ML20245B386

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Rev 1 to ANF-89-02, Washington Nuclear Power-2 Cycle 5 Reload Analysis
ML20245B386
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 03/31/1989
From: Krajicek J
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17285A407 List:
References
ANF-89-02, ANF-89-02-R01, ANF-89-2, ANF-89-2-R1, TAC-72251, NUDOCS 8904260084
Download: ML20245B386 (49)
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ANF-89-02 i g REVISION 1 i 1< ..

j jIc_'. ADVANCEDNUCLEARFUELS CORPORATION I

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l WNP-2 CYCLE 5 RELOAD ANALYSIS I

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l MARCH 1989 I

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ADVANCEDNUCLEARFUELS CORPORATION l ANF-89-02 l Revision 1 Issue Date: 3/10/89 ,

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WNP-2 CYCLE 5 RELOAD ANALYSIS I

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Prepared by I gal lw j> A/A /2 9. /9 R?

I ,// ' 7 J. E. Krajicek BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services I

I March 1989 I

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NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS I

DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development pro-grams sponsored by Advanced Nuclear Fuels Corporation. It is being submit.

ted by Advanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad-vanced Nuclear Fuels Corporation fabricated reload fuel or other technical services provided by Advanced Nuclear Fuels Corporation for light water power reactors end it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, information, and belief. The information con-tained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Corporation in their demonstration of compliance with the U.S. Nuclear Regulatory Commission,'s regulations.

Advanced Nuclear Fuels Corporation's warranties and representations con. E coming the subject matter of this document are those set forth in the agree-ment between Advanced Nuclear Fuels Corporation and the customer to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:

A. Makes any warranty, or representation, express or im-plied, with respect to the accuracy, completeness, or usefulness of the information contalned in this docu-ment, or that the use of any information, apparatus.

method, or process disclosed in this document will not infringe pnvately owned rights, or B. Assumes any liabilities with respect to the use of, or for e damages resulting from the use of, any information, ap-paratus, method. or process oisclosed in this document.

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ANF-89-02 Revision 1 Page i

SUMMARY

OF REVISIONS Revision 1 to ANF-89-02 was issued to address a reload batch size change from 144 to 136 assemblies and other minor text changes which describe the reload batch size change. Feedwater controller failure calculated results at 47%

power and 106% flow with normal scram speed and recirculation pump trip are also included for a 144 assembly reload batch size as are editorial comments suggested by the Washington Public Power Supply System.

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ANF-89-02 s Revision 1 Page ii TABLE OF CONTENTS Section Pace

1.0 INTRODUCTION

. . ... ......... .............. 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . ......... 3 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . 4 3.1 Design Criteria . . ...................... 4 3.1.3 Fuel Centerline Temperature . . . . . . . . . . . . . . . 4 3.2 Hydraulic Characterization . . . . . . . . . . . . . . . . . . . 4 3.2.5 Bypass Flow . . . . ................... 4 3.3 MCPR Fuel Cladding Integrity Safety Limit ........... 4 3.3.1 Coolant Thermodynamic Condition . . . . . . . . . . . . . 4

( 3.3.2 Design Basis Radial Power Distribution ......... 4 3.3.3 Design Basis local Power Distribution . . . . ...... 4 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 5 4.1 Fuel Bundle Nuclear Design Analysis .............. 5 4.2 Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . 5 4.2.1 Core Configuration ............. ...... 5

{ 4.2.2 Core Reactivity Characteristics . . . . . . . . . . . . . 5 4.2.4 Core Hydrodynamic Stability . . . . . . . . . .... 6 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . ... 7 5.1 Analysis Of Plant Transients At Increased Core Flow Condi 1ons . 7 5.2 Analyses For Reduced Flow Operation .............. 8 5.3 Analysis For Reduced Power Operation (SLO) . . . . . . . . . . . 8 5.4 ASME Overpressurization Analysis . . . . . . . . . . . . . . . . 9 5.5 Control Rod Withdrawal Error . . . . . ............ 9

, 5.6 Loading Error for Reload Fuels . . . . . ........... 9 5.7 Determination Of Thermal Margins . . . . . . . . . . . . . . . . 9

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, 6.0 POSTULATED ACCIDENTS . . ................ .. ... 12 l 6.1 Loss-Of-Coolant Accident . . . . . . . ...... ..... 12 6.1.1 Break Location Spectrum . ........ ....... 12 6.1.2 Break Size Spectrum . . . . .... ... ..... 12 6.1.3 MAPLHGR Analyses ....... ............ 12 f 6.2 Control Rod Drop Accident ......... .... .... 12

, 7.0 TECHNICAL SPECIFICATIONS ......... .... ....... 13 7.1 Limiting Safety System Settings ....... .. ... .. 13

{ 7.1.1 MCPR Fuel Cladding Integrity Safety Limit . . . ... 13 7.1.2 Steam Dome Pressure Safety Limit ... . ...... 13 7.2 Limiting Conditions For Operation ...... ... ... ... 13

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i 7.2.1 Average Planar Linear. Heat Generation Rate Limits For ANF 8x8 Fuel ......... . .. ..... 13 t 7.2.2 Minimum Critical Power Ratio . ... .. ... ... 13 1

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I ANF-89-02 Revision 1 Page iii 7.2.3 Surveillance Requirements . . . . . . . . . . . . . .. 14 7.2.3.1 Scram Insertion Time Surveillance ....... 14 7.2.3.2 Stability Surveillance . . . . . . . . . . . . . 15 7.2.3.3 Technical Specification LHGR Surveillance ... 15 9.0 ADDITIONAL REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . 29 APPENDIX A 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLIES (LFAs) ....... A-1 I

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1 ANF-89-02' (

Revision 1 J Page iv .q LIST OF TABLES Table Paae

'4' 1 NEUTRONIC DESIGN VALUES .......................

