation

4. Hot Leg Level. 1.2,3.4(m) 1/ loop low l 00 SR 3.3 2.1 (s 3 in, SR 3.3 2.4 above inside SR 3.3 2.5 surface of SR 3.3 2.6 the hoi legs) e 8

9 9

h a

w

E5fAS Instrumentation s a

+

3.3.2 NOTES (n abowwe7:117tmontereree vretmmtom c IP # " 5 e '" ' I # ~' ~

c) Time conetante used in the le*dAeg controner ere ,, = (sol seconde end ,, s (si ees.noe.

(c) wthv.e,bu whe resounsen d-xEn} 2 w.'~n 0 ':4 ': '

(d) Below the P 11 (Pressanter Pressure) interlock.

(e) Wth reactor shut down less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

(f) Below the P 6 (Intermediate Range Neutron Flux) interlocks.

(g) Wth upper mternals in place and refueling cavity less then fun.

(h) During movement of irrediated fuel essemblies.

(i) Wih the RCS pressure boundary intact.

(j) Wth the RCS not being cooled by the Normal Residust Heat Removal System (RNS),

ai Noi opniecoue if the usive ore ciosed, im) u.s...a athnehrowowe e ==m e4-RvMci wiTH ' * "

(n) . Wth the RCS being cooled by the RNS.

(d Not applicable when the startup feedweter now paths are isolated.

.ns--- l Al SE/ 7"&

Reviewer Note: .The values specified in brackets in the Nominal Trip Setpoint column are the SSAR Chapter 15 eafety analysie values and are included for reviewer informanon ordy.

The values specified in breckste fonowed by * *

in the Nominal Trip Setpoint column are typical values for the Function. No credit was assumed for these Functione (typiceny diwree tripetectusoonalin the SSAR Chapter 15 safety analyue end no safety analysie value is eveilable.

The 'Bettery Charger input Voltage - Low

Functione (15.c and 20.b) value specified is a typical value for the Function. The octual value wiu depend on the capabilities of the equipment selected with regard to its abdity to function with degraded voltage so won se the setpoint methodology.

Fonowmg the setpoint study, the values specified in brackete must be replaced with the actual Nominal Tno Setpoints. Upon seleccon of the instr 9 mentation the Nominal Tno Setpointo wiR be calculated in accoNiance with the setpoint methodology described in WCAP 14604. The setpoint calcuta.ione vnu refioct the design basis and incorporate NRC accepted setpoint methodology, Note 1: 405 peig is for a steemGne break outside containment,

, 525 peig le for e steamline break inside containment.

Note 2: 470'F is for e steemline break. 510'F is for CVS malfuncuen.

Note 3: 67% is the norrinal setpoint. ,

_ 74% is the snelytod setpoint.

9 R

s bAP600 a ,e, - ie.co - -

-3.3-44 08/96 Amendinent 0

. ~

'~ 5@dM LCO 3.3.2 INSERT ?MEI.J***

(a)

Above the P-11 (Pressurizer Pressure) interlocs. When the RCS boron concentration is below that necessary to meet the shutdown margin !

requirements an RCS temperature of 200'F. '

(c) Above the P 12 (Pressurizer Level) interlock.

(m) Not applicable for valve isolation functions whose associated flow path 1s l 1solated.

l l

(p) Above the P-19 (RCS Pressure) interlock.

f .

9 e

h e

b

Brira A. McIntyre, Oli27 PM 6/10/97, Reports 4 X Priority: 1 (Highest) -

Date: Tue,10 Jun 199713:27:17 0400 -

. To: TRQ@NRC. GOV _ '

From: " Brian A. McIntyre" <mcintyba@wesmail.com> -

Subject:

Repons - ,

Cc: meintyba@wesmall.com H

. Ted, .

1 If you did NOT receive the PCS PIRT and the revised shutdown evaluation repon today, please call.

- Brian A. McIntyre.

Bell'412.374.4334 .

! WIN 284.4334 l.-

FAX Bell 412.374.4887 FAX WIN 284.4887 Printed for " Brian A. McIntyre" <mcintyba@wesmail.com> _ 1!

Lovato, Janet M. _ _ . _ _

From: Nydes, Robin K.

Sent: Tuesday, June 10.1997 2:39 PM To: Nydes, Robin K ; Deutsch, Kenneth L; Suggs. Charles W.; 'wch@nte gov'; Lovato, Janet M.;

. McIntyre, Brian A; Winters, James W.; Vijuk, Robert M.; Schut::, Terry L -

i

' david.w bland @snc.com'

]

- Janet Would you please put a copy in the informal correspondence box? Thank you.

Executive Summary:

We had a successful meeting in Rockville on June 9 with the l&C Branch to discuss their Tech Spec comments (OITS items 2434 through 2463). Of these,1 is dropped,16 are Resolved,4 are Action NRC, and 8 are Action. .

W. The only difficult one to resolve is the common cause software failure root cause analysis program requested by the NRC and deemed unrequired by Westinghouse and utilir! helpers.

'The Specific Actions resulting from our meeting are defined below and will be reflected in OITS:

l Action - Nydes to update OITS status (with details),

2434,2438 To resolve the NRC concem with common cause software failures. We will submit our original approach formally for NRC feedback. That approach is that any such failures do not need an additional root cause analysis program since they are covered under existing programs. Action - Nydes 2436,2442,2443 To address NRC comment regarding the first inoperable l&C channel (we currently specify no action until the first " required" (i e., the first of three) channelis inoperabia but now plan to take action place the first channel in bypass if it becomes inoperable. This affects reactor trip and ESFAS.

Action - Nydes/Suggs/Deutsch 2451 To address NRC concern that ADS squib valve should have on-line surveillance (or equivalent).

Action: Nydes/Schulz 2024 and 2454 will be resolved as part of the DAS short term availability controlissue. Action: Schulz 2462 To determine appropriate title for trip setpoint column (NRC wants it to say " Allowable" but this is a enenc concern and AP600 doesn't want to set any inappropnate precendent for other plants.) Action:

N des NRC has the following actions:

2440 NRC has the action to confirm l&C definitions are acceptable 2455 and 2456 NRC accepts two qualified monitors but needs to determine acceptability of 7 day - - -

completion time with both monitors inoperable.-

2460 NRC agreed we addressed tester testing question but NRC has action to determine acceptability of 24 month surveillance frequency COT.

Angela Chu confirmed the Tech Spec Branch would not be sending comments such that we have now received comments from each branch. We are sending the Tech Spec markup tomorrow (Action - Robin) and continue to address the NRC containment systems branch comments (Action - Chip) and to incorporate resolution of l&C branch comments into the RTS and ESFAS Tech Specs and Bases (Action - Robin, Chip, and Ken).

Robin Page1

b Westinghouse FAX COVER SHEET e

RECIPIENT INFORMAllON SENDER INFORMATION DATE: 6>- H. n NAME: C., 4% !

TO: LOCATION: ENERGY CENTER -

" Toe h sb/ EAST l PHONE:

FACSIMILE: PHONE: Office: 4 i 2.-p 4 - 4.2w COMPANY: Facsimile: win: 284 4837 i U s WR.c. outside: (412)374 4887 LOCATION:

Cover + Pages 1+ a 3 .

e * ,

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, pkase call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125, m

COMMENTS:

k-ge,. ;,

+L4. t.wd. z. e AA Li;3 k +e %~,1 s a *~ %.b\4 r,. - k yac 6.b TMWs v.q,,,g t , 'Tk;s info. M% Ji\ be Im.acp.cM m b - b + + ra d $i m d d a. T R 6 + T=bia M - Z'i, Wk 'bs i

\b 4 % d mA dacW pop a.,e e.L% /d&l4% 6 desa.

%s s%sL 5 n Tu,u. m.tsj,% .% M Y n 4s # AA p

(.d5 Et~~ 3h wi\\ be. J A4A 4 4 M A cue + . 3 Ta W T i . 29.

ITr 4w e q . b %is inks, M w A N '4 p\u4d **

h . ~ + a . w .f A . v a.A , we. A 1 6 4 < - yo u s.-A4Lt IQ. & fML hs comM h %h M . aihtc h

k p l86 *4-

\ck. us4:Am v bu 6x wp AhA ~~

( Sj C &.

O cc m s..ba

e.

AP600 LEVEL 2 PRA BASED INSIGilTS INSIGHT DISPOSITIO:,

Ib. (cont.)

Depressurization of the RCS through ADS minimsics the poteritial for high pressure melt ejection esents.

Procedures will be provided for use of the ADS for depressurirauon of the Emergency l RCS after core uncovery. dui.mg r. c cic e.Jc .t af;ci cwc ur.co ci,

Response

Guideline; The ADS mitigates high pressure core damage events which can produce large uncertainties in containment integrity due to the following severe accident PRA Chapter 36 phenomena:

high pressure melt ejection direct containment heating induced steam generator tube rupture l , induced RCS piping rupture and rapid hydrogen release to containment Id. (cont.)

l The operator is instructed to open the gravity recirculation lines and Good the Emergency l reactor cavity in the esent that the core has uncovered and PXS or non safety Response injection cannot oc restored. Guidehnes l

l 25. The depressurization of the reactor coolant systern below 150 psi facilitates in. PRA Chapter 36 vessel retention of moltea core debris,

26. De re0cetise reactor vessel insulation provides an engineered flow path to PRA Chapter 39 allow the ingression of water and venting of steam for externally cooling the vessel in the event of a severe accident involving core relocation to the lower plenum.

The reDectise insulation can withstand pressuie differential loading due to the IVR boiling phenomena.

No coatings are cpplied to the outside surface of the reactor vessel which will inhibit the wettabiliry of the surface.

27. The reactor cavity design provides a reasonable balance between the regulatory PRA Chapter 39 requirements for sufficient ex vessel debris spreading area and the need to and Appendix B quickly submerge the reactor vessel for the in-vessel retention of core debris.

25. De design can witustand a best estimate ex sessel steam explosion wnhout PRA' Appendix B failing the containment integrity.

29. The containment design incorporates defense in depth for mitigating direct PRA Appendix B containment heating by providing no significant direct now path for the ,

transport of particulated molten debris fram the reactor cavity to the upper containment reh ions.

A i

V l

0 r- -

Certified Design hiaterial

30. He hydrogen control system is comprised of passise autocatylitic recombiners (PARS) and hydrogen igniters to limit the concentration of hydrogen in the containmer,t daring accidents and beyond design basis accidents, respectnely Emergency The operator action to activate the igniters is the first step in ERG FR.C 1 to Response ensure that tne igniter activation occurs prior to rapid cladding oxidation. Guidelines

31. The containment layout prevents the formation of diffusion llames that can SSAR I.2.

challenge the integrity of the containment shell. General Arrangement Vents from compartments where hydrogen releases can be postulated are away Drawings from the containment wall and penetrations or are hatched and locked closed.

IRWST vents near the containment wall are turned to direct releases away from the containment stell.

32. De containment structure can withstand the pressurization from a LOCA and PRA Chapter 41 the global combustion of hydrogen released in-vessel (10 CFR 50.34(f)).

Sesere Accident

33. The steam generator should not be depressurized to cooldown the RCS if water Ntanagement is not available to the secondary side. This action protects the tubes from Guidance

large pressure differential and minimizes the potential for creep rupture. Framework Severe Accident

34. Depressurizhg the RCS and maintaining a water level covering the SG tubes hianagement on the secondary side can mitigate fission product release, from a steam Guidance generator tube rupture accident. Framework G

e 2

o 8

- TX CONFIRMATION REPORT ** AS OF JUN !! '97 13:50 PAGE.01 APG00 DESIGN CERT DATE T l ?1E T0/FROM MODE MIN /SEC PGS STATUS

-01 6/11 13:48 #23tHRC G3--S 01'29 03 OK G

1 .

1

rs: Bill Hoffman sha

&nfaCha (Jvp s Terry Shuis

[en han /)1'bermot+

cc : % IUin/ cts 6n' an Ol'/nfyre a

Mhched e an on3 updak basedon our June 57 in Q(refer lo achims jii shku f'e/d).

1 60/- Wwid yev pTease ce Anyla n thk. 7ha hs.

Rneda luovW > ca// rra wth an A)2C s Er aem any ? Thanks.

bbin.A]feS 5

e

- Al%80 Open Item Tracking Sy; tem Database: Executive Summary 1 ' Date: M2/97

' Selection: ' ptem nol between 2024 And 2024 Sorted by hem #

item ' . DSI.R Section . Thic,1ksesiptke Resp (W) .NRC No IL -ich Questam - 13 pc : Iktml Status - Engy Staus -~ .Statah truer h I_ ' Danc '

2024 NRR/IllCil 16 DSI R-OiSO TICitSPEC/Sd:utz Dropped - Actam W

~ ~~

l28. tksign of the Iberse Actuation System ~'

! ~ he DAS has been identified by Westinghouse to be a RTNSS-anctutant system for A1WS camsaktmeses the staff needs addamuusi -

'information regarding the design and reliabilwy (See DSER Open leem 7.72-1)

' ~ ~ ~ ' ' ~

Closed - SSAR Chapeer 7 revised to addresk

,Per iII.*1 telecon. NRC has action so dncuss lack of DAS/ lech Spec relatannhep methin the staff rin 12/2.

!I orwarded report copy ofIhis item to NRC for con &matum that NRC simus should be Actum-N to deternune af thrs iecm is resolved sin 4/85N7

Dropped - Dropped per teleplume derective from Diane Jackson on 4/23S7. .

'During a (W9 W/NRC rnig in Rockwalle, this shoulJ not be dropped it reises on campiction of the DAS samut term atadabeldy controis (see setaned ,

lseem2454) rLn6/10 j 4

i I

i i

P.nge: 1 Total Records: 1

_ . _ . . . _ _ _ _ _ m ._.

AP600 Open item Tracking System Database: Executiv Sommisry. Date: W12197

' Selection: [iterit noj between 2434 And 2464 Sotted by item #

Item DSI R Sectke . Title /Descriptum Resp (W) NRC No Ilsanch Que> tam Typc Detail Stahes Evigineer Staties Status

._ _ _ _. __ _ _ . _ _ _ _ _._1.c_ ne. r No. /

. _ _ Dan..e 2434 NRR!IllCB 16 i M'IG4)I TEGISPEC Action W Actum W

~

Pmbde Westinghouse possten on admm{ontrols section requirements for aAnf}t Closed - Wsth issuance of ale Tech Spedin SSAR E9~

~ ~~

]re fas iot caine analysis program [ "

' ' ~ ~ ~ ~~~~]

As agreed during W/NRC sntg 6N in Rockville, Acison W to formally proviue posstion regard:ng NRC request for (conunon cause) software fadame get cause WWs pmgram p &l0 __ , , _ _ - _ _ _ , _ _ _,,

._,,__~,,,-_,3 i 2436 NRR/IllCil 16 i MTG-O! TEGISPEC/Busa Actum W Actum W

[ Evaluate the actions for one Ahanne[incpershic versus fingle c!Ermels multipit ms moperable due to reacsor inp gioikfadd

'Gosed - With issuance of the Tech Specs in SSAR Rev. 9.

.. . _ . . . . _ _ . . , -((]_ . . . . . - p

. As agreed durmg 6N W/NRC notg in Rockville, this is Action-W. This =di be addressed in a revision to the RTS and ESFAS TS en take actson to ibypass u ._given _ one

_ _inoperable

_ . _ . _. channel, . . _ . not _ _masteg _ . .inned _ _one. . acquwed channel is inoperable. This mail close mah mems 2438,2442 and- 2443.

rko &l

-.__s i

2437 NRR!!IICB 16 i MTG-OI TEO tSPEC/Busa Gosed Resolved !

d

~ = -. _. - . a . n..< _-.i

~ ~ ' ' ~ ' "

%dIWithIssuance of the Tech Spedin3SAR'Rev.'9 . - . . + ,]

jUtdsry input was sought but not used since STS completion tunes and survedlance frequencies were used This was statused resolved durmg 6N nagi ;

twith W and NRC in Rockvdle. rka 6/10 I w_ - _ _ _ _ , . . _ . _ . _ _ . _ _ _ _ . _ - . _ _ _ _ . - _ _ _ _1 TFOISPEC/Barsa Actum W Act on W I 2438 NRR!!!!Cil 16 i MTG-Oi

~ ~

l (Consider providing (in Bases')iat5anaE for Miisistem aIctionsIersus oper~ator acteoes (mhich may reser*e the system autornatac actkes)f e g ,

i

like when a channel trips in the auto function only to be bypass as inop by the operator based upon actkm steps) ;

=======---========:.==.;:

2-

,Oosed - With issuance of the Tech Specs in SSAR Rev. 9

' f As agreed during 6S W/NRC mag in Rockville, this is Action-W. This will be addressed in a resision to the R TS and l'SFAS TS to take action to i

bypass given one inoperable channel, not waitsng until one required channel is unoperabic. Thes edi close math items 2435. 2442 and 2443. rio 6(IJ ,

2439 NRR/IllOs 16 i MIG 4)I TEGISPEC/Busa G osed Resolved

~ ~

Defmstion of channel is froen the sensor through the output at the rector inp subgroup ' Need so include ehere applicable (10 defestkms. 3 3 h 3

[*'C k._ - - _ _ l

~

.Omed - Wah issuance of the Tech Specs En35AR Rev' 97_ . . . . . _ _ _ _ _ . _[

] l

{nis was incorporaged into the previous Tech Spec rev p B 3.34 Statused Resolved durmg 6N mig usth W/NRC in Rocksille, ska 6/10 )

2440 NRR/IllCil 16 i MTG-OI TEOtsPEC/Busa G osed Actam N ,

]

~

i

{ Verify that'the'NUREG defenst [ar~ e approprase and include en the Apis 00 _

' ' ~

~ ~'

~ - '~

[ Closed A With issuance of tie Tech Spedin 5SAERU. !

{At the 6N W/NRC mtg in Rockville, NRC took the action to cofum appropriateness of1&C defnetions rka 6/10 NRR/IllCD 16.I TidlSPEC/Busa Caused Resolved 2441 ,MTG-OI

~

fThe term auaomatic trip lo[ic* is utilUed sithEthe"T5mathMdcrind Det'ermee the appropriate.ern for use in the is if a '

is used, then derme the term. {

===:= = = -- ::: = =: . = : =:_ ' . . - ,= .=: ,

Closed - With issuance of the Tech Specs in SSAR Rev. 9. ,

{During the 6N mig mgW/NRC in RQs die, this change was made in alie previous TS revyn p B 3 3-7. Statined Resolved ska 6/10_

_ f I

t Prge: 1 Total Records: 30 )

Al'600 Opea Item Tracking System Datnbase: Execttiv2 Summary Date: 6/12/97 Selection: biem nol between 2434 And 2464 Sorted by item 0 Item DSL R Section 'litic/tkscriptwn Resp (W) NRC No liranch Question lype iktail Saatus Engencer Status Status letter No / Dase 2442 NRR/IllC11 16 3 M iG-OI lECitSPEClliersa Actmin W Actm.wi W twaluate the potential for fadures that could defeat the capabdity for placing functions into by' pass Thes needs to be considered in the dewtopment j of the actions (operator could be required to take an action to put the channel in bypass. and be unabic to perform the action) At scry mensmum, the bases should clearly esolain what is meant by placing the channel in bypass. Is taking the actkm (switch operstkm) without the sysicm succeedmg

oL' This action is there to go from 1/3 logic to 2/3 logic (which affewds operating fault toicrances) Staying ui the 1/3 conditam is not unacceptable _

fThis concem is vslid for teth the RPI and ESF. ~

^ ~

{

' Closed i With issuance of the Tech specs in SSAR Rev. 9 ~!

l As agreed during 6/9 W/NRC msg in Rockwdie, this is Act.on-W. This will be addressed in a revisson to the R TS and ESFAS TS to take actam to !

i bypass given one inoperable channel, not wasting untd one required channel is inoperable Thes will close with nems 2438.2442 and 2443 rLn 6/IQ 2443 NRR/lilCis 16 1, MTGe TTCitSPFC/Barsa Actam W Actum W

[Cornaunicatnon of fasied by passed condetsons is a concern within the estrumentation sectwn

~

j T

~

trioAd %;ith'issuancc of afETecESpecs in SSAR Rev[9 ~l l As agreed during N9 W/NRC mig in Rockville, this is Actam-W. Thss wdl be addressed in a revisam to the RTS and ESIAS TS to take action to {

MPass ggen one umpeggp_not wasng untyoneyeQ channelis umperable. Jhys win pse wnh a 24R 2442 and 240 de q 2444 NRR/lilC11 16 i MTC 4M TECilSPEC/Birsa Closed Resolved

~

'Descussing' ESF actuation cchelon~lf an ESFAC [AI or A2) fads, need tolenfy Iog5c cabinet is operable 4 I __ ._ _ __ _ _ _ _. _ _ _ _ _ _ _ _ _

! Closed - With issuance of the Tech Specs in SSAR Rev. 9.

Statused Resolved during 6/9 W/NRC mtg m Rock, die. This applies only to battery-backed cabincts and that es now Tsed p 113 3 w and SR

]j 33 2 2. An 6/10 .

NRR/lilCil 16 i MlG4)I TE CitSPI C/11ersa Closed Resolsed 2445

Re-consider development of ESF actuation table and consklerstkm of fadures Consideratkm needs to enclude fadures that defeat multiple f.sn

~

Miosed - With issuance of the Tech Specs in SSAR Rev. 9 ~ l

{Statused Resolved during 6/9 W/NRC mig in Rockwille. Coscred in ESFAS functkms 25 and 26 ran 6/10 1

16 i MIG 4M TI CitSPEC/Ilarsa Closed Resolved 2446 NRR/1llCil ~

~

{Agccment reached that LSF channel definstion is'frorn sensor through the outkot ESF subgroups (ISFI and i SF2)

~ ~ ~

~'~

I l

(Clos'c'd'-

Statused PesolvedWith issuance during 6/9 W/NRC mag inof the Appears Rockwdie TechonSpecs 15 p 3 3-54;in rLnSSAR 6/10 Rev[9 j 16 i M TG-Of TLCllSPI C/flersa Closed Rcwived 2447 NRR/1llCll

~

j l Assignment of ESF functions to ESF subgrups'and logic processors simuiJ be consistent with dacts.ty amoung functxes $Pecifkally, dn j functkms should be piaced so that fluid system dacrsity is mamtained thmugh the PMS where appropriate. Fudd system designer efhwt jCTs simuld help identify these functions. l lCloscUWith issuance of the Tech Specs in SSAR Rev 9. ' ~ f h"'*I._c g y punn yNRC mtg in ,Rottsille. We didn1 taic credit for iunctional dnerssty of 2 ISF processors rLa 6/10 MIGEOM TECilSPEC/Bersa Drepped Ikopped 2448 NRR/lilCil 16 i ~

o l'gi{ block (no irEludmg the output actuatkm segnals) l

[Line for logic miry t[dden the middle of the

~~

{No actErequired (lf t iident(sed for delAion[%esting! mess ~c cimc'uNedI

! Issue dropped frorn " Top 50* last_

jClosed - With issuance of the Tech Specs in SSAR Rev 9.

iC?*Id"*f8'*U dyd" during 6/9 W/NRC mig in Rocksdie. rLa 6/10 Page: 2 lotal Records: 30

AP600 Open item Tracking System Database: Executiva Summary Date: 6n2/97 Selection: blem no] between 2434 And 2464 Sorted by item #

Item WI R Section Titic/Dewriptum Resp (W) NRC No Branch fype Detad Status Engineer Status Status i etier No f 9 ,e Question 2449 NRR/IllCB 16 i MTG4M TI CitSPI C/Busa Closed Residsed

{ Manual actuations of I_SF should address the} dedicated cetrt,ls (system [ level manual actuatums) l

' ~ ^ ~ ~ ^~ ~ ' ~~

l Closed - With issuance of the Tech Specs an SS R kcEU ~ ~l ,

{Statused Resolved during 6/9 W/NRC mtg an Rockv}dle NRC agrees manual actuatums are smw mcluded sin &l0 j

2450 NRR/ll!CB 16 8 kfIG4)! WCllSPFC/Birsa Closed Resolved

~ ~

ConssJer card failures that affect multiple manual mauations (1/0 card faduret versus failures that afTect am*eveJual manual actuaisons (anpet dessce ' I

. l 1 fadure) _, ,

~~. _ _ _

i. _ _ _ _ _ _ _ _ ._

~ ~~

~

I i (

l Closed - With assuance of the Tech Specs en SSAR Rev.9Statused Resolved durmg 6/9 W/NRC mtg in Roc I

l(see 2436 and 2445 kw other multiple failure concerns) rim 910 NRR!!!!CB 16 i MTCMM TECI:SP1 Cischuli Actum W Action W 245i

~

[4tti Seage ADS valves will' need to be wd fuity{ Evaluate manual' actuation precedents few flWRs relatsve to squel5" valves j ,

^ ~

~

'j

~ ~

~

lCloscd - With issuance of Une Tech Specs an 5SAR ReEv 9' "

During the 6/9 W/NRC mig in Rockville Westmghouse agreed to consider an on 4rie survedlance (or equivalent) for the ADS squib valve. The f actim is to add an apprpate survedlance,in addition to the IST every 24 nasnths. rin us 0 j NRR/IllCB 16 i MICM)I TECilSPEC/Barsa Closed Residwed 2452 identify and verify ESF signals that may not cowhd % CAP 11633 Figure 2 7 architecture (radiation sagnals faw example) f

, ~

l Closed - With issuance of the' Tech Specs in $5AR Rei9 Resiewed subject figure during 6/9 W/NRC mig in Rockvil:e ar.d NRC considered this resolved No change to the DitHbersity % CAP-11633 lAlso, the right channels are in the TS when compared with the SSAR. rin &l0 l 1ECllSPE C/Hersa Closed Resolved 2453 NRR/litCB 16 i MIG 4)I I

Nf.S actuation / control room isolatoon technicai' specification sinnsid be encluded wette other ESI' fJnctums Assurned logic for separarmn an th:

standard specs is that these fucntions were typica!!v not included an the Westmgfwuse prosectaca speems f Closed - With issuance of the Tech Specs in SSAR Rev. 9. .__ - - i jStatused Resolved during 6/9 W/NRC meg in Rockwdle included m previous 15 rev as l'uncamm 20 p 3 3-41 (Bases p 113 3-92/93 =as marked in

. master markup durmg the msg to correct appiscable modes) rLn &l0 TECitSPEC/Schuli Actam W Actum W 2454 NRR/lilCH 16 i MIG 4)I

'~ '

Clarify NRC position on technical skcniscatEfd DAS l

--=========.: ,

Cloud With issuance of the Tech Specs in SSAR Rev. 9 .

!Actiors W to close with subnattal of DAS short term avadabdety controls (_cc related stem 2024) rLa (.10 l TECllSPEC/Itirsa Closed Actum N 2455 NRR!lllCH 16 1 MIG 4)I

Evaluate fadure of a desplay N al)DPS bcs diat' result in the unasadabdity of one disgday Need to consider ;

safety argume completion time to correct Spare parts censiderations could be significant._.__.__ .

~

l

[ Closed [Wsth issuanceAf'the TcciSpecs in SSI~IfRev. 9

'During 6/9 W/NRC mig in Rockwille NRC agreed two qualdeed numitors is adequate Imt took the actum to deterwinne af the7 day compictum twee es.

l

[acceptabic wsth thith monstors inoperable. See 2456. sia &l0 _ _ _

Page: 3 Total Records: 30

AP609 Open Item Tracking Syst.:n Database: Execttiva Summary Date: W12/97 Selection: blem no] between 2434 And 2464 Sorted by item 8 Item D$l R Sectkm ' Tale /Descripuon Resp (W) NRC

- No lirarKh Type. Detal Status Engmeer status - Status Questkm ___ __ lener_No / _ [ thne__

2456 NRR/IllCB 16 i MICA)I TECilSPEC/Busa Closed Action N

~

{Wath two failures of displays and iui s~afe' ty-related d splays av'astalk is the 7 day cumpletaon tame appropnale.

