ML20198G600

Summary of 750313 Meeting W/Util,United Engineers & Constructors & B&W in Bethesda,Md to Discuss Proposed Applicant Responses to Several Issues Identified as Technical Areas Where Util Did Not Meet Requirements

ML20198G600
Person / Time
Site:	Washington Public Power Supply System
Issue date:	04/15/1975
From:	Cox T Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
CON-WNP-1059 NUDOCS 8605290693
Download: ML20198G600 (59)

Text

T ,

t

,,4 7 , .., .

! 9

. A?P. 1 5 ' E Docket Nos: 50-460 l and 50-513

! )

l APPLICANT: 0 WER SUPPLY SYSTEM (WPPSS)

WASHINGTON PUBLIC P' PACILITY : WASHINGTON NUCLEAR ONE AND POUR (WNP-1,4)

SIRMART OF MEETING TO DISCUSS OUTSTANDING ISSUES On March 13, 1975, representatives of WPPSS, United Engineers and Constructors (UE&C), Babcock & Wilcox (B&W) and the NRC staff met in Bethesda, Maryland.

The meeting was held at WPPSS' request to discuss peoposed applicant responses to several issues which had been identified by the staff as technical areas in which WPPSS did not meet staff requirements. An attendance list is attached (Attachment 1).

A sumsary of the meating discussion is as follows:

i

1. Containment Spray System - In response to the staff conclusion that the spray system design included through PSAR Amendment 15 was inadequate, WPPSS presented preliminary flow disgrams and calculations describing a revised i

design. The proposed design, shown in the applicant's handout (Attachment 2) uses eductors to withdraw NaOH from a storage tank into the containment spray system piping. The design v6uld be similar to that reviewed and approved on the Trojan docket. W. Pasedag suggested that a standard chemical industry reference work be used for the titration curve of NaOH in water rather than the article by Callagher in Nuclear Technology. Preliminary staff evaluation of the proposed design was that it was an acceptable approach. W. Pasedag also pointed out that the PSAR amendment to be submitted should include

.s

1) curves showing minimun and maximum pH in the sump as a function of time from ECCS initiation, to be predicated on the various faulted conditions that could obtain, and

2) an updated, restatement of previous WPPSS committents concerning CSS preoperational testing.

W. Pasedag said that the system proposed appeared to merit an assumed elemental iodine removal coefficient of 10.0 hr-1 A#

J i

e,e.s e

• 8605290693 75041S evaaa=e> PDR ADOCK 00000460 4 PDR -

l

\ v

.ars e. I

• v. s. eovs=====r naarme e,,ie s. ..u.a.in Aac.ns ta Mn Ascu ouo

e

? ,

2. Offsite Doses - The staff had calculated doses in excess of Regulatory Guide 1.4 guidelines even with an artificially assumed effective containment

' spray system. A UE&C presentation discussed a proposed unsprayed containment I volume of 11% and a reduction in the containment leak rate technical specification

to 0.15%. L. Soffer of the NRC staff agreed that those changes should resolve

, the dose issue. A staff meteorologist, R. Kornasiewicz, briefly discussed

! - the staff treatment of applicant:s site data in. developing I/Q values for offsite

dose calculations. He said that the calculations outlined in the Standard Review Plan are u_ sed,rather than the Regulatory Guide 1.4 curves. The Regulatory i Guide [_ curves _aro, now considered too general and inappliesble to many western sites. A planned future revision to Regulatory Guide 1.4 will delete the

, curves. For the WPPSS application, the staff used the Hanford Mateorological Station (IDfS) data from the 200 foot level and modified those data logarithmically for the 33 foot level. The six: month data from the WNP-2 tower was reviewed to verify the conservatism inherent in the HMS data. The six month data was not used to calculate the X/Q's used to compute offsite doses. When full year data

> is supplied by WPPSS, the conservatism of HMS data will again be checked.

3. Component Leakage in ESF Area - Offsite doses as a result of ECCS f component leakage outside containment, post-LOCA, were discussed. The staff considered approximately 100 gallons total leakage at temperatures that would cause radioactive iodine release to the atmosphere such that Regulatory Guide 1.4 limits would be exceeded. This occured only because no credit was taken for iodine removal filters in the ESF area. WFPSS stated that a revised analysis would be submitted assmnf ng a 6.5 gpm leak for 15 minutes, with charcoal filters providing 95% removal of elemental iodine. Staff site analyst, L. Soffar, indicated that this approach should resolve this problem.

4. Control Room Habitability - WFPSS presented sketches showing the

! locations of an additional secondary (remote) air intake for control room l emergency air supply (Attachment 3). K. Murphy of the staff stated that the i

control room doses would be acceptable with the intake arrangement proposed.

It was agreed that WPPSS would include drawings of the revised remote intake j locations and some detail of the intake structures in PSAR hendment No.16.

Duct valving sud controls to assure system function even in the event of single active failures would also be addressed in the amendment. In discussing individual portable breathing apparatus, K. Murphy said that WPPSS should consider resupply of individual apparatus from air storage tanks stored near the control room. He also stated that acceptable breathing apparatus must provide a positive pressure within the face mask.

' rrees p o

eumMaasa > ,,,,,,,,,,

4 satt > ._

Foran AEC.318 (Rev. 7 33) AICM 0240 1lr v. e. oovenMusNt reentlN4 OPPICal 9974e588 000

i  !

I ,

! 5. Containment Pressure Analysis Long-Term Blowdown - Staf f analysts

[ with the containment Systems Branch had presented two questions to the applicant i as a result of review of recent applicant submittals. These questions, trans-j mitted to WPPSS prior to the meeting, are shown as questions 6.1-and 6.2 in i j Attachment 4 of this summary.

1 In response to question 6.1(1), B. Dunn (B&W) described the PRIT l code and its utilisation in conjunction with the CRAFT code to model mass

! and energy release from the reactor eeolant system in a LOCA (Attachments 5 and 6). The staff'aontainment systems reviewer, T. Greene, requested that l the anterial presented be incorporated into the PSAR and WPPSS agreed to -do this.

l J. Daniel (UE&C) addrnssed 6.1(2), explaining that the change in I B&W supplied data after 3500 seconds was for the purpose of conservatively l i

l calculating the sump temperature. UE&C considered this acceptable from

! the containment pressure . standpoint since the pressure peak is calculated i

before 3500 seconds. The staff stated that containment pressufe should be

{ conservatively calculated for the fuli 24 hourutime period. It was agreed ~ ~ ~

that the PSAR would be revised to include unmodified B&W blowdown data and '

j to include contai -ant pressure response curves calculated using the

! unmodified data.

I V. Mani (UE&C) addressed 6.1(3), explaining that recirculation .

time was determined for two limiting conditions of injection and spray

,  : flows. The selected 3500 second recirculation time was based on minimum i

ECCS flow and a 480.000 gallon useable BWST volume.(Attachinant 7). The

! actual BWST capacity of 680,000 useable gallons provides for a much longer injection period using relatively low temperature BWST water, which UE&C >

felt would significantly reduce the actual containment pressure and sump temperature. After lengthy discussion, it was agreed that providing the

pressure analysis with unmodified blowdown dsta (discussed above), the

! 3500 see recirculation time and additional definition of what failure assumptions were used to set the limiting flow conditions, should result in a design description complete enough for staff review.

t t

6. Subcompartment Differential Pressure Analysis - K. Niyogi (UE&C)

! discussed the results of new subcompartment analyses (Attachment 8) in response to staff questions 6.2(1), (2) (Attachment 4). The staff noted that these newer analyses results appeared to be similar to those calculated by the staff, although there is still some uncertainty about the steam generator subcompartment. Staff review of the applicant's latest submittal will continue. Mr. Niyogi also addressed UE&C's handling of vent loss i co_efficients_and friction factors _(Attachment 9)... Errors in the presented

!1 oss coefficients _were discussed._ It was agreed that the forthcoming

] PSAR Amendment 16 would show corrected tables 6.2-45 through 6.2-47 and would include some description in the text concerning how the coefficients vero conservatively e n l evi t n en tf .

r ernes> _ ___

evnaame >

oats > - -.

Fesus Aac St3 (Re., p.3M AaCM 0240 1lt v. e. eeve n=nesse enmeme errocas seve-eae.ies  ;

1

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

-m _

e t

l i .

7. combustible cas control - V. Mani presented the results of some recent work by UE&C on the potential for hydrogen gas accumulation in

the containment subcompartments. They have identified two subcompartments

! in which they plan to include Seismic Category 1 fans and ducting to assura l dilution and =iving of air in those compartments.

l 8. Decay Heat Removal System - Although not originally scheduled for this meeting, WPPSS asked the staff to discuss the requirement for remote manual operation of the DHR valves located insnediately outside containment i (V31A and V328). This requirement was first discussed with WPPSS in a 1 telephone conference on 1/22/75. V. Stello, Assistant Director for Reactor

! Safety, spoke to the group and explained the staff position on this issue.

He stated that staff interpretation of GDC 19 has led to the requirement i

that the DHRS design shall provide controls for the initiation and mainte-i nance of normal-shutdown cooling (either hot or cold shutdown) from the control room. The system must be capable of this function even in the event of a single activa failure of fluid components or any single active or passive failure of electrical components.

i t

f5f$

T. H. Cox, Project Manager i Light Water Reactors Branch 2-3 Division of Reactor Licensing Attachsents:

1. Attendance List

2. Containment Spray System

3. WNP-1,4 Plot Plans

4. Request for Additional Information

5. PRIT Provisions i 6. PRIT Description

7. Response to Staff Request 6'41(3)

8. Subcompartment Analyses

, 9. Estimation of Loss Coefficient

., , , , , , , x7886/ LWR 2-3

. ...m... TE -

re ,. 4//5775 _

,g Form ABC.)l3 (Rev. F$)) ABCM 0240 W u. o. eava.nusmv paintime opricas esta.ese see

7 1

6-

, , e ,

ITTACHMENT 1 ATTENDANCE LIST MEETING WITH WPPSS AND CONTRACTORS MARCH 13. 1975 NRC

• V. Stello

• T. Novak T. Cox

• L. Soffer

• W. Pasedag

• W. Jensen

• T. Greene

• J. Shapaker

• T. Su IT&C H. Jenkens K. Niyogi A. Friedman J. Daniel S. Moss C. Thornes I. Sargent WPPSS A. Hosler .