. 16 A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES ..................... .A-4 LIST OF FIGURES Fiaure Paae 3.1 RADIAL POWER HIST 0 GRAM FOR 1/4 CORE SAFETY LIMIT MODEL . . . . . . . 18 3.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL) ... 19 3.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL) . . 20 3.4 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL) . . 21 3.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL) . . 22 4.1 WNP-2 ANF 4 CYCLE 5 ENRICHED ZONE ENRICHMENT DISTRIBUTION ..... 23 4.2 WNP-2 CYCLE 5 REFERENCE LOADING PATTERN BY FUEL TYPE (ONE QUARTER OF SYMMETRICAL CORE LOADING) ........... . 24 5.1 WNP-2 CYCLE 5 CONTROL R0D WITHDRAWAL ANALYSIS INITIAL CONTROL R0D PATTERN .................... 25 5.2 REDUCED FLOW MCPR OPERATING LIMIT FOR NORMAL FEEDWATER TEMPERATURE . 26 5.3 REDUCED FLOW MCPR OPERATING LIMIT FOR FFTR OPERATION . . . . . . . . 27 7.1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 8X8 FUEL'. . . . . . . . . . . . . . . . . . . . . . . 28 A.1 XN-3 8X8 ENRICHED ZONE ENRICHMENT DISTRIBUTION . . . . . . . . . . A-6 A.2 9X9-IX ENRICHED ZONE ENRICHMENT DISTRIBUTION . . . . . . . . . . . A-7 I

A.3 9X9-9X ENRICHED ZONE ENRICHMENT DISTRIBUTION . . . . . . . . . . A-8 A.4 LHGR LIMIT FOR 9X9-IX FUEL . . . . . . . . . . . . . . . . . . . . A-9 A.5 LHGR LIMIT FOR 9X9-9X FUEL . . . . . . . . . . . . . . . . . . . . A-10 A6 ANF 9X9-IX AND 9X9-9X MAPLHGR LIMITS , . . . . . . . . . . . . . . A-ll i

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ANF-89-02 Revision 1 Page 1

1.0 INTRODUCTION

This revised report summarizes the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 5 reload for the Supply System Nuclear Project Number 2 (WNP-2). The original WNP-2 Cycle

'5 core was designed and safety analyses were performed to support a reload batch size of 144 assemblies. This revised reload report supports a WNP-2 l Cycle 5 reload batch size that has been reduced to 136 assemblies. l WNP-2 is scheduled to commence Cycle 5 operation in June 1989. This report is intended to be used in conjunction with ANF topical report \N-NF 19(A), Volume 4, Revision 1, " Application of the ENC Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(A), Volume 4, Revision 1.

Generic final feedwater temperature reduction (FFTR) analysis to support cycle extension was previously performed for WNP-2. This FFTR analysis is applicable for a condition with all the control rods out with normal feedwater temperature. That is, additional MCPR limit changes are applicable to Cycle 5 and future cycles when reactor operation is being extended by reduction of the feedwater temperature.

The WNP-2 Cycle 5 core will comprise a total of 764 fuel assemblies:

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including 132 ANF 8x8 unirradiated assemblies; 2 ANF 9x9-IX unirradiated Lead Fuel Assemblies (LFA); 2 ANF 9x9-9X unirradiated LFAs; 152 once irradiated ANF l 8x8 assemblies,148 twice irradiated ANF 8x8 assemblies; 128 thrice irradiated 3 ANF 8x8 assemblies; and 200 irradiated P8x8R assemblies from the Cycle 1 core fabricated by General Electric (GE). The ANF 9x9 Lead Fuel Assembly (LFA)

) licensing information is given in Appendix A. The reference core configuration is described in Section 4.2.

i Atg.89-02 Revision 1 Page 2

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The design and safety analyses reported in this document were based on the design and operational assumptions in effect for WNP-2 during the previous l

operating cycle which encompass core flow up to 106% of the design basis ,

value.

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ANF-89-02 Revision 1 Page 3 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 9.8 The expected power history for the fuel to be irradiated during Cycle 5 of WNP-2 is bounded by the assumed power history in the fuel mechanical design analyses.

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ANF-89-02 Revision 1 Page 4 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.1 Desian Criteria 3.1.3 Fuel Centerline Temperature The LHGR curve in Figure 3.4 of Reference 9.8 shows that the ANF 8x8 fuel centerline temperature will be below the melting point at 120% over power.

The LHGR curve in Reference 9.8 is greater than 120% above the LHGR limit curve in Reference 9.1. Therefore, fuel centerline melt is protected for all ANF 8x8 exposures within the bounds of the referenced LHGR curves.

3.2 Hydraulic Characterization 3.2.5 Bvoass Flow Calculated Bypass Flow Fraction (100% power /106% flow) 10.7%

3.3 MCPR Fuel Claddina Intearity Safety Limit I 3.3.1 Coolant Thermodynamic Condition Core Power 3950 MWt I Core Inlet Enthalpy 525.6 Btu /lbm Steam Dome Pressure 1021 psia I Feedwater Temperature 414*F 3.3.2 Desian Basis Radial Power Distribution See Figure 3.1.

I 3.3.3 Desian Basis local Power Distribution See Figures 3.2, 3.3, 3.4 and 3.5.

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I ANF-89-02 Revision 1 Page 5 1 4.0 NUCLEAR DESIGN ANALYSIS I 4.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment 2.62 w/o U-235 Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform 2.79 w/o U-235 with 6-inch

.I top and bottom natural uranium l

blankets Burnable Poisons Figure 4.1 Non-Fueled Rods Figure 4.1 1 Neutronic Design Parameters Table 4.1 Note: The reload includes 4 ANF 8x8 assemblies of the 2.64 w/o U-235 design loaded in Cycle 4 and described in the Cycle 4

I Reload Analysis Report ANF-88-02 and 4 9x9 LFAs of the 2.53/2.59 w/o U-235 design described in Appendix A.

