ML20072L710

Forwards BNL 821213 Memeo Re Degraded Core Accidents at Facilities,Per Task III.3 Defined in Design Basis for Hydrogen Control Sys in CP & OL Applications. No-cost Extension to Contract Requested

ML20072L710
Person / Time
Site:	Perry, 05000000
Issue date:	12/20/1982
From:	Pratt W BROOKHAVEN NATIONAL LABORATORY
To:	Long J Office of Nuclear Reactor Regulation
Shared Package
ML20072L675	List: ML20072L681 ML20072L685 ML20072L689 ML20072L699 ML20072L710
References
CON-FIN-A-3389, FOIA-83-31 NUDOCS 8303310489
Download: ML20072L710 (19)

Text

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BROOKHAVEN NATIONAL LABORATORY ASSOCIATED UNIVERSITIES, INC.

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\. . . [ .

Upton Long Is!cnd. New York 11973 1

(516)232s Copodment of Nuc!act Energy FTS 666' 2630 j] ,

~~

O December 20, 1982

? . Igo -y e o Dr. John Long D Reactor Systems Branch .

? Division of Systems Integration a U.S. Nuclear Regulatory Commission ~

i; ail Stop P-1132 j Washington, D. C. 20555

Dear John,

Please find enclosed a copy of a BNL memorandum entitled, " Degraded Core -

Accidents in the Perry Power Plant," (J. W. Yang to W. T. Pratt). This memo-randum satisfies the milestone for Task III.3 as defined in the 189, Design Basis for Hydrogen Control Systems in Construction Permit and Cperating Li-cense Applications (FIN A-3389).

~

This cemorandum satisfies the final milestone under this contract (FIN A-3389) which was due for completion on December 31, 1982. A request has been made for a no-cost extension to this contract to allow us to prepare-final reports on our assessments of degraded core accidents in PWR Ice Conden- __

ser Plants and BWR Mark III Containment Plants. These reports will collect together all of the assessments performed for the different plants considered during the course of this contract. I will send you these reports at the end I of January 1983.

d

[j If you have any questions on the enclosed memorandum, please cont'act me or Dr. Yang.

Very truly yours,

-A R

L.i Trred. -

o

< Wm. Trevor Pratt, Group. Leader - '

R Accident Analysis Group f

@ WTP:tr '

Encl.

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y cc: R. A. Bari (w/ enclosure) (4.4) a R. E. Hall "

y W. Y. Kato p J. F. Meyer "

y, N. Lauben Fj . B. Sheron l:;- 8303310489 830224 PDR FOIA

, HIATT83-81 PDR L __

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'aROOKHMEN NATIONAL LABORATORY

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$ MEMORANDUM j~ ,

J DAM: December 13, 1982 r,. "

1 -

To: .W. T. Pratt FROM: J. W. Yang

SUBJECT:

Degraded Core Accidents in the

-) .

'j Perry Power P1 ant f

l The Perry nuclear power plant is of the BWR6/ MARK III type. It.is simi-l lar to the Grand Gulf and Clinton power plants, for which analyses of degraded core accidents have been performed at BNL.[1 The FSAR for the Perry plant contains comparisons of design parameters with the Grand Gulf plant.

For reference, the following three tables from the Perry FSAR are attached to this memorandum:

Table 1.3-1 Nuclear Steam Supply System Design Characteristics Table 1.3-3 Engineered Safety Features Design Characteristics -!

Table 1.3-4 Containment Design Characteristics  !

The rated power of the Perry plant (3579 MWt) is 93% of the rated power -

.. +

j of Grand Gulf plant (3833 MWt). The steam flow rate, feedwater flow cate, l

j number of fuel assemblies (i.e., masses of UO2 and Zircaloy 2) and other de-- [

sign parameters of the Perry plant are also about 93% of the corresponding

,, parameters of the Grand Gulf plant. In the events of partially or full core

-1

!

t meltdown, the production and release of hydrogen should be similar to that re-ported for the Grand Gulf plant.

