ML20039G953

Cycle 8 Reload Analysis, Revision 1

ML20039G953
Person / Time
Site:	Dresden
Issue date:	12/31/1981
From:	Chandler J, Morgan J, Owsley G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17194A407	List: ML17194A406 ML17252A552 ML20039G953 ML20039G956
References
XN-NF-81-76, XN-NF-81-76-R01, XN-NF-81-76-R1, NUDOCS 8201190340
Download: ML20039G953 (61)

Text

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I I DRESDEN UNIT 3 CYCLE 8 RELOAD ANALYSIS I

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l XN-NF-81-76 I Revision 1 12/31f81 DRESDEN UNIT 3 CYCLE 8 RELOAD ANALYSIS I  ;

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Prepared by: d -

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J. C. Chandler Reload Fuel Licensing I

Concur: / / 2 !//, /

%."F. Owsley, Marpger /

Reload Fuel Licensing Approved: 3 W m c- 13[/7[T5(

J. C Morgan, Maiiager Li ensing & Safety Engineering Approved: 7d /

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,3 R. B. Stout, Manager jE Neutronics & Fuel Management

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Approved: , . e 1 G. A. Soft ~ uanac

Fuel E neering Technical Services

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.I EKON NUCLEAR COMPANY,Inc.

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U. S. CUSTOYER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLE ASE RE AD CAREFULLY Exxon Nuc! ear Company's warranties and representations concerning the I.

subject matter of this document are those set forth in the Ag eement be+ ween Exxon Nuclear Company, Inc. and the Customer pursuant to which this document is issued. According'y, except as otherwise ex pressly provided in such Agreement, neither Exxon Nuclear Company, Inc. nor any person acting on its behalf makes any warranty or representation, expressed or implied, v. Mh respect to the accuracy, comp!eteness, or usefulness of the information contained in this document, or that the use of any mformation, apparatus, method or process disclosed any liabilities with respect to the use of, or for damages resuf ting from the use of i any information, appara%s, method or process disclosed in this document.

The information contained herein is for the snte use of Customer, in ordw to avoid irrpairment of rights of Exxon Nuclear Company, Inc. -

in patents or inventions which may be includcd in the information con-tained in this document, the recipient, by its acceptance of this document i agrees not to publish or make pub'ic use (in the patent sense of the term) of such information until so authorized in writing by Exxon Nuc! ear Company, Inc. nr until aher six (6) months fo!!owing termination or expirat:on of the aforesaid Agreernent and any extens;on thereof, un!ess othe+w se expressly provided in the Agreement, No rights or licenses in or to any patents are irno!ied by the furnishing of this document.

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I i XN-hF-81-76

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Revision 1 1

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I TABLE OF CONTENTS i

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Page Section 1

1.0 INTRODUCTION

2 2.0 MECHANICAL DESIGN ..................................

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3.0 THERMAL HYDRAULIC DESIGN ...........................

3 j 3.1 DESIGN CRITERIA ...............................

3.1.1 Hydraulic Compatibility ................ 3 3.1.2 Therma'. Margic Performar.ce ............. 3 4

3.1.3 Fuel Centerline Temperature ............

4 3.1.4 Rod Bow ................................

3.1.5 Bypass Flow ............................ 4 4

3.2 HYDRAULIC CHARACTERIZATION ....................

3.3 HYDRAULIC COMPATIBILITY AND THERMAL MARGIN PERFORMANCE ............................ 5 7

3.4 FUEL CENTERLINE TEMPERATURE AT OVERPOWER ......

8 3.5 ROD BOW .......................................

l 8 3.6 BYPASS FLOW ...................................

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I 3.7 FUEL CLADDING INTEGRITY SAFETY LIMIT .......... 8 20 4.0 NUCLEAR DESIGN .....................................

20 4.1 BUNDLE NUCLEAR DESIGN .........................

4.1.1 Neutronic Design Parameters ............ 21 I

4.1.2 Enrichment Level and Distribution ...... 21 I

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l TABLE OF CONTENTS (Cont.)

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Section _

l 21 4.2 CORE NUCLEAR DESIGN ...........................

l 4.2.1 Core Configuration ..................... 21 l

4.2.2 Core Reactivity Characteristics ........ 22 4.2.3 Control Rod Patterns ................... 22 l

4.2.4 Core Stability ......................... 23 ANTICIPATED OPERATIONAL OCCURRENCES ................ 30 1 5.0 1

5.1 ANALYSES OF PLANT TRANSIENTS AT RATED CONDITIONS .............................. 30 31 5.2 ANALYSES FOR REDUCED FLOW OPERATION ...........

31 5.3 FUEL LOADING ERROR ............................

31 5.4 CONTROL R0D WITHDRAWAL ERROR ..................

32 5.5 THERMAL MARGIN DETERMINATION ..................

5.6 ASME OVERPRESSURIZATION ANALYSIS .............. 32 POSTULATED ACCIDENTS ............................... 43 6.0 43 6.1 LOSS OF COOLANT ACCIDENT ANALYSES .............

6.1.1 Break Spectrum Analysis ................ 43 6.1.2 Limiting Break Analysis ................ 43 44 6.2 CONTROL ROD DR0P ACCIDENT .....................

47 7.0 OPERATING LIMITS ...................................

47 7.1 LIMITING SAFETY SYSTEM SETTING 5 ...............

7.1.1 Fuel Cladding Integrity Safety Limit ... 47 7.1.2 Steam Dome Pressure Safety Limit ....... 47 I

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Section Page I 7.2 LIMITING CONDITIONS FOR OPERATION ............. 47 j

l 7.2.1 Average Planar LHGR .................... 47 I

7.2.2 Minimum Critical Power Ratio ........... 47

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8.0 REFERENCES

FOR EXXON NUCLEAR METHODOLOGY FOR BOILING WATER REACTORS ............................. 50

lg 9.0 ADDITIONAL REFERENCES .............................. 52 l I

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f tv X N-Ni -tti - /6 Revision 1 I LIST OF TABLES I

Table Page 3.1 Hydraulic Characterization Comparison Between 9

ENC XN-1 and GE 8x8R .............................

l 3.2 Dresden 3 Thermal Hydraulic Design Conditions .... 10 Thermal Mydraulic Analysis Results:

