W3P83-0636, Forwards Draft Amend to Fsar,Consisting of Results of Revised Large Break LOCA ECCS Performance Analysis.Amend Also Updates FSAR Sections 6.2.1.5,6.3.3.1 & 15.6.3.3.1. Info Will Be Included in Next FSAR Amend

Forwards Draft Amend to Fsar,Consisting of Results of Revised Large Break LOCA ECCS Performance Analysis.Amend Also Updates FSAR Sections 6.2.1.5,6.3.3.1 & 15.6.3.3.1. Info Will Be Included in Next FSAR Amend

ML20072C528
Person / Time
Site:	Waterford
Issue date:	03/03/1983
From:	Maurin L LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
W3P83-0636, W3P83-636, NUDOCS 8303080467
Download: ML20072C528 (47)

Text

)% _f C

LOUISIANA ,4a OnanONOe srnar P O W E R & L I G H T[ P. O BOX 6008 . NEW OnLEANS. LOutSIANA 70174

• (504) 366-2345 SuSN5vsY$

L.V. MAURIN Vice President Nuclear Operations

.'tarch 3, 1983 W3P83-0636 Q-3-A29.15 Director of Nuclear Reactor Regulation ATTENTION: Mr. G. W. Knighton, Chief Licensing 3 ranch No. 3 Division of Licensing U.S. Nuclear Regulatory Agency Washington, D.C. 20555

SUBJECT:

Waterford SES 3 Docket No. 50-382 ECCS Reanalysis

Reference:

W3P82-4063 from L. V. Maurin to T. M. Novak dated 12/22/82

Dear Sir:

The referenced letter documented LP&L discussion with the NRC which led us to perform an ECCS reanalysis addressing the as-built safety injection tank delivery line flow resistance and the effects of operation of the contain-ment purge system on the minimum ECCS containment backpressure using the CE ECCS evaluation flow blockage model developed in response to the require-ments of NUREG-0630. As noted in the referenced letter, we understand that NRC approval of this CE model will support the timely review of our submit-tal.

Please find enclosed a submittal of a draft amendment to the Waterford 3 FSAR. This draft amendment consists of the results of a revised large break LOCA ECCS performance analysis and updates FSAR Section 6.2.1.5 (min-imum containment pressure analysis), Section 6.3.3.1 (introduction and summary), and section 15.6.3.3.1 (large break LOCA). In addition, several references are added to Section 15.6 defining the ECCS performance model revision used in this analysis to conform with the required clad deforma-tion and flow blockage model guidelines of NUREG-0630. This information will be included in our next FSAR Amendment.

Yours very truly,

/

?? G L. V. Maurin LVM/RMF/ssd cc: E. L. Blake, W. M. Stevenson, J. Wils w (NRC), L. Constable 8303080467 830303 PDR ADOCK 05000382 A ppg

\ 1%'

WSES-FSAR-UNIT-3

~~'

Following closure of the MFIV's, there is an inventory of feedwater be-tween the MFWIV and the ruptured steam generator. As the ruptured steam

~

generator depressurizes , this inventory starts to boil. As steam in the line expands, this feedwater inventory is pushed into the steam generator and is boiled of f by primary to secondary heat transfer. The expansion of the feedwater inventory into the ruptured steam generator has been con-sidered in the analysis. The expansion is assumed to be isentropic.

6.2.1.4.5 Energy Inventories An energy balance for the nost severe secondary system pipe rupture is provided in Table 6.2-9 6.2.1.4.6 Additional Information Required for Confirmatory Analyses The flow area of the main steam lines are as indicated in Table 6.2-1. For the MSLB analysis, the postulated rupture is assumed to occur at the nozzle of one of the steam generators. Therefore, the fL/D from the ruptured unit to the break is conservatively assumed to be zero. In the MSLB analysis, the fL/D from the intact steam generatcr to the break is assumed to be 10.97 which is related to the flow conditice of the 32 in. i.d. pipe at the steam generator nozzle.

Feedwater flow to the ruptured steam generator for the most severe MSLB cases listed in Subsection 6.2.1.1 are chown on Figures 6.2-13c and 6.2-13d.

f s Minimum Containment Pressure Analysis for Performance I 6.2.1.5 Capability Studies on the Emergency Core Cooling System 6.2.1.5.1 Introduction and Summary features of Appendix K to 10Crr.30providestherequiredandacceptag{g) Included Emergency Core Cooling System (ECCS) evaluation models in this list is the requirement that the containment pressure assumed in the evaluation of ECCS performance not exceed a pressure calculated conser-vatively for that purpose. The ECCS performance analysis for Waterford-3 which is presented in Subsection 6.3.3, meets the minimum containment pres- l9 sure requirement of Reference 11, Appendix K. Paragraph I.D.2.

6.2.1.5.2 Method of Calculation The calculations reported in this section are performed using the large

~

break evaluation model described in {gjgrence 12, which was approved by the NRC in Reference 17. The CEFLASE-4A computer program is used to -

determine the mass and energy released to the gggjainment during the blowdown phase of a postulated LOCA, and the COMPERC-II computer program is used to determine both the mass and energy released to the containment during the reflood phase and the minimum containment pressure response to be used in the evaluation of the effectiveness of the Emergency Core Cooling System.

