ML19246A906

Proposed Change to Unit 2 Tech Specs Re Storage of 4.1% Enriched Fuel,Prepared by Nuclear Svc Corp

ML19246A906
Person / Time
Site:	Calvert Cliffs
Issue date:	06/11/1979
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML19246A901	List: ML19246A900 ML19246A902 ML19246A905 ML19246A906
References
QUAD-1-79-372, NUDOCS 7907090225
Download: ML19246A906 (22)

Text

.

.L QUAD-1-79-372 Attachment 3 BALTIMORE GAS & ELECTRIC CO.

LICEllSIfiG REPORT CALVERT CLIFFS liUCLEAR POWER PLAllT UllIT 2 STORAGE OF 4.1% EliRICHED FUEL Prepared By:

000i.303 33E0!!C33 COEPC2EIlCO CAMPB ELL, CALIFORNI A L_

Prepared By / '

/ # <7 Reviewed By / b N d O*; (

M. 'Re f e'r " ( K. Wong

]

Project Engineer 7 /go /n J. D/ Gil/ crest Revision fio. Date fiSC Job Number Issued By Date 0 6 'll /79 BGE-0103 okdt/:0 6/11/79 I

30 23 3 7907090 3 R:T 7 J'

QUAD-1-79-372 OUCt. EAR SERVICES CORPORATICi1 kil.lR DREX CD A AO A ATION TABLE OF CONTENTS Page

1.0 INTRODUCTION

1 2.0 NUCLEAR ANALYSIS 3

3.0 CONCLUSION

18 4 ,;"

274

QUAD-1-79-372 NUCLEAR SERVICES CORPORATIOD

. . v. , , , ~ e .

p/JADREX w CO A AQ A ATION

1.0 INTRODUCTION

The spent fuel storace racks presently in use in the spent fuel pool of Calvert Cliffs Unit 2 were designed by Nuclear Services Corporation for storage of spent fuel with a maximum enrichment of 3.7 weight percent U-235, which corresponds to 43.8 g/cm of U-235. It now anticipated that storage of fuel with 4.1 weight percent (48.5 g/cm) U-235 will be necessary in the future.

To ensure the safety of storage of 4.1% enriched fuel NSC has performed a new criticality analysis for the Unit 2 spent fuel racks. This analysis and report follow the same format as the previous analysis which was described in recort number NSC-BGE-0101-R001, Revision 1 The analysis for 4.1% enriched fuel is described in the following section.

.l. r.  ;> :c5

QUAD-1-79-372 DUCT. ERR SERVICES CORPORR7100 A.LIADREX CO A AC A ATION TABLE 1 CALVERT CLIFFS, UtlIT tl0. 2 K RESULTS eff Term Value Method k

0

.9199 Calculated for cell at 212 F ak)

.005 Estimated from similar design ak 2

.0082 Calculatea from sensitivity analysis ak 3

.0025 Calculated for 0.050" displacement ak 4

.0048 Critical experiment calculations ak g .0077 Critical experiments results, 95%

confidence level ak 6

.0039 Sensitivity analysis, 0.01" tolerance ak .0025 Estimated for 0.02% enrichment increase 7

2 eff b k0+ak)+ak2 + ak3 + ak4 + (ak5 + ak 6 + ak 7 ) 1/2

= 0.9398 r.

f

. QUAL-1-79-372 DUCT. ERR SERVICES CORPORATION

& DIw:5.ON OF

%IZjAD9EX CO A AC A ATION 2.0 NUCLEAR ANALYSIS The 12.5 x 13.0 inch spacing of the fuel assemblies is sufficient to maintain k eff below 0.95 for all normal and abnormal fuel storage conditions. The analysis results are described below.