]. -

The Engineered Safety Systems and containment constructions of the two h plants are basically the same. Both the Perry and Grand Gul f plants have the ~

%o same LPCS, HPCS, ADS, LPCI and'RHRS. The flow rates of HPCS, LPCI and RHRS of

1. [ the Perry plant are 95% of that of the Grand Gulf plant. The LPCI flow rate is, however, only 86% of that of. the Grand Gulf plant. The containment volume n

^

.. - - :- -...-.. ....--.:....-- . - - -- .. .. a - a =. -

a .

W~ . . Memo to W. T. Pratt 4* - - - . December >13, 1982 , -

g, Page 2

a .

e 7,

,t volume and the suppression pool ~ water volume are smaller in the Perry plant

!) compared with the Grand Gulf plant. But, the free ' air volume in the drywell j section of the Perry plant.is slightly larger (3%) than that of the Grand Gulf

{ plant. Because of th'e similarity between these plants, it is believed that

] the results of analyses reported in References [1-4] for the Grand Gulf and

.f ' the Clinton Power Stations are applicable to the Perry plant.

J

. }. - .

^

References ' -

1. Gasser, R.. "An Assessment of Postulated Degraded Core Accidents in the Grand Gulf Reactor Plant," BNL Draft Informal Report, (June 1982).

2. Gasser, R., " Analysis of Full Core Meltdown Accidents in the Grand Gulf Reactor P1 ant," BNL Draft Informal Report, (August.1982).

3. SNL Memo, J. W. Yang to W. T. Pratts "FARCH Analysis of Hydrcgen Burning During ' Degraded Core Accidents for the Clinton Power Station and the -

l Skagit' Power Station," (September 27,1982).-

l .,

4. Yang, J. W., "An Analysis of Hydrogen Combustion During Degraded Core

,j-Accidents in the Clinton Power Station," BNL Informal Report, to be l t- published.

i .

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i-TABLE 1.3-1 j C'OMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN Cl!ARACTERISTICS( )

PERRY GRAND GULF ZIMMER IIATCl! 1 - ,

BWR 6 BWR 6 BWR 5 llWR 4 I .

l 238-748 251-800 218-560 218-560 [

A. TlIERMAL AND 1!YDRAULIC DESIGN .

Rated power, MWL - . 3,579 3,833 2,436 2,436 I Design power, HWL 3,758 4,025 2,550 2,550 (ECCS design basis) '

I 6 -

6 6 6 t-Steam flow rate, Ib/hr 15.4 x 10 . 16.491 x 10 10.477 x 10 ' 10.03 x 10 ,

' 6

' 0 6 0 r Core coolant flow rate, lb/hr 104.0 x 10 .

112.5 x,10 78.5 x 10 78.5 x 10 i-w -

O 6- 6 6 6 i Feedwater flow rate, lb/hr 15.367 x 10 16.455 x 10 10.447 x 10 9.998 x 10 [t System pressure, nominal in 1,040 1,040 1,020 1,020  ;

steam dome, psia

• t.

Average power density, kW/ liter 54.1 54.1 50.51 . 51.2 Maximum thermal output, kW/ft 13.4 13.4 13.4' 18.5' II Average thermal output, kW/ft 5.9 5.93 , 5.40-7.11 Maximum heat flux, Btu /hr-ft . 361,600 -

361,600 354,255 - . 428',300 Average heat flux, Btu /hr-ft 159,500 159,800 144,032 164,410 I' Maximum UO2. temperature, F 3,435 3,435 3,325 4,380 I

m--------

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,; g- -, <-;; y s., 7 ,. , - .-.........>..wG".:,

s . x,

.a. ;. --

c.a - . , . - ,- .. ... . swL . . :- u - - - . . . -

@ $ @ }.

TABLE 1.3-1 (Continued) ,

PERRY GRAND GULF ZIta!ER I!ATCl! 1 BWR 6 BWR 6 BWR 5 BWR 4 -

238-748 251-800 218-560 ,

218-560 t

• r '

TIIERt!AL AND 1:YDRAULIC DESIGN (Cont.)