I 3.3 Dresden 3 Cycle 8 - Mixed Core ................... 13 Critical Power Ratio Results for Different I 3.4 Core Configurations (Dresden 3, Maximum Power Assemblies with 1.5 Radial Peaking) .............. 14 3.5 Uncertainties Considered in the MCPR Safety Limit ..................................... 15 I 4.1 Dresden 3 Reload Batch XN-1 Neutronic Design Values .................................... 24 5.1 Significant Input Parameters to the I Transient Analyses ............................... 34 5.2 Analytical Results for Plant Transients at I Rated Conditions ................................. 35 5.3 Control Rod Withdrawal Error Analyses ............ 36 I. 5.4 Limiting Transient ACPR for Resident Fuel Types ....................................... 37 6.1 Sumary of Resul ts of ECCS Analysis . . . . . . . . . . . . . . 45 6.2 Control Rod Drop Accident ........................ 46 7.1 MCPR Operating Limits ............................ 48 I

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j v XN-NF-81-76 E Revision 1 l I l I LIST OF FIGURES

!I Figure Page j 3.1 Axial Power Profile .............................. 16 t

i Central Orifice Zone Demand Curves ............... 17

3.2 a

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Assembly Voids versus Assembly Power ............. 18 3.3 3.4 Dresden 3 Cycle 8 Safety Limit Radial Power Histogram ........................................ 19

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1 4.1 E nr i c hme n t D i s t r i b ut i o n . . . . . . . . . . . . . . . . . . . . . . . . . . 27 i

i 4.2 Dresden Unit 3 Cycle 8 Reference Loading Pattern .......................................... 28 4.3 Decay Ratio vs. Reactor Power .................... 29 5.1 Scram Reactivity Used in the PTSBWR3 Analyses .... 38 l

5.2 Starting Control Rod Pattern for Control Rod Withdrawal Analysis .............................. 39 5.3 Generator Load Rejection Without Bypass .......... 40 S.4 Generator Load Rejection Without Bypass .......... 41 S.5 Generator Load Rejection Without Bypass .......... 42 7.1 Dresden Unit 3 MAPLHGR vs Hot Assembly Burnup I Results for Fuel Type XN-1 8x8 .................. 49 I

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1.0 INTRODUCTION

j This report describes the analyses performed by Exxon Nuclear Company

. I (ENC) in support of the Cycle 8 (XN-1) reload for Dresden Unit 3, which l I

is schedulen to commence operation in tr.: Spring of 1982. Dresden 3 is l

!B the first BWR/3 to be licensed on the basis of ENC analyses. Some of these analyses may be referenced in subsequent ENC jet pump BWR submittals iI as generic to jet pump BWR installations in general and to BWR/3 plants in particular.

This report addresses the following areas:

lE E Mechanical Design a) b) Thermal-hydraulic Design l c) Nuclear Design d) Anticipated Operational Occurrences i

1 e) Postulated Accidents jB f) Operating Limits

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[ 2 XN-NF-81-76 Revision 1 2.0 MECHANICAL DESIGN The mechanical design of BWR fuel fabricated by ENC is described on a generic basis in XN-NF-81-21(P), " Generic Design Report - Mechanical Design #,r Exxon N;. clear Jet Pump BWR Fuel Assemblies," dated November 1981 (Ro erence 9.1). This reference document addresses design bases,

- descriptions and design drawings, design evaluations, and plans for testing, surveillance, and inspection. The design bases and evaluations establish criteria for the determination of fuel system damage and assure (1) that normal operation and anticipated operational occurrences do not

. result in the violation of any of the established criteria, and (2) that ENC analyses of postulated accidents do not underestimate the number of fuel rod failures.

The mechanical design of ENC XN-1 8x8 fuel is covered by the generic analyses and evaluations in Reference 9.1.

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1 i 3.0 THERMAL HYDRAULIC DESIGN This section describes the thermal hydraulic design of the ENC XN-1 8x8 fuel which will be used for Cycle 8 and subsequent cycles in Dresden 3. Design criteria are identified and evaluations in support of those criteria are reported. The evaluations have shown that the subject fuel design satisfies the established criteria in the areas of hydraulic I compatibility, thermal margin performance, fuel centerline temperature,

{ rod bow, and bypass flow. In addition, this section reports the MCPR Fuel Cladding Integrity Safety Limit for Dresden 3 Cycle 8, which was l determined to be 1.05.

3.1 DESIGN CRITERIA Primary thermal hydraulic design criteria of Exxon Nuclear Company reload fuel for BWR's which are applicable to Dresden 3 are as follows:

3.1.1 Hydraulic Comnatibility The hydraulic flow resistance of the reload fuel assemblies shall be similar to existing fuel in the reactor so that there is no significant impact on total core flow or the flow distri-bution among the assemblies in the core.

3.1.2 Thermal Margin Performance Fuel assembly design shall minimize the likelihood of boiling transition during normal reactor operation and during anti-cipated operational occurrences. The fuel design shall fall within the limits of applicability of the XN-3 critical power con ciation.

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!I 3.1.3 Fuel Centerline Temperature Fuel design and operation shall be such that fuel center-line melting is not expected for anticipated operational occurrences.

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I 3.1.4 Rod Bow Anticipated magnitudes of rod bow shall not impact

! thermal margin in BWR fuel designs.

3.1.5 Bypass Flow Bypass flow frcs reload fuel assemblies shall match I design bypass flow from exining fuel to provide adequate flow in the ,

bypass region. i 3.2 HYDRAULIC CHARACTERIZATION Component hydraulic flow resistances for the ENC XN-1 8x8 l

reload design and the GE 8x8R design have been determined in single I phase flow tests of full scale assemblies. The GE 8x8R design is typical of 8x8 and P8x8R designs for purposes of hydraulic characterizations.

Table 3.1 summarizes the component flow resistances for the two designs.

In developing this table, the test results have been modified slightly to account for the differences between the tests and actual reactor I operation.

The total loss coefficient of the inlet hardware, including the effects of the side entry inlet orifice loss lumped together with the lower tie plate losses, was determined in the flow tests. The loss coefficient of 3.6 for the lower plate shown in Table 3.1 for ENC fuel I is adjusted for the geomatry of the ENC XN-1 fuel design. The difference in lower tie plate loss coefficients between the ENC and the GE fuel I

r 5 XN-NF-81-76' Revision 1 b designs is also indicated in Table 3.1. This result reflects the dif-ference measured in the flow tests between the total inlet hardwara loss of the GE design and the total inlet hardware loss of the ENC design.