2.1 Sumary of Results The value of k eff is determined as follows:

2

+ Ak +Ak keff = k0+ak)+ak2 + ak3+1 4 + (ak5

, 6 7 ) 1/2 where k0 = nominal calculated k (2-0 diffusion theory) ak) = transport correction ak2 = assembiy location effect (fuel in tube) ak3 = rack spacing tolerance effect (tube in rack) ak4 = methods bias ak5 = uncertainty in methods bias (95% confidence level) ak6 = channel thickness tolerance effect ak7 = fuel fabrication tolerance effect The value of k eff is maintained below 0.95 for all normal and abnormal storage conditions in accordance with Standard Review Plan 9.1.2. Results are sumarized in Table 1.

n' f]

QUAD-1-79-372 OUCi. EAR SERVICES CORPOR9710D A O'ViS10 4 C F

%7dADREX CO AP O A A*10 N 2.2 Method of Calculation Verification of k eff is obtained using a two dimensional (X-Y) diffusion theory computer code calculation. The calculational model covers both finite and infinite arrays of stored fuel assemblies. Neutron cross sections are developed for a four group energy range using the CHEETAH code which is an adaptation of the LEOPARD-CINDER code. The output of CHEETAH is used for a diffusion theory calculation using the CITATION code to establish the k eff. In order to verify the accuracy of the CHEETAH-CITATION diffusion calculation, a comparison was made with various critical experiments as shown in Table 2. The data in the table defines the actual fuel used in the critical experiments. This data was placed into a CHEETAH-CITATION calculation with the results shown in the last column for comparison. These results may be compared with a measured k eff f 1.000. Deviations vary from a few tenths of one percent to a little over one percent as indicated by the table. The standard deviation for the uranium oxide cases is 1.0047 1 0.005 for CHEETAH and 1.0050 1 0.005 for CITATION.

A transport-theory correction between 0 and 0.7% has been used for similar fuel storage configurations. (Refer to Wisconsin Electric Power Company, Publication March 28, 1975, Docket Nos. 50-256 and 5-301). The transport correction is not independent of other uncertainties, as it is in effect included in ak for critical lattices. We conservatively 4

estimate it to be 0.5%.

2. 3 Data and Assumotions Used in the Calculations Table 3 sumarizes the fuel design values which were used in the calculations.

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TABLE 2 (Continued) g RESULTS FRON OIEETAH CALCULATIONS

\

i 0F SELECil0 UO,/Puo, CRiflCAL EJtPER!l1ENIS DEllCifMRK EVALUATI0fl C AL CUL AT E D SQUARE l%JLIIPLI-D0 2 CRillCAL CATION Puo, ll,0:ll* Pu FUEL PitLEI CLAD CLAD L A) TICE PITCll MOLE FAciOR CASE REFER- ENRICllflENT V0ttAtE DEllSI TY DINi[IER CLAD OD THICKNESS in BUCt(ING HUf0ER INCE WT.% RATIO g/cm 3

IN. HATE RI AL IN. IN. IN. FRACil0N B ppm m' k lirc-WCAP- t aloy-4 0.391 0.02325 0.792 0 0 159.3 1.01994 B-3 1305-54' 6.00 11.78 10.99 0.3374 0.56 0 0 122.1 1.02551 P-4 4.50 0 56 0 337 112.3 1.03223 B-5 4.50 0.735 0 0 159.6 1.01708 8-6 9.75 0.52 0 0 103.8 1.01397 B-7 3.51 1

1.04 0 0 128.4 1.01704 7 B-0 22.28 V V V V V V B-9 V 4.48 0.56 0 0 120.9 1.02046

• E. G. Taylor, "Saxton Plutonium Pro 9 ram". Critical Experimcats for ttie Saxton Partfas Plutonium Core, WCAP-3385-54, Decemtier 1965.

Wt. Percent Pu-239 Pu-240, Pu-241, Pu-242 is 90.49, 0.57, 0.09 and 0.041.

