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# ^ WTED CORRESN@@

ISHAM, LINCOLN & BEALECOUNSELORS AT LAW

ONE FIRST N ATION AL PLAZA FORTY-SECOND FLOORCHICAGO, ILLINOIS 60603

TELEPHONE 312-558 7500 TELEX:2 S288

WASHINGTON OFFICE

August 19, 1981 "' ' "" ''''7E$ a's* '""'' " *suiWASHINGTON. D. C 20036

202-833 9730

UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSION

ATOMIC SAFETY AND LICENSING BOARD

In the matter of )) Docket Nos. 50-254-SP

COMMONWEALTH EDISON COMPANY ) 50-265-SP(Quad Cities Station, ) (Spent Fuel Pool Modification)Units 2 and 3) }

Dear Administrative Judges:

Please find ?nclosed Supplement 2 to Revision 1of the Report prepared by Joseph Oat Corporation forCommonwealth Edison entitled " Licensing Report on High DensitySpent Fuel Racks for Quad Cities Units 1 and 2." ThisSupplement provides revised sections 5 and 8, which shouldbe inserted into Revision 1.

Sincerely,

.

|Robert G. Fitzg.tbbons,'Jr.

*

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RGF:jp c), tEnc. DOCK 5TEDcc: Service List umm0 4ff,

2 AUG 2 41981 > 1

h 0ff!ceof theSecretagfjh k txketing & SeMees- ,

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Commonwealth EdisonOne First Nat'enal P' ara cNcaco. HhnoisAccress Reply to: Post Cffrce Sox 767Chicago. Illinois 6069o_

August 10, 1981

Mr. Harold R. Denton, DirectorOffice of Nuclear Reactor RegulationU.S. Nuclear Regulatory CommissionWashington, DC 20555

Subject: Quaa Cities Station Un its 1 and 2Transmittal of Supplement 2 toRevision 1 of the Licensing Reporton High Density RacksNRC Docket Nos. 50-254/265

Reference (a): T. J. Rausch letter to H. R. Dentondated June 24 19' 1 (transmittal ofRev. 1 Licensi.3 Report).

Dear Mr. Denton:

Enclosed is Supplement 2 to Revision 1 of the reportprepared by Joseph Oat Corporation for Commonwealth Edison entitled" Licensing Report on High Density Spent Fuel Racks for Quad CitiesUnits 1 and 2". This supplement provides revised sections 5 and 8,,

which should be inserted into the book provided in Reference (a).

Section 5, Thermal Hydraulic Considerations, now includesthe fuel heat generation calculations (part 5.1) ano the results ofthe pool thermal hydraulic analyses (part 5.2.2).

Section 8, Radiolooical Consecuences, now containsadditional information anc operating experiences concerningracionuclide concentrations in spent fuel pool water and fuel poolraciation levels.

Please address any questions you may have concerning thismatter to this office.

One signed original anc thirty-nine (39) copies of thistransmittal are provided for your use.

Very truly yours,

f W'

#Thomas kauschNuclear Licensing Aaministrator

Boiling Water Reactors

Enclosurecc: Region III Inspector - Quad Cities-

2406N \

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-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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5. THERMAL-HYDRAULIC CONSIDERATIONS-

A central objective in the design of the high-density fuel rackis to ensure adequate cooling of the fuel assembly cladding. In thefollowing, a brief synopsis of the design basis, the method ofanalysis, and computed results is given.

5.1 Decay Heat Calculations for the Spent Fuel

This report section covers requirement III.1.5(2) of the NRC "OTPosition for Review and Acceptance of Spent Fuel Storage and Handling

' Applications" issued on April 14, 1978. This requirement states that

calculations for the amount of thermal energy removed by the spentfuel cooling system shall be made in accordance with Branch TechnicalPosition APCSB 9-2 " Residual Decay Energy for Light Water Reactors forLong .erm Cooling"2 The calculations contained herein have been madein accordance with this requirements.

'' 5.1.1 Basis: 2

The Quad Cities 1 and 2 reactors ar- rated at 2511 Megawatt-Thermal (MWT) each. The core contains 724 fuel assemblies. Thus, theaverage operating power per fuel assembly, P is 3.468 MW. The fuelo,

assemblies are removed from the reactor after a nominal burn-up of25000 Megawatt-days per short ton of uranium (MWD /STU). The fuel

,

discharge can be made in one of the following two modes:

(i) Normal discharge - Mode (i)

(ii) Full Core discharge - Mode (ii).

As shown in Table 1.1 of Section 1, the average fuel assemblyremoval batch size for Mode (i) is 200 fuel assemblies. However, thecomputations are performed for a Latch as large as 240 fuel *

assemblies. The fuel transfer begins af ter 100 hours of cool-off time{

in the reactor (time after shut down). It is assumed that the timeperiod of discharge of this batch is 2 days,

|_

0' VpoM30lSI m

_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _

e

O I

.' .

Mode (ii) corresponds to a full core discharge (724 arsemblies).'

It is assumed that the total time period for the discharge of one fullcore is 6 days af ter 100 hours of shut down time in the reactor). Thedischarge rate to the pool is assumed to be continuous and uniform.

The heat dissipation from each pool is accomplished by two ,independent fuel pool cooler loops, each equipped with a pump rated at '700 gpm. In addition, the Residual Heat Removal (RHR) heat exchangersmay be used in conjunction with the fuel pool coolers to boost the heatremoval rate. For each unit, there are two RHR heat exchangers *

supplied by four pumps, three of which can deliver 14500 gpm at 360' Ihead. Despite the large potential capacity, it has been assumed thatonly 1000 gpm of this flow rate is available for the fuel pool, throughone six inch pipe line.

In the following, all relevant performance data for the spentfuel pool and RHR heat exchangers is given. '

,

i

Spent Fuel Pool Heat Exchanger: ,2a.i

TEMA type 21-197 BEU, 1020 sq.ft. surface on 196 U-tubes;5/8" diameter x 18 BWG arranged on 0.875" triangular pitch.Postulated fouling for both tube and shellside surfaces is0.0005 sq.ft- F-Hr/ BTU (each surface). Shellside (cooling) {and tubeside (pool water) flow rates are 800,000 and 350,000

| lbs/hr. respectively. The corresponding value of the re-;duced thermal flux (NTU) for fully fouled condition is

0.933; and the temperature efficiency is 0.55. '

.;

b. RHR Heat Exchanger:I

TEMA type 63-288 CET . 11000 sq.ft. overall surface on 2415I :

tubes, 3/4" diameter x 18 BWG arranged on 0.9375" triangular |>

pitch. Postulated fouling for the shellside of the tubes

; surface is 0.0005 sq.ft. OF-Hr./ BTU and that for the itubeside (river water) is .002 sq.ft. OF-Hr./LTU. |

,.

|

5-2

.

. i.

(

*

The shellside and tubeside design flow rates are 5.35 x 106''

6lbs/hr. and 3. 5x10 lb/hr. , respectively. The corresponding,

value of NTU is 0.745; and the temperature efficiency is0.385.

