ML20246D671

Criticality Analysis of Byron/Braidwood Fresh Fuel Racks

ML20246D671
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	06/30/1989
From:	Adams W, Fecteau M, Schmidt B COMMONWEALTH EDISON CO.
To:
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ML20246D661	List: ML20246D657 ML20246D671 ML20246D687
References
NUDOCS 8908280221
Download: ML20246D671 (20)

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CRITICALITY ANALYSIS OF THE SYRON/BRAIDWOOD FRESH FUEL RACKS r.

8 June 1989 M. W. Fecteau W. M. Adams B. W. Schmidt W. A. Bordogna

'$ I i ' 8908280221 890015 %

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.. r TABLE-OF CONTENTS 1.0 Introduction ............................................... 1 1.1 Design Description .................................1 1.2 Design Criteria. ...................................1 2.0 Criticality Analytical Method .................................. 2 3.0 Criticality Analysis of Fresh Fuel Racks ......................... 3 3.1 . Full Density Moderation Analysis ....................... 3 3.2 Low Density Optimum Moderation Analysis ..................S 3.3 Postulated accidents ................................ 6 4.0 Acceptance Criterion For Criticality ............................ 7 5.0 Conclusion ........................... ................... 8 Bibliography .................................................. 16 i

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. Table of Contents i {

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__________._.__..m__ _ _ _ _ . . _ . _ . . _ _ _ _ _ _ _ _ _ _ __ _ . . . . _ . _ _ _ . _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ __ _ _ _ . _ . . . _ _ _ _ _ _ _ _ _ _ _

.4'

. LIST OF TABLES Table 1. Benchmark Critical Experiments [5.6) ................. 9

- Table 2. Fuel Parameters Employed in Criticality Analysis ......... 10 Table 3. Fuel Parameters Employed in Criticality Analysis ......... 11 Table 4. Summary of Maximum K.n., including Blas and 95/95 Uncertainties 12 i

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List of Tables ii I i

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

+4' LIST OF ILLUSTRATIONS Figure 1. Byron /Braidwood Fresh Fuel Storage Cell Nominal Dimensions 13 Figure ~ 2. Byron /Braidwood Fresh Fuel Storage Array Layout ........ 14 Figure 3. Sensitivity of K.n to Water Density in the Byron /Braidwood Fresh Fuel Storage Racks ............................ 15 1,.

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List of Illustrations ili

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1.0 INTRODUCTION

The Byron /Braidwood fresh fuel rack design described herein employs an exist-ing array of racks, which will be analyzed at a higher enrichment. This analysis

. will reanalyze the fresh fuel array for. criticality to show that 5.0 w/o fuel can be stored in the rack in all storage locations. The fresh fuel rack design is an unpoisoned array, previously analyzed for storage of Westinghouse 17x17 fuel assemblies with enrichments up to 4.0 w/o U*" utilizing every storage to-cation.

. The' fresh fuel rack reanalysis is based on maintaining K.n 's 0.95 for storage of Westinghouse 17x17 OFA, STD, VANTAGE 5 or VANTAGE SH fuel at 5.0 w/o V" with an uncertainty of 0.05 w/o under full water density and optimum moderation conditions.

1.1 DESIGN DESCRIPTION The fresh ~ fuel rack storage cell design is depicted schematically in Figure 1 on page 13. The fresh fuel rack layout as used in the optimum moderation q analysis is shown in Figure 2 on page 14. i 1

1.2. DESIGN CRITERIA Criticality of fuel assemblies in a fuel storage rack is prevented by the design j of the rack which limits fuel assembly interaction. This is done by fixing the l minimum separation between 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 factor (Ken) of the fuel assembly array will be less than 0.95 as re' commended in ANSI 57.3-1983 and in Reference 1.

Introduction 1 L

l

, g 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 the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions j which include strong neutron absorbers, large water gaps and low moderator

' densities.

\

The design method which insures the criticality safety of fuel assemblies in the ,

fresh 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 ENDFIB-V data. The NITAWL program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The, Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM program which is a one-dimensional So 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

• iPlexiglas and air) that demonstrate the wide range of applicability of the method. Table 1 on page 9 summarizes these experiments.

The average K.n 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 0.0018 Ak.

- Criticality Analytical Method 2 L____-__-______------ _ - _ ------- - A

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I 3.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS l j

Since the fresh fuel racks are maintained in a dry condition, the criticality analysis will show that the rack Keve is less than 0.95 for the full density and low density optimum moderation conditions. The full density and low density optimum moderation scenarios are accident situations in which no credit can be taken for soluble boron.

'The following assumptions were used to develop the KENO model for the storage of fresh fuel in the fresh fuel racks under full density and low density optimum moderation conditions:

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. All fuel pellets are modelled at 96 percent theoretical density without dishing or chamfers to bound the maximum fuel assembly uranium loading.