The ci;ilarity tich  ;~:!: fe r t'- "??? cf t'^ 25 5 ""! 22cter ;12-t: _

m..c_. , _ _ . , , .s,.; c: ._. u,,m <,.4_, -_4

______.__a fr__

-- - -"' - "'~

' ~~ ~~- ~'

rilgr*2 2}

. _ ( 15 I6 )' _.2 a g:-'ri: b1:xd:u: : 1:u!: tic- 22: 7'r' rr^d frr 211 6.2-30 Amendment No. 9, (6/80)

r

a. ,

WS ES-FS AR-UNIT-3

_c ar, r_ ___,_r__,_, 2._:__

.. _ . . _ _ ..-...._. ..___.__,2_._._

__ .t..__ ____c_.____________

..c a; tFr

: ;!::::, th: ,Iri - :::::Irr: t prrrrrre tr Er cred in the SCCE

. perfr:-- cr : ri;ri_ 5 :

1: !:::3 :p::if!::117 f:: '?: t :: f r : f-2. .

6.2.1.5.3 Input Parame ters 6.2.1.5.3.1 Mass and Energy Release Data 20sEna A The mass and energy released to the containment for the most severe LOC ,

the 0.8xDEC/PD break, is listed as a function of time in Table 6.2-19. ,he l23 di ::: :d i- Erbrecti:- 5.2.' 5.2, th: ::: red e-- ;; relerrrf furi:;

51 erd ^- Err 5 r- crirrict > . -- ^ r ic " ' y f o r 211 '^- ""* pir-tc, : d the

> '-^rgy - 'errr! Arrir; refired Er 2:12 12 tro r ;--

• fir:11; f:

r

'..  :: c f ;4-4. The quantity of safety injection fluid assumed to spill from the break is discussed in Subsection 6.2.1.5.3.5.

6.2.1.5.3.2 Initial Containment Internal conditions The initial containment internal conditions which have been assumed for this analysis are:

Temperature - 80 F (Minimum) '

14 37 Pressure - 4-2 psia (Minimum)

Relative Hur;idity - 100 percent (Maximum)

,m For each parameter, the conservative direction with respect to ninimizing the containment pressure appears in parentheses.

6.2.1.5.3.3 Containment Voluue The net free containment volume assuced for this analysis is 2,677,000 ft . 3 6.2.1.5.3.4 Active Heat Sinks For this analysis, it is conservative to maximize the heat removal capacity of the containment active heat sinks; thus , both the containment sprays and all four containment fan coolers are assumed to actuate in the shortest possible time following the break and to operate at their maximum capacity, assuming the minimum temperatura of both the stored water and ecoling water. To minimize the actuation time , offsite power is assumed to be available for all active heat sinks. (It should be noted that of fsite power is assumed to be unavailable for the SIS.)

The Safety Injection System equipment assumed to be operable for this analysis is discussed in Subsection 6.3.3.2.1.

The heat removal rate of the containment fan coolers is shown as a function of containment temperature in Figure 6.2-30a. The operating parameters assumed for the containment sprays are as follows:

6.2-31 Amendment No. 23, (11/81)

s .

Insert A The mass and energy release during blowdown and reflood has been calculated specifically for Waterford-3. The blowdown calculation was performed using NSSS system design data generic for all 34XX Mwt reactor plants (San Onofre 2 and 3, Waterford-3, and Pilgrim 2) with the exception of safety injection tank flow resistance factors (K-factors). Based on results from Waterford-3 SIT blowdown tests, K-factors specific to the Waterford-3 SIT blowdcwn lines have been conservatively used for the CEFLASH 4A and COMPERC-II computer program cal culations.

I

\

l I

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

, s* .

WSES-FSAR-UNIT-3

' Flow - 4180 gpm (Maximum) ./

)

SS Temperature - -Pe F (Minimum) - l 6.2.1.5.3.5 Steam-Water Mixing The effect of mixing and condensation of containment steam with spillad ECCS water upon the containment pressure is calculated in the manner described in Section III.D.2 of Refarence 12. The effective ECCS spillage rate is shown as a function of time in Figure 6.2-30b.

6.2.1.5.3.6 Passive heat Sinks The surface areas and thicknesses of all exposed containment passive heat sinks are listed in Table 6.2-7. To conservatively maximize the heat transfer to these passiv'e sinks, the surface areas have bean assumed to be at the naxinum of their uncertainty ranges, and their thermal properties (conductivity and heat capacity) hava b*en maximized. The thermal proper-ties assumed for this analysis are:

haterial Thermal pinductivity Volumetric Heat Capacity

] '

(BTU /ht-f t-F) (BTU /ft'-F) l9 Paint - Contair. ment 1.67 26.5 Vrssal interior Paint - Containment 1.57 19.75 1 Vessel exterior -

Paint - Steel Structure 0.235 49.9 Paint - Concrete 0.156 ;47.1 Carbon Steel 25.9 53.57 Stainless Steel 9.8 54.0- 9 i

concrete 1.0 31.9 6.2.1.5.3.7 Heat Transfer to Passive Heat Sinks The condensing heat transfer coefficients between the containment atmo-sphere and the passive heat sinks have been calculated in the manner de-scribed in Section III.D.2 and Figure III.D.2-2 of Reference 12. The var- l9 istion of the condensing heat transfer coeffi:ients as a function of time is snown quantitatively in Figure 6.2-30c.