3.3 HYDRAULIC COMPATIBILITY AND THERMAL MARGIN PERFORMANCE Hydraulic compatibility as it relates to thermal margin per-formance and the relative thermal margin performance of the ENC XN-1 8x8 reload design and the GE 8x8R design have be3n determined with detailed E

thermal hydraulic analyses which have calculated critical power ratios (CPR's) of ENC XN-1 and GE 8x8R fuel types in difference core configurations.

The GE 8x8R design is typical of 8x8 and P8x8R designs for purposes of hydraulic compatibility and thermal margin performance. The configurations analyzed are as follows:

[ Case 1. The Dresden 3 Cycle 8 mixed core configuration in which 224 ENC XN-1 reload fuel assemblies are coresident with existing GE fuel types including 200 GE 8x8R assemblies.

The analysis of this configuration provides the relative thermal margin performance of ENC and GE 8x8R fuel designs

[ in a representative mixed core configuration where each

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fuel type experiences the same core pressure drop.

Case 2. A modified Cycle 8 configuration in which the 224 ENC assemblies are replaced with GE 8x8R assemblies to yield a an all GE core. Comparison of results from this case

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with those of the first case provides the impact of the ENC XN-i reload on thermal margins for GE fuel.

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Case 3. An all ENC configuration (i.e., Dresden 3 Cycle 11).

! Comparison of results from this case with those of the i

first case provides the impact of coresident GE fuel on thermal margins for ENC fuel.

! The peak assembly powers of both the ENC XN-1 8x8 design and 1

!3 the GE 8x8R design are assumed to be the same in all cases (5.235 MWt)

3 and correspond to an assembly radial peaking factor of 1.50.

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The analyses were performed using the methodology described in References 8.6, 8.10 and 8.11 as implemented in ENC's XCOBRA thermal hydraulic program (Reference 9.2). Critical power ratios are calculated within XCOBRA using the XN-3 critical power correlation (Reference 8.9).

Table 3.2 summarizes the input conditions for the analysis.

The core loading for the Cycle 8 mixed core configuration case (Case 1) is also defined. The analysis includes consideration of rod-to-rod I local peaking impact on criticai power for the ENC XN-1 and GE 8x8R fuel types. Figure 3.1 provides the axial power distribution applied to all fuel assemblies in the analysis.

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Table 3.3 provides a summary of calculated thermal hydraulic parameters from the Dresden 3 Cycle 8 mixed core configuration case. l The differences between ENC XN-1 and GE 8x8R parameters are small. As seen in Table 3.3, the flow to the maximum power ENC assembly is some-what larger than the flow to the GE 8x8R assembly at the same power ,

level. The higher flow for the ENC XN-1 8x8 design reflects the some-what lower hydraulic resistance of this design versus the GE 8x8R.  :

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Although the demand curves shown in Figure 3.2 are specific to the j Dresden 3 Cycle 8 mixed core case, the relative comparison of ENC XN-1 and GE 8x8R designs is applicable in any core loading where these fuels i

are coresident. Figure 3.3 shows active region axial average void fraction versus assembly power for the ENC XN-1 and GE 8x8R designs.

1 Table 3.3 gives CPR values of 1.536 and 1.486 respectively for i the maximum power ENC XN-1 and GE 8x8R assemblies in the Dresden 3 Cycle 8 mixed core loading case. The 3% higher operating critical power l

j j ratio for the ENC XN-1 design compared to the GE 8x8R design is primarily l

a result of its greater heat transfer area and higher flow.

I 3.4 ~ FUEL CENTERLINE TEMPERATURE AT OVERPOWER Fuel rod centerline temperatures are determined at 120% over-l power conditions as a check against the occurrence of calculated center-line melting during anticipated operational occurrences. The conditions l at 100% power for maximum temperature in the ENC XN-1 design are a rod i

exposure of 21,200 MWD /MT and a nominal peak power of 13.93 kw/ft. Fuel temperatures under these conditions envelope maximum temperatures expected at any exposure for ENC XN-1 reload fuel in the Dresden 3 reactor. The peak power at 120% overpower is 16.7 kw/ft. The peak centerline tempera-ture at the 13.93 kw/ft (100%) nominal design condition was calculated 1

to be 39090F. The peak centerline temperature at 120% overpower (16.7 kw/ft) was calculated to be 46070F. The melting point of 002 decreases 580F l

per 10,000 MWD /MT. At 21,100 MWD /MT exposure, the melting point is about 49000F so that the margin to centerline melt for the 120% overpower l

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i I design condition is 2930F. All other expected operating conditions have I

greater margin to fuel centerline melting.

These calculations have been performed with ENC's RODEX2 code.(8.13) i 3.5 R0D B0W Post-irradiation examination of BWR fuel fabricated by ENC has shown that the magnitude of fuel rod bowing is very small. No impact on I

thermal margins is expected from these small dimensional changes.

! 3.6 BYPASS FLOW As shown in Table 3.3, bypass flow for the Dresden 3 end-of-Cycle 8 mixed core loading case is calculated to be 10.4% of the total I core flow (98 x 106 lbm/hr). This result is within the usual range of bypass flow for BWR's and represents adequate bypass flow.

l 3.7 FUEL CLADDING INTEGRITY SAFETY LIMIT The Fuel Cladding Integrity Safety Limit was calculatad using 4

j the methodology reported in Reference 8.10. The design basis power i

l distribution histogram shown in Figure 3.4 was used to determine the safety limit. The Monte Carlo procedure used the uncertainties I summarized in Table 3.5 to calculate the limit. The detailed analysis 1

is reported in Reference 9.3.  ;

! The analysis demonstrates that a Fuel Cladding Integrity

E Safety Limit of 1.05 provides assurance that durir.g steady state operation r

at the safety limit, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition.