2. All fuel rods contain uranium dioxide at an enrichment of 5.0 w/o (nominal) and 5.05 w/o (" worst case") U'" over the entire length of each rod.

3. No credit is taken for any U'" or U"' in the fuel.

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

5. The fuel rack center-to-center spacing is assumed to be 21 inches.

3.1 FULL DENSITY MODERATION ANALYSIS in the 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 all directions which precludes any neutron leakage from the fuel array.

Calculations for fuel racks have shown that the Westinghouse 17x17 OFA fuel assemblies yield a larger Keet (approximately 1 - 2 %Ak/k) than does the

.. Westinghouse 17x17 STD fuel assembly when both fuel assemblies have the same U*" enrichment. Westinghouse 17x17 OFA fuel assemblies are also more reactive than similarly enriched VANTAGE 5 or VANTAGE SH fuel assemblies since the VANGATE 5 and SH designs incorporate 6-inch natural enrichment axial blankets at the ends of the 12 foot fuel rods. Thus only the Westinghouse Criticality Analysis of Fresh Fuel Racks 3

j fr '

17x17. OFA' fuel assembly-was analyzed under full' density moderator conditions

-(see Table 2 on page 10 and Table 3 on page 11 for fuel parameters).

The KENO calculation for the nominal. case resulted in a. K.o of 0.9046 with a 95 percent probability /95 percent confidence level uncertainty of 0.0070.

The maximum K.n under normal conditions arises from consideration of me '

chanical and material thickness tolerances resulting from the manufacturing

- process . and from : consideration of off-center positioning of fuel . assemblies =i within the individual storage cells. For ease of fuel handling,. the storage cells

. are built with a slightly larger inner dimension than the assembly outer dimen-sion as shown in Figure 1 on page 13. .Because of this extra space, a stored assembly can sometimes be. offset from center within a cell resulting in a group of assemblies clustered together. . However, studies of off-center positioning of fuel assemblies within the storage cells have shown' that centered fuel as-semblies yield conservative results in rack K n.-

Due to the relatively large cell ' spacing, the small tolerances on the cell I.D.

and center-to-center spacing are not considered since they will have an insig-nificant effect on the fuel rack reactivity. However, the sheet metal thickness is reduced by 0.010 inches. Furthermore, fuel enrichment is assumed to be 5.05 w/o U'" to conservatively account for enrichment variability. Thus, the most conservative, .or " worst case" KENO model of the fresh fuel storage racks contains the minimum sheet metal thickness with symmetrically placed fuel assemblies at 5.05 w/o U**. ,

Based on the analysis described above. the following equation is used to de-velop the maximum K.n for the Byron /Braidwood fresh fuel storage racks:

K.n= K rst + Bmein.e + (((ks)* r. +

(ks)*m.in.e ]

where:

K..r.: =

worst case KENO K n with full density water Bm.rn e =

method bias determined from benchmark critical comparisons k s .or.. = 95/95 uncertainty in the worst case KENO K.n ksm.inoe = 95/95 uncertainty in the method bias

)

Substituting calculated values in the order listed above, the result is:

K.n = 0.9135 + 0.0083 + /[(0.0094)* + (0.0018)* ] = 0.9314

)

- Since K.n is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met.

Criticality Analysis of Fresh Fuel Racks 4

____.____.___m_____.__.__._m.________ _____.__.'___..__________.__.-___ .___._-__.__.__.._______..m.__ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ . _ . _ __ _ . . _ _ . . . _ _ _ _ _ .___.___._____2

3.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS For the low density optimum moderation analysis. the fuel array is finite in all directions. The " worst case" cell configuration from the full density analysis is used to model the actual fresh fuel rack array, individual cells are arranged into arrays of 2 x 22, with the cells on 21" centers. The entire fresh fuel rack i is then composed of three such 2 x 22 arrays, arranged parallel to each other in the lengthwise direction. The nominal distance between the centers of as-semblies in adjacent arrays is 71.87 inches (see Figure 2 on page 14 for rack layout). Fuel assemblies were assumed in every one of the 132 cell locations.

The model includes a concrete floor and concrete walls on all sides. Under low water density conditions, the presence of concrete is conservative because neutrons are reflecte.d back into the fuel array more efficiently than they would be with just low density water. The area above the fresh fuel rack is filled with water at the optimum moderation density.

The Westinghouse 17x17 STD fuel assembly was anaQzed under low density optimum moderation conditions. Calculations have shown that the STD fuel assembly is more reactive by approximately 0.5 to 1.5 %Ak/k under low mod-erator density conditions. This is because the STD fuel assembly contains o higher Uranium loading than the OFA assembly, and when optimum moderation conditions are present, higher loadings result in higher reactivity. The STD fuel assembly analysis also bounds VANTAGE 5 type fuel assemblies since the calculations ignore grids and since the VANTAGE 5 type fuel assemblies contain less enriched uranium than the STD fuel assembly because of axial blankets (see Table 2 on page 10 and Table 3 on page 11 for fuel parameters).