~ ~ ~ '- ^ ~ ' ' ~ ~ ~ ~

]

Closed - Wah issuance of the'TcUiSpecs in SSAR iteIr$9 ~ j During W9 W!NRC enig in Rockville, NRC agreed two quahfied amenors is adeq sate but took the actam to determee if the7 day completum tune es!

with M nmmyrmgabicJecp_E Qg. . _ - _ - - - _ ,-

,l 2457 NRR/lilCB 16 i MTG-Of TECitSPEC/Busa Closed Resolved

~ ~ ~ ~

!For PAMS, standard' specs were determined uinig all hpe A, and category I varialdes ConsEle'r'usief actum Mno EEU 63) producing a specsal repat Need n consider how this apptjes to AP600 PAMS and AP600 RG I;97 categorization

~

~' '~

~ ~

[ Closed iWsth iduandof the Tc'ch Specs'En SSAR Rev 9 .

[

jDunng the &9 W/NRC nug in Rockville, this was statused resolved sance the AP600 has no Type A sanatdes and has completed the RG I 97

evaluation rLn #10 __.. ._ .___ ._ __ _.__

t_. . _ . . _ _

NRR/lHCB 16 i MICM)1 TECitSPEC/Basa Closed Resolved 2458 Determine completion tunes for functions from the remote shutwwn workstatims Completson tunes should seflect less IAc!dmed of use Closed - With issuance of the Tech Specs in SSAR Rev. 9 Along with 2459 and 2460,these stems are NRC suggestions for TS improvements that have been abandomes cts are STS-based,me design based

.This was statused Resolved during the 6/9 W/NRC satg in Roci ville. ran #10. - - . -- - - . - . -_ - ._.

16 i MIG 4)1 TECilSPEC/Barsa Clowd Resolved 2459 NRR/lliCB RC L ?"' *_0'?? $** "E*'0** 8"I*' *"0'**" W **O 0"_H "?'*S.?' **0 qda _

~~

~~ ~ ~' ' ~ ~ ~ ~

~ ~ ^ ~~l} '

I Closed! With' issuance oWTectiSpecs in SSAR RevT91 Along with 2458 and 24f.9 these nems are NRC suggestums for TS unprovements that have been abandoned. cts are STS-based, not design based ;

y y was statusef Qved during the gyC mig in Rocksille sace we are not taking credit for self4hagnosatics in TS. rLn W10 TICitSPEC/Barsa Closed Acuan N 2460 NRR/lliCB 16 1 MiGOI s .

L _

4

! Closed - Wuh issuance of the Tech Specs in SSAR Rev. 9 Along with 2458 and 2459, these items are NRC sanggestions for TS improvements that have been abandoned cts are S1S-based, n(m design based

During the 6/9 W/NRC mig in Rockville, NRC took the actum to determene the agde aacy of the 92 day and 24 moenh OOTs rLa ulo TECllSPEC Closed Rewdved 2461 NRR/lllCB 16 I M iG-OI ~

~

[NP50 detenNinc5ANiddthe credet g[ erie s$1fEinagibst feEurUfIIEduaEliased s[ stem wdlun ihe detelopment_

~'^~

] clic D

~ ~ ~ '

'C5Es~eUWith5Esirice~of tiie~TNE$ pees in SSAR Rev. 9.

[ Good idea but. this was statused Resolved during the 6/9 W!NRC msg in Rockedle given that we dsJ nut take credd for the selfahagr 4rLn &I0 . _ _ _ _ . _ . _ _ . _ _ _ _ _ _ , _ _ _ _ . _ . . _ _ ._

t TECalSPEC/Barsa Acthe W Rewdved 2462 NRR/lilCD 16 i MTG4)I ~~

!Determme approprimiciness ofAmrini:Eng EEc~luiciis statement (allowing for adjustmentAsuices'sary of the setporras are wnha the requwed rangt

[and accuracy) in the definitke for channel operational test. ^

~

MoEIl- Wah issuance of the Tc' ~liSgEcs' c in SSAR Rev. 9

'With confirmation that these STS words are aporr priate for AP6@ Qn that senpoents could be abasted at the transmitter or mdsrectly), thes mas statused Resolved during the &9 W/NRC nwg .a Roth,ille. sin 6/I2 _ ___ ,

Page: 4 Total Records: 30

.n .

. AP600 Open Iseen Tracidag Sysseen Desabase: Ezecesive C y Dece: WI2/97 Se4ectooe: inern nel betscen 2434 And 2464 Sorted in tscen # .

~

b D51 R Secsaus Talet)euripense RC5F tWB NtC No Ikaah Questam Type Deted %tanus Esguicer Stase Stama t eser N.s,.. /

. - - . - . - . - . - - , --- . . - . _tw 2463~ NRR!!aCB 16 i MIGa T10tsrt C/Besa Gosed Aame W

~

Coender NRC secenumendshen Ear uncludag ceeialushalkveable oshses er my senyonei Deternume she feasabdey of usang sadety analyse omd

$" M* I"' d'S'enyg specs _ _

__ ___3,_ _ ~_ _ _ _

~

~ ~ ~ ~ ~ ' ~ - ~ ~ ' ~ _ ~,

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

~l

~

!oosed - Wnh meance of she Tech specs uTSUR AN 9.~

NRC now wants she weed " Allowable

to appear in the columen W (we had eyeed for ene setynung Thss as a generic issue and Wesemi d==mse j has the acisse se desenumme what is moons appsername for Af600 waamma scenag me sempreaprune p+ fur echer yIames safety aussysn salues are 3

no semisme brackened. rks &t2 l-w...._..._.. _ _ _ . ._ _ . . _ _ _ . - _ _ . . ~ . . _ _ . _ . . _ ._

2464 NRR/5CSR 16 1' MTGame TECH 5PEC14cDennot Oosed lascarne

~

(The T5 shuund anchale cosisadershoe of pene! smal power h eone[dueZ_ aur beMe h should be smeed IInes the t'SNRC ss L*""'d '6"" *c'""5 '"'..'** P"'P' ""d

~

1""d"'7.M**'""""'***"* ~

~ ~

[ comed TNah issuance of Eiech' Specs is SSAR Rzw[9_.. '

jNRC to connne ihas is sesolved by sewieweg SR 3 6 6 5.page 3 &l6 Wc.J . __ beheves shes ss adersmed (wasch was chsiged fr == lacs so I M d"""EM "88

E " M "8ICI M 32 . . _ . _ _ . . _ _ _ _ _ . _ . _ _ . . . . _ _ _ _ . .

4 Page- 5 Total Records: 30

- - - . . _w- _ s

i 2

5 0 5 2

_50 4 5 9 2 8 6 0 3 6 9 1 1 6 1 9 5 4 6

-

2 4 i

m / / / / / / / /

/ /

e - _

e 2 5 T 5 2

1 0

8 5

6 1

2 6 _

0 1 0 4

3 9

6 5

5 4

9 9

6 2 4 l_

5 6

_4

_ 3 _ 06 0 _4 4 0

_ 5

_ 2 9 04 2 3 0

_2 9

/ _ / . /

d / _

_ e 6 _

/ / /

,_/ / /

h 0 7 7

2 6 -_ 0 __ 9 8 6 6 9

0 . 6 ,_1

_ 5 1 C 1 1

- 2 1

4 7 2 ._

3

,1_ 3 0 0 2 _. 2 J.

L_._ f 1

0 0 _

5 3 6 2 2

0 . 9 1

_ 5 1 1 2

0 d .

1 _

e h

s e

/ / / / / /

/ / / / _

e 5 0 0 .

R 0 __. 0 _

0 3 0 0 8 8 4 7 9 8 1 2 0 2 3 3 1 3 1 0 1 1 3 3 N 7 2 1 s

se / / / / / / / / / /

i m t 7 e c 4 1 1 0 7 0 4 6

2 t

s ) A 5 4 6 6 y C 1 S t R g ro N 2 4 8 2 2 n 2 5 2 0 i p / 1 1

8 0 3 2 2 1

4 2

3 da e W 1 R e l W

(

se / / / / / / / / / /

r i, mc T ut wc 7 0 2 2 9 5 7 e

e 7 2 4 0 6 5 2 2 A

ma e t l

s e

9

/

2 1

s u

7 2 3 1 1

h S i t

/

6 t

a n

t ru t e ec S

{ 0 0 0 0 0 0 0 0

0 O .

pj o f o n ._

O r s s i e M / / / / / / / / / /

e e 0 P d e

0 mc s u

m d 1 0 0 0 l u 0 0 0 0 6 e e a e s A _

0 1 P l e

A s S t

e _

R _.

N- 1 0 0 0 2 0 1 0 0 4 m

r / / /

f / / / / / / /

n o

C 9 0 0 0 4 0 2 4 0 9 1

1

- l

_ _ - O W -

3 1 o 1 0 0 0 4-0 0 8 m ._ 1 f

r e

/ /, / /

/ /

/ - / / /

e 4 _

C 1 0 0 . 0 _ 5 0 3 0 0 2

- 2 d

0 0 0 _ 06 1 0 1 0 0 2 e

r p / / / /

s / / / / / /

6 r

D 0 0 0 0 6 1 0 1 0 0 2 6

y r _. .

o _.

_ - yro _.

1

_. t a .

0 -

- im. ._

_. t r

_ _. a e c

f n _ 0 _m n

. fa l

. M 1 o _ o3 1 f

O r e

_ 0 C C. 0 I n o_ g .

- _. e e _

mn f

u

. R - RL R R _ U __ C o s s _

_ S _ S _ S _E

_. E t c l I l I e - e a

_ S A A _ e - l y e

t

. D D _ D D _. R R M.T e -

K T o _

41 '/2 9/

h Westinghouse FAX COVER SHEET W

RECIPIENT INFORMATION SENDER INFORMATION DATE: 7veJ 'c II , 19Ci '7 NAME: 66Vd~ [/[LIC A TO: g, LOCATION: ENERGY CENTER -

J OC FS fA$ l'/ EAST 32(,S PHONE: FACSIMILE: 6,l - 30).4 l5-200 2- PHONE: Office:(4123 3,H 53)o COMPANY: Facsimile: win: 284 4887

@N outside: (412)374 4887 LOCATION: DJ H tw Ric Mb Cover + Pages 1+ ' l '

\ .

The following pages are teing sent from the Westinghouse I:nority Center, East Tower.

l Monroeville, PA. If any problems occur during this transmission, please call:

l WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

f Joe ,

Oo L_ les ne sc s -ro +e cm,om- shwat f&.~E '5 ovi _fm 3TP cd,tates hs A .G= tau ~r or mr n a desa~ ~ es .e , A=

RucusnHn gc.779L , .3 s rin oHc Yo 0 ruf c ems es=we . _.

dem re knm cJax.<>rn nos. -

m o uw . ' Ccwr J

o, CN V}

f Westinghouse ' Energy Systems b n5 Electric Corporation " 5

N 8*8'5 N 55 NSD.NRC.97 5174 '

DCP/NRC0907 I Docket No.: STN 52 003 June i1,1997 Document Control Desk j U.S. Nuclear Regulatory Commission j Washington, DC 20555 l ATTENTION: T. R. QUAY

SUBJECT:

WESTINGilOUSE RESPONSES TO NRC FOLLOWON QUESTIONS REGARDING TliE AP600 INITIAL TEST PROGRAM (ITP)

Dear Mr. Quay:

Enclosed are the Westinghouse responses for additional information from the Containment Systems and Sesere Accident Branch, relating to the AP600 Initial Test Program (ITP) as requested in your letter of May 14,1997 from Mr. Joseph M. Sebrosky of your staff.

Enclosure I to this letter contains the followon RAl's related to ITP in your May 14,1997 letter and Westinghouse's responses to close RAls 260.118 through 260,137. (Responses to RAls 260.138 and 260.139 are still being prepared.) Enclosure 2 contains the proposed SSAR changes to Chapter 14 resulting from these RAls. Additions to the SSAR are shown in italics; deletions are shown with a strikeout line.

Note that these proposed SSAR changes to Chapter 14 include testing related to design changes incorporated for post.72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months functions in subsections 14.2.9.2.7, Spent Fuel Pool Cooling System Testing; 14.2.9.1.16, Long. Term Safety Related System Support Testing; and 14.9.2.1.3, Passive Core Cooling System Testing. Enclosure 3 to this letter is subsection 14.2.9.2.4, Normal Residual llent Removal System which has been revised to include testing to remove heat from the spent fuel pool; and subsection 14.2.9.l.4, Passive Core Cooling System Testing which has been revised to include testing of the IRWST gutter isolation valve controls.

Westinghouse requests the staff review these responses and inform Westinghouse of their status to be designated in the "NRC Status" column of the OITS. We suggest " Action N".

Westinghouse requests you provide any comtnents on these responses as soon as possible so that the milestones to SECY.97 051 can be met. ,

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

M Brian A. McIntyre, Manager ,,/d Advanced Plant Safety and Licensing

!ca Enclosures cc: W. C. liuffman, NRC (SEl, SE2, SE3)

N. J. Liparuto, Westinghouse (w/o Enclosures) nu.

e e

NRC REQUEST FOR ADDITIONAL INFORMATION Enclosure i NSD NRC 97 5174 DCP/NRC-0907 e e t

0 T Westinghouse l

NRC REQUF" FOR ADDITIONAL. INFORMATION Queston 260.118 (OITS 5296)

Re:

NRC Lener May 14.1997. The following is a general comment on Section 14.2.9.1.4. The purpose of the testing in this section is stated in terms of the safety related function "to transfer heat from inside containment to the environment".

Additional testing objectives need to be incorporated into the initial Test Program (ITP) to validate the expected PCS f wetting characteristics. Additional testing objectives need to be incorporated into the ITP to validate the oserall heat transfer characteristics used in the design basis accident evaluation model which are dependent on the as built i structures. (RAls 260.119 through 260,130 are speci6c examples of the above general comment).

Response

The response to RAls 260.119 through 260,130 address the speci6c aspects of this general comment.

SSAR Revision: NONE 9

e e

260,118 1

NRC REQUEST FOR ADDITIONAL INFORMATION Queston 260.' U (OITS 5297)

Re:

NRC Letter May 14,1997 The testing purpose needs to be expanded, it needs to be clear that there are distinct periods (three) of Cow which need to be evaluated as well as the period of performance,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Response

Item (c) has been clanned that testing will be condacted to cover the entire range of expected PCCS Dow rates.

SSAR Revison: See attached Section 14.2,9,l 4, Passive Containment Cooling System Testing e e 6

4 t

260.119 1 YMW

~ -

NRC REQUEST FOR ADDITIONAL INFORMATION "4

\

l 1

Queston 260.120 (OITS 6298) 1 Re' NRC Letter May 14,1997 The passive containment cooling system water storage tank (PCCWST)is now also used as a safety related makeup source for the spent fuel pool (see SSAR Section 6 2.2, Revision 11. February 28,1997, page 6 2 21). This should be stated under Purpose, as is the fire protection function. Also, the description of the new, isolated Gre protection tank within the PCCWST should be provided. Appropriate testing for the spent fuel pool makeup function needs to be developed and referenced in the Initial Test Program.

Response

Testing of the passive containment :ooling system, spent fuel pool cooling, and long term safety system support has been modined to appropriately redect the new NRC criteria regarding post 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months operations.

SSAR Revision:

See attached Sections 14 2.9.1.4,14 2.9 2.7, and 14 2.9.1.16 Passive Containment Cooling System Testing, Spent Fuel Pool Cooling System Testing, and Long Term Safety System Support Testing.

e 0

4 260.120 1

NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.121 (OITS 5299)

Re

NRC Letter May 14, 1997 . Under prerequisites, the quantity of water a$ailable in the PCCWST needs to reDect an amount sufucient to demonstrate that at the minimum lesel(volume) specined in the technical speci0 cations, the PCS will provide at leut 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continuous cooling water.

Response

The prerequisites state the requirements for the ability to Oli the tank. Item (c) of the passive containment cooling system testing will be modined to include a test to show that the tank has suf0cient capacity to lut 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

(

l SSAR Revision: .

See attached Section 14.2.9.1.4, Passive Containment Cooling System Testing.

4 e

e 260.121 1

NRC REQUEST FOR ADDITIONAL INFORMATION Queston 260.122 (OITS 5300)

Re:

NRC Letter May 14. 1997 Under General Test Acceptance Criteria and Methods, the reference to Sectbn 6.2 should be limited to Section 6.2.2. " Passive Containment Cooling System? only. This test does not coser the other sections.

Response: -

l The comment will be incorporated.

SSAR Revison: .

See attached Section 14.2.9.l.4, Passive Containment Cooling System

Testing.

4

?

1

. r 4

i i

260.122 1 I

$MW i

I L

-. -.-s-_.--.. . .. . . . _ . , _ _ . _ _ . . . . _ , _ . , ._ - -- , _. . . , - - - . _ - - - - _ _ _ - _ -- - - _ . . - - _ , . . - . _ . . - _

NRC REQUEST FOR ADDITIONAL INFORMATION' Question 260.123 (OITS 5301)

Re:

NRC Letter May 14, 1997 Under General Test Acceptance Criteria and Methods, the reference to " appropriate design speci0 cations" is unacceptable. SSAR Section 6 2.2, speci0cally Section 6 2.2.2.4. "S) stem Operation." needs to identify the relevant design speci0 cations which are directly veri 0ed by test. At a minimum these include, for each now phase:

a. The minimum acceptable now rate for each now phase, as measured just prior to the uncovery of each stand pipe.

b. The minimum acceptable water coverage area on the vessel side wall near the upper annulus drain elevation for each now phase, and the uniformity of the coverage around the circumference of the vessel, c, The time period for each now phase, which considers the design objectise of prnviding cooling water for a period of at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in three now phases to account for the reduction in the amount of heat to be, removed during each phase.

Response

As section 14.3 requires the AP600 design team to provide a scoping document that dennes the applicable design

requirements that must be validated, the use of " relevant" or " applicable" design specifications for a reference for more detailed design criteria for the acceptability of preoperational testing is warranted and has precedence in earlier standard subrnittals (i e. CEESSAR System 80+ and GE ABWR).

a. SSAR section 6 2.2 has been modined to include these now rates,

b. SSAR section 6 2.2 has been modined to specify the minimum acceptable coverage under the third phase of operation. .

c. SSAR section 6.2.2 has been modined to include 1he flow pro 0le for PCS operation which will provide the timing information requested.

SSAR Revison:

See Revision 13 of AP600 SSAR.

260,123 1

l i

NRC REQUEST FOR ADDITIONAL INFORMATION l l

Question 260.124 (OITS 5302)

Re. t

, , NRC Letter May 14,1997. Heat removal requires an adequate water Alm on the sestel exterior surface (suf0ciently I thick to assure stability based on the design basis accident evaluation model), as noted in SSAR Section 6.2 2 2.4 '

A non invasive method for approximating the Olm thickness during each Cow phase needs to be included in the 4

Init'al Test Program. Based on the known water delivery Cow rate and the water coserage area, near the upper ,

annulus drains, a method which measures the time for "a water particle" to travel from the vessel spring line to the upper annulus drain can be used to estimate the average Olm thickness over the covered sessel side wall.

7 4 Response:

The key parameters associated with proper PCS operation are total water now and containment surface area coverage, which are thoroughly tested for as part of the ITP. Provided that these parameters can oc veri 0ed, the

actual water (Um thickness need not be tested. .

s SSAR Revision: NONE 1

4 9

e 4

4 i

l g.

260.12 & 1

.i--.,, - _ - - - --c ,. - . - - - , , _ . . . , . - . . -

- , - - - . , - . . ,, . , -. n n,., . , - .

1 .

t .

i l NRC REQUEST FOR ADDITIONAL INFORMATION l

[T *H .

3 Queston 260.125 (OITS 5303) s Re. ;

NRC Letter May 14,1997 Under General Test Acceptance Criteria and Methods, c), references to item a, and c.

under RAI 260,123 nted to be incorporated. [

. I Response: !

See the responses to_ items 260.119 and 260,123.

SSAR Revir,en:- NONE j e * '

i I

l

4 4

4 i

1 260,125 1 Y W88tiflghouse 1 -

I

m, i NRC REQUEST FOR ADDITIONAL INFORMAitON

, Question 260.126 (OITS $304)

Re:

i NRC Letter May 14,1997. Under General Test Acceptance Criterir and Methods, c), the text refers to the PCCWST

" drain" flowpath. SSAR Section 6 2.2 refers to the PCCWST " outlet" piping or " discharge" piping. In other

, descriptions, for example the technical specidentions, references are made to the PCCWST " delivery" nowpath

. [ piping). Thre needs to be one term which is consistently used to identify the PCS piping which provides the

. cooling water to the distnbution bucket.

4

Response

l This SSAR has been revised to use the term " delivery flowpath" to refer to the PCCWST discharge piping.

j -

SSAR Revison.-

i ,

J See Revision 13 of AP600 SSAR and attached section 14.2.9.l.4, Passive Containment Cooling Water System Testing.

f i

" +

i '

260,126 1

i NRC REQUEST FOR ADDITIONAL INFORMATION

.p m=

l l

Questen 260.127 (OITS 5305) l i

Re:

I i

NRC Letter May 14, 1997 Under General Test Acceptance Criteria and Methods, d), reference to item b under i

RAl 260.123 needs to be incorporated.

Response

This comment will be incorporated, consistent with the response to RAI 260.123, SSAR Revision: NONE D

,e

I O

! +

260,127 1

0 NRC REQUEST FOR ADDITIONAL INFORMATION g

l .-

Question 260.128 (OITS 5306)

Re:

l NRC Letter May 14,1997 Under General Test Acceptance Criteria and Methods, d), in addition to senf>ing the uniformity of the wetted surface (proper operation of the water distribution busket and weirs), an estimation of the water film thkkness needs to be incorporated, as discussed in RAI 260,124. ;

Response

+

l See the response to RAI 260,124, SSAR Revision: NONE .

s 8

-f

. e r

'M 4

e 2

260.128 1 1

i f , .-.r-.- -, --------+w e.,we, , -, ,- -,w,w,,<m - - - - -. - - - - --- .- v4 w,e,m,--,-,ws..w.ye, --.#. .v,r-mw--wy

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

NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.129 (OITS $307) b NRC Letter May 14,1997. Under General Test Accepance Criteria and Methods, b), reference is made to features and equipment not identined in Section 6.2.2 of the SSAR. Dese features need to be included in the SSAR description:

a. Diverse actuation signals, those in addition to the ill 2 containment pressure signal, need to be included in SSAR Section 6.2.2.1. Attematively, the Initial Tett Program description would have to specincally address SSAR Section 7.3 to identify the appopriate features of the PCS actuation system that are covered by the testing,

b. The shield plate which protects the distribution bucket.

Respunse ..

a. Item (b) of test abstract 14,9.2.1.4 addresses the testing of system Interlocks including PMS and DAS.

Referentes to these SSAR sections will be incorporated.

b. De shield plate does not perform a passivt containment cooling system function and therefore is not discussed in Section 6.2. This plate settes as a shield for radiation from the containment and is shown in the general arrangement drawings in Section 1.2 and will be shown in Figure 3.8.4 7 in the Rev.14 issue of the SSAR.

MR Rnmun See attached Section 14.2.9.1.4, Passive Containment Cooling System Testing 9

9 4

260,129 1

' 0,

~ ~

NRC REQUEST FOR ADDITIONAL INFORMATION

. l l

Question 260.130 (OITS $308) l Re:

l NRC Letter May 14,1997. An additional test objective needs to be deseloped that will provide an estimate of the overall heat transfer process during the testing of the PCS. Consideration should be given to performing the test with

, a sufficient temperature difference between the PCCWST water temperature and the internal containment temperature to observe and measure containment cooldown. With no steam inside containment, this test will validate the overall thermal resistance of the vessel wall and its inorganic rinc coatirgs used in the design basis accident evaluation model. These data should, if practical, be obtained in cordunction with General Test Acceptance Criteria and Methods, O which provides information on the exterior boundary of the PCS (alt flow rates and temperatures).

Response

An additional test will be incorporated to require sample coupons from the containment shell to be laboratory tested to determine hs ronductivity with and without an appropriate coating of paint. .

I SSAR Revision:

See attached Section 14.2.9.1.4, Passive Containment Cooling System Testing 4 'g o

4 e

260,130 1

4 NRC REQUEST FOR ADDITIONAL,INFORMATION Question 260.131 (OIT3 5309) i Re: l l

l NRC Letter May 14. 1997 Is the preoperational test in 13.2.9.1.10 separate and distinct from the ASME Containment Structural Acceptance Test? Is it performed after the ASME Containment Structural Acceptance Test?

If this is the case, it should be clarined in Section 14.2.9.l.10.

Response

Section 14.2.9.l.10 will be clarined. The ASME Containment Structural Accegunce Test specined in Section 3.8.2.7 is a construction test and is separate from the testing specined in this section.

SSAR Revisl6n!

See attached Section 14.2.9.l.10. Containment Solauon and Leak Rase Tesdag.

e 4

100.151

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

NRC REQUEST FOR ACO(TIONAL INFORMATION p +;;

~

e l

Queston 260.132 (OITS 5310) ,,

Re; l

i l

NRC Letter May 14,1997. A requirement to verify that isolation valve divisional asognments for instrumentation l and actuation circuits are correct should be included. Also, instrumentation logic and terr.ote msaual operation capability should be verilled.

Response: .

Verificatbn that isolation valve divisional assignments for instrumentation and ac'tuation circuits are correct is included under Section 14 2.9.l.14 Class lE DC Power and Uninterruptible I ower Supply Testics SSAR Revisco:. NONE 4

260,132 1

l NRC REQUEST FOR ADDITIONAL INFORMATION Question 260.133 (OITS 5311)

Re:

NRC Letter May 14,1997. Fail open and fail close valve motions should be verified.

Response

4 This testing is performed under item (a). Inservice testing requirements include fail safe testing of safety related valves.

4 SSAR Revision: NONE 6 8 ,

9 9 4 e

e 0

O a

260.133-1

e NRC REQUEST FOR ADDITIONAL INFORMATION

-; m Queston 260,134 (OITS 5312)

Re:

NRC Letter May 14.1997 Stroke times should be verifled.

Response

This testing is performed under item (a). Inservke testing requirements include stroke testing of safety related valves. .

SSAR Revision: NONE t

i 4

4 4

e e

, 260.134-1

NAS Bit 9851135 AB0fD##41lufttEATitu I

Question 260.135 (OITS 5313)

Re:

1 NRC Letter May 14.1997 Plants have used their Type C test procedures for preoperational testing. The test abstract references ANS.$6.81994 for leakage testing methodology. The 1994 standard is permitted for Opdon B leakage testing programs to meet the requirements of Appendix J. Option A plants that want to use their Appendix J procedures and methods for preoperational testing have to use the 1972 standard.

Response

I This comment will be incorporated.

SSAR Revision:

See attached Section 14 2S.I.10. Containment isoladon and t.eak Rate Testing.

e tes.1 N (fg

1884 8t4988T P08 A38metM IntetEATitt Question 260,136 (OITS $314) l Re:

NRC Letter May 14. 1997 - The Purpose and the General Test Acceptance Criteria and Methods sections of 14.2.9.1.11 do not address the nonsafety related functions described in Section 6.2.4. Specifically, those aspects of the system that have been incorporated to meet the requirements of 10 CFR $0.34(f)(2)(la) need to be venfled by

, testing. Therefore, this test abstract needs to be modified to include testing that vertfles the operability of (1) all sisteen hydrogen sensors in their role of supponing proper actuation and operation of the hydrogen Igniters, and (2) l the' alternative power supplies to the hydrogen igniters.

Response

This comment will be incorporated.

( ..

SS AR Revision:

See attached Section 14.2.9.l.11. Containment Hydrogen Con:rol System Testing.

Hl.1864

1888 M49tST ME A80meth!lufttaATNN .

Question 260,137 (OITS $31$)

Re:

NRC Letter May 14.1997 The $$AR does not appear to describe when the hydrogen igniters are to be actuated and how they are to be operated. This information is needed to support test c) under General Test Acceptalte Cnteria and Methods.

l

Response

l This comment will be incorporated. Th9 SSAR will be revised accordingly.

SSAR Revision: .

See Revision 13 of the AP600 SSAR and attached Secdon 14.2.9.1.11. Containment Hydrogen Control Systern o

Ill.1N 1 L&'

NRC REQUEST FOR ADDITIONAL INFORMATION Enclosure 2 NSD NRC 97 5174 DCP/NRC 0907 e

e 9

w weeneous.