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ll 9 XN-NF-81-76 Revision 1 i Table 3.1 Hydraulic Characterization Comparison Between ENC XN-1 and GE 8x8R i

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! ENC GE 8x8R Lower Tie Plate Loss Coefficient (KLTP) 3.60 3.57*

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Upper Tie Plate Loss Coefficient 0.378 0.901 6.413 R e .165 Spacer loss Coefficient 1.734 Re .069 lI Bare Rod Friction Factor 0.208 R e .20 0.212 Re .20

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• This is at Reynolds number of 200,000. More generally, the difference (

l between GE and ENC lower tie plate pressure losses as referenced to the GE 8x8R bare rod flow area is given by:

" GE AKLTP KLTPGE _ KLTPENC BRENC 1

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47.59 Re .148 7.85

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10 XN-NF-81-76 I Revision 1 Table 3.2 Dresden 3 Thermal Hydraulic Design Conditions I l Reactor Conditions Core Power Level (MWt)(100%) 2527

! Core Exit Pressure (psia) 1026 Core Inlet Er.thalpy (BTV/lbm) 521.8 Total Recirculating Cnolant Flow (lbm/hr) 98.0 x 106 Core Loading l Central Region Peripheral Region 8x8 (GE), Type 1 (w/o channel seal) 32 36 I

8x8 (GE), Type 2 (with channel seal) 184 48 8x8R (GE) 200 --

l 8x8 (ENC) 224 --

Total 640 84 I

Core Power Distribution 1

Axial Power Shape Figure 3.1 Average Bundle Power (MWt) 3.49 Central Region (Average) 3.81 Peripheral Region (Average) 1.09

11 XN-NF-81-76 Revision 1 Table 3.2 Dresden 3 Thermal Hydraulic Design Conditions (Cont.)

Core Power Distribution (Cont.)

Average Bundle Power by Type (MWt) 8x8 (GE), Type 1, Central 2.31 8x8 (GE), Type 2, Central 3.19 8x8R (GE) 4.25 8x8 (ENC) 4.13 8x8 (GE), Type 1, Peripheral 1.01 8x8 (GE), Type 2, Peripheral 1.15 Maximum Bundle Power (MWt) 8x8R (GE) 5.235 8x8 (ENC) 5.235 Fuel Assembly Description Hydraulic Resistance Characteristics Table 3.1 Fuel Rod Diameters (inch) 8x8 (GE) .493 8x8R (GE) 8x8 (ENC) ,

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Revision 1

'I draulic Table 3.2 Dresden 3 Thermal Design Conditions H[ Cont.)

Fuel Assembly Description (Cont.)

1. Fuel Rods / Assembly 8x8 (GE) 63 8x8R (GE) 62 8x8 (ENC) 63

1. Spacers (all fuel types) 7 Active Fuel Length (feet) 12.1 Total Fuel Rod Length (feet) 13.1 I

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Table 3.3 Thermal Hydraulic Analysis Results Dresden 3 Cycle 8 - Mixed Core Core Average Results Exit Enthalpy (BTU /lbni) 609.8 Exit Quality (active region) 11.1%

Exit Void Fraction 0.583 Axial Average Void Fraction 0.316 Flow Hole Leakage Pressure Drop (psi) 5.6 Bypass Flow 10.4%

Core Pressure Drop (psi) 15.5 I

Naximum Power Assembly (1.5 Radial Peaking) Results ENC 8X8 GE 8X8R I Assemoly Flow (lbm/hr) 118.5 x 103 111.6 x 103 Exit Quality (active region) 19.0% 20.4%

Exit Void Fraction 0.742 0.757 I Axial Average Void Fraction , 0.465 0.482 Critical Power Ratio 1.536 1.486 I

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lI Table 3.4 Critical Power Ratio Results for Different Core Configuations (Dresden 3, Maximum Power Assemblies with 1.5 Radial Peaking)

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I ENC XN-1 GE 8X8R I Case 1 Cycle 8 - Mixed Core 1.536 1.486 I

Case 2 Modified Cycle 8 --

1.495 GE 8X8R replacing ENC XN-1 I

Case 3 Cycle 11 - ENC Core 1.509 --

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Table 3.5 Uncertainties Considered in the MCPR Safety Limit I

Parameter Standard Deviation

• Reference Feedwater Flow Rate 0.0176 8.10 Feedwater Temperature 0.0076 8.10 Core Pressure 0.0050 8.10 Total Core Flow Rate 0.0250 8.10 Core Inlet Enthalpy 0.0024 9.2 XN-3 Critical Power Correlation 0.0411 8.9 Assembly Flow Rate 0.0280 8.10 Power Distribution Radial Peaking Factor 0.0518 9.2 Local Peaking Factor 0.0246 8.1 I

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• Fraction of nominal value I

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Revision 1 il 4.0 NUCLEAR CESIGN This section provides a bundle nuclear design evaluation and a core l

nuclear ce'ign report.

4.1 BUNDLE NUCLEAR DESIGN The results of the neutronic design analysis for ENC's XN-1 8x8 fuel design for Dresden Unit 3 are presented in this section.

Analyses were performed using the methodology described in Reference 8.1.

The key nuclear design characteristics for the E*!C XN-1 8x8 fuel assembly are summarized below.

. The fuel asseably average enrichment, including a six inch top and bottom natural (0.711 w/o U-235) uranium blanket, is 2.69 w/o U-235. The average enrichment of the cen 'al portion of the fuel assembly is 2.87 w/o U-235 and is 133.24 inches in length.

. Five enrichments are utilized to yic'd a flat local power distribution which results in a balanced design relative to MCPR and MAPLHGR limits.

. The assembly contains five burnable poison rods contain-ing 3.0 w/o Gd 023 blended with 2.25 w/o U-235 to reduce the initial reactivity of the assembly.

. The fuel assembly contains 63 fueled rods and one non-fueled water rod.

I I

I XN-NF-81-76 I 21 Revision 1 I

4.1.1 Neutronic Design Parameters The key neutronic design parameters for the ENC Type XN-1 fuel design are presented in Table 4.1.

4.1.2 Enrichment Level and Distribution The nominal enrichment level (average fissile content) of the enriched lattice of the ENC XN-18x8 reload fuel assemblies is 2.87 w/o U-235. The maximum lattice km in the normal reactor core geometry at peak reactivity is 1.224.

The enrichment distribution of the ENC Type XN-1 reload fuel desion was selected on the basis of maintaining a balance between local power peaking factors, assembly reactivity, maximum average planar linear heat generation rate (MAPLHGR), and minimum critical power ratio (MCPR) considerations. The enrichment distribution of the ENC XN-1 8x8 reload design is shown in Figure 4.1.