' 2.i.5.2.5

. cr-t:*----t P;cr- -- INERT 8 L .ulyCre:ented i- thie sube-cr 4n h~ b--n parf arc-d se eunin c^r

_,_._ ___._:____. h - .n._ c__,.:___ , n..-  %.,,_, w,. ,u___c___

C. 1.' ' __ '_f.. ._ll_', ~ 1 ~ill..,_.~_~3', 1 .. - -.,

__J_ '__' . _ ' m..__...,

C ll ! '_

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

_ .. ._ pur;in; dur:n; - rr:1 ^per.ti^- dit ~ccu- riy duri ng ' -..._11 .!

3 6,2-32 Amendment No. 9, (6/80)

Insert B 6.2.1.5.3.8 Containment Purge System The analysis presented in this subsection includes the effects of the containment purge system which is assumed to be operating at the time of the postulated LOCA. The purge system isolation vent linas are 48 inches in diameter. The butterfly-type purge system isolation valves are mechan-ically limited to a maximam open position of 40 . From this 40 open position, the purge system isolation valves are assumed to be fully closed 5.0 secor.dt after actuation of the ESF relays in response to the high con-tainment pressure (18 psia) signal. It is conservatively assumed that only 3 dry air is removed from the containment atmosphere.

4

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

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

WSES-FSAR-UNIT-3 7: . - - - : ;- c f th: :21: f : y:::. Th e ch:::: ef : ::rj ig;_;;:5-bili;,

-!Lnb :uch :: : p : : : ri__. ' ' C_t :: n A;; :::i n; purging :p::::i;n:

~

-':,- C ::c::, 5: :;;r :::11 :. -

6.2.1.5.4 Results -

For the most severe LOCA, the 0.8xDEG/PD break, the minimum containment l23 pressure response is shown in Figure 6.2-31a. As required by 10CFR50 9 Appendix K, the containment pressure used in the ECCS performance evalua-tion does not exceed this pressure (see Figure 15.6- ). The responses ll23 of the containment atmosphere and SIS (recirculation) ump temperatures are shown in Figures 6.2-31b and 6.2-31c , respectively. containment l9 pressure responses for breaks other than the worst bres are presented in the figures in Subsection 15.6.3.3. ,

19Z 6.2-33 Amendment No. 23, (11/81)

, s- TABLE 6.2- /7 WATERFORD 3 BLOWDOWN AND REFLOOD MASS AND ENERGY RELEASE DATA 0.8 DEG/PD 14 Ass fbl 6e gelesse T4(oLof dass A T4goL(%l &

(Sec) (b/sec') (5b/sec0 (I m) ('i!TD) o O O O o

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FIGURE 6.2-30B WATERFORD 3 0.8 x DE GUILLOTIiiE BREAK IN PUMP DISCHARGE LEG EFFECTIVE SPRAY AND SPILLAGE TO CONTAINMENT 20000 -, , , ,

ONE TA!!K SPILLING z 16000 - -

S u? \ '

\

i i a

\ /

EFFECTIVE ANNULUS SPILLAGE 8000 W EFFECTIVE PUMP SPILLAGE

• " N N/

V

////

CONTAINMENT SPRAYS 0

0 50 100 150 200 250

! TIME, SEC

I FIGURE 6.2-31A WATERFORD 3  ;

0.8 x DE GUILLOTINE BREAK IN PUMP DISCilARGE LEG MINIfiUM C0flTAlflilENT PRESSURE "

I i

L60.000 i I

i l 1 50 000 -- ; l 1  !

l i

^

40.CGC -

- -i  !

<K ' '

WA co -

c_  :

i

, i . t.

u.!

c_ 30.000 : --

,i 23 *

[ {

v) [,

en ~,

UJ i c: i  :

n_ --

l 1 .:

20.000 I

.---i i i

i 10.000 l

I 3

i i .

e l l

O.000 o o o a 73 -

o o a

• o o C) - *

• o a '

o a o a

• 'D

. C LO o a e

o - m  : o

ps p q = ;; v

, , ; .ta vs :.

a .: i .. a . . . . .t- '

c::

J .. u .

4 6 8

FrGURE 6.2-31B WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG TEf1PERATURE, CONTAINMEllT ATMOSPHERE 300 -

250 -

4 200 --

o u_ s\ N i

f 150 h

E 100 -

50 -

0 0 50 100 150 200 250 TIME, SEC

FreuRE 6. 2-31C WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG TEMPEPATURE, SUMP 300 -

250 -

f 200 -

si R 150 -

Ei 5

N 100 -

50- -

' ' ' ' i 0

0 50 100 150 200 250 TIME, SEC

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

WSES.?SAR-UNIT-3 The worst single failure of an active component for a large break is the failure of a LPSI pump. .For conservatism, it is assumed that the operating LPSI pump is connected to the broken leg and to one intact leg. As ex-plained above, the flow to the broken loop is the same as that to the in-tact loop.

In the Small Break LOCA Evaluation the pressure between the three (3) in-tact legs and the broken leg is no greater than about 5 psi. Therefore, all lines see essentially the same back pressure and the flow is split evenly between them.

The worst single failure for a small break is the failure to start of one diesel generator. Therefore, only one HPSI and one LPSI pump will operate.

It is assumed that the operating LPSI pump is connected to the broken leg and to one intact leg. As in the larFe break, the flow to the broken loop is the same as that to the intact loop.