4.2 Core ..uclear Desian Analysis 4.2.1 Core Configuration Figure 4.2 Core Exposure at E0C4 (mwd /MTU) 16,900 Core Exposure at B005 (mwd /MTV) 12,500 Core Exposure at E005 (mwd /MTV) 18,300 4.2.2 Core Reactivity Characteristics B0C Cold k-eff, All Rods Out 1.1135 BOC Cold k-eff, Strongest Rod Out 0.9869 Reactivity Defect (R-Value) 0.0 I Standby Liquid Control System (SBLC) 0.9635 660 ppm Boron, Cold k-eff I

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4.2.4 Core Hydrodynamic Stability

% Power /% Flow State Points Decay Ratio (C0TRAN) 65/45* 0.49 -

47/2).6** 0.89 42/23.8*** 0.82 I

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  • 45 percent flow - APRM Rod Block intercept point.
    • Two pump minimum flow - 47 percent power.
      • Natural circulation flow - APRM Rod Block intercept point.

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ANF-89-02 Revision 1 Page 7 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Transient Analysis Report Reference 9.3 5.1 Analysis Of Plant Transients At Increased Core Flow Conditions References 9.3 and 9.11 Limiting Transient (s): Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LOFH)

Transient analyses for WNP-2 Cycle ,2 anticipated operational events showed that delta CPR values at design basis conditions are bounded by delta CPR values at design basis power (104%) and increased core flow conditions (106%). Thus Cycle 5 analyses results at increased core flow conditions are conservatively applicable to rated flow conditions.

I Cycle 5 specific analyses of transient events were performed for two recirculation pump operation conditions, with the recirculation pump trip I (RPT) in service and out of service, and for two scram conditions which are normal scram speed (NSS) a r.. technical specification scram speed (TSSS). Analyses were performed at end-of-cycle exposures. T.he results shown in following table support a Cycle 5 reload batch size of 136 as sembl ies. Specifically, the LRNB case with iSS and RPT operational was rerun, and a 0.01 increase in delta CPR for GE fuel was required. For the remaining three LRNB cases, a 0.01 was added to both the GE and ANF MCPR limits to conservatively account for delta CPR changes due to the reduced reload batch size. Generic analyses we e performed for FFTR to extend cycle operation (Reference 9.11).

The loss of feedwater heating event was analyzed on a plant specific bounding value basis and the delta CPR results are bounding values for WNP-2.

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Maximum Delta CPR

% Power / Maximum Maximum Pressure GE ANF Transient *  % Flow Heat Flux % Power % osia Fuel Fuel B S 104/106 121 406 1169 0.29 0.25 LRNB, NSS 104/106 127 501 1181 0.36 0.32 RPT Inoperable LRNB, TSSS 104/106 127 454 1174 0.36 0.32 RPT Operable LRNB, TSSS 104/106 132 594 1189 0.42 0.36 RPT Inoperable 4 FWCF, NSS 47/106 54 163 1026 0.23 0.20 RPT Operable FWCF, NSS 47/106 56 217 1023 0.29 0.25 RPT Inoperable LOFH N/A N/A N/A N/A 0.09 0.09 5.2 Analyses For Reduced Flow Ooeration References 9.3 and 9.11 Limiting Transient: Recirculation Flow Increase 5.3 Analysis For Reduced Power Ooeration (SLO) References 9.12, 9.13, and 9.14 ANF has performed analyses for WNP-2 which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended period of time. These analyses were performed for the most limiting g>

transient events, the pump seizure accident and the loss-of-coolant-accident 5 (LOCA) for the maximum extended power state during WNP-2 single-loop operation (SLO). The transient analysis and pump seizure accident analysis are documented in Reference 9.12, and the LOCA analysis is documented in Reference 9.13. The conclusions presented in these documents are applicable to ,

future cycles with ANF fuel and have been reviewed by the U. S. Nuclear Regulatory Commission, Reference 9.14; the SLO limits from the USNRC review are summarized below.

  • Normal scram speed (NSS) is based on measurei plant scram insertion data, see l Section 7.2.3.1.

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ANF-89-02 Revision 'l Page 9 SLO MCPR Operating Limit for ANF and GE fuel 1.35 Two-loop MAPLHGR limits which are shown in Section 6.1.3 for ANF fuel apply duri,9 SLO. For GE fuel the reduction of the MAPLHGR limit to a value of 0.84 times the two recirculation loop operation MAPLHGR limit for SLO remains unchanged.

5.4 ASME Overpressurization Analysis References 9.3 and 9.11 Limiting Event MSIV Closure Worst Single Failure MSIV Position Scram Trip Maximum Pressure 1315 psig Maximum Steam Dome Pressure 1286 psig 5.5 Control Rod Withdrawal Error l Initial Control Rod Pattern for CRWE Analysis Figure 5.1 Rod Block ANF Fuel GE Fuel Monitor Settina Distance Withdrawn Delta-CPR Delta-CPR (ft) 106%* 5.5 0.18 0.18 107% 6.0 0.19 0.19 108% 7.5 0.22 0.22 l

5.6 Loadina Error for Reload Fuels l

With Correctly loadina Error Loaded Core I Maximum LHGR, kW/ft 16.6 13.3 Minimum MCPR 1.26 1.44 i

5.7 Determination Of Thermal Marains j Summary of Thermal Margin Requirements

  • Rod Block Monitor Setting (RBM) of 106%

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ANF-89-02 Revision 1 .

Page 10 1 All system transient results were analyzed at the more limiting I'

increased flow conditions (106%) rather than rated flow conditions (100%). LRNB results for the more limiting power (design basis condition - 104%) were used for this transient.