/ 2,130 2,781 Average volumetgic fuel 2,185 ,2,185 ,

temperature,, F j i

Average cladding surface 565 566 566 566 { i temperature, .F

?!inimum critical power ratio (tiCPR) 1.20 1.20 1.24 N/A

\' - . V Cool'nt a enthalpy at core inlet, 527.7 ,

527.9 (527.4 523.7  !'

Btu /lb .

+

1 u >

l- Core maximum' exit voids within 79.0 76.1 75 75  !

assemblies I s e Core average exit quality, % steam 14.7 14.6 13.2 .12. 7 "'

Feedwater temperature, F 420 420 420 387.4

\l Design Power Peaking Factor ,

~

llaximum relative assembly power 1.40 1.40 .

.1.40 1.40

! . j:

i. Local peakirig factor 1.13 1.13

• 1.24 1.24.

l Axial peaking factor - 1.40 1.40 , 1.4 1.50

i. j 2.21 2.21 ,' 2.43 2.60 Total peaking factor

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TABLE 1.3-1 (Continued)

PERRY GRAND GULF ZIMMER IIATCll 1 BWR 6 BWR 6 BWR 5 BWR 4 . t 238-743 251-800 218-560 218-560 ,.

1. B. NUCLEAR DESIGN (First Core) I .

Water /UO 2 y lume ratio (cold) 2.70 2.70 2.55 2.53

. n Reactivity with strongest <0.99 <0.99 ' <0.99 <0.99 f '

control rod out, k,gg )

Dynamic void coefficient at .

core average voids, %, and 40.95 -41.31 40.54 38.0 i rated output, C/% -7.16 .-7.14 . -8.57 -10.74  !'

Fuel temperature doppler coefficignt, .

{ cnd of cycle hot operating, *C -0.412 -0.396 -0.419 -0.40

1 i

Initial average U-235 enriclunent 1.90 1.70 1.90 2.34 t wt. % t i

Initial cycle exposure, 9,138 7,500 9,200 9,413 -

HWd/short ton . (Ave. Iirst core) .; ,

C. CORE MECIIANICAL DESIGN , L

1. Fuel Assetbly ,,. j Number of fuel assemblies 748 800 560 ,

560 .

Fuel rod array ,

8 x 8' 8x8 8x8 7x7 ,

l f

8

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.- -- ._T. . .i, 4. . ~. .

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-m.:e fg. @#v.. ~.

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i i ! TABLE 1.3-1 (Continued) ,

, PERRY GRAND GULF ZIMMER HATCll 1 i BWR 6 DVR 6 BWR S BWR 4 .

238-748 251-800 218-560

• 218-560 ,

i t' '

CORE MECllANICAL DESIGN (Cont.) '

1. Fuel Assembly (Cont.)

Overall dimensions, in. 176 176 176 176 Weight of UO2'Per assembly 456 458 .466 466 .

Ib (pellet type) (Chamfered) .

Weight.of fuel assembly, Ib 697 ) *, ,697 698 675 s .

2. Fuel Rods

'. \

Y Number per fuel assembly 62 62 , 63 49 as .

Outside diameter, in. 0.483 0.483 0.493 0 563 ,

..e Cladding thickness, in. 0.032 0.032 0.034 ,

0.032 Cap, pellet to cladding, in. 0.009 0.009 0.009 ,

0.0012 .

Length of gas plenum, in. 10 10 14 16 ,

q * . ,

Cladding material (free Zircaloy-2' Zircaloy ,

Zircaloy-2 Zircaloy-2 standingloadingtubes) -

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TABLE 1.3-1.(Continued) .