. i r:~

+ 14. I:itial Test Prograa 1 14.2.9.1.4 Passive Containment G .llag System Testing Purpose The purpose of the passive containment cooling system testing is to verify that the as installed components perform properly to accomplish their safety related functions to transfer heat from i inside the containment to the environment, as described in 9"em ne 7 9 e ceeMr:

! re" ; r e :rer;: ! M :!:2 pri S: : : '-5!!; p!mM rr of " . -- e r + %

l ._a 7 q . . ..'_ g .:.; g .u r _ . g g e g g :. . g .:e- ' P 2 M Section 6.2 2.

i The passive containment cooling water storage tank also provides a safety related source of i makeup waterfor the spentfuelpool, andprovides a seism.cally quallfled source of waterfor the

! I fire protection system. Testing of thesefunctions are discussed in subsections 14.2.9.1.16 Long I Term Safety System Support Testing and 14.2.9 2.8 Fire Protection System Testing.

l Prerequisites The construction testing of the passive containment cooling system is successfully completed. The preoperational testing of the Class IE de electrical power and uninterruptable power supply systems, the non Class IE electrical power supply system, the compressed and instrument air system, and o'her interfacing systems required for operation of the above systems is available as needed to support the specified testing and system condgurations. Additionally, a sumcient quantity of acceptable quality water for filling the paesive containment cooling water storage tank and draining onto the containment is available, and a means of Alling the tank is available.

General Test Acceptance Criteria and Methbds Passive conteinment cooling system performance is observed and recorded during r. series of individual component testing that characterizes passive containment cooling system operation. The following testing demonstrates that the passive containment cooling system operates as described in Section 6.2 and appropriate design speci0 cations

, a) Proper operation of safety related valves is verined by the performance of baseline in service tests as described in subsection .

3.9.6.

b) Proper calibration and operation of safety related, defense in depth, and system readiness instrwnentation, controls, actuation signals and interlocks as discussed in SSAR

sections I 7.J and 7.5 are verined. This testing includes the following:

Normal rse.go containment pressure

High range containment pressure '

Passive containment cooling water now rate

Passive containment cooling water storage tank level

Passive containment cooling water isolation valve instrumentation and controls

Diverse actuation system passive containment cooling initiation Draft Revision June 5.1997 gg,g

i

. 14. Initi:1 Test Progra e

Passive containment cooling water storage tank water temperature I

Air inlet and shield plate,20 r9 " 'f rese protection heater controls This testing includes demonstration of proper actuation of these functions from the main control room.

I c) ":: n !c ar:: ef : 9 7r ' c e : ' - "r, - - - " :;;r ten t J: 9 ae t 7:"5 2,4 I ab n e: "m;;r 19 " 'h j "eq: 9:^

7.a _ _7.nder:::ir

.cr. m %fh

, :.t . 9 69:q".M:r:o:

ge z . .t m .a c :.:g. g .g 3;cr . _. .. g. g.'g pg I 9 : rd SS 29 && ":ndire f Flow testing is performed to demonstrate proper systemflow rates by draining the passive l containment cooling system water storage tank This testing demonstrates the proper I resistance of thefour passive containment cooling water storage tank deliveryfloupaths.

I This testing also demonstrates that water is supplied at the spec $edflow rates and times I for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months consistent with the design basis analyses presented in Section 6 2.1.

d) The proper operation of the passive containment cooling water distribution bucket and weirs is serified and proper wetting of the containment is observed and recorded during

,d,raindown testing in item c, above. ,

e) The proper operation of the drains in the upper containment / shield building annulus to each drain of the containment cooling water from the annulus Door is verined.

O The resistance of the passive containment cooling air dowpath is verified by measuring the wind induced driving head developed from the air inlet plenum region of the shield building to the air exhaust at several heaJons along the Dow path and at several circumferential locations, and measurement of the induced air now velocity. Temporary instrumentation is used for this testing.

I g) Sample coupons from the containment shell with end without an appropriate coating of 1 ,

paint are laboratory tested to determine their conductivity.

l l h) The proper operation of each of the PCS water storage tank recirculatiodmakeup pumps I to perform their recirculationfunction.is ver$ed.

Draft lievision Jue 5,1997 W Westinghouse

.:m .a , _

.. F ik

14. Initial Test Prograa

~

i 1

14.2.9.1.10 Containment Isolation and Leak Rate Testing

. Purpose The purpose of the coi;t:.mment isolation and leak rate testing is to demonstrate that the as. installed containment isolation valves, piping and electrical containtrent penetrations, and hatches, and the containment vessel properly perform the following safety functions as described in Section 6.2:

Automatic isolation of the piping penetrating containtnent required to assure containment integrity

The containment vessel, penetration, and isolation valve leakage is 'essl than the design i basis leakage at or near the containment design pressure consistent with 10 CFR J0,

. l Appendix Jpressure test requirements.

I "e e-ts:-- s* 'erre! :M pr : t6 : m " ':t:M :%ntei !- -- ! p em:: cc : : :-

I !& 'O CP 50 ipp M!- 1 p ::: 2 ::* p f:--- :

Prergquisites The construction testing of the containment, containment hatches /airlocks and containment l penetrations, including A* containment pressure test as specsfied in Section 3.8.2.7 has been

I completed. @7he matruction testing of the piping and isolation valves or electrical wiring through the penetrations, has been completed. The instrumentation to be used in performing the Type A, B, we C testing is calibrated and available, including their associated data processing equipment. The required creoperational testing of the protection and safety monitoring system, plant control sysfw, %: lass IE electrical power uninterruptable power supply, and other interfacing system = wo .d Dr operation of the containment isolation devices and data collection

, is available.

General Test Acceptance Criteria and Methods Containment isolation functions, leak rate, and structural integrity performance are observed and recorded during a series of individual component and integrated system testing. The following testing demonstrates tiv the containment functions as described in Section 6.2 and the appropriate i design specifications e.re achieved. The testing is in accordance with the ? 92- Wib-M i Standeed combined license applicant's Containment System Leakage Testing' ? p ---t: -

1 Program, which meets the requirements of ANSilANS.56.81994 or ANSUANS-N4341972, as I appropriate.

, a) Proper operation of safety related containment isolation valves, listed in Table 6.2.31, is verified by the performance of baseline in-service tests as specified in subsection 3.9.6.

Draft Revision Juna 5,1997 W Westinghouse

d i~b k

', 14. Initial Test Progra a b) Proper calibration and operation of safety related containment isolation instrumentation, controls, actuation signals and interlocks is verined. This 'esting includes actuation of the containment isolation valves from the main control room, r.ad upon receipt of a containment isolation signal.

c) The appropriate Type C leakage testing is performed for each piping path penetrating the containment boundary, verifying the leakage for each containment isolation valve (listed in Table 6.2.31) or set of isolation salves. This testing for individual isolation valves may be performed in conjunction with the associated system test.

d) ne appropriate Type B leakage testing is performed for each containment penetration whose design incorporates seals, gaskets, sealants, or bellows. This testing incluoes door or hatch operating mechanisms and seals, c) A baseline in service test / inspection of the accessible interior and exterior surfaces of the containment structure and components is performed as specified in subsection 3.8.2.

I O " :t reure!:- r;9y ef'he ent9 e-* : :!ud!,;p - rie : !: med by p er:us!ng I e e-t9 e-' e e pre::ure repred by !0 CP 50 ippe91 te iett 'hei- fun: tie l , qt dergn pre- e -

I f g) A Type A integrated leak rate test is performed to verify that the actual containment leak rate does not exceed the design basis leak rate specified in the Technical Specifications.

i 19.1.11 Containment Hydrogen Control System Testing Purpose I The purpose of the containment hydrogen control system : eft 'ed testing is to verify that the I system properly performs the following safety-related and non safety defense-in depth functions described in Section 6.2:

Prevent the concentration of hydrogen in containment from reaching the Hammability limit I

Prevent the concentration of hydrogen.in containmentfrom reaching the detonation limit.

Monitor the containment hydrogen concentration as required by Regulatory Guide 1.97 Prerequisites ne construction testing of the containment hydrogen control system is completed. The Class IE de electrical power and uninterruptable power supply systems, the non-Class IE electrical supply system, and other interfacing systems required for operation of the above systems and calibrated data collection instrumentation are available as needed to support the specified testing.

Draft Revision

$ Westingflouse June 5,1997 l

.- 14. Initial Test Program I ,

General Test Acceptance Criteria and Methods f crformance of the containment hydrogen control system is observed and recorded during a series

. of individual component testing. The following testing veri 0es that the system operates as i described in Section 6.2.4 and as specified in the appropriate design speci0 cations:

i' I a) Proper operation of both the Class IE-4he safety-related and non Class IE containment hydrogen concentration instrumentation and alarms is verined.

b) The ability of the passive autocatalytic recombiners to achieve their specified plate temperature when exposed to a specified atmosphere containing hydrogen is veri 0ed by I testing a portion of the installed recombiner plates ex containment. This testing map include l certilled manufacturing tests of the plates performed in accordance with the recombiner l qualification requirements.

4 l c) The p ;: manual actuation and operation of the hydrogen ignitersconfirm that the igniters I are supplied by two power groups from two subsystems of the non class lE de and f

I uninterruptible power supply system h m 'd .

e e 9 4

4 e

.l e

Draft Revision June 5,1997 3 Westingh00S8 t

- v r

. W M EiE

. 14. I:itial Test Program =- 'E 14.2.9.1.16 Long-Term Safety Related System Support Testing Purpose The purpose of this testing is verify the capability to perform the following functions for i maintaining the extended operation of the safety related systems and components as described in Section 1.9:

Supply makeup water to the passive containment cooling system l .

gypp;i mt.up ,,....e.u : :u. g e .q----.

. Supply makeup water to the spent fuel pool I . Provide electrical power for post accident instrumentation, control room lighting and I ventilation, division B and C and l&C room ventilation, passive containment cooling system l . pumps, ancillary generator room lights, ancillary generatorfuel tank heaters, f . Supp!: & tc Provide ventilation cooling to be main control room ,

I . Provide ventilation cooling to the Class IE cabinets for post accident instrumentation Prerequisites l '

l Re ' construction tests of the safety-related systems and/or components designed for long term actions have been successfully completed. The preoperationel testing of these systems and/or components, including instrument calibrations, has been completed as required for the speciGed testing, system conGgurations, and operations. Equipment required for data collection is available and operable. Water used in this testing should be of a quality suitable for 611ing the speciGed components. Equipment used to provide the required long term actions is available.

General Test Method and Acceptance Criteria The ability to perform the required long term actions is observed and recorded during a series of individual component and integrated system testing. He following testing verines that the long-term actions can be performed as discussed in Section 1.9 and as specined in appropriate design specifications:

I O Se d!'!"; te pe"!& - eup ' 2te- : cc-t !- :- ' 9: p!;!:; cc- r"^ Advh

--Mur' H rd :;ne . r &::+d :- Sr"e- O --d :9: c"c- 5M "' WmGod.

I a b) The ability to provide makeup water to the passive containment cooling water storage tank as described in subsection 6.2.2 is veri 6ed.

I b -4) The ability to provide electrical power io the post accident monitoring instrumentation.

I control room lighting and ventilation, division B and C 1&C room ventilation, passive i containment cooling system pumps, ancillary generator room lights, ancillary generator i tank heaters, by using perm'r ;! e & 1en r the ancillary diesel generators as described in Section 8.3 is verified.

Draft Revision y Westingt100S8 June 5,1997

= ==i e -- ,

7

14. Initial Test Program d) Ne d!"*;qe pre Me ' er'e'e :^ pr--eed 2!: re e - :- --+ ! ec ! rupp!y and pre ~" 'e! n ; e 1:4;; pe- d!e :^ rpre- ed 2:- be*t!:: 1: de : : bed : c c.:r- 6 * ::

t er mi e I c e) The ability to provide main control room and 1!: e-:' cu':t in ventilation cooling using a i pe- d'e ancillary fans as described in Krubsection u 9.4./ is verified.

I d 4) The ability to provide ventilation cooling to post. accident monitoring instrumentation I equipment rooms using pens!: ancillary fans as described in Esubsection u 9.J./ is verified.

I e g) The ability to provide makeup water to the spent fuel pool via the safety related makeup connectionfrom the passive containment cooling system water storage tank. as described in subsection 9.1.3, is verified.

s Draft Revision June 5,1997 W W85tingh00S8 W

p ==

14. Initial Test Prograu 11.2.9.2.7 Spent Fuel Pool Cooling System Testing Purpose The purpose of the spent fuel pool cooling system testing is to verify that the system properly performs the following defense in depth function described in subsection 9.1.3:

Remove heat from the spent fuel stored in the spent fuel pool Prerequisites

, The construction testing of the spent fuel pool cooling system has been completed. The spent fuel pool is filled with water of acceptable quality and chemistry. The ac electrical power and distribution systems and other interfacing systems required for operation of the pumps and for data collection are available as needed to support the specified testing and system con 0gurations.

General Test Acceptance Criteria and Methods Spen,t fuel pool ec,oling system performance is observed and recorded during a series ofindividual component and integrated system testing. The following testing demonstrates that the system properly performs its defense in-depth function as described in subsection 9,l.3 and appropriate design specifications:

a) Proper operation of the spent fuel pool cooling pu.nps, valves, and strainers is verified.

b) Proper operation of the instrumentation, controls, actuation signals and interlocks is verified, including:

a Automatic pump actuation if an operating pump stops

Pump flow rate

Pump discharge pressure

. Spent fuel pool water level and control

+ Spent fuel pool water temperar m

. Water return temperature -

This testing includes operation of the system pumps from the main contro1 room.

c) ne capability of the pumps to provide the expected cooling flow rates to and from the pool is verified; with both pumps operating, with either individual pump operating, and with either heat exchanger operating.

d) In conjunction with Item c above, the pump (s) runout flow rate is verified to be properly limited, and adequate net positive suction head is verified to be available during the appropriate operating modes.

Draft Revision June 5,1997 $ Westirighouse !

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

= _

', t4. Iaisial Test Program --

e). The proper operation of the spent fuel pool siphon breakers is verified.

i

)) The proper operation of the spentfuel pool gravity drain makeupfloupathfrom the cask 1 washdoun pit is verified 4

e 4

S 6

9 e

6 4

Draft Revision June 5,1997

.4 4,.J-J_Ada_.4 ,A 434 a#4-A.hes,. sea 4s,ups_ ..pmA,,, A4 %p,M,A,4.g4.J_of4..,w,,.4,ae d%" _mh-4 4 M 4 s.h h.X4- -

- M d dE W A_ h_4 mh-6,.44ti,,,46.*am .g 1 ?,

(-

e n

4

-- NRC REQUEST FOR ADDITIONAL INFORMATION 4 SW t

4 ~ **

F 4

7 e

I i

8 Enclosure 3 i

!. - NSD-NRC-97-5174

, DCP/NRC-0907 i

c l . . .

1 1

4.

i 1

i f

I i .

k 2

e 4

i

(

4 i ,

i i

l l

t

, [ "5!

" 14. IIitial Test Program 14.2.9.1.3 Passive Core Cooling System Testing Purpose The purpose of the passive core cooling system testing is to verify that the as installed components and their associated piping and valves properly perform the following safety functions, described in Section 6.3:

Emergency core decay heat removal

+ Reactor coolant system emergency makeup and boration

Safety injection

Contaimnent pH control Prerequisites The construction testing of the passive core cooling system, or of a specific portion of the system to be tested, is successfully completed. The preoperational testing of the reactor coolant system, normal residual heat removal system, chemical and volume control system, the refueling cavity, the C, lass IE de and uninterruptable power supply, the ac electrical power and distribution- ,

systems, and other interfacing systems required for operation of the above systems is completed as needed to support the specified testing and system con 6gurations. A source of water, of a quality acceptable for filling the passive core cooling system components and the reactor coolant system, is available.

General Test Method and Acceptance Criteria The performance of the passive core cooling system is observed and recorded during a series of individual component testing and testing with the reactor coolant system, ne following testing demonstrates that the passive core cooling system operates as described in Section 6.3 and appropriate design specifications.

a) Proper operation of safety related valves is verified by the performance of baseline in service tests as described in subsection 3.9.6. Also, the proper operation of non-safety-related valves is verified including manual valve locking devices. This testing does not include actuation of the squib valves, which is discussed in item t, below.

b) Proper calibration and operation of safety related instrumentation, controls, actuation signals, I and safety related interlocks ar specified in section 7.6 is verified. This testing includes the following:

. Passive residual heat removal heat exchanger flow ,

. Core makeup tank level

= In containment refueling water storage tank level

. Containment floodup level Draft Revision June 5,1997 3 W85tlfigt10US8

'i 14, Ixitial Test Program

. Core rnakeup tank inlet / outlet valve controls Passive residual heat removal heat exchanger inlet / outlet valve controls

. In containment refueling water storage tank outlet valve controls Containment recirculation valve controls

. Automatic depressurization valve controls l . In containment refueling water storage tank gutter isolation valve controls 1

- This testing includes demonstration of proper actuation of safety related functions from the main control room.

l' c) Proper calibration and operation of instrumentation, controls, and interlocks required to demonstrate readiness of a safety related component is verified. This testing includes the following:

. Accumulator pressure and level and alarms 1 . Passive residual heat removal heat exchanger temperatures

. Passive residual heat removal heat exchanger high point vent level

. Core makeup tank inlet line temperatures

. Core makeup tank inlet line high point levels

,. . Direct vessel injection line temperatures j l . In containment refueling water storage tank level and temperatures d) Proper calibration and operation of temporary instrumentation and data recording devices used in this testing is verified. This testing includes the following:

CMT level '

CMT Dow and balance line temperatures

PRHR supply line temperatures

- Accumulator wide range level

In-containment refueling water storage tank and sump-recirculation How

. ADS piping differential pressure The passive core cooling system emergency core decay heat removal function is verified by the following testing of the passive residual heat removal heat exchanger.

I e) During hot functional testing of the reactor coolant system, the heat exchanger supply and i return line piping water temperatures are recorded to verify that natural circulation How initiates.

i f) The heat transfer capability of the passive residual heat removal heat exchanger is verified I by measuring natural circulation Dow' rate and the heat exchanger inlet and outlet l temperatures while the reactor coolant system is cooled to s 400'F. This testing is I performed during hot functional testing with the reactor coolant system initial temperature i 2 540*F and the reactor coolant pumps not running.

Draft Revision June 5,1997 ggg

t

, + =g

~

" ' 14. Initial Test Program I, g) The proper' operation of the passive residual heat removal heat exchanger and its heat transfer capability with forced Gow is serified by initiating and operating the heat exchanger

! with all four reactor coolant pumps running. This testing is performed during hot functional i testing with the reactor coolant system at an elevated initial temperature between 350*F, and 400*F. The heat exchanger heat transfer is determined by measuring the heat exchanger How rate and its inlet and outlet temperatures while the reactor coolant system is cooled down.

I h) The heatup characteristics of the in-containment refueling water storage tank water are verified by measuring the vertical water temperature gradient that occurs in the in containment refueling water storage tank water at the passive residual heat removal heat exchanger tube bundle and at several distances from the tube bundle, during testing in item

. e), above. Note that this verification is required only for the first plant.

The passive core cooling system emergency makeup and boration function is verified by the following testing of the core makeup tanks.

I i) The resistance of the core makeup tank cold leg balance lines is determined by filling the core makeup tanks with Dow from the cold legs. This testing is performed by filling the

. cold, depressurized reactor coolant system using a constant, measured discharge now from' .

the normal residual heat removal pumps. The reactor coolant system is maintained at a constant level above the top of the cold leg balance line(s). The normal residual heat removal sy stem flow rate and the differential pressure across the cold leg balance lines are used to determine the resistance of the balance lines.

I j) During hot functional testing of the reactor coolant system, the core makeup tank cold leg i balance line piping water temperature at various locations is recorded to verify that the l' water in this line is suf0ciently heated to initiate recirculation Oow through the CMTs.

i k) Proper operation of the core makeup tanks to perform their reactor water makeup and

boration function is verified by initiating recirculation flow through the tanks during hot functional testing with the reactor coolant system ata 530cF. This testing is initiated by simulating a safety signal which opens the tank discharge isolation valves, and stops reactor coolant pumps after the appropriate time delay. The proper tank recirculation now after the pumps have coasted down is verified. Based on the cold leg temperature, CMT discharge temperature, and temporary CMT Dow instrumentation, the net mass injection rate into the reactor is verified. Note that this verification is required only for the first plant.

The passive core cooling system safety injection function is verified by the following testing of the core makeup tanks, accumulators, in containment refueling water storage tank, containment sump, automatic depressurization, and their associated piping and valves.

L 1) Proper now resistance of each of the core makeup tank injection lines is verified by gravity i draining each tank filled with cold water through the empty direct vessel injection now path, i while measuring the CMT level (driving head) and discharge flow rate. Air enters the top Draft Revision June 5,1997 3 Westiflgt10US8

+ . -

L i*

. 14. Initial Test Program 4

of the draining tank from the reactor coolant system cold leg via the cold leg balance line.

I- If necessary, the flow limiting orifice in the core makeup tank discharge line is to be resized, and the core makeup tank retested to obtain the required line resistance.

, i m) The proper now resistance of each of the accumulator injection lines is veri 0ed by I performing a blowdown from a partially pressurized accumulator through the empty direct

! vessel injection flow path, while measuring the change in accumulator level and pressure.

If necessary, the flow orifice in the accumulator discharge line is to be resized and the 4 accumulator retested to obtain the required discharge line resistance.

. i n) Th: proper now resistance of each of the in containment refueling water storage tank

- i injection lines is verified by gravity draining water from the tank through the empty direct I vessel injection flow path, while measuring the water level (driving head) and discharge

I How rate using temporary instrumentation. If necessary, the flow orince in the ,in-4 containment refueling water storage tank injection line is resized and retested, until the 4 required line resistance is achieved, i

1 I o) The Dow resistance of each of the flow paths from the in containment refueling water 4

storage tank to each containtnent sump, and from each containment sump to the reactor is

. v.eri0ed by a series of tests. These tests gravity drain water from the in containment- ,

- I refueling water storage tank to the containment sump, and from the sump through the empty I direct vessel injection now path, while measuring the storage tank water level (driving head)

I and injection How rate using temporary instrumentation. This testing is performed using temporary piping to prevent Gooding of the containment. A spool piece with prototypical resistance may be used to simulate the squib valves in the now paths tested.

I p) The resistance of each eutomatic depressurization stage 1,2, and 3 Dowpath and flowpath combination is verified by pumping cold water from the in containment refueling water storage tank into the cold, depressurized, water Olled reactor coolant system; and back to the in containment refueling water storage tank using the normal residual heat removal i pump (s). The resistances are determined by measuring the residual heat removal pump flow I rate and the pressure drop across the Dow paths tested using temporary instrumentation, i q) The resistance of each automatic depressurization stage 4 Howpath and their flowpath combinations is verified by pumping cold water from the in containment refueling water storage tank into the cold, depressurized, water filled reactor coolant system using the I normal residual heat removal pump (s). The resistances are determined by measuring the i residual heat removal pump flow rate and the pressure drop across the flow paths tested i using temporary instrumentation. The automatic depressurization stage 4 squib valves are not required to be included in this test.

I r) The proper operation of the vacuum breakers in the automatic depressurization discharge lines is verified.

Draft Revision June 5,1997 9)t/ ggg

_ . _ _ . ~ -

t

14. Initial Test Program i s)' During hot functional testing of the reactor coolant system, proper operation of automatic -

depressurization is veri 6ed by blowing down the reactor coolant system. This testing veriGes proper operation of the stage 1,2, and 3 components including the ability of the spargers to limit loads imposed on the in containment refueling water storage tank by the blowdow n. Proper operation of the stage 1, 2 and 3 salves is demonstrated during blowdown conditions. Note that this verification is required only for the first plant, t) The proper operation of at least one of each squib valve size and type including a i containment recirculation, in containment refueling water storage tank injection, and a stage 4 automatic depressurization squib valve is demonstrated. The squib valve performance and the Dow resistance of the actuated squib valves is compared to the squib valve qualitication testing results.

u) The proper operation of the containment sump instrumentation is demonstrated by simulating the containment Good-up water levels.

v) The proper operation of the CMT level instrumentation is demonstrated during the draindown testing of the CMTs, speciGed in item 1) above.

I w) . In conjunction with the verspcation of the core makeup tanks to perform their reactor water' .

I makeupfunction and borationfunction described m item k) above, the proper operation of I the core makeup tanks to transition from their recirculation mode of operation to their l draindown mode of operation after heatup will be verified. This testing will also versfy the l proper operation of the core mak up tank levelinstrumentation to operate during draining l of the heated tankpuid. The in containment refueling water storage tank initial level is i reduced to at least 3ft, below the spillway level as a prerequisite condition for inis testing

'\ in order to provide suficient ullage to accept the mass dischargedfrom the reactor coolant i system via the automatic depressuri:ation stage 1, I

! The recirculation operation in item k) above, should be continued until the core makeup i tankpuid has been heated to 2 35TF. The core makeup tank isolation valves are then

! closed, the reactor coolant pumps are started, and the reactor coolant system is reheated 1 up to hotfunctional testing conditiort This testing is initiated by shutting of the reactor .

' coolant pumps, opening the core makeup tank isolation valves, and by opening one of the automatic depressuri:ation stage Ipow patlu to the in containment refueling water storage tank This will initiate a large loss of mass from the reactor coolant system, depressuri:ation of the reactor coolant system to the bulkpuid saturation pressure, and additional recirculaiion through the core makeup tank Core makeup tank draindown initiates in response to the continued depressuri:ation and mass loss from the reactor s coolant system. The automatic depressurization stage I pow path is closed after the core I makeup tank level has decreased below the level at which stage 4 actuation occurs. Note

. I that this versficbtion is required onlyfor theprst plant.

Draft Revision June 5,1997 3 Westingh00S8

t

, 14, IIitial Test Progran 14.2.9.2.4- Normal Residual Heat Removal System Testing Purpose The purpose of the normal residual heat removal system testing is to verify that the as installed components and associated piping, valves, and instrumentation properly perform the following defense in depth functions, as discussed in Section 5.4:

Remove reactor core decay heat and cool the reactor coolant system during shutdown operations at low pressure and temperature Remove reactor core decay heat from the reactor coolant system during reduced reactor coolant inventory operations in Modes 5 and 6 Following actuation of the automatic depressurization system, provide makeup to the reactor

. coolant system at low pressure l

e

Circulate and cool water from the containment after draindown of the in containment water

, storage tank ,

( .

Provic'e low temperature overpressure protection for the reactor coolant system l .

Remove reactor core dreay heat and cool the spentfuel pool during refueling operations l when the core is ofloaded):om the reactor vessel to the spentfuel pool.

Prerequisites The construction testing of the normal residual heat removal system is completed. The required preoperational testing of the in-containment refueling water storage tank, reactor coolant system, passive core cooling system, component cooling water system, service water system, a: electrical power and distribution systems, and other interfacing systems required for operation of the above systems and data collection is available as needed to support the specified testing and system configurations. The reactor coolant system and the in containment refueling water storage tank have an adequate water inventory to support testing.

General Test Acceptance Criteria and Methods Normal residual heat removal system performance is observed and recorded during a series of individual component and system testing, that characterizes system operation. He following testing verifies that the normal residual heat removal system performs its defense in depth functions as described in subsection 5.4.7 and appropriate design specifications: ,

a) Operation of valves to open, to close, or to control flow as required to perform the above defense in depth functions is verified.

Draft Revision 3 W85ti!1gh00$8 June 5,1997

4

14. Initial Test Progrars I

b) Operation of system controls, alarms, instrumentation, and interlocks associated with

performing the above defense in depth functions is verined. In addition.. the proper operation of the normal residual heat removal system / reactor coolant system isolation valve ,

interlocks specified in Section 7.6 is versfied.

c)_ The normal residual heat removal system pumps testing includes verincation that the pump flow rate corresponds to the expected system alignment, proper pump mininow operation, and verincation that adequate net positive suction head is available for the con 0gurations

tested. The following system con 0gurations are tested with each pump operating individually and with two pumps operating

.- . Recirculation from and to the reactor coolant system with the reactor coolant system at mid loop hot leg water level and atmospheric pressure

!' . Makeup to the reactor from the in containment refueling water storage tank with

approximately 4 feet of water in the tank

\ . Recirculationfrom and to the spentfuel pool with the pool at normal minimum level.

I d) , During the verincations of normal residual heat removal system now to the reactor coolant- ,

I- system, verify that the pumped now provides suf6cient back pressureio maintain a water i level in the CMT.