These calculated results are based on end of cycle conditions and increased core flow (106%). The LRNB results, which determine thermal limits, have been revised to reflect a reload batch size of 136 assemblies.

Delta CPR MCPR Limit Equipment GE ANF GE ANF Event Operational Status Fuel Fuel Fuel Fuel Model LRNB RPT Operable, NSS 0.29 0.25 1.35 1.31 COTRANSA/XCOBRA-T LRNB RPT Inoperable, 0.36 0.32 1.42 1.38 E NSS 5 LRNB RPT Operable, TSSS 0.36 0.32 1.42 1.38 LRNB RPT Inoperable, 0.42 0.36 1.48 1.42 "

TSSS FWCF RPT Operable, NSS 0.23 0.20 1.29 1.26 FWCF RPT Inoperable, NSS 0.29 0.25 1.35 1.31 LOFH N/A 0.09 0.09 1.15 1.15 XTGBWR Note: For cycle extension with reduced feedwater temperature, add 0.02 to delta CPR/MCPR LRNB and FWCF transient results in the above table.

MCPR Operating Limits At Rated Condition For Cycle Exposures Less Than g EOC -2000 mwd /MTV are based on the CRWE (100% To 106% Flow) g Fuel Tvoe MCPR Limit (106% RBM)

ANF 1.24 GE 1.24 I

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l MCPR Operating Limits At Rated Condition From E0C -2000 mwd /MTV To E0C (100% To 106% Flow) With Normal Feedwater Temperature I Fuel Tvoe MCPR Limit ANF 1.31 l GE 1.35 5 MCPR Operating Limits At Rated Condition Beyond All Rods Out With Reduced Feedwater Temperature (100% To 106% Flow And Thermal Coastdown)

Point (E005)

Fuel Tvoe

  • MCPR Limit ANF 1.33 GE 1.37 MCPR Limits at Off-Rated Conditions Figures 5.2 and 5.3 Reduced Flow MCPR Limit References 9.3 and 9.11 9

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ANF-89-02 Revision 1 Page 12 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident 6.1.1 Break location Spectrum Reference 9.4 l 6.1.2 Break Size Spectrum Reference 9.4 6.1.3 MAPLHGR Analyses (ANF Fuel - Two-Loop Operation and SLO)

References 9.5,

, 9.13 and 9.14 Limiting Break: Split Break in the Recirculation Suction Piping With an Area Equal to Sixty Percent of the Double-Ended Cross-Sectional Pipe Area Bundle Average l Exposure MAPLHGR Peak Clad

  • Peak Local *
(mwd /MTM) (kW/ft) Temperature. 'F MWR, %

0 13.0 1779 0.50 l 5,000 13.0 1755 0.45 10,000 13.0 1761 0.46 15,000 13.0 1765 0.46 1

20,000 13.0 1771 0.51

/ 25,000 11.3 1659 0.32 30,000 9.4 1513 0.16 35,000 7.9 1385 0.09 l

Heatup analysis shows insignificant changes in PCTs and local MWR, but no [

change in MAPLHGR limits, from thb MAPLHGR analysis for the earlier ANF 8x8 fuel design which is shown in Reference 9.5.

6.2 Control Rod Drop Accident Reference 9.7 Dropped Control Rod Worth, mK 6.23

Doppler v
'ficient dk/kdT, 1/*F -10.0 x 10-6

) Effective velayed Neutron Fraction 0.0050 Four-Bundle Local Peaking Factor 1.18 Maximum Deposited Fuel Rod Enthalpy (cal /gm) 90.1 f

  • For the ANF-4(6Gd2) fuel design PCTs and MWRs.

ANF-89-02 Revision 1 Page 13 7.0 TECHNICAL SPECIFICATIONS .

7.1 Limitina Safety System Settinas 7.1.1 MCPR Fuel Claddina Intearity Safety Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1346 psig 7.2 Limitina Conditions For Operation 7.2.1 Averaae Planar Linear Heat Generation Rate limits For ANF 8x8 Fuel Bundle Average Exposure MAPLHGR <

i (mwd /MTU) (kW/ft) 0 13.0 5,000 13.0 5 10,000 13.0 15,000 13.0 '

20,000 13.0 I' 25,000 30,000 11.3 9.4 35,000 7.9 For single-loop operation these limits also apply to ANF Fuel when using a SLO MCPR limit of at least 1.35.

7.2.2 Minimum Critical Power Ratio I Rated Condition MCPR Operating Limit Up To E0C -2000 mwd /MTV Exposure (100% To 106% Flow)

I Fuel Tvoe Limit (106% RBM)

ANF 1.24 I GE 1.24 I

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Rated Conditions MCPR Operating Limits From E0C -2000 mwd /MTV To E0C (100%' To 106% Flow) 5 E i Fuel Tvoe Limit ANF 1.31 GE 1.35 Thermal Coastdown and FFTR Rated Condit io MCPR Operating Limit Beyond All Rods Out Point With Reduced Feedwater T uperature (100% to 106% Flow)

I Fuel Tvoe Limit ANF 1.33 ,

GE 1.37 Reduced Flow MCPR Limit (all cycle exposures) Figures 5.2 and 5.3 Single-Loop Operation (SLO) MCPR Limit (all cycle exposures)

Fuel Tvoe Limit ANF 1.35 GE 1.35 7.2.3 Surveillance Requirements 7.2.3.1 Scram Insertion Time Surveillance The ANF reload safety analyses were labeled NSS (Normal Scram Speed) performed using the control rod insertion times shown below which are based on plant data. In the event that plant surveillance shows these scram insertion times may be exceeded, the plant thermal margin limits are to default to the g values which correspond to the technical specification (TSSS) control rod E ,

scram times (see Section 5.7).