MG. k,gg = 1.022010.E8 (SID. DOI ATION) e s'

. QUAD-1-79-372 DUCLERR SERVICES CCRPORATIOD

% Y/J A D R E X CO R A C A ATIO N TABLE 3 FUEL DESIGN PARAMETERS Fuel Assembly Array 14 x 14 Number of Fuel Rods 176 Number of Water Rods 5 Rod Pitch 0.580" Fuel Pellet 0.D. 0.3805" Clad 0.D. 0.440" Clad I.D. . 0.388" Clad Thickness 0.026" Clad Muterial Zircaloy 4 Pellet Density, % T.D. 95.0 Maximum Bundle Enrichment, Wt. % U-235 4.1 Nominal Active Fuel Length 137.18" r, $h

QUAD-1-79-3',.

DUCLSAR SERYlCSS CORPORAT100 WUADREX CO APC A ATION The following assumptions were used in the calculations:

a. Pool is filled with fuel at the highest enrichment stored in an infinite array.

b. The water in the fuel storage pool is clean and unborated.

c. The pool temperature is 212 F. (100 C)

d. No creait is taken for the fuel assembly support structure.

Howe.ar, the neutron absorption of the 3/16" (4.7625 mm) thick stainless stell tubes in which the fuel assemblies are supported while in the storage racks is considered in the calculation.

e. Neutron absorption in fuel assembly grids is excluded.

f. Axial neutron leakage is included.

g. No credit is taken for U-234 and U-236 in the fuel .

. The above assumptions are considered as a conservative base for the calculations.

2.4 Nuclear Design Criteria The following criteria were used to evaluate the results of the nuclear calculations:

Normal Storace and Handling Maximum keff = 0.95 Included as r.ormal conditions are fuel reactivities up to the maximu, pool water temperature 1 212 F and eccentric positioning of the fuel in the proper location.

.n. '3

Q E 79-372 DUCLCAR SERVICES CORPORATION MUADREX CO APO A ATION Abnormal Storage and Handling Maximum kgff = 0.95

" Abnormal" refers to a single operator error. Included as abnormal conditions are dropping of a fuel assembly, dropping a fuel handling grapple, movement of the refueling bridge before the fuel is out of the rack or the fuel grapple is released, and placement of a single fuel assembly outside the storage racks.

2.5 Results of Calculations The results of the k eff calculations for various storage and handling conditions are listed as follows:

Condition eff

1. Normal positioning in the spent fuel storage array 0.9199 (Figure 1)

2. Eccentric positioning in the spent fuel storage array (clumped in groups of two or four)

a. Fuel at side of channel, x-direction 0.9239 (Figure 2)

b. Fuel at side of channel, y-direction 0.9232 (Figere 3)

c. Fuel at cornar of channel 0.9281 (figure 4, Case 1)

3. Eccentric positioning in the spent fuel storage array with the channel offset 1/8" (3.175 rmi) in both X and Y directions 0.9336 (Figure 4, Case ?)

4. One extra fuel assembly at side of rack 0.9402 (Figure 5, Case 2) e,

,4

DUCLEAR SERVICES CORPOR97100

& Divalic'u 0F Q(JADREX CO R AQ A ATION The value of k eff is not expected to change significantly under abnormal storage situations. Consideration was given to a fuel assembly lying on top of ? rack. Due to the small axial neutron leakage reactivity of 0.2% Ak, and the separation of the fuel assembly from active fuel in the rack, the increase in reactivity is <0.002.

The drop of a fuel assembly onto the rack and abnormal 'uel handling machine loads of 1650 pounds vertical and 825 pounds horizontal are also considered. The fuel rack is designed to prevent any plastic deformation in the fuel region for these loads. Therefore, the reactivity effect of these abnorinal loading conditions is insignificant.

Sensitivity studies were performed to evaluate the influence on k eff of fuel tube spacing, pool t< sture, and fuel tube thickness. The results of t' ise anal- are summarized below.