4.2 CORE NUCLEAR DESIGN This section provides a description of the core configuration established for Cycle 8 operation of Dresden Uniti 3. Core nuclear design analyses were performed using the methodology reported in Reference 8.1.

4.2.1 Core Configuration The reference Cycle 8 core loading pattern and fuel assembly inventory are shown in Figure 4.2. All the listed assemblies are irradiated except for those identified as XN-1 8x8, which are unir-radiated at the beginning of Cycle 8. No 7x7 fuel remains in the core for Cycle 8.

I

I I 22 XN-NF-81-76 Revision 1 I The nominal end-of-Cycle 7 core average exposure is

. 21,292 MWD /MTV. The beginning and end of Cycle 8 core average exposures are 12,945 MWD /MTV and 21,127 MWD /MTU respectively. For the Cycle 8 cold shutdown reactivity calculations, the end of Cycle 7 exposure is 20,806 MWD /MTU.

4.2.2 Core Reactivity Characteristics The calculated 80C8 cold (680F) core k-effective values at all rods out and all rch in are 1.099 and 0.944 respectively.

The Technic ' Specifications require the reactor core to be subcritical by 0.25% Ak in the most reactive condition with the strongest control rod stuck out of the core. The most reactive cold shutdown condition for Cycle 8 occurs at BOC. The calculated core k-effective with the strongest rod out is 0.984 resulting in a shutdown margin of 1.6% Ak. The R value for Cycle 8 is 0.04% Ak to account for the effect of B4C settling in the absorber tubes.

The standby liquid control system is capable of bringing the reactor from full power to a cold shutdown assuming none of the withdrawn control rods can be inserted. With a boron concentration of 600 ppm in the reactor water, the mcximum core keff is 0.944 at cold, xenon free conditions. The calculated shutdown margin (Ak) of the liquid control system is 0.0565 compared to the required Technical Speci-fication minimum value of 0.03.

4.2.3 Control Rod Patterns Operating control rod patterns are not expected to vary significantly from those typically used in the past.

I

I 23 Xii-NF-81-76 ,

Revision 1 )

I g 4.2.4 Core Stability i 3 The stability of the Cycle 8 core was determined analy-tically. The resultant decay ratios are shown as functions of core l power in Figure 4.3. The calculated value of the decay ratio at the intersection of the natural circulation flow line and the 100 percent rod line is 0.45.

I lI I

I

'I lI il I

I I

I

.I

1 i

i XN-NF-81-76 24 l Revision 1 l 1

Table 4.1 Dresden 3 Reload Batch XN-1

,I Neutronic Design Values i

I jW Fuel Pellet Fuel Material UO2 Sintered Pellets Density, g/cc , 10.36

% of TD 94.5 Di amete r, inches 3.57 w/o U-235 pellets 0.4055 lI i

Others 0.4045 Pellet Axial lieight Ref. 9.1 Dish Volume (total), % of Pellet Volume l Enriched 1.0 l Natural 0.0 l Reference Fuel Temperature, *F I Enriched Natural 915 740 Fuel Rod Fuel Length, inches 145.24 Fuel Stacked Density, g/cc 10.26 I Diametral Pellet-to-Clad Gap, inches 3.57 w/o U-235 fuel Others 0.0085 0.0095 Cladding Material .Zircaloy-2 Claa I.D., inches 0.414 Clad 0.D., inches 0.484 Initial Pressurization Ref. 9.1 I

I E

ll J 25 XN-f1F-81-76 Revision 1 I Table 4.1 Dresden 3 Reload Batch XN-1 Neutronic Design Values (cont.)

l lI j Fuel Assembly i

l Number of Rods, Total 64 Fuel Rods I Low-Low Enrichment (1.35 w/o)

Low Enrichment (1.90 w/o) 4 1

Medium-Low Enrichment (2.25 w/o) 19 Neoium-Low Enrichment with Gd 0 (2.25 w/o + 3.0 w/o Gd 23 0 ) 2 3 5 Medium Enrichment (3.34 w/o) 16

{ High Enrichment (3.57 w/o) 18 g Inert Water Rod 1 Fuel Rod Pitch, inches 0.641 i

l Fuel Assembly Loading, Kg 002 197.3

!E j5 Fuel Assembly Loading, Kg U 173.9 Core Data j

, Number of Fuel Assemblies 724 Rated Thermal Power Level, MWt 2527 Rated Core Flow, Mlbm/hr 98.0

! Core Inlet Subcoolina, BTU /lbm 24.6 Moder ator Temperature,

• F 546 Channel Dimensions j Th ianess , in. 0.080 Internal f ace-to-f ace aimension, in. 5.278 Fuel Asse:r.bly Cell Dimensions

, Assembl y Pi tch, in. 6.0 i

j Wide Gap Th ickness, in. 0.750

! Narrow Gap Thickness, in. 0.374 I

5 1

i iI I3 jW 26 XN-NF-81-76 Revision 1

'1 I Table 4.1, Dresden 3 Reload Batch XN-1

]

Neutronic Design Values (cont.)

!I 1

Control Rod Data Total Bl ade Span, in. 9.750 i Total Blade Support Span, in. 1.5630 Blade Thickness, in. 0.3120 B1ade Face-to-Face Internal Dimension, in. 0.200 j

I Number of B4C Rods per 81ade 84 B4 C Rod 0.D., in. 0.188 l- B4 C Rod I . D. , in . 0.138 Percent of B4C Theoretical Density 70 I ,

I .

I I

I I i I i

I 27 XH-NF-81-76

• Devis10n I l

f I  :

L

ML

._e.

ML

M

M  : M

ML  : ML I

e

ML

ML* : M  : H  : H  : ML* M  : ML  :

ML  : ri  : H  : H  : H  : H  : ML* : M  :

1  : ML  : M  : H  : H  : W  : H  : H  : M  :

I

ML  : M  : H  : H  : H  : H  : H  : M  :

j

ML  : ML  : M  : H  : H  : H  : M  : ML  :

i I

2 . . . .

L  : ML. : ML  : M  : '1  : M  : ML*

• ML  :

y _________________________________________________

I .- . . . .

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• L  :

il 4 I DE i

l LL ---

1.33 W/0 U235 W L ---

1 . 9 ': . '/0 U235 ML ---

2 25 W/0 U235 I M 3.39 W/0 U235 H ---

3.57 W/0 U235 ML* ---

2.25 W/0 U235 + 3.00 W/0 GD203

• W ---

INERT WATER R00 I -

I 4

' I FIGURE 4.1 ENRICHENT DISTRIBUTION I

i

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

uw p -

..s.