The above data enables complete characterization of the thermal per-formance of the heat exchangers.

Reference ( is utilized to compute the heat dissipation require-ments in the pool. The total decay power consists of " fission

products decay" and " heavy element decay." Total decay power P for afuel assembly is given as a linear function-of P and an exponentionalgfunction of t and t 'o s

ie: P=P f (tg,ts)g

where

P= linear function of P"'

o

P= average operating power per fuel assemblyo

ta= cumulative exposure time of the fuel assembly in the 2

reactor

ts= Time elapsed since reactor shutdown

The uncertainty factor K, which occurs in the functional

relationship f (to,ts) is set equal to 0.1 for ts >10 sec in the

interest of conservatism. Furthermore, the operating power P isg

taken equal to the rated power, even though the rea'ctor may be opera-ting at a fraction of its total power during most of the period of

| exposure of the batch of fuel assemblier. Finally, the computations

and results reported here are based on the discharge in the year 2005

I

A

6

5-3

-._

-,, . .

.

5

(ref. Table 1.1). This is when the inventory of fuel in the pool will^

be at its maximum resulting in an upper bound on the computed decayheat rate.

In the past, Quad Cities reactors have operated on what is

commonly referred to as "18 month cycle." Quite often, system plan-

ning requires extended reactor coastdown operation (sometimes to 40%of rated power) af ter the end of full power reactivity (19000 MWD /STU)has been reached. The batch average discharge burn-up of current fuelbatches is approximately 25000 MWD /STU. In the future, due to present

lack of spent fuel reprocessing in the U.S., it is conceivable that

the average discharge exposure can approach 30,000 MWD /STU due to,

higher initial enrichments and longer coastdowns. A longer coastdown

period implies a greater value of t in the foregoing equation, itoalso implies a smaller value of P It can be shown that an exposure.

period, t equal to 4.5 years (3-18 month refueling cycles) alongo,

with the rated reactor power produces an upper bound on the value of P.This is due to the fact that f (to,ts) is a weak monotonicallyincreasing function of t Hence, the reactor operating time is- g.

8assumed to be 4.5 years (t =1.42x10 secs). 2g

Having determined the heat dissipation rate, the next task is toevaluate the time temperature history of the pool water. Table 5.1.1identifies the loading cases examined. The pool bulk temperature time

,

history is determined using the first law of thermodynamics (conserva- |tion of heat). The system to be analyzed is shown in Figure 5.1.1. I

i

A number of simplifying cssumptions are made to render the |analysis conservative. The principal ones are:

,

1. The cooling water temperature in the fuel pool coolerand the RIfR heat exchangers are based on the maximum

postulated values given in the FSAR.

.~

5-4

.

y,-- yy - , . - - - ,-:

_

o i

.

p

2. The heat exchangers are assumed to have maximum foul !''

the|ing. Thus, the temperature ef fectiveness, S, for'

heat exchangers utilized in the analysis are the lowest fpostulated values: S= 0.52 for fuel pool coolers, 0.385 ;for RHR heat exchangers. S is cr.lculated f rom FSAR andheat exct. anger technical data t..eets. '

:,

3. No heat loss is assumed to take place through the con-crete floor.

4. No credit is taken for the improvement in the filmif

coefficients of the heat exchangers as the operatingjtemperature rises. Thus, thefilmcoefficientusedin|

fthe computations are lower bounds.

S. No credit is taken for evaporation of thc pool water.

The basic energy conservation relationship for the pool heat exchangersystem yields: 2--

C =0 -0 -0t 1 2 3 (5.1.2)

where

C: Thermal capacity of stored water in the pool.

t: Temperature of pool water at time,r

( O: Heat generation rate due to stored fuel assemblies inythe pool. 0 is a known function of time, r f rom the

1

j preceding section.

|i

0: Heat removed in the .wo fuel pool coolers.2I ~

i| 0: Heat removed in the RHR heat exchanger ( 0= if RHR3 3

i is not used),

i -s ,

5-51

:_ . - _. .-- . -_

. . . . . . . . . . . . . . . . . .

- - - - - - - - - . - . .- -

I } } =,

.

TABLE 5.1.1 -

LIST OP CASES ANALYZED

Case No. Condition No. of No. of No. of Total Time Cool off timefuel spent fuel Ri!R' s to transfer before transfer

assemblies pool llXS in-service fuel into begins, hrs.N the pool

th, hrs.

1 Normal discharge 240 2 0 48 100with oversizebatch f

T 2 Same as One 240 2 1 48 100

j 3 Normal discharge 200 2 0 48 1001

4 Normal discharge 200 2 1 48 100

5 Full Core 724 2 1 144 100| _

Discharge

>

i. +

1. .

l.

*

iThe pools for Unit 1 and 2 in the Quad Cities installations have '

^ ,

total water inventory of 44887 and 44471 cubic feet respectively whenall racks are in place in the pools and every storage location isoccupied.

5.1.2 Decay Heat Calculation Results:

The calculations were performed for the Quad Cities Unit 2 pooldisregarding the additional thermal capacity and cooling systemavailable in the other pool. The use of the lower water inventory of jthe Unit 2 pool thus is the bounding case.

For a specified coolant inlet temperature and flow rate, the ,

lquantities Q and Q re shown to be linear function of t in a recent I2 3paper by Singh(3) As stated earlier, 0 , is an exponential function.

1of r Thus Equation (5.1,2) can be integrated to determine t.

directly as a function of r The results are plotted inFigures |

.

(5.1.2) - Figures (5.1.11) and show that the pool water never

approaches the boiling point under the most adverse conditions. These--

i

figures also give Q as a function of r Two plots are generated for7 .

each case. The first plot for each shows temperature and powergeneration for a period extending from where r isr=0a r=2r n n

the total time of fuel transfer. The second plot shows the same 2

quantities over a long period. The long-term plots are produced toindicate the required operating time for the heat exchangers,

i Summarized results are given in Table 5.1.2.

Finally, computations are made to determine the time interval toboiling after all heat dissipation paths are lost. Computations are

made for each case under the following two assumptions:

(i) All cooling sources lost at the instant pool bulktemperature reaches the maximum value

-

5-7

._. ._. . . --. .-.

- -

. .,

.

e

(ii) All cooling paths lost at the instant the heat dissipa-^

tion power reaches its maximum value in the pool.

Results are summarized in Table 5.1.3. Table 5.1.3 gives the,

bulk boiling vaporization rate for both cases at the instant theboiling commences. This rate will decrease with time due to reducedheat emission f rom the fuel.

5.2 Thermal-Hydraulics Analyses for Spent Fuel Cooling

This report section covers requirement III.1.5(3) of the NRC "OTPosition for Review and Acceptance of Spent Fuel Storage and HandlingApplications" issued on April 14, 1978. Conservative methods havebeen used to calculate the maximum fuel cladding temperature asrequired therein. Also, it has been determined that nucleate boilingor voiding of coolant on the surface of the fuel rods does not occur.

