Text

Criticality Analysis of Prairie Island Units 1 & 2 Fuel Racks
	ML20248J888
Person / Time
Site:	Prairie Island
Issue date:	03/31/1989
From:	Boyd W, Cobb R, Krieg D; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20248J877	List: ML20248J874; ML20248J882; ML20248J888;
References
	NUDOCS 8904170133
	Download: ML20248J888 (39)
	v • d • e

_ _ .

,, i

.o:

l Exhibit C i I

. Prairie Island Nuclear Cenerating Plant License Amendment Request Dated April 6, 1989 Criticality Analysis of Prairie Island j Units 1 and 2 Fuel Racks Prepared By-Westinghouse Commercial Nuclear Fuel Division March 1989 .

J l

.i I

i i

%"1$ P

6 6 \ *

=, O i

\ )

\ . . . ... .

, ,. . .......+,...

. . ... . . ~. ~ * = . . +

+*.4 e y . *4 {

. . . . .. .. .m...,

p ....,.,s.."r*=

.. ate . ... * * ** . . . . * * * " *

a

. : ., .. .. trese r._..

,,,,., m' e me.n.j ; * *J. . *.e-e.*

gT.,'w .e.

e e e. y

.ger~ . s .p e e

y*e

.'..*.'7***....

q.. .% %~'

e J pr vg j ~..

.t-gpgf;,g p -

,s s..e

m . .

. 7=. . r. a .

. - . ,.* 9 g' ** li E :p'." =

aaes=.

- - l..'

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

.**r/ "1.'+*" *P.,""

. e. . a = ** * *m a**.1. . **. * * .

t . J. C.'

e' T .* l.t ='/ =* * ."*'. . ' * * =d.D. EI.=e.T.

..,o **s'-** -..# .

..m.,..

m- + . w. . .. ,4. ..

.117*!:. T&.~ *J. .' *4t . ***.t.

.-.s*

). .

- ******'* .u.. .%. .,,q_. . _ _ _

.a. J4..s w, .e.t* 1.*"+

.m . . ~. . . r s .- - . - - , .-

.e

=T t.J "k. .,." .*, res e.-,***m 4<< t a.e.vgu1.

.*<.. ss.**

,.,,a.. e. vere.=a.-

g '..e......- . 4*re..

_...w.~ . .. m>

Tag - CRITICALITY ANALYSIS OF PRAtRIE ISLAND G.M_ ._ r y.,e"ss.ca [.i',

p,C w.er

.=7

--a- - =~ '

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

UNIT $ 1 AND 2 FUEL RACKS .. .~a.../. ~ . sic ..-..-w--

~~ - *

.. . . ~.- . .. ,

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

u

* * ' .-. - . . , ...-....~

.,... , , . . . . ~

~ . . ~ . . -. .ir .- . ; ~ . ..

.....2...

..,,,...~.e,

.w =..

. eu. . e s -r o w m ? ss 'e r ~. ._... , .;. n .x;;.,

.o .- . ..

..m.

.~.

w ? L;~.* ~.~ *.: "'* * ' j f - . , . . .

.. ...--,gf.a. p ,r e .r. w -

.,; . .<*s '; ,

)/e;.a.s ., l 4:

, . . ,..d....w-

. . . .. ..~. . .

f'... 1,." ..

~

m... - -. .

~

'. T a..M."_. J. .'=TT ' 17 '._L.'J.. '.W. . - .

_ , = *. .*, , , ,.*****=.~**,,'*,.**..%,n

.. .. . . a ;r ; y ';;; ;-n::P. **

.. ..,.u.wi......++

7 . , ,

*'*** ** 3, _g .;p.,y y; p ,,, g . , , , , , ,,

). ( s ..,,s ur ,L.,44 :wJA .. ,a . e. 5 . . p.*

3%s. msg .4. . C.ed-J ** e

,. .......a. .* u-....,a..

. . .... . . -se. en. .. a. s,'s s e-~-. - * * * .g. q r .r,. ;y s . .: . .y _

.e

w.~.

. - * ,+ . . .

..,,.o.

.. . ..,...4,.....~,,%

...7

j

-4 . . , , , * ,

.. . . . . \ , , , , _ . . .

..- 'm e;

. . . -. -=

- ., .. .. .. ,.... . . .. .... . . ... a,. .-.- . . .-

.y.

g

. - , . . ...3. - -"- .

...~..++;. : r.* *

m. , . . . [ y,' *;,4 4 y.,, J.,,

~~p m- ?.,.. . . . ~ . . .,..,..~4..[,~

* ~

.,,,.%...*~.:. . .

., .. n,,,rir,.J5,,g.,,..'. *. g. ... 1

~. ,- - ~ - . ~ . .~. .... m.-- . s e * - .. . . - . .

. ~

\ ..,a... .,.....--..

. ._ ..;. g . . . , . . . - . . . . .

...e, . .. .- >

.'.% . :2 M, . 9 *, '*,a..

s

.-. .~ . .- . .

\ ..-.L .

j t

i

. .....,-. . . . ..- . . . . . - . . . - . - . 3 .

.. . .. , .. . m m.. . m .~ t .._.n _.,f.,.

s

. s. 3

= . _ - _a ._

s .. e. =a..

,..:. * ^

J. D'.;~ M '- , m.

. _ ~. m *&c ut.4 is n y n.% me.t*n s

. = - : ,,,,

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

,... ._.. ...._..__,,..-y . - . - -a _ -.,,,,r-e

+ - . . --e ._ . . ..

.****~*r**.; * * * * * .-

=... . .. . .

e.* /a . . i . .

,,...,,.,,....e. . . . . ~ .

~

aie.=.m = -. - . +-".*~- a .e. =

. ..m..___

. . .. m . m __ _.-... . . . .. *~.- .

.=:s. y =-~ =- m.=. - . ..-

, = =.=.==

E ME E E5 E m

W --

Westinghouse w U.. Commercial _ Nub.. =fear Fuel s_ Division 53 E BE h.

E E E 55 5 ME r_...-

. . , . E E E E

.n.... , . . _ . ..n.

m m. -

. _ . - . _ _ _ _ _ . - _ - ~ - < ~ ~ . - = =_-

A EME E E BEE

. *= = .. -

= = . *

= .

. i ".*. *f. .

E5 E E5 m

.* * ] 3

, yyp g . . - * * '

4

\

D t

i CRITICALITY ANALYSIS OF PRAIRIE ISLAND UNITS 1 AND 2 FUEL RACKS March 1989 W. A. Boyd D. J. Krieg R. C. Cobb l

) W. A. Bordogna

O ;

)- .e t-I t

1 o

l l TABLE OF CONTENTS 1.0 Introduction ............................................... 1 1 1.1 Design Description .................................2 1.2 Design Criteria ...................................2 ,

s 2.0 Criticality Analytical Method .................................. 3 3.0 Criticality Analysis of Spent Fuel Racks ......................... 4 3.1 Reactivity Calculations ..............................4 3.2 Postulated Accidents ...............................5 4.0 Criticality Analysis of Spent Fuel Racks With Poison Gaps .......... 7 4.1 Fuel Burnup Credit .................................7 4.1.1 Reactivity Equivalencing .........................7 4.1.2 Analytical Methods' ............................ 8 4.1.3 Reactivity Calculations ..........................9 4.2 Soluble Boron Concentration ........................... 11 4.3 Checkerboard Fuel Loading ............................ 11 5.0 Criticality Analysis of Fresh Fuel Racks ......................... 13 5.1 Full Density Moderation Analysis ....................... 13 5.2 Low Density Optimum Moderation Analysis ................. 14 l 6.0 Acceptance Criterion For Criticality ............................ 16 Bibliography .................................................. 33 .