Analysis of the Byron /Braidwood racks has shown that the maximum rack K.,v under low density moderation conditions occurs at 0.045 gm/cm' water density.

The calculation of the Byron /Braidwood fresh rack reactivity at 0.045 gm/cm' water density resulted in a peak Lee of 0.7738 with a 95 percent probability and l 95 percent confidence level uncertainty of 10.0084. Figure 3 on page 15 shows the fresh fuel rack reactivity as a function of the water density.

Based on the analysis described above, the following equation is used to de-velop the maximum K.e, for the ByrordBraidwood fresh fuel storage racks under i low density optimum moderation conditions:

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Criticality Analysis of Fresh Fuel Racks 5 j l

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K.ev= Ke... + Sm.inoe + / [(k s)'.... + (ks)'m.inoe ]

where: l l

Kc... - = maximum KENO K.ft with optimum moderation Ben.inoe =

method bias determined from benchmark critical comparisons kse.... = 95/95 uncertainty in the maximum KENO K.vf ksm.ince = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:

K.ev = 0.7738 + 0.0083 + /[(0.0084)* + (0.0018)* ] = 0.7907 Since K.ve is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met.

1 3.3 POSTULATED ACCIDENTS Under normal conditions, the fresh fuel racks are maintained in a dry environ-ment. The introduction of water into the fresh fuel rack area is the worst case accident scenario. The full density and low density optimum moderation cases analyzed above are bounding accident situations which result in the most con-servative fuel rack Keve.

Other accidents can be postulated which would cause some reactivity increase (i.e., dropping a fuel assembly between the rack and wall or on top of the rack).

For these other accident conditions, the double contingency principle of ANSI N 16.1-1975 is applied. This states that one is not required to assume two un-likely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these other accident conditions, the absence of a moderator in the fresh fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.

Generic studies have shown that the maximum reactivity increase for postulated accidents, such as those mentioned above, will be less than 10 %Ak/k. Since '

the normal, dry fresh fuel rack reactivity is less than 0.70, the maximum rack K.ve for postulated accidents will be less than 0.95.

t E 'j Criticality Analysis of Fresh Fuel Racks 6

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I 4.0 ACCEPTANCE CRITERION FOR CRITICALITY l

l The neutron multiplication factor in the fresh fuel' racks shall be less than or equel to 0.95, including all uncertainties, under all conditions.

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 N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality 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."

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Acceptance Criterion For Criticality 7 Lc

5.0 CONCLUSION

For the Byron /Braidwood fresh fuel storage racks, the acceptance criterion for.

criticality is met for the storage of Westinghouse 17x17 OFA, STD, VANTAGE 5 or VANTAGE SH fuel at 5.0 w/o U*". with an uncertainty of 0.05 w/o.

All locations of the fresh fuel rack array can be utilized to store any number of Westinghouse 17x17 OFA, STD, VANTAGE 5 or VANTAGE SH fuel assemblies at any enrichment up to a maximum of 5.05 w/o U'". There are no restrictions on fuel assembly positioning or on mixing of different fuel assembly. types within the fresh fuel storage rack array.

A summary of calculated K.m for the Byron /Braidwood fresh fuel storage racks is given in Table 4 on page 12.

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H General Enrichment separating Soluble Description w/o U235 Reflector Material Boron ppm. Keff

1. UO2 rod lattice 2.46 water water- 0 0.9857 +/- 0028

2. UO2 rod lattice 2.46 water water 1037 0.9906 +/- 0018 3.'UO2 rod 1sttice 2.46 water water 764 0.9896 +/- .0015 4 UO2 cod lattice 2.46 water P C pins O O.9914 +/ .0025

5. UO2 rod lattice 2.46 water -74C pins O O.9891 +/- .0026

6. UO2 rod lattice 2.46' water d4C pins O O.9955 +/- .0020

7. UO2 rod lattice 2.46 water 84C pine ' O O.9889 +/- .0027

8. UO2 rod lattice 2.46 water B4C pins O 0,9983 +/- .0025

9. UO2 rod lattice 2.46 water water O O.9931 +/- 0028

10. UO2 rod lattice 2.46 water water 143 0,9928 +/- 0025

11. UO2 rod lattice 2.46 water statniess steel 514 0,9967 +/- .0020

12. UO2 rod lattice' 2.46 water stainless steet 217 0.9943 +/- .0019-13..U02 rod lattice 2.46 water borated aluminum 15 0.9892 +/- .0023=