'

5.2.1 Basis:-

In order to determine an upper bound on the maximum fuel claddingtemperature, a series of conservative assumptions are made. The most

!

important assumptions are listed below:

a. As stated above, the fuel pool will contain spent fuel withvarying " time-af ter-shutdown" ( ts). Since the heat emissionfalls off rapidly with increasing t it is obviously con-s,

servcitive to assume that all fuel assemblies are fresh ( ts =100 hours), and they all have had 4.5 years of operatingtime in the reactor. The heat emission rate of each fuelassembly is assumed to be equal.2

b. As shown in Figures 2.1 and 2.2 in Section 2, the modulesoccupy"an irregular floor space in the pool ~. For purposes

of the hydrothermal analysis, a circle circumscribing the

-

5-8

, _ _ _ _ _ _ _ _ - _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ - . .

. . _ ._

,_

) ) ) -.

.

TABLE 5.1.2 -

MAXIMUM POOL BULK TEMPERATURE t, COINCIDENT TOTAL POWER O andy

COINCIDENT SPECIFIC POWER FOR THE HOTTEST ASSEMBLY

-6Case No. No. of n Time Maximum Coincident Coincident O x10 BTU / hourtAssemblies to transfer pool bulk time (since specific

fuel into temp.0F initiation power q,pool, hrs. of fuel BTU /sec.

transfer, hrs.

1 240 48 134.6 64 10.24 10.85*

2 240 48 121.2 58 10.39 10.993 200 48 130.6 64 10.24 9.37

4 200 48 118.5 58 10.39 9.48

5 724 144 145.8 150 8.72 24.89

:

,

>

. .-. . . _ . . . _. . _ _ . ___ . _.

) ) ) -.

. '.

TABLE 5.1. 3

TIME (Hrs) TO BOILING AND BOILING VAPORIZATION RATEFROM TIIE INSTANT ALL COOLING IS LOST

Case No. CONDITION 1 CONDITION 2Loss of Cooling at maximum Loss of Cooling at maximum

pool bulk temperature power discharge rateTime (IIrs) bap. Rate Time (fir s ) Vap. Rate

Ib./hr. Ib./hr.

I 20.3 10759 20.4 11038

2 23.7 10723 24.5 10955

3 24.8 9203 24.7 9461

E 4 28.2 9224 29.1 9389

5 7.7 25147 7.64 25384

>

-

. o

.

+

actual rack floor space is drawn. It is further assumed-

that the cylinder with this circle as its base is packed

with fuel assemblies at the nominal pitch ot 6.22 inches

(see Figure 5.2.1).

c. The downcemer space around the rack module group varies, asshown in Figure 5.2.1. The nominal downcomer gap (9 inches)available in the pool is assumed to be the total gap avail-

able around the idealized cylindrical rack; thus, the

maximum reaistance tc downward flow is incorporated into theanalysis.

e

d. No downcomer flow is assumed to exist between the rackmodules.

In this manner, a conservative idealized model for the rack

assemblage is devised. The water flow is axisymmetric about the

vertical axis of the circular rack assemblage, and thus, the flow is

two-dimensional (axisymmetric three-dimensional). The governing''

equation to characterize the flow field in the pool can now bewritten. The resulting integral equation can be solved for the lower

plenum velocity field (in the radial direction) and axial velocity

(in-cell velocity field), by using the method of collocation. It

should be added here that the hydrodynamic loss coefficients which

enter into the formulation of the integral equation are also taken4from well-recognized sources and wherever discrepancies in reported

values exist, the conservative values are consistently used..

i

! After the axial velocity field is evaluated, it is a straight-

forward matter to compute the fuel assembly cladding temperature. The

knowledge of the overall flow field enables pinpointing the storagelocation with the minimum axial flow (i.e. , maximum water outlet temp-erature). This is called the most " choked" location. It is recog-

nized that some storage locations, where rack module supports are

located, have some additional hydraulic resistance not encountered in

n

5-11 *

|t

I

.

. .

t

,

other cells. In order to find an upper bound on the temperature in''

such a cell, it is assumed that it is located at the most " choked"

location. Knowing the global plenum velocity field, the revised axial

flow through this choked cell can be calculated by solving the

Bernoulli's equation for the flow circuit through this cell. Thus, an

absolute upper bound on the water exit temperature and maximum fuel

cladding temperature is obtained. It is believed that in view of the

preceding assumption, the temperatures calculated in this manner cver-

estimate the temperature rise that will actually be obtained in the

pool.

The maximum pool bulk temperature t is computed in Section 5.1.3 and

reported in Table 5.1.2. The corresponding average power output from

the hottest fuel assembly, q is also reported in that table. The

maximum radial peaking factor, ranges from 1.6 to 1.8 for Quad Cities

installations. Thus, it is conservative to assume tha t the maximum

specific power of a fuel assembly is given by

9 = q a,-A

where a = 1.8-

r

The maximum temperature rise of pool water in the most disadvan-

tageously placed fuel assembly is given in Table 5.2.1 for all loading 2

cases. Having determined the maximum " local" water temperature in the

pool, it is now possible to determine the maximum fuel cladding

temperature. It is conservatively assumed that the total peaking

factor a is 3.1. Thus, a fuel rod can produce 3.1 times the averageT

heat emission rate over a small length. The axial heat dissipation in

a rod is known to reach a maximum in the central region; and taper off

at its two extremities. For the sake of added conservatism it is

assumed that the peak heat emission occurs at the top where

-

5-12

_ - _ _ . .. .-_- . . _ .

.- -. . . . - _ - _ _ . . _ - .

) ) ) *,

.

.

'.

TABLE 5. 2.1.

MAXIMUM LOCAL POOL WATER TEMPERATURE ANL- LOCAL FUEL

CLADDING TEMPERATURE

Case No. Max. Local Pool Maximum Coincident Local CaseWater Temperature OF Cladding Temperature OF Identified

1 157.8 183.6 240 AssembliesCoolingMode A

7

2 144.7 170.8 240 AssembliesT Cooling3 Mode B

3 153.8 179.6 200 Assemblies -

CoolingMode A

4 142 168.t 200 AssembliesCoolingMode B

5 166.6 189.0 724 AssembliesCoolingMode B

* Cooling Mode A mean only two f uel pool IIxs working.

Cooling Mode B means two fuel poa. and Portion of 1 RiiR working,w

.. .. . -. __. __ .__ _

j- o *

e

TABLE 5.2.2,

-

POOL AND MAXIMUM CLADDING TEMPERATURE AT THE,

INSTANCE FUEL ASSEMBLY TRANSFER BEGINS.,

4

4.

Case No. Cladding Coincident Pool,

O OTemp. F Temp, FBulk Local

1 178.3 110.5 133.7,

2 173.3 105.2 128.7i

3 178.3 110.5 133.7

4 173.3 105.2 128.74

5 170.6 105.2 126.0

i

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e

|

i

!

