ML20041C396

Large Break Loca/Eccs Performance Results W/Addl Plugged Steam Generator Tubes

ML20041C396
Person / Time
Site:	Millstone
Issue date:	02/28/1982
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20041C395	List: B10425, Forwards Large Break Loca/Eccs Performance Results W/Addl Plugged Steam Generator Tubes ML20041C396
References
TAC-47355, TAC-47389, TAC-47473, TAC-47577, NUDOCS 8203010277
Download: ML20041C396 (27)

Text

_-

Docket No. 50-336 Millstone Nuclear Power Station, Unit No. 2 Large Break LOCA/ECCS Performance Results with Additional Plugged Steam Generator Tubes February, 1982 8203010277 820219 PDR ADOCK 05000336 P

PDR

The Loss of Coolant Accident (LOCA) has been reanalyzed for Millstone Unit 2 with 9.4".; (800 tubes) steam generator tube plugging.

The following

)

information amends the Safety Analysis Report section on Major Reactor Coolant System Pipe Ruptures.

The results are consistent with acceptance criteria provided in Reference [1].

The descriotion of the various aspects of the Westinghouse LOCA analysis methodology is given in Reference [2].

This cocument cescribes tne major pnencmena modeled, the interf aces among the computer cooes, anc the features of the codes which ensure ccmpliance with tne Acceptance Criteria.

The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV cooes wnicn are used in the LOCA analysis are describec in detail in References [3]

through [6]; code modifications are specified in References [7] througn

[13].

These codes are used to. assess the core heat transfer geometry and to determine if the core remains amenable to cooling througnout ano subsequent to the blowdown, refill, and refloca phases of the LOCA.

The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS curing blowdown, and the_WREFLOOD computer code is used to calculate this transient Wring the refill and reflood phases of the accident.

The COCO computer code is used to calculate the Containment pressure transient throughout the LOCA analysis.

Jimilarly, the LOCTA-IV compu-l ter code is used to compute the thermal transient of the hottest fuel rod iring the entire analysis.

SATAN-VI is used to calculate the RCS pressure, enthalpy, density, ano the mass and energy flow rates in the RCS. as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA.

SATAN-VI also calculates the accunulator water flow rates and internal pressure ano the pipe break mass and energy flow rates that are assumed to be vented to tne Containment curing blowdown.

At the end of the blowdown pnase, these l

1

i

ata tre 'trintf erm :: the AREFLOOD c ce.

The mass anc energy release

-nes carinc Dic<,cawn ire utilizec in :ne COCO :oce for ;se in tne

stermination of :ne Containment pressure response caring nis first

nase of the CCA.

Acaitional SATAN-VI ou tput cata incluaing tne care "cw rates anc entnaipy, tne core pressure, anc tne core power aecay transient, are transferreo to the LOCTA-IV coae.

With initial information from the SATAN-VI coce, WREFLOOD uses a system thermal-hydraulic' moael to determine the core ficoaing rate (i.e., tne rate at which coolant enters the bottcrn of the core), the coolant pres-i sure and temperature, and the core aater level curing the refill anc reficod phases of the LOCA.

WREFLOOD also calculates the mass ano energy flow audition to the Containment througn the creak.

Since tne mass flow rate to the Containment depenas upon tne core flooding rate and the local core pressure, which is a function of the Containment backpressure, the WREFLOOD and C0C0 codes are interactively linkea.

WREFLOOD is also linked to the LOCTA-IV code in that thermal-hydraulic I

parameters fran WREFLOOD are used by LOCTA-IV in its calculation of the fuel temperature.

LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel clad ternperature and metal-water reac-tion of the hottest rod in the core.

l The analysis presented here was performed with the 1981 version of the evaluation model which includes the NUREG-0630.

This evaluation model was formally approved by the NRC on December 1,1981.D3] Reactor Coolant pumps are assumed to continue to run during blowdown unless otherwise noted.

2

^

esu ::

~he ina'.js:s :f tne icss ;f ::cian; at:icen is :erf:rmec at :.02 :ercen:

f :ne licensac : re :cwer rating.

The :eax linear ;cwer anc :::ai : re

cwer usec in the analysis are given in Tacie 2.

Since there is margin

etween the value of ;eak linear power density used in this analysis and the value of the peak linear power censity expected curing plant opera-tion, the peak claa temperature calculated in this analysis is greater

nan the maximum ciao :Ex erature excec:ec :o exist.

Taole 1 presents the occurence time for various events througnout the accident transient.

Table 2 presents selected input values and results frcm the hot fuel red thermal transient calculation.

For these results, the hot spot is defined as the location of maximum peak clad temoeratures.