6.3.3 PERFORMANCE EVALUATION 6.3.3.1 Introduction and Summary The ac.:eptance criteria for Emergency Core Cooling Systems for Light Water Ccoled Reactors are set forth in 10CFR 50.46 (Reference 1). The analyses presented in this subsection and in Subsection 15.6.3.3 demonstrate that the Waterford-3 ECCS desi F n satisfies theoc criteria.

~

The ECCS performance was evaluated for a spectrum of b eak sizes ranging from a full double ended guillotine break to a 0.01 ft3 brean. At 13.7 Kw/ft, thebreakyieldingthehighestpeakj}adtemperatureandlocalclad 23 ,

,, oxidation was identified as the 0.8xDEG/PD -

Nsce.T C, i

(a) DEG/PD = Double Ended Guillotine at the Pump Discharge. 23

~.

6.3-17 Amendment No. 23, (11/81)

Insert C The ECCS performance was reevaluated for the 0.8 x DEG/PD break at 13.4 Kw/ft using actual flow resistance K-factors for the SIT injection lines and assuming that the containment purge system is in operation at the time of the postulated LOCA.

l l

l l

WSES-FSAR-UNIT-3 The results of the ECCS performance analyses show that the plant meets the . S ['

10CFR50.46 Acceptance Criteria at a peak linear heat generation rate of j ll 23 M Kw/ f t . Conformance is summarized as follows:

13.4 .

Criterion (1) Peak Clad Temperature. "The calculated maximum fuel

- element cladding temperature shall not exceed 2200 F."

2itt-The analysis yielded a peak clad temperature of 4H+ F f23

.for the 0.8xDEG/PD BREAK.

Criterion (2) Maximum Cladding Oxidation. "The calculated total oxidation of the cladding shall nowhere exceed 17 per-cent of the total cladding thickness before oxidation."

The analysis yielded a local peak ' clad oxidation per- 23 centage of 6 for the 0.SxDEG/PD BREAK. l so criterion (3) Maximum Hydrogen Generation. "Ihe calculated total amount of hydrogen generated from the chemical reaction of the

- cladding with water or stesm shall not exceed one perceht of the hypothetical anount that would be generated if all of the metal in the cladding cylinders surtwnding the fuel, excluding the cladding surrounding the plenum ,

volume, were to react." ,,gg The analysis yielded a peak core-wide oxidation of- g3 for the irexDEG/PD BRFAK.

94 \

Coolable Geometry. " Calculated changes in core geometry .

Criterion (4) shall be such that the core remains amenable to cooling."

The clad swellinggd accounts rupture model which is part of the for the effects of changes evaluation model in core geometry if such changes are predicted to occur.

With these core geometry changes, core cooling was enough

to lower temperatures. No further swelling and rupture can occur since the calculations were carried to the point

[

at which the temperatures were decreasing. Thus, a cool-able geometry has been maintained.

! Criterion (5) Long-Term Cooling. "After any calculated successful intitial operation of the ECCS,' the calculated core temper-I' ature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

23

-3 J

I 6.3-18 Amendment No. 23, (11/81) 1

WSES-FSAR-UNIT-3

(. 15.6.3.3 Loss of Coolant Accident (LOCA) 15.6.3.3.1 Identification of Causes and Frequency Classification The estimated frequency of a IDCA classifies it as a limiting fault as defined in Reference 1 of Section 15.0. A LOCA is' defined as a hypothe-

. . tical break in a pipe in the reactor coolant pressure boundary resulting in the loss of reactor coo in excess of the capability of tha coolant makeup system.gt For at athis rate analysis, the particular breaks 9

assumed are described in Subsections 6.3.3.2.3 and 6.3.3.3.3.

4 15.6.3.3.2 Sequence of Events and Systems Operations i

The transient behavior during a LOCA is as follows. During the blowdown .

i phase, the primary system depressurizes as primary coolant is ejected Y; through the break into the containment, and the reactor is shutdown either by moderator voiding, or by CEA insertion. Following depressurizatica, ,

emergency cooling water is injected into the cold legs, flows into the '

downcomer, fills the lower plenum, and refloods the core. When the core has been completely recovereo, the long-term cooling mechanisms described in Subsection 6.3.3.4 will maintain acceptable core temperatures until the plant is secured.

! The sequence of important events which occur in the short-term is listed in ' .

Table 15.6-12 for large-break LOCA1 sad in Table 15.6-12a for small break 9 LOCAs. The sequence of events for long-term cooling is discussed in Sub-section 6.3.3.4.

l y, 15.6.3.3.3 Core and System Performance 15.6.3.3.3.1 Large Break LOCA

~ 2 4 t a.T

• g t '

l 9

15.6.3.3.3.1.1 Mat 'ematical Model

. yW f L .y TheAcalculations reported in this section are performed using the CE large break evaluatigmodel described in References 3 and 4. In the CE model, the CEFLASH-4A computer program is used to determine the primary system flow parameters during the blowdown phase, and the COMPERC-II(5) computer program is used to determine the system behavior during the refill and reflood phases. The core flow and thermogamic parameters from these two codes are used as input in the STRIKIN-II program, which is used to calculate the hot rod clad temperature trgnsient and peak local clad oxidation percentage. Except fo r the 0. 5 f t S/PD BREA the steam cooling heat transfer coefficients calculated by the PARCH code were used for the time interval during which the reflood rate was' less than 1.0 inch /

trans-second. For the 0. 5 f t S/PD BRpK, a minimum steam cooling heat #

fer coef ficient of 5.0 Btu /hr-ft -F was used for conservatism, as described in CENPD-132, Supplement 1 (Reference 3) for the same time inter-val. The STRIKIN-II version used is identified in CENPD-135, Supplement 5 (Reference 6). The core-wide clad oxidation pp3'reeg *igobtained from the results of both the STRIKIN-II and COMZIRC computer programs.