,

I

t

i

;-

;

I

,

5-14|

. . . . . _

_ __

. .

.

.

the local water temperature also reaches its maximum. Furthermore,no|^

credit is taken for axial conduction of heat along the rod.Thehighly|Iconservative model thus onstructed leads to simple algebraic

equations which directly give the maximum local cladding temperature,D'

) c

5.2.2 Results

Table 5.2.1 gives the maximum local cladding temperature,tc, at

the instant the pool bulk temperature has attained its maximum value.It is quite possible, however, that the peak cladding temperature 2

occurs at the instant of maximum value of g , i.e. , at the instant whengthe fuel assembly is first placed in c. storage location. .able 5.2.2'

gives the maximum local cladding temperature at =0. It is to benoted that there are wide margins to local boiling in all cases. Thelocal boiling temperature near the top of the fuel cladding is 2400F.F'Jrthermore, the cladding temperature must be somewhat higher than the

'

boiling temperature to initiate and sustain nucleate boiling. The; above considerations indicate that a comfortable margin against the+

initiation of localized boiling exists in all :ases.

.

A

5-15

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

. _ _ _ __ _....__.____.___.m. . _ _ _ _ _ _ = _ _ _ . _ . _ . . . _ _ _ _ _ . _ _ _ _ . _ _ . _ .

. .

.

.

REFERENCES *O SECTION 5n

1. FSAR, Quad Cities, Section 10, Auxiliary and Emergency Systems.; 2. U.S. Nuclear Regulatory Commission, Standard Review Plan, Branch

Technical Position, APCSB 9-2, Rev. 1, November 1975.

3. Journal of Heat Transfer, Transactions of the ASME (c. 81), "SomeFundamental Relationships for Tubular Heat Exchanger Thermal;; Performance," K.P. Singh.

24. General Electric Corporation, R&D Data Books, "Ecat Transfer and

i Fluid Flow," 1974 and updates.!!

}

i

,

. M

1

.

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1

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5-16

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

- - - - - - - - - - - - -

) ) ) -

.

.

.

POOLREMOVABLE SPOOLPIECES (2)

COOLING WATERRHR HEAT AT 95* F(MAX)

EXCHANGER(I)_ _ _ _

/ -t i f 4_- / (C i

'

t ! ! !--

, , o- y 4 +-

l III I I I| TO HEAT

| ! ! ! ! !;l

,

SiNx] n =r -.

FUEL ( '

POOLSECONDARY COOLERCOOLANT (2) I fATlO5*F

- 1 I

| -*

TO gR E ACTOR

CAVITYN.Cr NORM ALLY -FROM REACTOR CAVITY

CLOSED

FI G 5.l.i iDEALIZZD POOL COOLING SYSTEM

- .

_

. ,

,

,

lh hl50-

_

|| - L -

.. POWERPEAK VALUE =

DISCHARGED 71.12 x 10 BTU /HRiO A AT 48 HRS -

14 0

_

9- 4 -

- CCIN

g a PE AK VALUE= 13 4 6 F A _. 30$ AT 64 HRS

|'

.'o-

u. --@*

7POOL BULK TEMP-

-

a-@ 2< W

E ICASEI x-12 0O

m_ _;- Q NUMBER OF ASSEMBLIES = 240 3

@ TIME OF DISCHARGE = 48 HRSSPENT FUEL EXCH ANGERS = 2o:g:5 o_RHR EXCHANGER =0 o

0. Q.IO

4 -- 11 0

-

3 -

n'

TIME, HRS >

iM00' ' ' ' ' ' '2 0 20 4] so 80 40 0 izo_

FIG 5.1.2 ; NORMAL DISCHARGE,240 ASSEMBLIES

.

__

.

. .

.

124 p

^

i

_

i -11 -

-

10- PEAK 14 0-

VALUE=7~

I.12xlOBTU / DAY POWER9 - AT 48 HRS -

DISCHARGED_

'

N u_8 - -13 0o

PE AK VALUE= s_ 134.6 F W

AT 64 HRS7 - h d-

3 m-

g POOL BULK TEMP ge o

6 - 'O '-12 0o

- WEh CASEI

5~ - - 3 NUMBER OF ASSEMBLIES = 240~

m TIME OF DISCHARGE =48 HRS- QSPENT FUEL EXCHANGERS= 2

4_ $ RHR EXCHANGER =0it

- liOo

|_ o_

3-!

_

TIME, DAYS =

20 4 8 12 16 20 24 28_

,

FIG 5.l.3; NORMAL DISCHARGE,240 ASSEMBLIES

i

!.. - _ _. ... . . -- - _ _ . -

. ..

.

12 44

_

-15 011 -

_

_

10 PEAK VALUE=POWER 7BISCHARGED 1.12x10 BTU /HR

AT 48 HRS 4- 14 03

9 -

_

-

r

8 Z

E CASE 2 E._D NUMBER OF ASSEMBLIES = 240 %$ TIME OF DISCHARGE = 48 HRS CT 13 0

F7 -? ~ SPENT FUEL EXCHANGERS=2_

9 RHR EXCHANGER =1 3,

-5 a'

6 BULK 8g_I - TEMP 1o ~

12 0-

co'

5 0[uJ

-3 PEAK VALUE= 121.2 F _

_2 AT 58 HRS4

|_

-

- IlO|

3-!

/ TIME, HRS =| v

_

|

2EO 4Q 6O 80 lbO INO I4C

'~^

0

FIG 5.1.4; NORMAL.DISCHA.RGE,240 ASSEMBLIES.

:L

. .

,

.

14 hY-

110-15 0

T-67( Jr

gIO

5-IfM ! PEAK VALUE=

- 14 0g

1L2x 10" BTU / DAY o'IAT 48 HRS 4 (

ae !"x

~

POWER [ 33 0

6. DISCHARGED CASE 2NUMBER OF ASSEMBLIES = 240_

TIME OF DISCHARGE = 48 HRS-

SPENT FUEL EXCHANGERS= 2 -

RHR EXCHANGER = 16 -

_

POOL 12 0-

5- / BULKTEMP/-

PEAK VALUE=_

4 - 121.2 F AT 58 HRS

_

- 11 0

3.

_

2 05'

0 4 8 12 1 20 24 2_

fig 5.l.5', NORMAL DISCHARGE,240 ASSEMBLIESTIME, DAYS *

_.

. .

,

.

l% P~

-

9_ _

6~

PEAK VALUE= 9.65 x 10 BTU /HRPOWER AT 48 HRS8-BISCHARGED - 14 0

,

.

te7I -

E POOL* BULK.