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Page 15 Position Inserted From Average Rod Time in %conds Fully Withdrawn As Defined in Footnote

  • i Notch 45 0.404 Notch 39 0.r560 Notch 25 1.504 I Notch 5 2.624 7.2.3.2 Stability Surveillance Core hydrodynamic stability analyses are bounded by the Technical Specifications which preclude operation in specified power / flow regions. The results of these analyses support operation below a line defined by the following power / flow points: 42% Power /23.8% Flow, 47% Power / 27.6% Flow, and 65% Power /45% Flow (see Section 4.2.4).

Surveillance requirements remain unchanged for Cycle 5, e.g., s'urveil-lance is required when operating in a power flow region above the 80% rod line and less than 45% core flow.

7.2.3.3 Technical Specification LHGR Surveillance The Technical Specification linear heat generation rate (LHGR) limit versus average planar exposure for ANF 8x8 reload fuel is shown in Figure 7.1.

This figure was developed from information contained in Reference 9.1, and the region of permissible operation is shown.

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  • Slowest measured average control rod insertion time to specified notches for each group of four control rods arranged in a 2x2 array.

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ANF-89-02 Revision 1 Page 16 2-TABLE 4.1 NEUTRONIC DESIGN VALUES E

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

% of T.D. 94.5 Diameter, inch Enriched Fuel 0.4055 Natural Fuel 0.4045 Fuel Rod Fuel Length, inch 150 Cladding Material Zircaloy-2 Clad, I.D., inch 0.414 Clad, 0.D., inch 0.484 Fuel Assembly Number of Fuel Rods 62 Number of Inert Water Rods 2 Fuel Rod Enrichments Figure 4.1 Fuel Rod Pitch, inch 0.641 Fuel Assembly Loading, kgU 176.0 5

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ANF-89-02 Revision 1 Page 17 E

TABLE 4.1 NEUTRONIC DESIGN VALUES (Continued)

Core Data Number of Fuel Assemblies 764 Rated Thermal Power, MW 3323 Rated Core Flow, Mlbm/hr 108.5 Core Inlet Subcooling, Btu /lbm 19.0  ;

Reactor Pressure, psia 1008.0 Channel Thickness, inch 0.100 Fuel Assembly Pitch, inch 6.00 Water Gap Thickness (symmetric), inch 0.522 ,

Control Rod Data Absorber Material 8C 4

Total Blade Span, inch 9.75 ,

Total Blade Support Span, inch 1.58 Blade Thickness, inch 0.260

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l Blade f ace-To-Face Internal Dimension, inch 0.200 Absorber Rods Per Blade 76 Absorber Rod Outside Diameter, inch 0.188 Absorber Rod Inside Diameter, inch 0.138 {

Absorber Density, % of Theoretical 70 E

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.936 : .977 : 1.023 : 1.015 : 1.011 : 1.041 : 1.076 : 1.052 :

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.977 : 1.011 : .907 : 1.042 : 1.035 : .932 : .962 : 1.075 :

4

1.023 : .907 : 1.017 : .988 : . 974 : .996 : .931 : 1.040 : '
1.015 : 1.042 : .988 : .000 : . 850 : .972 : 1.033 : 1.009 :
1.011 : 1.035 : .974 : .850 : . 000 : .985 : 1.038 : 1.011 :
1.041 : .932 : .996 : .972 : . 985 : 1.012 : .901 : 1.043 :

4

1.076 : .962 : .931 : 1.033 : 1.038 : .901 : .976 : 1.078 : i

. . . . 1

1.052 : 1.075 : 1.040 : 1.009 : 1.011 : 1.043 : 1.078 : 1.054 :

....#..*............g.............e...__ e ........._w .e... ....e . . .

FIGURE 3.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)

E E

I  !

i

ANF-89-02 Revision 1 Page 20

.944 : .962 : 1.011 : 1.044 : 1.043 : 1.010 : .960 : .943 :
  • : .962 : .980 : 1.064 : .894 : 1.033 : 1.059 : 1.034 : .961 :
1.011 : 1.064 : 1.010 : .994 : .982 : 1.002 : .915 : 1.010 :
  • : 1,044 : .894 : .994 : .000 : .907 : .980 . 1.032 : 1.042 :
1.043 : 1.033 : .982 : .907 : .000 : .988 : .952 : 1.041 .

f * . . . .

1 . . .

1.010 : 1.059 : 1.002 : .980 : .988 : 1.004 : 1.060 : 1.065 : .

r .

960 : 1.034 : .915 : 1.032 : .952 : 1.060 . .966 : 1.053 :

1 . . . . . .

.943 : .961 : 1.010 : 1.042 : 1.041 : 1.065 . 1.053 : 1.019 :

I FIGURE 3.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS

, (ANF XN-3 FUEL) l

}

I ANF-89-02 Revision 1 Page 21 E

I

  • : .950 : .963 : 1.000 : 1.027 : 1.026 : .999 : .963 : .950 :
.963 : .981 : 1.052 : .920 : 1.033 : 1.049 : 1.020 : .963 :
1.000 : 1.052 : 1.017 : 1.005 : .997 : 1.011 : .936 : 1.000 :
1.027 : .920 : 1.005 : 600 : .935 : .996 : 1.033 : 1.027 :

i

1.026 : 1.033 : .997 : .935 : .000 : 1.002 : .971 : 1.027 :
  • : .999 : 1.049 : 1.011 : .996 : 1.002 : 1.016 : 1.054 : 1.042 -

. .963 : 1.020 : .936 : 1.033 : .971 : 1.054 : .973 : 1.029 : I

. . . . . . \

. . . . . l

.950 : .963 : 1.000 : 1.027 : 1.027 : 1.042 : 1.029 : 1.003 : 1 1

1 de . _*_....m.e. ..e.................es .............e e . ...............