-I e) The capability of the normal residual heat removal heat exchangers to provide the required heat removal rate from the reactor coolant system is verined by testin6 Perfoimed with flow from and to the heated reactor coolant system, with each normal residual heat removal '

pump / heat exchanger operating individually.

.i j) ' The capability of the normal residual heat removal heat exchangers to provide the required ,

l heat removal rate from th spent fuel pool is vertfied. Since the spent fuel pool is not I heated during pre-operational testing, this verification can be made based on theflowrote

\ ,

from item c and heat removal capabilityfrom item e, above.

I g 4) Operation of the normal residual heat removal system relief valve which provides low temperature overpressure protection for the reactor coolant system is veri 0ed by the performance of baseline in service testing, as specified in subsection 3.9.6.

1- h -g) Operation of the system to facilitate draining the reactor coolant system water level to near the centerline of the hot leg for reduced inventory operations is verined. This test is performed in conjunction with the chemical and volume control system, and is used to demonstrate the performance of the reactor coolant system hot leg level instruments as discussed in subsection-14.2.9.1.1.

Draft Revision June 5,1997 3 Westingh0U$8

l l

h Westinghouse FAX COVER SHEET e ,

RECIPIENT INFORMATION SENDER INFORMATION DATE: 4 / i $ /9 7 NAME: 3,L, , guy TO: LOCATION: ENERGY CENTER -

3 . 7 Ae.k.soes EAST PHONE: FACGIMILE: PHONE: Office: /,, g 3 p, . y g g COMPANY: Facsimile: win: 284 4887 t). 6. pet c outside: (412)374 4887 LOCATION: g u u, t , p Cover + Pages 1+/

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmicalon, please call:

WIN: 284-5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

THE AT'TA c H Th A1 A An lip oF i AB 2. 2 I to C. ?-

( 2., ~3 ,$ MC W S W8 M-AAlf,Ji~5 TC> A DD/T E 6 $ NAc

&com m GNTS i10 '7"~y6 PhN6 C A 2 2- oP hHE i 2. 1997 (perswc,,) ri4s anwiEs faa sms snnw r. e waue

~6 G ?'Ao vj D8D LAT2st , Nor E T HAr- r-M E '2)AA Pr c)) A N4 FS _

IN L F TT'f f m iP/p;3 c. 08(6 p h '/ j 3, 19 9 7 p> EtR E A/ Q T"'

JNc L Jb 2'b iN R FJusca l7 oF THESSAk

&~ s 6 u as min,$

e A A, sews

7

5. Reactor Coolant System a:d Cennected Systems Table 5.2 3 ASME CODES CASES Code Case Number Title N-4 i l Special Type 403 Modified Forgings or Bars, Section 111, Division 1, Class 1 and Class CS.

N 20-3 SB 163 Nickel Chromium Iron Tubing (Alloys 600 and 690) and Nickel tron-Chromium Alloy 800 at a Specified Minimum Yield S'sength of 40.0 ksi and Cold Worked Alloy 800 at Yield Strength of 47.0 ksi,Section III, Division I, Class 1.

N-60-5 Material for Core Support Stnictures, Section W. Division I, (*)

N 71 15 Additional Material for Subsection NF, Class 1,2,3 and MC Component Supports Fabricated by Welding, Section Ill Division 1 N 122 2 Stress Ind.ces for integral Structural Attachments Section m, Division I, Class !

N 249 ll Additional Materials for Subsectica VF, Class I, 2, 3, and MC Supports Fabricated Without Welding, Section !!!, Division I (D) l N-284-[ ' ' Metal containment Shell Buckling Design Methods, Section m. Division 1 Class MC N 318-4 Procedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or I 3 Piping Section W. Division N 319-1 Alternate Procedure for Evaluation of Stresses in Butt Welding Elbows in Class 1 Piping Section W Division 1 N 391 1 Procedure for Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 1 Piping Section m, Division i N 392 2 Procedure for valuation of the Design of Hollow Circular Cross Section Welded Attachments on Class 2 and.3 Piping Section W. Division 1 (C)

N-474-2 Design Stress Intensities and Yield Streagth Values for UNS06690 With a Minimum Yield Strength of 35 ksi, Class 1 Components,_Section m, Division 1, 2142 F-Number Grouping for Ni-Cr Fe, Classification UNS N06052 Filler Metal,Section IX.

~

2143 F-Number Grouping for Ni-Cr Fe, Classification UNS W86152 Welding Electrode,Section IX.

(a) Use of this code case will meet the conditions for Code Case N 60-4 in Regulatory Guide 1.85 Revision 30.

(b) Use of this code case will meet the conditions for Code Case N-24910 in Regulatory Guide 1.85 Revision 30.

l (c) Use of this code case will meet the conditions for Code Case N 392-1 in Regulatory Guide 1.84 Revision 30. '

LO) c ft *TBA s e- A"D M Gru*0s us GD f*R f u a s.uknen o f co ~ r Au n M fnf ~du et u N '

Co N"biToon s rd A r A R E ^1o Y i n c u u b2"D r n R Eusstoa e o r 60% E CAS E

, AI - 2.8 4 ara s. PR o u n D ED , ,5 v 5 S S crio ^' y 3,8, 2 ,

Revision: 5 S February 29,1996 5.2 36 y "l;;Jugistse

4 i,

IL W

em l Westinghouse FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: ((/6/97 NAME: 1 h abaM A TO: LOCATION: ENERGY CENTER -

. A o k.s = e) EAST PHONE: FACSIMILE: PHONE:

Office: [q, g r > 4 y g COMPANY: Facsimile: win: 284 4887 t), S , A) /\ C outside: (412)374-4887 LOCATION: Qo g v, g g 1+

Cover + Pages l The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmisolon, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

1 A iv 5 TH I S M kR n o? SHoolh 4%'DAFSS YH 2- YtPsNfo

@ 01= T77w u YN- A 'T" s 6AmE oP % u8 iWG '7~ H E 1 '8 E CALL. , TLii ks E Wh vE '7~ME S TAPE B E thew ,

m _

bes M a Isyngg u.sa N T S A S "L 4,vs os w mong 6 vhv5

's

3. Design cf Structures, C:mpone ts, Eq;lpme:t, and Systems e

if the component has a distributed mass whose dynamic response will be single mode dominant, the equivalent static seismic load for the direction of excitation is dt0ned as the product of the component mass and the seismic acceleration value at the component natural frequency from the applicable Door response spectra times a factor of 1.5. A factor of less than 1.5 may be used if justified. Static factors smaller than 1.5 are not used for piping systems. A factor of 1.0 is used for structures or equipment that can be represented as unifonnly loaded cantilever, simply supponed, fixed-simply suppone3 or fixed fixed beams (References 10 and 11). If the frequency is not determined, the peak acceleration from the applicable floor response spectrum is used.

3,7.3.5.2 Multiple Mode Dominant Response nis procedure applies to piping, ic.strumentation tubing, cable trays, and HVAC that are multiple span models. The equivalent static load method of analysis can be used for design of piping systems, instrumentation and supports that have significant responses at several vibrational frequencies. In this case, a static load factor of 1.5 is applied to the peak accelerations of the applicable floor r sponse spectra. For runs with axial suppons th acceleration value of the mass of piping in its axial direction may be limited to 1.0 times its calculated spectral acceleration value, ne spectral acceleration value is based on the frequency of the piping system along the axial direction. He relative motion between suppon p>ints is also considered.

3.7.3.6 T'hree Components of Earthquake Motion Two horizonta' components and one vertical component of seismic response spectra are employed as input to a modal response spectmm analysis. He spectra are associated with the safe shutdown earthquake. In the response spectrum and equivalent static analyses, the effects of the three components of earthquake motion are combined using one of the following i methods:

The peak r~ponses due to the three earthquake components from the response spectrum analyses are combined using the square root of the sum of squares (SRSS) method.

WHO AAL R' M W T AL M ML htMW"D/ '/

( @ p A m g a t-A 1. FA 6d.u g ^> c. y' *

A MM DAN *A EA M YO 3 3 H id'i~5 g) ~

Revision: 13 3 WBSilngh00S8 3.7-24A May 30,1997

Q' Westinghouse FAX COVER SHEET e

RECIPIENT INFORMATION SENDER INFORMATION DATE: /8[/ 7 /97 NAME: 1 4,mpgg TO: LOCATION: ENERGY CENTER -

. TAcRAe d _

EAST PHONE: FACSIMILE: PHONE: Office: ([/sj y 24 yg y COMPANY: Facsimile: win: 284 4887 u$ p /:t. d outside: (412)374 4887 LOCATION: g c,e gy,2,t g.

Cww+Pgn 1+ / ,

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

QA1wwa4 >c WE A)s tu tuf iy-p ap

%E D/2miSF% PoVL YWF 2 /3 8 WNfmum CTtTh V lY W / 2- _ t 8f Isv&Lub EN

/N S SAL R E tJ. )V _

. m k a, t

N s. In i'y A E-f v AN.S

~l l 1

- i AP600 SSAR Appendix 3B Section 3B 7 Sensitivity study for the constraint effect on LBB Westinghouse performed a sensitivity study on a 6 inch diameter pipe to demonstrate that the leak before-break evaluation margins are not significt.ntly affected when constraint effects of pressure induced bending are included. The analysis used a finite element model of a 6 inch diameter pipe welded to a nozzle with a fixed end condition.

This conservatively represents the bounding conditions for AP600 piping. The normal and maximum stresses were used from a representative AP600 6 inch line bounding analysis curve. The material properties for the base metal and TIG weld were considered in the analysis. The stability analysis was performed using the J integral method. This analysis was developed in consultation with the NRC.

The concluslor. of this sensitivity study is that the leak before-break margins for 6-inch and larger piping on AP600 are not significantly affected by the constraint effect and application of leak before break to such piping is acceptable.

38.8 References t

9 e

4 9

M Westinghouse FAX COVER SHEET w

RECIPIENT INFORMATION SENDER INFORMATION DATE: 6/ /g /9 7 NAME: %, L,4gp yg TO: LOCATION: ENERGY CENTER -

1 d A cx s o r/ EAST PHONE: FACSIMILE: PHONE: Office:(4,d ;;ggg COMPANY: Facsimile: win: 284 4887 L). $ . N A C outside: (412)374 4887 LOCATION: 9 c)M Vj L L L

(

Cover + Pages 1+ Af ,

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

'!/IN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

YAS FoL L ow w s m AAw uP is pao vahsh 55 A A.

) Al 9 9 si% s g ro rnr Rea.uvsr To w as os s no os,

) Al fed M A 1*ld N A bour TWE us E oP HE A7' \/ 6 p r 6/ALovs YO m mdAY6 rP P SSOR WF/L oaFA f)LL. 'T'His tv> L.r_

B& J NSLuT> E'b ) AJ R S V . ) */ . S7 pro $ c) P oi rs e vrs - co-man w l

)

h c c.

52. V e "JtM l' u e, mws l A1 - l+> v'1A E do RLW l S.aw$ 1

e,

. :q e 5. React:r Coolant System cnd C:nnected Systems 5.4.12 Reactor Coolant System liigh Point Vents The requirements for high point vents are provided for the AP600 by the reactor vessel head vent valves and the automatic depressurization system valves. He primary function of the reactor vessel head vent is for use during plant startup to properly fill the reactor coolant system and vessel head. Both reactor vessel head vent valves and the automatic depressurization system valves may be activated and controlled from the main control room.

The AP600 does not require use of a reactor vessel heac v:nt to provide safety related core cooling following a postulated accident, The reactor vesselhead vent valves (Figure 5,4-8) can remove noncondensable gases or steam from the reactor vessel head to mitigate a possible condition of inadequate core cooling or impaired natural circulation through the steam generators resulting from the accumulation of noncondensable gases in the reactor coolant system. He design of the reactor vessel head vent system is in accordance with the requirements of 10 CFR 50.34 (f)(2)(vi).

3 Westinghouse 5.4 69

e.

.?

@i 5, Re:ctor Coolnt Syst:m cud Connected Syst:ms l The reactor vessel head vent valves can be operated from the main control room to provide l an emergency letdown path which is used to prevent pressuriier overfill folicwing long-term i loss of heat sink events. An orince is provided downstream of each set of head vent valves I to limit the emergency letdown flow rate.

I The first stage valves of the automatic depressurization system are attached to the pressurizer and provide the capability of removing noncondensable gases from the pressurizer steam space following an accident. Venting of noncendensable gases from the pressurizer steam space is not required to provide safety-related core cooling following a postulated accident. Gas accumulations are removed by remote manual operation of the first stage automatic depressurization system valves.

. The disci.arge of the automatic depressurization system valves is directed to the in containment refueling water storage tank. Subsection 5.4.6 and Section 6.3 discuss the automatic depressurization system valves and discharge system.

He passive residual heat removal heat exchanger piping and the core makeup tank inlet piping in the passive core cooling system include high point vents that provide the capability of removing noncondensable gases that could interfere with heat exchanger or core makeup t,artk operation. These gases are normally expected to accumulate when the reactor coolant system is refilled and pressurized following refueling shutdown. Any noncondensable gases that collect in these high points can be manually vented.

The discharge of the passive residual heat removal heat exchanger high point vent is directed to the in-containment refueling water storage tank. The discharge of the core makeup tank high point vent is directed to the reactor coolant drain tank. Section 6.3 discusses the passive residual heat remo"al heat exchanger and venting capability, which is part of the passive core cooling system.

5.4.12.1 Design Bases The reactor vessel head vent arrangement is designed to remove nor.condensable gases or steam from the reactor coolant system via remote manual operations from the main control room through a pair of valves. The system discharges to the in containment refueling water storage tank (IRWST). -

The reactor vessel head vent system is designed to vent a volume of hydrogen at system pressure and temperature almost equivalent to one half of the reactor coolant system volume I in one hour. In addition. the reactor vessel head vent is designed to provide an emergency I letdown path that can be used to prevent long term pressurizer overfill following loss of heat I sink events. The reactor vessel head vent is designed to limit the emergency letdown flow I rate to within the capabilities of the norme! makeup system.

The system vents the reactor vessel head by using only safety related equipment. The reactor vessel head vent system satisfies applicable requirements and industry standards, including Ju 5.4-70 3 Westlflgh00S8

5. React:r Coolut Syst:m cnd Cecnected Syst:ms ASME Code classifications, safety classifications, single failure criteria, and environmental qualification.

The piping and equipment from the vessel head vent up to and including the second isolation valvs . .e designed and fabricated according to ASME CodesSection III, Class I requirements.

The remainder of the piping and equipment are design and fabricated in accordance with ASME Code, Section lil, Class 3 requirements.

The supports and support structures conform with the applicable requirements of the ASME Code.

The Class I piping used for the reactor vessel head vent is 1 inch schedule 160. In accordance with ASME Section 111 it is analyzed following the procedures of NC-3600 for Class 2 piping.

The piping stresses meet the requirements of ASME Code, Section 111 NC 3600, with a design temperature of 650 F and a design pressure of 2485 psig.

{

j The automatic depressurization system functions as a part of the passive core cooling system.

The first stage automatic depressurization system valves are connected to the pressurizer, The valves are designed, constructed, and inspected to ASME Code Class I and seismic Category I requirements. Subsection 5.4.6 and Section 6.3 discuss the design bases for the automatic depressurization system and automatic depressurization system valves.

The primary function of the passive residual heat removal heat exchanger and core makeup tank high point vents is to prevent accumulation of noncondensable gases from the reactor coolant system that could interfere with operation of the passive core cooling system.

Section 6.3 discusses the design bases for the passive residual heat removal heat exchanger, the core makeup tanks, and their vent lin,:s.

5.4,12,2 System Description The reactor vessel head vent arrangement consists of two flow paths, each with redundant I isolation valves. Orifices are located downstream of each set of head vent isolation valves I to limit the reactor vessel head vent flow rate. Table 5.418 lists the equipment design parameters. The reactor vessel head vent arrangement is shown on the reactor coolant system piping and instrumentation diagram (Figure 5.1-5).

The head vent arrangement consists of two parallel paths of two 1-inch, open/c! se, solenoid-operated isolation valves connected to a 1 inch vent pipe located near the center of the reactor vessel head. The system design with two valves in series in each flow path minimizes the possibility of reactor coolant pressure boundary leakage. The solenoid-operated isolation valves are powered by the safety related Class IE DC and UPS system. The solenoid-operated isolation valves are fail-closed, normally closed valves. The valves are included in the valve operability program and are qualified to IEEE-323, IEEE-344, and IEEE-382.

[ Westiligh0USe 5.4 71

5. Reart:r Cool:nt Syst:m cnd Connected Syst:ms The vent system piping is supponed such that the resulting loads and stresses on the piping and on the vent connection to the vessel head are a:ceptable.

The automatic depressurization system valves are included as part of the pressurizer safety and relief valve module attached to the top of the pressurizer and are connected to the pressurizer nozzles. The automatic depressurization system includes a group of valves attached to the reactor coolant system hot leg that are not used to vent noncot.oensable gc,es. The pressurizer safety and relief valve module is supponed by an attachment to the top of the pressunzer and provides support for the automatic depressurization syvem valves. The automatic depressurization system valves are active valves required to pro ide safe shutdown or to mitigate the consequences of postulated accidents. Subsection 5.4.6 discusses the function control and power requirements for the automatic depressurization system valves.

Ju AH 5.4-> 2 w westirlatiouse

s

Enri H Novendetem 3.s . .

no girim 4m , . Westinghouse i r x girimo" AP600 Project 4 ema.noee-ononou com 0 s JUN 101997 Brian A. Mc Intyfe To: Bill Huffman (NRC) From: Eari H Novendstem l

CC: See Below Date: June 17,1997 j Re Ch.15 Discussion iterri # 39 Pages: Cover + one l

l l

l O urgent @Fornewsw O Menee comment O Messe Reply O messe Recycle i

j l

comments:

J Bill Attached are advance discussion dem responses on Chapter 15 SSAR. It is marked draft, only because they haven't been issued with a Westinghouse Cover Letter. Please give copy to Summer. ;

i i Thanks.

cc: B. McIntyre (NRC Informal Correspondence File), S. Fanto, l l E. Cariin, J. Winters, New File 7.5.1.7 (SAR), W. Carison l c (/

b l ,j . - l

.?

l 4

e 1

f i

06/17.'971003 AM 97D617 Fax on CA 15 Commeram 4

) m -

N NUCLEAR REGULATORY CO31311SSION CON 151ENTS CONCERNING WESTINGilOUSE AP600 SSAR CllAPTER 15 ACCIDENT ANALYSES Inadsertent Loading and Operation of a Fuel Assembly in an Irnproper Position (SS AR 15.4.7)

Question 39:

The staff has reviewed the consequences of the spectrum of postulated fuel loading errors. The analyses show for each case considered either the error would be detectable by the available instrumentation (and hence remediable) or the error would be undetectable, but the offsite consequences of any fuel rod failures would be e small fraction of 10 CFR 100 guidelines. A COL action item should be included in =

the SSAR to use the available in core instrumentation before the start of a fuel cycle to search for fuel loading errors.

Response to Question 39:

This item has been addressed in Chapter 14 of the SSAR (Section 14.2.10.4.2).

DRAFT G 1

l i

2 of 2 1

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

e Eart H Novendstern i

j Phone (410 374 4790 Faa. (412) 3744011

ama roe.neminou...rn - - s j JUN 5 01997 j

O' Brian A. Mc Intyre j To: Bill Huffman (NRC) From: Earl H Novendstem i

CC: See Below Date: June 18,1997 Mei LTC Meeting Ols . OITS 5135 to $137 Pages: Cover + two -

l

\

i \

.i l 0 Urgent for Review 0 Messe comment O Messe Regdy O Messe Recycle

]

i e ammentes

.I l

4 Bill . At our March 12* LTC meeting this ) ear, inere were three meeting items identified for Westinghouse to take acton. Since they are OITS items, a formalletter should not be needed I plan to enter the attached text into OITS and status them as acton NRC. (Note: I will enter OIIS # 5135 words j after we transmrt 14601, Rev 1).

I Please gNe a copy to Lambrose. Thanks.

i i

64,t'#

(

t

] -( l j ', v ,

4 4 i <

l cc: B. Moinspe 94RC Informal Correspondence Fife), S. Fanto,

R. Kemper, File 7.5.1.6, T. Schutz Otit697110 PM . 970618 Fas on LTC Ols **

'g l 1, In response to the questions posed by NRC Staff personnel at the March 12, 1997 meeting on AP600 long-term cooling in Rockville.

Maryland, the following responses are ps.ovided for Open Items.

OITS 5135: Westinghouse needs to describe the treatment of mass and l energy releases calculated by GOTHIC out of ADS-4 for LTC core boil-off calculations.

l Response The core boil-of f calculations performed as a part of the AP600 long-term cooling methodology are described in WCAP-14601, Revision 1, Section 5.

CITS 5136: Determine applicability of Regulatory Guide 1.1 for DBA LTC Chaptcr 15 calculations.

Regulator / Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps,' has been reviewed in regard to its applicability to the AP600 SSAR Chapter 15.6 long-term cooling ECCS performance analysis. As indi:ated by its title and confirmed by the introductory text, this Regulatory Guide is concerned with "the proper performance of system pumps'.

Because the AP600 has no safety-related pumps, cavitation due to inadequate NPSH is of no concern. The Regulator / Guide 1.1 stated positidn*that " systems should se designed so that adequate NPSH is '

provided to system pumps" has no direct relevance to AP600 long-term cooling design basis analyses. The nonsaf ety-related RNS pumps are designed to pump saturated water from the containment sump.

Nevertheless, the AP600 ECLJ performance during long-term cooling is somewhat af fected by the containment pressure (Ref. response to RAI 440.645). In that context, the regulatorf guide emphasis on containment pressure bears further discussion. The AP600 design depends on heat transfer through the ;ontainment shell to remove decay heat to the ultimate heat sink, the environment. For this reason, and because the loss of-containment integrity is a beyond design basis scenario, in design basis scenarios for long-term cooling the AP600 containment will always be pressuri:ed to some extent above atmospheric pressure. The Regulator / Guide 1.1 guidance to assume 'no increase in containment pressure" is not appropriate for AP600. The caution in Regulatory Guide 1.1 against relying v. an overly high containment pressure to =tchiev,e adequate ECCS performance is addressed for AP600 by applying in the SSAR Chapter 15.6.5.4C long-term cooling analyses a conservatively low containment pressura. This pressure is calculated as described in WCAP-14601, Revision 1, assuming a maximum PCS flowrate external to containment. The 'SSAR long-term cooling ECCS performance analysis is based on a conservative (low) containment pressure boundarf condition. Furthermore, beyond design basis thermal-hydraulic uncertainty analysis scenarios which assume the loss of containment integrity show adequate core cooling of the AP600 core at 14.7 psia containment pressure, as reported in WCAP-14800.

4 ve

f t

Regulatory Guide 1,1 also calls for the use of maximum expected fluid temperatures. Consistent with this statement, the SSAR LTC analyses consider the containment sump to be saturated liquid at the prevailing containment pressure for breaks in the RCS loops.

For double-ended DVI line breaks, water spills directly from the IRWST onto the containment floor, and the sump water temperatures

' are well below 200F at the time recirculation begins; they are input at conservatively high values in the WCOBRA/ TRAC SSAR LTC analyses. These temperatures are consistent with the concept of using " maximum expected temperatures" for the DEDVI breaks in the

, AP600 SSAR. Overall, appropriately conservative, not srbit ra ry, containment temperature and pressure boundary conditions are l applied in the AP600 SSAR analyses of LTC ECCS performance even i

though Regulatory Guide 1.1 is not applicable, l

OITS 5137 : Are the check valves f rom the accumulator subcompartment to the sump safety grade and operable during LTC?

Response A four-inch pipe connects the bottom of each of the two accumulator subcompartments in the AP600 containment with the I

flooded sump region below the reactor coolant loops. Its purpose is to drain the accumulator room (also known as the PXS valve room) into the primary sump following a break within the confines of that room. To prevent backflow into an accumulator room when a LOCA break occurs elsewhere in the AP600 containment, these drain lines are equipped with redundant series check valves. The check valves are safety-related and are operable during the long-term cooling period.

e 4

e

Lindgren, Donald A._ -

From: Lindgren, Donald A.

Sent: Friday, June 20,19971:14 PM To: ' Jackson, Diane' Cc: McIntyre, Dnan A: Winters, James W.; Israelson, Gordon A.

Subject:

Chapter 9 Open items I have identified a few Chapter 9 open items that can be closed based on information already in the SSAR. They are as follows.

. OITS #253 l

Action from 11/5/96 telecon between Westinghouse and NRC Plant S) stems Branch: Add text or sketch or both to l SSAR section 9.3.5 to indica'e the number of sumps and the general areas serviced by the equipment and Door i drainage subsystems. See NRC letter of 12/9/96.

Response

Figure 9.3.51, General Arrangement of Drainage System was included in SSAR Revision 13. This includes the information requested in the 11/5/96 phone call.

OITS 5358 i

The scenario is as'fo'llows: The AP600 unit undergoes a normal refueling outage where 1/3 of the core fuel assemblies are deposited into the spent fuel pool, the plant is operated at full powerjust long enough to allow the new core to become fully irradiated, then something happens to the AP600 plant that necessitates a full core offload. Should a station blackout or a seismic event then occur, the time to pool boiling could be shorter than the times presented in SSAR Table 9.14. Theretore, the staff is requesting Westinghouse to evaluate this bounding scenario and document in SSAR Table 9.14, the time to spent fuel pool saturation and the height of water above the fuel at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Response

The information requested is included in SSAR Table 9.14, Revision i1 OITS 5359 in SSAR Section 9.1.3.5, the staff requests that:

a. Westinghouse clearly describe in the SSAR how the gravity drain process works following a loss of spent fuel pool cochng given a total core ofnoad. This description needs to include when the fuel transfer tube isolation valve and the fuel transfer gate are opened and how the status of these components will be controlled following a total core of00ad (i.e. Technical Specifications). .

b. Westinghouse clarify in the SSAR that the spent fuel pool level instrumentation is safety related.

Response

a. The design for spent fuel pool dling has been changed so that water inside the containment does not have to be relico on to provide a safety related water supply. See note 2 in Table 9.14. The staff question is obviated with the design change, b, The description of the spent fuel instrumentation requirements in 9.l.3.7 includes the requirement that it is safety related.

I will make the W status of these Action N.

Page1 i

se o Ewt H Novendown

% ai,or w oo r u ais o n e ti Westinnhouse AP600 Project tw-e vva =m JUN24'i337 Bnan A. Mc Intyf e l

To: Bill Huffman (NRC) From: EariH Novendstem CC: See Below Dete June 19,1997 no OITS henis Pages: Cover Page + Five O uty.nt For news.w OM comment O Menee n.Wy OM n cycl.

Comme $ts: Bill, Westinghouse has just issued Rev 13 of the SSAR. As a result, I mil change status of all these items to Action N. AdditionaHy, we mil be shottty issuing a number of final repons. Once they are issued, I will change the OITS to status NRC. I mit use the Italcized words. Comments?

cc: B. McIntyre (Informal non-proprietary NRC file), S. Fanto, E. Carlin, R. Kemper Ofi1997 229 PM 97t419 Fan on CIT 3 Ota doc

.. o Juro 19,1997 LOFTRAN DSER 0121.6.1.7 2 (OfTS 3134)

Westinghouse needs to identWy the information provided in RM response that will be incorporated into the LOFTRAN final vertfication and validation (V&V) document (WCAP 14307)

, or the code applicabilNy document (WCAP 14234).

This information has been incorporated in the latest revtscns to these reports, whch were issued to the NRCin June 1997.

RAl: Meeting Open item (ORTS 5%

i 3) loclude ccpies of approp6*4 W4 vui the 'beckground" summary in the final version of l the LOFTRAN V&V RepoM (WhW)FTRAN CAD.

This informaton has been incorporated in the latest revtsions to these reports, whch were issued to the NRCin June 1997, NOTRUMP DSER 21.6.2.21 (OfTS 3140) ,

Westinghouse needs to identify which information from the NOTRUMP related RM response

- will be formally incorporated into NOTRUMP related documentation such as the final verification and validation report, the code applicability document (WCAP 14206), or the SSAR.

This informaton has been incorporatedin Rev. 2 of the Validabon Report, which was issued to the NRC on June 1997.

21.6.2.2-2 (OfTS 3141)

Westinghouse needs to submit the final vertfleetion and validation report.