B&W J. Happell B. Dunn K. Suhrke

• Denotes part-time attendance o

i a

s 4

5 A?R 15 '475 Docket Noe: 50-460 l and 50-313 I

APPLIC&lff WASHINGTON FUBLIC 70ER SUPPLY SYSTEM (WPPSS) i PACILITY : WASHINGTON NUCLEAR ONE AND FOUE (WNP-1,4)

! St39fARY OF ISETING TO DISCUSS OUTSTANDING ISSUES i

on March 13, 1975, representatives of WPPSS, United Engineers and Constructors (UF.4C) Babcock & Wilcox (B&W) and the NRC staff met in Bethesda, Maryland.

The meeting was held at WPPSS' request to discu's e psoposed applicant respouses to several issu'ee which had been identified by the staff as technical areas in which WPPSS did not meet staff requirements. An attendance list is l sttached (Atcachment 1).

i

A summary of the meeting discussion is as follous

)

i 1. gatainment Sprar Systen - In response to the staff conclusion that the spray system, design included through PSAR Amendment 15 was inadequate.

WPPSS presented preliminary flow diagrams and calculations describing a revised l design. The proposed design, shown in the applicant's handout (Attachment 2) uses eductors to withdraw Na0R from a storage tank into the containment spray system piping. The design would be similar to that reviewed and approved on l the Trojan docket. W. Pasedag suggested that a standard chemical industry reference work be used for the titration curve of NaOH in water rather t.han j the article by Gallagher in Nuclear Technology. Preliminary staff evaluation j of the proposed design was that it was sa acceptable approach. W. Pasedag also pointed out that the PSAR amendment to be submitted should includet

, 1) curves showing minimum and maximum pH in the sump as a function I

j of time from ECCS initiation, to be predicated on the various i faulted conditions that could obtain, and

2) an updated, restatement of previous WPPS$ constitments concerning CSS preoperational testing.

i W. Pasedag said that the system proposed appeared to merit an assumed elemental iodine removal coefficient of 10.0 hr-1 o

' petee p j .wename > __ . . , , _ ,_

eats > _ .. -

Pesas ABC.)ls (see. 9 SH ASCM 0240 $ v. s. eevennuent amistine eeriese e,te.see e.e j

I

l

- I i

2. _Offsite Doses - The staff had calculated doses in excess of Asgutstory Guide 1.4 suidelines even with an artificially assumed effective containment

, spray system. A UE&C presentation discussed a proposed unsprayed containment l volume of 112 and a reducties in the containment leak rate technical specification to 0.13%. L. Seffor of the NRC staff agreed that those changes should resolve

! the dose issue. A staff ascesrologist,1. Kornamiewies, briefly discussed l

the staff treatmoet of applicants site data in devoleping R/Q values for offaite i dose calculations. 'He said that the calculations outlined in the Standard Review Plan are used rather than the hogulatory Guide 1.4 curves. The Regulatory Guide ;~clic rei are~ Tow considered too general and Anapplicable te many western sites. A~ planned' future revision to Regulatory Guide 1.4 will delete the curves. yer the WFPSS application, tha staff us,d the Manford Metectological Staties (HMS) . data from the 200 foot level and imodified those data logarithmically i for the 33 foot level. The siz_asnth data from the FNY-2 tower was reviewed to

' verify the conservatise inhereer in the HMS data. 'the six mouth data was not used to calculate the X/Q's used to compute offsite doses, When full year data is supplied by VPPS3, the cosmervatism of IDai data will agaia he checked.

3. c --:rt Leakaas in F.3F Area - Offeite doses as a ree. alt of ECCS i component leakage outside containment, post-LOCA, were discussed. The staff

' considered approrimetely 100 hallons total leakage at tamperatures that would

eause radioactive iodine release to, the staosphere such tt.at Rag 21 story Guide 1.411anies wouli t be exceeded. This occured caly because no credit was taken for iodine removal filters in the RSF area. WPPFS stated that a revised analysis would be submitted assuming a 6.5 gym laak for 13 minutes, with charcoal filters providing 95% removal of elemental iodine. Staff site analyst, L. Soffer, indicated that this approach should resolve this problem.

4. _Copt,r A 2com Habitabil h - WPPSS prosegted sketches showing the locations of an additional asocondary (remote) air intake for control, room i emergency air supply (Attachneet 3). K. Murphy of the .= toff stated that the i i control room doses would be acceptable with the intake arrsagement proposed.

. It was agreed that WPPS3 would incivde drawings of the revised remote int'ake i

locaticas and some detail of the intake sf.ructures in PSAR Aniandment No.16. ,

l Duct volving and controls to assure system functit.n even in the event of single

(

active failures would also be addressed in the amendment. In discussing

! individual portable breathing apparatus, K. Murphy said that WPPSS should l consider resupply of individual app.oratus from air storage tanks stored near l

the control room. He also stated that acceptable breathing apparatus must i provide a positive pressure within the fac.e mask..

\

I l .

• = =a a = a > . _ _ ._.- _ __

em * -. - ._ _.- .._._ __ _

r Asc.ne as .s m Ascu oua h...........n...n.....i.,.....

l ,

l r l

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

n 3-t l -

5. Containment Pressure Analysis Long-Term Blowdown - Staff analysts with the Containment Systems Branch had presented two questions to the applicant as a result of review of recent applicant submittals. These questions, trans-mitted to WPPSS prior to the meeting, are shown as questions 6.1 and 6.2 in I

Attachment 4 of this summary.

In response to question 6.1(1), B. Dunn (B&W) described the PRIT l code and its utilisation in conjunction with the CRAFT code to model mass and energy release from the reactor coolant system in a LOCA (Attachments 5 and 6). The staff' containment systems reviewer. T. Creene, requested that l the material presented be incorporated into the PSAR and WPPSS agreed to do this.

J. Danial (UE&C) addressed 6.1(2), explaining that the change it.

B&W supplied data after 3500 seconds was for the purpose of conservatively .

calculating the sump temperature. UE&C considered this acceptable from the containment pressure standpoint since the pressure peak, is calculated before 3500 seconds. The staff stated that containment presstife M6fdNe -

conservatively calculated for the full 24 hourntime period. It was agreed -

that the PSAR would be revised to include unmodified B&W blowdown data and to include containment pressure response curves esiculated using the unmodified data.

V. Mani (UE&C) addressed 6.1(3), explaining that recirculation time was determined for two limiting conditions of injection and spray flows. The selected 3500 second recirculation time was based on minimum ECC3 f. low and a 480,000 gallon useable BWST volume (Attachment 7). We actual BW5I capacity of 680,000 useable gallons provides for a much longer injection period using relatively low temperature BWST water, which UE&C felt would significantly reduce the actual containment pressure and sump temperature. Af ter lengthy discussion, it was agreed that providing the pressure analysis with unmodified blowdown data (discussed above), the 3500 see racirculation time and additional definition of what failure assumptions were used to set the limiting flow conditions, should result in a design description complete enough for staff review.

6. subcompartment Differencial Pressure Analysis - K. Niyogi (UE&C) discussed the results of new subcompartment analyses (Attachment 8) in response to staff questions 6.2(1), (2) (Attachment 4). n o staff noted that t!iese newer analyses results appeared to be sinilar to those calculated by the staff, althouah there is still some uncertainty about the steam generator subcompartuent. Staf f review of the applicant's latest submittal will continue. Mr. !!1 yogi also addressed UESC's handlin; of vent losa coefficients and, friction factors. (Attachment.9)._ Errors .in the presented

' ions coef ficients were discussed. It was agreed-that the forthcoming PSAR Anendment 16 would show corrected tables 6.2-45 through 6.2-47 and

~

would include soen description in the text concerning how the coefficients wu comer meiwi f caeiimend.

errie s em ,,. . , , , _ _

eusaaase >

• mate >

Forms ABGils ( A.v. 9 5H AECM UJ40 TT u. s. eevennesamt reimtine opriese to,a.ese. tee

i l

4-l 7. Combustible cas control - V. Mani presented the results of mone

recent work by UE&C on the potential for hydrogen gas accunulation in i j the containment subcompartments. They have identified two subcompartments

• l in which they plan to include Seismic Category 1 fans and ducting to assure ,

l dilution and miring of air in those compartmenta.

! 8. Decay Heat Removal System - Although not originally scheduled for this meeting, WPPSS asked the staff to discuss the requirement for ramote a i manual operation of the DER valves located immediately outside containment (V31A and V325). This requirement was first discussed with WFPSS in a  !

telephone conference on 1/22/75. V. Stallo, Assistant Director for Raactor i i Safety, spoke to the group and explained the staff position on this issue.

He stated that staff interpretation of CDC 19 has led to the requirement that the DHR3 design shall provide controls for the initiation and mainte- ,

nance of normal shutdown cooling (either hot or cold shutdown) feczt the control room. The system must be capable of this function even in the event of a single activa failure of fluid components or any single active or passive failure of electrical components.

I

, oth,u.M S:ma bi T. H. Cox, Project Manager Light Water Reactors Branch 2-3 Division of Reactor Licensing Atcachments:

1. Attendance List

2. Containment Spray System

3. WNP-1.4 Plot Plans .

4. Request for Additional Information

5. PRIT Provisions

6. PRIT Description 7, Response to Staff Request 6;1(3)

8. Subcompartment Analyses

9. Estimation of Loss Coefficient

" " "

• x2886/LMR2 J.

• waam

• TCaxerm _._

eats

• 4/ 175 Fore ABC488 (Rev.f.5)) ABCM 0240 # u. s. esvammessuf pasurme oppiase esta.ese.e.e

l 6 ATTACHMENT 1 ATTENDANCE LIST MEETING WITH rPSS AND CONTRACTORS MARCH 13. 1975 NRC

• V. Stallo

• T. Novak T. Cox

• L. Soffer

• W. Pasedag

• W. Jensen

• T. Greene

• J. Shepalter

• T. Su UE&C H. Jenkens K. Niyogi A. Friedman J. Daniel S. Moss G. Thornes I. Sargent WPPSS A. Hosler '

B&W J. Happell B. Dunn K. Suhrke

• Denotes part-time attendance e

r

.A racwasr 2 UWT-0142 March 11, 1975 Telecopy CONTAINMENT SPRAY SYSTEM g g g .75 Y. MAffl

SUMMARY

The redesigned CSS system features relatively constant spray pH during the injection phase, total independence of spray pII during injection, and a controlled pH within an acceptable range during recirculation, regardless of postulated casualties to related systems. Additionally, final sump pH is within an optimum range, regardless of transients and casualties to related systems, assuming that the entire contents of the NaOH tank are added to the sump, DESIGN FEATURES -

The CSS system consists cf one (1) NaOH tank, two (2) Eductor / Pump Systems and associated piping with the following characteristics. .