..m

, W SELT D 15.6-17 Amendment No. 23, (11/81)

Insert ***

The large break LOCA calculations described in the following subsections pertain to both a spectrum of large breaks and to a worst break (0.8 x DEG/PD) reanalysis. The spectrum analysis contains consistent base data, input, and mathematical models. The worst break reanalysis utilized containment purge data, actual SIT discharge line flow resistance K-factor input, and the latest C-E ECCS Evaluation Model Flow Blockage Analysis in Reference 13. Based on the spectrum results, which show that containment pressure and SIT injection play a relatively minor role in differences in PCT for different break sizes or break locations, it is concluded that only the worst break (0.8 x DEG/PD) need be reanalyzed due to the effects of containment purge and changes in SIT l

discharge line X-factors.

l l

I b

l

f Insert D The worst break (0.8 x DEG/PD) ECCS reanalysis reported in this section is performed using the C-E ECCS Evaluation Model Flow Blockage Analysis described in Reference 13. In this C-E model, new rupture temperature, rupture strain, and flow blockage models, adopted from NUREG-0630 (Referen e 14), are used in the STRIKIN-II and PARCH codes. Also the steam cooling heat transfer coefficients calculated by the PARCH code, for use during the less than 1.0 inch /second reflood rate time interval, are calculated using an explicit method for redistribution of steam flow around the blockage region, described in Reference 13. The core-wide clad oxidation percentage is obtained from the COMZIRC results of the spectrum analysis. Improvements in steam cooling heat transfer using the model of Reference 13, reduced the amounts of cladding oxidation such that the results of the spectrum analysis are bounding.

. WSES-FSAR-UNIT-3

~

REFERENCES (Cont'd) m 4

7. CENPD-138, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant fleatup" (Proprietary). d CENPD-138, Supplement 1, "PARCll, A FORTRAN IV Digital Program to Eval-unte Poel Boiling Axial Rod and Coolant Heating" (Modifications),

Feoruary, 1975.

CENPD-138. Supplement 2, " PARCH, A FORTRAN IV Digital Program to 9 Evaluate Pool Boiling, Axial Rod and Coolant Heatup", January,1977 (Proprietary).

8. J. J. DiNunno, et. al ., " Calculation of Distance Factors for Power and Test Reactor Sites , " TID-148a4, Division of Licensing and Regulation, AEC, Washington, D.C. ,1962.

9. " Calculative Methods for the CE Cmall Break LOCA Evaluation Model",

CENPD-137, August, 1974 (Proprietary).

" Calculative Methods for the CE Small Break LOCA Evaluation Model",

CENPD-137, Supplement 1, January 1977 (Proprietary).

10. Letter, 0.D. Parr (NRC) to F.M. Stern (CE), June 13, 1975. 9 11 Letter, K Kniel (NRC) to A.E. Scherer (CE), September 27, 1977.

12 ANSI N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", 1973. g 1

13. h I-P to LD- fi- ois , ' C- E EC45 b #"-t-1 M JM  % 0- f , # = k y l%tI(P ).

Iq, f), A . Pm a- JL fl. G. ,

CJL: ' ' h J- a~. t- f MM L.o c. A b___ 0 ,

mj st-( 40 AErr -06 3o , j m o. -

s' 15.6-24 Amendment No. 9, (6/80)

WSES-FSAR-UMIT-3

- TABLE 15.e-12 TIME SEOUENCE OP IMPORTl;;T C'.'ENTS FOR LARGE LOCA (SECONDS AFIF.R BREAK)

SI Tanks End, of Start of SI Tanks SI Pumps Hot Rod Bypass Rcflpod Empty On Ru p t t.. e Break On 1.0 DES /PD(a) 13.7 19.8 32.S 76.1 76.1 63.1 0.8 DES /PD 13.7 20.0 33.0 76.3 76.3 63.3 0.6 DES /PD 14.8 21.2 34.3 77.6 77.6 72.9 0.5 ft 2S/PD 134.0 142.88 155.7 201.3 201.3 220.5 23 1.0 DEG/PD 13.6 19.7 32.7 76.2 76.2 60.5 0.8 DEG/PD 13.9 20.1 33.1 76.5 76.5 59.0 0.6 DEG/PD 15.5 21.9 35.0 78.3 78.3 70.9 n,

(a) See Table 15.6-15 for an explanation of these abbreviaticas.

/~)

,j ( O LO M u e .a A c ,,p_? .. +

• .JL =M 1 srr I

%h y A .. g- .

i 1

_. o.e ocG/rn 13.3 19.9 *s.t ias.3  : 3. 3 m. ,

l l l

l 15.6-39 Amendment No. 23. (11/81)

WSES-FSAR-UNIT-3 e .