Rev. 2 of the Validaten Report has been issued to the NRC in June 1997 RAl: Meehng Open item 2(OfTS $147)

1) include copies of appropriate RAls and the " background" sumerwy in the final version of the NOTRUMP V&V Report (to be issued subsequent to the ACRS meeting).

This informaton has been thcorporated in Rev 2 of the Valedaten Report, whch was issued to the NRC in June 1997. It will be used as the techncalbasis for the ACRS meeting on NOTRUMP.

4 RAl: Meeting Open item (OfTS 5144)

Page 2

Juro 19, t 997

2) Incorporate RAl 440.440 and 440.441 Into the text of the final V&V report (to be leeued out::yd to the ACMS meeting).

This information has been incorporatedin Rev2 of the Valdation Report n+1ch was issued to the NRC in June 1997.

CHAPTER 15 SSAR DSER-Ol Section 15.21 (OITS 1200)

Westinghouse should re submit the Chapter 15 accident anstysle when its analyale codes have 1 been vertfled and validated.

Submntalof Rev.13 of the S4R bcluded an updated Chapter 15, using final validated codes. This was submnted bJune 1997.

l APRIL 19,1996 (HSif) DISCUSSION ITEMS (OITS 2264) l

8. Sete Shutdown (SSAR Section 1.9):

e. The July 22,1994 response to Q440.106 describes the ceiculation of a single pesolve break that occure in the long term phase, and demonstmtse that there was sufficient water inventory in the containment to support adequate containment tecirculation.

Provide more ceiculationaldetail.

This was iricludedin submMtalof the Ch.15 of the SS4R (Rev.13), transnutted to the NRC in June -

1997.

DSER 0121.6.1.61 (OITS 3132)

Westinghouse needs to doeortbo, in Chapter 15 of the SSAR, the PRHR heet trenefer option it has selected for each analysie in which LOFTRAN le applied and explain why the option la conenvative for that appinostion.

This was included in submittalof the Rev.1 of the LOFTRAN CAD (WCAP 14234), transnytted to the NRCin June 1997.

4 e Page 3

.. e o

June 19,1997 DSER Of 21.6.4 2 (OITS 3172)

Wootinghouse needs to justify the failure of one ADS 4 valve as the worst case single failure for an LTC anatyone of a double ended cold leg gulliotine (DECLG) break This was included in submittalof the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER Ol 21.6.4 3 (OITS 3173)

Westinghouse needs to perform analyste of containment and vessel pressure versus time in the evolution of the LTC soonarlo to justify pressure values for an LTC analysis of a DECLG l

break This was included in submittalof the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER Ol 21.6.4 4 (OfTS 3174)

Westinghouse needs to provide a detailed account of the pressare variations in the vessel and the pressure losses due to the flow in the vessel to support the LTC analysis of a DECLG break This was included % submittalof the Ch.15 of the SSAR (Rev.13), transmitted to the NRCin June 1997.

DSER Ol 21.6.4-6 (OfTS 3175)

Wootinghouse needs to addrese a discrepency between the total in)oction flow and total outflow from the vessel in ite LTC analysis of a DECLG broek This was included in submittal of the Ch.15 of the SSAR (Rev.13), transmitted to the NRCin June 1997.

DSER 0121.6.4 6 (OfTS 3176)

Weetinghouse noods to clarify the void fraction predictions used in its LTC analyste of a DECLG break This was included in submittalof the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER of 21.6.4 7 (OfTS 31T7)

Westinghouse must address the peak cladding temperature and pressure distributions if the water in the sump le estureted during LTC and the void fraction was higher in its LTC anatysle of a DECLG break _

This was onckmindin submnalof the Ch.15 of the SSAR (Rev.13), transnytted to the NRCin June 1997.

O e Page 4

Juno 19,1997 DSER 0121.4.4 4 (OfTS 3178)

Westinghouse needs to choose additional windows for the DECLG-broek LTC analyene to demonstrate the effectiveness of the pesolve cooling system.

This was includedin submittalof the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER 0121.6.4 9 (OITS 3179)

Weetinghouse needs to address boron precipitation in its Chapter 15 LTC SSAR analyels.

This was includedin submittal of the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June

. 1997.

l DSER 0121.6.410(OITS 3100)

Wootinghouse needs to address the effect of water holdup or diversion in containment, water lose through conteinment leakage, or the nood to replenish any water lost form containment because of leakage to support the LTC analysis.

This was includedin submittal of the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER 0121.6.411 (OITS 3181) _

Westinghouse needs to address the adequacy of the windows selected as toprosentative of the LTC event. ,

This was included in WCAP 14776, Rev. 2, transmitted to the NRC in June 1997.

DSER Ol 21.6.412(OITS 3182)

Westinghouse needs to identify how the portion of a L8LOCA LTC tronalent between the end of an L8LOCA anetyens and the stort of an LTC analyens will be treated.

This was included in submittal of the Ch.15 of the SSAR (Rev.13), transmitted to the NRC in June 1997.

DSER Ol 21.6.414 (OITS 3184)

Wootinghouse needs to integrets the containment response into the AP600 elmulated transient ,

calculations.

This was r\cluded in WCAP 14601, Rev.1, transmkted to the NRC in June 1997.

e Page 5

b Westinghouse FAX COVER SHEET e 1 i

RECIPIENT INFORMATION SENDER INFORMATION OATE: (, - a f c37 NAME: d,.,d, M g TO: LOCATION: ENERG/CENTh .

Win MvNme EAST PHONE: FACSIMILE: PHONE: Office:

COMPANY: Facsimile: win: 284 4887 U$ taRc outside: (412)374 4887 LOCATION:

Cover + Pages 1+a53 ,

The following pages are being sent from the Westinghouse Energy Center, East Tower, ,

Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

T3M - Here m 4 Ai m mua 6 ~ ,mrid m n e M ~c r G 4 tJt d b 4 m u y / w o r a u s e b c A ,n m. h RMs. TI%.a c42w a p-h 6 A u t w;,. A\ \ d ~ k ne w

( D c rts M @n h 5 (t,\v . ow OT*T S h th w o,,4 b bt b Al

>& ( tD. des A\ w ki n,(- d r<. n.d deu Al '911ht hit pu b m ? 76 nct ,p\m9 caQ1 %

a s w3a d L.,wng +b. ioh5/16 kLt<

W a< 'i t ut un '

b kho h L AL cLcm 0

Adt h% ruo G- as o s. % %'s

' ' " " ' C; 4 '

O 6 w a.a u. a. s 1 sO M tud s e n ut ,n d, s nSo n9b.

s Resp (m6es to ressaining actions related to 4 MAAP4/NOTRUMP Benchmarking RAls A telecon was held between Westinghouse and the NRC on October 15,1996 to address questmns the stafI had on responses to thermal / hydraulic uncertainty RAls (primarily related to the MAAP4/NOTRUMP benchmarLing eifort).

The staffs summary of the telecon and remaining actions on four RAls is provided in an NRC letter dated October l 28,1996. Below are the Westinghouse responses to the remaining actions from that telecon.

Remalaina Action of RAI 492.21 (OITS #3381)

The NRC's question is why the core heat transfer is not ranked "high" in the PRA PIRT7

Response

The core heat transfer is not of high importance in the PHA PIRT because Westinghouse did not benchmark MAAP4's capability to predict the core temperature. MAAP4 was used to determine trends. Core lesel and duration of uncovery from M AAP4 are believed to be enough to assess that core cooling occurs. Westinghouse has used the high importance items in the PRA P!RTs to defmc the parameters for MAAP4 versus NOTRUMP comparison. l Westinghouse is not confirming the validity of the MAAP4 PCT prediction in the MAAP4/NOTRUMP benchmarking l effort. The NRC is requested to review WCAP 14869 and WCAP 14800 to determine if these topical reports address the staffs concerns.

OITS Status:

Westinghouse.. Changing the Westinghouse status column of this item to " Action N." .

I NRC Please confirm that the NRC status column can also be changed to " Action N."

Remainina Action of RAI 492.24 (OITS #3384)

De discussion during the 10/15/96 telecon between Westinghouse and the NRC concluded with a the action of needing further discussions on how high importance items in the PRA PIRT will be substantiated as bounded.

Response

Although there have been no further discussions on this specific issue between the staff and Westinghouse, there is a discussion in Section 9 cf WCAP.14869, MAAP4/NOTRUMP Benchmarking to Support the Use of MA AP4 for PRA Success Criteria Analysis, that does provide results of parameters for each of the high importance parameters listed on the PRA PIRT, it is requested that the staff review Section 9 of WCAP 14869 to determined if the information provided addresses their concerns. .

. OITS Status:

Westinghouse Changing the Westinghouse status Olumn of this item to " Action N.*

NRC Please confirm that the NRC status column can also be changed to " Action N."

4 I

t hemainina Action of RAI 492 25 (OITS a3385)

The staffs question, as discuued during the 10/15/96 telecon, centered on the interpretation of trends in the compariwn of h1 AAP4 and NOTRUht?. The h1 AAP4/NOTRUh1P benchmarLing ef fort w as not y et completed w hen the telecon was held, thus it v as agreed during the telecen to wait to see the results of the benchmarking to see if we agree on the interpretation of trends, and whether differences are significant.

Response

The staffs question pertained to how Westinghouse would judge the trends of the comparison between the h1AAP4 and NOTRUMP results. The results and trends are provided in WCAP.14869, MAAP4/NOTRUMP Benchtnarking to Support the Use of MAAP4 for PRA Success Criteria Analysis.

OITS Status:

WestinEh ouse Changing the Westinghouse status column of this item to

Action N."

NRC , Please confirm that the NRC status column can also be changed to " Action N."

Remainina Action of RAI 492.30 (OITS #3390)

Based on the 10/15/96 telecon between Westinghouse and the NRC, Westinghouse agreed to address the difference between smal[ break LOCAs and SGTRs in terms of the potential for containment bypass in the T/il uncertainty documentation.

Response

SGTRs sersus small LOCAs were discussed in Section 3 of WCAP 14869 MAAP4/NOTRUMP Benchmarking to Support the Use of MAAP4 for PRA Success Criteria Analysis, in additmn, SGTR bypass is addressed in WCAP-14800, AP600 PRA Thermal /llydraulic Uncertainty Esaluation for Passise System Reliability, by considering the potential impact of each success sequence on the large release frequency. This assessment of the SGTR success sequences impact on the large release frequency was also provided to the NRC in the draft T/ll uncertainty report submitted to the NRC on January 2,1997 (DCP/NRC0695).

OITS Status:

Westinghouse Changing the Westinghouse status ce"umn of this item to " Action N."

NRC Please confirm that the NRC status column can also be changed to " Action N."

2

Linderen, Donald A.

From: Lindgren. Donald A.

l Sent: Monday, June 30,1997 2:54 PM

, To: ' Jackson, Diane' Cet _

McIntyre, Brian A; Winters, James W.; Orr, Richard S.

Subject:

Structural Open items Update Based on changes in SSAR Revisions 13 and 14 I have updated the Westinghouse Status for the following open items from Confirm W to Action N. Please call if you have questions.

Don Lindgren .

3.7.2.8 6 (OITS #663)

Changes to address this issue are included in 3.2.1 SSAR Rev.13 3.7.2.121 (blTS #668)

_.Rasponse provided in letter NSD NRC 97 5105, dated 5/2/97. SSAR Changes are included in SSAR Rev.13, .

subsection 3.7.2.1.1 3.8.4.4 3 (OITS #751)

Letter DCP/NRCb894 Dated 6/2/97 provided response and draft SSAR revision. SSAR changes were included in SSAR Rev.13 3.8.5 9 (OITS #767)

Letter DCP/NRC0894 Dated June 2,1997 provides the response for this issue. Changes to SSAR subsection 3.7.2.1.1, 3.8.4.1.1 and Table 3.7.2 16 were included in SSAR Rev.13 RAI 220.106 (OITS #5242)

Letter DCP/NRC0908, dated June 11,1997 provided a response. SSAR Revision 14 added information about the fire water storage tank to subsection 3.8.4.1.1 of the SSAR RAI 220.107 (OITS #5243)

_ Letter DCP/NRC0908, dated June i1,1997 provided a response. SSAR Revision 14 added a paragraph to subsection 3.8.4.4.1 to provide information about the tank liner RAI 220.108 (OITS #5244)

Letter DCP/NRC0908, dated June 11,1997 provided a response. SSAR Revision 14 revises subsection 3.8.4.4.1 to provide additional information about the model.

Page 1

~ ^

AP600 Open Items TraddagSyssene Databesc Executive Semissary pese g3em Selecties: [isem no] between 5248 And 5248 Sorted by leem 0 l'em DSER Secean Tale /Drscngnson R"p (W) NRC No Branch ' Quemme Type Detal Semens Engsmeer Suses sem** tseer Ne # Duse 5248 NRit/T58 16 i TEIA)I TECH 5FEC Acenon W Actasm N II~eelecon$nh A Eactor Syssems BU('SUBb as held Apnt 30.1997, so discuss WM M se'5RXB Td5pec 1 lisose ther ese NRC acasons: .

l The seaposses so quessmas1. 2. 3. 4. 5. 8. t 0, t 1. I s. I9.20 are ok (for some. Wesangbew meeds se fan the spec e :arter).

i Three of shrse are NRC acnon.

/'

I) To cbc 6 (3 2 I A $ wner Som es seccerum accepentmIwy.

'~

jl } i k (\ 2) To conrwm cknesse of I4 (3 413 and 3 414L NRC has acnce ao sev ee the SD Evalemeen Repaws

3) Te che 24 (3 414L NRC =sil se<vahnme to endensand Wesnaghou. pmenee ihm STS tess Ier PtVs as tona.ed Tk eehers are We - f attaca, big enes fant:

qA' .

l 5 UIJ p h $4) To ckne 7. 8 3.16.17 (and all other T5 applicable um teode 4 bue endaag in neede 4L Wessagh ee ..it devek.p sensormenom semes and scyisir the new acace so "smessese assion to sesease._.* =uh an actice "se sesaove .? t lTimsas BIG.

5) To clame 21.Wesneshaine needs so act no sesolve clu equsb valveoperabehry issues Tk desemasne expends also seclak =alves Th as BIG A samuff part of 21 shes I dade's omsd te get lose es to secogsmac she ADS saages are one equnalene and tweak eus the anoines,cuesmaene each sw M[ . appsomch developed Apnl 9 & 10.

I I '4) To close 82 (TS 3 4 IOL Westinghoone will pnswete she TS smarkup esh Rev I of the sespamw so RAI 4 to 17. I'm um suse thus win chne se a e

% \ast0% K J ,

b .

(,-)#~~c"~'~~

To clow 9 (3 4 3) wesanghaine =su deserswee an appsepnase peessure and faz a spee markup em NRC for mar i branch ve=se.

59699)Wd -

!.> To cio.c > $ <3 S ix em .homea.d d.c-ss annerverve than 1.00 3 0.3

.e. a s em ,.ess e i eeds Cr<Ses. . mci deres..f Act.o. o as oTra8MW . 3A> # !,,1.

.se m>.2,..es mm , . .a , ,,

. .oe - mNR--.h_.--d_

~ .o .,, . .- - -

-.c f'[ j p(? h & W , h ,~~ 23 az' ~" .'-.- - ~ ~~a e -- d'~ - v'- ~ -.- ~ is 1s^= > >

css "

, 4 1

c,#f 9 C,W d t m -c 7,,,,,, c - '_- " 'r- *=M__ _ _ .

5'"' "*- - .. i 1 ( ,, .2. Actsom NRC

{ }M/ -

( , 3. Actson NRC(telecem set for July 1) h 4. Closed wah June t8 Tech Spec snartup setzesaal

5. Act.o..: u .ai.e ra , .e aAires.si a,e. r5T. a c m e 1I l6.

t Closed TS auwks.p provieed wwh RAI 41017 Rev 1.

\ )

h 1 j7. Closed Piessure knas a < 500 pus j8. Closed Delened narogen psessure condaios. g

9. Dame by faz em Apnl 30 d/ io . Cimed_ p 3 9_I63_famames a SSAR Rev 13 are enesuaens wah Tech specs __ _ . _ _ .

t _

B Page: 1 Total Records: I

FAX COVER SHEET h) c Westinghouse RECIPIENT INFORMATION SENDER INFORMATION <

l DATE: h/141/y'/ NAME: 7/g; ly[ftpfn TO: LOCATION: ENERGY CENTER .

t hhdAV '

EAST PHONE: FACSIMILE: PHONE: Omce:

COMPANY: Facsimile: win: 284 4887 Mb outside: (412)374 4887 LOCATION: gfjpI.qf6 1 g g Cover + Pages 1+ l The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

w 9

y

, t' e

i O

h I %

S I R

~ %

E R

$ $ ! E

N O 4 $ 4 E $ $ S

I R *

R w l R $ $ E w N

l .

O N 5,0i

$ @S O O N O O O n O V

~

c%O O O

% %- % % % % % % % i% % -% % % % % % % % % %

1O y . A% R # $ E E E w 3 0

R A *

@ $ H $ 2 a 2 O O S M C 2 3

O 3 O O E n O c $ 3 $ O O O 3 O O

O O O O O O O O O O O O O N O O O O O !

s

-l I l l I I O'e 5 m}NIef@! U w U O $ $IC I*IN @ p IT * $

I

<O e Q

O *

D O M O O N O O @ Q O O 4 O N & w O # 4 $ O e O w Q @ $ O 9 y . .- -

ClO D.O O e jw im OjN

O $}* O !$]N !h M O 'O gr

i

i i ;

i

% g O!O O Ofo

t O O O O O O i I j O O O O - 0 1 0~OjolC O Oj '

Ih I w' s! '

{ % % % . % % % % % ' % % ' % % !s' % r s' % % - w N '%

> O O

O O O O O O O O O O O O O O O O O O O li,. I4

, ' E O O N O O O O O O O O O O O O O w O O O O O *

{%

% %- % t% % % % !% j % - % % % % % . w'

( Y O.O N OlOje Ojo O O O CICIO O O N OiOlC O O l 1 l l j i i e

f i : .

, O OI;lO, -_lw O O!O!OlO O;O o O O e 0: O'O O C e % % .% %

%- % %. wi % %

4

% : -w % % % % % % %+ % - %

e VO O N O w S N

O O O O O O O N M N

O O O O N N *

w.n N O w $ ON *-O -

e e O e 4 1 0 O aw % % , w % % ,: % % % % % % , % % % % w.s %

O N N * * . n .< O w ;ON

O e e i e e , O .O 1

y i Q Qi e swi = i

W = r ri e,rc ee..s, i . i n.

isirigi r ,

gi>5.,gigi 8

z E Z-I.

Z dZ-Z

- - Y. >i W i$ E = 3 ZZ Z: Z-ZiZ'Z l , =l d.

Z Z 3- 2. - 2.Z Z: . E.

Z,Z

6. E. . . -

Z- Z: Z Q

l, 71-terjsames W.; lj t3# AM W30N7, Wednesday 1 minder of items Ne 1 ]

Frorn: *Wmters, James W " <w interjw@ westinghouse.com>

To: "' des t @nre. gov:" <desi @nre. gov >, "'trq@nrc. gov'" <trq@nre. gov >.

"'jms3 @nre. gov'" <jms3 @ nre. gov >, "'wch@ nrc. gov'" <wch @ nre. gov >

Cc: "'dtj @ nre. gov'" <dtj @ nre. gov >, "'tjk2 @ nrc. gov'" <tjk2 @ nre. gov >.

"McIntyre, Brian A" <mcint> ba@wesmall.com>,

"Cummins. Ed"

<.cumminwe@ westinghouse.com>,

"Vijuk, Robert M,"

<vijukirm@ westinghouse.com>,

Rarig. Bruce li"

<ratigbe@ westinghouse.com>.

"Nydes, Robin K." <nydestk@ westinghouse.com>,

"Lindgren, Donald A."(lindg1da@ westinghouse.com>.

"llaag, Cynthia L."

<hangIcl@ westinghouse.com>,

"' mms @nrc. gov'" < mms @nte. gov >,

"Tupper, Robert B," <tupperrb@wesmall.com>

Subject:

Wednesday Reminder of items Needing NRC Acknowledgment Date: Mon,30 Jun 199711:38:08 0400 X Priority: 3

> The background for each of these items is outlined in the Open items

> Tracking Systent. They ALL have been addressed by Westinghouse.

> NRC is requested to please acknowledge receipt of information (lated

> to each of the following Open items. The reviewer in each case should

> have a submittal from Westinghouse as identified in OITS for the item.

> Recognizing that reviewing for completeness of the response in each

> case constitutes an NRC action, we recommend that receipt

> acknowledgement be accompanied by direction to change their "NRC

> Status" to " Action N" " these are truly " Action W", please provide

> a description of the action Westinghouse is expected to take. We know

> of no action required, Many of these are very (over 6 months) old.

> We are not asking for resolution or even NRC review at this time,just

> acknowledgement that you have received the information as outlined in

> the OITS Status Detail. If your investigation sh,ws I goofed and

> didn't remove item numbers that I should have, please let me know.

> SCSB 972,973,984,988,1002,1007,1008,1009,1012,1458,1461,

> 1633,1638,1639,2415,2416,24I8,2485,2487,24%,2502,2503,

> 2504,2$05,2512,2514,2718,2883,2884,2885,2887,2888,2889,

> 3070,3078,3079,3124,3125,3126,3127,3128,3129,3197,3198,

> 3200, 3202, 3208, 3210, 321 1, 3215, 3419, 3422, 4162, 4299, 4300,

> 4301,4302,4303,4304,4305,4306,4307,4308,4542,4543,5153,

> $ 154, 5155, 5156, 5157, 5158, 5159, 5160, 5161, 5162, 5163, 5164,

> 5165, 5166, 5167, 5168, 5169, 5237, 5238, 5239, 5240, 5241, 5292,

> 5293, 5294, 5295, 52%, 5297, 5298, 5299, 5300, 5301, 5302, 5303,

> $ 304, 5305, 5306, 5307, 5308, 5309, 5310, 5311, 5312, 5313, 5314

> 5315,5351,5353,5354,5373,5493,5494,5495,54%,5498,5499, Printed for " Brian A. McIntyre" < mci _ntyha@wesma_il.com> __ ._ _i. ~ ~ T ilij

s' v5501,5502,5503,5508,5511,5512,5513,55 5520,5521, p 5522,5523 - 14 items have been removed nd 42 new i ms have been f29 e added since the last repon.

~

d SRXB 955,1260,1261,1400,2237,2238,2239,2258,2980,3172, p3173,3174,3175,3176,3177,3178,3179,3180,3182,3381,3384,

> 3385, 3390, 3961, 4150, 4524, 4973, 5079, 5080, 5081, 5082, 5083,

. p$084,5085,5086,5087,5088,5089,5090,!O91,5092,5093,5094,

> 5138,5139,5140,5141,5142, $143,5144, $145, $146,5170, $171, p5172,5173,$174,$175,$176,5177,$178,5179,5180,5181,$182, t5183,5184,5185.5186,$187,5188,5189,$190,$191,5192,$193, p5194,5195,5196,5197,5198,5199,$200,5201,5202,$203,$204, 1486,44& 488,5489,5506,5507,5537 - 27 items l

r 5205,5206,5485,I nd 90 new it "p

> have been remove Q u have been added since the last l > report.

> ECGB 628, 649,662,664,668,745,750,751,766,767,768,769,

> 772, I 885, 2515, 3432, 3437, 4159, 4160, 5028, 5029, 5030, 5031, 5032,

> 5033, 5059, 5060, 5061, 5062, 5063, 5064, 5065, 5066. 5067, 5068,

> $069, 5070, 5071, 5072, 5073, 5074, 5075, 5076, 5077, 5078, $ 150,

> 5242,5243,5244, $245,5246 .. 7,5288,5289,5290,5291,$$25

> - 3 items have been remove and 29 it have been added $1nce the 4 'a 6

> last report.

3 '

> SPSB 1406,1423,1425,1432,1447,1448,1450,1452,2795,2939,

> 2942, 2945, 2958, 2959, 3007, 3009, 3943, 4185. 5014, 5015, 5016,

> Si19, $120,5129, $130, $134,5497,5505 - NO items have been removed

> and 4 items have been added since the last report.

> SPLB 172,174,253,264,372,1024,1025, i142,1171,2032,2419,

> 2892,2893,2894,2895,2896,3053,3122,3482,4170 4190,4195,

> 5346, $358,5359 - NO items have been remove nd 7 it s have been p7 L added since the last report.

d TSB 1278,1279,1280,1281,1282,1283,1284,1285,1970,2074, p 3054,4189,4224,4225,4226,4227,4971,4972,5250,5526 - NO items

> have been removed and I item has been added since the last report. *f v

v HICB 5095,5096,5097,5098,5099,5100,510l,5102,'5103 - NO

> items have been removed and 1 item has been added since the last *I d report.

d No Branch identified 4145,4151,4152,4153,4154,4155,4156,4157, n4182,419& 4197 1 .8,4615,5109,5110,5111,5112 1 item has

> been deleted 12 it s have been added $1nce the last report. *#

> EMEB 608,798,800,801,805,814 - 1 items have been deleted and I o

> item has been added since the last report.

> PERB 1023,2387,3520,3521,3522,5106,5107 NO items have been

~ ~ ~ ~ ~

TPrinEd f$r "BrianLA,_Milntyre"jmcintyba@kesmailhom> __ _ _ _

_ _2j

e 6

5

> deleted and 2 items have been added since the last report. AL p PDST . 501,502. $N,505,1763.. NO items have been deleted since

> the last report.

> liliFB - 1525,3941,5247.. NO items have been deleted since the last p .eport.

p ADT . 459,1991.. NO items have been deleted since the last report.

l D liQMil $104. 5105.. I item has been deleted and 2 items have been *'

t added since the last report.

l > g-sPEPil.1019 t fp

> itVill . 5510 D '

EELil 5108

> Thanks

> Jim Winters

> 412 374 5290

{ > ..

~

Prinied for " Brian A. McIrityre" <mcintiba@wesmall.com> _. E_~~~ T3]

f

Lindgren, Donald A.___

From: Lindgren, Donald A.

Sent: Wednesday, July 02,1997 4:32 PM To: ' Jackson, Diane'; 'Kenyon, Tom' Cc: Orr, Rchard S.; Schulz, Terry L.; Winters, James W.; McIntyre, Brian A: Vock, Preston A.; Matty.

Terrence J. I 1

I have changed the following open items from the Confirm W status to the Action N status for the Westinghouse l status since the information has been included in SSAR revisions. Please let me know if they can be closed or resolved.

Don Lindgren (412)374-4856 DSER 3.6.3.41 OITS #608 Action N SSAR Rev.14 includes additionalinformation in 311.7 DSER 3.9.6.21 OITS #798 Action N SSAR Rev.13 includes changes in subsection 5.4.8.

DSER 3.9.6.2 3 OITS #800 Action N SSAR Rev.13 includes changes in subsection 5.4.8.

DSER 3.10.2 OliS #814 Action N SSAR changes are included in subsection 3.10.2.2.

DSER 5.4.12.4 1 OITS #955 Action N SSAR Rev.14 includes changes in subsection 5.4.12.

RAI 210.219 OITS #3509 Action N SSAR Table 3 2 3 includes reference to requirements in 6.2.2.2.3 for design and fabrication requirements for selected PCS components.

Meeting commitment OITS #4257 Action N Changes are included in SSAR subsection 7.1.7.

t 4

Page1

O) FAX COVER SHEET e Westinghouse RECIPIENT INFORMATION SENDER INFORMATION DATE: '7, // o /9 7 NAME: 3 A j g gfpg,y TO: LOCATION: ENERGY CENTER .

3 IAeM3 rw EAST PHONE: FACSIMILE: PHONE: Omce:fam) 3 fa .pg g, COMPANY: Facsimile: win: 284 4887 d 5, AJ /26 outside: (412)374 4887 LOCATION: R Uc.K 1/8J-A 6 l

Cover + Pages 1+ L '

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please esil:

! WlH: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

wHE Fou <>.u o ruc m A e n n - u? 'PA 6'rs tui L u R G s ot- u E OPFw 'YTat*1 )905 W 6 $$A/L c/} AA6 2 wiLL B F isi RGV 1 f , YHf oTNfe YAC f Ls A- tha n o P of weAP D o r 4'

. 0 71 r

A -

M) L u m pows Me/MA G.