NaOH Tank (1) Contents: 20 v/o NaOH in solution with demineralized water. An inert gas blanket is used to preclude solution decomposition.

(2) Capacity (useable): 7500 gal.

(3) Safety Features: Redundant Icw level alarms, redundant isolation valves (parallel) for injection initiation, and isolation valves (NO series) for terminating NaOH addition at chemical additive exhaustion. (See flow diagram.)

(4) Construction Features:

Stainless Stect Safety Class 2, WPPSS Quality Class I Vacuum Breaker, Relief Valve Pump / Eductor (2)

• Total pump capacity at design head: 3000 gpm Spray flow: 2700 gpm Eductor bypass flow: 300 gpm Spray additive addition rate: 30 gpm Safety Class 2 WPPSS Quality Class I Each 100% train redundant and supplied from separate emergency diesel

• generator channel.

PH Values (calculated)

==

• UWT-0142 e,

'N Final Sump pH: 9.0 - 9.25 (Variable function of initial RCS Boron concentration) 1 Injection' Spray pH: 9.3 Recirculation Spray pH: 9.3 - 10.0 3

E*.

(Total range under all normal and faulted conditions, actual values under any specific conditions will be a function of: (1) initial RCS boron concentration, (2) failure of related components - HPI or LPI or CSS pumps and (3) time af ter start of recirculation.)

E51175 V. MAm i

o e

?

I ,

l i

t

4 , . .,

t l Il, 'I l Ii . i ,; i; Ii :,.

i l I ii  : i. l . i-e l l! i !t- . i lI . .

.l...i!  ;

',' I. .  ; !

e  !

8 '

. , t, . a o ii. .

. . . i'!

I . i i ,. i i . .

I

. , , i , , ,i ,. , , j i

x: ! . . i i i j i , , i ,i.. .

l' li I, i l Jo l i. .26.!. .l ,l  !, i l 'i ij'li.j, i . .i I i I .

F F-

.c! li' ' I 1 8 I i i t t i i j j g 0! -

! -g.i l .. . . .l...

! 'l . ' ..'..I[. i '_  ; .'.,. ,

.fP. ;1l . i . ' ! ! ! I

e. ! .f4 u

. . . . i t i i  !

'. .. , . i

. I .. , ..

D

b. l l 8o :k i (L A

i .

('  ! .! .

.I [I .i j l .

! li n

lu

. i

' ' *  ! 'O g' j ' 'i

! I

' I i i

!j .  :

i 1

I i- -l-  :- - l- ! -[ i;/! il t

> $5 [ ~2:' Il ! !I

'o ii.i  ;

il 7" r-i-f b.

.I 1 _1 l - f: ^ll_

i 4

~T hJ1i. *l' !!! ii M . .-L! .- L. .b-_

(J) .

i

~-

l: [. ,

05 ,

• A fE  ; oli 1

0 '

i b -

.g o . .

S 1

l l

! go y- .' .

p. . ... ,

i i , i i  ! ,, i . .

m o q i. ._. :_ . ' ..:._: . . .

y O  :

u  ;; 1y .. ,

u. i 1 -

.3 .

._ i V) ,

g 3 u .o -

, _ _ . . . . i .. . _ .. _ . . . . . . . . .

.I.__. .

8 i.

2 . . ... . _

, . i i

.;.p....

i  :

} -- } _ .l. -. . . . .

. . , .~. W. &

i

. t m

....... o . . .: . .. .... ..a....... ..

. W .o . .. . .

i

.? .' ,

' I ' '

2  ! . I

.p l .

g

.l g lf { -- T $ , ,. . ..

, ; ) 7, i. , .7 ( -  ;-

. . . . -. . . _ . . _ Q .

S. ,

4 i ,. ..4 ...

.. .i. .. . h . . . . . . .

!  ! I

.i . ih 9

.! .i!_[..f..!._I.!,..[...I-!, .f 9

T

.' .'i .

,l .

i e

\

g -- , i. ..

i  ! i ,

-t-

. ..O y .

A. . j _; .m . . g . . . _. ...q l

, .. i . .

. . , ., .. ._7, . . . . . - -

. - - ; . , j . . ..I ._.7..  ;

gg g .

....,.7 . ..

_. . lp .. v. .

r-- . ,

2 ix . . . .

,o . i.1..._. t .o .. , .

.7p/

27 sc4-e.-

45'o l--df o

.4 d. -

. ;i...-

.L . ; .; .

a _.'  ;

0 . .a_..._..

.I O M' - - - - -- _ _ , _

[/hop p.A- ', ho t  : -

i yr , . . , .

y r . .

. . ,t

{y

' ' ' ._3 . '

w LL <-  ; g ' Q g., ,3

'~ ~ ~ ~ L 4

-oitz y  :  ; , . ~ ~ ~ ~ ~ ' ~ ~ ' ~ ~ ~ " ~ '

2' \6 1- '

j ..! ; j . ! .. ' . i .

\[')  ; J ._ L _i.- __!...'..i._ _.;

.J V, t i a . 7. _ .. -  : i

. m . . t .,

ni ... __ }>,; _ t__[ _a ..i. i

.o. O

. __ j q.- r.1- I i 1---r.-

. . ' r i .

. :.. . .. . t . . .,; . . t . d, g

{.__.J ..'. '

g i

4-..j_.. -_ ._.4- . .- - F -- g gy*

I :s I . ._ . . .

g g! .. . .j_ ' _;

pp i

- j . . } ._.. . _ . . . . . . .., ._. _ .

... . g__i J'l .

i I

I liIi i i I l

yn

-e .

.s .

l gg g

t, . .

_.__ % ga

__ p . _ . _ __.

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

4 i ~.

/  :

'. @46 = a-s-

a is i .e n

x 8 8%c_v az3- l t, e

. . _ .g . _ _ _ _ _ _ _ _ . . _ . _ . _ _ _ _ _ .

R .;.

4 4r; g 16 i I * \

sr

i.

• a lI .

g ,l $

/

l

y.. n .

, q

~

l o

/ g. .\, _ . ..:' [

m v -

3

-l

\ T- i.

4 t

i t .

i a ,,

Ok,g; gy[

. . _ _ _ _ . /(. ,,

3 a ,S i O N _..' ,

i  :

' )

p \ -'~l{

'  !  !  !! I j ijj i;

\

,c

. [ i!!lil,l  :

>~

e3 u

s e _ '.:-;-

1 s Y, 5 sd< a>

._P] o.E4 4pr.

1.;j .-. -

i.a .

t h i x,,4 l g8-

/ -- . - "1 - ,=!

N h .; . . -

f.s WJ I .9 oE{

( 'I /1 1 i LJ .i 's j.4 .

9i'

=

I8 y lll iI' 5W 1 ! 2eg:

at i.'  : I ill$-3  !

3L M_l$  !.

. . b jil'_,i... . _. _.2' 5

(1 H

,- _ W,1 _ y/ Hi'ih? - . _ . .

'=.._,

~ ~ '

(; j/ j j

~I c(j a  :

- s -  : l t

W 1 I 3 1 I

. _ _ _ _ _ . . . . . _ . . . ._ m. ,

I I i

3 . .c I I

s

. . _ _ . _ _ . _ . .i _. _ _ . . . . _ . . . .. .__._ . I , ,,

1

), r

.l 1

i r 1 .

Nu AcHMENr 3 .

9..

.' Vi

. y~

• Eb a wo  ;*

. .u_ = . . .

, ._.2.. .

gg ,

. MYE pg l. '

J 16i g) $ C

. I

' }P 7. ,c  : 'h M9  ! 5 -

l to it. 37

- l l -

l N= **

r

! I D h

.i g,

. If

's 5' I d s 3 h .. . ...

L

-). >

E i f g ,i

, s ,

3 1 ,/ '

. y I

.l pg g

• , \

gg

" x _t

.l I

T 'i i

il r

e!! .e

[, ih .I q

- d. ][*., $a 3$$

;j

< - =

".G.

l ( 4,,- }

q ( 2 ]

> gr i

f % ' N . W . 1,.

F I h K -____

' 'gge

( .

1 i

, e. j 6

/1 5

c,, L.J. j.'

1 q i ',- 6 m ti

=

i .

i: a h t irt

. a l-I ll .I ,,. '

' ' - E

.i\

't i

3 I i j

i i i

1 . t

i 1 I "-

i 1

[

, .i

- l- D

.. . . . - . . . _ ..!. m ,

~' '

I l 1 1 I I E I I I I 4

~ _ . - _

s [

t A77MCHfdG W 4 REQUEST FOR ADDITIONAL INFORMATION UPPSS 1 & 4 DOCKET NOS. 50-460 & 50-513 6.1 The following requests are related to the containment pressure analysis:

(1) For the containment pressure analysis the CRAFT computer code was used to calculate the mass and energy release rate to the containment from the time of rupture to about 1000 seconds.

At this time, another computer program was used to predict the mass and energy rates to the containment. Provide a detailed description of this~ computer code and the assumptions regarding decay energy and steam quenching. Discuss the applicability of your method to both hot and cold leg breaks, and for both partial and full ECCS operations.

(2) At approximately 3500 seconds, you have modified the mass and energy release rates to the containment that were predicted by Babcock and Wilcox. Provide a detailed description of the method and assumptions used to determine the mass and energy release rates to the containment after 3500 seconds. Include in your discussion the assumptions made regarding the temperature of fluid entering the core and temperature of spray water. Discuss the applicability of your method to both hot and cold leg breaks, and both partial and full ECCS operations.

(3) We have noted that for the containment pressure response that the pressure is still relatively high at the time of recirculation.

um

1

. , ,. . s  !

o.

.) l 1

At this time the containment spray system will become less effective in reducing the containment pressure than it was before because of the high temperature water that is being drawn from the sump. There could be a significant pressure increase in the containment at this time. Therefore, provide a discussion of how the recirculation time was determined assuming various single failure modes in the ECCS and containment spray system. In your discussion, specify the amount of water remaining in the borated water storage tank when you switch to the recirculation mode.