'

6 -13 0

f ~'g ,

5@ / " PEAK VALUE= 130.6 F a-

/ AT 64 HRSo-

s_.no

4@ - 12 0

3 CASE 3 1

i@NUMBER OF ASSEMBLIES =2OO o

~

TIME OF DISCHARGE =48 HRS 23!- SPENT FUEL EXCHANGERS=2 N-

| RHR EXCHANGER =0 x- _J

B3-11 02 -

_

1

| l-

-

,

!_

TIM %, HRS=

' 'C3 20 40 60 80 10 0 l 14 0'

._

FIG 5.l.6; NORMAL DISCHARGE,2OO ASSEMBLIES

_

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

. .

,

.

g 50_

-

N'|L 'N Jgh POWER -14 0

o fEAK VALUE= DISCHARGEDi 9.65 x 106

g BTU / DAY

r AT 48 HRS_

$ u_An ~a65 4 2 - 13 0g5 -32 POOL BULK TEMP 5

o-PEAK VALUE =130 6 F

- '

- AT 64 HRS g4 -I20

CASE 3_

NUMBER OF ASSEMBLIES = 200TIME OF DISCHARGE = 48 HRS -3

-| SPENT FUEL EXCHANGERS = 2- RHR EXCHANGER = 0

2 -- 110

_

i _

~

Tl ME , DAYS >' ' ' ' ' '% 4 8 12 16 20 24 2g00!

-

t

| FIG 5.l.7; NORMAL DISCHARGE,200 ASSEMBLlES|

s

. .

..

.

104 n

-

- ('

.I 150--

9- /L_,

- POWER PEAK VALUE= 9.65 x 106DISCHARGED / BTU / HR AT 48 HRS [8 -'

Cc

.- ? -

p - 14 0/

7-ce ,'-

w |

-E / CASE 4'o

x NUMBER OF ASSEMBLIES =200 C6$ TIME OF DISCHARGE = 48 HRS 3 '

[$ SPENT FUEL EXCHANGERS: 2 F-

y / RHR EXCHANGER = 1 Yl30'5,-M 55 _;

^ '

O. cc@~: tu

4' o3U_- POOL

BULK - 12 0

3- TEMP' '

~__

PEAK VALUE = 1185 F-

2- AT 58 HRS

-

.

- Il0| -

_

T I M E, HRS 2 "

' ' '

Cg 20 40 60 80 I 120 14 0'0 "-

FIG 5.l.8', NORMAL DISCHARGE,200 ASSEMBLIES

. . ._ _ .- . -- . .-

. .

,

'

103 A

Jni~

-l50g_

-

\ E_'

8- a:.

3_ PEAK VAUJE =

D SCHARGED9.65 x 10-6 BTU / DAY '\ y- 14 0

7- AT 48 HRS 'N dx aN.s~

.

a-

O_6- A k

-w

E CASE 4-

13 05 NUMBER OF ASSEMBLIES = 200-

5- w TIME OF DISCHARGE = 48 HRS@ SPENT FUEL EXCHANGERS=2_

- yg RHR EXCHANGER = 1_

4 --d Ooa

>- x& ..w m

lb- 12 0

,

3o- POOL BULK TEMP

~

_

2 -

PEAK VALUE =l|8.5 FAT 58 HRS_

- 11 0

,

-,

TIM E,, DAYSr -

b 8 12 16 20 24 28' '

10 5_

FIG 5.l.9NORMAL DISCHARGE,200 ASSEMBLIES

..

,

\.

|2Gg o

_

15 0-

PEAK VALUE:6-

24.7 x 10 BTU /HRPOWER AT 144 HRS

DISCHARGED _

20 -

['

A

1 3',/ PE AK VALUE= 145.4 ''F -{40

TI ^AT 150 HRS'

7, &

W~ o,_.Q-

15h O 2--

Wx

i a

POOL BULK TEMP s-13 0;s- < ki b _J-

', (n Oi 5 0-

10'-- $ CASE 53 NUMBER OF ASSEMBLIES = 724>

2 TIME OF DISCHARGE = 144 HRS 12 0-

SPENT FUEL EXCHANGERS=2-

RHR EXCHANGER =1

'

_

5-

!

IlO-

_

TIME , HRS 2

l0M' ' ' '

0 40 80 NO i0~ 20 240 28c0 '

, _

FI G 5.1. lOFULL CORE DISCHARGE,724 ASSEMBL!ES

I. . - . . -

. .

,

.

A \ PEAK VALUE = 2.5 x 10 BTU / DAY7-

AT 148 HRS- 60

|- Dea ( -

20 -$, POWERP DISCHARGEDg-

PEAK -14 0-0 VALUE =

,

@ 145.8 F

y AT 150 H RS_

15

y POOL -13 0.- g BULK

gTEMP-

9- ^10 C ASE 5 o'-

NUMBER OF ASSEMBLIES = 724 3TIME OF DISCHARGE = 144 HRS H-- 12 0SPENT FUEL EXCHANGERS = 2 y-

RHR EXCHANGER = 1 ym_.J -

5 0I a.

-

11 0

*TIM E, DAYS _.

' ' ' ' ' ' '0 E0 5 10 15 o 25 o 3D_

FIG 5.1. l|FULL CORE DISCHARGE,724 A3SEMBLIES

._ _ __ -

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ . . . -_________ . _ -_ _ _

_

) ) ) -,

.

IDE All Z ED OU T LIN EOF POOL BOUNDARY

IDE ALIZED OUT LIN E 7-- A C T U A L OU T LIN E OF<

RACK AS S E MBLYOF R ACK ASSEMBLY /

AC TU AL OUTLINE /OF POOL

/ /,,

'''//,

j ,/ .,,

'

,

/ '/ / s'' ,. /,/ ,/ ,/ ,/ / ,

,- / '/ ,

RACK ASSEMBLY, / / ,/,,'

-

'-

'''' 'j 's , i

' '', / ' '/ ,,

,

| / // ,'

/ |// //

h-. .f3'

1.t.. i_. .W.]- ASS UMED ADDED:_. _ .. p ) . _

FUEL ASSEMBLIES,

/

Figure 5.2.1 Rack Space Enveloping Cylinder(Quad Cities Station Units 1 and 2)

. .

r.

8. RADIOLOGICAL CONSEQUENCES-

8.1 Objectives and Assumotions

The radiological consequences of expanding the storage capacity

of the spent fuel storage pool have been evaluated with the objective

of determining if there is any significant additional radiological

impact, onsite or offsite, relative to that of the currently author-

ized spent fuel storage pool. The principal factors considered in

evaluating the additional radiological consequences were the fol-

lowing:

o Operating experience and measurements.

o Reduction in decay heat generating rate and fuel tempera-

tures with time following removal from the reactor.

o Age and nature of the additional fuel to be stored.

-

o Analyses of radionuclide releases to the pool water from

failed fuel.

In addition, the radiological impact to operating personnel has

been evaluated to ensure that such exposure remains as low as isc

! reasonably achievable.[

Each spent fuel pool is currently authorized to store approxi-

mately two full cores. By comparison, each expanded spent fuel

storage pool can accommodate more than five full core loads. The 2

|additional storage capacity will be used for aged fuel which has been

out of the reactor 5 years or more. It is important to note that the

difference between the radiological impact for the currently

authorized storage pool capacity and the expanded storage pool

capacity is attributable entirely to the presence of additional aged

fuel in the expanded spent fuel storage pool.