We nTFAJ-60MNS

t Criterla Referenad AP600 Section Crlierta Position ClartfkatioWSumman Description of Enceptions C.I Conforms C.2 Code Case 1644 Conforms C.3 Exception Design margins of two for flat plates and three for shells are unnecessarily restrictue for normal, upset, and emergency conditions, as well as inconsistent with ASME Code requirements. For these loading conditions, the AP600 limits the allowable buckling strength to 24 of the entical buckling strength, C,4 ASME Code, Section 111. Exception his regulatory position recommends that design NF 3221.1, NF 3221.2, stress hmits be used in conjuncuon with a NF 3222, NF 3262.2,111400 loading combination that includes operating basis canhquake. D'eASME Code rules (in which Level B stress limits are typically used for the upset load combination) provide a consenative design basis. he AP600 uses the latest rules (as of 1/90) without further restriction or justification.

De operating basis earthquake has been clintinated from the AP600 design basis. ,

Refer also to the discussion on Criteria Section C.3.

C.S.a ASME Code Secuon Ill, Exception Refer to the discussion on Criteria Secuon C.3, NF 3224 C.3.b-c ASME Code, Section Ill, Conforms NF 3262.2.111400 C.6.a ASME Code,Section III, Conforms F 1323.l(a), F 1370(c)

C.6 b ASME Code, Section 111, Exception ne limit based on the test load given in this NF 3262.1 regulatory position is overly conarvative and is 1

inconsistent with ASMF C6de riquirements. ','..

I AP600 uses the provisians of the ASk,@ coa l

Section III, Appendix l .v determine 4hed I

condition allowable loads for supports designed i by the load rating method n: ?.W wee + the tema = : .!:f" by TL : O S SV .

C.6.c Conforms C.6 d ASME Code Section III, Conforms F 1324. F 1370(c)

RevWon: 15 Draft lA 58 $ Westingh0088

9 SRP Chapter 3 - DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS .g -%

.m jj Criteria Referenced AP600 Section AP600 Criteria Position Comments / Summary of Exceptions R_G.1.130 Exception "Ihe limit based on the test load given in this regulatory position is overty conservative C6A and is inconsistent with ASME Code requirements. S .*_"E - i ' ' e

--L

' -! byTL-n48 S' '% -~T*g A?CCd gjsGS TffE TRousSIoNS RG.1.130 Arrepr hfe c>s;,{ssE code, SFev-~ lli'> A??FM1><X F ro %rrgouve C6.c. C6.d. C7 go gpg gnm 4 gq g gg ,

%YSt&su fi> E Y ^'i~HL Lo K1> % ATM4 JL4 f*N lb% C5, 3 b.(1) Acceptable 3A(2) Acceptable 3A(3) Acceptable 3b.(4) SRP 3.9.2 Acceptabic 3A(5) SRP 3.9.2 Acceptable 3A(!) A w +bk l 3A(7) Acceptable -

SRP f 3.9.3, Appendix A - Stress F w. for ASME Code Class 1,2, med 3 C-

--- -- and Ceanpement O_;;: 2 er Safety-Relesed Sysessas and Class CS Cece Sepport Structures Under Speerse Service r - m; em,t- (a,,, 3, es4)

C I.I Acceptable 4

i W- wesnnehouse

-u tm .,t ==

3-60 O

g

Ecinar an ,, totis am'77Fs/e7 , tele:e _ open_ tems m21cted to 4 1 Date: Tue.-15 Jul 1997

10:25:37 -0400 To: jms30nre. gov-From Cindy Haag <haagc19wesmail.com>

Subjects Telecon Open Items Related to MAAP4 and MELCOR 1.8.4 i Calculations Cc: SCOBELJH0 westinghouse.com, RAAGCL Subjects. Telecon Open items Related to KAAP4 and MELCOR 1.8.4 Calculations

Dear Joe,

As discussed in the NRC's telecon summary _of July 9, 1997 on the subject matter, Westinghouse agreed to provide the staff with the val"e vault room elevation, the valve vault floor drain elevation, and a suggested list of parameters that should be used in the future to compare the MAAP4 and MELCOR calculations that were discussed during the June 30, 1997 telecon between Westinghouse, NRC, Sandia, and ITS. Below is the Westinghouse response to these'telecon open items..

1. -Valve vault floor elevation = 87' 6'

2. Valve vault floor drain is at the floor elevation in the valve vault and

_ empties just above the 83' elevation in the loop compartments.

-3. Suggested list of parameters for use in comparing MAAP4 and MELCOR calculations for the 3BE-FRF1 cases

a. The plotted parameters for the MELCOR runs:

Break Flowrate (water, steam) ;

CMT Injection Flowrate Accumulator Injection Flowrate CMT Level l ADS Stage 1,2,3 Flowrate (water, steam) l ADS Stage 4 Flowrate (water, steam)

RCS Collapsed Level heactor Vessel Mixture Level Core-Exit' Gas _ Temperature Hot Core Temperature Mass of Core Debris in Lower Head-

-Containment Water Level (Containment, Valve Va' alt)

b. Parameters for Chronology:

Rx Scram RCP: Trip .

CMT Activation Accumulator Actuation (P < 700 psi)

ADS Stage 1 Actuation ADS Stage 2 Actuation

ADS Stage 3 Actuation ADS Stage 4 Actuation CMT Empty-Accumulator Empty

-Core Uncovery-

Core Reflood Initiation (water injection through break)

Cavity Flooding _ Initiation (core-exit temperature > 2000 F)

Initial Debris Relocation to Lower Head

!-Printed for'Cindy saag <heagc19wesmail.cces 1 l

i O

If you have questions on the above information, please call me at (412) 374-4277.

A copy of this email will be placed in the informal correspondance file.

Regards, Cynthia Haag Advanced Plant Safety & Licensing Westinghouse Electric Corporation l l i

I l

t l ifnted for Cindy Haag <hangclDwcomail.ccm> 2 "j

(Wb Ylestinghouse -

TAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION OATE: L/ ll, /997 TO:

NAME: L Q,,,3 LOCATION:

% ,$w/04 ENERGY CENTER -

EAST PHONE: FACSIMILE: PHONE:

ome,:gjf, g ,y, g COMPAf#: ~

Facsimito: u;n; A/RC pg44g7 cutside: (49)374 4g7 LOCATION:

Cover + Pages 1+/

The followin9 Pages are btIng sent from the Westinghouse Energy Center, East Tower.

Monroeville PA. .lf any problems occur during this transmisslor, please call:

WlH: 284 5125 (Janice) or Outside; (412)374 5125.

~- - =-_ . , - _

COMMENTS:

ryc Avnwco 01Aecu A gueua Mma &cw rin L 93, kn wr Rn L.,e

f. It/3 non,n m ps e c F w an wr I.140 cou se r isain.,cr see, w gnu ,,,,,,

Dus una e.t i070 Mees r sceco n,/ Ak3 venmesu sfr>> w a spur e

gSAt. *Tbt o c ms c w<c c co ignu /2eurs,,o tr c,nt e.,) cu ,: ,/m ,m,,

ik U.

<c L.~nuo

,,, , rn ., - -

C. w m W 1

/lso v th t wiviw piv 7C M 'NU f(o%lli 0./ ^4I)

[. MVdlb b Oftf, iCO)

j i

/

9. Ausulary Splems 1

I are balanced to maintain a constant exhaust design air flow through the fans, ne exhaust fans are located in an equipment room on Elevation 100' 0* in the northwest comer of the radwaste building.

ne exhaust fans discharge to a common duct which is routed to the plant sent. A radiation monitor records activity in the discharge duct and activates an alarm in the main control room when excess activity in the effluent discharge is detected, ne radiation monitoring system is described in Section 11.5.

The exhaust air collection duct inside the radwaste building exhausts air from areas and rooms where low levels of airborne contamination may be present. Exhaust connection points are provided to allow the direct exhaust of equipment located on the mobile systems. Where potential for significant airbome release exists, mobile systems include HEPA filtration. Back draft dampers are provided at each mobile system connection to prevent blowback through the equipment in the event of exhaust system trip. C,.hm $r m4,/4 gueu.; phs m ,a c/ le,/ ,4 r.,% // 2 w //.4 9.4.8.2.2 Component Description ne radwaste building HVAC system is comprised of the following major components. Dese cotnponents are located in the non seismic radwaste building.

Supply Air Handling Units Each air handling unit consists of a plenum section, a low efficiency filter bank, a high efficiency filter bank, a hot water heating coil, a chilled water cooling coil bank, and a supply fan with automatic inlet vanes.

Supply and Exhaust Air Fans ne supply and exhaust fans are centrifugal type, single width single inlet (SWSI) or double width double inlet (DWDI), with high efficiency wheels and backward inclined blades to produce non overloadir.g horsepower characteristics. The fans are designed and rated in accordance with ANSI /AMCA 210 (Reference 4), ANSI /AMCA 211 (Reference 5), and .

ANSI /AMCA 300 (Reference 6).

Low Emelency Filters and High Emclency Filters l The low efficiency (25 percent) filters and high efficiency (80 percent) filters have a rated dust spot efficiency based on ASHRAE 52 (Reference 7). The filters meet UL 900 (Reference 8) Class I construction criteria.

Hot Water 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 Revision: 12 April 30,1997 9.4 52 T Westinghouse

^~~ ' e. '-

~~' -~ ~

u 4 3 ( , rJ I 4164 44,JJ f .t A 444 Jet JJJJ 4, w w w essessessessessessess i ese T1 REPORT ***

essessesseesseesseees TRANSMISSION UK T1/R1 No 4:27 CONNECTION TEL st3014152300 St'BADDRESS CONNECTION ID NRC ST. TIME 07/17 11:12 USAGE T 01'13 PGS. 2 H E Sl't.T OK

/ Westinghouse -

TAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION DATE: fuc/ /6, /99 7 NAME: __ L Lj,u7 pas TO: -

LOCATION:

% l{entfod ENERGY CENTER -

EAST PHONE: FACSIMILE: PHONE: omes: f/fe.374fa go COMPANY:

Facsimite: Mn: 284 46e7 A/RC outside: (412)374 4087 LOCATION:

Cover + Pages 1,/

The followl09 Pages are being sent from the Westl6ghouse Energy Center, East Tower, Monroeville, PA if any problems occur during this transmission, pleau call:

WIN: 264-5125 panico) or Outsidei (412)374 5125.

~

COMMENTS: .

-7ys mfntMoo inAdcue gueuso hereuer Q*ev iIFm E 93. Nurr neA /Eer l.aa.

I.1D pa,n m seee n,wr, I. ik ooe.v nr w w aarw na _

On eum er iaTo smes cearge M A% veanx.o str>> rosmayrw

<;-rg, T.it.5 CHM r tuta Co wTu lEsausreu /T Oatm W WrM %

1 WESTINGHOUSE ELECTRIC CORPORATION PO BOX 355 E-'

PITTS8URGH, PA 15230 0355 ~

ADVANCED PLANT SAFETY AND LICENSING FAX NO: 412 374-4487 (WIN 284)

CONMRidATON NO: 412 374 4237 DATE:

TO: ,

d #M LOCATION:

FAX NUMSER:

FROM: Briar A. McIntyre PHONE: WIN: 284 4334 BELL: 412-374 4334 NUMBER OF PAGES (INCLUDING COVER SHEET):

D COtedENTS / MESSAGE:

ZN 17a ' dar-- S w m df- M+e m Eww [ n % % ,n en a-A A AS.e f of 7 &nu+ Arm

~ '

78+.~ & >- 4 LpK ?

ua

~ ~ jy Q /Qwp o *a UNITED STATES

! j NUCLEAR RECULATORY COMMISSION - '

$ f WASHINGTON, D.C, snaam m MJ{*,*ggg7 June 9, 1997 s

c Brian A. Mc Intyre Mr. Nicholas J. Liparulo, Manager

De U Nuclear Safety and Regulatory Analysis Nuclear and Advanecd Technology Division Westinghouse Electric Corporation

. P.O. Box 355 Pittsburgh, PA 15230

SUBJECT:

REGULATORY TREATMENT OF NON-SAFETY RELATED SYSTEMS (RTNSS) FOR AP600 7*

Dear Mr. Liparulo:

y On May , 199ft Westinghouse and the Nuclear Regulatory Comission (NRC) held a meet AP600 RTNSS issues. Discussions during this meeting included alternative methods of regulatory oversight the NRC staff would consider if Westinghouse wished to take credit for non-safety related systems within the RTNSS process and thereby expedite the staff's review of issues related to the focused probabilistic risk assessment (PRA) and thermal-hydraulic uncertain-ties. It is the staff's understanding that Westinghouse may be willing to -

recommend some administrative availability controls on certain non-safety related systems which the combined license (COL) applicant would be responsi-ble to implement. The staff would find such an approach acceptable provided the following conditions were met:

Details on the controls for the particular systems, inc19 ding limiting

conditions for operations, applicability of those conditGns, required l actions, completion times, and surveillance requirements are reviewed I

and agreed to by the technical staff.

The system controls include commitment to satisfy the Maintenance Rule (550.65) for those systems (i.e. the RTNSS systems are within the secpe of the Maintenance Rule), in addition, the availability and reliability goals used in the resolution of the RTNSS issue must form the basis for the Maintenance Rule performance goals.

The above information is included in the AP600 standard safety analysis report (SSAR) and the associated design control document, together with a descrtption of the RTNSS process and how the selected equipment is credited under the RTNSS process, and that a COL commitment is included in the $$AR to implement operating procedures consistent with the SSAR RTNSS availability controls. The staff would include a provision in the AP600 design certification rule whi:h would make the RTNSS availability controls binding on an applicant or licensee that referenced the AP600 design certification.

Westinghouse is requested to identify the specific systems which it plans to place under the above RTNSS controls and provide the detailed availability controls and maintenance rules goals which will de placed in the SSAR.

W h Westinghouse FAX COVER SHEET {

ame f RECIPIENT INFORMATION SENDER INFORMATION l

OATE: 1-I'7-9') NAME: tch n N ycArs TO: LOCATION: ENERGY CENTER .

R Abic, sky' EAST PHONE: FACSIMILE: PHONE:

Office: 3~N- 4 ) 26

- COMPANY: Facsimite: wm: 284 4887

. Mdb outside: (412)374 4887 LOCATION:

Cover + Pages 1+L'_

The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

doe 'b are 0% reeth & Nec s+zcsm o

\ (1 %* 0 [fC(CAIf I~ k \//n)Y QAQ

~

kb'as-Asaats%bi1 kokrnor (I t

! M rs h n n e k rn A ) DOY d(#' h '

mmn eAP n ed tiv Sb cva repm+

%e nk vm o I

LWM

\; A -

~

s AP600 Open Iten TrackingSysten Database: Executive Susaniary page: 7/37/y7 Selectimet: [rx st codej=" Action W* And [ resp engj like *mmis** Sorted by item #

Itern DSER Section TiticAhpion Resp (W) NRC No Branch Quesaion Type Detal Status Engmeer Status SJens W No / Dase l

1525 NRR/HilfB 20 4-2 DSER4)I -

~

g TRGs Closed For Issue I C I, ehe staff Mu5s that the AfdshiAEitds~ar[MEoS$sIy alErNhMs. Supportag analyses necessary to desesarase the effecs, veness of operasor actums am resgene to transsents and ah sliculd also be prontedgWesteghouse ___ ,,,_ _ __

Rewisson i ofilEat-powEC. y Response G'umielsnes was subnuned by DCP/NROO376 on 8/985U ateorinal . ,and adnuantrative plant procxdures are the responssinlay of the Combened Lxense applicant as indicased an SSAR Section 135.

pian W- See NRClener dased 12S/96 Cloned with ERG Rev 2 subnurial. sin 1/15/97.

Forwarded seport copy of thes nem so NRC for confirmatson that NRC status should be Actum-N to 7fthis acus as secolved. rks 4/65/97__

~

NRR/lHUB RAIOl Actium N ' Action NSD-NRC-97-5119 '

5247 18 QS/Kerch

[ Respond to NRC *Edmorial Cornments on the AP600 Human fears EndDucumentarson received 15y'lener dard apn124,1997._ ]

[Teledwnh Kerds/llongarra held and SSAR and WCAP changes agreed upm Markups acre UM'ylor summagement sevied l IC- u sec'd and feel changes bemg incopormaed. Will provide a letter so NRC transmettag snarkups and WCAPs; nestups so be incorporased h SSAR Rev 13,end May. rks S/8/97 Westinghouse action complete with SSAR enasimp and W_APs sobrantal by NSD-NRC-97-5119. rLa S/19 Acuan N - Change _s included in the leuer are included in the $$AR __ ._ , , _ _ _ . . _ , _ _ _ , , _

3 Q( N.

- 'y ab, 7A, r prom . -

g , (Ocv O, GU DfdV DN M p% g SMU5 u dCM *

~

7 g % M rfMS-Page: 1 Total Records: 2 "naia ou, a

x __ -_.- -__ - -

AP600 Open Iteni Tracking Systesu Database: Executive Seemanary . Date. 7/17s7 Selection: [nrc se ctalej=*Actaan W* And [ resp engl hke

rap ** Sated by Item #

licm DSER Section - Tarle/Dewnption Resp -

(W) NRC No firarah Question Type Detaal Stases Engmeer Sames Sysms g ,,,, g, f gg, 3943 NHR/SPSH 16 2 MTG-Of RAP / Canton.nnke Ckned Action W NTD NRC-%-48M

.. - . . -. . . - . . - . . . . . . . _ , . - - . _ . . - . . . . _d . - . . . . . . _ _ -- .,

jHased on an Augma 16.1996. meeting with the NRC so nesolve their commesas on the Rehat= hey Assarnace Prugrans (SSAR Secteun 86.2). we have

!placed the hydrogen ignsors and contamment fan <oolers in the "Rask Segaaficant SSCs under the Scope of RAP

table wah TBDs as justaficanon jfor seclasson. Upon completion of the I~veused PRA, there udl te an evalemson so deternene af these components should be escluded as time RAP l table. A SSAR Secison 16.2 markup edi he transmarted by DCP/NRCU612 (NSU NRC-%4830L Thrs surkup will be sevued based on the r 5 S E -

2. - n1=_ . .. m.1. r -- -- - - -

fSSAR markup =as e-- ==-d To close.this martup sfumid be nevned to reflect replacement w deletion of TBIA sin 10fl666 l jSSAR confirmasury to ensure the marimp provided an NSD-NRC-97-4958 as encorporated uno SSAR Rev 12 NRC did out onafum tius maskap, frather. they revised the cruena for selectang SSCs for the R AP. Suggestion mas made 4/I5 to NRC that stus seem be (kmed wish NSD-NRC 97-4958 ~ ~ ~ ~

'!and the follow 4m actions be tratked by OITS asem 4852. sin 4/15N7 - #%^ ~

4n52 NRiullQMH 16"2 MTG 6 RAP /Carnum Mb bion N NSD NRC-97-5822

~ '

[Evalsascimpact of NRC INU/16N7).Unsena for Enaabhshmg Rask SignifEaN Structase[Sysnems 'and Componeers kS555 en tE2ansadened lfw the AP6M rem!ay Asswanu Pmgmn. __g

= = == = , = = =

la psogress. See NSD NRC-97-4958, espimamg thes lener as hemg evalmased via 2/I(W97.

The RAP tahic of SSCs has been revned based on the mannsenance rule cntena and espert panel review and a draft =di be sens to NRC om May 2.

With NRC-m on the draft, the saasus wdl be changed to ConferneW so ensure at gets uno the cad May SSAR sev. rLa 5/1/97 Ttus draft was not sens due to cznensive engt seview comments. The RAP will go as SSAR Rev I4 A markup es tesng vent en the NRC(the letter was ready 6/18) ran 6/25 j

, . =y Action N - Response provided by DCP/NRCtMm62 of 6/26/97g31_ ._ . _ . _

_j

o. & Bebcos 9 Q;w 4 bps o a k

s Joe- vo c O&% P*l'[

h '

. , a m -

A

[g h: Y ! #

% t

/

i Page: 1 Total Records: 2.

= ._..m___.-___u-

,- gj AP600 Opes Item Tracking Systens Database: Executive " - - -- -- y pose: 7f37jy7 o Selection: [nrc st codej=*Actwo W* And [ resp engj like *SD** Sorted by item #

Incm DSER Set 1eun Title /lxscrrpsson R'5P (W) NRC N2 Branch Question Type Detal Samus Engmeer Samus A g,,,,g ,f gg,, - -

2919 NRR/SPSB 19 i RAlOf SIVBuctedWallace Closed Actson W - SD-NRC-96 46HO Shutdown PRA followen questson so DSER Of 59.l.33-1:

,Open seem 19 I.33-1 sequested Westinghouse to ify the low burnen error rate for snadvenent drasning of seacsur vessel seventary tisough the

Nonnel Resadual Heat Removal (RHR) system In response. Westinghouse quannfsed the laitchlumd of the operasor ovd- ; the scactor coulant b system dunng drasa down operassons so scach medioup condsesons. Wessinghouse also quannised the lacialumd ehnt a LOCA could occur by '

anadvenent opensag of Nannal RHR valve V024. The staff needs the followsag informanson to conclude chas she frequency of overdramag the i \ reactor vessel to reach audloop condetsons is on the order of E-6 per year, winch as much lower than cunent operasmg espenence iB iq 045

@q' t 0 a. westanghouse shouid use operating espenence so desernene the frequency of the operasar inadecnently m.t ._.g the RCS denng nedloop.or pestify thm cunens operating expenence is not appiscable by desenbang any Al%UO design ? - -- u over cunear plants

()h , 'b Westmghuuse needs to add snuse enformasson en the shusdona PR A about the availaNe Bew! sestrunnentaewn dunng the drama down

. process. A descnpesan of lum the guessuruer wade range level sastrumentatson is conneaed so the RCS would be tecipful jl, .

. c. Westmghouse needs to clanfy in the PRA how the two nus leg instrunnrnts are conneced and clanfy whether they share coenmum erference h.

d. Wesunghoene needs to documens in the PRA the basis for she beta factor of 005 for the hun leg ansarunents. Tius value is na hstoJ se Chapter 29 or Scanon 54 7 of the PRA.

C e. For drase down scenano 2. Wessinghouse needs tojustify the likeldsoud that the aar operased valves fad so clo=c on demand. Westmghtwe b A (\h =needs to (1) documens the sesting sneerval for these valves and (2) calculme valve unavalatesisty using Hstandby fadure rase)%esung enerrval)f2)or a

).

t d } demand fadese rase (such as IE-3 hssed in Table 54-58). = = := - : a = :, w =:

((

_==:-- =_2.----

Q ' ') t

' qloned - Response provsded by NSD-NRC-96-4680 and sevmon I of RAI sespnase NSD-NRC-96 4739 -

). ,C Closed with Shutdown Evalueeson Repurt. WCAP-14837 Rev 0,transnused by N_SD;NRC-97-5062 _a_4/I5 ,

i ,

^

RAlOI SIWtzwas Ckned Acteue W NSD-NRC-96-46h0 3007 NRR/SPSB 19 i -- - -- . -

Tine followeg quesnons pertain no shusdown opermison auh the RCS open t a Accordmg so the SSAR Chapeer 6. Stages I.2.and 3 of ADS are saanuaDy opened PRKMt to enshating RCS draumlown operanons to nedbup condetsons. However. no inforznasson is provided an the shutdown PRA as to when ADS as opened pnur to drase duwe operatnues Please docaensen8 m the shusdown PRA how tius $$AR assumpuan will be ases (i e Tech. Specs, minun controls.etc_)?

b. Dunng RCS drasadown operasson wah Stages 1. 2. and 3 open. af Nonnal RHR ctmisag is lost. the operator has to enanually initsase gravey snjection froen the IRWST. If the operasar aannees gravsty injection AFTER the RCS beges to bust.could surge Inne fluudsag occur and cause

.gr.vey injectson so stop? The staff sequests Westmghouse to provsde analyses venfying that surge line floodsag is nun a problem,assunung any RCS

,j hel _ ,,, _,,. .- -_ _ _ , _ _ , _ , _ , , _ , , _ _ , _ _ _ _ _ _ - _ .-__

~. - - ,

l Closed - Response provided by NSD-NRC-96-4680

Reopened pendmg cosapletion of the less of RNS annlysis for the Shusdown Evatuation Report. WCAP-I4837. Actam W to revene 4 8 5. See NSD NRC-97-5062. ska 4/15 (Westinghouse issued 6f6 by IX'P/NRC9t97. Closed. rha . _ _ _ _ _ . _ _ _ ._ ,

Page: 1 Total Records: 3

. . . . . A... . . . . .

O . r!

l AP609 Open item Tracking Systeen Database: Executive f-- i Dise: 7/1737 )

Selecties. [nrc sa code]=* Action W" And [ resp eng] like *SD** Sorted by item # -

hem DSER Sectwn Tale /Descripuin Resp (W) NRC Engmeer No Branch Quesnon . Type. Detail Samos Stmas 3 % % No f tw 0865 NRR/SPSB 15 KEY ISSUE SIVlewts Gosed Actwo W m , _ _ _ . . . . . . . - . -_ _ . . . . . . _ - . . . . _ _ _ _ . _ _ . . _ . . . . _ ._ ._

Key issue Noenber

22. Shutdown and Im Power Operations (SPSI4 Espenence wah events occurnes dunng shutdowsopee- a indscues thm substanraal safety .m..- ;s ase warrameed for low power and shutdown operassons. Westanghouse responses to RAF- ;gardmg the shusdown nsk issue are mostly qualemsve wndios quesnamsve amelyses. The staff has also requessed Westanghouse to provide a sysaematic evaluatsue of the Al%00 desage against the issues adentified sa NUREG-1449.

Included in this issue is whether the proposed APM10 TS comply with SECY-93-190 *Regulmory Apprumcb to Shutdown and I.aw-Pbwer Operatsnas." and NUREG-1449 *

"Shutduwe and Im Power Opersion a C- -a Nuclear Pawer Plants an the Uened Semes "jsee Sectum_19 I of the DSER )-. _. _ _ _ __ __

_..._J.

Remanas open pending complessoa of the Emss of RNS analysis for the Shutdown Evalearson Report. WCAP-14837. Artice W so sevne WCAP-

- 14837 Sectson 4 8 5. See NSD.NRC-97-5062. rha 4/15 g Q,Westmshouse issued 6M by DCP/NRC-897. Ched_ rka__ _ _ .__ , _ _ _ _ . _ _ _ _ _ _ _

t i

Page: 2 Total Records: 3 .

^

'W)ase Westinghouse -

TAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION OATE: %/ t I, f f,7 NAME:

TO:

L g,gy LOCATION:

lem l<e~/cN ENERGY CENTER .

EAST PHONE: FACSIMil.E: PHONE:

' Ott,c8: Vr 2 -3 7V- ra 9o COMPANY:

Facsimite: do:

i __ M .I A/4 c 284 4647 LOCATION: outside: (412)374 4087 Covor + Pages 1+/

The follow 109 Pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville. PA. If any problems occur during this transmission, piease call:

WlH: 284 5125 (Janice) or Outside: (412974 5125.

_ _ - y- _

% %,u, COMMENTS:

Tm pin e o A e w ee n.~ n we m e er nenw9.[se~rw ._

lcu em rA i*l ]fttf9 7 8</ WDry. Blt) c H Au c s' w n e_ t der in s-sArt fl e w sao u ( C .

cc' .ov cc. d enJ Hw v'@lc D .

Lu wTD J GAC TN WAG 7EA,vus Ewnf -'- -

710MTD (0(ig.t 3(4)

W h

a

3

, 9. Auxiliary Systems i

Table 9.5.1 1 (Sheet 27 of 31)

AP600 FIRE PROTECTION PROGRAh! CON 1PLIANCE WITil IITP CS1 Ell 9.51 IITP C51 Ell 9.51 Guideline Paragraph Comp'" Remarks 203. Portable extinguishers and manual hose C.7.g C stations should be readily available outside the battery rooms.

Turbine llullding 204. The turbine building should be separated C 7.h C from adjacent structures containing safety-related equipment by 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months rated fire barriers.

205. The fire barriers should be designed to C.7.h C maintain structural integrity in the event of collapse of the turbine structure.

206. Openings and penetrations in the fire barrier C.7.h C should be minimited and should not be located where the turbine oil system or generator hydrogen cooling system creates a fire exposure hazard to the barrier.