6.2 Subcompartment Analysis We calculate significantly higher pressures for the steam generator compartment, pipe penetration, and reactor cavity. As we have discussed with you, the pressures you calculate for the pipe breaks assumed are significantly lower than what we calculate; however, we have not been able to determine the reasons for these differences. Therefore, we will require the following:

(1) Justification of the methods and assumptions used in the analysis; and, (2) Justification of the vent loss coefficients and friction factors provided in Tables 6.2-45 through 6.2-47 including identification of each component used in determining the overall loss coefficient and the flow area to whic.h this loss coefficient applies. Also provide a sample calculation of a loss coefficient and friction factor that typifies your method.

4

6

- - . 9rrr)cHmedr 5 I%

1 4

n .

!- 1 l

=D L l u

C%

c:Z J .

".3 CQ N ces r.AJ c5 g- -

m C"D

==

==a- M W

gjp ..

O MJ a

_- w-p l>=

Cw p Cg ft.

g, O- gs .O- - n

~p b, a., R,J 4 CD Ch,= C=Ue g** o L.:LJ g  %

C'.';

"~ m: ".,,

C' La R1

'e Q W w

[me, a U c.0 &  %-

R! *e -

'"=I R

CEO RA H g Cd lE2 b g C0 c:2 e- RJ., R.J

~,

g a w , .c. ,. ~

c- - - g

!= - ""I"l ""--

RJ El CD y-w 4 es t-w hg c:'Z C3 CR C3 M gg a  != -~- "0"

. LeJ M l CD R5 W' r.---- EJ2 U e"

<e---

I l

CO CQ d C u-  :. .: a l

w CN en =2&' t3D cc tw l .

9 e

O i

l

KEY PRIT EQUATIONS -

[1 13-1

' + (Ac ) (T c -T)s Qcs "

li

.Q cs C) c .

(MCp )g dTF= PF (gN T +' QBG - QFC .

dt .

=

. (MCp )C .QFC + QMW - 9cs - Os dt - .

t2 ERel "

(Q cs + Eaux)dt t ti +t 2' 1 .

1+ 2 MRel = ,

H Steam - H water tt i 2 b inj dt I

MVessel = Mint -[Mrel +' ,

, s S A M+ *0 ,

d e

p, O gp a

{l 9

• % i e

9

\

G r .-

a

~ c', n. hf

. ~

M

),

? h ,. ' est."em% ewe r , m' ?a.woreusse

'c%  ?

V sC,d [,/J i ( -- - - --

, g, ,c;g L_.gi .' L-- -.

wp W

"Il e 'a 3 F -

N ._

g m g . ?7 J .g-- c.O gj .'?lI w c.o u.

E e

g

/ ,t _

,$ 6 g( ~. \s s._ . - n-. ,

'm i

O 9 .

fi LOW 8 TEE COLD LEG DREAK l.

'l

.. 5 , /, ,

..: .'.... . . s, .

a s *. ; . .'

.s*.*, ..

...s..s .

. .r. s .

8 ' .,: .*. s . .'

,..:.s.'.'.,***.

y '. ,. . s', r .. ; ... . ".. .,. .' ,:.;. . i.g ;_ .-

.... ..g..

k- **** ** , , = ' * ...g , ' .[ff/,1

.e*

ECC 3

• .a *e.

• p i :r'.*a r,. $.:.

', 3* e* *. , , I*;*

.~. 9*.'.*'f a . . . . . . ... *. . * . .. ' . .. .. .: ..... , 1 .,

• ' ' ;: 'j.5,;y' ; .y.g.p a. :q

.s.,.... ;. . . .

. . ..s . .. .. .. .. .- .. . .=+ .. . .. . : ., ... ...s..: * . [ ,

v.u...,:.. .1,g... ,a. . .ai 3,

. .  ; . e..

.6...,: .q. .... * ...f.;.'yt. g.l_: g. i.g. g. .

.n - -

L.a n  :

F, . .

s,

, * , ? *... [. *.,,..* ,. *, . ], ., .; .. ,. 'J ,

.. . . . .-..,.c....

& ' G1 /D 2 l;.A -s:i  : * :.*;:.a*.*].Q}i:*. '.:'f,i:.2.Q:.'.6;;<

o ,s

• e- *

,n .;4.2 g O*

. n

• s;. ..o

, g. . .

• \be .. ae

,% Q v e. .

. '.'a).,y;;.

,g .

Q-)

V ,

,  ;} . ...

t .:

t -3 p c; C* C g .+4 4

ii l

" N. t / ,

+

i

\

N e s' l l' ;g r

\* / 4M r.

3./

r

• Q ja, /

/ O

, y

• r

/ '.

jt l x, \, i .

l.t s

M g

){ . /  %. N .

! (e / .

'N s

_s,..

I t I t t

l. .

1. b

. \ /

% s . /

x.,, .

.s 1

S sa . . . . . . fan

O e

• h g

asama

• n, c.3 22:

neA D LEJ C4  != -3

~ N Q d N

-2 s =3 CD N l.ams E A 3 C 1-e i-a.

~

b uJ t- l.tA >< CD w J

w N -

d g A r- A D < e g Lt.! C3

. w a Lt.J g a W cB .

M q LN IJJ asllg, c  % d Lt.J J b' RJ .I L1""J a W. lll:s l *"'"* M

& Q O a.

4 .

gaa a 4. O V< .

F=--

I.1J W

W W CD tam 3

- ~

W &, @,

, E

+

e I

j '- -

AT771 CHM 5dr 6

.PREUMiNARY.

1. Introduction Long term mass and energy releases for reactor' building design are determined by limited application of the program PRIT. This paper provides a general description of the program and its mathematical basis.

PRIT is a general calculation program and has options which are not utilized for containment analysis. This paper also presents the application of the program, the options utilized, the assumptions made, .end the utilization and interpretation of the output.

2. General Description The PRIT program was developed to simulate the reactor mass and energy release during long term cooling. It contains provisions for core stored energy

~

release, decay heating, metal-water reaction heating, tabular input to allow

. for other sources of energy such as primary metal heat, initial fluid inventory,*

injection mass and energy, and transient fluid inventory.

The basic model structure provides 50 equal-volume core regions with input .

provisions to allow any choice of power distribution. The model may be used to simulate the entire core or any subdivision of the core. Therefore, the core geometry may be detailed to the degree consistent with the results desired.

The PRIT kinetics model for simulating core power may be bypassed by selecting the tabulan power option. If the kinetics model is used, then heat from neutron, beta, and gamma sources is assumed to be generated within the fuel If according to the pre-accident power distribution and infinite irradiation.

the tabular power option is selected, then the' heat ~ from neutron, beta, and gamma sources and the proportion of heat from each generated in the fuel must bc determined on a case-by-case basis. The nuclear kinetics required for both the kinetics model and the tabular power option are determined by a suitabic kinetics program.

The circonium-water reaction heat is assu.ted to be generated completely within the clad node. The heat necessary to increase .the steam temperature from the bulk temperature to the reaction temperature is transferred from the clad at .

Y e

_2 the point of reaction. The progrs= has provision for steam rate-limiting to any degree desired.

Within each of the regions there is a singic fuel node and a single clad node with simulation of thermal resistance according to the normal fuel rod- geometry.

Provision is made to develop the heat transfer from the clad node to the fluid

. sink node by specifying the time-dependent surface coefficient and sink temperature. The surface heat _ transfer coefficient and sink temperature input data are determined from calculations based on flow and water inventory as f

furnished 'from a performance analysis (by some suitable program such as CRAFT) of the blowdown phenomena and the effects of core flooding tanks and emergency injection. Provisions have also been made for inputting the fuel-clad gap coefficient for each s'gment.e The determination of fluid inventories and mass and energy releases is made by a special inventory routine.

The calculations begin by computing ^-the energy availabl,e to boil off steam from the reactor vessel. The tabular heat is added to that transferred from the core to. arrive at the total heat available to gencr' ate steam. The injection

~

mass is added to the mass lef t in the vessel. The release rate is determined by dividing the heat available by, the enthalpy'of saturated steam minus the enthalpy of the reactor vessel fl'uid. The mass is released to the re' actor building in the form of saturated steam until the water lef t in the vessel equals that. required to reach the leak plane. Thereafter, the full injection flow rate and total heat for boiloff is added directly to the reactor buildir.g.

The program outpu) includes the following:

1. Average fuel temperature of each region.

'2. Average clad temperature of each region.

3. Percent metal-water reaction in each region.

4. Heat rate to fuel pin.

5. Integrated heat to fuel pin.

6. Heat rate from clad.

7. Integrated heat from clad.

e

+

8. Reactor fluid inventory.

. 9. Mass released to containment.

10. Energy released to containment.

3. Mathematical Model The equations that follow describe the model used for long term mass and energy release. The symbols are defined in the table at the end of this section.

Except for the summation functions, which.are not shown below, the entire set of equations is solved for each core segment. Equations 1, 2, and 3 are solved for the entire core and then divided by the number of segments and multiplied by -

the peaking factor for each segment. If the equation is of an elementary form, no explanation is given. In all cases the dimensional form is shown.

If the tabular power option is elected, the total heat (power) generated is determined in accordance with 3.1. 'However, if the tabular power option is not elected, the neutron generation and the decay heat generation are determined.

as in 3.2 and 3.3,.respectively.

3.1 Total Heat Generation (Tabular Power Model)

QSEG = (9.48055 x 10-4).(P)-(power fraction) e see watt (watts) (1)

The power fraction is th,e time-dependent fraction of the initial steady-state

, power P, including neutron, beta, and gamma sources, which is determ'.ted by a suitable kinetics program such as CHIC-KIN.13 3.2 Neutron Heat Generation (Kinetics Model)

A one-delay group simulation is used, for which a widely accepted equation 5 is

-Q g =(0.9ff 9.48055 x 10-4)(P)(2) exp -g exp - 1, t

" -(se '"att) < watts) (2>

6 87# *#' e

3. 3 Decay Heat Cencration (Kinetics Model)

T he curve of beta-gamma decay heat can be fit bp a series of straight-line segments on a log-log paper according to the formula below.6 QBG = (9.48055 x 10-4). pat-a

= (watts) (3) see see - watt 3.4 Metal-Water Reaction Heat Tbc form of the equation 7 in current use depicts the parabolic rate of the clad radius reaction for unlimited steam availability:

K h- -aE d,

dt r ,, , [f,ro - rj RT c em em2 /sec sec cm ,

(4)

For steam limiting, the ' steam flow rate is controlling:

Na de

-(0.1875) AC b F/ g / lb zire  !

cm *

(ftJLlb steam [1b/secj sec lb/ft3 \ft 2 ) (5)

The heat generated is a function of the exothermic reaction constant and the clad mass rate of reaction:

dr Qg = (1.015 x 10-3) e (Qe). (Pc)* (r) (L). E Btu = h Btu ypm_ ft2 g cal) lbnJ ] cm sec (6) sec cal lba cm2 in gm / ft/

It should be noted that in equation 4, r will equal r oand dr/dt would therefore equal infinity until some reaction took place. In the program, ro -r is

- artificially held to a minimum of 10-4 ,

At the threshold of the reaction and for a few sections thereaf ter, the reaction rate will be high but will drop back until the temperature effect becomes i

significant--af ter which the rate will increase very rapidly. Equations 4 and

- 5 are compared, and the one having the minimum value at a solution time is

o

~

selected for the rate of change in the cladding radius calculation. To simulate unlimited steam availability, W, in equation 5 is set to 0 and equation 4 is controlling. To simulate complete steam limiting, Ws is set' to a very small number (10-6) and equation 5 is controlling. To prevent the premature selection of steam flow rate limiting (if this option is chosen by input), the comparison of the parabolic and the steam flow rates is suppressed in 'the program until equation 4 reaches its minimum value.

