ML20012B095

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Criticality Analysis of South Texas Project Fresh Fuel Racks
ML20012B095
Person / Time
Site: South Texas  
Issue date: 12/31/1989
From: Biswas D, Fecteau M, Krieg D
HOUSTON LIGHTING & POWER CO.
To:
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ML20012B092 List:
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NUDOCS 9003130501
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l' CRITICALITY ANALYSIS OF THE SOUTH TEXAS UNITS 1 & 2 FRESH FUEL RACKS I

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I December 1989 I

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M. Fecteau 3

D. Biswas

'g D. Krieg W. Bordogna I

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I 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 Rocks.........................

3 3.1 Full Density Moderation Analysis 3

3.2 Low Density Optimum Moderation Analysis

.................5 I.

3.3 Postulated Accidents

...............................6 4.0 Acceptance Critorion For Criticality............................

7 13 Bibliography I

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Table of Contents i

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LIST OF TABLES Table

1. Benchmark Critical Experiments [ SA),.................

8 Table

2. Fuel Parameters Employed in Criticality Analysis 9

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5 List of Tables il I

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.I LIST OF ILLUSTRATIONS Figure

1. South Texas Fresh Fuel Storage Cell Nominal Dimensions 10 Figure
2. South Texas Fresh Fuel Storage Array Layout 11 Figure
3. Sensitivity of K.o tes Water Density in the South Texas Fresh Fuel I

12 Storage Racks I

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List of Illustrations ill I.

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1.0 INTRODUCTION

The South Texas fresh fuel rack design described herein employs an existing arrsy of unpoisoned racks, which will be analyzed for the storage of Westinghouse 17x17 STD, XL, OFA, and VANTAGE 5 fuel assemblies.

This I

analysis will show that Westinghouse 17x17 STD, XL, OFA, and VANTAGE 5 fuel assemblies with nominal enrichments up to 4.5 w/o U'" can be stored in the fresh fuel rack array utilizing every storage location.

The fresh fuel rack analysis is based on' maintaining K.ee 5 0.95 for storage of Westinghouse 17x17 STD, XL, OFA, and VANTAGE 5 fuel with nominal I

enrichments up to 4.5 w/o U'" under full water density and optimum moderation conditions.

I 1.1 DESIGN DESCRIPTION The fresh fuel rack storage cell design is depicted schematically in Figure 1 on page 10. The fresh fuel rack layout as used in the optimum moderation analysis is shown in Figure 2 on page 11.

l 1.2 DESIGN CRITERIA Crincality 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

' 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 ef fective multiplication f actor (K.te) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.3-1983 and in Reference 1.

Introduction 1

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,4-1 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 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 safety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codes for cross-section I

generation and KENO IV"' for reactivity determination.

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

for all cross-sections used for the benchmarks. and the storage rack analysis is generated from ENDFIB-V data. The NITAWL program includes, in this 11-brary, the self-shielded resonance cross-sections that are appropriate for each I

particular geometry.

The Nordhel'm Integral Treatment is used.

Energy and spatial weighting of cross-sections is performed by the XSDRNPM program I

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 8 summarizes these experiments.

The average Keet of the benchmarks is 0.992. The standard deviation of the bias value is 0.0008 Ak, The 95/95 one sided tolerance limit f actor 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, i

Criticality Analytical Method 2

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I 3.0 -CRITICALITY ANALYSIS OF FRESH FUEL RACKS 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 water dent.ity and low water density'(optimum moderation) conditions. The full density and I

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 r:o credit is taken for any natural enrichment axial blankets or burnable absorbers in the fuel rods,

=

2.

All fuel rods contain uranium dioxide at an enrichment of 4.50 w/o (nominal) and 4.55 w/o (" worst case") U'".

3.

All fuel rods are modelled with a fuel stack height which is infinitely long for the full density moderation scenario and 168 inches long for the opti-mum moderation scenarlo.

l 4.

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

[

5.

No credit is taker, for any U'" or U'" in the fuel.

8.

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

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 lateral and axial extent which precludes any neutron leakage from the array. Figure 1 on page 10 depicts the fresh fuel rack cell nominal dimensions.