TABLE 15.6-13 GENERAL SYSTEM PARAMETERS AND INITIAL CONDITIONS (LARGE BREAK ECCS ANALYSIS)

Quantity value Reactor power level, MWt (102% of nominal) 3,458 e 23 Average linear heat rate, kW/ft (102% of 5. 6 -

nominal)

Peak linear heat rate, kW/ft is.4 13.7.

23 Cap condugtance at peak linear heat rate (*' t ,*6ot. 1,447 Btu /hr-ft F Fuel centgine temperature at peak linear J,319 7 3,377.8 ll 23 heat rate F Fuel averg temperature at peak linear z.,is t .4 2,157.6 ll 23 heat rate F Hot rod gas pressure *), psia 1,113.3 23 Moderator temperaturd coefficient at initial +0. 5 x 10 4 density, ap /F l23 System flowrate (total), ib/hr 148.0 x 10 0 Core flowrate, Ib/hr 144.15 x 10 6 23 Initial system pressure, psia 2,250 Core inlet temperature, F 557.5 Core outlet temperature, F . 618.6 Accive core, height, ft. 12.5 Fuel rod OD, in. 0.382 Number of cold legs 4 Number of hot legs 2 Cold leg diameter, in. 30 Hot leg diameter, in. 42

a. These quantities correspond to the burnup (676 MWD /MTU, hot rod 23

, average) yielding the highest peak clad temperature.

~5

Q 15.6-40 Amendment No. 23, (11/81)

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

WSES- FSAR-UNIT- 3 TABLE 15.6-13 (Cont'd)

- Quantity value Safety injection tank pressure, psia 609 Safety injection tank gas / water volume, ft $99/1,679 57A.

e A

i

(

l

. . .s l _

l 15.6-41 L

WS IS- SAR-UNIT-3 9 AB I 15.6-14 i

PEAK CLAD TFMPFRAT G$S AND OXIDATION PERCENTAGES FOR THE LAEGE BFFAK SPECTRUM Peak Clad Clad Oxidation -

Tenipera ture(a ) (g) ' '

23 Break ( F) Local Core-Wide ~

1.0 DES /PD 2,107 16.2 0.787 0.8 DES /PD 2,108 16.2 U.796 0.6 DES /PD 2,092 15.3 0.740 0.393 23 0.5 ft S/PD 2,049 14.2 1.0 DEG/PD 2,115 16.6 0.805 0.8 DEG/PD 2,118 16.7 0.805 0.6 DEC/PD 2,094 15.1 0.685 .

Y e .

(a) Acceptance Criteria is $2200 F 23 (b) Acceptance Criteria is $17%

(c) Acceptance Criteria is $1.0%

(d) h)M Lu.A. ey 30.

• _, A e,.p _+ ,,,4 m ;y: ; 5 2 7 y a p g A y c -

6-m t_p . .

cal

- c.f DEG /rp 2)f f f. T 4-0.f05 i

i l

l i

15.6-42 Amendment No. 23, (11/81)

I

WSES-FSAR-UNIT-3 TABLE 15.6-15 LARGE BREAK SPECTRUM Break, Size, Type and Location Abbreviation Figure 1.0 x doubic-ended slot break in 1.0 x DES /PD 15.6-36 through pump discharge leg 15.6-44 0.8 x double-ended slot break in U.8 x DES /PD 15.6-45 through pump discharge leg 15.6-53 0.6 x double-ended slot break in 0.6 x DES /PD 15.6-54 through pump discharge leg 15.6-62 23 2

0.5 ft slot break in pump discharFe 0.5 ft S/PD 15.6-83 throuFh leg 15.6-91 1.0 x double-ended guillotine , break 1.0 x DEG/PD 15.6-92 through in pump discharge 1er 15.6-100 0.8 x double-ended guillotine break 0.8 x DEG/PD 15,6-101 through in pump discharge leg 15.6-109k 23

~~

0.6 x double-ended guillotine break 0.6 x DEG/PD 15.6-110 through

. in pump discharge leg 15.6-118 Peak clad temperature vs. break 15.6-128 area

^ ^^ ^ ^

_- ^

_- ^

%9,

, [^. ^ _ _ _ _ _

oga M .. 2. A . .'.00 . % M o,3 = occc /p, is, c. - sp6

?

(4 i A 15. 4 - 2.o 5 t

a ,gg 3,7 (4 WJ u - i.7 = .A Ak _ + +() . . - - _Gl

.% % y e% K-p .

15.6-43 Amendment No. 23, (11/81) o

.o FIGURE 15,6-128 -

PEAK CLAD TEMPERATURE vs. BREAK AREA 2200 -

6 l O DISCHARGE LEG SLOT BREAKS 2160 _

13.7 KW/FT.(w/o containment purging)

--- O DISCHARGE LEG GUILLOTINE BREAKS O 6 DISCHARGE LEG GUILLOTINE BREAK 13.4 KW/FT cre e K a tor) d (W 2120 -

/

O 5

r o

2080 -

i

! W 2040 -

2 0.5 FT 0.6 DE 0.8 DE 1.0 DE 2000 l

! 0 2 11 6 8 10 12 14 BREAK AREA, FT 2 l

I FIGURE 15.6-186

~

WATERFORD 3 0'.8 x DE GUILLOTINE BREAK Ifl PUMP DISCHARGE LEG CORE POWER 1 2001 .

