ML20141M120

Criticality Analysis of Maine Yankee Spent Fuel Storage Racks to Allow Up to 3.95 W/O U-235 Fuel

ML20141M120
Person / Time
Site:	Maine Yankee
Issue date:	12/02/1991
From:	Adli D, Cacciapouti R, Napolitano D YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20141M115	List: ML20141M112 ML20141M118 ML20141M120
References
YAEC-1828, NUDOCS 9204010192
Download: ML20141M120 (27)

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Criticality Analysis of Maine Yankee's Spent Fuel Storage Racks to Allow 885 Up to 3.95 w/o U ruel lievember 1991 by D, G. Adli D. G. 11apolitano a

Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachus3tts 01740 9204010192 920324 l

PDR ADOCK 05000309 P

PDR

Prepared byI f/AA)dO

(( ~

1/SAI D. G. Adli, 11uclear Engineer (Date)

Reactor Physics Group Huclear Engineering Department

$7n s14 n*'N Q *L<>

ll,%Tbl}

D.

G. Napolitano, (ey' lor Engineer

/ Wate) neactor Physics Group liuclear Engineering Department D/'[/

Approved by:

N R/A. Cacciap4cti, Manager

.t e )

(ractor Phydts Group Nuclear Engineering Department l

L,' $l 4W h

E. C. Slifer, P4 rector 1Date)

Nuclear Enginehing Departmftnt 11

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DISCLAIMER OF PESPONSIBILITY This document was prepared by Yankee Atomic Electric Conpany

(" Yankee").

The use of information contained in this document by anyone other than Yankee, or the Organization f or which this docur.ent was prepared under contract, is not authorized and, with resreet to any unnuthorized

,qtt, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

4 i

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ABSTRACT This report presents a revision to the criticality analysis 10r Maine Yankee's spent fuel storage racks.

The original analysis justified the placement of fuel with enrictaent up 4.13 w/o O'" in the phase I racku, 3.72 w/o v'" in the phase II spent. a 1 racks and 5.$0 w/o O'" in the new fuel vault.

This analysis justifies the place',tont of fuel with initial enrichment up to 3.95 w/o U"' anywhere in the spent f uel pool. The present analysis is perforrned with KENO-Va Monte Carlo and CASMO-3 integral transport theory.

Av P

s, PJ;g OF CONTENTS Pace DISCIJ2MER OF RESPONSIPTLITY iii ABSTFM-iv t AELE Ob' CONTEN' v

LJ97 OF TAB.:..

v1 LIT

• OF FIGURES vif 1.0 10TROKCTION 1

1.1 RF ctic

. Basis 1

+

1.2

't %

,y Methoos 1

41 O SPEN ~ F-F

.Y ANALYSIS 3

2.1 Opent Fual chanical Design 3

3 0,1 Fuel Assemb.

.',s 2.3 KENO-V4 Mode);..,

4 2.4 CASMO-3 Modolling 5

'2. 'i - Y.r c?. Enrieme r-5

'.4 Sensitivity Analysis 5

7 Maximum Ftsah Fuel Enrichment 5

i c' Accident Situations 6

3.0 CONCLUSION

S 19

4.0 REFERENCES

20 v

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LIST OF TABLES i '

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ILuoter.

Titig Pace 2.1 The Phase II Canister Type 3 Dimensions and Tolerances 7

s l

2,2 Nominal Fuel Assembly Design Specifications 8

a 2.3 Sensitivity Analysis Resulto 9

2.4 Spent Fuel Rack-K,u,5 vs. Enrichment 10 4

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LIST-OF FIGUREfg Number ~

.T111.1 Page 2.1 The Maine' Yankee Spent Fuel Pool Arrangement 11 2.2 Storage Rack Module, Radial 12

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2.3 Storage Rant Module, Axial 13 2.4 Storage Rack Unit Cell For Criticality Analysis 14 1$

2.5 Storage Rack XEt:0-Va Model 26 Storage Kr.ck CASHO-3 Model 16 2.7 Storage Rack K.rr vs. Enrichment, CASMO-3 ar1 KENO-Va Comparison.