I The Westinghouse 17x17 OFA fuel assembly yields a larger K.ve (by approxi-5 mately 1 to 2 %Ak/k) than does the Westinghouse 17x17 STD/XL fuel assembly under full density moderation conditions when both fuel assemblies have the 7g Criticality Analysis of Fresh Fuel Racks 3

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same U"' enrichment and fuel stack height. The VANTAGE 5 fuel design pa-rameters relevant to the criticality analysis are the same as the OFA parameters and will yield equivalent results. Thus, for the full density optimum moderation scenario, an infinitely long Westinghouse 17x17 OFA fuel assembly was ana-lyzed (see Table 2 on page 9 for fuel parameters).

The KENO calculation for the nominal case resulted in a K.ve of 0.9044 with a 95 percent probability /95 percent confidence level uncertainty of 10.0082.

The maximum Kees under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing process. Due to the relatively large cell spacing, the small tolerances on the I

cell 1.D. and center-to-center spacing are not considered since they will have an insignificant effect on the fuel rack reactivity. However, the sheet metal thickness is reduced to its minimum tolerance. The assemblies are symmet-I rically positioned within the storage cells since the relatively large cell-to-cell spacing causes the reactivity effects of asymmetric assembly positioning to j

be insignificant. Furthermore, fuel enrichment is assumed to be 4.55 w/o U"'

B to conservatively account for enrichment variability. Thus, the most conserva-tive, or " worst case" KENO model of the fresh fuel storage rackc contains the I

minimum sheet metal thickness with symmetrically placed fuel assemblies at 4.55 w/o U"'.

B Based on the analysis described above, the following equation is used to de-velop the maximum K.ve for the South Texas fresh fuel storage racks:

K.et = K.o..i + Bm.ineo + / ((ks)'....i + (k s)'m.inoe )

where:

K. orsi worst case KENO K.tv with full density water

=

method bias determined from benchmark critical Bm.iwoo

=

comparisons 95/95 uncertainty in the worst case KENO K.ee ks.or i

=

kam.inov 95/95 uncertainty in the method bias

=

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

K.et = 0.9080 + 0.0083 + /[(0.0087)' + (0.0018)' ) = 0.9252 Since K.et is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met.

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Criticality Analysis of Fresh Fuel Racks 4

l 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 'in modelling the actual fresh fuel rack array which is depicted in Figure 2 on page 11. Concrete walls and floor are modelled. Under low water I

density conditions, the presence of concrete is conservative because neutrons are reflected back into the fuel array more efficiently than they would be with I

just low density water. The area above the fresh fuel rack is filled with water st the optimum moderation density.

i The Westinghouse 17x17 STDIXL fuel assemtily was analyzed in the model with I

I a fuel stack height of 168 inches (see Table 2 on page 9 for fuel parameters).

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The STDIXL fuel assembly is more. reactive than the 17x17 OFA or VANTAGE I

5 fuel assembly (by approximately 0.5 to 1.5 %Ak/k) under low moderator den-sity conditions when the fuel assemblies have the same U'" enrichment and fuel stack height. This is because the STD/XL fuel assembly contains a higher ura-I nium loading than the OFA assembly, and when optimum moderation conditions are present, higher loadings result in higher reactivity.

.g Analysis of the South Texas fresh fuel racks has shown that the maximum rack e-K.ve under low density moderation conditions occurs at 0.043 gm/cm' water den:ilty. The K.e# of the South Texas fresh rack at 0.043 gm/cm' water density I

is 0.9190 with a 95 percent probability and 95 percent confidence level uncer-tainty of 10.0086. Figure 3 on page 12 shows the fresh fuel rack reactivity as 1

a function of water density.

Based on the analysis described above, the following equation is used to de-velop the maximum K.ve for the South Texas fresh fuel storage racks under low

. density optimum moderation conditions:

Bm.inos + /((ks)'d... + (ks)'m.inoa )

l K.ev= Kb...

+

where:

Km...

maximum K es with optimum moderation

=

Bmeros a

method bias determined from benchmark critical comparisons k se...