-

! 8-1|

_ _ _ _ . __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. .

.

The radiological consequences of storing the additional quantity| of aged fuel have been evaluated. To ensure a conservative evaluation

of the storage of failed fuel, it was assumed that the spent fuelstorage pool is entirely filled with high-burnup spent fuel (28,500Mwd /MtU burnup), ranging from newly removed fuel (1 core load of 724fuel assenblies) to aged fuel witt a cooling time of approximately 18years. The maximum fission-product inventory in the stored fuel in

2each pool would result from an idealized fuel cycle in whichapproximately 181 spent fuel elements were removed from the core andplaced in the pool annually. With this fuel cycle, the expandedstorage pool capacity, when completely filled, would contain thefollowing:

(1) For currently authorized 724 newly removed assembliesstorage capacity (full core load) and 4

refueling discharges of 181

assemblies with storage

periods of 1, 2, 3, and 4"'

years, respectively.

(2) Aged fuel in expanded 13 refueling discharges of

storage capacity 181 assemblies with storage

periods of 5 to 17 years and2

any remaining capacity (up to170 assemblies), containing

fuel stored for 18 years.

Reduced fuel burnup or increased cycle length would result in a -

lower fission-product inventory or longer storage (decay) periods.Thus, the assumed storage pool composi tion. should result in a con-

servative estimate of any additional radiological inpact due to theexpanded storage capacity.

n

8-2

_ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _

. .

.

.

8.2 Operating Experience,,

8.2.1 General Industry Experience 2

1'

In a survey of spent fuel storage pool experience, Johnson, at

Battelle Pacific Northwest Laboratories, has shown that typical con-

centrations of radionuclides in spent fuel pool water range from 10-4-2pCi/ml, or less, to 10 p ci/ml, with the higher value associated

with refueling operations. Isotopic measurements of the nuclides con-firm that a major fraction of the coolant activity results from

activated corrosion products dislodged from fuel element surfaces

during refueling operations or carried into the spent fuel pool water

(with some fission-product radionuclides) by mixing the pool water

with primary system water during refueling. These sources of storage

pool radionuclides depend upon the frequency of refueling operationsand are basically independeat of the total number of fuel assembliesin storage.

Once fuel-handling operations are completed, the mixing of pool''

water with primary system water ceases and these sources of radio-

nuclides decrease significantly; only dissolution of fission-products

absorbed on the surface of fuel assemblies and low levels of erosionof corrosion-product (crud) deposits remain. With aged fuel (5 or

more years storage), neither of these latter sources would be expected4

to contribute significantly to the concentrations of radionuclides in

the storage pool.

In view of the above, it is concluded that the additional storage

capacity of the expanded spent fuel pool will not measurably alter thecurrerfly approved radiological impact or impose any significant

additional burden on the cleanup system as a result of corrosion-

product radionuclides or fission-product carry-over from the primarysystem during refueling operations.

'

During storage, the level of gamma radiation from fission

products in the fuel decreases naturally due to radioactive decay.,s

-

8-3 5,

. _ . .. ._ - - - - .

. . +,

.

Because of this decay, the contribution of the ag 2d fuel to the dose,

rate at the pool surface by direct radiation will be very small (< 5%)compared to that from the more-recently-discharged fuel. Thus, it is

concluded that the occupational dose rate above the surface of thepool from direct radiation will be essentially the same as that forthe currently autho.rized storage pool.

I

8.2.2 Related Plant Experience f!

8.2.2.1 Radionuclide Concentrations in Spent Fuel Pool Water<

Measurements have been made of the principal radionuclide

concentrations in both Quad Cities fuel storage pools during reactoroperations. Table 8-1 summarizes these measurements. As shown in,

Table 8-1, the pool water radionuclide concen tra tions are not

significantly affected by the number of fuel assemblies stored in thepool; over three (3) times as many fuel assemblies are stored in the QCunit 1 pool as in QC Unit 2 pool, but both pools have essentially the

''same Cs-134, Cs-137, and Co-60 radionuclide concentrations. This 2

observation lends credibility to the expected low contribution fromaged fuel in storage.

Similar measurements made at the Dresden Unit 2 pool (which issimilar to the Quad Cities pools) indicate that the contribution, ifany, from aged fuel will te very small or negligible in comparison tothe higher activity levels (especially during refueling) of freshlydischarged fuel. The Dresden measurements also show that the higherradionuclide concentrations which are measured during refueling

operations rapidly (within 2 months) drop to near the pre-refuelinglevels even though the pool contains the freshly discharged fuelremoved from the reactor.

With the expanded spent fuel storage capacity, the contributionfrom the aged fuel is correspondingly expected to be very low ornegligible in comparison to that from recently discharged fuel or from*

primary system carry-over during refueling.,,

8-4

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

. _ __

.

,

.

Table 8-1 Observed Radionuclide Concentrations In~

Spent Fuel Storage Pool Water

Fuel

Plant Assemblies FCi/ccDate Status in Pool Cs-134 CS-137 Co-60

QC Unit 14/27/81 Optg. 1139 2.6 x 10-4 7.9 x 10-4 7.1 x 10-55/18/81 Optg. 1139 5.5 x 10-5 1.8 x 10-4 1.8 x 10-4 '

5/24/81 Optg. 1139 4.5 x 10-5 1.7 x 10-4 1.9 x 10-56/1/81 Optg. 1139 9.2 x 10-5 3.0 x 13-4 2.9 x 10-4

i

QC Unit 24/27/81 Optg. 353- 2.4 x 10-4 7.8 x 10-4 2.7 x 10-4 |5/18/81 Optg. 353 7.1 x 10-5 2.2 x 10-4 9.5 x 10-45/24/81 Optg. 353 6.5 x 10-5 2.6 x 10-4 6.1 x 10-56/1/81 Optg. 353 8.5 x 10-5 2.8 x 10-4 2.1 x 10-4

8.2.2.2 Pool Cleanup System Operation_

2In the Quad Cities spent fuel storage pools, operation of the ,

cleanup demineralizer system and frequency of resin replacement isdetermined primarily by requirements for water clarity rather than the

loading of fission product radionuclides. The amount of suspendedparticulate material that must be removed to maintain the desired

vater clarity is determined by the frequency of refueling operations

and is independent of the number of fuel assemblies stored. Thus, the

expanded capacity of the Quad Cities storage pool will not

significantly alter either the frequency of resin or filter media

replacement above that currently experienced, or the personnel

radiation exposures during. maintenance operations.

-

8-5

. .,

.