17 2.8 Storage R4ck L,ini vs. Initial Enrichment 18.

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l.0 JJEPODUCTION 1.1 Peculations and Desion Basis The applicable codes, standards and regulations of for spent fuel and new fuel storsge include the followingtcriticality safety

• General Design Criterion 62 - Prevention of Criticality in Fuel Storage and liandling

• NUREG-0B00, USNRC Standard Review Plan, Storage and Section 9.1.1, New Fuel Storage.Section 9.1.2, Spent Fuel

• ANSI /ANS-57.2-1983, Facilities At Nuclear Power Plants, Section 6.4.2. Design Requirements f

• ANSI /ANS-57.3-1983, Facilities at LWR Plants, Section 6.2.4. Design Requirements for New Fue3 Storag These regulations and guides require that for spent fuel racks the trv.imum calculated K.,,, including margin for uncertainty in calculational method and mechanical tolerances, be less than or equal tO 0 95 with a 95%

probability at a 95% confidence level.

In order to asstre the true reactivity will always be less than th calculated reactivity, the following cor.servbtive assumptions are made in e

calculating the criticality safety limitu for the spent fuel racks:

+ pure, unborated water at 68 'F is used in all calculations, a 2t infinite array with no radial or axial leakage is moCelled e

neutron absorption from spacer grids is heglected, i.e. replaced by

=

water, 1.2 The YAFC Criticality Safety Methods g

Yankee Atomic Electric Compary (YAEC) combination of criticality cafety methods based on:has developed and validated a KENO-Va Monte Carlo"'28, CASMo-3 LWR lattice integral transport *, PDQ-7 fine mesh diffusion theory"' and SIMULATE-3 nodal burnup credit analysis"'"

This pertaits cM tivality analysis by several independent methods and allcwa the flexibility to handla various LWR fuel types, fuel storage arrays and criticality safety assump';, ions.

described in more detail below. These methods and their applications are In the NITAWL-S/ KENO-Va methodology, the NICAWL-S code prepares a 1

im W A Em I

. - ~.- - -

working nuclide library and performs resonance self-shielding for anU. In this analysis, the 123 group data is used in all KINO-Va calculations. The working nuclide library along with case specific compositions and rack geometry data are input to KENO-Va. KENO-Va performs a multi group, Monte Carlo eigenvalue calculation.

The results from KENO-Ve analysis are (gg vs. generation, fluxes and reae. tion rates. Since Monte Carlo is stochastic in nature, results will always have some uncertainty.

In this analysis, KINO-Va is used to verify the CASMO-3 spent fuel rack criticality results, i

I CASMO-3 is an integral transport lattice code with a hierarchy of i

energy condensation and spatial detail leading to a seven-group, transmission probability model of the fuel rack unit cell, COXY.

The 40 micro-group ruclear data is used in all CASMO-3 calculations.

CASMO-3 is flexible encugh to Mndle up to a 19x19 feel assembly array with storage canister regions, poison sheets, and water gaps.

CASMO-3 can perform transport theory burnup crodit analysis. Hot full power lattice depletions can be executed, and cold zero power restarts in rack geometry can be

{

- performed. CASMO-3 can produce few-group cross sections for PDQ fine mesh diffu ?lon theory ana?ysis, Also, CASMO-3 can produce two-group homogenized cross section for noda) burnup credit criticality analysis on futel storage arrays.

CASMO-3 is used to study: rack K,gs vs. fresh fuel enrichment and unit cell sensitivity to mechanical perturbations. Since the results of CASMO-3 i

calculations are deterministic, K.re vs, enrichment is monotonic and smooth.

Also, a reactivity change, AK, from mechanical perturbatf.on is not overwhelmed by stochastic uncertainty like the AK from Monte Carlo would be.

In previous analysis, KENO-Va was used as the refereace calculation in the determination of maximum fresh fuel enrichment.