1.0000 1

8000 ,

oc uJ. -

Z o

Q_

6000 1

F--

o l I-l

.4000 2000 0 000% o a a a a 8 8 8 8 8 8 9 9. 9 9 9 9 0 - a m e m TIME IN SEC

B 1

FIGURE 15.6-137 .

llATERFORD 3

~

0.8 x DE GUILL0TIllE BREAK IN PUMP DISCHARGE LEG-i PRESSURE IN HOT ASSEMBLY N0DE e

2400.0 2000 0

, G 1600.0 k c

c-i uJ oc a

W 1200 0 N -

m Q_

r 800 0 ,

. .e \ .

e 0.0 a o o a 8 8 8 8 8 8 2 2 5 d 5 N TIME IN SEC

FIGURE 15.6-188 WATERFORD 3 3.8 x DE GUILLOTIliE BREAK Ili PUf1P DISCllARGE LEG LEAK FL0ll 120000 100000 PLNP SIDE o --- REACTOR VESSEL w SIDE en S

80000 w .

W C

E -60000 2

o

_J

\

40000 \

\.

\..

N N

~20000 \

s s

h N

\

CK -

s ' -

0 g O a g 8

o 8 8 8 8 8

, a . . . .

l 0 s I s 8 8 TIME IN SEC

.-_ - - ~

FIGURE 15,6-189 .!

WATERFORD 3 0.8 x DE GUILLOTIHE BREAK IN PUMP DISCllARGE LEG FL0ll IN ll0T ASSEMBLY - PATH 16, BELOW HOT SPOT 4

30.000 .

20.000 l

u to u>  !

m 10.000 I

'h tu l--

CE E 0 000

' ^

Nf 8

_J LL

-10.000 f

a l

i

-20.000

-30.000 a a -o a 8 8 8 8 8 8 0 0 5 0 b b TIME IN SEC

FIGURE 15.6-190 UATEP. FORD 3 0.8 x DE GUILLOTIllE BREAK Ill PufIP DISCllARGE LEE FLOW IN 110T ASSEf1BLY - PATil 17, ABOVE HOT SPOT i

30.000 g I

20.000 i u .j t.u en .

10.000 -\

ca

-.a

&h)V L1.2 .

w C

E O'000

~

l

\f 8

_a L1_

-10.000 ,

- L

[

-20.000

! -30 000 -

!6 !J  % N E $

l l

TIME IN SEC

FIGURE 15,6-191 IIATERFORD 3 0.8 x DE GUILLOTIllE BREAK Ill PUfiP DISCHARGE LEG HOT ASSEilBLY QUALITY __

1.0000 ,

7 i j l ge i

! I lI f 1 I i 1 .i 'I I

8000  ; f .

7 l I l l 1 1

-l 1

' I I \

I I

'I

\ ";

! 'h s

6000 I

j j./

,j 1

}\! .).

0

.4000 l I

. I I \

l I

.2000 j l NODE 13, BELOW HOU EST REGION l ---NODE 14, AT HOTTEST REGION

- -N0DE 15, ABOVE HOTTEST REGION 0 0000' o a .c a 8 8 8 8 8 8 0 U 5 N N N TIME IN SEC

! FIGURE 15.6-192 l!ATERFORD a >

0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG l .

C0f1TAINf1ENT PRESSURE l t

I i

i I

-60.000 i  !

i l  !

i 50 000 --

40.000 0, -

m l

u.I

! cc 30.000 -

l D '

w I w -

$ N _ _

i 20 000 , -,  !

l i  !  :

I i

' I 10.000 i, i

I

' l '-

i l'  :

l 0.000 o o a o o

o o o o o o . . . . ,

o o -

o o o o ,

..o o a: a u, I

i o -4 M .* LO E-i

~

p. ,

[$ Q C C.fi?/ C C "'

,i I */.t.

.~. . { ~. [.1 6 s*..: e o_L L- -_ ___

r 4

FIGURE 15,6-193 ilATERFORD 3 O.8 x DE GUILLOTINE BREAK IN PUi1P DISCl!ARGE LEG MASS ADDED 10 CORE DURIllG REFLOOD l

150000.- " -

i' 125000 TIME, SEC REFLOOD RATE .

0.0 - 4.8 2.070 IN/SEC 4.8 - 51.6 1.410 IN/SEC

'51.6 - 600.0 0.654 IN/SEC

$ 100000 .

I b

S o

• 75000 '

g -

8

$ 50000 ,

l 250C0 I

, l .

I ,

i .

i

/

/ .

l i . <

l ',-

/ i o

i o

! i j o o o i o o. a o o o .

a

, o o o o o .

- ce -

u) m o o - a m v e

[ IIME 5. TIE.R CONICCT, SEC ,

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

FIGURE 15.6-194 IIATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE 2200

[' N.  !

2000 ,

1800 , f 1600 O

d 1400 l i=

E l i g .

W 1200i  ; 4 s

= 1 5

u 1000 800 60-0 400 t O' 100 200 300 400 500 600 t -

FIGURE 15,6-195 UATERFORD 3 O.8 x-DE GUILL0TIllE BREAK Irl Puf1P DISCHAREE LEG ilID Afi!iULUS FLOU 15000.