Diesel Generator Areas 207. Diesel generators should be separated from C.7.i AC The standby diesel generators each other and from other areas of the plant are separated from each other by 3. hour rated fire barners. by a 3-hour rated fire barrier and are housed in a separate structure. remote from safety-related areas.

. 7he ancillary diesel gener-ators are separated from other U*Mhof the plant by 3-hour rated fire barriers. The ancillary diesel generators are not separated from each other, but can be easily replaced with transportable diesel generators.

Revision: 15 DRAFT July.1997 9.5-56 W Westiflgh0USB

4 s

- TX CONFIRMATION REPORT ** AS OF JUL 22 '97 9136 PAGE.01 APG00 DESIGH CERT DATE TIME T0/FROM MODE MIN /SEC PGS STATUS 01 7/22 09:35 301 504 2300 G3--S Ol*00 02 OK l

$ g I

e e

?e 4

FAX TO JOE SEBROSKY (NRC)

July 22,1997

Subject:

Markup of SSAR section 13.3.1 and response to RAI 720.391 Joe, Attached are the following two items:

l

1. Markup of SSAR subsection 13.3.1. He COL item will be changed as shown to address Bob Palla's (NRC) l comment that he needs post 72 hr actions stated within a COL item. He placement of this statement is '

consistent with Westinghouse's commitment in our letter of March 14,1997 (NSD-NRC-97 5024) to NRC, and the NRC's reply in their letter of Junc 5,1997. He change will be made in SSAR Rev.15 unless the NRC informs Westinghouse otherwise. This closes, from die Westinghouse perspective, the accident management topic discussion we had with Bob Palla during the 7/17/97 telecon, if further information is need by Bob, please inform us immediately.

- 2. Response to RAI 720.391 (OITS #5504). This response is formally being sent to you, but I ask that you pass a copy to John Flack and Nick Saltos (PRA group) in the meantime, his RAI pertains to Nick's

- question about the PCS drain system functioning during a seismic event.

If you have questions, please call me, A copy of this fax wiIl be placed in the Westinghouse informal correspondance file and will be formally included in a transmittal of that information.

Regards, -

. y

>~

Cy la Haag Advanced Plant Safety & Licensing ec: J. Winters D. Lindgren .

J.- Evans 4

e 4

^

e y

13. C:nduct ef Operation t

e

^

for the high level requirements for the technical support center and the operational suppon center. See Section 7.5 for identification of plant variables that ars provided for interface to the emergency planning areas. -

Communication interfaces among the main control room, the technical support center and the emergency planning centers are the responsibility of the Combined License applicant.

13.3.1 Combined License Information Item Combined License applicants referencing the AP600 cenified design will address emergency 3 planning,and its communication interface.

3.e 13 pot.- M kca u%s 13.4 Operational Review '

Dis section is the responsibility of the Combined License applicant.

13.4.1 Combined License Information Item Combined License applicants referencing the AP600 certified design will address each operational review.

13.5 Pla'nt Procedures i

l Plant procedures are the responsibility of the Combined License applicant. References to

(

applicable combined license information are included in Section 1.8. His includes, for example, reference to guidelines on inservice inspection in Chapters 3 and 6, and initial testing in Chapter 14.

I Reference 2 provides input to the Combined License applicant for the development of plant operating procedures, including information on the development and design of the AP600 emergency respanse guidelines and emergency operating procedures. Also included in Reference 2 is information on the computerized procedure system, which is the human system interface that allows the operators to execute the plant procedures.

13.5.1 Combined License Information Item Combined License applicants referencing the AP600 certified design will address plant procedures including the following:

Normal operation

Abnormal operation Emergency operation l

1 L

Revision: 13 May 30,1997 13-2 3 W852 gh00S8

NRC REQUEST FOR ADDITIONAL INFORMATION pun m u.4 lMl I

Question: 720.391 j A possible failure mechanism of passive containment cooling is the blockage of the baffle as a result of a seismically induced blockage of the drain system (not included in the SMA). De opening of the PCCWST air.

operated vahes (due to loss of compressed air upon loss of offsite power) would release water which would block the bafne if the drain system fails due to the seismic event. Westinghouse should evaluate and discuss the feasibility of this containment cooling failure mechanism.

Response

As stated in SSAR subsecuon 6.2.2.2.4, "the (PCS] cooling water not evaporated from the vessel wall flows down to the botton of the inner containment annulus into floor drains. De redundant floor drains route the escess water to ' storm drains. The drain lines are always open (without isolation valves) and each is sized to accept maximum passive containment cooling system flow, he interface with the storm drain system is an open connection such that any blockage in the storm drains would result in the annulus drains overflowing the connection, draining the annulus l independently of the storm drain system."

The annulus floor drains are essentially pipes embedded into the wall of the shield building. De water travels through the pipe to the outside of the shield building, ne interface with the storm drain system is an open connection, meaning the water can either travel down a pipe (like a downspout) to the storm drain or if the storm -

drain or "downspout" pipe is blocked, the water simply overflows through the open end of the drain pipe that is located outside the shield building, and dumps onto the ground. Rus, success of the safety related drains is independent of the availability of the storm drain system. Rather, success of the annulus drains is dependent on the shield building integrity, he PCS, including the annulus drains, is part of the shield building structure which is qualified as Seismic Category I.

Since the annulus drain is embedded within the shield building concrete wall, it is covered by the shield building HCLPF. Seismic failure of the PCS is modeled in the seismic margin analysis, and it is usumed (see Table 55 3, sheet 4 of 4) that the shield building wall HCLPF is conservatively the same as the shield building roof HCLPF.

De shield building HCLPF is dominated by the shield building roof and has a calculated HCLPF of 0.58g.

Rus, the sequence postulated in this RAI, of the drain system failing due to the seismic event leading to water blocking the PCS air baffle, is not a feasible seismic failure mechanism of the PCS Water will still drain out the shield building onto the ground if the storm drain lines are btbcked or failed due to the seismic event. De annulus drain can only fail when the shield building wall itself fails.

PRA Revision: None.

T westinghouse

r

, Ettl H Nonnd tern Phone (412)374 4790 '

Fax (412) 3744011 Email. novenceh 4 *estinghouse com k' pil 2 2 'i337 y( A" '

Brian A. Mc Intyre l To: Bill Huffman (NRC) From: Eari H Novendstem CC: See Below Date: July 2.1997 Ret Ch.15' Discussion items Pages: Cover + 2,, S, 3ej l

O urgent O ror Review O Please comment O Please Repey O Please Recycle e Comments:

Bill Attached are advance discusson items responses for the majonty of the non-LOCA Chapter 15 SSAR. It is marked draft, only because they haven't been issued with a Westinghouse Cover Letter.

Please give copy to Summer.

Thanks.

0 OY %

g6 6Ag cc: B. McIntyre (NRC Informil Correspondence File), S. Fanto, E. Cariin. P. Rosenthal, J. Winters, New File 7.5.1.7 (SAR), W. Cartson t

07C171149 AM 970712 Fax en Ch. t5 Commrmczx

r

'f PRELIMINARY

- Question 5.

Chapter 15 should include a table or similar summary documentation of system actuation times, valve closure time and systems parameters assumed in each safety analysis.

Response

Tables 15.0-4 and 15.6.511 of the SSAR have been expanded in Revision 13 of the SSAR to include the limiting actuation times, valve closure times and system parameters used in the safety analyses.

e *

-- W 5

em O

6

=.

2_

PRELIMINARY

<e

, increase in Feedwater Flow (SSAR Section 1512) 6 The analysis should include the calculated transient DNBR, RCS and steam generator pressures to support the conclusion for the full power case that it meets the acceptance critena of the Condition 11 events.

Response

Revision 13 of the SSAR provides updated analysis for the excessive feedwater flow event that includes an explicit value for the predicted minimum DNBR As discussed in Revision 13, that value is 2 63, using the WRB 2 equation, which is well above the design limit defined in Section 4 4 of the SSAR.

Attached Figures 61 and 6 2 present 'ie predicted pressurizer and steam generator pressures for the same feedwater malfuncuon case reported in Section 15 l 2 of Revision 13, except that the case reported here provides a full 800 second transient. The SSAR Revision 13 case terminates shortly after the reactor trip (rod motion at 488 4 seconds). - Both the subject cases are mitia'ted from full power with maximum reactivity feedback coef6cients and assume that the rod control system is in 11anual. The lengthened analysis time used m the current case allows a more complete characterization of the steam generator pressure transient than the case in Revision 13 of the SSAR, that is primanly intended to address DNB concerns. , ,

The resuits'in Figures 61 and 6 2 show that the peak predicted pressures are 2320 and 1107 psia for the pressurizer and the steam generators, respectively. Therefore, both the primary and secondary systems remain well below 110% of their design values. In conjunction with the DNBR results discussed above, these results ensure that all Condition 11 acceptance entena are met for the excessive feedwater flow event.

Also attached are Figures 6 3 through 6-5 that present ihe predicted nuclear power, loop temperature nse (AT), and core coolant mass flow for a fuit 800 seconds Revision 13 of SSAR Section 151.2 presents the same results, but only for about 500 seconds. When observ ng the response of the parameters in Figures 6 3 through 6 5 during the additional 300 seconds, it should be regier6bered that reactor trip, turbine tnp, loss of offsite power, and reactor coolant pump coastdown all take place at about 490 seconds.

I e

e

o. '

PRELIMINARY 2500

._.__.._.._._._A, w.-.~._. . ,

l ~

i l ^ 2000 - -

o -

m -

m. .

w c 15 0 0 - -

,m- . ,

u .

c. 10 0 0 - -

500 l l :

0 200 400 600 800 Time (seconds) i 4

Pressuriser Pressure Figure 61 1

6 4

49

6

~

PRELIMINARY l

--- Leep 1 steem generater pressure Loop 2 (preseurlier loop) eteam generster greesure l

1200 .

1100 -2,

~

n o .

. 1000 - -

m :

o. ;

900 -.

o : -

m .

e 800 -y ,

m .

m -

e u

700 - .

c. :

600 - -.

~

500 0 260 460 660 800 Time (seconds)

Steam Generator Pressure Figure 6 2 O

e ip.,

,---.---my-,-- . - - . . , - - - - -- - - , - . - . - - , , - - - - - - - . - . - - - . -

PREllMINARY

! - 12 l # .

e -

,_f._._._._

E' O .

e 1

.. . l M $

U

= .6- -

e w =

, e a he * -

. .4- -

9 .

O L

.2- -

l

\-

E ' ' ' ' ' ' ' ' ' ' *- ' ' '

0 0 260 460 860 800-Time (seconds)-- .

6 g

Nuclear Power Figure 6 3 e

N e

4

-d

%O

" w.

Q 4

PRELIMINARY l

12

_ i.

__.. _. _ .___. _. _ . _ . _ . _ .)

g 8--

l

. .i

. .1

.s. .

~ .

. . l

=

4 . -

i -

\

2- -

u .

\.

0 0

' ' ' ''' x'''''"'*-- '

0 200 400 800 -800 Time (seconds)

Core Coolant Mass Flow Figure 6 5 '

9 0

6-

%9 y

7

' RELIMINARY

!. 80 ,

be

, yg .: ._._._._._._._._. -

w .

l l sg . . <

, e -

i_ e : {,

l a . i v 50 -;

l. .

~ E

' \

i 40 - -

o : '{

~

l

e 30 -

- 1 o -

a. 2 0 -j 'g o .

o . \.

-J 10 - -

N,N.

"t i I t t t 1 i 1 e t I I i

,,****=n n o 0

0 260 460 660 000 Time (seconds)

Loop Delta.T Figure 6 4 e

t 4

e S

PRELIMINARY B Inadvenent Operation of the Passive Residual Heat Removal System (SSAR 15 216) 13 Explain how the zero power case is bounded by the inadvenent opening of a SG relief or :afety valve. ,

Response

Both the zero power case for the inadvenent operation of the PRHR heat exchanger and the inadvertent opening of a steam generator relief or safety valve (SSAR 15 2.1.6) are initiated with the plant at hot zero power conditions. The analysis for inadvertent opening of a steam generator relief or safety valve assumes that the PRHR heat exchanger is actuated co. incident with the opening of the steam system valve. This assumption is conservative, since the combined effect of the PRHR heat exchanger and the open relief or safety valve produces a more severe RCS cooldown than either component alone. A more severe cooldown increases the reactivity insenion which increases the possibility that the reactor will return critical.

The zero power case for the inadvenent operation of the PRHR heat exchanger event only cools down the RCS by operation of the PRHR heat exchanger. The resulting maximum cooldown and reactivity insertion from this event will be less than that produced by the 1 .

inadvertent opening of a steam generator relief or safety valve, as described above.

Therefore, the results for the inadvertent opening of a steam generator relief or safety valve bound those for the zero power case of the inadvertent actuation of the PRHR heat exchanger event.

4 e

e e

9 5

Turbine Trip-(SSAR 15.2.3) PREUMlNARY Question 14.

This analysis does not address compliance with the GDC 17 requirements. To satisfy GDC:

17, the effects of a loss of offsite power on the turbine trip event should be considered.

Response: .

1 Revision 13 of the SSAR provides updated analyses of the turbine trip event with a loss of offs'te.

power assumed. The revised analyses address compliance with GDC 17, O

e e

/o-.

Loss of ac Power to.the Plant Auxillaries (SSAR 15.2.6)

Question 15. Provide a DNBR' transient curve for a loss of ac power event.

Response

A DNBR curve has been supplied as Figure 15.2.6 12 in Revision 13 of the SSAR.

l l.

I 9

4 4 e t

Y 9

e 4

4 m-_

9

//

i PRELIMINARY Feedwater System Pipe Break (SSAR 15.2.8) j Question 18; The loss of offsite power, resulting in a RCS flow coastdown, is assumed at the time .

of reactor trip. Westinghouse has stated that this is more limiting than the case where power was lost at the initiation of the event. An earlier RCS flow coastdown, which reduces the capacity of the primary coolant to remove heat from the core, may result in a larger increase in the peak RCS pressure. This analysis should include the i - techt.ical basis needed to show that a coincident loss of power with a reactor trip will result in a highest peak RCS pressute during a feedwater line break event.

. Response:

During a feedline break, a RCS flow coast down due to a loss of offsite power is more limiting if the loss of offsite power occurs at reactor trip than at the initiation of the event. This is demonstrated by comparing the following analyses:

Case ! - SSAR feedline break case. Offsite power is assumed lost at reactor trip.

Case 3 - Feedline break with offsite power lost at the initiation of the event.

- Figures 181 and 18 2 show the results of the analyses. In both cases the feedline break is initiated at 10 seconds. In the Case 1 (SSAR case ) the reactor is tripped on low steam generator water level.

The rods begin dropping into the core and offsite power is lost at 85.1 seconds.

Case 3 assumes that offsite power is immediately lost at the start of the event concurrent with the feedline break. Reactor trip in Case 3 occurs within

I second of the start of the event on the reactor coolant pump under speed trip function. Case 3 is less severe that Cue I because of the early reactor trip.

Figure 181 shows core power during both cases. The core power in Case 3 rapidly decreases within a few seconds of the start of the transient. In Case 1, the reactor remains at approximately full power for over i full minute. During this time period a beatup of the reactor coolant system occurs and the-inventory in the steam generators is significantly depleted. When Case I finally reaches the reactor trip setpoint and offsite power,is lost,' the plant condition are much worse that those of Case 3.

Figure 18 2 shows the pressurizer pressure for both case. The peak pressure for Case 1 is 2620 psia, while for Case 3 the peak pressure is only 2491 psia. -

Due to the differences in the time of reactor trip Ind the heatup of the RCS prior to reactor trip, more severe results occur if offsite power is lost at reactor trip rather than at the initiation of the event.

i

% .~ g -

.y ---

/L -

PRELIMINARY Case 1 offaite power foot at time of reotor tria _

Case 3 offaite power lost at event i n i.t l e t i o n I ~l.2 1- i

- l t ,

2-

, * '8 -

~ -l 6- i-4 - -i

.4-I'i

-i

2- \

s

~. %,

5

, i i$ ,1,7i,' , ,, .. , , , ..,,, , , , ,,,,,

, 0 , , ,

1 2 3 4 5 4

10 .

TO 10 10 10 10 Time (S)

Figure 18-1 Nuclear Power During Feedline Break Events ,

9 J

4 A

N S

~

PRELIMINAR'f J

t 4-I 4

Case 1 offsite power lost at t ic a e' rooter trip

- - - - Case 3 offsite power lost at e v s .i t enitiotion

- 2800

= :

s i ?a. 2600 -: -

- ~

g 2400 - [s \s ,'"

c

. . i '

" ~ #

2200 - '- -

s ,

4 e - ,

o- -

. m

- 2000 - -

~

4

1800 -

Is1600 - -

a :

a o

1400 - -

a. ~

1200 ''l ''l '"l 1 2 3 4 5 6 7 10 ' ,10 10 10 10 10

.c T.ime s Figure 18-2 Pressurizer Pressure During feedline Break Events 4

1

.h 1

i 1

, , n - ,.

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

If Feedwater System Pipe Break (SSAR 15.2.8)

PRELIMitiARY Question 19, This analysis (page 15.217) assumes a double-ended rupture occurs after reactor trip on the low low steam generator level signal. This is inconsistent with the sequence of events in Table 15.21 (sheet 5 of 5) that shows initiation of the feedwater line break before the reactor trip. His inconsistency should be corrected.

Response

While the text in the SSAR may be ambiguous, the text for the SSAR is not inconsistent with the methodology used for the AP600 feedline break. The feedwater line break is initiated before reactor trip. After reactor trip the characteristics of the break are changed. The purpose of doing the analysis in this manner is to develop a conservative case which bounds all break sizes and interactions e of the main feedwater control system with the break flow.

The following general characteristics and phenomena for large full double ended and small feedline breaks are combined as described below, to perform a bounding analysis of this event.

Full double ended ruorure characteristics

A full double ended rupture results in a rapid blow down of the faulted steam generator, ne fluid inventory in the faulted steam generator is rapidly depleted and steam generator pressure also decreases rapidly. He low steam generator level reactor trip signal is reached very early in the transient. The depressurization also causes the low steamline pressure setpoint to be reached and the PRHR heat exchanger to be automatically actuated early in the transient.

If the break is between the steam generator and the feedline check valves, reverse flow from the faulted steam generator will occur,

All main feedwater flow will spill from the break.

Check valves in the feedline will prevent reverse flow from the intact steam generator, ne intact steam generator experiences a loss of feedwater (i.e. zero feedwater flow into or out of the steam generator) because of the response of the feedwater control

, system to the depressurization of the faulted steam generator.

As the feed ring in the faulted steam generator uncovers, reverse flow from the steam generator will include steam. This produces a time period when the feed line break is similar to a steamline break ad over cooling of the RCS occurs. ,

The faulted ano intact steam generators behave differently. At the time the low steam generator level reactor trip setpoint is reached in the faulted steam generator, the intact steam gene,rator may still contain significant inventory which can be used for removal of decay heat from the RCS.

Small feedline break characteristics

During a small feedline break, interaction's with the main feedner system may affect I the consequences of the event. Depending upon the break size, the main feede,ater control system may be wie to continue delivering now to the steam generators.

With a very small break the main feedwater system may be able to deliver nominal now to both steam generators and spill Guld from the break. With slightly larger break, the faulted steam generator may not receive any feedwater Dow, ne intact steam generator may receive a reduced amount of feedwater dow. With a still slightly larger break, reverse now from the faulted steam generator may occur. The intact steam generator may receive no feedwater now. Check valves in feedlines will prevent reverse 00* from the intact steam generator.

Following a small feedline break, the time to reactor trip on low steam generator le"el will be delayed in comparison to a full double ended break,

With the reactor trip delayed, steaming to the turbine by the intact steam generator will result in depletion of the Guld inventory in the intact steam generator. ne intact steam generator will have less inventory available for post reactor trip decay heat removal.

Following reactor trip, the main feedwater control system could th.ottle feedwater now. Rus, suddenly at reactor trip, reverse now from the faulted steam generator could suddertl y start or increase because the main feedwater system is no longer

upplyit g fluid to the break. Similarly if offsite power is assumed lost at reactor trip, main feedwater now would suddenly stop.

Following reactor trip, the faulted steam generator will continue to blow down, but a small break size will limit the rate of inventory loss from the steam generator. During a small foodline break, the faulted steam generator may be able to provide some post reactor trip decay heat removal from the RCS, .

ne methodology used for the,AP(,00 feedline break analysis merges characteristics of small and large feedline breaks into a single case with results that bound large and small feedline breaks. De methodology used for the AP600 feedline break analysis uses the following break assumptions.

"a") At start of accident -

Small feedline rupture occurs, No feedwater flow into or out of either of the steam generators occurs, De main feedwater control system interacts such that all now released from the break is supplied by the main feedwater system. Both steam generators continue to supply steam to the turbine which depletes their inventories.

h.

PREllMl NARY b") At reactor trip on .

Main feedwater is terminated. Reverse now from the faulted low steam steam generator starts. The reverse now is calculated generator usuming a full douHe endNi feedline rupture occurs. ne water level reverse now is assumed to be saturated water until the faulted steam generator liquid invertory is completely depleted.

i Zero feedwater now into or out of the intact steam generator continues.

i Combining assumptiona "a" and "b" into a single event, produces a transient with the most limiting aspects of both a large double ended and a small feedline break. Assumptions 4" and "b" increase the severity of the foodline break by delaying Hactor trip, minimizing the cooldown effects of a feedline rupture, minimizing inventory in the steam generators at the time of reactor trip and

. minimizing the post reactor trip decay heat removal ability of the steam generators.

Assumption "a" in essence results in : complete loss of feedwater Cow to both steam generators at the start of the event. This delays the time at which the faulted steam generator reaches the low level reactor trip serpoint. When the reactor trip occurs, the intact steam gene'ator Guld inventory will also be depleted to the low level trip setpoint. This minimizes the amount of Guld available in the intact steam generator after reactor trip for decay heat removal. Prior to tuctor trip, with no

depressurization of the faulted steam generator or reurse flow from the faulted steam generator, the ,

i RCS initially experiences a hutup instead of a cooldown. At the time of reactor trip, the RCS temperatures and pressure are elevated above nominal values and higher amounts of energy ara stored in de RCS, After reactor trip, use of assumption "b" quicidy depletes the remaining liquid inventory in the faulted steam generator before it can be used for any significant decay hut removal. Within approximately 5 seconds the liquid blowdown through a full double ended rupture depletes the remaining Ilquid

inventory in the faulted steam gerwrator and impairs the faulted stum generator decay heat removal ;

ability By ass.iming a liquid blowdown, the steam generator depressurization induced cooling of the RCS is also minimized.

The SSAR analysis does assume a break at the start of the trardient. Interactions between the break in the feedline and the main feedwater control system are assumed to initially sesult in a complete loss of feedwater to both steam generators such that no feedwaar flow is deliver to or lost from the feedwater

. nozzles. At reactor trip the assumptions are modified such that the faulted stum generator blows down through a double ended rupture and reverse now occurs from the faulted stum generator.

Revision 13 of the SSAR includes additional text to clarify the assumptions related to the break Cow i- and feedwater system lateractions. .

d

-.- --y ,.., .,.y. - - , --y _ ,, - .,

y ,n.., , , , , .- , --.-----v, ---

I7 PRE LiMIN ARY Feedwater System Pipe Break (SSAR 13.2.8)

Question 20.

De staff noted that the non safety related startup feedwater system (SFWS) and the pressurizer spray (PS) were credited (page 15.219) for heat removal to limit the increue in l

the peak RCS and steam generator pressures. Use of these non safety related systems to mitigate the consequences of the feedwater line break event is not acceptable to the staff.

Either reanalyze the feedwater line break event without credit of the SWFS and PS or demonstrate that the effects of the initiation of both non safety related systems on the feedwater line break event is insignificantly small. Also, it wu noted that the pressurizer safety valves were set at a minimum value. Justify the use of the lowest pressurizer safety setpoint for overpressurization prediction for an Fl_B event. I

Response

The non safety related startup feedwater system is not credited to mitigate the consequences of a feedwater system pipe break, ne text in the SSAR was summarizing actuation logic which could potentially actuate the non safety related startup feedwater system. The analysis in the SSAR does not credit actuation of the startup feedwater system and the ambiguous text discussing the startup l

feedwater system has been removed in Revision 13 of the SSAR.

The nonqafety related pressurizer spray and a low pressurizer safety valve setpoint were used in the .

analysis presented in the SSAR. The impact of thwe usumptions on peak pressurizer pressure is insignificantly small. To quantify the impact of pressurizer spray and the pressurizer safety valve setpoint on the peak pressure during a feedline break, analyses for the following two cases were performed:

Case 1 - Feedline break with pressurizer spray operable and a low pressurizer safety valve setpoint (case presented in the SSAR)

Case 2 - Feedline break without pressurizer spray operable and the normal pressurizer I safe'.) valve setpoint l

Figure 201 shows the pressurizer pressure from Cues I and 2. Case I with usumptions which minimize pressurizer prenure has a peak pressurizer prusure of 2620 psia at 89 seconds. This peak pressure occurs several seconds after the reactor trip which occurs at 83.1 seconds on low steam generator level. Case 2 which has pressurizer spray turned off and a higher safety valve set point hu a peak pressurizae pressure of 2624 psia. De peak pressure and the reactor trip occur in Case 2 at the same time as in Cau 1. The difference between the peak pressures is insignificantly small and in both cases the pressure remains below 110% of design prenure.

It should be noted that the high pressurizer pressure reactor trip setpoint (2460. psia) was reached at 48 seconds in Case 2, but no credit for this reactor trip function was usumed. Case 2 continued to usume that the reactor tri5l ped on low steam generator level at 83.1 seconds, if the high pressurizer pressure trip is credited the peak pressure in Cue 2 would be lower. In Cue 1, the high pressurizer pressure reactor trip setpoint was not reached until after the reactor was tripped on low steam generator level.

De feedline break is less limiting with respect to margins to RCS design pressure than other events

jg

.. PRE LIMIN ARY such n a locked reactor coolant pump rotor event. Ti.erefore, usumptions are used in the analysis which emphaske other riuantitles which may approach design limits. The feedline break is a loss of heat sink eved and cors . design limits may be challenged following reactor trip if adequate decay hat removal is n:t provided. Verification that the core remains in place and intact with no loss of core cooling capab.iity can be demonstrated by showing that no boiling occurs in the active primary coolant system prior to the time that the PRHR heat removal capability exceeds the core decay heat generation rate, if subcooling is maintained in the RCS, then the core remains covered at all times and fuel clad temperatures will remain low and no core damage will occur.

The usumptions used in the analysis cue pruented in the SSAR emphula RCS subcooling margin.

i Assumptions used in the analysis were chosen to minimize RCS subcoollng margin. The purpose of l

i using pressurizer spray and a lower pressurizer safety valve setpoint is to maintain the RCS pressure and saturation "mperature lower and minimize the subcooling margin.

d m

9

l li j ..

PRELIMINARY l

I l

Case 1 with proesuriter sprey aveliable

Case 2 without proesuriter sprey ovollable 2800 ,

2600 ~

O ~

r-e-

'" T 2400 2500 L

2000 = 3P

1800

~ \,\I I2 1600 * '

M . I a '

1400

1200 ~

1

' ' ' ' ' " ' ' ' ' ' ' " ' ' ' ' " " ' '\'"" ' ' ' ' ' ' "

2 3 4 5 6 10 10 to 10 10 to Iim8 (S)

Figure 201 Pressuizer Pressure Sensitivity During a Foodline Break 4

2.0 Fe.dwaier Sysiem eine area <SSAR is.:.i> PRELIMINARY Question 23 Figure 15.2.84 shows that the water level reaches the top of the pressurizer during a period of 6000 to 20000 swonds after an FLB event. Discuss the function of the pressurtzer safety valves (PSV) assumed in the analysis during and after the period when the pressurizer is filled with water. Justification should be provided if the PSVs are assumed to reclosed during or after the period when the pressurizer was filled with water.

Response

The version of Figure 15.2.84 which shows the pressurizer filling during a feedline break event is from an earlier version of the SSAR and is no longer correct. In the current feedline break analysis presented in Revision 13 of the SSAR, Figure 15.2.84 showa that the pressurizer does not fill and 1 there is no water tellef from the pressurizer safety valves.

S I

a 4

O e

9

tI

.. PRELIMINARY Partial Loss of Forced Reactor Coolant Flow (SSAR 15.3.1)

Question 24. The staff found that no discussion was provided for the events with loss of one or three reactor coolant pumps. A discussion of the results of the loss of one and three reactor coolant pumps events should be provided.

