ML20207G022

Thermal-Hydraulic Analysis of DC Cook Spent Fuel Pool

ML20207G022
Person / Time
Site:	Cook
Issue date:	07/31/1988
From:	Odell L SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17325A898	List: ML20207F998 ML20207G007 ML20207G022
References
ANF-88-09, ANF-88-9, NUDOCS 8808230358
Download: ML20207G022 (22)

Text

-

A N F-8 8-0 9 ADVANCEDNUCLEARFUELS CORPORATION THERM AL-HYDR AULIC AN ALYSIS OF THE D.C. COOK SPENT FUEL POOL f

JULY 1988 R

O OO 15 P

PNU

ADVANCEDNUCLEARFUELS CORPORATION ANF-88-09 Issue Date: 7/14/88 l

l THERMAL-HYORAULIC ANALYSIS OF THE 0.C.COOKSPkNTFUELPOOL Prepared by y

e

'U

//

f 1

t L. D. O'DelT, Team Leader PWR Safety Analysis Licensing ~& Safety Engineering Fuel Engineering & Technical Services l

Contributors: D. L. M1ynarczyk D. A. Prelewicz July 1988 9f

=

I CUSTOMER DISCLAIMER ft$PORTANT NOTICS REGAROfNG CONTENTS ANO USE OF THIS DOCUMENT Pt.1ASE READ CAPEPULLY Advanced Nucear Fume Corporacon's warrances and recreeentanons cork coming tne suotect matter of the cocument are moes set form o tne Agreement benveen Advanced Nucteer Fuete Corporsoon and the Customer oursuant to wfucn me oocurnent ts ' souse. Accortungry, exceot as omerwee exoressly aro-viced in sucn Agreement. nenner Acvancac Nuc: oar Fws Corporanon nor any person acung on its berlet menos arty warranty or reoresentanon, exorossed or irnched, witn roepect f) the accuracy, compietenose, or usetuinees of tne infor.

macon containec in mte cocurnent. or mat me use of any informaten, socaratus.

memod or process ceciosed in mis cocument wul not intnnge onvarevy cwned ngnte; or assumes any lisonsbee with resoect to me use of any information, ao-paresus, rnemod or process cocosed in mis docurnent.

The informenon contened heresi is for me scpe use of Customer.

In croer to avoed imomrment of ngnte of Advanced Nuc:eer Fuees Corporsoon in patente or irtvormons weien may De ec:uced in me informanon contained in this occument, me racioesnt. Dy its accootance of mre document, agrees not to pucken or meme puceic use (in me patent use of me term) of such informanon unni 30 autnorued in wretng Dy A4vanced Nucear Fuste Corporsoon or untd after six (6) montne todowing terminanon or expiracon of me aforesaid Agreement and any extensen moreof, unsee omenwee exoroesty providee e me Agreement. No ngnte or liconess in or to arty potents are ernphed Dy ne furresning of this cocu-ment.

I ANF-3t45 472A ;1117)

ANF-88-09 Page i TABLE OF CONTENTS Section Pace

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 ANALYTICAL METHODOLOGY.....................

4 3.0 CALCULATIONS..........................

7 3.1 Linaa'r Heat Generation Rates..................

7 3.2 Calculations of Pressure Drop vs Flow 8

3.3 Solution for Total Flow 8

3.4 Pool Heatup Ra t e........................

10 3.5 Clad Surface Temperature....................

11 4.0

SUMMARY

OF RESULTS.......................

14

5.0 REFERENCES

16

'n

ANF-88-09 Page ii LIST OF TABLES

(

)

Table EA91 1.1 S uma ry o f Re s u l t s.......................

3 3.1 Parameters Used in Analysis 12 13 3.2 Assembly Parameters G

l l

9 V

4 J

mu m in p

u

ANF-88-09 Page 1

1.0 INTRODUCTION

AND

SUMMARY

The purpose of this report is to demonstrate that the storage of fuel with 50,000 mwd /MTU exposure in up to one-half of the D. C. Cook spent fuel pool storage locations will not violate thermal limits.

The current analysis for the D. C. Cock so:nt fuel pool high density storage racks (References 1 and 2) considered the storage of fuel with an average exposure of up to 40,000 mwd /MTU.