In this r'evisud analysis, CASMO-3 is used as the reference calculation. This deterministic approach eliminaues th3 stochastic uncertainty inherent in KENO Monte Carlo calculations.

The net result is a slfghtly higher allowable fresh fuel enrichment in the spent fuel racks.

The use of KENO-Va and CASMO-3 for fuel storye criticality analysis has been validated by comparison to 21 B&W fuel storage critical

. experiments.A" The methodology bias and uncertainty determined from this validation will be used in the calculation of K,eg at a 95/95 probability /

confidence level.

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2.0 _ SPENT 5'UEL RACK CRITICALITY ANALYSTA 2.1 Spent Fuel Back., Mechanical Des g The Maine Yankee spent fuel pool arrangement is shown in Figure 2.1.

In early. 1985, the spent fuel pool was reracked with 10.25" center-to-center, high

density, BORAL poisoned spent fuel tacks manufactured by GCA/ PAR.

The reracking was done in two phases.

Modules from the two phases occupy rcughly half of the spent fuel pool apiece.

There'are a total of 26 high density fuel storage rack modules with a total of 1476 storage locations.

Criticality analysis in support of the licensing of these racks justified the placement of fresh fuel with enrichments up-to 4.13 w/o U'" in the Phase I racks and the placement of fresh fuel with enrichments up to 3.72 w/o U235 in the Phase II racks '"-

This revised analysis will-justify the placetent of up to 3.95 w/o U 28' anywhere in the spent fuel pool.

The spent fuel modules comprise canisters which are welded together at the top to each other and at the bottom to a frame forming rectangular arrays or modules, A 7 X 9 module is shown in Figure 2.2 and a canister axial profile ic chown in Figure 2.3.

The criticality analysis will concentrate on unit analysis of the canister at the active fuel length, see v

Figure 2.4.

Since it is the most limiting, canister type 3 of the Phase II modules used in determining the criticality safety limits. The canister I

type 3 diiaensions and tolerances are given in Table 2.1.

This canister design will be used in this revision to the criticality analysis for the Maine Yankee racks.

2.2 Fuel Assembiv Desion The Maine Yankee fuel assembly is a 14x14 array of fuel pins with 5 large -(2x2 fuel pini guide tube positions. This fuel ascembly design has been manufactured by both C-E and ANF.

The fuel assembly design is summarized is Table 2.2.

Due to its slightly higher fuel loading the C-E design is slightly more reactive than the ANF design; therefore, the C-E design _will be used in all criticality calculations.

Enrichments for Maine Yankee fuel assemblies have varied in the past from about 2.0 w/o U235 t6 the 3.7 w/o U"5 in Cycle 12, the present cycle.

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Since"it is anticipated that sligntly higher fuel assembly enrichments may be needed for fine tuning cycle lengths or that split enrichments maybe needed for fine tuning radial peaking, spent fuel rack criticality is 3

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reanalyzed recognizing these changes in fuel management.

Normally, some burnable poison (BP) pins are present in the assemblies of a fresh batch of fuel.

BP pins are solid A1,0 -B.C pellets 3

clad with.ziraaloy.

These pins displace fuel pin positions and are an integral part of the fuel assembly.

Credit is allowed for such bps in spent fuel ract criticality analysis.

However, Maine Yankee loading pattern optimization has reduced BP requirements to the point where credit for bps is not viable.

Thus, the criticality analysis will assume all unshimmed assemblies are placed in the spent iuol racks.

The two major assembly tolerances impacting spent fuel rack criticality are the f.2el (UO,) density deviation and the enrichment (w/o) deviatien. The tolerance on fuel dt;nsity is 10.07 g/cc baseo ?n the C-E spec on pellet theoretical density and 1 0.08 g/cc based on the ANT spec on theoretical density. The more conservative 10.08 g/cc will be wed in the criticaltty analysis.

The maximum (100%) variation on enrichment is by DOE contractual agreement 10.013 w/o U'".