l l

l Table of Contents i

Q 6

.s f 'v I.

q l

!1 i

)

LIST OF TABLES 4

i Table - 1. Benchmark Critical Experiments (5,6) ................ 17 l

. Table 2. Fuel Parameters Employed in Criticality Analysis ......... 18 Table 3. Prairie Island Units 1 and 2 Fuel. Assembly Minimum Burnup vs

. inital-U'" Enrichment for the Sp'ent Fuel Rack ........... .19 Table 4. Comparison of' PHOENIX lsotopics Predictions to Yankee' Core 5 -j Measurements ............................... 20 (

Table 5. Benchmark Critical Experiments PHOENIX Comparison ...... 21- 1 Table '6. Data 'for U Metal and UO2 Critical Experiments .......... 22 i Table 7. Prairie Island Units 1 and 2' Spent Fuel Rack Maximum Reactivity vs Initial U'" Enrichment and Poison Panel Gap Size ...... -24 -l

.I 6

I i

i, 1

i :(

1. 1 i.

l l

j.

List of Tables ll l:

D ,

- - - - - _ _ _ _ . - - - . . _ _ a

e i L ..

) ..

L; l

LIST OF ILLUSTRATIONS !

' Figure 1. Prairie Island Units 1 and 2 Spent Fuel S'orage Cell Nominal Dimensions ................................. 25 :

Figure 2. Prairie Island Units 1 and 2 Spent Fuel Rack Layout ...... 26 Figure .3. Prairie island Units 1 and 2 Fresh Fuel Storage Rack Axial Layout 27 ;

Figure 4. Prairie Island Units 1 and 2 Fresh Fuel Rack Radial Layout .. 28 Figure 5. Maximum Spent Fuel Rack K.te versus H2OlUO2 Volume Ratio For Heavy Load Drop Accident ....................... 29 i Figure . 6. Prairie Island Units 1 and 2 Fuel Assembly Minimum Burnup vs. j Initial U Enrichment and Poison Gap Size for Storage in the f

Spent. Fuel Racks ............................. 30 Figure 7. Prairie Island Units 1 and 2 Region 1 Three of Four Fuel Assembly Loading Schematic ..................... 31 Figure 8. Sensitivity of K.e# to Water Density in the Prairie Island Units 1 and 2 Fresh Fuel Storage Racks ................... 32-1 I

List of Illustrations lii l

p 1

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

i

\

[ .

q l

l l

b

1.0 INTRODUCTION

The Prairie Isla'nd Units 1 and 2 fresh and spent fuel rack (SFR) designs de-I scribed herein employ . arrays of racks, which will be reanalyzed at a higher !

enrichment under various preconditions. The spent fuel rack design . is a poisoned rack, previously analyzed for storage of 14x14 fuel assemblies with enrichments up to approximately 3.9 w/o U'" (39.0 gm U*"/ axial em) utilizing every storage location.

This analysis will reanalyze the fuel rack arrays to determine the following:

1. The maximum fresh fuel enrichment that can be stored in all storage to- f cations in the spent fuel rack and maintain K.ve 5 0.95.

2. The spent . fuel rack reactivity increase that will result from postulated gaps ,

of up to four inches in the poison panel material. i

3. The fuel assembly burnup as a function of enrichment and poison gap size i that will maintain the spent fuel rack reactivity less than the 0.95 K.ve limit.

4. The soluble boron concentration needed to maintain K.ve s 0.95 with 4.27 w/o U'" fuel stored in all storage locations and four inch axial gaps in the poison panel material.

5. The maximum spent fuel rack K.#e with 4.27 w/o U'" fuel stored in three of four storage locations and four inch axial gaps in the poison panel material.

6. The spent fuel rack. Keet following a heavy load drop which could crush the fuel in the spent fuel pool with 1800 ppm of soluble boron.

7. The fresh fuel. rack reactivity that will result _with the storage of 4.27 tv/o U'" fuel under full moderator density and optimum moderator dent,ity conJitions.

Introduction 1 l

l

o 1.1 DESIGN DESCRIPTION The spent fuel storage cell design is depicted schematically in Figure '1 on page 25 with nominal dimensions given on the figure. The spent fuel rack layout is ;

~

shown in Figure 2 on page 26. The fresh fuel rack layout is shown in Figure 3 on page 27 and Figure 4 on page 28.

1.2 DESIGN CRITERIA Criticality of fuel assemblies.in a fuel storage rack is prevented by the design of the' rack which limits fuel assembly interaction. This is done by fixing the i minimum separation between assemblies and inserting neutron poison between l assemblies.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level ,

that the effective multiplication f actor (K.n) of the fuel assembly array will be~ ;

less than 0.95 with full density moderation as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and in Reference 1 and less than 0.98 with low density mod- .j eration as specified in the plant Technical Specifications.

1 i

l 1

i' 1.

i l

l i l~ .I l

l l

l.

Introduction 2 t

i

_ _ _ _ _ _ _ _ - _ - _ . _ _ _ _ _ - _ _ _ _ _ _ . _ - _ _ k

3 r

1 i

i l

l l

2.0 CRITICALITY ANALYTICAL METHOD The criticality calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which 'het racks are designed. This benchmarking data is sufficiently diverse to l I

establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.

The design method which insures the criticality stfety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codes for cross-section !

. generation and KENO IV"' for reactivity determination.

The 227 energy group cross-section library that is the common starting point {

for all cross-sections'used for the benchmarks and the storage rack is generated from ' ENDF/B-V data. The NITAWL program includes, in this library, the selfishielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim ilntegral Treatment is used. Energy and . spatial j weighting of cross-sections is performed by- the XSDRNPM program which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KENO IV"' which is. a three dimensional Monte Carlo theory program designed for-reactivity calculations.

A set of 33 critical experiments has been analyzed using the above method to !

demonstrate its applicability to criticality analysis and to establish the method bias and variability. The ' experiments range from water moderated, oxide fuel arrays separated by various materials (B4C, steel, water, etc) that simulate LWR fuel shipping and storage conditions

to dry, harder spectrum uranium metal cylinder ' arrays with various interspersed materials * (Plexiglas and air) ' that

- demonstrate the wide range of applicability of the method. Table 1 on page 17 summarizes these expe 3ments.

The average K.vf of the benchmarks is 0.992. The standard deviation of the bias value is 0.0008 Ak. The 95/95 one sided tolerance limit factor for 33 values is - 2.19. Thus, there is a 95 percent probability with a 95 percent confidence

- level that the uncertainty in reactivity, due to the method, is not greater than )

0.0n 9 ? N l

p Criticality Analytical Method 3

l I

I I 3.0 CRITICALITY ANALYSIS OF SPENT FUEL RACKS I A study was performed to determine the maximum enrichment of the most re-activity fuel assembly type that could be stored in the spent fuel racks and maintain a K.te $ 0.95. The spent fuel rack poison material was assumed to be intact with no axial gaps. Results of the analysis show that 4.07 w/o U'"

is the maximum enrichment which can be placed in the spent fuel racks. The model development and results are discussed in the following section.

3.1 REACTIVITY CALCULATIONS I The following assumptions were used to develop the nominal case KENO model for the spent fuel rack storage of fresh fuel using all storage locations:

I 1. Calculations for spent fuel racks analysis herein have shown that the i Westinghouse OFA fuel assemblies yield a larger K.ef than does the Westinghouse STD fuel assembiy when both fuel assemblies have the same U'" enrichment. Thus, only the W 14x14 OFA fuel assembly was analyzed (see Table 2 on page 18 for fuel parameters).