.l PERRY . GRAND GULF '~ZIMMER HATCll 1 i BWR 6 BWR 6 BVR 5 BWR 4 ,

238-748 251-800 218-560 218-560 CoxE MECHANICAL DESIGN (Cont.) {. . ,

3. Fuel Pellets Material UO UO UO UO 2 2- 2 2 Density, % of theoretical

{

95 95 .95, 95 -

Diameter, in. 0.410 0.410 0.416 0.487 Length, in. 0.410 0.410: 0.4i0 0.5 r 4. Fuel Channel ,

Y I N Overall dimension, length, in. 167.36 167.36 166.9 166.9 Thickness, in. O.120 0.120 0.100 0.080 Cross section dimensions, in. 5.45 x 5.46 5.45 x 5.45' 5.48 x 5.48 5.44'x 5.44 Material Zircaloy-4 Zircaloy-4 Zircaloy'-4 Zircaloy-4

5. Core Assembly c Fuel weight as UO2 , Ib 341,640 365,693- 260,551- ,

272,850 ,

Core diameter (equivalent), 'in. 185.2 191.5 160.2 160.2 Core height (active fuel), in. 150 150 .146 144' l 1

op .- . w . &

.- -s-

. x_... h'. ,u.-.:,d.

. w . ~ .u ' n .. . -

a =

O

@ @ j. ,

TABLE 1.3-1 (Continued)

PERRY GRAND GULF ZIMMER IIATCll 1 BWR 6 BWR 6 BWR 5 BWR 4 .

238-748 251-800 218-560 218-560 CORE MECllANICAL DESIGN (Cont.) I

6. Reactor Control System Method of variation of Moveable control Movable control Movable control Movabic contryl reactor power. rods and variable rods and variable rods and variable rods and vari 3ble forced' coolant forced coolant forced coolant forced coolant flow flow flow flow .

Number of movable , .

control rods 177 g* *'. 19 3-1

'137 137

- \.

u, Shape of movable '

s control rods Cruciform Cruciform Cruciform Crucif.orm Pitch of movable e , ,

12.0 12.0 12.0 12.0. - ~'

control rods Control material in B C granules 4

Bg C granules B4 C granules B C granules 4

movable rods compacted in compacted in compacted in , compacted in SS tubes SS tubes SS . tubes SS tubes ,

Type of control rod . Bottom entry Bottom entry Bott,om entry Bottom entry. -

drives , locking piston locking piston - locking piston locking piston Type of temporary Burnable poison; Burnable poison; Burnabic poison; Burnable poision; reactivity control gadolinia-urania gadolinia-urania gadolinia-urania gadolinia-urania for initial core fuel rods fuel rods fuel rods fuel rods

/

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4 *

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w . . . .

-.a..

. . ~ _ . . . -

. . ::~ ~w. , x . . . . + .c a .. w ... ............ .- -

. i O @' @o.I -

TABLE 1.3-1 (Continued) ,

. l

. PERRY GRAND GULF ZIMMER HATCl! I g

! BWR 6 BWR 6 BWR 5 BWR 4 '

l 238-748 251-800 218-560 218-560 i

+

CORE MECllANICAL ON / (Cont.) { ,

1

7. Incore Neut p ..,ggrymentation ,

j Number of incore neutron 164 176 - 124 124 i

detectors (fixed) f l 1

Nun.ber of incore detector 41 44 i 31 31 .i

. assemblies l

Total number of LPRM 164 .176 124 124  !

detectors  !

(J .

& Number of incore LPRM 41 44 31 31

. penetrations Number of LPRM detectors 4 4 4 4 ,

per penetration Number'of SRM penetrations 4 6 4 , 4 Number of IRM penetrations -

8 8 . '8 8 l Total nuclear instrument 53 58 ,

43 .

43 penetrations  ;

Source-Range -

Shutdown Through Criticality Monitor, Range 4 6 4 4 l Intermediate Range Prior to Criticality to Low Power ,

l Monitor, Range ,

8 8 8 8 '

j

=  !

.i

_. . =y ._ _,,,_ _ ,' ,,_.

• , y ; . +- .' v2^ ' _- 2_.m_

r t? "fe

-[. m U, . - .4-

- . , .y g j]hy, . , 3 g, ,~ g J Q g,g ; , , , y A g , ,.,,,,,,

, , , ,. , g']g*g , g y , y' _ * . , _J *.. i ,

l' ,

s

@ @ O. ..