3.5 Heat Transfer, Fuel to Cladding 9FC " (K FC) + (CA ) * (Ty - TC )

sec

=

sec-ft Z -F (ft2 ) (F) (7) 3.6 Heat Transfer, Clad to Fluid $1nk QCS " (KCS)*~(A C )*(TC - Ts) ,

Btu , Btu f 2) (y)

. see sec-f t Z- F 1 1 1-1 l KCS " Mcs + CC)

Btu sec-ft -F "

Z I 1 Btu

+ 1 Btu

\ -l'

\sec-ft-F Z

, sec-ft-FjZ (8) 3.7 Heat Transfer to Reaction Steam (O d Since the average temperature of the steam environment decreases over approxi-mate range (600-300*F) as a tabular input versus time, and the zirconium-water reaction takes place over a clad temperature range of 1800 to 4800*F, heat must be transferred frem the fuel pin to the steam in order for the reaction to take place. It is assumed that the heating takes place without resistance or lag and is therefore dependent only on the rate of the zirconium-water reaction as dictated by the equation shown in section 3.3. Since Qs:

Q3 = Ci e(TC - Ts)-(Qg.g).(C2)

Btu Btu sec lb steam - F () Btu see lb steam required Btu (0.56) O.397 2805- ,

Q3 = (7.925 x 10-5) * (TC - T 3). Q:r; (9)

O s.-- 6*Le

In order to preserve a sensible heat balance between the reactants and the products, some means must be included to allow the QS to "re-appear" as a heat source af ter the reaction. ,

Actually, Q3 should be included as part of the combined mass-specific heat product of the Zircaloy and hydrogen. Since PRIT does not simulate these reaction products, the accumulated Q3 is stored and added to the containment as a function of the total metal-water reaction in the event of corn melt.

Over the reactio'n range of 1800 to 4800*F, the steam on the average is heated by a AT of 3000*F. Therefore, the accumulated QS for a molted segment is:

/Q3 dt = (7.925 x 10-5). (3 coo)./Q3g de (10)

Btu = (h ) (*F) (Btu) 3.8 Radiation Heat Transfer .

This capability, though provided in the program, has never been used because tl5e uncertainty in determining the view f actors seems to outweigh the possibic advantages of the results. However, the equation that is programmed is:

7 QR"KRij (Th+459.69)4-(T{+459'.69)4,a Btu Btu (11)

- see " see - F4 ,

3.9 Fuel Node

' ' d (MC p )p g (Ty) = PF(l).QN + QBG - QFC

[ Btu)[FkBtu (12)

(1bm-F)\ sec) sec 3.10 Clad Node d

(MCp )C Te- (TC) " 9FC + EMW - 9CS -O S 3g(\1bm-Fj(sec)htu\fFQBtu see (13) 3.11 Inventory Routine for Mass and Energy Release .

The inventory routine integrates and stores the energy release rates and periodically calculates the mass and energy to the containment. Total

, _7_

energy release is computed by:

+

ETOT " NCS a'ux -

Btu = see + sec (14)

The first integral represents the core and decay heat energy, and the second integral is used to account for all other sources. E aux is input as a table by the user and normally accounts for vessel metal, primary piping metal, steam generator energy, etc. The energy released between any two times is given by:

Erel M1** t2) = Ecoe (t2) - E t ot (tl)

Btu = Btu - Btu The mass is determined by division by the increment, is,enthalpy between that for saturated steam and the reactor vessel fluid enthalpy.

Erel (tid t2)

M rel (t d t ) " Hsteam - Hves,sel i 2 Ibn = Btu Btu ~

Btu lbm lbm (16)

To account for a probable difference between the enthalpy of the water initially in the reactor vessel and the enthalpy of the injection water, the two fluids are segregated and the initial mass removed first (livessel =i H nitial) . At such time that the incegrated mass released equals, the initial mass, the program allows the release of injection water; that is Hvessel = Hinjection in equation 16.

The mass in the vessel is computed from:

Mvessel " Minitial~hrel+# inj dt Ibm = lbm - lbm + lba (17)

To allow for vessel overflow, Mvessel is ' compared to a preset value and when it exceeds this value, Mrel is replaced by Minj. This simulates the overflow 6

situation and places either two phase or, at extended times, subcooled water i

into the containment. . ,

f 3.12 Symbol Definitions for Math Model The engineering symbols used in section 3 arte shown below:

A = coefficient for decay heat curve fitting.

AC = clad surface area per segment, ft2 ,'

a = exponent for decay heat curve fitting. .

C1 = specific heat of steam, Btu /lb-F.

C2 = steam required per Btu of reaction heat, lb/ Btu.

2 C = clad conductance for 1/2 the clad thickness, Btu /sec-Ft _p, e

Eaux = Auxiliary Energy Source Rate, Btu /sec.

E rel = Energy released to containment between print intervals, Btu.

E got = Total energy re' leased to containment, Btu.

AE = zirconium-water reaction activation' energy, cal / mole.

H = heat transfer coefficient clad to sink, Btu /sec-Ft 2 _y, CS H inj = enthalpy of injection water, Beu/lbm.

Hinitial = enthalpy of vesse1 ~ water at start of problem, Btu /lbm.

Hsteam = enthalpy of saturated steam, Btu /lbm.

Hyessei = enthalpy of reactor vesse.1 fluid, Btu /lbm.

K = zirconium-water reaction rate law constant, cm2 /sec.

KCS = heat transfer coefficient, clad to sink (including half of clad plus surface film), Btu /sec-F-ft2, KFC = heat transfer coefficient, fuel to clad, Btu /sec-F-ft2, 4

KRij = radiation heat transfer coefficient, Btu'/sec-F . .

1 = prompt neutron lifetime, sec.

L = total pin icngth per segment, in.

(MCp)y = heat capacitance of fuel per segment, Btu /F.

(MCp)c = heat capacitance of clad per segment, Btu /F.

Mi nj = Injection mass rate, Ibm /sec.

Mrel = mass released to containment between print intervals, ihm.

- No = initial vessel inventory, Ibm. .

NSEG = number of segnents used. ,

.P = initial total core power, watts. .

' PF(1) = segment peaking factor.

• • QBG a beta-gamma heat generation per segment, Btu /sec.

g

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

---

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

-9 QC = metal-water reaction heat, cal /gm (QMTUf).

QFC = heat transfer, fuel to clad, per segment, Btu /sec.

Qg = heat generation by metal-uator reaction per segment, Btu /sec.

QN = neutron heat generatic, per segment, Btu /sec.

QR = heat transfer by radiation per segment, Etu/sec (QRAD).

QS = heat transfer to reaction steam per segment, Btu /sec.,

QSEG " QN + QBG .

r = clad radius 3 cm (RADUC) .

R = gas constant, 1.987 cal / mole-K.

t = time, sec. .

TC = temperature of clad node, F.

Ty = temperature of fuel node, F.

Ts = temperature of fluid sink, F.

W3 = steam flow par segment, lb/sec (SFL0if).

Z = excursion power factor (Z 1).

~

8 = total delayed neutron fraction (BETA).

p = reactivity unit (RHO) (must be negative) .

pC = clad density, lb/ft3 (RHOC).

1 = equivalent single group delayed neutron precursor decay, sec-1 (DECAY).

4. Application

• PRIT is used to model the primary' system for times larger than those 'modeled by CRAFT. For quenching analysis, this is 500 to 86,500 seconds; for non-quenching i.t is 1000 to 86,500 seconds. Ihe initial conditions for PRlT are determined from the end conditions for CRAFT. The kinetics option is not employed; but power is tabular and determined from the ANS-5 decay heat curve (published October 1971) times 1.2. Zirconium-water reaction is allowed but because of the core temperatures in this type of analysis is never encountered.

Heat transfer coefficients used in the core are sufficient to keep the cladding within a few degrees of the fluid temperature.

s Hot Leg 3reaks For hot leg breaks, vessel overflow is set to the level of the hot Icg pipe at the bend above the steam generator (880,000 lbm) . ' Overflow occurs at around 1500 seconds for partial ECCS and 1200 ' seconds for naximum ECCS. At e

recirculation, 3500 seconds for partial'ECCS and 1800 seconds for full ECCS, mass is no longer released and energy is placed directly in the sump. At ,

these times, the overflow is subcooled water. Whatever mass is removed from

- the sump is returned af ter being cooled by the decay heat cooler and heated by the core. Because there is no loss of water, it is valid to ignore the mass recirculating and merely add energy to the sump.

Cold Leg Breaks For cold leg breaks the vessel will overflow at the cold leg nozzle, but this spillage will not be heated water. Cold injection water will be spilling and steam produced in the core will be flowing over the top .of the spillage. To model this the vessel mass required to initiate overflow in PRIT is set to infinity so that only steam is released. Spillage is accounted for by

, comparing the PRIT calculated vessel mass to the value which could actually exist and flowing the difference to the sump at the injection enthalpy. This is a land manipulation. Af ter recirculation, spillage would be returned to the sump at the sump enthalpy and thus need not be accounted. A certain amount of recirculation water will be converted to steam and this amount must be removed from the sump. The sump enthalpy is assumed at 260 Btu /lbs. After recirculatien, the primary system boiloff mass is removed from the sump at an enthalpy of 260 Btu /lbm. The boiloff rate is controlled by the enthalpy of saturated steam (1177) minus 260 Btu /lbm. As PRIT does not allow a switch in injection enthalpy, the boiloff rate just prior to recirculation is controlled by 917 Btu /lbm. This is con'servative by 202 Btu /lbm. Typically the total conservatism here is 22 million Btu.