95/95 uncertainty in the maximum K.es

=

ksm.inoa 95/95 uncertainty in the method bias

=

^

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

K.,e = 0.9190 + 0.0083 + /[(0.0086)' + (0.0018)' ) = 0.9361 I

Since K.ev is less than 0.95 including uncertainties st a

95/95

- probability / confidence level, the acceptance criteria for criticality is met.

Criticality Analysis of Fresh Fuel Racks 5

S 3.3 POSTULATED ACCIDENTS Under normal conditions, the fresh fuel racks are maintained in a dry environ-s 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 ;ases are bounding accident situations which result in the most conservative fuel rack K.et.

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-

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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.

The maximum reactivity increase for postulated accidents (such as those men-tioned above) will be less than 10 %Ak/k. Furthermore, the normal, dry fresh

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fuel rack reactivity is less than 0.70. As a result, for postulated accidents, the maximum rack Keet will be less than 0.95.

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Criticality Analysis of Fresh Fuel Racks 6

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I 4.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication f actor in the fresh fuel racks shall be less than or equal 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 ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."

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i Acceptance Criterion For Criticality 7

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v Table 1.

Benehmark Crit 6 sal Experiments [BA]

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4 General

. Enrichment Separating

. Soluble Description w

Reflector Material Boron ppm Koff

..................../.o U235 I

1. UO2 rod lattice 2.46 water water O

O.9857 +/-.0028

2. UO2 rod lattice 2.46 water water 1037 0.9906 +/-.0018
3. UO2 rod lattice 0.46 water water 764 0.9896 +/-.0015 I

4 UO2 rod lattice 2.46 water Bec pins O

O.9914 +/-.0025

5. UO2 rod lattice 2.46 water B4C pins O

O.9891 +/-.0026

6. U02 rod lattice 2.46 water 84C pins O

O.9955 +/

.00tO

7. U02 rod lattice 2.46 water B4C pins O

O.9889 +/-.0027

8. UO2 rod lattice 2.46 watee 84C pins O

O.9983 4/-.0025 I

9. UO2 rod lattice 2.46 water water O

O.9931 + /-.0028 10.~ UO2 rod 1sttlee 2.46 water water 143 0.9928 +/-.0025

11. UO2 rod lattice 2.46 water stainless steel 514 0.9967 +/*.0020
12. UO2 rod lattice 2.46 water stainless steel 217 0.9943 +/-.0019
13. UO2 rod lattice-2.46 water borated aluminum 15 0.9892 +/-.0023 14 U02 rod lattice 2.46 water borated aluminum 92 0.9884 +/-.0023 I
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 1sttico 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 B
20. UO2 rod lattice 2.46 water borated aluminum 320 0.9920 +/-.0020

-21. UO2 rod lattice 2.46 water borsted aluminum 72 0 9939 +/-.0020

22. U metal cyltnders 93.2 bare air O

O.9905 */-.0020

23. U metal cyltnders 93.2 bare air O

O.9976 +/-.CD20

24. U metal cyttnders 93.2 bare air O

O.9947 +/-.0025 I

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 cylinders 93.2 bare plexiglass O

O.9941 + /*.0030

29. U metal cylinders 93.2 paraffin plexiglass O

O.9928 +/-.0041 I

30. U metal cylinders 93.2 bare plexiglass O

O.9968 +/-.0018

31. U metal cyltnders 93.2 paraffin plexiglass 0

1.0042 +/-.0019

32. U metal cylinders 93.2 paraffin pleutglass O

O.9963 +/-.0030

33. U mets) cy)inders 93.2 paraffin plexiglass O

0,9919 +/-.0032 I

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'y Talple 1.

Puol Parameters Employed in Crtt6 canty Analysis Parameter W 17x17 OFA-W 17x17 STD/XL

& VANTAGE 5 I

Number of Fuel Rods per Assembly 264 264 I

Rod Zirc-4 Clad 0.D. (I nch) 0 360 0 374

'C1ed Thlekness (Inch) 0.0225 0.0225 Fuel Pellet 0.D. (Inch)'

O.3088 0 3225 Fuel Pellet Density

(% of Theoretical) 96 96 Fuel Pellet Dishing Factor 0.0 0.0 I

Rod Pitch (Inch) 0.496 0.496 Number of Zirc-4 Guide Tubes 24 24 Guide Tube 0.D. (Inch) 0.474 0.482 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 (Inch) 0.016 0.016 I

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BASIC FUEL CELL 21' X 21*

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.88 9e

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.86

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.08 H2O DENSITY (G/CC)

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Notet Error bars represent 95/95 tolerance about the keno calculated Keff I

Figure 3.