TABLE 1.3-1 (Continued) i PEDRY GRAND GULF j ZIMMER llATCl! 1 BWR 6 DVR 6 BWR 5 BWR 4 238-748 251-800 218-560

• 218-560, t

CORE MECIIANICAL DESIGN (Cont.)

7. Incore Neutron Inst. (Cont.) ,,

Power Ra'nge Monitors, Range .

Approximately 1% Power to'15% Power 5

~

Local power 164

~

Range Monitors 176 124 124 Average power i* '. - -

8 i. 8 i 6 Range Monitors t

" ' 6 w .

.L . Number and type of 7 Sb-Be 7 Sb-Be 5 Sb-Be incore neutron . 5.Sh-Be sources \ #

D. REACTOR VESSEL DESIGN Material Low-alloy steel / ' Low-alloy steel / Carbon'st'ecl/

stainless clad stainless clad ' Carbon steel / ~

stainicss clad stainless clad.

Design pressure, psig -

1,250 1,250 -

1,250 1,250 Design temperature, F. 575 575

'575 .575

, Inside diameter, ft-in. 19-10 20-11 .

18 -2 .- 18 Inside height, ft-in. 70-5 70-10 . 69-4 69-4 i

b

~

') =. -

I

pes;'.TrJr %g g 3 _ __, 73.,; g , ,1,7

,. ,  ;;,, g ,c ,g,cf71 ',  :

,3C;7 g. W A ,j n _

i  ;

I .

i .- ,

TABLE 1.3-1 (Continued) i PERRY GRAND GULF ZIMMER HATCl! 1.

BWR 6 BWR 6 BWR 5 BWR 4 238-748 251-800 218-560 218-560

D.  ? .

REACTOR VESSEL DESIGN (Cont.)

Minimum base metal thickness 6.00 6.77 i

(cylindrical section), in. 5.375- 5.375 .

J Minimum cladding thic'kness, in. 1/8 1/8

~

1/8 1/8 E.

REACTOR COOLANT RECIRCULATION DESIGN I

Number of recirculation loops 2 2 2 2

,- Design pressure u .

.L Inlet leg, psig

- 1,250 1,250 1,250 1,148 Outlet leg, psig 1,650(2);1,550(3) 1,650(2);1,550(3) 1,675(2);1,575(3) 1,274 Design temperature, F 575 575 575 562 Pipe diameter,, in. 24 24

• 20 28 Pipe Material, ANSI -304/316 304/316 304/316 -

304/316 Recirculation pump flow rate, gpen 42,000 44,600 ' 32,500 45,000 Number of jet pumps in reactor 20

~

24' 20 20 ,

F. MAIN STEAMLINES Number of steamlines - 4 4 4 4 Design pressur'e, p,sig 1,250 1,250 1,250 1,118  !

p. l '

I

.: > ~ i e ~ A ' .a;g2O6-  ?.1;ckcgc.,,;.&,  :.2,,L,: , alii,_, . . . . ..::;

,'  ; i, &X3gg.rLL-ag gb3gzumyi_ :.*h W:& . .ii 1

.1 e .

e,' }-

.j .-

i k.

TABLE 1.3-1 (Continued) 4 PERRY GRAND GULF ZIt!MER IIATCl! 1 BWR 6 WR 6 BWR 5 BWR 4 .

238-748 251-800 218-560 . 218-5_60' F. HAIN STEAHLINES (Cont.) f Design temperature, F '575 575 575 560 .

Pipe diameterf in.i 26 28 , 24 24 p

. . )

Pipe material Carbon steel Carbon steel- Carbon steel Carbon steei k

NOTES: .

) ,. . .

• t

1. Parameters are related to rated power output for a sibgle plant unless oth,erwise noted.

w

,8 2. Pump and discharge piping to, and including, discharge block valve.

n -

3. Discharge piping from discharge block valve to vessel, s e O

F e

o b

e O

e r

l .