2. All fuel rods contain uranium dioxide at an enrichment of 4.07 w/o U'" over the entire length of each rod.

3. No credit is taken for any U or U** in the fuel, nor is any credit taken for the buildup of fission product poison material.

4. The moderator is pure water at a temperature of 68*F. A conservative value of 1.0 gm/cm' is used for the density of water.

5. No credit is taken for any spacer grids or spacer sleeves.

6. The array is infinite in lateral extent and finite in axial extent which allows Ie neutron leakage in only the axial direction from the array.

7. The minimum poison material loading of 0.040 grams B-10 per square cen-timeter is used throughout the array.

'ne "cmwm Km mfer norir . ' conditi ons mis e t from a rMderw%a of irw-

' chu:le- ' act tm ' 4 ! ' U .t a n t, .,.; luierances resuiung i, s.r: .h1 eruwnct n. 3 proces.s in addition to asymmetric positioning of fuel assemblies within the I Criticality Analysis of Spent Fuel Racks 4 I -

l storage cells. Studies of a6ymmetric positioning of fuel assemblies within the l- storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.tv . The sheet metal tolerances are considered along with construction tolerances related to the cell I.D., and cell center-to-center spacing. For the spent fuel racks this resulted in a reduction of the nominal 0.752" water gaps to their minimum values. Thus, the " worst case" KENO model j

-of the spent fuel storage racks contains minimum water gaps of 0.562" with symmetrically placed fuel assemblies.

I Based on the analysis described above, the following equation is used to de-velop the maximum K.fi for the Prairie Island Units 1 and 2 spent fuel storage racks with fuel storage in all locations:

~

K.v t = Kwor i + B m.inoo + B .ri + ( (ks)' wor.i + (ks)'m.inoa ] ,

where:

Kworsi = worst case KENO K.fr that includes material tolerances, and mechanical tolerances which can result In spacings between assemblies less than nominal

=

Bm.ince method _ bias determined from benchmark critical l comparisons l Bo.ri = bias to account for poison particle self-shielding kswersi = 95/95 uncertainty in the worst case KENO K.ve ;

ksm.inoo = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:

K.es = 0.9331 + 0.0083 + 0.0010 + ( ( 0. 00 50 ) * + (0.0018)* ] = 0.9477 Since K.et is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met with fuel enriched to 4.07 w/o.

3.2 POSTULATED ACCIDENTS Most accident conditions will not result in an increase in K.tv of the rack. Ex-amples are the loss of cooling systems (reactivity decreases with decreasing l

water density) and dropping a fuel assembly on top of the rack (the rack -

L structure pertinent for criticality is not excessively deformed and the dropped assembly has me c ther twalva inches of water separating it from the --t'vr fuel harghr. of .stomo ..uuntbt'r whee *uuct'.id..s incet e .c % l l

Criticality Analysis of Spent Fuel Racks 5

_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ i

t, .,

f t However, accidents can be postulated which would increase reactivity (i.e., or f dropping a fuel assembly between the rack and pool wall). For these accident conditions, the double contingency principle of ANSI N16.1-1975 is applied. This states that- one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be I assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

An analysis was performed to show that the presence of 1800 ppm of soluble i boron in the spent fuel pool would maintain the K.## $ 0.95 following a heavy load drop accident. It was conservatively assumed that all of the fuel storage cells in the spent fuel rack would be crushed together following a heavy load drop accident. The spent fuel rack H20/UO2 volume ratio was reduced by first reducing the storage cell spacing, then the inner cell space around the fuel as-sembly and finally the fuel rod pitch. This methodology is the same as that used in the existing fuel cask drop evaluation completed by Quadrex"'. The assumptions used in the model development are the same as those discussed in Section 3.1 except the model is infinite in all directions. The maximum spent fuel rack reactivity was calculated with and without soluble boron as a function of the H20/UO2 volume ratio. Additional uncertainties of 0.005 AK are added to the no boron results and 0.010 AK to the boron results tor. conservatism.

The results, shown in Figure 5 on page 29, show that the maximum fuel rack reactivity will be less than 0.90, including uncertainties at a 95/95 probability / confidence level, with 1800 ppm of boron.

The presence of approximately 1800 ppm boron in the pool water will decrease reactivity by about 20 percent AK. Thus, for . postulated accidents, should there be a reactivity increase, K.## would be less than or equal to 0.95 due to the effect of the dissolved boron. i I

I l

1 I

l Criticality Analysis of Spent Fuel Racks 6 j

I 1

l CRITICALITY ANALYSIS OF SPENT FUEL RACKS WITH 4.0 POISON GAPS ;

l A poison panel sensitivity study was performed to determine the reactivity in-crease that would result as a function of the postulated axial gap size in the poison panels. The methods and model used to calculated the spent fuel rack reactivity as a function of polson gap size are discussed in Section 4.1.3. The gaps are located at the axial mid-plane of all of the poison panels in the spent i fuel racks. Table 7 on page 24 shows the maximum reactivity calculated as a f function of gap size and enrichment. As can be seen from this data, many of the enrichment and poison gap size combinations will result in a fuel rack K.fr greater than the 0.95 limit.

The following sections discuss the analytical methods and models used to show that the K.tv limit of 0.95 can be met by taking credit for:

i

1. Fuel depletion.

2. Soluble boron in the spent fuel pool water.

3. Checkerboard loading of the fuel assemblies.

4.1 FUEL BURNUP CREDIT The data shown in Table 7 on page 24 was used to determine which fuel enrichment and poison gap size combinations will result in a fuel rack K.tv greater than 0.95. Fresh fuel assemblies at the selected enrichments are then depleted to determine the fuel burnup which will reduce the fuel rack K.vv to 0.95. This analytical technique is known as reactivity equivalencing.

l 4.1.1 REACTIVITY EQUIVALENCING The concept of reactivity equivalencing is predicated upon the reactivity de-crease associated with fuel depletion. A series of reactivity calculations are performed to generate a set of enrichment-fuel assembly discharge burnup or-dered pairs which eli ylc'c' an coulva!ca+ K.ee when the fuel is stored in $c ep%;

.Yes. b- Qim ener Jhe nov'vu n V . " i r. 9 %.

I Criticality Analysis of Spent Fuel Racks With Poison Gaps 7

O 1

J Figure 6 on page 30 shows the constant K.n contour generated for the Prairie

! Island Units 1 and 2 spent fuel racks as a function of the poison panel gap size. Note in Figure 6 on page 30, for the 2 inch gap size, the endpoint at 0 MWOIMTU where the enrichment is 3.97 w/o and at 1,451 MWD /MTU where the l

enrichment is 4.27 w/o. The interpretation of the endpoint data is as follows:

the reactivity of the spent fuel racks, with 2 Inch poison panel gaps, containing fuel at 1,451 MWD /MTU burnup which has an initial enrichment of 4.27 w/o is equivalent to the reactivity of the spent fuel racks containing fresh fuel having an initial enrichment of 3.97 w/o. It is important to recognize that the curves j in Figure 6 on page 30 are based on a constant rack reactivity of 0.95 and not on a constant fuel assembly reactivity. The data in Figure 6 on page 30 is also provided as Table 3 on page 19.