10000 1

co gd li s

_.a 5000 f

tu F-C i e 0 --

1

., o

-5000 3

-10000 l ! Y I

-15000 O a a a 8 8 8 8 8 8

? E b b b -

TIME IN SEC

k FIGURE 15.6-196 ilATERFORD 3 0.8 x DE GUILLOTIf!E BREAK Ill PUMP DISCllARGE LEG OUALITIES AB0VE AND BEL 01! CORE 1.0000 g i r.

l I pl

\ l I i 1 I I /

1 I

' g/

8000 l II I I I I )

l ,

0 '

>- 6000 j  ! ,

I D \ l

\

l '

' I \

h' v a I U I

.4000 I y ,

I .

I I

2000 .

I l ABOVE THE CORE l r

/g

/ - - - BELOW THE CORE

\

/  :

0.0000 I 'I '

a 2 a a 8 8 8 8 8 8 i 2 0- 5 d 5 N  !

TIME IN 5EC L -

1 FIGURE 15.6-197 WATERFORD 3 0.8 x DE GUILLOTIflE BREAK Ifl PUfiP DISCllARGE LEG CORE PRESSURE DROP

(

30.000 .

20.000 G

n.

w 10 000 f x A 1 5

w Y \'

\

~

\n O.000 , s jv F- Il

_a W

.! o -10.000 t

-20.000

-30.000 a o .a 0 8 8 8 8 8 8 a 3 5 d 5 N TIME IN SEC v- -- - - , , , , - + n n- ,-. ---e--- n- +- -

FIGURE 15.6-198 WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG SAFETY INJECTION FLOW INTO INTACT DISCHARGE LEGS 6000 -

5000 -

B R

$ 4000 -

i S

u_

E 3000 -

t3 w

2

$ 2000 -

i 1000 -

0' 0 40 80 120 160 200 TIME, SEC

r ' --

FIGURE 15.6-199 lIATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PURP DISCHARGE LEG l ilATER LEVEL IN DOWNC0flER DllP,ING REFLOOD I

[_ _

l 18.000 ^ '

i i i l~ ( {  ! i  !

'5.000 -

12 000 w

u_

_.3 y

J 9 000 g i  ;

l l

i a:

ua

' il t

H.

I 2 .

l 6.000 )-

i .4 i e i

1 f I i

' e

!. i  !  ! i 3;C00  ;

j j ,

8 l  !

. 5 i i l  !

! 0.00C a o o o a  ;

a o* o o o i

' o ' ~ * ' '

o o. o o o o' '

e. y e e o O ~ N " v "-

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

l fgg~_ f* I' ! 7_ L U ' *4 I 't 'v ' ' )c .-I. b I *

.' Fisuas 15.6-200 WATERFORD 3 1800 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT SPOT GAP CONDUCTANCE 1600 1400 1

1200 O

f 1000 t

E s

E5

, 800 d

E t3

, g e g 600 -

a k

e i 400 l -

r j i

! A ' '

200 f N l

'f

! 0 0 100 200 300 400 500' 60C TIME, SECONOS

--w gw-,-e a- ---T - " -

m+- - '+~ ~$ - e' e '. -- w-'

, .~,

FIGURE 15,6-201 WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUf1P DISCHARGE LEG ~

8- LOCAL CLAD OXIDATION 16 14 12 s 10 2

S

% /

0 3 ,,

l 0

6

.f .

4

/

2 0 .

O' 100, 200 300 400 500 SCO TINE, SECONDS

. ,so FIGURE 15.6-202 o WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG 2700 -

CLAD TEMPERATURE, CENTERLINE FUEL TEMPERATURE, AVERAGE FUEL TEllPERATURE AND C00LANTiTEMPERATURE FOR HOTTEST N0DE

' l l 2400 FUEL CENfERLINE a~

s AVEilAGE FUEL 2100' -

x 1800 CLAD '\ Nb f ss

~

O 1500~

L' y

f X_

ui

$1200,I '

E i i

I 900 I

600 h .

COOLANT 300 O --

0 100 200 300 400 500 600 TIME, SECONDS

. - .. FIGURE 15.6-203 o WATERFORD 3 0.8 x DE GUILLOTINE BREAK lii PuilP DISCHARGE LEG HOT SPOT HEAT TRANSFER COEFFICIENT 100 9

O 160 i .

140 u_

j 120 w

E s

EE 100 s

5 C .

II. -

b 80-S l 5 1  %

z g 60 r

m  !

! 40 .

l ,!

f El

! si l ..

I

, -./

I l

\

- i O -

0 100 200 300 400 500 600 l TINE, SECONDS . _

FIGURE 15.6-204 WATERFORD 3 0.8 x DE GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT R0D INTERNAL GAS PRESSURE 1200 -

PINITIAL = 1113.3 PSIA 1000 -

- RUPTURE = 43.12 SEC 800 1

g 600 -

a u

c_

400 l

200 -

i i

' ' ' ' ' i 0

0 20 40 60 80 100 120 i

TIf4E, SEC l

l . . - - - _ - . __

< < %).

FIGURE 15.6-205 WATERFORD 3 0.8 x DE GUILL0TIllE BREAK Ill PUMP DISCllARGE LEG CORE BULK CHAllNEL FLOW RATE 30000 ,

20000 lj CORE INLET w k --- CORE EXIT en ll s 11 m 10000  !

\

\ /

$ 0-

~

l J '

u

-10000. p' BULK CHANNEL REPRESENTS 98%

OF TOTAL CORE FLOW AREA

-20000.

-30000 O a o a 8 8 8 8 8 8 s 5 5 5 N TIME IN SEC