A e

~

i .

l > -

.v d&:.:%.CNfiitia:h u.u- . .s ' .... ..u'32 .;a. : . ~:.s L ., m < . ~ ,a :n~ -a%:.,.1k:.: ~ . A MLu A~'M"n'[

\ @ @

6 ), -

~k 2 l .

i i

i TABLE 1.3-3 l COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CllARACTERISTICS

} .

f .

l PERRY GRAND GULF ZIMt!ER  !!ATCll 1 BR6 BWR 6 BWR 5 BWR 4 I -

. 238-748 251-800 218-560 218-560 i EMERGENCY CORE COOLING SYSTEMS .

(Systems sized on design power) .,

f

(See Section 6.3) -

)

i

! 1. Low Pressure Core Spray Systems i

. Number of loops 1 1 1 2

-- Flow rate, gpm 6,110 at 7,115 at 4,725 at 4,625 at

, y 128 psid 128 psid 119 psid 120 psid C'.

2. liigh Pressure Core Spray System Number of loops 1 1 1 1(N Flow rate, gpm 1,550 at 1,650 at 443 at 4,250 1,147 psid 1,147 psid 1,160 psid 6,000 at 7,000 at, 4,725 at 200 psid 200 psid 200 psid ,,,

3. Automatic Depressurization System

~

Number of relief valves 8 8 6 7 ,

4. ' Low Pressure Coolant Injection ( }

3 3 3 2 Numberoflooks Number of pumps 3 3 3 4 i_ l .

~;- - jl.Zhg5L ? i

~: a-Q,*LXL[.5;i,;;", .,

,; .a ._ g g 3. ;p_ . == _ _ .

~ '

@ S' @ [. -

i

.,q. - TABLE 1.3-3 (Continued)

', PERRY GRAND. GULF ZIR!ER HATCl! 1 4 BWR 6 BWR 6 DWR 5 BWR 4

., j 238-748 251-800 218-560 .

218-560 -

~

j EMERGENCY CORE COOLING SYSTEMS (Cont.) .{

} 4. Low Pressure Coolant Injection (Cont.)'

.. ~ ;..

~

i Flow rate, gpm/ pump 7,100 at 7,450 at ~5 ,050 at *

{

?

20 psid 24 psid 20 psid 9,200at{

20 psid

5. Auxiliary Systems -

[

(See Sections 5.4 and 9.2) -

. I

6. Residual Heat Removal System i*

N. , 'l

- i  : I [

Reactor Shutdown Coolin. g Mode: -

.i.

\ l Number of loops 2 2 2 2..

-t l.

Number of pumps 2 4 2 4,s y;. I'

. l Flow rate, gpm/ pump (3) 7,100 7,450 5,050 7,700 {

6 Duty, Btu /hr/ heat exchanger (4) 46.9 x 10 50 x 10 6 30.8 x 10 6 30.8 x 10 6 (

j Number of heat exchangers 2 2 2 2 .

Primary containment cooling diode:' -

'I i

Flow rate, gpm .7,100(5) 7,450(5) 5,050(5) *11,500(5) f 4

9 I

4 e

1 +#.< + a( .g . 1 1 , j , 3 3,,,;; ; { y . _

, , ,, ,,,,, ,,g; a egygg_gE -e .

b . b @ j ',

4 i

TABLE 1.3-3 (Continued) i PERRY GRAND GULF ZIMMER IIATCl! 1 l BWR 6 BWR 6 BWR 5 ~ BWR ' 4 -

j -

238-748 251-800 218-560. 218-560 EMERGENCY CORE COOLING SYSTEMS (Cont.)'