Quenching // Non-Quenching PRIT is a non-quenching code. For hot leg breaks only steam is released until the entire primary system is full of water. After this time, a z phase or subcooled effluentis predicted. This is characteristic of a flow heat removal process not quenching. For cold leg breaks steam is continuously released to the atmosphere. The steam is produced from an enthalpy higher than the injection enthalpy and it is not allowed to mix with spillage before entering the containment.

4

4777KHMEdr 7 ,

r. .. . , s .

TIME OF RECIRCULATION NRC Reauest: We have noted that for the containment pressure response that the pressure is still relatively high at the time of recirculation.

At this time, the containment spray system will become less effective in reducing the containment pressure than it was before because of the high temperature water that is being drawn from the sump. There could be a significant pressure increase in the containment at this time.

Therefore, provide a discussion of how the recirculat' on i time was determined assuming various single failure modes in the ECCS and Containment Spray System. In your discussion, specify the amount of water remaining in the borated water storage tank when you switch to the recirculation mode.

Response: The capacity of the BEST has been increased from a useable capacity of 480,000 gallons to 680,000 gallons. The recirculation time of 3500 sec. for minimum ECCS corresponds to the 480,000 gallon capacity.

Therefore, there is still 200,000 gallons available for LPI and CSS operation. However, it is conscevative from a containment analysis standpoint to' assume a 3500 sec. recirculation time. The conditions at recircula, tion (3500 sec.) are as follows: .

Min. ECCS bbx. ECCS Containment Pressure 32.4 psig 38. psig Containment Temperature: 252 F 261. F

, Sump Rater Temperature: 159 141.

Spray Temperature

• 90* 114** 90* 109**

Sump Water tbss: 4.8 x 106 lb 4.17 x 10 6 Sump Water Energy 611. x 10 6BTU 460 x 10 6

• Changes due to switching to sump.

• • Spray Temperature af ter sump water being cooled. -

l t.

0

• The times of recirculation are as follows:

Containment Analysis Actual Minimum ECCS 3500 5100 Maximum ECCS 1800 2600 Since the additional 200,000 gal. useable capacity allows 600 sec, more of 900 F injection into the containment for the min. ECCS case and 800 seconds longer for the max. ECCS case, it is logical to assume that both the containment pressure and sump water temperature would be ,

reduced.

The recirculation peak would, therefore, be less severe.

O G

e G

9 4

7-s e

._ A7 m cfinic ur~ 8 O

9 SUBCOMPART'ENT ANALYSES

~

9 9

9 e

n e

1  ; .

• e.

TAE!.E 6.2-29

• SimCOMPARnfENT At:At.YSF.S Stae:ARY Prux/ time Mr.ax/ tire Pressure Break Aree (>st /see / node s) Response Type clowdown Nodaltration (cci a / c ee Ya eJ.l Cn parteent (Ft2) 1.ocation l #j S/ O.if b !O*08 FM,-s/vM- 4&ft0't- T6.2-53 Splic T6.2-34 T6.2-81 T3.2-83a i P4 actor Cavity 4.276 SE Cold Les (1) -

(1-6)

• ' Equivalent T6.2-835 327 - *i GI)lo.4-T6.2-82 31G/,03i/a.4 d.nG h & W' T6.2 *4 7.875 SE Hot Leg split T6.2-37 *

(1-6) T6.2 (i")

• Equivalent T6.2-bok

- 16. 2.Ja c 16.9/o.34

-35s/db== Er:Strtt T6.2-57 t T6.2-40 T6.2-79 15.75 DE Hot Leg Split with

• r 76.2-57A Steas Pressufizer F a

Cecerator Equivalent Discharging

(,/- li )

(51t' F(.2-5lb 3 into Primary ,

Iq.r/0 3s E - System -

• 3h5'frG4-- 4fr:fh 1* T6.2-79A Hot Leg Split with T6.2-23 T6.2-79 76.2-793

15.75 DE F (1-15)

• Equivalent Pressurizer

.- Discharging sa into Contain.

E X T6.2-80 -37147/-iOY NA T6.2-80a Cold Leg Split T6.2-33 h Piping - 8.55 DE /G3 o,g(

Equivalent

• " Penetration ,

o 8.3/o.G 5 SG.U/ 12= - =S . ;/- .? T6.2-57C Double Ended 46.64 T6.2-57b'

• E Pressurizer 1.79 Surge Line 34* - (13-3)

Cutilotino .

)  % 8.*

L em.

" *-Previous tv-reaching aquilibr-lum-with-overall-Containment Tressure

• y Reaches Co rtlo s'nvYiC 6d pre SS W c M Cg6 DIE bd " *

. .E

• y;*

l

- .;p

. F3

• ? *./

t ~

.: g ?;

I .,

[.

l , . . . . -.

s . .

. t -

j ' *

.j -

l'.

j ~ '

~

COMPARTMENT- .

1 0 . .

-l. 2 c) 4 s .

. 9 m

6 + -

~ -

o -

• - l v

CC

• e-. . .

. .uy .

A o . . .

es

• ~

• o

. ~ .

h1 .

xo -

c - . .

. . U) -

?

U) uJ .

. M -

.L 9

• ~

s .

e

^

o - , *

• 1[ hh h [ [ *

• TIME (SEC) ,

..~ Figure .8'.2-56 I..

4 Reactor Cavity Subcompartment Analysis ~

2

. , 7.875 Ft Hot Leg Single' Ended Split .

1 m4+u . e, . . --+ -- --u- - y- +w , -r--

. ,3

-3ii. ,

. 2 .

g -

. I.

e F , .

i g

u .

r e .Q gas megw o ,nm ~ CC _.

. 6

.o ao _o .

m*o uo o nm 9 . .

+

2 . 1 t, 5

6 0-2 -

A .

i a

i _+ .

.y.

)

+

i .

i s _.

6 .

R 6 '

7 e 1 1

a .

8 c 0 7

5 t

o r

F .

m- t C .

2 a v

Ti ._ .

- I I .

i

. l o t M t y E .

i 1 S g

u b

( ._

• c Si S o E i m Cs n p -

g a )

s l r .

e t $ .

E e m 6 . r_ 1 1 1 3C 3 n n 6 - 21 0 O d t M .

e 1' mP

• d S

A n

a 0- 2

.+ A0 A p l . .

R l y T i s M t i f

s i E N

, T s

i s

a i

d ... .

s .

6 .

i t * .

o.

F .

e

• .!e * .

. . . . > .\

i t a . .

~

t

- . s

' ~

COMPARTMENT j

- f - -

sa .

7 6

• 8 A .

+

o -

9

• w -

o.

'r

- ~ .

. ~

C .

~ .

w .

A o.

• a .

s - .

W * .

xo .

2 uy (n

W .

x .

c_

o

~ .

o . .

3 A S E i E 5 4 0-8 E 5 I .5 E i EE W .-

10-* i 5 I 5 5 5 6 0 '1 0-2 i TIME (SEC) .

Figure 6.2-568 Reactor Cavity Subcompartment Analysis

• 7.875 Ft 2 Ilot Leg Single Ended Split 6

e e

F . <,=,~..x.,,t1.

4 i . ~i., '

. g .

u ,

r e -

6 '

2 .

5 .{

6 .

C .

a6 mw~-

a uMkzs a J ~ a- s eDgmtmMw,

n 4

• o n.a ;o v> a .

. .o 9 ~ -

i -

1 .

0 D R 4 -

i e .

f a .

f c -

e t r o i .

e r n

• t C .

i a * -

a vi 5 - .

l P y r

t i

/ .

e s

s u

S u

b c

s 6

V .

r o e p m i .

R' a s . '

e r s n s t i .

p m .

o e o-2 - .

n n

s t .

e .

A -

n a Ti .

- l y I - -

s M .

. i '

7 s E s 8 ( ' -

7 5 S4 .

E F ^ ~'

t 2

Cs .

) s I

l o

t i

s s

+ ,N. . .

1 1 1 1 C.

M O

I.

e z .

TTTTP h

g OOOOA

- S i

0' ' - R T

g n 6732M E

A.

l 5 e E N E

n

+A00T d -

e $ . .

d . .

S' .

p l

i t

' 4 5 w .

~ .

5 , '

3 . <

' <t 6 - *..

0 .

- ' ]! '

0* .

. i

t; ' i -

. , ' . i -!a: ::.:..tg

.  :  ? ..% ' .

y r s t- *

s. -

t; q

~ '

COMPARTMENT . .  ;

1 6 .

2 0

. 3 A

. o 15 +

g.- .

.i

- s o.

w .

= , . .

. a- -

m -

m ,

Q_

o. - .

c3 e w

- M o O

m a-U) y e __  ;

M * -

k.

o O

' 4 i $ A $ S iSS I O-2 i 4

$ A SSi0$iO i s A 5 SiSSIO' i $.A$siASIO*'

. '10 l

TIME (SEC) .

Steam Generator Compartment Analysis

,\ Figure'6.2-57 ,

15.75.Ft2Leg Hot Leg DE Pressurizer Discharging into Primary System e ,

i

'j

, 's ,

. STEAM GENERATOR COMPARTMENT ANALYSIS " '

. 15.75 Ft2 IIot Leg D.E. - .

t -

' Fressurize'r Discharging into Primary ,

System .

g Fig. 6.2-57a -

COMPARTMENT *

- ~

2 TO 15 8 '

~

3 T0- 15 0 "

1 TO 15 A -

. . o

, - u 6.

U) . . '

g .

l W

Z o 4~ .

.a u y3 u) '

La .

g  !

a. o. . . -

m . n

=

. CE . .

p .

z .

l

{

td o -

Z w f- -

ti.

LL. - ,

e.-< -

a .

i

o. .

m .

o '

• -E~ . . . . ... . . . . .. . . ... . . . . . . .

' 0-* 2 1 3 4 5 6789IO-2 2 3 4 s s765I0-5 2 3 4 s siss\0 2 3 4 5s.78910' TIME (SEC)

. e 57A  : .

9

i

/ 4, e -

- .i(

l '

, 'V , PRESSURIZER SUBCOMPARTMENT ANALYSIS PRESSURE RESPONSE 1 4(-

. Fig. 6.2-57c. - .