Spacing Effects Table 4 lists the effects of fuel spacing.

t s%BLE4 K VERSUS SPACING, 212 F eff K

Cell Pitch, Inches eff 9.25 x 9.75 1.238 10.00 x 10.50 1.133 10.75 x 11.25 1.048 11.50 x 12.00 0.9822 12.25 x 12.75 0.9331 12.50 x 13.00 0.9199 12.75 x 13.25 0.9084 22 7

QUAD-1-79-372 DUCl.SRR SSRVICES CORPORATION O/_lADREX CO APC A ATION TABLE 4 (continued) 13.50 x 14.00 C.8839 15.50 x 16.00 0.8436 17.50 x 18.00 0.8304 Temoerature Effects The reference rack unit cell was analyzed at several temperatures in order to determine the temperature effect on k The eff.

fuel pellet, clad, moderator, and rack pitch were assumed to expand normally. The results are shown in Table 5. The change .s less than one (1) percent. The numbers quoted in this report are the values for the highest temperature (212 F).

TABLE 5 Infinite lattice eff Temoerature , C 0.9138 4 0.9186 60 0.9199 100 Effect of Variation in Steel Channel Thickness For the reference 13" x 12.5" (330.2 x 317.5 mm) pitch rack, the steel channels are 3/16" thick. In order to estimate hforthesteelchannels,wehavethefollowingdata:

Steel Thickness K t, in. (mm) eff 0.1875 (4.7625) 0.9199 0.l'T75 (4.28625) 0.9273 ao k 2 - k:

g= 0.01875 = 0.4627 per inch (0.0182 per mm) k)xk2 f reduced thickness n r

HO B C D 2

I A

HO 2

l i

I .

B HO 2

C SS-304 E HO 2

NOTE: NOT TO SCALE 212*F INCH 4.06573 0.37798 0.18776 1.62728 1.87753 (CM) (10.32694) (0.96007) (0.47692) (4.13329) (4.76918)

-(, ~; r> V

> i Figure 1. Unit Cell Model for Norr.al Fuel Positif ni. g a

31.79118cm 4.13329 'd 20.65392cm 1.92014 g 4.13329 y

~

c-; ' g ~ cr:F

~

cm "

5 1.47528 cm 5 l.

l

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ll ll l l l l l 1 l I i 16.53311 cm l l l l l l l l l l HO 2

55-304 y HO 2 ,

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NOTE: NOT TO SCALE d

Figure 2. Eccentric Positioning , ;8 X Direction Offset 10.32696cm _

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m. v. .

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bi

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NOTE: NOT TO SCALE CASE 1 - AS SHOWN. FUEL. OFFSET DIAGONALLY.

CASE 2 - AS SHOWN, EXCEPT: A = 4.13329cm .3175cm = 3.81579 B = 4.13329cm + .3175cm = 4.45079 C = 4.76918cm .3175cm = 4.45168 D = 4.76918cm + .3175cm = 5.08658 BOTH FUEL AND CHANNEL CFFSET DIAG 0tnLLY n

Figure 4.

' '  ?'9 D Eccentric Positioning Diagcnal Offset

99.19908cm ,

(13.0lS25X3) ili 4.76918cm 94.42990cm 15.88474, 30.48cm cm -

l l5.57033cm I II i EA O G

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N p C O

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~

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$ 09

$ N0 2

d) =0 y ,

HO 2

8 -

9 G =0 CASE 1 - BASE CASE CASE 2 - EXTRA ASSEMBLIES AT EACH I LJ OF THE TWO LONG SIDES I I MODEL I I SHOWN l l AB0VE NOTE: NOT TO SCALE Figure 5. Extra Fuel Assembly at the

r. g Side of the Rack 29 f QUAD-1-79-372 DUC!. ERR SERVICES CORFORATIOD WCADREX CO APO A ATION

3.0 CONCLUSION

The criticality analysis shows that with all uncertair..ies included the calculated k eff f r storage of iuel with 4.1% U-235 is still well below the limit of 0.95. Therefore, storage of this fuel in the Calvert Cliffs Unit 2 spent fuel racks presents no safety problem.