However, the variation in fuel essembly enrichment for a fresh batch of fuel can be as high as 10.05 w/o U'".

This deviation will

used in the criticality analysis.

2.3 KENO-Va Model1ina NITAWL-S is used to process che raw 123 group data into a working library and perform resonance calculations for U'"

as a function of enrichment. The wora.ing library along with KENO-Va input data are used to create a model of the rt:k canister with a fresh fuel assembly.

In this analysis, the model is used to study the criticality of the storage rack canisters as a function of fresh fuel enrichment. 'the KENO-Va model of the spent fuel rack unit cell is an explicit pin by p10 model (see Figure 2. C.).

Reflecting boundary; conditions are applied at the sides, top atid bottom simulating a two dimensional infinite array. A 123 group working library is created by NITAWL for criticality analysis of the racks versus fresh fuel enrichment and for fresh fuel alternatir.g rows analysis.

Separate resonance calculations are performed for U'"

at each enrichment with t

Dancoff factors calculated by Sauer's method. "

The Ket calculat' ion for i

each enrichment is an average of three independent calculations with:

l' different starting seeds for the random number generator, a cosine starting j~

distribution and 300 neutrons per generation for 103 generation, skipping the first three.

Thus, each case is a result of 90,000 histories.

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?.4 CASMO-3 Modelling The two ditaensional CASMO-3 model of the Maine Yankeo spent fuel racks is rhown in Figure 2.6.

This model is based on the seven group transmission probability routine, COXY.

Reflecti.ng boundary cor.ditions simulating an infinite two dimensional array are implicit j n COXY. In this fuel storage rack model, the pin cells representing i.he fuel assembli are appropriately homogenized square ceLis surrounded by an explicit inner water gap, steel canister wall, BORAL sheets, steel outer wrapper, and flux trap water gap.

The model makes use of half diagonal fuel storage symmetry.

2.5

,K,,,, r, v s. Enrichment e

Calculations of rack K.e, vs. enrienment were perforrad using both KENO-Va and CASMO-3.

Enrichment was varied from 3.00 to 5.00 s/o U " in 2

increments of 0.25 w/o U ". The results, without uncertainties added, are 3

given in Figure 2.7.

Agreement between KENO-Va and CASMO-3 is withjn a standard deviation.

Also, this validates the reactivity performance of CASMO-3 at enrichments up to 5.0 w/o U n, 2

.6 Rack Sensitivity Analysis Perturbations of the mechanical and compositional tolerances are parformed using CASMO-3.

A nominal configuration of the spent fuel rack an canister with fuel of 3.5 w/o v was perturbed t one tolerance at t time.

Perturbations considered are: center-to-center spacing, inner canister dimension, outer wrapper dimension, BORAL pl;te width, BORAL plate

~

thickness, BORAL core thickness, BORAL core B,C loading, asserbly U density and assembly UO2 en21chment. The sensitivity analysis results are shown in Table 2.3, and the final root sum of squares is 0.00636.

This a

is defined to be the 95/95 mechanical uncertainty.

Sensitivity of rack K.tr to fuel assembly radial enrichment zoning was performed with CASMO-3 using ennchment splits of 4.3 and 3.4 w/o U ". The 2

lower enrichment pins were placed next to guide tubes and the corners of U n, g

2 the assembly.

The average aseembly enrichment was 3.93 w/o reditction of 0.0018 6K was calculated from the uniform case whe?. zoned assemblies are place in the racks.

Similar reductions in reactivity are seen for in-core lattice calculations. Thus, uniform assembly enrichments calculations can be considered bounding.

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Maximum. Fresh Fuel Fnrichment A

determination of r.aximum fresh fuel enrichment without administrative controls ir r..ade by adding all uncertaintiss to tt;e nominal K rr values vs. enrichment and then solving for the enrichment at which K.tr e

= 0.95, the NRC limit.