14.- U02 rod lat tice 2.46 water borated aluminum 92 0.9884 */- 0023

15. UO2 rod lattice 2.46 water borated aluminum 395- 0.9832 +/- 0021

16. UO2 rod lattice 2.46 water borated aluminum 121 0.9848 +/- .0024

17. UO2 rod lattice 2.46 water borated aluminum 487 0.9895 +/- .0020

18. UO2 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. UO2 rod lattice 2.46 water borated aluminum 320 0.9920 +/- 0020

21. 002 rod' lattice' 2.46 water borated aluminum 72 0.9939 +/- 0020

22. U metal cylinders 93.2 bare air O O.9905 +/- 0020

23. U metal cylinders 93.2 bare air O O.9976 +/- .0020

24. U metal cylinders 93.2 bare air O O.9947 +/- .0025

25. U metal cylinders 93.2 bare air O 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 cyttnders 93.2 bare plexiglass O O.9941 +/- .0030

29. U metal cylinders 93.2 paraffin plexiglass O O.9928 +/- .0041 20..U metal cylinders 93.2 bare plexigless O O.9968 +/- .0018

31. U metal cylinders 93.2 paraffin plextglass 0 1.0042 */- .0019

32. U metal cyttnders 93.2 paraffin plexiglass O O.9963 +/- .0030

33. U metal cylinders 93.2 paraffin plexiglass O O.9919 +/- .0032 1

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Table 2.- Fuel Parameters Employed in Criticality Analysis -

Parameter.- W 17x17 OFA' W 1'7x17 STANDARD Number L of = Fuel Rods L

.per Assembly 264 264-

. . Rod . Z'i re-4 Clad 0.D. (inch) 0 360 0 374 Clad . Thickness (inch) 1 0.0225 0.0225

+

Fuel Pellet 0.D. (inch) 0 3088 0 3225.

Fuel; Pellet Density

(% . of Theoretical) 96 96; Fuel Pellet' Dishing Factor ~ 0.0 0.0 Rod Pitch (inch) 0.496 0.496 1

l Number . of- Zire-4 Guide Tubes 24 24' Guide Tube 0.0. (inch) 0.474 C.482 l'

Guide Tube Thickness (inch) 0.016 0.016 Number of Instrument Tubes 1 1 Instrument Tube 0.D. (inch) 0.474 0.482 Instrument Tube Thickness (i nch) 0.016 0.016 l

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4 Table 1. Fuel Parameters Eraployed in Criticality Analysis Parameter W 17x17 W 17x17 VANTAGES VANTAGE SH Number of Fuel Rods per Assembly - 264 264 Rod Zirc-4 Clad 0.D. (inch) 0 360 0 374 Clad Thickness (inch) 0.0225 0.0225 Fuel Pellet 0.D. (Inch) 0 3088 0 3225 Fuel Pellet Density

(% of Theoretical) 96 96 Fuel Pellet Dishing Factor 0.0 0.0 Rod Pi tch -(inch) 0.496 0.496 Number of Zire-4' Guide Tubes 24 24 Guide Tube 0.0. (inch) 0.474 0.474 Guide Tube Thickness (inch) 0.016 0.016 Number of instrument Tubes 1 1 Instrument Tube 0.D. (Inch) 0.474 0.k74 Instrument Tube Thickness (inch) 0.016 0.016 Enriched Zone Length (inch) 132.0 132.0 Axial Blanket Lengths -

(Inch) 6.0 6.0 Axial Blanket Enrichment

- (w/o) 0 74 0 74 f l ,

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f l Table .4, Summary of Maximum K.m. Including Blas and 95/95 Uncertainties 9 -_

U/s-- K.ev= Km... + Bm iaea ' + (((ks)'o... - (ks)*m.,nos )..

m i

Fu'll' Density Analysis-

. o Nominal. -0 9201 Worst Case 0 9314 Optimun Moderation Analysis H 3 0 0.020 g/cc O.7286 H 0 0 0.045.g/cc o.7907 4 . ,

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• H2O e o.070 g/cc o.7708 3

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CELL CENTER TO CENTER (21.0" Figure 1. Byron /Braidwood Fresh Fuel Storage Cell Nominal Dimtasions

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U 21" L I ,

71.87u 21" TYP--- ---

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- 21" j l h 71.87" 21" TYP-- ---

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' BIBLIOGRAPHY

1. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent l Fuel Storage 'and Handling Applications,, April 14, 1978.

2. W. E. ' Ford \\l, CSRL-V: Processed ENDflB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNLICSDITM-160, June 1982.

3. N. M. Greene. AMPX: A Modular Code System for Generating Coapted Multigroup Neutron-Gamma Libraries from ENOFIB, ORNLITM-3706, March 1976.

4. L. M. Petrie and N. F. Crcss, 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 fue/, 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.

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Bibliography. 16

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