. e

. * +

COMPRRTMENT 6.

. 1 e

. 13 l

o. .

W . .

9_

o N ,

e .

=-

d -

c ex . .

, O a .

CL -

v l Y * '

QC o D

w o'

• (O

. td -

0:

L .

o .

tb ~

2 5 4.5 $iSSIO- i s i 5 $ i s5'1 0-' i 5 i S SiSS310" , -

1 0-* i $ i SSiS$310-2 i

. TIME (SEC) 57t;

~' ~

., .;.- y.

V6 .

l STEAM GENERATOR COMPARTMENT l.NALYSIS

• 15.75 Hot Leg D.E. Pressurizer *

- - Discharging Into Primary System Fig. 6.2-57d . .

COMPARTMENT 11 0 -

i

12. D .

A

. 14 ..

i

o. . ,

. o. .

t * . ,

• 1 .

o.

v cc . ~ .t in . . .

e a. _ . ..

• try . .

-O~ . . . . .

e LLI ..

M o a g-u)

(D ,

j-LLI c 0 - .

'M

a. ,

o E~ - *

. o E $ .A 5 5iS$h0*

i i A S$6640-* i 5 A $ 5i0$'10-5 i $ A SSiESTO i . .

. 10-3 i

.. TIME (SEC) ,

e

. .s s.c . .

- STEAM GENERATOR COMPARTMENT ANALYSIS 15.75.Ft 2 Hot Ihg with

' Pressurizer Discharging into C*.tainment - . .

Fig. 6.2-79a .

COMPARTMENT t g 2 0 - .

3 3 ,,

15 +

. o . .

. u> - -

s .

o .

i .

m . ..

a- - .

~ .

U3 .

. a. o .

O * ,

> - a .

w

-a Of g- o .

v2 .

U3 ,

• w e .

ad ..

o-o .

o 3 4 5 678940' S

5 S A S 6 ISS'1 0-5 i S A 5 SiSS'10P.

10-' i S A 5 S i SS I O-2 i TIME (SEC) 79k  :

I g I g ! :i * , -

. It a.h.uwak2s

.- - :J c-cJ

• m g au m u,O x e a_ m -

P S r T g, W.O go - o - uao uao e s

E A

- _ - - - - Fs1M iu5

. 1c gr.G 0-2 . i7E z5N

. . - 6e E

. rFR 2 2 tA

, . D 2T

. 7i O 3

- 9sI R l

. bco

. htC 4 .. .

a O rLM geP

s. .

ig A 6

' n R 7 .

. gwT 8 . i M

s. - itE .

- nhN 11 t T

. o 0-2 C A

N o A n L 1 t Y a S i I

- 3 n S m

e 4 n t

T s. , .

I 8 M7 e.

E s. .

1I -

( ,

0 S - 8

.E C

)

2 3 . 1 32C 4

O M

s. TTTP

,s. OOOA 7 R 8 T

s. 1 1 1 M I .

555E .

O" N

, T 2

A' 0 0

. 3 - .

4 . .

s. -

s. . .

7 .

8 '

s. I 1

0' . -

7 .

4 .

8 -

' . I .- ' ,

.I 2 .i - ' .  : s:.,i i . ie I , ,. -: , ll * ,, :

!.  ?..

['*g g s 3- '4 c 4G */!J':0 -j% A f T

' O 'sih,.' s.sw.47a s a e a:vce e.s . , u. ..

s currn. o.esec's cr. .

l .

t i.

I 8

f 2 3 4 5 67899 2 3 4 S 67098 2 3 4 5 67sst I t 2 3 4 5 s 7838 2 3 4 5 6 7894

..I~~ I h - __. - _. .

y. **l  ;;. J.;; __ .. . _ . .. . .

. .Ly g .7 .

.j 7

.i n _.._ __ . . . : T_ . _. _ .. ._. . _ . _ ._ _

..  : n T - ..

1 M .._[: , ./Ztg (gf7 /75;/.

s _

I

!i ii3' t ':I l'd 'O. 2:.

5

-- 2

i. 6

.:$j u,.lj .:.

p- .:!- -

t'!

- .r

.,;.- r. g- -

l;p j;is- l _j .

.:. _. L

.w  ; .: . .; . . . .

._;;.y .=

.=..

i . .

.. , .. 6.- __ ._. . .. .

.,, ,j .rg . ._

. _ _. ..- . . . ... _ _ , . .p ,

.... . o. -

3 _.. __ ___ _ _ . .. _ . .

.a; ... r. ,. u.p ..

-i

.% a _ _ _ _ _ . . .

y . . _ . .... ._. .

j  ; ,.a . . :. .. .

g . 2

!!:; Illi 3. .. .

+ .

li- ... .  :: .

w ._. ._ _- . _ _ _ .

z '

i 4!i  : la i4 . T..L i-

.' iid-

.!p: .g , t..

.. . . .a. .

4.: . .

3 1, ,

. ,. . g

..e4., .... .. . . ..

.s 10g

[g,

. p'. ,

+ . I e .: ,,

+

ll: ' 8 l ,, .

.t e p..; ;e-i ,1. ,

gg1 6*

4 . g .:

g ;. s .

l 3

.t :

.c .I 1,. ,;; .e . .. .. ,.,  : s.3 ."t

.l , l-

4. . J y - . ._ .. .

g. .. _. _ . . _ . _ . .

9

-.-_.r

~~ ~

q__ . _.

3 . __

.._. . f_

a -

l -..s a . .-. l - _- I.-. i, g

. . , . _ . ._. . a . .

}

i .

/. . .,_. ? ).: .. l . ...- . .. _ .. .

s

.h b

, !IIl l* t  !?i:

..'. z. . . .

-: 3, !- :,  ::

f . _. . .

j t

!p ..;t.:

w . .. .. _

jls :; . l l

s l'.;.

t:-

..l

]::.

.;_: a. 4

.'. k. .;

1:

4,l* jj*f .,. .

g_ .. .

.a _. .. ._. . . .

L ..! .*.. J . -.. _ .. _. _ .. .

T'-

. . ;. s .;. _ .

'"N  :#: ..

_m'4

_i _i eo.. :

I t;- 3 . .

-., ~ '

-  !), -- - - - - - ,

m -+ +

s :3;I rj r-

}-

.l 4. ;-

i e-- -i-

, 3 +

i: ... .. .e .  :,

./_ .

a _ . . __. ,, .,

.L.....

, I l j..i ,. .. .: ... ..-. I. _.

,,i

..:,..i..... .. ., .., .

.. . . ,,,,- t .

..i.

. ... . t. ..

t -

PIPING PENETRATION PRESSURE RESPONSE

' t * - -

m j o, -

'--- ~-"-

Q *

l. -rh COLD LEG SPLIT.

. . .. . . .... . . .... ...... I, r r s

Figure 6.2-80A -

' lb s _

s' l l i .... . . . . .

_j_- :_. .. _,.

1 ._. .. ... . . . ....

i. . t. _.,.

i 4 _

&,. .d. .g e  ! . ..- . .

I

..g3.

.. i Ih. . ., .. .. . ..e . .

3 -

.'  :. _[_

i '.iT, ;

t.n,: .;i._ ._ .. .. . ._ _ .

l _._ ._ .

m .. .

-l

. . ;;: :!,i . . . . . .

. .  ?. :. t.:. :.

,:. .i

.  ;  ;. f: a. : . :  :: i- .

.. ...--n. t...t. ;. :

n. . . .... ;i . . i;. . . . , .

.i 11 .- I.;;, . .

g - . . . ,. . .

...l 4, ..,.

i .

.... Il . . .,ll.

4,, .gt, sq g .. . ... .

. ) . .

l g, . .. .

!. g a,... - a.. ;P.. - -

m g .

.e.

,n

. . . . 6

..:ii  ?,,. , .. - - .

f.. ,, ,9 - i n .

t. -

l  !..

, g 'O 0 01 al -

/

i .c. cool o. col 1 -

j .

TIME W c) .

O* O

• 8 o-REACTOR CAVITK SUBCOMPAR1NENT ANALYSIS

• 4.276 Ft- Cold Leg Split , ,

i - Fig. 6.2-83

) .

] .

~

• COMPRRTMENT .

i - -

. 1 0 2 .O 3 A l

o*-

. 4 + .

- O ,

. s u ,

• e

• G

^

9_ -

o g . ,

e .

CI~ . .

U) .

. - c .

o. . . ,

e. . '

(t' O. 0 0 _

o. f, -

n v. =

Q M .

o;- .

0, . .

o, g

j-A A A S 585IG^ 5 * ' ' .". . .

-4. gp a i tis.t6A.io;' t 1 i Aiif8S'10" i '

. IlttE (SEC1 ,

t .

85

. \* .

REACTOR CAVITY SUBCOMPARDENT ANALYSIS . ' *i

• - ~

4.276 Ft2 Cold Leg Split Fig. 6.2-83a ,

~ '

COMPARTMENT ..

B

~ '

. 5

. 6 6 -

. O. .

• O

.-e t . . .

O e O. . - ,

CD y , ,

g ' '

~ -

h g

= .

a_ O.

. CD O .. , .

. w . .

bJ .

E o *-

a . .

m (f)

• hJ -

\ x .

Q.

• O. .

4-

~

. t ,. .

. o a i s aih6'10= i i i s sib 6'10-: i s i s s i66i o" i 5 A. S S N 10-* i TIME (SEC) g3a .

.=

1

.. 1

~ . . <

l DIFFERENTIRt. PRESSURE I10 PSI) 50 10 0 15.0 20 0 25.0 00 f f I p f

g i

.G o,* .

' e 1

y p ny

• N' .

l 03

• nOd en

. w- .

.os . o. n w t-2

+- tu e

ut-os

. u co

• • cr to 4- *o co- w w- . .

. H

. O e

en .

to- *- .

4 tn H

M t.9- .

4

. 4- ' -

y m-Cn-

\ .

.J-3 co-R w-e r" -

mO m g m

O w

ro- -

w-

,t e-a e-a e- a O I t O s- 3 -

"~ -t -I --I 7

\ O O O "D C)~ l y

4- *d

.co. *-a tDa . .J TQ 3 e x- m. m o,. ,

z n -

\

>GGw .

• U  !

.g. f .

\ .

~ "

. 01-

.J- () fI . a co- -

w - , t q w -

oo Cle .

M -

e.