K. t r is calculated at 95/95 probability / confidence level by the following equation:

Kn,,3

• K

+6K,, *[EaK ) 3 + (% 13 (1) m where:

(

Km = K.r r of the nominal configuration, AK

= calculational bias, o

95/95 calculational uncertainty, and AK,

=

AK, 95/95 mechanical uncertainty, a

For CASMD-3 based fuel storage criticality enlculations, AK

-0.00251 a

0.00853."'

From Table

2. 4, AK, and AK, 0.00636.

Thus, the total

=

=

uncertainty, AK, appl'.ed to the nominal M.et values is:

3 AK, =. 0 0 51 +y' (. 0 0 8 5 3 ) 2 + (. 0 0 6 3 6 ) 2., 0 0 g 13 (2) m K,3,,3 vs. enrichment is given in Table 2.4 and is plotted in Figure i~

2 2.8, Interpolating between the Kn,,3 values for 3.75 and 4.0 w/o U " gives U"

for K 2

a maximum fresh fuel enrichment of 3.95 w/o 0.95 with

=

uncertaintie;.

2.8 Accident Situations Accident situations include: an assembly on top of the racks, and an assembly next to the sides of the racks.

Credit is allowed for the

,g presence of soluble boron (1720 ppm) in accident situations. This refueling c

concentration of soluble boron provides a 30% redaction in reactivity over the unborated sit'ntion and more than adequately suppresses reactivity effects free. the above accident situations.

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s Table 2.1 The Phase II Canister Type 3 Dimensions and Tolerances Item M nsion (In. )

Canister Center-to-Center 10.25 1 0.10 Spacing Inner Canister Dimension 8.75 0.10 Inner Canister Thickness 0.105 02ter Wrapper Dimension 9.747 1 0.10 Outer Wrapper Thickness 0.036

~

Flux Trap Thickness 0.503 Loral Plate widtb 8.00 1 0.125 Boral Thickness 0.177 0.012 Boral Core Thickness 0.0921 1 0.0027 Boral Core w/o B C 35 t 1.6 4

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Tabis 2.2 l

Nominal Fuel Assembly Design Specifications Type E Types L & M Types N

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P

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Fuel Vendor C-E ANF C-E J

Fuel Assembly Creerall Length 150.718 156.71; 156.718 x

Spacer Grid Size 8.115 8.115 8.115 No. Zirc Grids 0

0 8

No Ine;'nel Grids 1

0 1

No. Bimetallic Grids 0

9 0

Spacei' Material 18.2164 14.2239 18.2364 (g/cc axial)

Fuel Rod Active Length 136.7 136.7 136.7 Clad CD 0.440 0.440 0.440 Clad ID 0.384 0.378 0.384 Clad Material Zr-4 Zr-4 Zr 4 Fellet OD 0.3765 0.370 0.3765 I

Pin Pitch 0.580 0.580 0.580

!=

Stack Height 10.0458 10.1994 10.0453 Density (g/cc)

Guide Tube Tube CD 1.115 1.115 1.115 Tube ID 1.035 1.035 1.035 Tubo Material Zr-4 Zr-4 Zr-4 3

c

• Ala dimensions are inches unless otherwise specified E

M K

8 I

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Table 2.3 Sensitivity Analysis Retalt.s 6K Center-to-Center Spacing li).00420 Inner Canister Dimension 10.00273 Outer Wrapper Dimension v0.00006 BORAL Plate Width 10.00143 DORAL Plate Thickness 10.00120 BORAL Core Thickness 10.00094 BORAL B C inding s0 00137 Assembly UO2 Density 20.00084 Assembly UO F.nrichment 10,00290

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Root Sum of Squares 0.00636 9

4 Table 2.4 Spent W el Rack K,,f,, vs. Enrichment w/o U"5 L.,,,,_.,

K., u,3,___,

3.50

0. 91'f 9 6 0.92609 3.75 0.93163 t).93976 4.00 0.94412 0.95225 4.25 0.95543 0.96356 4.50 0.96594 0.97407