A new type of higher enriched fuel (referred to as Type 2) is capable of burnup to 50,000 mwd /MTV, and, therefore, a higher decay heat

~

generation rate.

y This report demonstrates that the storage of fuel with a slightly higher V

assembly heat generation rate, potentially interspersed with lower heat 4

generation rate and lower flow resistance fuel, is within acceptable limits.

The criteria used to determine this are: (1) no local boiling following the 7

full core offload, and (2) maintenance of adequate cooling following a loss of forced circulation cooling or partial flow blockage.

The thermal analysis consists of several portions, involving determination of the maximum heat rate per assembly, the maximum heat input to the pool, natural circulation cooling of the fuel in the pool under normal and accident conditions and the pool water heatup rate following loss of pool cooling.

A,.

g.,

3 Nlf _.,y /

A Peak fuel clad temperatures were calculated for three cases: a normal cooling case, an accident case with loss of pool cooling and an accident case with lf M D partial (90%) flow blockage at the limiting assembly outlet.

Results of these

~

analyses are presented in Table 1.1.

A summary of parameters used in the analysis is given in Table 3.1, and a summary of the fuel types considered is given in Table 3.2.

Local boiling does not occur in tha 1bniting assembly even ender 90% flow blockage conditions.

Adequate cooling is maintained even when forced circulation cooling is accidentally lost. When forced circulation cooling is lost, the minimum time to reach 212*F bulk boiling conditions is 8.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

_,. v. 3-

.y.,. g. ;1. n.,. i.,-.,,.c...,,...

w.. y..g.. = p _. - ; ;.jp,,c

. e.

e. ;,.7, ;,y.;. _.

c.,> 4 f

., n 7

., :: 1.,,

..~

e e, ; n_.

ANF-88-09

' '["?;..f-Page 2 4.7,jl f'

.';3 _ a

.n

}l Q s .

^

.6 The maximum linear heat rate calculated in this analysis increased by only 6f.

1 Y,.

4.65%, compared to the analysis in References 1 and 2.

Due to the very

'y,

.r conservative assumption in the previous analysis that all fuel in the core was y

exposed at 40,000 mwd /MTU, the maximum heat rate following a full core offload o,.,

,~. '...

calculated in the prior analysis is almost identical to that calculated in J.;'.

.r 3

Y h.

this analysis.

The conservative assumption made in this analysis is that one 4.d

]<:

half of the fuel is exposed to 50,000 mwd /MTV and the other one half of the L.; ; *.

fuel is exposed to 25,000 mwd /MTV prior to full core offload.

This was shown

..y.

f -. ' ;..

'4 to bound the case of 1/3 of the core at 50,000, 1/3 at 33,000 and 1/3 at

e.

-4 p:.

-(

17,000 mwd /MTV burnup.

Since there is no change in maximum heat rate, the d 6.*g

}

l$'

~$

heat removal capability of the cooling system has been shown to be adequate by the analysis presented in References 1 and 2.

n.

.$r 3

e'.'

I. ' ;;;:

^$

(

(

Gl

?

s I

s' fe'

• hy j

JE *

(g }A.;:

4. h,

'..4.

. ',* y g g. -

v; v

,.,. 5

?.., L '.

?.

401';?./

'k

• ' y ?p]. -

M.

c.

y g O

/

Of 1,~

A ( y '_, ' ',

h. -

t.

[,, * ' pl

' j,.*

g,

'{

, g.

1

~

' +.

ANF-88-09 l

Page 3 1

Table 1.1 Summary of Results CASE DESCRIPTION NORMAL LOSS OF FORCEO 90% FLOW COOLING CIRCULATION _

BLOCKAGE Bulk Temperature 150 *F 212 *F 150 'F Assembly Inlet Temperature 150 'F 212 'F 150 'F Assembly Maximum Fluid 181 'F 232 'F 198 'F Temperature Peak Fuel Clad Temperature 187 'F 236 'F 205 'F Exit Quality 0.0 0.0 0.0 Exit Void Fraction 0.0 0.0 0.0 9

4 89 9

r v

b E

I ANF-88-09 i

Page 4

!=

2.0 ANALYTICAL METHODOLOGY I

I The methodology described in Branch Technical Position ASB 9-2 '(Ref. 3) was used to calculate the decay heat for the fuel pool loaded with fuel assemblies discharged on a normal schedule followed by a full core offload 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> after shutdown.