28 XN-NF-81-76

-Revivor 1 C1 A2 C1 A2 C1 B2 C1 C1 B2 C1 C1 C1 C1 B4 A4 A2 C1 00 C1 D0 C1 00 B2 00 B2 DO B2 DO B2 A4 C1 DO A2 00 B2 D0 B2 00 C1 D0 C1 D0 C1 B2 A4 A2 C1 00 C1 DO B2 DO B2 DO B2 D0 C1 DO B2 A3 C1 D0 B2 D0 B2 D0 C1 DO C1 DO C1 00 C1 B3 A4 B2 C1 D0 B2 DO B3 D0 B3 DO B2 DO C1 B2 B3 C1 00 B2 DO C1 D0 C1 DO C1 D0 C1 B3 B3 C1 B2 DO B2 DO B3 DO B3 DO C1 DO B2 A3 B2 DO C1 DO C1 DO C1 00 C1 DO B2 B2 B4 C1 B2 DO B2 00 B2 D0 C1 DO C1 B2 84 C1 DO C1 DO C1 00 C1 00 B2 B2 A3 C1 B2 00 C1 DO C1 B3 B2 B2 B4 C1 DO C1 00 C1 B2 B3 A3 B4 B4 B2 B2 B2 B3 B3 A4 A4 A4 A3 A4 I XY X = Fuel Type Y = Cycles Irradiated Fuel Number of Type Assemblies Description A 72 GE 8x8 2.50 w/o U-235 B 228 GE 8x8 2.62 w/o U-235 C 200 GE P8x8R 2.65 w/o U-235 D 224 XN-1 8x8 2.69 w/o U-235 Figure 4.2 Dresden Unit 3 Cycle 8 Reference Loading Pattern l (Cne Quarter of Symmetrical Core Loading)

i I

29 XN-NF-81-76 Revision 1 i

~

l 1.0 - - - - - - - - - - - - - - - ~ - - - - - ~ ~ ~ ~ ~ ~ ~

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d 0.4 i

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/ 100% '

!3 Rod Line 0' ' ' ' ' '

O 20 40 60 80 100 j Percent Power Figure 4.3 Decay Ratio vs. Reactor Power 4

lI 4

4

I 30 XN-NF-31-76 Revision 1 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Analyses are made to demonstrate the ability of the fuel to perform satisfactorily during infrequent and moderately frequent operational events and to establish appropriate operating limits for the reactor.

The generic methodology used for the analysis of these anticipated events has been reported in References 8.1, 8.6, 8.8 and 8.12 The purpose of this section is to report the results of the anticipated operational occurrence analyses performed in support of the operation of Dresden 3 Cycle 8.

5.1 ANALYSES OF PLANT TRANSIENTS AT RATED CONDITIONS The generator load rejection without bypass transient was determined to be the limiting transient for full power, full flow opera-tion. This determination was made after evaluating a number of transient c,ents for the change in thermal margin associated with them using the in:wt values reported in Table 5.1. The considered transients and the primacy results of the analyses are reported in Table 5.2. Uncertainties in the input parameters for these other transients were assumed to be at bounding values. These analyses are reported in detail in Reference 9.3.

The generator load rejection without bypass transient was evaluated to determine thermal margin requirements using the generic statistical methodology described in Reference 8.12. Results of that analysis are reported in Section 5.5.

I -

I l r

I

[{

4 31 XN-NF-81-76 Revision 1 5.2 ANALYSES FOR REDUCED FLOW OPERATION The MCPR reduced flow multiplier, kf , was reevaluated. The cod,ination of the MCPR Operating Limit and the kf curves as established in Reference 9.5 provides adequate protection of the MCPR Fuel Cladding Integrity Safety Limit during anticipated operational cccurrences from all attainable power-flow combinations. Analyses in support of the revised kr curves are reported in Reference 9.5.

5.3 FUEL LOADING ERROR The inadvertent loading of a fuel bundle into an incorrect core location and the inadvertent rotation of a fuel bundle 180 degrees from its intended crientation were analyzed using the methodology described in Reference 8.1. The largest calculated ACPR for the fuel loading error is 0.16 for both the ENC XN-1 and GE 8x8 fuel types. This ACPR was determined by the fuel misloading error analysis; the fuel misorienta-tion resulted in a smaller ACPR value.

5.4 CONTROL R00 WITHDRAWAL ERROR The consequences associated with the inadvertent withdrawal of a high worth control blade until its motion is halted by the rod block was evaluated using the methodology described in Reference 8.1 using the limiting control rod pattern shown in Figure 5.2. The results are reported l

parametrically with rod block setting in Table 5.3. The rod block monitor setting for Cycle 8 was selected to be 110% as indicated in the table; at this setting, the largest ACPR for the rod withdrawal error is 0.15.

I I

I i 32 XN-NF-81-75 Revision 1 I

5.5 THERMAL MARGIN DETERMINATION The thermal margin requirements for Cycle 8 operation were determined from the consequences of the generator load rejection without bypass transient using the methodology described in Reference 8.12. The statistical predictor variables selected for the statistical analysis were scram delay time, scram insertion speed, scram reactivity, and void reactivity. Application of the statistical methodology to the parameters of the limiting transient resulted in the ninety-fif th percentile ACPR of 0.25; this value was use6 in determining the MCPR Operating Limit.

This methodology was applied to each fuel type resident in the core, and the results are reported by fuel type in Table 5.4. The results for GE 8x8R and P8x8R fuel types were calculated to be identical because of equal values for gap conductance.

The MCPR Operating Limits were determined from the values for the Limiting Transient ACPR as noted in Table 5.4 and the Fuel Cladding Integrity Safety Limit as determined in Section 3.7. The operating limit values are reported in Section 7.0. Plant responses to the limit-ing transient at nominal input conditions are shown in Figures 5.3-5.5.