~~

4.75 0.97550 0.98363 5.00 0.98445 0.99258

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PHASE I RACK PHASE II RACK MODULES MODULES i

rigure 2.1 The Maine Yankee Spent Fuel Pool Arrangement 11 l

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0.0 FLOOR Figure 2.3 Storage Rack Module, Axial 13 l

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Figure 2.4 Storage Rack Unit Call For Criticality Analysis 14

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Storage Rack KIDIO-Va Model i

15

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U L FWP. Parameters 4 N

Half riux HFsT=14, s= 1.4632 cm, Cww 22.2250 cm, STW=.3745 en, CAW.GAN 4632 cm Trap Gap CRD Parameters:

G AW !+--

CRT=.23393 cm, CR5 1.1902, AsL=20.32 cm r$7 Fateseters:

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a. 4285 cm, ba 6388 en

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Figure 2.6 Storage Rack CASMO-3 Model 16

e

+

1.0

/

/

o KENO-Va

</

0'95 -

O CASMO3

/

c 2

h 0.9 -

0 B's 0.8 -

35 3.75 4.0 4IS 4.5 4.75 5.0 9/01/91 Ennchment (wlo U-2a5) w Figure 2.7 8torage Rack N,, vs. Enrichment., CASMO-3 and luCNO-Va Comparison 17

+

1 r

'e 1.0 0.98 -

7 y 0.96 -

f 0.95 NRC LIMIT

/

A 9

0.94 -

W 0.92 -

~

3.95 w/o U-235 Fresh Fuel Limit 0.9 3.0 3.5 4.0 4.5 5.0 9/01/9.1 Enrichment (w/o U-235)

Figure 2.8 Storage Rack K,sf., vs. Initial Enrichment 18

_.. _.. _ _ _ ~.. _ _ _ _ _. _ _...

I o

4

3.0 CONCLUSION

S Conservative criticality analysis of the Maine Yankee spent fuel racks shows that they can accomodate f uel assemblies with nomial enrichment up to 3. 95 w/o U'".

The analysis is bounding for radial enrichment zoned assemblies up to this average' enrichment.

9 0

1 19

.e J

4.0 PEFERENCES 1.

ORNL/NUREG/CSD-2/V2, "NITAWL-S, SCALE System Module for Performing Resonance Shielding and Working Library Production", R. M. Westf all, L. M. Petrie,-N. M. Greene and J. L. Lucius, October 1981.

2.

ORNL/NUREG/CSD-2/Vl/R2, " KENO-Va, An Improved Mcnte Carlo Criticality Program with Supergrouping", L. M. Petrie and N. F. Landers, December 1984.

3.

STUDSVIK/NFA-86/7, "CASMO-3, A Fuel Assembly Burnup Program", User' Manual, M. Edenius, A. Ahlin and B. Forssen, November 1986.

4.

EPRI/ARMP Documentation, PDO-7/ HARMONY User's Manual",

B.

M.

Rothleder, March 31, 1983.

5.

-STUDSVIK/SOA-88/02,

" TABLES-3P, Library Preparation Code for SIMULATE-3P,"

D.

M.

Ver Planck, K.

S.

Smith and J.

A.

Umbarger, February 1988.

6.

STUDSV1X/SOA-88/01, " SIMULATE 3P Advanced Three Dimensional Two Group Reactor Analysis Code,"

D.

M.

Ver Planck, K.

S.

Smith hnd J.

A.

Umbarger, February 1988.

7.

B&W-1484-7, " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel", N. M. Baldwin, G. S. Hoovler, R. L.

Eng a:.d P. G. Welfare, July 1979.

8.

YAEC-1622, " Validation of the YAEC Criticality Safety Methodology",

D. G. Napolitano and F. L. Carpenito, January 1988.

9.

YAEC-1637, " Criticality Analysis of Maine Yankee's Spent Fuel Pool and New Fuel Vault," D. G. Napolitano and L. G. Adli, February 1988, 10.

"EPRI-CELL Code Description," ARMP Package, Part II, Chapter 5, Pages i

5-14, 5-15, 5-76 and 5-77, October 1975.

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