{

In particular, the follcwing steps were taken to analyze the D. C. Cook spent fuel pool:

1 h

1.

Determine the kw/ft per assembly for the full core offload consisteat P

with Branch Technical Position ASB 9-2 (Ref. 3), "Residua', Decay Energy f

for LWR's for Long-Term Cooling".

2.

Determine the total power input to the pool.

F 3.

Analyze one row of 21 assemblies and determine the flows in each

[

assembly.

E a)

Use the code XCOBRA-IIIC (Ref. 4) to generate a AP vs flow rate curve.

1 5

b)

Iterately solve for the total flow by balancing buoyancy and flow

[-

friction forces.

Spent fuel pool loss coefficients are from Reference 2.

r-L 4.

Perform a hot channel analysis using XCOBRA-IIIC to determine the maximum temperature of the coolant which will occur in the 21st assembly farthest from the pool wall.

5.

Using the results from Step 4, determine the maximum clad temperature.

6.

Using the heat generation rate determined in step 2, calculate the time 5

- w_g

?

,.,,.g m.,.

u.g.....,, %. 7.. %

..,.,m

.. y 3......

t.

.?

E b -: X) ' f

.a 1y

^*

w. ; ;. ;

.4 J. _..:

5.O, y.

.?

Y ANF-88-09

.)

.h _:

Page 5

. ^

.. ~..

to reach bulk boiling following loss of forced circulation cooling for N;.-'*.w normal discharge and full core offload conditions.

,g I

Assumotions

[:. -r

~ -c.

The following major assumptions were made:

7 f.y 7

p.f.. e..

Y

-[h,~[

f

')g 1.

The 193 full core offload assemblies all come from 0.C. Cook Unit 2 and

'^;'

are Type 2.

The spent fuel pool contains 2050 fuel assembly locations

.'c ef e

'y and is used for both 0.C. Cook units.

3k.

4 3 -

s.

~ ~ 'i 2.

For purposes of calculating the pool heat load, 1/2 of the full core

~,.. i

[4 offload is assumed to have 50,000 mwd /MTV axposure, and the other 1/2 to

.,. t.. ' p v,J. v.

g have 25,000 mwd /MTV exposure.

This exposure was compared to the case of

.- ;..:;J y

I.k 1/3 of the core at 50,000 mwd /MTU, 1/3 at 33,000 mwd /MTU, and 1/3 at

.I.[ ~ $

,f.

17,000 mwd /MTV, and the former is conservative.

O ;. ' '.i.

-q<.

r z.

e-.

p,_, ;4 -

c' 3.

The remaining (2050 - 193) assemblies in the spent fuel pool arrived on a I' ; ' #

,'g schedule which assumes a 12 month cycle for Unit I and an 18 month cycle

,S E.; :.

.?

p for Unit 2.

This does not exactly fill the pool, so the oldest

.- - !f -f

,j assemblies are assumed to be 17 from Unit 1.

Di y'E

(,. (..;

, i.* p.3.*

J

..t 4.

The peak clad temperature analysis considers a limiting row of 21 M..ff ;

f assemblies. Two potentially limiting situations are analyzed: (1) all 21 Q,[fl.'

3 assemblies 50,000 mwd /MTV exposure Type 2 fuel, and (2) alternating Type

.Q.

1 and Type 2, with 11 Type 2 and 10 Type 1 assemblies. Thi^ second case

' O #f

+

- <p i

results in the higher power Type 2 assembly being farthest from the wall and hence the most flow limited.

5.

The Type 1 assemblies were assumed to all be 15X15, since this assembly has a lower pressure drop and hence lower flow resistance than the 17X17.

This will maximize the flow through the Type 1 assemblies and hence

~

minimize the flow thr< ugh Type 2 assemblies, including the limiting k,.

f l Nth.

L gll y. f.Wp?QYEf wl f.% lbyMp%g 7_fp/ QQif,Q4.:$%::,jQ,j-4 GQ.1 % ; Q. ' b. L '.