4.1.2 ANALYTICAL METHODS The data points on the reactivity equivalence curve were generated with a transport theory computer code, PHOENIX *. PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMS*

library is used with PHOENIX. ]

j

. q A study was done to examine fuel reactivity as a function of time following #

discharge from the reactor. Fission product decay was accounted for using CINDER" '. CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years af ter discharge. The fuel reactivity was found to reach a maximum at approx-imately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months af ter discharge. At this point in time, the major fission c product poison, Xe'" , has nearly completely decayed away. Furthermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 30 years I following discharge. Therefore, the most reactive point in time for a fuel as- i sembly after discharge from the reactor can be conservatively approximated by ,

removing the Xe'" . i The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been examined following discharge from a reac-tor, in addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 4 on page 20.

The measurements were made on fuel discharged from Yankee Core 5"". The i data in Table 4 on page 20 shows that the agreement between PHOENIX pred-lctions and measured isotopic compositions is good.

i Tho ngrr r n W. tw treer, reactivities,r cr' w.tre wm FH3E""' nj th: resu"J. > '

l @ crmcb. L . cl.raci mw ' we 'a ,ra. .a ... i abia o un pago 2. n, parameters describing each of the 81 experiments are given in Table 6 on page Criticality Analysis of Spent Fuel Racks With Poison Gaps 8

(

l

22. ' These reactivity comparisons again show good agreement between exper- 1 iment and PHOENIX calculations. !

l An uncertainty associated with the burnup-dependent reactivities computed with l PHOENIX is accounted for in the development of the burnup requirements. An '

additional bias of 500 MWD /MTU is included in the results. This is considered a very conservative since comparison between PHOENIX results and the Yankee i Core experimen's and 81 benchmark experiments indicates closer agreement. :

4.1.3 ' REACTIVITY CALCULATIONS The maximum K.ft f'or storage of the spent fuel with poison panel gaps is de- l termined using .the methods described in Sections 2.0 and 3.0. The KENO-IV

~

computer code is used to calculate the storage rack multiplication factor as a function of frer.h fuel enrichment and poison gap size.

The following assumptions were used to develop the nominal case KENO model i for the spent. fuel rack storage of fresh fuel using all storage locations with "

post'ulated axlel gaps in the poison

panel material: )

1. The W 14x14 OFA fuel assembly is at its most reactive point in life, and -l no credit Js taken for any burnable poison in the fuel rods.

2. All fuel rods' contain uranium dioxide at 3.87, 4.07 or 4.27 w/o U*

enrichment over the infinite length of each rod.

3. No credit is taken for any U' or U* in the fuel, nor is any credit taken for ved buildup of fission product poison material.

4. The moderator is pure water at a temperature of G8"F. A conservative value of i.0 gm/cm' is used for the density of water.

5. No credit is teken for any spacer grids or spacer sleevos.

6. The array is infinite in, lateral extent and finite in axial extent which allows neutron leakage in only the axial direction from the array.

7. The _ minimum poison material loading of 0.040 grams B-10 per square cen- ,

timeter is used throughout the array. !

8. The axial gaps in the poison panels are positioned in the axial center of the ective fuel in all poison panels in the spent fuel racks.

To determine the maximum Ken under normal conditions, the worst case me-chanical and material tolerant a= are urari in the model. The same worst case M % moocI ttis5 med e. Sectnen .s.1. n unto bn to 0 nermi' > r.e m m n1 K.o as a function of the pomun poom gup size-

)

Criticality Analysis of Spent Fuel Racks With Poison Gaps 9

_ _ - -- 1

-. +, .

~

Ba' sed on the analysis . described above, the following equation is used to de-

~

velop~ the maximum K.## for the-storage of spent fuel in the Prairie liland Units

,1. and ~2 spent fuel storage racks:

4 K.ve = K or.i + Bm.inoo + B,.gy discussed above, the fuel burnups needed to reduce the spent fuel reactivity 'below the K.,# limit of 0.95 including all uncertainties at a 95/95 probability / confidence level are determined and shown in Figure 6 on page

30. Therefore, with. the appropriate spent fuel burnup, the acceptance criteria for criticality are met for. the storage of spent fuel with poison panel gaps up to four inches at the axial mid-plane and an initial enrichment up to 4.27 w/o U *". i 4.2 ' SOLUBLE BORON CONCENTRATION As the data in Table 7 on page 24 shows, the maximum K et of the spent fuel racks with 4.27 w!o U'" fuel and four inch ax'al gaps in all of the poison panels is greater than .the design lirnit of 0.95. A calculation was performed to de-termine the spent fual pool soluble boron concentration that would be required to reduce this K.tv below the 0.95 limit. PHOENIX was used to determine the L suhNc e.>oren cent.entratran wMr.amwouH #iMat t rM aper: Ti.nl t w nmtivit f u'y a nK . o f 0.031. As a sw!..; # thi! U.. .inon a 6 pent fuel pool soluole
)
boron concentration of 250 ppm was determined to be sufficient to reduce the 6 Criticality Analysis of Spent Fuel Racks With Poison Gaps 10 i
-__a_._-___--__--_-__-- - - _ _ . _ _ _
. I spent fuel rack K.e# below the 0.95 limit including all uncertainties at a 95/95 !
. probability / confidence level. This includes a 10 percent conservatism added to the soluble boron concentration. j 4.3 CHECKERBOARD FUEL LOADING A calculation was also performed to show that the 4.27 w/o U fuel could be stored in the spent fuel racks using three out of four storage locations with four inch axial gaps in all poison panels. The following assumptions were used !
to develop the nominal case KENO model for the spent fuel rack storage of i fuel using three out of four storage lerations with postulated axial gaps in the poison panels: ,
J
1. The W 14x14 OFA fuel assembly is at its most roactive point in life, and. !
no credit is taken for any burnable poison in the fuel rods. ;
1
2. All fuel rods contain uranium dioxide at an enrichment of 4.27 w/o U"' over i the entire length of each rod.
)
3. No credit is taken for any U' or U* !!. the fuel, nor is any credit taken
.for the buildup of fission product poison material. ;
4. The moderator is pure water at a temperature of 68'F. A conservative value of 1.0 gm/cm' is used for the density of water.
5. No credit is taken for any spacer grids or spacer slaeves.
6. Fuel assemblies are loaded into three of every four storage cells in a checkerboard pattern as shown in Figure 7 on page 31.
7. The array is infinite in lateral extent and finite in axial extent which allows 1 neutron leakage in only the axial direction from the array.
8. The minimum poison material loading of 0.040 grams B-10 per square cen-timeter is used throughout.the array.
9. The axial gaps in the poison panels are positioned in tne axial center of the active fuel in all poison panals in the spent fuel racks.
The maximum K.## under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies vithin the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield con-3.n : n :c mt'*n i ~ ryk V x "e sums m o s., tow,: cu u e r-int,rcci &m
. V '; f i f'r. 0 * * ".'c b O'; . I L ; LILT.L,C. . . $f 0:! ^
I" CP .
.. **- OLidu. W 1,i,.ih C.
spacing. For the spent fuel racks this resulted in a reduction of the nominal 0,752" water gaps to their minimum values. Thus, the " worst case" KENO model Criticality Analysis of Spent Fi.iel Racks With Polson Gaps 11
e
']
l , .
1 f.:
of' the spent fuel storage racks contains minimum water gaps of 0.562" with .,
symmetrically placed fuel assemblies. -
Based on the analysis described above, the following equation is used to de- '
velop the maximum K.tr for the Prairie island Units 1 and 2 spent fuel storage ,
racks with fuel storage in three out of four locations: l K.ve = Kwor + B meinoe + B' pari + [ (ks)' wor i + (ks)'m.ince ]
. where: :
Kwori = worst case KENO K.fr that includes material tolerances, and mechanical tolerances which can result in spacings between assemblies less than nominal Bmeinoe = method bias determined from benchmark critical comparisons i B,.ri = bias to account for poison particle self-shielding kswer i = 95/95 uncertainty in the worst case KENO K.et ksm.inod = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K ve = 0.9016 + 0.0083 + 0.0010 + [ ( 0. 004 6 )' + (0.0018)' ] = 0.9158 Since K.ft is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met with fuel enriched to 4.27 w/o, four inch axial gaps in the poison panels and using three out of four storage locations.
l l
l Criticality Analysis of Spent Fuel Racks With Poison Gaps. 12
_1______.--______ _ _ _ i
.a 5.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section describes the analytical techniques and models employed to per-form the criticality analysis for storage of fresh fuel in the Prairie Island Units 1 and 2 fresh fuel racks.
l Since the fresh fuel racks are maintained in a dry condition, the criticality analysis will show that the rack K.et is less than 0.95 for the full density and
0.98 for the low density optimum moderation conditions. The full density and low density moderation scenarios are accident situations in which no credit can I be taken for soluble boron. The criticality method and