That loca-tion is specified in Table 2 for the worst break case analyzed. The location is indicated in feet which presents evevation above the bottom of the active fuel stack.

I Table 3 presents a summary of the various cantainment systems parameters i

and structural parameters wnicn were used as input to the COCO c:mouter codeN usec in this analysis.

i 4

t Cigures 1 througn la present tne carane*.ers of orincical interes*.

1. ccm tne large creak ECCS analysis.

The following items are noted:

Figure 1:

Hot spot clad temperature.

Figure 2:

Coolant pressure in the reactor core.

Figure 3:

Water level in the core and downcomer during reflood.

l Figure 4:

Containment pressure transient Figure 5:

Core flow during blowdown Figure 6:

Fuel rod heat transfer coefficients.

Figure 7:

Hot spot fluid temperature.

Figure 8:

Mass released to Containment during blowdown.

t.gure 9:

Energy released to containment during blowdown.

Figure 10:

Fluid quality in the hot assembly Figure 11:

Mass velocity Figure 12:

Safety injection tank water flow rate into RCS during blowdown (per tank).

Figure 13:

Pumped safety injection water flow rate during reflood.

Figure 14:

Core reflooding rate.

l 4

in_c;3sient

.7 ema : W vsi:

r :rea.<3 ;

inc :nc:ucing :ne :Cucie encec severance :f a reactor

an: : ice, the Emergency Care C:oling System vill teet ne Acca::ance Criteria as presented in 10C. R50.4. ' 13 That is:

\\

U 1.

ne calculated peak clad terccerature does not exceed 22CO F based On a peak core linear po'.ser of 15.6 kw/ft.

c.

.ne amount or r.uei etement claccing that reac:

chemica iy alth water or steam dces not exceed 1 percent of the Octal amount of Zircalloy in the reactor.

i 3.

Tne clad temoerature transient is terminated at a time wnen the : re gecmetry is st:ll amenable to cooling.

ine clacding oxida: Ton limits of 17% are not exceeded during or af ter quenching.

I 4.

The core temperature is reduced and decay heat is removed for an i

extended period of. time, as required by the long-lived radioactivity renainina in the care.

r I

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.m. c..um;er ;, sanucry a, E' r a.

2no :oraan T.

A.,

"Westingnouse ECC5

(

2.

?:-;eion, M.,

Massie, 4 va:uaticn *'ocel - Summary," WCAP-6229, July 1974 t

3.

Soroeion, :.

M.,

et 31., " SATAN-,'I Progr am: Comprenensive Space-Time j

nalysis of _]ss of Cociant," WCAP-8302 (Proprietary) ano Decencent a aCAP-8206 (Non-Proprietary), June 1974.

J.

<elly, R.

D., et al., " Calculational Mocel for Core Refloocing after a Loss of Coolant Accicent (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-3171 (Non-Proprietary), June 1974 5.

Bordelon, F. M. ano Murpny, E.

T., " Containment Pressure Analysis Code (CGCO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 6.

Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Tran-sient Analysis," WCAP-8301 (Proprietary) ano WCAP-8305 (Non-Proprietary), June 1974 Ferguson, K. L., and Kemper, R. M., ECCS Evaluation Mooel for 7.

Westingnouse Fuel Reloaos of Comoustion Engineering NSSS, WCAP-9526 (Proprietary) ano WCAP-9529 (Non-Proprietary), June 1979.

3.

Ferguson. <.

L.,

ano,emoer, R.

M., Aodenoum to ECCS Evaluation Mocel for Westingnouse Fuel Reiouas of Comoustion Engineering NS$5, OctoDer 1979.

n al., "',est a.gnouse ECCS Evaluation Mocel - Sup-I 9.

Earcelon, F.

M.,

et piementar. Inf orma tion,".,cAF -04 71 ;Proorietary ) ano 4 CAP-8472 f ilon-3~:orier. ry, lori

.975.

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4CAF-i621

;r:etary, Inc aCAF-5622 icn ;nor t etar:.,

.c.e ter i.~ 7 5.

'.1..etter ;S-CE-924, catec Janu ary 23, 1976, C. Eicneloinger ;oesting-nou se) to 3. 3. '!assallo (t;RC).

12. Eicneicinger, C., '"..estingncuse ECC5 Evalu ation v.ocei,, Feoru ary 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-P-A (Non Prcorietary '!ersion), Feoruary 1978.

13.

fluclear Regulatory Commission Letter, James P,. Miller (l'RC) to E. P. Rahe (Westinghouse), " Acceptance for Referencing of the 1981 Version of the Westinghouse Large Break ECCS Evaluation Model," December 1,1981.

TABl.E 1 LARGE BREAK TIME SEQllEttCE OF EVEtlIS C =0.6 DECI.G D

(Sec)

SIART 0.0 0.69

5. 1. signala ci) 5.