ANF-88-09 Page 6 assembly at position 21.

6.

A uniform axial heat flux profile was ' assumed.

This is conservative for determining margin to bciling since it results in a high local heat flux at the channel exit where the fluid temperature is greatest.

Unlike the situation in the core, PWR spent fuel axial peaking factors for decay heat are generally considerably less than 10% (Reference 5).

7.

For the determination of heatup rate, conduction to the pool floor and wall, and evaporative cooling at the pool surface were neglected.

(

8.

For the case of lost forced circulation pool

cooling, the bulk temperature of the pool water was assumed to be 212*F, the saturation temperature at atmospheric pressure.

While the pressure increases with pool depth, it is not possible to sustain a significant temperature

gradient, i.e., temperature increase with depth, due to buoyancy forces which cause any water heated above 212*F to rapidly rise to the surface and boil.

9; The fuel rack inlet tempe-ature for the normal cooling case was assumed to be 150*F to permit comparisons with the original analysis.

The bulk temperature following a full core offir;ad is only 130*F, so a 20'F conservatism is applied.

}. 'l5.g :/ f. y. Q;: ?:.L :,; ';p. Mr.d.; y h)y 4 L >.; &i 4 :. b..u:g., p g y',* 7';. ; : ;-? Q : (.,,.,,

~..

• • * ; g y.1, );

_.. ; a

.. ~

.{

j. ;

y ANF-88-09

e.. - t" Page 7

?'P g

<y 5

' i!

3.0 CALCULATIONS 4

J

~

C Table 3-1 gives a list of parameters used for this analysis. Table 3-2 gives 7)?

a list of fuel related parameters for ANF 17X17 (Type 2) and Westinghouse e-15X15 (Type 1) fuel assemblies.

These two fuel types will be the thermally 3

[.h limiting fuel.

This is because they are the most recently offloaded fuel and Of will, cont.equently, have the highest decay heat rates.

Further, the Type 1

/w 4.c fuel also has the icwest AP which, when taken in combination with the Type 2

". i f. J.

vg

f fuel, will result in the lowest flow rate in the Type 2 fuel.

3

.J A

U j.>..

,[.

Following is a description of the calculations:

bf f96.e

.:1 P'.-

3.1 Linear Heat Generation Rates

,. f p

ASB 9-2 was used to determine the decay heat fraction per discharge and for N-the full core offload of Unit 2.

Unit 2 will have the worst case offload, with type 2 fuel.

The linear heat generation rate was then determined for the

M.i N'g

,, e t

'.? A' Unit 2 (17x17) offload assemblies and for the most recent Unit 1 (15x15)

'g. '-.

l

'}

discharge assemblies as follows.

.g ';j,

!.[

p,:.

=%

a) 17 x 17 Fuel

.J

&{

DECAY HEAT FRACTION, P/P 0.003053

'i' o

d j,

Po, MWt PER CORE (193 assy) 341]

i DECAY HEAT, MWt PER CORE 10.4137 b !.s'l 5$

DECAY HEAT, KWt PER ASSEMBLY 53.957 Nd Q)4 w

g.

f-FUEL TYPE ANF 17 X 17 ACTIVE LENGTH, INCHES 144 LINEAR HEAT GENERATION RATE, KW /ft 4.50 t

b) 15 x 15 Fuel DECAY HEAT FRACTION, P/Po (64 assy) 0.001021 Po, MWt PER CORE (193 assy) 3250 DECAY HEAT, KWt PER RELOAD 3318 C., _.. _ _ _ _ _ _

f '- "

-w

,4 f

.,,)

+ =

,..,kTy i

• 5 y,

.f

..Md f

ANF-88-09 Page 8 DECAY HEAT, KWt PER ASSEMBLY 51.848 FUEL TYPE H 15 X 15 ACTIVE LENGTH, INCHES 144 LINEAR HEAT GENERATION RATE, KW /ft 4.32 t

3.2 Calculations of Pressure Oroo vs Flow The linear heat generation rates determined above were used as input to XCOBRA-IIIC.

An XCOBRA hot channel run was made to determine the pressure drop versus flow relationship for each of the two channel types, 15x15 and 17x17.

For each fuel type three cases were run.