FIGURE 3.4 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS en, xs.2 ru m g

I' I;

l l

ANF-89-02 Revision 1 Page 22

  • .967 : .969': .997 : 1.019 : 1.019 : .996 : .968 : .966 :

.969 :

.981 : 1.044 :

.932 : 1.030 : 1.042 : 1.013 :

.968 :

.997 : 1.044 : 1.017 : 1.008 : 1.001 : 1.012 :

.944 :

.997 :

I

  • 1.019 : .932 : 1.008 : .000 : .947 : 1.000 : 1.030 : 1.019 :
  • 1.019 : 1.030 : 1.001 : .947 : .000 : 1.006 : .976 : 1.020 :
  • .996 : 1.042 : !.012 : 1.000 : 1.006 : 1.017 : 1.047 : 1.032 :
.968 : 1.013 : .944 : 1.030 : .976 : 1.047 : .975 : 1.020 :

1 . . . . . . .

i ...................... ........................_.................

.966 : .968 : .997 : 1.019 : 1.020 : 1.032 : 1.020 : 1.003 :

FIGURE 3.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS

> (ANF XN-1 FUEL)

}

ANF.89-02

                                                    • Revision 1 Page 23
LL  : L  : ML  : M  : M  : M  : ML  : L  :
L  : M  : ML* : H  : H  : M  : ML*  : ML  :

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L  : ML  : M  : M  : M  : M  : ML  : L  :

I .................................................................

I LL RODS ( 1) ---

1.50 W/0 U235 L RODS ( 5) ---

2.00 W/0 U235 ML RODS ( 9) ---

2.50 W/0 U235 M RODS (21) ---

2.64 W/0 U235 I H RODS (20)

ML* RODS ( 6)

W RODS ( 2) 3.43 W/0 U235 2.50 W/0 U235 + 2.00 W/0 GD203 INERT WATER R0D I

I FIGURE 4.1 WNP-2 ANF4 CYCLE 5 ENRICHED ZONE ENRICHMENT DISTRIBUTION I

I _ __ _,_,_,__

I ANF-89-02 l Revision 1 5 Page 24 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 1 D C F C D C F C D C D C D F B 2 C H B H D H C H D H B H D B B 3 F B E B F B F B F D F D D F A 4 C H B F C H B H C H B H B N A 5 D D F C G C F B F D F D D F B 6 C H B H C k C H C H B H C D B 7 F C F B F C H B E D F D F D A B C H B H B H B F C H B H B A 9 D D F C F C E C I D E F A 10 C H D H D H D H D C C A A 11 D B F B F B F B E C C 12 C H D H D H D H F A 13 D D D B D C F B A A 14 F B F H T D D A 15 B B A A B B A Fuel Type Number of Tvoe Assemblies Description A 56 GE 8x8 Type 11 1.76 w/o U-235 (Cycle 1)

B 144 GE 8x8 Type III 2.19 w/o (Cycle 1) E l C 128 ANF 8x8 2.72 w/o U-235 (Cycle 2) 5 D 148 ANF 8x8 2.72 w/o U-235 (Cycle 3)

E 24 ANF 8x8 2.72 w/o U-235 (Cycle 4)

F 128 ANF 8x8 2.64 w/o U-235 (Cycle 4)

G 4 ANF 8x8 2.64 w/o U-235 (Cycle 5)

H 128 ANF 8x8 2.62 w/o U-235 (Cycle 5)

I 4 ANF 9x9 Lead 2.53/2.59 w/o U-235 (Cycle 5) .

FIGURE 4.2 WNP-2 CYCLE 5 REFERENCE LOADING PATTERN BY FUEL TYPE l (0NE QUARTER OF SYMMETRICAL CORE LOADING)

( '

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ANF-89-02 Revision 1 Page 25 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 -- -- -- -- -- - --

59 55 -- --

00 --

40 --

00 -- --

55 51 -- -- -- -- -- -- -- -- -- -- --

51 47 -- -- 36 --

16 --

00 --

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47 43 -- -- -- -- -- -- -- -- -- -- -- -- -- -- --

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16 -- 00 --

36 --

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19 15 --

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15 11 -- -- -- -- -- -- -- -- -- -- -.

11 7 -- --

00 --

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7 3 -- -- -- .- -- .. --

3 l

2 6 10 14 18 22 26 30 34 38 42 46 50 54 58

\

  • Control Rod Being Withdrawn Rod Position in Notches Withdrawn Full In = 00 l

Full Out = --

r FIGURE 5.1 WNP-2 CYCLE 5 CONTROL R0D WITHDRAWAL ANALYSIS l INITIAL CONTROL R00 PATTERN

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ANF-89-02 Revision 1-Page 29 9.0 ADDITIONAL REFERENCES

.9.1 S. F. . Gaines, " Generic Mechanical Design- for Exxon Nuclear Jet Pump BWR Reload Fuel," XN-NF-81-21( A), Revision 1, Exxon Nuclear Company, Inc.,

Richland, WA 99352, January 1982.

P 9.2 R. H. Kelley, " Exxon Nuclear Plant Transient Methodology for . Boiling Water Reactors," XN-NF-79-71(P), Revision 2, Exxon Nuclear Company, Inc.,

Richland, WA 99352, November 1981.

. 9.3 J. E. Krajicek, "WNP-2 Cycle 5 Plant Transient Analysis," ANF-89-01, Revision 1, Advanced Nuclear Fuel s Corporation, Richland, WA 99352, March 1989.

9.4 J. E. Krajicek, "LOCA Break Spectrum for a BWR 5," XN-NF-85-138(P), Exxon Nuclear Company, Inc., Richland, WA 99352, December 1985.