Response

Selection of the cases presented in SSAR Section 15.3 for loss of forced reactor coolant dow events is based on the reactor coolant pump motor electrical configuration and the configuration of the protection system functions used to mitigate these events.

The AP600 has two electrical busses to supply power to the RCPs. Each of the two busses supplies electrical power to two RCPs. De RCP: are connected to the busses, such that the two pumps sharing an electrical buss are from opposing RCS loops. With this electrical arrangement, the following loss of forced reactor coolant flow events can be postulated.

Case A -

One out of four RCPs coast down due to a RCP fault or a breaker fault Case B -

Two out of four RCPs coast down due to a buss fault, ne two RCPs coasting down are on opposing reactor coulant loops.

CaicC -

Four out of four RCPs coast down due to a complete loss of ac power to RCP busses.

With the electrical arrangement of the reactor coolant pumps, an event with loss of three of the four RCPs is not a credible event. Two independent faults would have to occur to cause a loss of three reactor coolant pumps. Therefore this event is not considered to be a realistic event and no analysis is presented.

Mitigation of loss of forced RCS flow events is accomplished by prompt tripping of the reactor before conditions in the RCS approach DNB limits, ne low reactor coolant flow and the low reactor coolant pump speed trips are used as the primary reactor trip functions for loss of forced RCS flow events. Each AP600 cold leg has four flow sensors. He reactor is tripped on low reactor coolant flow in any cold leg if two of four flow sensors in that cold leg measure flow below the setpoint.

Each of the reactor cc)lant pumps has one speed sensor. De reactor is tripped if the speed in any two reactor coolant pumps is below the low speed setpoint. The AP600 Technical Specification allow the plant to be operated with one PMS channel set or division out of service. Design basis accident analyses must also consider the possibility of a single failure of another PMS channel set or division.

Considering an inoperable PMS channel set or division and single failures, the low reactor coolant flow reactor trip function is used to detect only partial loss of flow events and the low reactor coolant pump speed reactor trip function is used to detect complete loss of flow events.

Analysis for Case C, the loss of four out of four reactor coolant pumps is mitigated by the low reactor coolant pump speed reactor trip. An analysis demonstrating that acceptable results are obtained is presented in Section 15.3.2 of the SSAR.

Cases A and B are partial loss of flow events and the low reactor coolant flow in any cold leg trip function is used for event mitigation. Both cases would trip at essentially the same time on low

PRELIMir4ARY 2.t. l

.. l reactor coolant now. Comparing Cues A and B, at the tims of reactor trip, the net core flow will be l much lower for the cue where two reactor coolant pumps are lost (Cue B) than for the cue where :

only one reactor coolant pump is lost (Cue A). Therefore, the results for an event with a loss of two reactor coolant pumps are more severe and bound those of an event where only one reactor coolant pump is lost.

Section 15.3.1 contains the results of an analysis where two reactor coolant pumps are lost. These I results bound those of a partial loss of flow event where only one reactor coolant pump is lost. The !

loss of three reactor coolant pumps requires multiple independent failures to occur, is not considered j credible and not presented in the SSAR. Section 15.3.2 contains the results of an analysis where all four reactor coolant pumps are assumed lost.

(i i

a 0- -%

t E

e e

4k 4

?

u

.. PREllMINAR'( -

L'ncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (SSAR 15 4 2) 29 The withdrawal at power analysis does not address compliance with the GDC17 requirements To satisfy GDC 17. the effects of a loss of offsite power on this event should be considered l

Response

Revision 13 to the SSAR includes a modified Section 15 4 2 that now addresses compliance with the GDC 17 requirements for the rod withdrawal at power event.

O e

9 e

9 ,

PRELIMINARY Boron Dilution (SSAR 15 4 6) 36 Discuss the technical bases for the charging Dow of 200 gpm assumed in the analysis Resoonse The Chemical and Volume Control System (CVS) has two centrifugal makeup pumps with flow control provided by a control valve located in the common discharge line from the pumps. A cavitating venturi located upstream of the makeup control valve, in the common discharge line, limits the makeup Dow and provides protection from excessive pump runout dows. It is this cavitating venturi that limits the charging flow to the maximum value of 200 gpm that is assumed in the boron dilution analysis, The actual design of the venturilimits .

- now to 175 gpm, so the analysis value is conservatively bounding.

1 d

4 O

4.

e d .

1 ..

'28 PRELIMINARY Boron Dilution (SSAR 15 4 6) 37 The value of the setpoint (60 percent increase /lomin) of the source range nuclear tiux instrumentation used in the analysis to isolate the demineralized water storage isolation valves is inconsistent with that specined in item 15 of Technical Speci0 cation Table 3 3 2 (60 percent increase /$0 min)

Resoonse i

The Technical Speci0 cation Table 3 3 2 value (60 percent increase /$0 min)is correct and the '

analysis has actually been performed using this value. The entry in the SSAR referring to "60 percent /10_ min" was inadvertently left in the text from a previous analysis. The SSAR-text has been modined to correctly reflect $0 minutes and that change has been included in Revision 14 to the SSAR.

4 4

e s * .

6 Y

e O

O 4

e

% 4

- PREl.lMIN AR Y Doron Dilution (SSAR 15 4 6) 38 Discuss the time it would take from the start of the dilution to the loss of shutdown margin and then to reach enticality for Modes 3 through 5. Discuss the magms (the times from the DWS isolation to core criticality) with consideration of uncertainties associated with the calculational methods, input parameters, signal processing time and salve closing time.

Besconse l To define the existing analysis margins for the AP600 boron dilution event during operating Modes 3,4. and 5. Table 381 provides the following results for the limiting cases reported in the SSAR (with the dilution starting at time zero)-

l Table 381 Operating Nominal Total Time to Alarm to. Post Alarm Mode Dilution Dilution Alarm Criticality System Flow Time to (minutes) (minutes) Response (gpm) Criticality (minutes) , .

(minutes) -

5 200 10 39 8 85 1.54 1.331 4 200 12.04 8.93 3.11 1.331 3 200 117.83 116 32 1.51 1.331 The " total dilution time to criticality" dennes how long aner the start of the dilution the core will return to criticality, if the dilution is not terminated. The " time to alarm" is the time after the start of the dilution when the microprocessor generates an alarm and the signal to initiate the automatic actions that will terminate the bcron dilution. The timing of the audible alarm, itself, is not actually a concem with the automatic mitigation system, since an operator -

action is not being relied upon to terminate the dilution. Rather, it is the timing of the signal to initiate the automatic protection functions that is critical to the system response. The time of " alarm to-criticality" defines how long after the initial microprocessor signal is generated the core is predicted to become critical, if the dilution continues uninternipted. The

" post alarm system response" time dennes how long aRer the microprocessor signal is generated that all the necessary operations to terminate the dilution have been completed.

In this analysis, the restilts are acceptable (i e. the ccre remains suberitical)if the post alarm system response time is less than the time of alarm to-criticality. As indicated in SSAR Section 15 4 6, the automatic protective actions generated for the analyzed cases terminate the dilution and maintain the plant in a suberitical condition, without operator action. Following these automatic prote'ction system actuations, the operator may take action to restore the

n _

PRELIMINARY 27 1e l Technical Speci0 cation shutdown margin.

l These results redect the assumed delay times denned in Table 38 2 Referring to Table 38 2, l the algorithm delay is a product of the sampling logic modeled for the AP600 boron dilution mitigation system. He associated 90 second delay is already included in the times :hown l

above for time to alarm and it is not included as part of the post alarm system response.

When added together, the delays listed as items 2 through 5 in Table 38 2 produce the post alarm system response.

As discussed in SSAR Section 15.4 6, the realignment of the CVS valves to terminate the dilution is a safety related function, while the realignment of pump suction to the boric acid ,

tank is a nonsafety related operation. The CVS pumps, themselves, are nonsafety related, so l their operation is not credited in the analysis. The analysis does consider the initial portion of this boration phase by treating it as a continuing dilution until any unborated water in the CVS lines is purged. The calculated purge time (44.88 seconds)is a function of the purge i volume and the dilution now rate.

A primary input assumption to the analysis is the maximum dilution now rate. As discussed in the response to Question 36, the CVS has two centrifugal makeup pumps with Gow control provided by a control valve located in the common discharge line from the pumps. A .

cavitating ventari, upstream of the makeup control valve, in the common discharge line, limits -

, the make'up Dow and provides protection from excessive pump runout flows. It is this cavitating venturi that limits the charging How to the maximum value of 200 gpm assumed in the boron dilution analysis. The actual design of the venturi limits now to 175 gpm, so the analysin value is conservatively bounding.

Another denning item for the boron dilution analysis is the Oux multiplication setpoint that is modeled. The Oux multiplication setpoint for the AP600 boron dilution analysis explicitly models substantial setpoint uncertainties in the process of calculating the response of the automatic boron dilution mitigation system. As defined in the AP600 Technical Speci0 cations (SSAR Section 16.1 Table 3.3.2-1) the nominal Oux multiplication setpoint is 1.60. This setpoint means that on a nominal basis, the beton dilution mitigation system will actuate when the measured neutron Dux has increased by 60% during ths specined sampling period (50 minutes). .However, the actual boron dilution cases reported in the SSAR for operating Modes 3,4, and 5 assume that the, automatic protection system response is delayed until a predicted flux increase that is substantially larger than the nominal 60% value.

For the' Mode 5 cases, the analysis models a 55% setpoint uncertainty, which means the automatic boron dilution mitigation system does not actuate until the measured Oux reaches 2 48 times (1.6 x 1.55) the reference value, over the specified sampling period. For Modes 4 and 3 the analysis models a 25.5% setpoint uncertainty, which means the automatic protection system response is dela ed until the the measured aux reaches 2.008 times (1.6 x 1.255) the reference value, over the specined sampling period, ne net effect of the setpoint uncertainties is to delay the actuation of the boron dilution mitigation system, thereby conservatively minintizing the time from automatic protection actuation to predicted criticality.

d .

r PRELIMINARY

. Table 38 2 Description of Delay Time (seconds) 1 Conservative fixed delay from the flux multiplication setpoint 90 until the boron dilution protection system microprocessor responds (algorithm delay: already included in automatic l protectien actuation time) l

2. Fixed delay time from the microprocessor to the to the mitigaion 10 actuation circuitry (signal delay: included in acceptance criterion) 3 Closure time for the Chemical & Volume Control System (CVS) 10 outlet isolation valves to the volume control tank (included in the acceptance criterion)

4. Opening time for the CVS isolation valve to the IRWST 15 (included in the acceptance criterion)

5. Purge time to eliminate unborated water from the CVS piping 44.88 , ,

, ,onnecting the IRWST to the RCS (incisded in the acceptance c

entenon) e 4

e k

4 e

4

W I Westinghouse ems FAX COVER SHEET RECIPIENT INFORMATION SENDER INFORMATION i

DATE: 7 A3-97 NAME: M$n khs TO: LOCATION:

C48 CA hu ENERGY CdNTER .

EAST PHONE: FACSIMILE: %i.q\g .3Og PHONE:

Office: 412 D4 4125 COMPANY: Facsimile: win: 284 4887 NkC outside: (412)374 4887 LOCATION:

Cover + Pages 1+) ,

l The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

Fw deJverv % Da\e. Tkcathec.

i Mk \hl .

1 h

9 e

l

- TX CONFl.RMATION REPORT ** AS OF JUL 22 '97 14:03 PAGE.01 APG00 DESIGN CERT 1

DATE TIME T0/FROM MODE MlH/SEC PGS STATUS 01 7/22 14:02 #23sNRC G3--S 01'01 02 OK l

=

0 4

4

Westinghouse FAX COVER SHEET D

RECIPIENT INFORMATION SENDER INFORMATION OATE: ~) . 1 0 4 ~? NAME; kNn l\)y(hg TO: LOCATION: ENERGY CENkER -

bOke YAbec exsr PHONE: FACSIMILE: PHONE: 0ffice; qip.3 y q _ q g gg COMPANY: Focsimile: win: 284 4867 NK outside: (412)374 4887 l LOCATION: hQ l

Cover + Pages 1+l '

The following pages are being sent from the Westinghouse Energy Center, East Yower, Monroeville, PA. If any problems occur during this transmission, please call:

WIN: 284 5125 (Janice) or Outside: (412)374 5125.

COMMENTS:

DG\e - krf 15 O1f %QQfSh?C M(0Arooki b w ar N ecq N N >c MSFS b stu\~ifv' 4

,J (

f 0 (. [6Th D1 DTC YO'S

. b bt bb (

I

-3 ,

I k_ b(

r t l$ e f ,0b k b' I

o.

, Nl M

./

e s

AP600 BASES 3.8.1 INSERT PAGE 3.8.4 Al if one of the Class IE DC electrical power subsystems is inoperable, the remaining Cissa 1E DC electrical power subsysk no have the capacity to support a safe shutdown and to mitigate all design basis accidents, based on conservative analysis.

Because of the passive system design and the use of fail safe components, the remaining Class 1E DC electrical power subsystems have the empncity to support a l snfo shutdown and to mitigate most DBAs following a subsequent worst case single failure. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is reasonable based on engineering judgement balancing the risks of operation with three DC subsystems against the risks of a forced shutdown Additionally, the Completion Timo reflects a reasonable time to assess plant status; attempt to repair or reptw, thus avoiding an unnecessary shutdown; and, if necessary, prepare and effect an orderly and safe shutdown.

[ The 0 hour0 days 0 hours 0 weeks 0 months Completion Time is also consistent with the time specified for restoration of one (of four) Engineered Safety Features Actuation Cabinet (ESPAC)

,t or Protection Logic Cabinet (PLC) division (LCO 3.3.2, ESFAS Instrumentation).

Depending on the nature of the DC electrical power subsystem inoperability, one supported division ofinstrumentation could be considered inoperable, Inoperability

, of an ESFAC or PLC is similar to loss of one DC electrical power subsystem. In both cases, actuation of the safety functions associated with one of the four subsystema/ divisions may no longer be available.

The installed spare battery bank and charger may be used to restore an inoperable Class 1E DC electrical power subsystem; however, all applicable Surveillances must be met by the spare equipment used, prior to declaring the subsystem OPEllABLE.

9 9

g j _ .iil j T^'

l('

}\ .

t 9

1

')

.' [D) WWestinghouse -

TAX COVER SHEET RECIPlENT INFORMATION SEHOER INFORMATION DATE: L / a1 /997 NAME:

I tv. (gg, TO:

, LOCATION:

Te A's Ny'o r.' ENERGY CENTER .

EAST l PHONE: FACSIMil.E: PHONE:

ome, q,2 u v n eo COMPANY:

Foc$imite: en; L4Jd/2c. pg,46'OI LOCATION: M$ide: (412)374 4087 l Cover + Pages 1+l The followlog Pages are being sent from the Wettinghouse Energy Center, East Tower, Monroeville. PA! If any probitme occur during this transm!ssion, please call:

WIN: 284 5125 (Janice) or cutside: (412)374 51?s,

--==-

COMMENTS:

I/3 ruA4C t.# _f#c L4 t 0 C ( Ot e' CMdd TR-N'1 3 9 6 7. Mc ,3- fgr gp r,1 e:. <m .vt u e e~r A . a s u e rnm feL .f- 7. ] es At e w ,,y J cat gu,,,,,w

) I l. Y Nits to o u $ e in W av e a r- Scu,1,,u , , - ng S Ts1,7 g e ,, , ,-

J./ & nt ?w kt) . .

(c : u cc.4 H (a, pg .- g

.h -- -

(y n .ul .

th>v Dd tu tuM I

,/uTCt* U f(~r% vi t W NI.

o s

i, 1 -

111gh Pressure Air Subsystem ne high prenure air subsystem consists of a high pressure air compressor with an integral air purification system, controls, and a receiver.

l De high pressure air subsystem is manually operated and may be loaded on an onsite standby 1-diesel generator. His subsystem supplies air to the main control room emergency habitability system, the generator breaker, and the fire fighting apparatus recharge station. He isolation vahes to these locations are normally closed and are opened on an as.needed basis to refill the specified equipment air storage reservoirs. The high pressure air subsystem is shown I

schematically in Figure 9.3.1 1 and major system components are described in Table 9.3.14.

9.3.1,2.2 Component Description

. Instrument Air Subsystem ne instrument air subsystem consists of two air compressor trains. Each compressnr train consists of a multistage, low pressure, rotary screw, air compressor package, a desiccant dryer with a prefilter and afterfilter, and an air receiver. Each compressor package includes an intake filter, rotary screw compressor elements, silencer, intercooler, aftercooler, moisture separators, bleed off cooler, oil cooler, oil reservoir, automatic load controls, relief valves, and

a. discharge air check valve. Each compressor train produces oil free air.

Two instrument air receivers function as storage devices for compressed air. The receivers continue to supply the instrument air subsystem following a loss of the instrument air compressors until the receiver pressure drops below system requirements. Each air receiver is equipped with an automatic condensate drain valve and a pressure relief valve.

Two air dryer assemblies are provided for the instrument air subsystem. Each dryer assembly consists of a desiccant filled, twin tower design. One tower may be used to dry air while the other tower goes through regeneration. When instrumentation senses a high dew point, the towers switch. The former operating tower then undergoes regeneration while the regenerated tower drys the instrument air.

Each dryer assembly includes a coalescing prefilter that removes oil aerosols and moisture droplets, as well as an afterfilter to remove desiccant dust.

, ~ % m s H.,a,& a,> sknk 2.pkn th orMM M 3 h} ("N{ '" *'O V

Table 9.3.12 provides design information for the main components associated with the instrument air subsystem.

Service Air Subsystem De service air subsystem consists of two air compressor trains. Each compressor train consists of a multistage, low pressure, rotary screw, air compressor package, and a desiccant dryer with a prefilter and aherfilter. A common air receiver is provided for the two trains.

Each compressor package includes an intake filter, rotary screw compressor elements, silencer, Revisiont 13 3 hstinghouse 9.3 3 May 30,1997

}

- TX CONFIRMATION REPORT ** AS OF JUL 24 '97 8:49 PAGE.01

. APG00 DESIGN CERT

. DRTE TIME T0/FROM MODE MlH/SEC PGS STATUS 01 7/24 00:47 301 504 2300 G3--S 01'20 02 OK I

l .

l O

e 4

4

^

n 4

July 24,1997 Marylee Slosson Ted Quay -

The attached table is a chapter by chapter breakdown of the Westinghouse open item status. The good news is that outside of the code V&V in chapter 6.2.15 and 21, there are only 56 Action W items for us to respond to. I expect that Revision 15 to the SSAR and the submittal of the $hort term availability controls will close about half of these. We are pushing on this end to get as many of these out in the very near term as possible, especially for those FSER sections that are July and August.

Submittal of the PXS PIRT/ Scaling report and the WGOTillC Applications repon will close from a Westinghouse perspective the vast majority of the open code V&V items.

Jim's last email to you on July 23 also attached. shows 474 items where the NRC has not acknowledged receipt of information Westinghouse has provided. Your attention to closing the gan between the Westinghouse and NRC status reports will be greatly appreciated.

Thanks!

l fy V /Y ,

l l

06MH AM wPF/ July 24.1997 l

6 f

AP600 OpEN ITEM STATUS July 24,1997 Chapter Confirm W Action W Total FSliR Date 1 0 2 16 September 2 0 0 11$ August 3 3 6 597 August 4 0 0 29 Complete 5 0 6 192 August 6 0 1 73 September 6.2 0 427 727 November j 7 1 1 30 August / September 8 0 0 43 August l

l 9 1 7 280 September / October i 10 2 80 1 September / October 11 0 0 83 August 12 0 1 28 July 13 0 1 43 September 14 0 3 137 September /l'BD 15 0 37 493 September 16 *

0 14 85 October / August 17 0 0 6 October 18 1 0 111 Complete 19 0 13 639 July / August / September 20 0 0 149 September / October 21 0 151 725 July / August / September / November Total 8 671 4681 4

4 4

I 0617H AM WPF/ July 24,1997 2

)

i

. Wintt% ,lanws W.[10L19fi 7/23M7, Wednesday Remindeto(ft lje_ms Ne

]

From: Winters, James W." <w interjw@westingbouse.com>

To: "'dco l @nre. gov'" < des t S ure. gov >, trq@ nte. gov'" <trq @ nte. gov >,

jms3 @nre. gov'" <jms3 @nre. gov >, "'wch@nre. gov'" <wch@nre gov >,

"'tjk2 @ nrc. gov'" <tjk2 @ nrc. gov >

Cc: "'dtj@nre gov'" <dtj@nrc. gov >, ** mms @nrc. gov * < mms @nre gov >,

"McIntyre, Brian A* <meintyba@wesmall.com>,

i "Cummins, Ed"

<cumminwe@ westinghouse.com>,

"Vijuk, Robert M." <vijukrm@wesmall.com>,

"Rarig. Bruce E." <rarigbe@w estinghouse.com>,

"Nydes Robin K."

<nydesrk@westingbouse.cor >,

"Tupper, Robert B " <tuppertb@wesmall.com>,

llaag, Cynthia L." <haagel@wesmall.com>,

"Lindgren, Donald A."

<lindglda@ westinghouse.com>,

"Fanto, Susan V."

<fantolsv@ westinghouse.com>,

" Winters, James W."

<winterjw@ westinghouse.com>

Subject:

Wednesday Reminder of items Needing NRC Acknowledgment Date: Wed,23 Jul 199710:19:29 0400 X Priority: 3

> The background for each of these items is outlined in the Open items

> Tracking System. ney ALL have been addressed by Westinghouse.

> NRC is requested to please acknowledae receipt of 13.rmation related D to each of the following Open items. The reviewer in each case should

> have a submittal from Westinghouse as identified in OITS for the item.

> Recognizing that reviewing for completeness of the response in each

> case constitutes an NRC action, we recommend that receipt

> acknowledgement be accompanied by direfiiin to chan[e~their "NRC o Status" to " Action N". If these are truly " Action W", please provide

~

. b a deRripii3E briheTittl6iiWestingh6uiereQected (6 Tall WeTnow

> of no action required, hiany of these are very (over 6 months) old;

a We are not asking for resolution or even NRC review at thTiTme,just

> acknowledgement that you have received the information as outlined in d the OITS Status Detail. If your investigation shows I goofed and

> didn't remove item numbers that I should have, please let me know. ,

t SCSB 972,973,984,988,1002,1007,1008,1009,1012,1458,1461,

> 1464,1467,1633,1638,1639,2415,2416,2418,2485,2487,24%,

> 2502, 2503, 2504, 2505, 2512, 2514, 27 I 8, 2883, 2884, 2885, 2887

> 2888, 2889, 3070, 3078, 3079, 3124, 3125, 3126, 3127, 3128, 3129, l 3 #'

> 3197, 3198, 3200, 3202, 3208, 3210, 321 1, 3215, 3419, 3422, 4162, d4186,4299,4300,4301,4302,4303,4304,4305,4306,4307,4308, a4542,4543,5054,5153,5154,5155,5156,$157,5158,5159,5160,

>$161,5162,5163.5164,5165,5166,5167,5168,5169,5237,5238,

~ PUnied]or f'_ Brian $hicIntyie'T<mcintiba_@ wdmAiEinn> .i _ ' TZ ' _. _ _ ' 1l A

e

> 5239,5241,5292,5293, $294,5295,5296,5297,5298, $299,5300,

> $ 30 l , 5392, 5303, 5304, 5305, 5306, 5307, 5308, 5309, 5310, 531 1

> $312,5313,5314, $315,5351, $352,5353,5354,5373,5493,5494,

> 5495, 5496, 5498, 5499, 5500, 5501, 5502, 5503, 5508, 551 1, 5512.

D $$13,5515,5516,5520, $521,5522, $523 NO items deleted and 2 mitems added since last repon.

D d SRXB 945,955,1260,1400,2258,2275,2300,2980,3140,3141,3172,

> 3173,3174,3175,3176,3177,3178,3179,3180,3181,3182,3184,

> 3381, 3384, 3385, 3390, 3961, 4146, 4150, 4187, 4465, 4468, 4471 p4472,4473,4475,4477,4478,4479,4482.4483,4484,4487,4488, d 4490, 4491, 4492, 4493, 4495, 4496, 4497, 4498, 4500, 4502, 4506, ?'4 o4508,4509,4510,4$11,4512,4513,4517,4521,4522,4524,4681, n4682,4689,4694,4973,5017,5018,5019,5020,5021,5022,5023,

> $024,5025,5026,5079,5080,5082,5083,5085,5093,5136,5137, d 5147, $148,5485. 3486,5487,5488,5489,5506,5507,5519, $537 NO

> items deleted and Sjitems added since last report.

> ECGil 628, 649,662,664,668,706. 708,745,750,751,766,767, -

> 768,769,772,1482,1485. I885,1888,2515,3268,3269,3432,3437,

> 4159, 4160, 5028, 5029, 5030, 5031, 5032, 5033, 5062, 5063, 5064, (, e r 5%5. 5%6, 5067, 5068, 5069, 5070, 5071, 5072, 5073, 5074, 5075, p5076,5150,5229,5230,5234,5235,$236,5242,5243,5244,5245, c $246,5286,5287,'5288,5289,5290,5291, $525 NO items deleted and t 3 items added since last repon z SPSB 1406,1423,1425,1432,1447,1448,1450,1452,2795,2939 D2942,2945,2958,2959,3007,3009,3943,4183.4I85,5014,5015, 7'

> 5016,5119,5120,5129,5130,5131,5134,5497, $504,5505 NO items p deleted and 1 item added since last report.

> S PLB 174, 224, 233, 253, 264, 275, 276, 281, 282, 283, 286, 289, p292,293,302,304,338,372,1024,1025,1090,1103,i104,i105,

> 1107, i108, i109, i110,1112,1142,1171,1766,2032,2419,2892,

> 2893,2894,2895,28%,3053,3085,3086,3087,3122,3424,3482, d 3%7,4170,4195,5004,5346,5'358,5359 NO items deleted and 27

> ltems added since last tepon.

z

> TSB 1278,1279,1280,1281,1282,1283,1284,1285,1970,2074, t2351,2352,2353,2354,2355,3054,4189,4224,4225,4226,4227, M

> 4971,4972,5250,5526,5529 NO items deleted and I item added since

. b last report.

D

> li!CB 2024, 2434, 2454, 5095, 5096, 5097, 5098, 5099, 5100, 5101, iL

> $102,5103 NO items deleted and 3 items added since last repon.

> No Branch Identified 4145,4151,4182,4198,4615 2 items deleted #

> and NO items added since last repon.

y z Branch "Unidentined" . $109,5110,5111 NO items deleted and NO - 3 Printed fMBil' a nMicI'ntyre'j <mdni[ba@heimali.'cinf ' ZZ

~(('\ f2]

_ /

e t

> iterns added since last report.

p EMEli 608,798,800,801,805,807,814 NO items deleted and NO 7

> items added since last report. ,

p peril - 1023,2387,3475,3476,3477,3478,3479,3480,3520,3521, /6

> 3522,3962,3963,3964,5106, $107 NO items deleted and 6 items p added $1nce last report.

p PDST 501. 502,503,504,505,1763 NO items deleted and 1 item 6

> added since last report.

p lillFil - 1525,3941,5247,5528 NO items deleted and NO items added y

> since last report.

p ADT 459,1991 NO items deleted and NO items added since last L d report.

D

> llQMll 4852,5104, $10$ NO items deleted and I itern added since last J r report.

l

> PEPI) 1019 NO items d leted and NO iterns added since last report. /

z

> RVill 5510 Nd items deleted and NO items added since last report. /

r EELil 5108 NO items deleted and No items added since last repon. /

l >

I

> Thanka

> Jim Winters #'

l > 412 374 5290 0/77 b

D D

b l

~

^~

Printed for "llrian'd,LMcIntyre'[<nicirit}tIa'Gi~vicsmall.c_o_mE[ ~ jl

NSD-NRC-97-5308, Forwards Correspondence Previously Sent Informally Over Period of 970605-970724

Text

SUBJECT:

Subject:

Subject:

Response

SUBJECT:

Dear Mr. Quay:

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Subject:

Response

Response

Response

Response

Response

Response

Response

Subject:

Subject:

Dear Joe,

SUBJECT:

Dear Mr. Liparulo:

= = == = , = = =

Subject:

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Subject:

Navigation menu

Search