8.2.2.3 Fuel Pool Radiation Levels

Measurements of the radiation levels above the spent fuel storagepool in both Quad C ties and the Dresden plants confirm that the doserates are essentially independent of the number of fuel assemblies

stored. On April 24, 1981, the measured dose rata above the QuadCities unit 2 pool was 4 mr/hr with 353 assemblies in the pool and also4 mr/hr above the Quad Cities Unit 1 pool with 1139 assemblies stored.

The average radiation dose above both Quad Cities pools duringthe period from January to April 1981 was 4-6 mr/hr. "omewhat higher

radiation dose rates (up to 15 mr/hr) were observed above the QuadCities pools during refueling operations, decreasing soon after

completion of ref ueling to the 4-6 mr/hr range. Expanding the storagecapacity of the spent fuel pools is thus not expected to significantlyalter the radiation dose rates over the pools above that currentlyexperienced.

In order to ascertain if there were crud depositions on the pool--

walls, measurements made above the center of the storage pool and atthe pool edge were essentially the same, indicating that there are nosignificant crud depositions on the walls of the pool that mightcontribute to a higher dose rate at the pool edge. Visual obser-!vations also confirm the absence of any significant crud deposition on!

2! the pool walls.

6

!

Radiological surveys in the vicinity of the spent fuelstoragef| pools indicate that the major sources of the observed dose rates are jt

derived from miscellaneous pieces of equipment in the vicinity of the I| pool or utilized in fuel handling operations (e.g. , sipping equipment, i

grapple attachments, vacuum hoses, etc.). None of these miscellaneoussources of radiation are affected by the number of fuel assemblies|stored. Consequently, expanding the storage capacity of the Quad

Cities spent fuel pool will not significantly alter the radiation dose

to personnel occupying the fuel pool area.1

mem

!

8-6.

.-

. .

.

8.2.2.4 Airborne Radionuclides-

Because of radioactive decay, Kr-85 will be the only significant

contributor to any potential increase in airborne radionuclide

concentrations above that currently authorized. For the current Quad

Cities storage pocls, Kr-85 has not been detected at the reactor

.

building vent (i.e., any Kr-85 present is less than the minimum10-6 pC/cc to 9 x 10-6 C/cc). AsI detectable concentration of 6 x

discussed in Section 8.3.3 below, no significant increase in Kr-85

concentration f rom the aged fuel is expecteo. Consequently, expanding,

the spent fuel storage capacity will not impose any significant

radiological burden from airborne radionuclides.

8.3 Consecuences of Failed Fuel

Escape of fission-products from failed fuel stored in the spentj fuel pool will contribute to the radionuclide concentrations in the

pool water. However, calculat!.ons described below indicate that the,

|"' radionuclide concentrations from failed fuel are considerably less

than the concentrations of corrosion-product radionuclides and, there-

fore, the aged fuel in the expanded storage pool will not contributesignificantly to the onsite or offsite radiological impact.

The decay heat generated in spent fuel rapidly decreases (by;

radioactive decay) following removal from the reactor and, in the agedfuel, will be very small ( < 5% of that in freshly-removed fuel). Fuel

temperatures and internal gas pressures will correspondingly decreasewith time. Johnson also cites evidence to confirm that UO2 is inertto the relatively-cool water of spent fuel storage pools. Therefore,

the release rate of fission-products f rom any defective rods among theaged fuel is expected to be negligibly small.

Release of fission-products from failed fuel probably resultsf rom water leaching or dif fusion of material plated out or absorbed in

-

8-7|

|

|'

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

*~

..

.

the fuel-clad gap of the fuel element during operation in the reactor.~

Once the material in the gap is depleted, further release will be very

small. Most of the fission-products are absorbed (retained) in the

fuel matrix and can escape only by diffusion through the UO . At the2

temperatures of the fuel in the spent fuel pool, the diffusion

coefficient will be extremely small.2,

In his survey, Johnson indicates that numerous fuel assemblies

with one or more defects have been stored in several spent fuel pools

without requiring special handling. Detailed analysis of the spent

fuel pool water confirmed that fuel elements with defects do not

continue to release significant quantities of radionuclides for long

periods of time following removal from the reactor. Nevertheless, the

calculations described here in Sections 8.3.1 and 8.3.2 were based on

the very conservative assumption that the rate of fission-product

release remains the same as the rate for newly discharged fuel.

Both Johnson, at Battelle, and Weeks, at Brookhaven National

Laboratory, have reviewed the corrosion properties of Zircaloy.-

cladding and the integrity of spent fuel elements stored for long

periods of time. They conclude that the co :rosion of Zircaloy

cladding in spent fuel pool water is negligibly small and that there

is sufficient evidence of satisfactory fuel integrity to justify

expanded storage. Consequently, there is not expected to be any

significant deterioration of stored fuel that might lead to additional

fuel failures in the expanded-capacity spent fuel storage pool.

8.3.1 Methods of Analysis

To assess the maximum potential radiological contribution from

failed fuel, the inventory of fission-products in the spent fuel was

calculated with the ORIGEN code, conservatively assuming that all

fuel was discharged from the core at 28,500 Mwd /MtU burnup. Experi-

mental values of escape rate coefficients in cool water shortly after

-

8-8

J

.. - . = _ .

,

..

..

.

discharge, as derived by Westinghouse,5 were used to calculate the^

fra<.'.ional release of fission-products from failed fuel, and it was

assumed that there were it fuel element failures. These escape rate

coefficients (listed in Table 8-2) were assumed to be constant | 2throughout the storage period, although it is known that fission-

product release from failed fuel is strongly dependent upon the

temperatures within the fuel pin. As natural radioactive decay

occurs, decay heat generation in the fuel becomes less and, as a

consequence, the fuel temperatures and internal gas pressures are

reduced. Furthermore, the inventory of leachable fission-products

becomes depleted and release f rom the bulk U02 by diffusion becomesextremely low.3

Thus, within a few months after discharge, the fuel temperatures

and effective leak rate coefficients decrease, and further leakage is

: educed to relatively insignificant levels.1

The percentage of failed fuel that exists in the stored fuel,

averaged over a large number of reactor cycles, is uncertain.-

1Johnson estimates that an average of 0.013 should be achievable. TheNRC, in NUREG-0017, cites 0.12% as a representative value.

; Nevertheless, to establish a conservative upper limit, the calcula-

tions reported here were based on the assumptions that 1% of all

stored fuel is failed and that constant leak rate coefficients, cor-;

responding to those measured shortly af ter shutdown, apply over the

storage periods. Concentrations of released fission-products were

calculated f rom the dynamic balance between the source term (leakagefrom the assumed failed fuel) and the rate of' removal by (1) radioac-

tive decay and (2) the spent fuel pool cleanup system (using

demineralizer cleanup efficiencies cited in NUREG-0017). This methodof analysis is similar to that used in NUREG-0017.

j

.. s

8-9

-.

.

. .