5.6 ASME OVERPRESSURIZATION ANALYSIS In accordance with the provisions of the ASME Code, an over-pressurization analysis was performed using the COTRANSA plant transient simulation code. The analysis showed that even if failure of the most critical active component were to be assumed (i.e., the scram associated with the closure of the Main Steam Isolation Valve were not to occur and the event were to be terminated by the APRM high flux scram), and if no I

[g 33 XN-NF-81-76 i3 Revision 1 lI i

!g credit were to be allowed for operation of the four electromatic relief ig

{ valves, the inadvertent MSIV closure event would not result in reactor i

vessel pressure exceeding 110% of the design pressure. The maximum vessel pressure observed in the analysis was 1346 psig, which corresponds to 108% of the reactor vessel design pressure. The steam dome pressure i

! associated with the maximum pressure was 1324 psig, which indicates that a pressure safety limit of as high as 1353 psig as measured by the steam space pressure indicator would provide assurance that the 110% of design

! pressure criterion was not exceeded.

The MSIV closure transient evaluated for ASME code compliance resulted in a higher maximum pressure than that noted for the turbine j trip without bypass transient evaluated under the same assumptions.

!.I

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i 34 XN-NF-81-7A i Revision 1 l lI i

Table 5.1 Significant Input Parameters to the l

j Transient Analyses l l

i Reactor Thermal Power (MWt) 2527 l Steam Dome Pressure (psia) 1020 FeedwaterEnthalpy(BTU /lbm) 304.1 j Scram Reactivity Fig. 5.1 90% Scram Insertion Time (sec) 3.5 Nominal Void Reactivity Loefficient ($/VF) -15.89 I

Nominal Doppler Reactivity Coefficient ($/0F) .0026 Core Average Gap Conductance (BTV/hr-ft2_oF) 893 Limiting Assembly Gap Conductance (BTU /hr-ft2_oF) 1430 I

I I

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i ,

Table 5.2 Analytical Results for Plant Transients at Rated Conditions (l) l Analytical Maximum Maximum Maximum Transient Model Exposure Power Flow Heat Flux Neutron Flux Pressure (2) ACPR i

' Generator Without Bypass Load Re (4}jection COTRANSA E0C8 100% 100% 114.3% 372% 1294 psia 0.25(3)

. Loss of 1450F Feedwater i Heating PTSBWR3 E0C8 100% 100% 119.2% 120% 1056 psia 0.16 i

Feedwater Controller Failure COTRANSA EOC8 100% 100% 116.8% 293% 1214 psia 0.21 NOTES:

1. Values associated with bounding uncertainty values; applicable to all fuel types present in core ,

2. Maximum transient pressure calculated for lower plenum

3. Statistically determined change in MCPR; applicable to all fuel types present in core mx

?Y

4. Parameters associated with nominal uncertainty values reported for Generator Load 7%

Rejection Without Bypass {h "h

I 36 XN-NF-81-76 Revision 1 I

Table 5.3 Control Rod Withdrawal Error Analyses I

ACPR Rod Block Withdrawn GE ENC Reading, Rod Position, 8x8 8x8 5  % Feet I

105 3.5 0.09 0.09 106 4.0 0.10 0.11 107 4.5 0.12 0.13 0.12 0.13 108 4.5 109 5.0 0.14 0.15 110* 5.0 0.14 0.15 I

I

• Selected rod block monitor setting for Cycle 8 I

t I

I 1 I I

,5 XN-NF-81-76 I

37 Revision 1 i

f i Table 5.4 Limiting Transient ACPR for Resident Fuel Types I

!I

g Transient 8x8(XN-1) aCPR 8x8R(GE)* 8x8(GE) i3 2

Generator Load Rejection .25 .25 .25 (w/o bypass) 3 l .21 .21 .21

Increase in Feedwater Flow l

Loss of Feedwater Heating .16 .16 .16

!I

!I l

!I i

I

• Typical value for 8x8R and P8x8R fuel types.

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!I 38 XN-NF-81-76 I

Revision 1 >

l

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d Calculated I 5 C

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Value l ,, -20 -

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N N

I -10 -

• 9" Value

-5 -

I l I 0_ .5 1.0 FRACTION OF CONTROL ROD INSERTION Figure 5.1 Scram Reactivity Used in the PTS 8WR3 Analyses I

I I 39 XN-NF-81-76 Revision 1

! 31 30 40

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27 24 6 4 24 1

23 36 1

I 19 0* 10 I

15 38 24 11 30 24 30 I 07 34 l

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30 34 38 42 46 50 54 58 Note:

• Control Rod Being Withdrawn, Rod Positions in Notches, Full In = 0, Full Out = Blank or 48 3 Figure 5.2 Starting Control Rod Pattern for l

Control Rod Withdrawal Analysis B

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I Jg 43 XN-NF-81-76 l5 l

Revision 1 il l 6.0 POSTULATED ACCIDENTS This section describes the analyses that were undertaken to assure

,g the capability of the reload core to withstand the effects of postulated 1g accidents.

6.1 LOSS OF COOLANT ACCIDENT ANALYSES This subsection describes the analyses performed in accordance with 10CFR50.46 and Appendix K to 10CFR50.

6.1.1 Break Spectrum Analysis A spectrum of potential LOCA break locations and sizes was reported in Reference 8.14. The conclusion in that document was that the limiting break was the double-ended guillotine recirculation line break at the suction to the recirculation pump. That potential break was selected for further analysis as the limiting break.

6.1.2 Limiting Break Analysis The selected limiting break was analyzed using the ,

methodology reported in References 8.2, 8.3, 8.4, 8.5 and 8.7. The fuel rod stored energy was evaluated with the methodology reported in Reference 8.13. The detailed ECCS analysis was reported in Reference 9.4.

The results of the ECCS analysis are reported in summary form in Table 6.1, which contains values for MAPLHGR, peak cladding temperature and peak local oxidation for several representative exposure points in the life of the fuel. These results are used to define the MAPiiiGR operating limits for fuel type XN-18x8 in Section 7.0 and are valid for fuel type XN-1 8x8 within the specified exposure limits so long as it re7ains in the Dresden Unit 3 core.

I

i 1

i 44 XN-NF-81-76 f Revision 1 i

!I

l 6.2 CONTROL ROD DROP ACCIDENT l

}

j The control rod drop accident was performed on a generic basis I

in Reference 8.1. The pertinent parametric values and the resultant

! deposited enthalpy are reported in Table 6.2. The calculated deposited enthalpy of 151 calories / gram is well within the allowable 280 calories / gram jg specified in the Dresden Unit 3 Technical Specifications.