Sensitivity of K.et to Water Density in the South Texas Fresh Fuel Storage Racks 12

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i BIBLIOGRAPHY 1.

Nuclear Regulatory Commission, Letter to All Power Reactor i

' Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent l

Fuel Storage and Handling Applications., Aprll 14, 1978.

j

2.. W.

E.

Ford lll, CSRL-V:

Processed ENOFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORML/CSDITM-160, June 1982.

3.

N.

M.

Greune, AMPX: A Modular Code System for Generating Coupled I

Multigroup Neutron-Gemma librarles from ENDFIB, ORNLITM-3706, March 1976.

I 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 Stologe I

of Power Reactor Fuel, BAW-1484-7, July 1979.

6.

J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, I

Nuclear Science and Engineering, Volume 52, pages 350-359,1973.

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Bibliography 13

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-a ATTAcitMENT 4 MARK-UP OF THE UPDATED FINAL SAFETY ANALYSIS REPORT l

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0 STPECS UFSAR i

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-products accumulate, this restriction is relaxed.

However, for the reference final core design described in this chapter, no such withdrawal limit is required.

Ejected rod worths are given in Section 15.4.8 for several different conditions.

Allowable deviations due to misaligned control rods are discussed in the

)

Technical Specificacit,ns.

]

A representative calculation for two banks of control rods withdrawn simultaneously (rod withdrawal accident) is given on Figure 4.3 37.

Calculation of control rod reactivity worth versus time following reactor trip involves both control rod velocity and differential reactivity worth. The rod position versus time of travel after rod release assumed is given on Figure 4.3 38.

For nuclear design purposes, the reactivity worth versus rod position is calculated by a series of steady-state calculations at various control rod positions assuming all rods out of the core as the initial position in order l

to minimize the initial reactivity insertion rate.

Also, to be conservative, the rod of highest worth is assumed stuck out of the core and the flux distribution (and thus reactivity importance) is assumed to be skewed to the bottom of the core. The result of these calculations is shown on Figure 4.3 39.

The shutdown groups provide additional negative reactivity to assure an adequate shutdown margin.

Shutdown margin is defined as the amount by which the core would be subcritical at hot shutdown if all RCCAs are tripped, but assuming that the hi hest worth assembly remains fully withdrawn and no 8

changes in xenon or boron take place.

The loss of control rod worth due to the material irradiation is negligible since only bank D may be in the core under normal operating conditions.

The values given in Table 4.3 3 show that the available reactivity in withdrawn RCCAs provides the design bases minimum shutdown margin allowing for the highest worth cluster to be at its fully withdrawn position. An allowance I

for the uncertainty in the calculated worth of N 1 rods is made before t

determination of the shutdown margin.

4.3.2.6 Criticality of the Reactor Durine Refueline and Criticality i

l-of Fuel Assemblies. The basis for maintaining the reactor suberitical during refueling is presented in Section 4.3.1.5 and a discussion of how control requirements are met is given in Section 4.3.2.4 and 4.3.2.5.

Criticality of fuel assemblies outside the reactor is precluded by adequate i

design of fuel transfer and fuel storage facilities and by administrative control procedures. This section identifies those criteria important to criticality safety analyses.

4.3.2.6.1 New Fuel Storare:

For Unit 1, w

e is red-i d in.,

center to-center racks in the new fuel storage f.aciut11 dry condition.

Prior to initial core loading, new fuel wa(stored wet in the 14 in., center-

~

to center spent fuel ra peksFo rsubsequent refuelings, new fuel may also be stored in the flooded Tondition in the 10.95 in, center to center high density spent fuebratkE For the flooded condition (with unborated water assuming jaw-fGiifor the highest antici iated enrichment [4.5 weight percent 4.3 28 Revision 0

[

STPECS UFSAR uranium 235) in the new or hi h density spent fuel racks) the effective multiplication factor does not exceed 0.95.