~.*e.*-*

• e

• ~

4

//e 1161CHA461*[ 9

~ -

Estimation of I.oss Coefficient Each flow path between subcompartments is censidered as com-posed of one or more of the following flow elements: .

. (i) Sudden contraction in area, .

(ii) Straight section with wall friction, and (iii) S'udden expansion in area. ,

The loss coefficient for a' sudden contraction in flow area is conservatively considered as 1.0, although actually in some cases the change in area is not so large and the velocity of approach is non-zero. , ,

The loss coefficient (fi/d) corresponding to the wall fric-

^ . tion of a flow path is calculated estimating its ceutvalent length and

( .-

diameter. In oost of the cases this loss coefficient is observed to be negligibic, but has be conservatively estimated as 0.1 . In some cases larger values have been uscd. ,

The loss coefficient for a sudden c'xpansion in flow passage is conservatively considered as 0.5, although actually in some cases

- the change in area is not so large. .

It is to be noted that for a flow area where the direction of flow is not obvious, the larger loss coefficient is used, as the con-tribution from the contraction and the expansion are unequal.

l j

c-

~

• < 3 .

STEAM CENEPJdOR COMPARTMENT Flow Path Loss - Coefficient .

Contraction + Friction _ + Expansion = Total

~'

0.0 0.1 0.0 0.1 1-2 .

0.1 1-3 0.0 0.1 0.0 0.1

'. . 2-3 0.0 1.0 0.1 0.1 0.0 0.5 1.6 3 - 14 .

0.5 0.6 3 - 15 0.0 -

0.1 1.0 0.1 -

0.5 1.6 4-5 0.5 1.6 4 - 14 1.0 0.1 .

0.0 0.1 0.5 0.6 4 - 15 0.5 0.6 5 - 15 0.0 0.1 .

1.0 0.1 0.0 1.1 6-7 '

0.0 1.1 6-8 1.0 O.1 0.0 0.1 0.5 0.6 6 - 15 .

0.0 0.1 0.0 0.1 7-8 '

0.5 0.6 7-9 0.0 . 0.1 .

O.0 0.1 0.5 0.6 8-9 ,

0.6 9 - 10 0.0 0.1 0.5 ,

1.3

• 10- 11 1.0 0.3 0.0 0.0 0.3 0.5 0.8 11- 12 0.8 11- 13 0.0 0.3 0.5 1.0 0.1 0.5 1.6 12- 13 1.6 13- 15 1.0 - 0.1 0.5 -

0.0 0.1 0.5 0.6 14- 15 ,

e e

• . e

. g . e t .

e ,

g D C

e e

~

.. a * ,

. c% *-

(/%

s_ .

REACTOR CAVITY .

loss - Coefficient Flow Path

• Total Expansion =

Contraction + Friction +

0.1 0.5 1.6 1-2 1.0 1.6 1.0 0.1 . 0.5 1-3 .

0.1 0.0 O.1 1-7 0.0 0.1 0.0 0.1 0.0 1-8 0.1 0.0 1.6 1 - 15 1.0 1.6 0.1 - 0.5 2-4 1.0 0.1 0.0 0.1 0.0 2-8 0.1 O.5 1.6

'2 - 15 1.0 1.6 1.0 0.1 0.5 3-5 .

0.1- 0.0 0.1 3-7 0.0 1.6 ,

1.0 0.1 0.5 3 - 15 0.1 0.5 1.6 4-6 1.0 0.1 0.0 0.1 0.0 4 - 10 0.1 0.5 1.6 4 - 15 1.0 1.6 1.0 0.1 0.5 5-6 0.1 0.0 0.1 5'9 '0.0 1.6 1.0 0.1 0.5 5 - 15 0.0 0.1

([]

6-9 0.0 0.1 0.1 0.0 0.1 6 - 10 0.0 1.6 1.0 0.1 0.5 6 - 15 0.1 0.5 0.6 7 0.0 0.6 0.0 0.1 0.5 8 - 11 .

0.1 0.5 0.6 9 - 11 0.0 .

0.6 0.0 0.1 0.5 10 - 11 0.3 0.0 1.3 11 - 12 1.0 0.8 0.0 0.3 0.5 12 - 13 0.3 0.5 0.8 12 - 14 0.0 1.1 -

1.0 0.1 0.0 13 - 14 0.1 0.5 1.6 13 - 16 1.0 0.6 0.0 0.1 . 0.5 15 - 16 e

g

• a e

l '

C' .

r= -

. . N .... .

{)

.w PRntARY SilIELD WALL PEMETRATION Flow Path Loss - Coefficient Contraction _ + Friction + Expansion == Total 1-2 O.0 1.3 0.5 1.8 1-6 0.0 1.3 0.5 1.8 2-3 0.0 0.1 . 0.0 ,

0.1 2-4 0.0 0.1 0.5 0.6 2-5 0.0 0.1 0.5 0.6 3-4 0.0 0.1

~

0.5 0.6 3-5 0.0 0.1 0.5 0.6 4 '- 7 . 0.0 0.1 0.5 0.6 1.0 0.1 -

0.5 1.6 5'- 7 . ,

6-7 0.0 0.1 0.5 0.6 4

8

• e D .

. i D

e e

e .

C' l

e

. . l

.,s, . ,

WNP- 1 Amandecnt 11 PSAR October 1974 g,3 .

V ') s

• i -

\l TABLE Q6.65-2 STEAft CENERATOR COMPARTMENT NODALIZATION

. Subcompartment Volume

_ ^(ft 3)_ Nomenclature (s. Number L

1 _

18,069.0 Steam Cencrator-East 2 23,463.0 . Steam Generator-East 3 88,857.0 Steam Cencrator-East 4 130,920.0 Steam Cencrator-West 5 2,6,280.0 Pressurizer  ;

6 57,525.0 Reactor Pool Arca .

7 '. * '

963.0 Reactor Cavity Nozzle Arca-East 8 963.0 Reactor Cavity Nozzle'Arca-West 9 ,

1,630.0 }tidsection Annular Arca

. 10 .

8,112.8 Reactor Bottom Area 11 .

3,072.0 -

Cable Chase 12 .

7,597.0 Incore Instrumentation Comp

. 13 67,194.0 First Floor Volume-East 14 41,847.0 Core Floodir.a. Tank Compartment 15 2,701,150.0 Containment b

. O . g ,

e 9

q y ...

. i . ,

n. ,

'q~ ,

r _ _ _ _ _. _ _ _ _ T _ _ _ _ _ ,

l 1 2 l I I I I I I i< l I 1 I

4 I

1 i 5 .

3 ,

. I ~

I I I

. l i I .

4

_________._J L _ _ __ _ _ _ _ .__J m

14

. r 15 . . ' , .

6 i , '.S s

i i 7 8 13 i .

. . 9 .

l .

i 10 11 12 AMEtlDMEf1T 9 AUGUST 1974 ,

WASHINGTOtt PUBLIC POWER SUPPLY SYSTEM STEA.! CE!iF.RATOR k- WPPSS NUCLEAR PROJECT tJO.1 '

CO:!PART:E:;T !;0DALIZATIO!!

Preliminary Safety Analysis Report i

FTC. 06.65-1

. O

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

-a g-

• ej

=0 3Y tln .'.

m

$w b E I """"""

) a

_c n - 8

>- 4

. v a .I a __

3

~

3 $

(f r

e. ,.

a -..

g 9 .g

-CQ V ,

e Q- o ug s

5-0=", - 5  !

~0 pi3

-- O - OT-- >- l -

v a.. .i S t. r

• a

_on  ; i

~ sv Sh s lNE f l -

@ 2 h 4 5--- _ ,.

(.w )

oc-g 23g48800c3c83 .

q=gnmeTTS*.co noe ""G :: M3033y s

E. .C. R 4 R. ;. 3 t. n. '3' C-.2 n.., R

e. H...4.,

""3 *22*

,} . .

• mp - - - -

v g,9 c sg Ac : : H ;.U' g C" - - . g d wy'www]d"M s WE5u  %

, i h 44.de&E54I5555II

. ... .nnan, i6 "

s  !

• d g d 4 4 4 Nc ;er;9Mb Ss W ,1 .

< 5

. ffea,6"85

• i 87*e sum. S, = 3

% Essssssssss.,ff.~"-

htVthttttt? 5g =

4 n s s a c c au s ,a n a c - S 'a aE 5

5 .".,5 .,e 5 5,===5's 5 5 5 5 5 v g='dU" y n n ** ss an n ar.n* nna n

%4a N

r II.VVVVXVWWVM

• f R H " f! ?eMr8 :! ? 9 9rr:rrrrrrrr:5:8h 7 .

O

• e C * .. .

'.. ..e:eb ee -ZE44 ecene z c e s

- - = p. =<<<<<<44444g .

4 i . . .

. . *,'..~ '

'i ' - . .

WNP- 1 Amendm:nt 13 -

PSAR D:ccmbar 1974 ,

TABT.E QG .66-1_

O' s - -

U' * - - Cold Leg Piping Ponceratica Nodalization .

Subcompartment Number Volume (ft ) Nomenclature ,

~

1 37.52 Primary wall sleeve 2 963.0 Reactor cavity nozzle area adjacent to

• penetration ,

3 963.0 Reactor cavity nozzic

• area opposite to penetration 4 . 5752'5.0 Reactor pool area '

5 9743 Reactor midsection

& bottom cavity Steam generatoc arca

/3 6

130920.0 7 2891400.0 Containmcat ,

d) Area Openings Between Compartments

• 2 Opening Between Compartments ,

Area (ft )- _

~

1 and 2 - 1.34 1 and 6 . 8.60 .

2 and 3 ,

50.24 . .

2 and 4 -

103.50 ~

2 and 5 22.00 3 and 4 103.50 3 and 5 22.00 4 and 7 1405,00

$ and 7 96.00 6 and 7 ..

1805.00

/0 4

O O . ,

t l

. so-

• t

_ p. -~ _,

4 e

1

- s, .

t

'-- -) .

) . .

I . .

e 4

O 1 .

1 .

b .

3 .

. y t

6 s

AMENO.'. TENT 10 s .

SEPTEf.tutin 1974 i

(J '

• PRU!AP.Y S!!!CLD t.'ALL '

WASillNGTON PUGtlC POWER SUPPt.Y SYSTEfA PENET A' MON N%

WPPSS HUCL. EAR PROJCCT f40.1 Ptcliminary Safety Analysis fleport I FIG. Q4.4A-1 . _

w..