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i IE 45 XN-NF-81-76 1 IE Revision 1 i i

8 l

I i Table 6.1 Summary of Results of ECCS Analysis I

l lI l Bundle Average Peak Cladding Peak Local Metal-Water I Exposure (MWD /MT)

MAPLHGR (kw/ft)

Temperature (OF)

Reaction

(%) ,

0 13.0 1879 0.8 10,000 13.0 1942 1.0 15,000 13.0 2089 2.8 18,000 12.85 2134 4.1 20,000 12.6 2156 4.1 25,000 11.95 2088 3.4 30,000 11.2 1937 2.1 35,000 10.45 1819 1.4 I

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1 l I -

l l 46 XN-NF-81-76 l

! Revision 1 l

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Table 6.2 Rod Drop Accident Analysis l

!I i Dropped Control Rod Worth = 11.2 mk Do;;cler Coefficient (/73 F) = -10.2x10-6 Jp ,

K A ( F)

T  !

= 0.0058 6effcctive Four Bundle Local Peaking (P4BL ) " I'129  !

l Maximum Deposited fuel Rod Enthalpy = 151 calories / gram lI l

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I I 47 XN-NF-81-76 Revision 1 I

7.0 OPERATING LIMITS I This section sumarizes the results of the analyses reported in the earlier sections of this report by collecting the proper operating limits as indicated by the analyses.

7.1 LIMITING SAFETi SYSTEM SETTINGS 7.1.1 Fuel Cladding Integrity Safety Limit (Specification 1.1.A)

Operation with a Minimum Critical Power Ratio of less than 1.05 shall constitute violation of the Fuel Cladding Integrity Safety Limit.

7.1.2 Steam Dome Pressure Safety Limit (Specification 1.2)

The reactor coolant system pressure shall not exceed 1353 psig at any time when irradiated fuel is present in the reactor vessel.

7.2 LIMITING CONDITIONS FOR OPERATION 7.2.1 Average Planar LHGR (Specification 3.5.I)

During steady state power operation, the Average Planar Linear Heat Generation Rate (APLHGR) of all the rods in any fuel assembly of type ENC XN-1 8x8 at any axial location shall not exceed the value indicated by the exposure-dependent function shown in Figure 7.1.

APLHGR limitations for other fuel types resident in the core are not changed by this report.

7.2.2 Minimum Critical Power Ratio (Specification 3.5.K)

During steady state operation, MCPR shall be greater than or equal to the values given in Table 7.1 times the appropriate value of k f.

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I XN-NF-81-76 Revision 1 I

Table 7.1 MCPR Operating Limits I

f Fuel Type MCPR Operating Limit I

E XN-1 8x8 1.30 l

I I 8x8R/P8x8R 1.30 l

0*0 1.30 l

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I I 50 XN-NF-81-76 Revision 1 I

8.0 R_EFERENCES FOR EXXON NUCLEAR METHODOLOGY FOR BOILING WATER REACTORS The following reports describe the ENC methodology for the analysis of jet-pump boiling water reactors. They are incorporated in this sub-mittal by reference.

8.1 XN-NF-80-19(P), Volume 1 (Supplements 1 and 2), May 1980 Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods for Design and Analysis I 8.2 XN-NF-80-19(P), Volume 2, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors EXEM: ECCS Evaluation Model, Summary Description 8.3 XN-NF-80-19(P), Volume 2A, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors I RELAX: A RELAP4 Based Computer Code for Calculating Blowdown Phenomena 8.4 XN-NF-80-19(P), Volume 28, Revision 1, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors FLEX: A Computer Code for det Pump BWR Refill and Reflood Analysis I 8.5 XN-NF-80-1c,(P), Volume 2C, June 1981 Exxon Nuclear Methodology for Boiling Water Reactors Verification and Qualification of EXEM 8.6 XN-NF-80-19(P), Volume 3, Revision 1, April 1981 Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology, Summary Description I 8.7 XN-CC-33(A), Revision 1, November 1975 HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option 6a XN-f:F-79-71(P), Revision 2, November 1981 Exxon Nuclear Plant Transient Methodology for 3 oiling Water Reactors 8.9 XN-NF-512(P), Revision 1, March 1981 The XN-3 Critical Power Correlation I 8.10 XN-NF-524(P), November 1979 Exxon Nuclear Critical Power Methodology for Boiling Water Reactors 8.11 XN-NF-79-59(P), October 1979 Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies I

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i I 8.12 XN-NF-81-22(P), September 1981 Generic Statistical Uncertainty Analysis Methodology l 8.13 XN-NF-81-58(P), August 1981 le RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model il 8.14 XN-NF-81-71(P), October 1981 Generic Jet-Pump BWR3 LOCA Analysis Using the ENC EXEM Evaluation Model I

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Revision 1 l

! 9.0 ADDITIONAL REFERENCES Although not specifically identified as part of the Exxon Nuclear i

m Methodology for Boiling Water Reactors listed in Section 8.0, the l5 1

following documents provide background information, j 9.1 S. F. Gaines, " Generic Design Report - Mechanical Design for Exxon l Nuclear Jet Pump BWR Fuel Assemblies," XN-NF-81-21(P), November l 1981.

I 9.2 T. W. Patten, "XCOBRA Code User's Manual," XN-NF-CC-43(P),

Revision 1, January 1980.

lI l 9.3 R. H. Kelley, "Dresden 3 Cycle 8 Plant Transient Analysis Report,"

l XN-NF-81-78, Revision 1, December 1981.

) 9.4 J. E. Krajicek, "Dresden Unit 3 LOCA Analysis Using the ENC EXEM l Evaluation Model--MAPLHGR Results," XN-NF-81-75(P), October 1981.

9.5 R. H. Kelley, "Dresden Unit 3 Analyses for Reduced Flow Operation," l

{ XN-NF-81-84(P), November 1981.

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! AN-NF-81-76 i Revision 1 13 Issue Date:

i3 12/31/81 DRESDEN UNIT 3 CYCLE 8 RELOAD ANALYSIS lI l

Distribution JC Chandler I RE Collingham GC Cooke NF Fausz LJ Federico JW Hulsman l

SE Jensen 15 WV Kayser 5 RH Kelley JE Krajicek

,3 TL Krysinski 15 JL Maryott JN Morgan l GF Owsley l GA Sofer

RB Stout LC O'Malley/ Ceco (60)

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