ross-section library are the same as those discussed in Section 2 of this report.
The following assumptions were used to develop the nominal case KENO model l for the storage of fresh fuel in the fresh fuel racks under full density and low
density optimum moderation conditions:
i
1. The fuel assembly contains the highest enrichment authorized, is at its most reactive point in life, and no credit is taken for any burnable poison in the fuel rods.
2. All fuel rods contain uranium dioxide at an enrichment of 4.27 w/o U over the entire length of each rod.
3. No credit is taken for any U or U"' in the fuel, nor is any credit taken for the buildup of fission product poison material.
No credit is taken for any spacer grids or spacer sleeves.
I 4.
5.1 FULL DENSITY MODERATION ANALYSIS lg l
l In the nominal case KENO model for the full density moderation analysis, the moderator is pure water at a temperature of 68*F. A conservative value of 1.0 gm/cm' is used for the density of water. The fuel array is infinite in lateral and axial extent which precludes any neutron leakage from the array. The W 14x14 OFA fuel assembly was analyzed. (See Table 2 on page 18 for fuel pa- ,
remeters) ,
I + evmm rr <- .ic;.c no. :s. :;.r.tiivo,m erm i;o o c ... * .. r ' n ,, i . . . .
chanical and n'aterL l thickness tolerances resulting from the manufacturing i
process in addition to asymmetric positioning of fuel assembiles within the ,
Criticality Analysis of Fresh Fusi Racks 13 _
(
s 1 ..
storage cells. Studies of asymmetric positioning of fuel assemblies within the j storage cells has shown that symmetrically placed fuel assemblits yield con- ]
servative results in rack K.vf . The most conservative, or " worst case", KENO I model of the fresh fuel storage racks contains no fuel rack steel blith sym-metrically' placed fuel assemblies.
Based on the analysis described above, the following equation is used to de- l velop the maximum K.1# for the Prairie Island Units 1 and 2 fresh fuel storage
- racks: I K.ve = Kwor.i + Bm.inoo + [ (ks)* wor.i + (ks)'m.inoo ]
1
)
where: l Kwor. = worst case KENO K.fr that includes material tolerances, and mechanical tolerances which can I result in spacings between assemblies less than nominal ;
1 Bm.inoo = method bias determined from benchmark critical comparisons k s.or. = 95/95 uncertainty in the worst case KENO K.fr k sm.inoo = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is: !
K.ev = 0.8736 + 0.0083 + [ ( 0. 00 74 )' + (0.0018)' ] = 0.8895 Since K.e# ie less than 0.95 including uncertainties at a 95/95 probability confi- j dence level, the acceptance criteria for criticality is met. j i
5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS i in the low density optimum moderation analysis, the fuel array is finite in the j radial and axial extent. The W 14x14 STD fuel assembly was analyzed. (See i Table 2 on page 18 for fuel parameters) The STD fuel assembly is the most reactive under low moderator density conditions. The " worst case" model of the fresh fuel storage racks used in the full moderator density analysis is also used in this analysis.
Analysis of the Prairie Island Units 1 and 2 fresh fuel racks under low density moderation conditions shows that the maximum rack reactivity will exceed the j design limit of 0.99 if the full complement of 88 fuel assemblies is assumed l
-to :n. m ci.e A*ter furthe evalWim c' 'h* frub fod rm * 'v:- 9t~- ]
r.c,7 yne ab r o m of 11 ine.J. e c w ai r..- e rw + ,. t re ' d.
1 criticality K.ft limit. .As a result the fresh fuel rack analysis is performed with ,
55 fuel assemblies in a 5 by 11 array.
Criticality Analysis of Fresh Fuel Racks 14
(
i i
9 Use of this 5 by 11 fresh fuel array in the Analysis of the Prairie Island Units I 1 and 2 racks has shown that the maximum rack K.tv under low density mod-eration conditions occurs at 0.075 gm/cm' water density. The KENO calculation of the Prairie Island Units 1 and 2 fresn racks at 0.075 gm/cm' water density resulted in a peak K.et of 0.9634 with a 95 percent probability al.195 percent confidence level uncertainty of 10.0075. Figure 8 on page 32 shows the fresh fuel rack reactivity as a function of the water density. The uncertainty bars shown on the figure are KENO one sigma error bars.
I Based on the analysis described above, the following equation is used to de-velop the maximum K.## for the Prairie Island Units 1 and 2 fresh fuel storage racks under low density optimum moderation conditions:
K.# r = Ko... + Bm.inoe + [ (ks)'e... + (ks)'m.inoo ]
where:
Ko... = base case KENO K.fr that includes nominal +
mechanical and material dimension Bm.inae = method bias determined from benchmark critical comparisons <
k s e... = 95/95 uncertainty in the base case KENO K.t.
ksm.inoo = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K.et = 0.9634 + 0.0083 + [ ( 0. 0 0 7 5 )' + (0.0018)' ] = 0.9794 I Since K.+t is less than 0.98 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met under low moderator density conditions. -
I c I ..
i
'I Criticality Analysis of Fresh Fuel Rocks 15
i'
! 6.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication factor in spent fuel pool and fresh fuel vault shall be less than or equal to 0.95 and 0.98, including all uncertainties, under full density and low density moderator conditions respectively.
The ' analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI 57.2-1983, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality q Safety," NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage"; and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