The first case used temperatures of 150*F for normal conditions, assuming a full core offload. A 212*F case for accident conditions (both pumps unavailable) was also ren, as was a case with 907. blockage at the outlet and a 150*F inlet temperature.

Results from these XCOBRA runs were used to generate two AP vs. flow tables from each of the three cases run.

These tables were used in an iterative procedure to determine the total flow into the racks.

3.3 Solution for Total Flow The pressure drop versus flow tables were input into a FORTRAN program written l

to perform an iterative procedure to determine the total flow through the l

spent fuel racks and through each cell of the racks.

This procedure is identical to that used for the original analysis, but is implemented by a FORTRAN code.

Two potentially limiting loading schemes were considered for a row of 21 assemblies.

There are 41 assemblier across the pool width, so the row of 21 assemblies constitutes a limithg situation.

That is, the cooling water circulates down along tne pool wall and under the racks to feed the assemblies out to the farthest assembly from the wall.

Two arrangements were considered: a row with all 21 assemblies 17x17, and a row with alternating 17x17 and 15x15 assemblies with a 17x17 assembly in the 21st location farthest from the wall.

Since the 15x15 assemblies have a lower flow resistance, there

y j+ w vy...m

,..,3 ;.f.

3.-9 7; 7.

y 3

...y y.

,,....,..,,,u..,m..

,.,y, 1

e

.t

. g.,.

+

,=.

t,

~

a L.

y

} - :y

~

.W Q.l N

ANF-88-09 e.. dl..

Page 9

' "l t

• t-

...y. -,

o,

t f

a v

Y 1

is the possibility that the fl ow could be higher in a 15x15 assembly even

c.';

t. a..

though the linear heat generation rate is lower.

f

.;.. 3 An initial guess was made for the total flow into the first cell of the

rack, IL f

and the program calculates the delta-pressures and flow rates for eacn of the z 'j j

Y.b..i'.

-)

twenty-one cells.

If the fl ow into the 21st cell does not match the total

..l.;.;. D. I~

jf flow minus the flow inside the first 20 cells, the program tries a new total

'/.,'. f l. '#

.. n flow rate and recalculates all 21 pressures and flowrates again.

This

.m

(

continues until the final flowrates match.

': ';..q,

' _ * (

3

+.

The converged flownte through the 21st cell is then used in a final XCOBRA i?.'

i.

~ t.. ~

u case which gives the temperature at 12 axial locations along the assembly.

A

.t

.e

't separate calculation is performed to find the fuel cladding temperature, i.e.,

s

,, 4 4

to account for the temperature rise from the water to the clad surface.

. ] f.. *

.e.

.{

- n j..

+ g: -

Flow differences for the 15x15 and 17x17 assemblies were very slight.

The

.... ' ~

.1 e

g.

-]

?'.

a}

17x17 case was slightly more limiting at 150*F, whereas the alternating 15x15

~

~-

and 17x17 case was more limiting at 212*F.

However, the differences between

./'.

,.c h

the cases was very small.

This indicates that any mixture of 15x15 and 17x17

' [.}.{

7.

assemblies is acceptable from a thermal standpoint.

L*

I".

4 -l.

. j. j..J.,. '

s - c -

.i The mass flux used, and the resulting maximum fl u i d temperature, are given

[,1 below for each case analyzed:

. ~ ' !; '!"," '.

l;'.. l' k h'

.'(

y FINAL (COBRA RUNS i:.t 1 ? /

k.! Y,h;:

g[,&,s,

(

'~Y MASS FUI,1 FLOW uaX. ;;UID TEMP.

CAST (lbm/sec) 4.

0.01917 1.6752 180.51 ';

50'; nor a: :co' ng 0.02986 2.509 231.30 ';

212*C. no cool :Jong 0.01218 1.0643 197.96 ';

50';

9C', o'oc< age s

L.

N.

,.g I j',' [-i -[ { ',:.

N 5-k kll }4.(h:tf4l. $, Y D Ib.W "~~ N ? l l.'_Q 'f1 W.?%,L; f f hl &.l$ *liikN

~

i f

ANF-88-09 Page 10 3.4 Pool Heatuo Rate Should all pool cooling by forced circulation be lost, the pool will heat up slowly due to the large thermal inertia.