.-

e

Table 8-2 Escape Rate Coef ficients into Cool Water for2Spent Fuel in Storage Pool-

Element Escape Rate Coefficient (sec~1)

I 1.7 x 10-12Rb, Cs 3.0 x 10-12<

Mo 1.8 x 10-12Te (0.9 x 10-12)*;

-15Sr 8.5 x '.0-16Ba 5.8 x 10,

**Zr 1.2 x 10-16<

* Escape rate coef ficient for Te assumed to be in same ratio to Mo, asgiven in NUREG-0017.** Assumed applicable to all other nuclides.

.

8.3.2 Fission-Product Radionuclide Concentrations

Based upon the method of analysis described above, the concentra-tions of fission-product radionuclides in the spent fuel pool werecalculated at several times following unloading of a full core into--

the spent fuel pool, with the remainder of the pool assumed to befilled with older fuel. Results of these calculations are summarized

2in Table 8-3, assuming continual operation of the spent fuel pool: water cleanup system. These calculated concentrations of . fission-

product radionuclides are directly proportional to the assumed it

failed fuel and would be a factor of approximately 8 lower for the

0.12% failures estimated in NUREG-0017 as a typical weighted averagevalue based on operating experience in a number of reactors. Of thefission-product radionuclides released, Cs-137 is the dominant

-6activity from the aged fuel (calculated to be a maximum ot 2.2 x 10-6pCi/ml with 1% failed fuel) . Low levels of I-131 (8 x 10 FCi/ml) and

-6Mo-99 (3 x 10 pCi/tti) are calculated to be present as a result of

leakage from a full core load of newly removed fuel (with 1%

failures). However, in the aged fuel, these nuclides have decayed andthere are no significant quantities of I-131 or Mo-99 remaining.

-

|

! 8-10

!!

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

. -

. .;

.~.

Table 8-3 Fission-Product Radionuclide Concentrations 2'

in Fully-Loaded Spent Fuel Storage Pool with it Failed Fuel

Concentration (HCi/ml)For Currently 11cremental Additions

Approved Storage due to ExpandedTime (days) Capacity * Capacity * Total

5 2.08 x 10-5 2.62 x 10 5-6i 10 1.43 x 10-5 2.62 x 10-6 2.34 x 10 5

-6 -6 1.69 x 10 520 9.35 x 10 2.61 x 10 1.20 x 10-

-6 -6 -530 7.54 x 10 2.61 x 10 1.02 x 1050 6.23 x 10-6 2.60 x 10-6 6

-6 -6 8.83 x 10 675 5.67 x 10 2.59 x 10 8.26 x 10-100 5.37 x 10-6 2.57 x 10-6 7.94 x 10-6

*See Section 8.1 for a description of composition.

Even with 1% failed fuel, the radionuclide concentrations in thespent fuel pool water are dominated by those from corrosion productsand carry-over from the primary coolant system during refueling.Furthermore, since the release rate from the aged fuel will be consid--

erably smaller than that indicated in Table 8-3 (due to the lower 2

temperatures and release rates in the fuel elements), the actual

contribution from the aged fuel will be negligibly small in practice.It is also expected that the percentage of failed fuel, averaged overthe reactor lifetime, will be considerably less than 1%. Thus, it is

concluded that the expanded-capacity spent fuel storage pool will notincrease the radionuclide concentrations in the pool water signifi-cantly above those for the currently approved sp<;nt fuel storage pool.Consequently, expanding the storage capacity of the spent fuel poolwill neither alter the onsite or offsite radiological impact norsignificantly increase the burden on the spent fuel pool cleanupsystem, as a result of fciled fuel.

R.3.3 Gaseous Releases from Failed Fuel

Because of the half-lives of the noble-gas radionuclides, onlythe release of Kr-85 (Tg of 10.76 years) has the potential of_

8-11

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

o. .

.."

increasing the radiological impact to the reactor building atmosphere | 2~

as a result of expanding the ecpacity of the spent fuel storage pool.(Short-lived noble-gas radionuclides and other volatile Fission-

1products, such as iodine, are not present in the aged fuel.) Johnsonconcludes that the radioactive fission gases will have been largelyexpelled from defective fuel rods during reactor operation and,

therefore, are not available for release during fuel storage. This is

expected, since the noble gases are chemically inert and there are noplate-out or hold-up mechanisms in the fuel-clad gap of the fuelelement. Measurements above the Quad Cities storage pools failed todetect any Kr-85 above the minimum detection level. 2

The small amount of chemically inert Kr-85 that might be absorbedon the surface of a fuel assembly and released slowly during storage,is believed to be insignificant, particularly in the aged fuel. SinceUO2 is chemically inert to cool water, diffusion of Kr-85 entrappedwithin the UO2 fuel matrix would be the remaining source for Kr-85release. Based on the method outlined in the proposed ANS 5.4

2s tat.da rd on fission gas release, the diffusion coefficient in the--

aged fuel at spent fuel pool temperatures will be negligibly small (ofthe order of 10-40). Consequently, diffusion release of Kr-85 from

aged fuel will be negligible in accord with Johnson's findings.1

It is concluded that the incremental radiological impact f rom therelease of Kr-85 with the expanded-capacity spent fuel storage poolwill be negligibly small.

8.4 Exposure for the Installation of New Racks

The existing spent fuel racks will be removed, and the new rackswill be installed in a manner which will maintain occupational expo-sure to levels as low as reasonably achievable (ALARA). The following

-

8-12

~ -. - . . . ...

. . .

...-,e

methods for the disposal of the existino racks are currently under-

review by CECO:

o Crating and shipment of the racks in "as-is" condition.

o Decontamination and shipment,

Dismantle and volur e reduction, with or without prior decon-o

tamination, and shipment of waste.

The final decision concerning the disposal of the existing spent

fuel racks will be based on project needs and experience gained from

rack disposal at Dresden.

8.5 Conclusions

Based on operating experience and the analysis cf potential

releases, it is concluded that expanding the capacity of the spent

fuel storage pool will not significantly increase the onsite or off-^

site radiological impact above that of the currently authorized

storage capacity. Similarly, the expanded storage capacity will not

impose any significant additional burden on the spent fuel pool

cleanup system, and no modifications to the cur:ent radiation2

protection program are needed.

|

|

:

|

|

,

i

!

;

m

T

8-13

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

I. . .

.-o'

REFERENCES TO SECTION 8-

1. A. B. Johnson, Jr., " Behavior of Spent Nuclear Fuel in Water PoolStorage," BNWL-2256, September 1977

2. ANS 5.4 Proposed Standard, " Method for Calculating the FractionalRelease of Volatile Fission Products from Oxide Fuel."

3. J. R. Weeks, " Corrosion of Materials in Spent Fuel StoragePools," BNL-NUREG-2021 (Informal Report), July 1977.

4. M. J. Bell, "ORIGEN-The ORNL Isotope Generation and DepletionCode," ORNL-4628, May 1973.

5. J. M. Wright, " Expected Air and Water Activities in the FuelStorage Canal," WAPD-PWR-CP-1723 (with addendum), undated.

n

A

8-14