1. Iank injection 15.7 21.7 foil of Blovulnwn 34.4 HotLam of Core Recovery S.

1. Tank Empty 64.2 21.7

t. net or Ilypass a r ritai cont ainment pressure sensor

I l Alit.l 2 LARGE BREAK C[)=0.6 [l[CLG H esu l ts Peak Clad Teirip. ~F 2045 Peak Clad Location,Ft.

7.0 I oca l ir/110 Rxn(iiiax) %

4.4 2

7.0 Ir/Il 0 Location,Ft.

1ocai 2

<0.3 iota i Ir/110 Rxn, %

2 ilot Roil thirst Iime, sec 28.8 llot ltod thirst Locat ion, FL.

5.7 Ca lcii t at iori As siiriipt. iosis 2700 l

h$$S Power,Mwt,102t of i'eak Core 1.inear Power, kw/ft 15.6 l

1

$.I. lant Actisatiori Piessiire, psia 215 3

S. I. T arik Water Voliiine, f t per tank 1080 I

l.. :.

iili.stCne. nit _

3 1r ' 'eisr5

. n ; a l " ~ d " ~. 2"y:'C3 1.331 s

. 5 -.;

.e'.

J'if,Clame

ntainment

nit a 20nciticns:

99 '

umicity 60*F

^ cntainment Temoerature 60*F Enclosure Bulicina Temoerature 40 F 3rcunc Temperature 14.7 osia Initial Pressure

nit'al Time for:

25 secanos 5 pray Flow 0.] seconos Fans (3) 14.J ieC nos Acaitional Fan Containment Spray Water:

50*F Temperature Flow Rate (Total, 2 pumps) 3300 gpm Fan Cooling Capacity (Per Fan)

Vapor Temoerature (*F)

Caoacity (BTU /Sec) 0.0 60 145 3360.0 165 5280.0 300 28800.0 350 32400.0 Containment Heat Absorbing Surf aces 1.

Surf ace Areas and Thicknesses Shell and dome - 71,870 Ft2 a.

(1) P aint - 0.003 In. (one side exoosed to containment atmosonere)

(2) Caroon steel - 0.25 In.

(3) Concrete - 3.0 Ft. (one siae exposea to enclosure cuilaing atmospnere)

Uniinea Concrete - 62,300 :t2 c.

(1) Concrete - 2.0 Ft. (one siae exposeo to containment atmosonere, one siae insulatea) c.

Galvanizec Steel - 120,000 Ft2 (1) Zinc - 0.0036 In. (one sice exoosea to containment atmosanere)

'2) Caroon steel - 0.20 :n. tone sice insulatea) 10

4.

,Is i C -

ni,.it;ne 3 3rimeters I n;ainment ;n vs:c3, 2 i'"~.ec '91n itee ' - 56,350 F-d

" 1100 - J.oU3

'.n.

.Une ilCe exoOscG 10 CDnta1G"enE itmossnere; El Caroon steel - 0.2 In. i.one slae insulatea) e.

3aintec Steel - 32,500 Ft2 C-j C lint - 0.003 In. (one siae exposea to containment itnoosnere)

(2) Caroon steel - 0.26 In. (one sice insulatea)

Paintec 3 teel

'2,125 et2 (1) Piint - 0.002 :n. t one sice exposea to containment itmoschere)

(2) Carcon steel - 0.36 In. (one sloe insuiatea)

Painted Thick Steel 4,230 Ft2 9

(1) Paint - 0.003 In. (one sice exposeo to containment atmosphere)

(2) Carbon steel - 2.94 In..(one siae insulatea) h.

Containment Penetration Area - 3,000 Ft2 (1) Paint - 0.003 In. (one sice exposec to containment atmosphere)

(2) Caroon steel - 0.75 In.

(3) Concrete - 3.75 Ft. (one siae exposeo to enclosure aulicing atmospnere) i.

Stainless Steel Line Concrete - 8,340 Ft2 (1) Stainless steel - 0.25 In. (one sice exposeo to containment atmosphere)

(2) Concrete - 2.0 Ft. (one siae insulatea) j.

Base Slab - 11,130 Ft2 (1) Concrete - 8.0 Ft. (one siae exoosea to containment sumo, one siae exposeo to grouna) k.

.eutron Shiela - 1400 Ft2 (1) Stainless steel - 0.024 Ft. (botn siaes exposeo to containment itmosonere) 1.

CEDM Cable Support Structure - 1330 ft-(1) Paint - 0.006 In.

(2) Stainless Steel - 0.1094 f*

'noth et ex: esad to cantainment atmospnere) 11

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