Atcachment b

. . . , . Ptge 1 cf 2 l DESIGN FEATURES

(_ -

1 VOLUME I

5.4.2 The total water and steam volume of the reactor c clant system is 10,614 + 460 cubic feet at a nominal T avg of 532*F.

5.5 METEOR 0LCGICAL TCWER LOCATION 5.5.1 The meteorological tower shall be located as shewn on Figure 5.1-1.

l 5.6 FUEL STORAGE CRITICALITY - SPENT %~L 5.6.1 The spent fuel 3 rage racks are designed and shall be maintained with a minimum 12.5 x to incn center-to-center cistance between fuel as emblies placed in the storage racks to ensure a k,,, equivalent te

< 0.95 with the storage pool filled with unborated water. The k,,,of

< 0.95 includes the c:nservative allcwances for uncertainties desd 2ed in Section 9.7.2 of the FSAR. La additicn, fuel in the storage pact snall have a U-22.5 loading of < h6.5 crams of U-235 per axial centimeter o'

(*.

~

fuel assembli. .

CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 inch center-to-center cistance between new fuel assemblies such that K ,,will not exceed 0.98 when fuel having a maximum enrichment of 4.0 weig,nt' percent U-235 is in place and acueous foam moderation is assumec. The K ,, of < 0.9S includes the conservative allewanceforuncertaintiesd!iiribecinSection9.7.2oftheFSAR.

DRAINAGE 5.6.3 The scent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of tne pool below elevation 63 feet.

CAPACITY 5.6.4 The fuel st: rage ecol is designed and shall be maintained with a ccmbined *crage ca:acity, for both Units 1 and 2, limitec to no mera than 1 .c fuel assemblies. \

g

~

Cdd n

, 5.7 COMPONENT CYCLIC CR TRANSIEN LIMITS L

5.7.1 The comecnents identified in Table 5.7-1 are desicnec and shall be maintained within the cyclic or transient limits of Table 5.7-1.

CALVERT CLIFF 3 UN'T 2 5-5 Amendment No.12

Attachment h Page 2 of 2

. v. ,

i DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is gg of 532 F.

10,514 + 460 cubic feet at a nominal T 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be loca* ed as shown c7 Figure 5.1-1.

l 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL t

S.S.1 The spent fuel storage racks are designed ar.d shall be maintained with a minimum o.7s inch center-to-center distance between fuel l assemblies plac~ec iii the storage racks to ensure a k,,, equivalent to

< 0.95 with the storace pool filled with unborated water. The k,,, o f

< 0.95 includes the conservative allowances for uncertainties deidribed In Section 9.7.2 of the FSAR. In addition, fuel in the storage pool shall have a 0-235 loading of < L6.6 grams of U-235 per axial centimeter l

'of fuel assembly. -

CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained with e nominal 18 inch center-to-center distance between new fuel assemblies such that k g will not exceed 0.98 when fuel having a maximum enrichment of 4.0 weigh percent U-235 is in place and aqueous foam moceration is assumed. The k,,, of < 0.98 includes the conservative allowance for uncertainties desdribed in Section 9.7.2 of the FSAR.

DhAINAGl 5.6.3 The spent fuel storage pool is designeo and shall be maintaiaed to prevent inadvertent draining of tne pool below elevation 63 feet.

CAPACITV 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than JG5f fuei assembiies .

SG$ _,

02 5.7 COMPONENT CYCLIC GR TRANSIENT LIMI Q v

5.7.1 The components i.dentified in Table 5.7-1 are designed and shall be maintained within che cyclic or transient limits of Table 5.7-1.

CALVERT CLIFFS - UNIT 1 5- 5 Amendment No. 27

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