A very conservative estimate of the time to reach bulk boiling at 212'F can be made based on a low estimate of pool water inventory and neglecting conduction to the pool walls and evaporative heat transfer at the pool surface.

The most rapid heatup to boiling will occur following a full core offload when the heat generation rate and the bulk pool temperature are at maximum values.

To reach bulk boiling at the pool surface, the initial pool temperature must be increased to 212'F.

In.the limiting case of a full core offload, the initial bulk pool temperature is 130*F and the required temperature increase is 82*F.

The decay heat fraction for the pool, assuming the full core offload, is 0.003 based on the Branch Technical Position ASB 9-2 (Ref. 3).

7 Btu /hr which, based on a This produces a pool heat input of 3.493 x 10 minimum pool volume, results in a pool heatup rate of 9.54'F/hr.

The minimum time to raise the temperature by 82*F is therefore:

time = 82*F/(9.54*F/hr) 8.6 hrs.

=

Anain, this is a very conservative estimate and corresponds to the abnormal case of a full core offload and initial bulk temperature of 130'F.

Following a normal dischaise of fuel with all fuel cells filled, except the 193 spaces reserved for the full core offload, the initial bulk pool temperature is 120*F and' the decay heat fraction is 0.00163 (Ref. 3).

This 7

corresponds to a pool heat input of 1.9 x 10 Btu /hr and a heatup rate of 5.19'F/hr, again assuming a minimum pool volume.

For this case, the minimum time to raise the temperature to 212*F is:

17.7 hrs.

time - (212 - 120'F)/(5.19'F/hr)

=

ANF-88-09 Page 11 3.5, Clad Surface Temoerature l

The clad surface temperature will be greater than the maximum fluid temperature due to the temperature rise across the surface film.

For the normal cooling case, a laminar flow film coefficient is used. The temperature rise from the fluid to the clad surface is then 6.5'F.

In the abnormal case of boiling in the assembly, a boiling heat transfer coefficient must replace the convective coefficient.

This yields a AT of 4.0*F.

1

2

.c ANF-88-09 Page 12 Table 3.1 Parameters Used in Analysis PARAMETER MA.LE Number of Assemblies in Pool 2050 Number of Assemblies in Full Core (Units 1&2) 193 Maximum Pool Temperature 120*F Max. Pool Temper =t"ea (Full Core Offload) 130'F Max. Pool Temperature (Full Core Offload 165'F With One Pump Out)

Full Power (Unit 2) 3411 MWT Full Power (Unit 1) 3250 MWT Batch Average Burnup (Type 1) 40,000 mwd /HTV Batch Average Burnup (Type 2) 50,000 mwd /MTV MTU/ assembly (Units 1 & 2) 0.467 Cooling Time After Shutdown Before 156 hrs.

Placement in Racks (Full Core Offload)

No. of Type 1 or 2 Assemblies in Spent 1025/1025 Fuel Pool Reload Schedule, Unit 1 12 months Reload Schedule, Unit 2 18 months Assemblies Per Discharge, Unit 1 64 Assemblies Per Discharge, Unit 2 88

-m

.:( ~.:. m.

n.,.

.....c ;n., y

..cc..

-:..x -<

o~:..

../...->

.. w. :

s.

., -+, _

k * :.]

~*

... -4

a. ; -

] < * :... {

'E g

M' i

1. 1 ANF-88-09

(. ~

Page 13

{ l':

, ",l.

a-... -p z.:

Table 3.2 Assembly Parameters PARAMETER W 15 X 15 ANF 17 X 17

~ 'i

~

.a.

(TYPE 1)

(TYPE 2) e

.e i

Clad 0.0. (In.)

0.422 in.

0.360 in.

'7 F*

Fuel Rods Pe-Assembly 204 264

. i $}

.';? -.QCr i.;

Top Of Heated Length To Bottom Of U.T.P.

8.40*

8.40 J.'-

Top Of Heated Length To Top Of U.T.P.

11.98*

11.98

,."q ] '-

l.

[:

Rod Pitch 0.563 in.

0.496 in.

-~.'.fl.

~

N Flow Area Per Bundle 43.32 in.2 45.30 in.2

(/

Fuel Length (Heated) 144 in.