For the normally dry co tion in the new fuel storage racks, the effective multiplication facto.' do not exceed 0.98 (with fuel of the highest anticipated enrichment in ace and assuming possible sources of moderation such as aqueous foam mist).

For Unit 2, new fuel is stored in 21 in., center-to-ceyr racks in the new fuel storage facilities in a dry condition.

Prior t initial core loading, new fuel can be stored dry in the 10.95 inch nomi

, center to center high density spent fuel racks.

For subsequent refu ngs, new fuel may also be stored in the flooded condition in the 10.95 n.,

center to center high density spent fuel racks. For the norma 11 dry or flooded condition (with unborated water assuming new fuel of tbe' highest anticipated enrichment [4.5 weightpercenturanium235)inthe)t,Sh density spent fuel racks), the effective multiplication factor fees not exceed 0.95.

For the new fuel racks the effective multiplication 6 tors for the dry and flooded conditions do not exceed 0.98 and 0.95, resp ively, as discussed above for Unit 1.

In the analysis for storage facilities, the fuel assemblies are assumed to be in their most etive condition, namely fresh or undepleted and with no control rods o emevable neutron absorbera present.

Credit is taken for the inherent ne3 on absorbing effect of the construction materials of the racks.

AssemblieVcannot be closer together than the design separation provided by l

the stpriige facility, except in special cases such as in fuel shipping l

con,t4Tners where analyses are carried out to establish the acceptability of design.

The mechanical integrity of the fuel assembly is assumed.

4.3.2.6.2 Soent ruel storame:

4.3.2.6.2.1 Unit i Hnterim Desien) - The following describes wet spent fuel storage in the spent fuel pool in the 14 in, racks in the event spent fuel storage is required prior to their replacement with the 9.15 inch and 10.95 inch nominal high density spent fuel racks. Unborated water of 1.0 g/cm8 is assumed in the analysis.

Over the range of water densities of interest (corresponding to 60*F through 212'F), full density water is a conservative assumption since a decrease in water density will cause the effective multiplication factor (k...) of the system to decrease.

The design basis for wet fuel storage criticality analysis is that, considering possible variations, there is a 95 percent confidence level that the effective multiplication factor (k,,) of the fuel storage array will be less than 0.95 per ANSI Standard N18.2 1973. The possible variations in the criticality analyses are in three categories:

1) calculational uncertainties,
2) fuel rack fabrication uncertainties, and 3) transport effects.

The results of comparing standard calculationc with 101 critical experiments-as summarized in Table 4.3 4 (Ref. 4.3 14) indicate that:

1.

The average difference between the calculations and experimental results or bias in the computations, was 0.1 percent Ak which is denoted as the calculational bias, and L

2.

The standard deviation in the difference between the calculations and experimental results was 0.86 percent Ak.

Multiplying the standard deviation by the appropriate one sided upper tolerance factor results in a calculational uncertainty valid at the 95 percent confidence level.

4.3-29 Revision 0

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/ /VSE/27 i

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' ew~Fue'l Storace:

New fuel is stored in 21 inch center to N

"4.3.2.6.1 conter racks in the new fuel storage facilities in a dry condition.-

For the flooded condition and for the low water density optimum j

moderator condition (with unborated vater assuming fuel of the highest cnticipated enrichment of 4.5 w/o U-235).the effective multiplication l

fcctor does not exceed 0.95.

j i

In the analysis for the storage facilities, the fuel assemblies are Ocsumed to be in their most reactive condition, namely fresh or undepleted and with no control rods or removable neutron absorbers i

present.

Credit is taken for:the inherent neutron-absorbing effect of 1

l the construction materials of the racks.

Assemblies cannot be closer 1

together than the design separation provided by the storage facility, L

sxcept in special cases such as in fuel shipping containers where l-cnalyses are carried out to establish the acceptability of the design.

In-the case of an accident that would increase reactivity, such as an f"

cosembly drop in the normal dry condition ( k gg 5 0.70), the maximum' k gg will be less than 0.95."

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