i Acceptance Critericn For Criticality 16 1
T .e t
a Table 1. Benchmark Critical Experiments (5,6]
I I
Separating Soluble I General Enrichment Matertal Boron ppm keff Desertption w/o U235 Reflector
1. UO2 rod lattice 2.46 water water O O.9857 +/- .0028
2. (l.12 rod lattice 2.46 water water 1037 0.9906 +/- .0018
3. UO2 rod lattice 2.46 water water 764 0.9896 +/- .0015
4. UO2 rod lattice 2.46 water 84C pins O O.9914 +/- .0025
5. 002 rod lattice 2.46 water 84C pins O O.9891 +/- .0026 6, UO2 rod lattice 2.46 water B4C pins O O.9955 +/- .0020 2.46 water B4C pins O O.9889 +/- .0027 I
7. UO2 rod lattice O.9983 +/- .0025
8. UO2 rod lattice 2.46 water B4C pins O
9. UO2 rod lattice 2.46 water water O O.9931 +/- .0028
10. UO2 rod latttee 2.46 water water 143 0.9928 +/- .0025
11. UO2 red lattlee 2.46 water stainless steel 514 0.9967 +/- .0020
12. UO2 rod lattice 2.46 water stainless steel 217 0.9943 +/- .0019 I 13. UO2 roc 14- UO2 rod
15. 002 rod
16. 002 rod lattice lattice lattice lattice 2.46 2.46 2.46 2.46 water water water water borated aluminum borated aluminum borated aluminum borated aluminum 15 92 395 121 0.9892 0.9884 0.9832 0.9848
+/-
+/-
+/-
+/-
.0023
.0023
.0021
.0024
17. UO2 rod lattice 2.46 water borated aluminum 487 0.9895 +/- .0020 18 U02 rod lattice 2.46 water borated aluminum 197 0.9885 +/- .0022
19. UO2 rod lattice 2.46 water borated aluminum 634 0.9921 +/- .0019
20. U02 roc lattice 2.46 water borated aluminum 320 0.9920 +/- .0020
21. UO2 rod lattice 2.46 water borated aluminum 72 0.9939 +/- .0020
22. U metal cyltnders 93.2 bare air O O.9905 +/- .0020
23. U metal cylinders 93.2 bare air O O.9976 +/- .0020
,I 24. U metal
25. U metal cylinders cylteders 93.2 93.2 bare bare air air O
O O.9947 +/- .0025 O.9928 +/- .0019
26. U metal cylinders 93.2 bare air O O.9922 +/- .0026
27. U metal cylinders 93.2 bare air O O.9950 +/- .0027
28. U metal cylinders 93.2 bare plexiglass O O.9941 +/- .0030
29. U metal cylinders 93.2 paraft'n plexiglass O O.9928 +/- .0041
30. U metal cylinders 93.2 bare plextglass O O.9968 +/- .0018
31. U metal cylinders 93.2 paraffin plextglass 0 1.0042 +/- .0019
32. O metal cylinders 93.2 paraffin plexiglass O O.9963 +/- .0030
33. U metal cylinders 93.2 paraffin plextglass O O.9919 +/- .0032 I
'I
'I 17
c.
j- :.
l v ,
f-
' Table. 2. Fuel Parameters Employed in Criticality Analysis l
l Parameter W 14x14 STANDARD W 14x14 OFA i g Number ' of Fuel Rods
!- per, Assembly 179 179 Rod Zirc-4 Clad 0.D.- (inch) 0.422 0.400 i;
Clad Thickness (inch) 0.0243 0.0243 !
Fuel Pel16t 0.0. (Inch) 0 3659 0 3444 Fuel Pellet Density ;
(% of Theoretical) 95 95 l Fuel Pellet Dishing Factor (%) 0.658 0.658 Rod Pi tch . (inch) 0 556 0 556 Number of.Zirc-4 Guide Tubes 16 16 Guide Tube 0.0. (i nch) 0 539 0 526 Guide Tube Thickness (Inch) 0.017 0.0170 -
Number of Instrument Tubes 1 1 i Instrument Tube 0.D. (Inch) 0.4220 0 3990 instrument Tube Thickness (inch) 0.0240 0.0235 V'" Number Density at k.27 w/o (atom / barn-em) .0009980 .0009980
-l
. i I l I
! l I
l L
r .- l Acceptance Criterion For Criticality 18 l'
L l V
l Table 3. Prairie Island Units 1 and 2 Fuel Assembly Minimum Burnup vs inital U'" Enrichment for the Spent Fuel Rack l I
Enrichment Posion Panel Gap Size (w/o) 0" Gap 2" Gap 4" Gap !
! 3.87 0 0 1846 4.07 0 536 3694 4.27 719 1451 4272 Note: Burnup in MWD /MTU l
{
l l
[
)
i i
l i Acceptance Criterion For Criticality 19 l
b
L Table 4. Comparison ' of PHOENIX isotopics Predictions to Yankee Core 5 Measurements
[ l b Quantity (Atom Ratio) % Difference
- U235/U -0.67 -
U236/U -0.28 l U238/U -0.03 -I l
PU239/U + 3.27 PU240/U +3.63
]
PU241/U -7.01 PU242/U -0.20 PU239/U238 +3.24 Mass (PU/U) + 1.41 FISS-PU/ TOT-PU -0.02 i
-l l'
i f
Acceptanco Criterion For Criticality 20 t
I
i i
f <
l
. Table 5. Benchmark Critical Experiments PHOENIX Comparison Description.of . Number of PHOENIX K r# Using Experiment Experiments . Experiments Bucklings
( UO2 l
l. Al clad 14 0.9947 l
SS clad 19 0.9944 Borated Hr0 7 0.9940 Subtotal 40 0.9944 U-Metal -
Al clad 41 1.0012 TOTAL 81 0.9978 i
l 1
l l
! l I
l~
f I
Acceptance Criterion For Criticality 21 1
.~
e i
Table 6. Data for U Metal and UO Critical Experiments (Part 1 Of 2) ;
1 1
t l.
1
(
Fuel Pellet Clad Clad Lattice.
Case . Cell A/O H20/U Denstty. Otameter Matertal 00 .Thtckness Pttch Boron Numoer.)ype U 235 Ratto (G/CC) (CM) Clad (CM) (CM) (CM) PPM 1 Hexa 1.328 3,02 7.53 1.5265 Aluminum t.6916 .07110 2.2050 0.0 2 Hexa 1.328 3.95 7.53. 1.5265 Aluminum 1.6916 .07110 2.3590 0.0 3 Hexa 1.328 4.95 7.53 1.5265 Aluminum 1.6916 .07110 2.5120 oO . O 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.5580 0.0 5 Hexa 1.328 4.89 7.52 .9855 Aluminum 1.150C .07110 1.6520 0.0 6 Hexa 1.328 2.88 10.53 .9728 Aluminem. 1.1506 .07110 1.5580 0.0 7 Hexa 1.328 3.58 10.53 .9728 Aluminum 1.1506 .07110 1.6520 0.0 8 Hexa 1.328 4.83 10.53 .9728 Aluminum 1.1506 .07110 1.8060 0.0 !
9 Souare 2.734 2.18 10.18 .7620 55-304 .8594 .04085 1.0287 0.0 10 Square 2.734 2.92 10.18 .7620 55-304 .8594 .04085 1.1049 0.0 ft Souare 2,734 3.86 10.18 .7620 SS-304 .8594 .04085 1.1938 0.0 12 Souare 2.734 7.02 10.18 7620 55-304 .8594 .04085 1.4554 0.0 13 Souare 2.734 8.49 10.18 .7620 SS-304 .8594 .04085 1.5621 0.0 14 Souare 2.734 10.38 10.18 .7620 $5-304 .8594 .04085 1,6891 0.0 15 Souare 2.734 2.50 10,18 .7620 55-304 .8594 .04085 1.0617 0.0 16 Souare 2.734 4.51 10.18 .7620 55-304 .8594 .04085 1.2522 0.0 17 Souare 3.745 2.50 10.27 .7544 55-304 .8600 .04060 1.0617 0.0 18 Souare 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 0.0 i 19 Souare 3,745. 4.51 10.37 .7544 SS-304 .8600 04060 1.2522 0.0 1 20 Souare 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 456.0 !