144 in.

'~

Bottom Of Inlet Orifice To Heated 3.98 in.*

3.98 in.

Length

1. c

??

Top Of Inlet Orifice To Bottom Of 3.73 in.*

3.73 in.

' l.v.. :

Heated Length O. '..

Instrument Tube Per Bundle 1

1 Instrument Tube 0.0.

0.544 in.

0.482 in.

~ -n

}..

Wetted Perimeter Per Assembly 341.4 in.

371.3 in.

c,.

Guide Tube 0.D.

0.544 in.

0.482 in.

g.. [. :

~

Guide Tubes Per Assembly 21 24

.--g..-

Fuel Rod Heated Perimeter Per Assembly 270.5 in.

298.6 in.

3.-:.,

l

i

u..

.s j..,

2 %. i ;;

s.,

P, en-yQ y[

a -

.;. f J.

The values for the 15x15 fuel are assumed to be equal to the 17x17 fuel; not a significant parameter.

.4 7

s mm _

,' ~ %.,; :.t.n.;f. j'] ; -L,.; 7 F_;;\\.g._s,lJ. e ;N,%..,. ! 1. 9, b s c't ' G_.'y-f ; 8..L &, J

• 6.6[,, c * ' L l. *[.

. L :.. -

+

y

..,.3.-

. ~.

r

- i 7-

.~ *f

g s.

s ANF-88-09 EW..

Page 14 f.

_ ', 3..

l t..h( 7. /

l 4.0

SUMMARY

OF RESULT 1

-J f

The peak assembly linear heat generation rate is increased by 4.65". over that used in the or {inal spent fuel pool analysis due to extending the burnup to W

n -

50,000 mwd /MTU compared to 40,000 mwd /MTV.

The previous analysis assembly

[..~

linear heat generation rate was 4.3 kw/ft whereas the new value is 4.5 kw/ f t.

e Que to a large conservatism in the prior analysis, in particular the

.( :

W assumption that all of the fuel in a full core offload is exposed to 40,000 i -

( -f

(

mwd /MTU (a physically impossible situation), the calculated total pool neat

~

load is essentially unchanged.

A conserf ative assumption of 50,000 mwd /MTU for one-half of the fuel core offload and 25,000 mwd /MTU for the remaining

} : <f. '.,,

4 half has been.usumed for the present analysis.

Since there is no ca'iculated k \\;

increase in thal poo~

heat load, the cooling system performance, i.e.,

' ef I. ~

n.

ability of the forced ulation cooling system to remove heat from the pool, is covered by prio) ilysi s,Ref. 1 and 2).

r Both a normal cooling case and a loss of forced circulation cooling case were

' 1.' f..

?

.. J analyzed following a full core of fload 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> af ter shutdown.

For the

~ -

g normal cooling case, the maximum fuel clad temperature is 187'F, a value well

- U below the onset of local boiling temperature.

A partial blockage (90*,) at the

'h1 top of the limiting fuel ce'l was also analyzed.

In this case, the maximum fuel clad temperature was 205'F, still well below t.ie saturation and onset of

c., J ',g. *

, Egl..

local boiling temperatures.

. lc-f ',:.f n.-

R' Should forced circulation cooling of the spent fuel pool be lost following a s'. [

full core offload C th the pool totally filled (2050 assemblies!. it C'l

. I

1. M take more that 8.5 nours for tne pool sater to reach 3 oulk Do, 'ng 2'2':

condition.

Even in this

ase.

.he ax um Se'

'aa e ceratare

~as calculated to be 236*F. 3

.e pe r at are ae De' w *.nat at an'cn 5e'

'ntege'*.

is challenged.

I a-f.-0 W. D.f94 4;'..;_14% 7 %g :rf ; \\. [ ';f ;7r9[Wy a ; ' 7 ; yy.g[.n*f.._Q',sl

c.. o..,

v z.s. y -

o, a,.; g %.

og :,,.g,

. qm_.. - e... g...,.....,;.. ;__.,,,,,, o. g. ;, y,

,.,1 gy,,.

p s.i

. -..: w

w. * '

~

~

, t.

a)

.:. \\ '..

..,-\\..'.