9.5 D. J. Braun, "WNP-2 LOCA-ECCS Analysis, MAPLHGR Results," XN-NF-85-139, Exxon Nuclear Company, Inc., Richland, WA 99352, December 1984.

9.6 M. H. Smith, " Generic Mechanical Design for Exxon tP clear Jet Pump BWR Reload Fuel," XN-NF-81-21(P), Revision 1, Supplement 1, Exxon Nuclear Company, Inc., Richland, WA 99352, March.1985, 9.7 " Exxon Nuclear Methodology for Boiling Water Reactors-Neutronics Methods for Design and Analysis," XN-NF-80-19( A), Volume 1 and Supplements, Exxon Nuclear Company, Inc., Richland, WA 99352,.May 1980.

9.8 " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN-NF-85-67(A), Revision 1, Exxon Nuclear Company, Inc., Richland, WA 99352, September 1986.

9.9 " Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods for Design Analysis," XN-NF-80-19(A), Volume 1, Supplements 1 and 2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.

f 9.10 J. B. Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.

I 9.11 J. E. Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction," XN-NF-87-92 and XN-NF-87-92, Supplement 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, June 1987 and May 1988.

9.12 J. E. Krajicek, "WNP-2 Single Loop Operation Analysis," ANF-87-119, Advanced Nuclear Fuels Corporation, Richland, WA 99352. September 1987.

) i

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I ANF-89-02 Revision 1 Page 30 9.13 J. E. Krajicek and T. Tahvili, "WNP-2 LOCA Analysis For Single Loop I

Operation," ANF-87-ll8, Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1987.

9.14 Letter, R. B. Samworth, USNRC, to G. C. Sorensen, WPPSS,

Subject:

Issuance Of Amendment No. 62 To Facility Operating License No.

NPF-21-WPPSS Nuclear Project 2 (TAC No. 67538), August 5,1988.

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ANF-89-02 Revision 1 Page A-1 APPENDIX A 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLIES (LFAs)

A.1 INTRODUCTION Evaluations have been performed consistent with ANF methodology (" Exxon Nuclear Methodology for Boiling Water Reactors", XN-NF-80-19) to establish a licensing basis for two ANF 9x9-IX and two ANF 9x9-9X Lead Fuel Assemblies (LFA) in the WNP-2 Cycle 5 core. Justification is provided which demonstrates the applicability of the WNP-2 Cycle 5 operating limits to these four LFAs unless stated otherwise.

The insertion of only four ANF 9x9 LFAs in the Cycle 5 core will have negligible effects upon core wide transient performance. However, some 9x9 LFA specific analyses have been performed to assure that the Cycle 5 operating limits are also applicable to the LFAs. Fuel specific LHGR and MAPLHGP, limits have been developed for these LFAs and are presented in this appendix.

A.2 FUEL MECHANICAL DESIGN l

A mechanical design analysis showing that the 9x9-IX and 9x9-9X fuel meet approved criteria will be documented in ANF-89-014(P).

l l The dynamic response of the LFAs is expected to be almost identical to that of the 8x8 already in the core. This is due to the fact that the fuel assembly stiffness is provided by the assembly channel, which is the same in both designs. The mass of the LFAs is very close to that of the 8x8's. It thus follows that the dynamic response should be the same.

I A.3 THERMAL HYDRAULIC DESIGN The 9x9 LFAs are hydraulically compatible with the co-resident ANF 8x8 fuel assemblies based on a comparison of fuel component hydraulic resistances.

l Steady state thermal hydraulic analysis has shown that even though the ANF 9x9 LFA design has a somewhat smaller flow area than the ANF 8x8 design, no reduction in thermal margin is experienced in the Cycle 5 core. This is due

ANF-89-02 Revision 1 Page A-2 to the increased critical power performance of the ANF 9x9 LFA design relative I

to the ANF 8x8 design at WNP-2 Cycle 5 conditions.

A.4 NUCLEAR DESIGN g1 l The average enrichment and enrichment distribution for the 9x9-IX and E '

9x9-9X fuel assemblies have been selected to match, as closely as possible, 3 the neutronic performance of the four 8x8 XN-3 2.64 w/o U-235 reload assemblies included in the Cycle 5 reload. The fuel assembly average I enrichment, including six-inch top and bottom natural uranium blankets, is 2.53 w/o U-235 for the 9x9-IX design and 2.59 w/o U-235 for the 9x9-9X design.

The average enrichment of the 138 inch central portion of the fuel assembly is 2.69 w/o U-235 for the 9x9-IX and 2.75 w/o U-235 for the 9x9-9X. Each 9x9 assembly contains six fuel rods containing Gd 023 blended with 2.51 w/o U-235.

The 9x9 fuel assembly contains 72 fueled rods and one central water channel displacing nine rod positions. The key neutronic design parameters for the ANF 9x9 LFA designs are presented in Table A.1 along with the corresponding values for the ANF XN-3 8x8 reload fuel design.

The nuclear characteristics of the 9x9 LFAs are similar to the Il i characteristics of the ANF 8x8 fuel. The effect of replacing four ANF 8x8 assemblies with the four. ANF 9x9 LFAs on the Cycle 5 core neutronics is negligible. The maximum cold uncontrolled non-voided km of the 9x9 fuel is g

1.215 compared to the maximum km of 1.229 for the XN-3 8x8 fuel; thus the 9x9 m fuel is compatible with the 8x8 fuel for fuel storage in the spent fuel pool.

I!

The LFAs were included in the core-wide stability analysis reported in Section 4.2.4. Local instability tests were performed on 9x9 leads in a BWR-3, ORNL/TM-9054; no detectable difference was noted in stability performance relative to the co-resident 8x8 fuel.