21 Souare 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 709.0 1 I
22 Souare 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1260.0 23 Souare 3.745 4.51 10.37 .7544 55-304 .8600 .04060 1.2522 1334.0 24 Square 3.745 4.51 10.37 .7544 SS 304 .8600 .04060 1.2522- 1477.0 25 Souare 4.069 2.55 9.46 1.1278 55-304 1.2090 .04060 1.5113 0.0 26 Souare 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04060 1.5113 3392.0 ;
1.1278 S$~304 .04060 1.4500 0.0 '
27 Souare 4.069 2.14 9.46 1.2090 28 Souare 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 -1.5113 0.0 29 Souare 3.037 2.64 9.28 1.1268 55-304' 1.1701 .07163 1.5550 0.0 3^ Souare 3.037 8.16 9.28 1.1268 5S-304 1.2701 .07163 2.1980 0.0 31 Souare 4.069 2.59 9.45 1.1268 55-304 1.2701 .07163 1.5550 0.0 32 Souare 4.069 3.53 9.45 1.1268 SS-304 1.2701 .07163 1.6840 0.0-33 Souare 4.069 8.02 9.45 1.1268 SS-304 1.2701 .07163 2.1980 0.0 34 Souare 4.069 9.90 9.45 1.1268 SS-304 1.2701 .07163 2.3810 0.0 35 Souare 2.490 2.84 10.24' 1.0297 Aluminum 1.2060 .08130- 1.5113 1677.0 36 Hexa 2.096 2.06 10.38 1,5240 Aluminum 1.6916 .07112 2.1737 0.0 37 Hexa 2.096 3.09 10.38 1.5240 Aluminum 1.6916 .07112 2.4052 0.0 l 38 Hexa 2.096 4.12 10.38 1.5240 Alumtnum 1.6916 .07112 2.6162 0.0 39 Hexa 2.096 6.14 10.38 1.5240 Aluminum 1.6916 .07112 2.9891 0.0 40 Hexa 2.096 8.20 10.38 1.5240 Alumtnuu 1.6916 .07112 3.3255 0.0 41 Hexa 1.307 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 i 2.4054 42 Hess 1.307 1.51 18.90 1.5240 Aluminum 1.6916 .07112 0.0 43 Hexa
307
. 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 I
t i
i 22 1
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 0
- -e .
i 6
Table 6. Data' for U Metal and UO Critical Experiments (P8rt 2 "Of. 2) 1' l
l 1 ,
l l
1 I. Fuel Pellet Clad Clad. Latttee .l l Case Cell A/O H20/U Density Otameter Matertal '00 .
Thickness Pitch Boron ;
Numoer Type U-235 Ratto (G/CC) (CM) Clad (CM) .(OM) (CM) PPM. l 44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2I9896 0.0
' 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 LO.O 46 Hexa 1.160 1.01 18.90 1.5240 Aluminum 1.6916- .07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916 .07112 '2.4054' O.O 48 Hexa 1.160_ 2.02 18.90 1.5240 Aluminum -1.6916 .07112 2.6162 0.0 49 Hexa 1.160' 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0
' 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 51 Hexa 1.040 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 52 Hexa 1.040- 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 1 53 Hexa 1.040 2.02 18.90 1.5240 Aluminum 1.6916 .07112. 2.6162 0.0 1 54 Hexa 1.040 3.01 18.90 1.5240 Aluminum 1,6916 .07112 -2.9896 0.0 l 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 l 56 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 - 0.0 3 57 Hexa 1.307 1.52 18.90 .9830 Aluminum 1 1506 .07112 1.5926, 0.0 j 58 Hexa. ' 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07312 1.7247 0.0 l 59 Hexa t.307 3.02 18.90 .9830 Aluminum .1 1506 .07112 1.9609 0.0 ;
60 Hexa 1.307. 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 l 61 Hexa 1.160 1.52 18.90 .9830 ' Aluminum 1.1506' .07112 1.5926- 0.0 l 62 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506- .07112 1.72d7 - 0.0 63 Hexa 1.160 3.02 18.90 .9830 Aluminum. 1.1506 .07112 1'9609 0.0 64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.17(2 0.0 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1,1506- .07112- 1.4412 0.0-66 - Hexa 1.160 1.52 18.90 .9830 Aluminum, 1.1506 .07112 1.5926 0.0 !
67 . Hexa 1.160. 2.02 18.90 .9830 Aluminum 1.1506' .07112 1.7247- 0.0 l 68 Hexa 1.160 3.02 18.90' .9830 A lumi num 1.1506 .07112 1 9609 0.0 i 69 Hexa 1.160 4.02 18.90 .9830 Aluminum. 1.1506 .07112 2.1742 0.0 1 70 Hexa 1.040 1.33 18.90- 19.050 Aluminum 2.0574 .07620 2.8687 0.0 1 71 Hexa 1.040 1.58 18.90' 19.050 -Aluminum 2.0574 .07620 3.0086 0.0 72 Hexa. 1.040 1.83 18.90 19.050 Aluminum 2.0574 .07620 3.1425 0.0 73 Hexa 1;O40 2.33' 18.90 19.050 Aluminum 2;O574 .07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574 .05620 3.6284 0.0 75 Hexa 1.040 3.83 18.90^ 19.050 Aluminum 2.0574 .07620 4.0566 0.0 76' Hexa 1,310 .2.02 18.88 1.5240 Aluminum 1.6916 .07112 .2.6160- 0.0 77 Hexa 1.310 3.01 18.88 1.5240 Aluminum 1.6916. .07112 2.9900' .O.0 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 ,07112' 2.6160 'O.O
. 79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900- 0.0
' 80 Hexe 1.312 2.03 18.88 .9830 Aluminum 1.1506 .07112 1.7250 0.0 81 Hexa 1.312 3,02 18.88 .9830 Aluminum 1.1506 .07112 1.9610 0.0
.r
, .(
t :
b l i 23 ;
.---_-__m.-___a ____-..__--__.m ____m. _ , _ _ _ _ _ _ _ , , _ _
l Table 7. Prairie Island Units 1 and 2 Spent Fuel Rack Maximum Reactivity vs Initial U'" Enrichment and Poison Panel Gap Size I
Enrichment Posion Panel Gap Size (w/o) 0" Gap 2" Gap 4" Gap 3.87 0.9361 0.9407 0.9646 4.07 0.9477 0.9509 0.9774 4.27 0.9542 0.9615 0.9810 ,
. s .-.
aus e
lI l
l I
I I -
I Acceptance Criterion For Criticality 24 __
. :o 4
1 i
i .
l' I 8.20" -
P P 0.752"
+
i f i
e 8.27" , ,
, , l
,l ,
~
A -
{
O.090" Inner Casiry Cell Center to Center (9.5") '
O.125" Absorber ;
0.024" Outer Casirq l
I I
i Figure 1. Prairie Island Units 1 and 2 Spent Fuel Storage Cell Nominal Dimen-sions 25
o
- 4
/
1 .
0 _
1
'8 1
1
\ \ ,' _
e .
e . '-.
s .
- ~
1 I/
.
s _
I ' a ,
l n ,
D a ,
E : ,
D ,
I e , "8 E . /- 2 i
5 _-
3 s 4 l o
o 3 P -
' e . 4 -