.t'.; ;'t ANF-88 0g

'. d.I;'~ :

Page 15 4,,

e

").,;

g' i.. ;l' i.

The design of the D. C. Cook fuel racks is such that no interstitial gaps J

.s-

?

exist in the spent fuel pool. Therefore, no analysis of intercell voiding was

' V.

.. ~

performed.

',b ^,.?'

d

7...-.,...

In sumary, boiling will not occur in the fuel pool in the limiting case of a

• ~ *,

full core offload with the pool filled following the offload.

Even if forced I ' i. M).'

..:- g circulation cooling is lost, the fuel clad temperatures will not approach

. M Qy,

-3

'i ' (',

I values which threaten clad integrity.

rt _ b...

< 4.ty -

a u

-, 3.g-t J

r,-

.e 4

y

. +'

?

'aQ*,

.]. - I. * {-

g-

... i C.

~4-

%1

, j 9,

-1 e

,5

,c

.N

1. .-- L/.

,p

}.

A

.g :

..,p.

1

'.fi :

^

f,

- ',=

3..-

y[/;

'.4'

-k.

=

~,

..'+e g q..-;*.

\\

4 g'{

.y. ;]ig? E

,f

{

l

. ~

y l

l l

SP *'

k..

.Aw> -.y

.t.,

y 3, ;..;.7,,,.,. _r

. ;. n. $. y,m..,,.w 4 : _,...

..,,y..v.

y,,7.u

-y.

,;. ~., a u,,,..,,.

,... e.

v. a r

,v...

o.~...

..,. ~

u.

. 3 p

3 a_

.?

~'}

I

'J '.,,,.,

,j

,;} -

l

.f' ANF-88-09 h.,.--

. h 'h '.a ".

Page 16

',f

?,i. s m

g,c

5.0 REFERENCES

f 9

1.

XN-RF-3, Rev.

O, "Design Analysis Report: Replacement of Spent Fuel

~

Storage Modules for the Donald C. Cook Nuclear Plant Units No. I and 2,"

Exxon Nuclear Company, Inc., Richland, WA, April 1980.

2.

XN-RF-3, Rev.

O, "Design Analysis Report: Replacement of Spent Fuel

.? c Storage Modules for the Donald C. Cook Nuclear Plant Units No.1 and 2, Appendix E: Thermal Hydraulics Calculations." Exxon Nuclear Company,

..., '4

~..

Q.

Inc., Richland, WA, April 1980.

? s.l i. - i, ^

'}.G.~

4.

h 3.

U. S. Nuclear Regulatory Commission, Branch Technical Position ASB 9-2, O

"Residual Decay Energy for Light Water Reactors for long Term Cooling,"

4 Rev. 2, July 1981.

.~

4.

XN-NF-75-21(P)(A), Rev. 2, "XCOBRA-IIIC: A Computer Code to Determine the

(/.' ' '

C.-

Distribution of Coolant During Steady State and Transient Core n.

Operation," Exxon Nuclear Company, Inc., Richland, WA, January 1986.

c.

n.,.

.,f 5.

R. B. Davis, "Data Report for the Nondestructive Examination of Turkey

. '. : U!

Point Spent Fuel Assemblies 802, B03, B17, B41 and B43," HEDL-TME 79-68 UC-70.

.. ?..: '

,i.;

.jf..-

4

_.;.f, 3_. yn C

1, ' *

?

. S p.*

g. 5

. *f V '. *.

5*

w; f.

.v=1

\\

_~. Q ib';~

Y:

t :.,i.% 'g_q..f;,g i g g;. p f..; g ;3g 'g.c.s;g: = Q n ; 4 4 QQ. j: - N..

s ANF-88-09 Issue Date: 7/14/88 4

THERMAL-HYRAULIC ANALYSIS OF THE D. C. COOK SPENT FUEL POOL Distribution TH Chen LA Nielsen MD DeGraw i

NF Fausz RC Gottula JC Hibbard JS Holm JW Hulsman JD Kahn TH Keheley TR Lindquist LD O' Dell BD Stitt BE Schmitt SR Wagoner RT Welzbacker HE Williamson BD Webb SC Yung AEP/HG Shaw (10)

USNRC/LA Nielsen (20)

Document Control (4) '

e ii.,,.