I -

f 7. Emergency Service Water System Flow rate, gpm"/ heat exchanger 25,300 5,000 8,000

}

.{

', Number of pumps / unit 2/1 2/1 4 4 Flow rate, gpm/ pump 11,900/900 12,000/1,300

8. Reactor Core Isolation Cooling System

,s.. a.

t y Flow rate, gpm 700 at 800 at 400 at 400 at-

,O 150-1,177 psig 150-1,147 psid 1,120 psid 1,120 psid

9. Fuel Pool Cooling and Cleanup System 6 0 Capacity, Btu /hr 20 x 10 11.8 x 10 6.9 x 10 6 8.5 x 10 6

NOTES: *

1. liigh pressure coolant injection system used.

2. A mode of the RllR system. _

Capacity during reactor flooding mode with more than'one pump running.

3.

4. lleat exchanger duty at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown. '

5. Flow per heat exchanger. -

i h

a

acTTlM.cG wrrPim 1 W m Z u x.^u, d .i. .[ . . f .., ... , n ..-,1: d ; i lich f l] @ j2 e' C O

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4-i . ,

r TABLE 1.3-4 l COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS ( )

~

PERRY GRAND GULF ZIMMER BiILLY f i BWR 6 BWR 6 BWR 5 BWR 5 -

i 238-748 251-800 218-560

Type , , MARK III. Steel Mare III. Mark II. Over- Mark II. Over- -

[.

containment, Reinforced and-under primary and-under pri ary -

with pressure concrete, con- containment, en- containment, n-suppression, tainment, but closing drywell closed drywell and

? enclosed by. with pressure and suppression suppression pool.* ,,

reinforced suppression. pool. Enclosed Enclosed by reactor concrete reactor)* Containment ,by reactor building. p building. Con- encloses drywell building.

• 9 L' tainmentenclosesk*andsuppression 3 $.

, drywell and sup- ' pool. t

- pression pool. '

S p

Leak rate, %/ day 0.20 0.35 0.5 0.5 s . , f Containment (l p

t-Construction Steel shell. Reinforced con- Not applicable. Not applicabic f enclosed by crete cylindrical .

. f reinforced con- structure (not -

E crete cylindri- prestressed) with .

t

. cal structure ' hemispherical

• i .'

(not prestres' sed)- head; steel lined. *

, -[

with hemispherical *

. head. .

t L

Internal design 185 185

• Not applicable Not applicable temperature, *F . l I

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i l TABLE 1.3-4 (Continued)  !

i i

PERRY GRAND GULF ZI!MER BAILLY BWR 6 BWR 6 BWR 5 BWR 5 j 238-748 251-800 218-560

. Design pressure, psig +15, -0.8 15 Not applicable Not applicablq ,

2 3 6 6 Free (air) volume, it 1.2 x 10 1.67 x 10. Not applicable Not applicable Drywell -

4 e

Construction Reinforced con- Reinforced con- Prestressed con- Prestressedco\ n-crete. Basically crete. Basically. crete. Drywell crete. Drywell in cylindrical; flat cylindrical; flat is frustum of a frustrum of a conc; concrete roof concrete roof conc; steel steel lined, with a steel with a steel lined.

~ refueling head. refueling head. '

~~

w

,L Internal design 330 330 340- 340 temperature, 'F ,

i Design pressure, psig +30, ~21 30 +45, +45, -2 j r

Free (air} volume, 278,000 270,000 287,000 263,800 i total, ft .

[

Suppression Pool -

Construction Reinforced con- Reinforced con- Prestressed con *

• Prestressed con-crete, steel crete, steel crete. Pool is crete. Pool is lined. Basically lined. Basically cylindrical; cylindrical; steel cylindrical. cylindrical. steel lined. lined.

i Internal design, 185 185 340 340 i temperature, "F ,

Design pressure,8psig 15 15 +45, -2 +45, -2 Nb '

4- _

O' @- ..

TABLE 1.3-4 (Continued)

PERRY GRAND GULF ZItRIER' BAILLY BWR 6 BWR 6 BWR 5 IMR 5 ,

238-748 251-800 218-560 Water volume, ft 120,000 136,000 105,000 73,500 (

Break area / total vent area 0.010 0.008 0.008. 0.012 3 '

NOTES: ,

1. Refer to Chapter 3.

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