' a .
' 8 .
' e .
' a .
l a
n a ' e .
_ C ' a .
' s .
r e -
f -
s =
n a . :
r . "O -
T '
. 4a
. =- .
e s ,
n n . -_
s a .
" 1 s ,
3 e - l
- s s ,
'8 o s a , o __
1 P 1
- a i I n s .
/ I , .
I
=
t u %l ,=! ,t.t4. ;- L* i
=_* & C3.E ^ ua n ntv.3* mC. 3mn m2C*
8 . . 8 2ec . p 7".1 n
wO
[l l lli1II l l,li! !
4 .
l i' 4 : ;r
'. l l l 1
i
. 1--f- i
. 'l i 3
'2 ! !
( . .l ' /- h e v - r r
a* r3 R
i l
" =..
,;,, 4, ,.
t c .
l'
'l i
' I
' 3
me r
,7) 4 c c . A os. t.: a as l
3 - s.vb e,s ),,N-4.7% !.
I % -
S. 5 ,
If[]
t ts e ..
t ,
/ ,
. . .a.u'u...m,
, i j
J ,
-l /
l
/
1 : ,
E $ $' $ T[I 7
[~ '
h .']. f $ i f l
. it
c I j
,';:?
. amma susi * *c '
m 24ec 6 s (.1 '1 == a s t. I r me 3, s 2 , c . .; *
. l s es.s 's .
j , s G".6C.M " i e .4,'
i N
Ie ? h*?
~ !
f* .
I .;:6 l
r l
. l [
th aex
. i
== =::
l
.i 1 1l l l} l i l
L _= ,,
e n' I-9 i I f
rie tsa"9ste*
. s .J , - - ., . . - - -
3, n 5,,, u -
a c c *lON
3 i ~~i
~
, w e o :-
C3uu TG MC Fsu O l ;.,
1 1
i i
Figure 3. Prairie island Units .1 and 2 Fresh Fuel Storage Rack Axial Layout
,~/
k
- ' - i __m_ _ _ _ _ _ . _ . _ _ _ . . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _
a t.
\ _- i-l 1-h I
l 1
-1
]
i I
(
e
.5 0'a c" ,
4 To A TbRICCiG R AIL S
. I J
__, _. r l - -
y, .
g,.,--
7 '- -
. -i J q;
g ..
t,U
,, 1+
e r i i, i
o r ' i i
-l, {
ii 0 0 0 0 0 O, O O O 0 0l j! ". i_
e,
.: i n n n n n n n m m n nl -,. t i
( km ---,,, - l i: 1 i
. UU UUU UUU w LJ Uj f er n n n n n n n m n n Djiu. i
'3 l l *** .
'~uttunas.. 't'l l E I ,,,
LJ L. l UU UW W w O U l p.e4 ., ' i.
jU .
n n n n m m n.m n n mi?. i j
l lMtt .
1:r.: asis. l ', a};.2-a g '
' eg. g-t U U'U UU lu U U UUU f,' I :'
.O O O O O b O O 0 ,0 0:jA 1. ,
,CJo sucessAs .ir.c- ee _ l <.s -j
! *. t. .. ._ is' 1* --, I; I ,i ;,
. p. . . . . .
g4 ,_i.
l i
Figure 4. Prairie Island Units 1 and 2 Fresh Fuel Rack Radial Layout 28 1
i 1
4 '- .
T j
1.10 , , . ,
i i l i I I i I-
! ____J____..____L____ ____.1____..____I_____
l I .I I
.I I I l- l-1 I I i 1.05 . . . ,
I I l 'l i I 1 1 I I I I I I I I I I I I I I I 1. l 1.00 . l l 1 1 I I I I I l
6 I I I l b
x ' '
l NO BORON l i i l i I I j l
.950 i l i I I I I I I I I i
____q____..-~~ r---- --~~7---- l-----
1 I I I I I I I
.900 I I i 1 I I I l' I I I I ;
l _ _ _ _ , _ _ _ _ .
l-----
i
- r l 1800
- BORON I i i i i
I I I i t
850 I 1.5 2.0 . 2.5 3.0 3.5 Hg0/UO.YOLUnlERATIO 2
Figure 5. Maximum Spent Fuel Rack K e# versus H20/UO: Volume Ratio For Heavy Load Drop Accident i
1 20~
(
e 4- =y
.'s i
4500 i , i i i !
i i i i I l r--' X --
r--~~- 7--- --l--- -- i---
4000 I I I I -- / I i
l-i 1
i I
i I
7i I I4INCHGAPl l
___7_-..- 3--- - - , - - - -- r- --
T- I 1 I I I l-3500 i i i i i i l i I l' ' - - - - - -
l I T--'
3000 1 I I / I I I l i
,i i i i i i i i I i I a
l i I I I 0:: l 1 I I l l 1
o2500 I I I I !
co i .I i i l I y I i f O I I I I I I ck:
<c I I I I I I 2:2000 I I o i I / I I e i __2___ _ _ _i_ _ _ ___t-- _- t--.. i C I I i i l i i I i l l 2 INCH GAPI 1500 l l l l l. l/
-- L--..- l--- --l--- -- L-- -- L--.. ,c 'l - - -
I I I I 1 I ,
i i I t i - t 1000 <
l i I I 4
.- L--..- A--- --l--- -- L-- ,4'L-l0INCHGAPl' 500 l l l l/
l / - 1 I I I /I I I
__ L--..- l--- --l--- ,l G-- -- --- - l---
I I I I i I e , , / , / i i b.7 3.8 3.9 4.0 4.1 4.2 4.3 U235 ENRICHWENT(W/0)
L l
1 l
l: Agut e L. Freirie island Units 1 and 2 Fuel Assembly Minimum clurnup vs. Initial U28* Enrichment and Poison Gap Size for Storage in the Spent Fuel l Racks l
l l
AcceDtance Criterion For Criticality 30 ;
1
\
(
a s
l ll l
1 I
l
\
l.
l l
l l
Figure 7. Prairie Island Units 1 and 2 Region 1 Three of Four Fuel Assembly j Loading Schematic 31
o- _ - _ _ _ _ _ - - _ .
I, ,
l I
I =
i i
I I
i
__J_.._ L---
---d -
-f--- - - - ',- - - ~
I I 1
I I
.95 .
1
{
t 1 '
l l I
____!___..__ L-_, ---;------I --
i I i 1 I I i i l
.95 .
+
i { , i w I i u i j i i w I l
__.._ _ _ "~ - ---I ----
- - - f - -- ---i
< t I c: 1 I
c' n i i i
< l
{I .94
I I i l I
i i l ' ~ ' - - ~
__ _..__----- ---,---- ---f--- ---i i I i i !
93 I I I i l l I
---i----
---h--- ---i l I
---~
i i i
.92 .06 .08 . 10 .12
.02 '04 WATERDENSITY(GW/CC)
I I Figure 8. Sensitivity of Ken to Water Density in the Prairie island Units 1 and 2 Fresh Fuel Storage Racks
=
1 a
.s i
i l BIBLIOGRAPHY l
1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications , April 14, 1978.
2. W. E. Ford Ill, CSRL-V: Processed ENDFIB-V 227-Neutron-Group and l Pointwiss Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSD/TM-160, June 1982.
3. N. M. Greene, AMPX: A Modular Code System for Generating Coupled ;
Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706, March '
1976.
4. L. M. Petrie and N. F. Cross, KENO IV--An Improved Monte Carlo Criticality Program, ORNL-4938, November 1975.
5. M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor fuel, B AW-1484-7, July 1979.
6. J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
7. Elllott, A. J., Wong, K., Licensing Report For Prairie Island Nuclear Generating Plant Units 1 and 2 Spent Fuel Cask Drop Evaluation, QUAD-1-83-017, October 1984.
8. A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Bol/Ing Water Reactors, WCAP-10106, June 1982.
9. Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of the 1-Lattice Code W/MS, Journal of British Nuclear Energy Society, 5, pp.
564-584, 1966.
I h 10. England, T. R., CINDER -
A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
11. Melehan, J. B., Yankee Core Evaluation Program Final Report, WC A P-3017-M? ",, ., a,o F. "*
Bibliography 33
_ _ _ _ _ _ L

ML20248J888

Text

1.0 INTRODUCTION

Navigation menu

Search