A.5 ANTICIPATED OPERATIONAL OCCURRENCES Analyses of the WNP-2 Cycle 5 limiting transients have been performed for l ANF 8x8, ANF 9x9 LFAs, and GE P8x8R fuel.. It has been shown that using the l XN-3 ANF CHF correlation, the bundle power required to produce transition I!

ANF-89-02 Revision 1 Page A-3 boiling in an ANF 9x9 LFA is higher than that for an ANF 8x8 bundle. That is, when an ANF 9x9 LFA bundle is modeled as an 8x8 bundle with equivalent conditions, there is margin to the MCPR safety limit during all A00s. The Cycle 5 Safety Limit Analysis considered the LFAs such that the MCPR safety limit of 1.06 is also applicable to the LFAs. Therefore, the ANF 9x9 LFAs can be monitored to the ANF 8x8 fuel limits.

A.6 POSTULATED ACCIDENTS Since heatuo is primarily a planar and not an axial phenomena, the l appropriate bundle power limit that is derived from. a LOCA analysis is the peak bundle planar power. The ANF 9x9 LFAs have better cooling during LOCA conditions relative to an ANF 8x8 fuel assembly due to the lower stored energy in the fuel rods, a greater surface area provided by the larger number of fuel rods, and more inert surface from the central water channel. Thus, a LOCA analysis for the ANF 9x9 LFAs would yield lower Peak Cladding Temperatures (PCTs) and metal-water reactions than an ANF 8x8 assembly at the same bundle peak planar power. The MAPLHGR limits for the ANF 9x9 LFAs restrict the peak bundle planar power to that analyzed for the ANF 8x8 fuel and assure that the USNRC criteria are met for the ANF 9x9 LFAs in Cycle 5.

The fuel loading error was analyzed for the ANF 9x9 LFAs. Results show that if the loading error went undetected, the offsite consequences would remain well within the guidelines specified in 10 CFR Part 100.

f A.7 TECHNICAL SPECIFICATIONS All operational limits used for ANF 8x8 fuel are applicable to the ANF 9x9 LFAs except for fuel type specific MAPLHGR limits and the 9x9-IX and 1

9x9-9X LHGR limits. The LHGR limits for the 9x9-IX and 9x9-9X LFAs are shown in Figures A.4 and A.5 respectively, and the MAPLHGR limits for the LFAs are shown in Figure A.6. The numerical values of Figure A.6 are 0.861 (62/72) times the MAPLHGR values of Section 7.2.1. The LFA single-loop operation (SLO) limits are bounded by the two-loop operation limits. l

ANF-89-02 Revision 1 Page A-4 s TABLE A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES 9x9-IX

- Reload XN-3 Except 9x9-9X Fuel Pellets 8x8 Gd Rods and IX Gd Rods Fuel Material UO2 Sintered U02 Sintered UO2 Sintered Pellets Pellets Pellets Density a g/cc 10.36 10.55 10.36 94.5 g  % of TD 94.5 96.26 Diameter, ' inch

" Enriched Fuel 0.4055 0.3740 0.3665 .

E 0.3740 0.3665 W Natural Fuel 0.4045

_F,uel Rods Fuel. Length, inch 150 150 150 Cladding Material Zircaloy-2 Zircal oy-2 Zi rcal oy-2 Cladding Liner Material N/A Zirconium N/A Clad I.D., inch 0.414 0.381 0.373 Clad 0.D., inch 0.484 0.431 0.431 I

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l ANF-89-02

. Revision 1 Page A-5 l 1

~

TABLE A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES (CONTINUED) 9x9-IX Reload XN-3 Except 9x9-9X Water Rods 8x0 Gd Rods and IX Gd Rods 0 Number 2 N/A N/A Cladding I.D., inch 0.414 N/A N/A l

Cladding 0.D., inch 0.484 N/A N/A l Central Water Channel I:

Outside Width, inch N/A 1.65 1.65 Thickness, inch N/A 0.0285 0.0285 l Fuel Assembly Data Number of Fuel Rods 62 72 72 Fuel Rod Enrichment Figure A.1 Figure A.2 Figure A.3 Fuel Rod Pitch, inch 0.641 0.569 0.569 )

Fuel Assembly Loading, KgU 176.0 176.8 167.7 I

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ANF-89 02 Revision 1 Page A-6 i

  • i LL  : L  : ML  :

M  : M

'ML  :

L  : LL  :

L  : ML

H
ML* : H  :

H  : M

L  :
  • ML  : H  : H  : H
H  : H  : ML*  : ML  :

M  : ML*  : H  : W  : M

H H  : M  :
  • M : H : H : M . . .
W  : H  : M  : M  :
  • ML  : H  : H  : H  :

H  : H  : H  : H  :

I

L  : M  : ML* :  :

H

M  : H  : ML*  : ML  :

I LL  : L  : ML  : M  : M  : M  : ML  :

:  : L  :

I LL RODS ( 3)

L RODS ( 7) 1.50 W/0 U235 1.94 W/0 U235 ML RODS ( 9)

I 2.50 W/0 U235 M RODS (16) ---

2.86 W/0 U235 H RODS (22) ---

3.43 W/0 U235 ML* RODS ( 5) ---

2.50 W/0 U235 + 2.00 W/0 GD203 W ' RODS ( 2) ---

INERT WATER R0D I

FIGURE A.1 XN-3 8X8 ENRICHED ZONE ENRICHMENT DISTRIBUTION I

I -

. 1, j

ANF-89-02

  • * * * * * * * * * * * * * * * * * * * * * * * * * * * *
  • Re v i s Page i o nA-7 1 !1 l
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M 1.92 W/0 U235 I l

R005 (12) --

2.51 W/0 U235 H RODS (50) ---

2.82 W/0 U235 M*1 RODS ( 5) -..

2.51 W/0 U235 + 1.80 W/0 GD203 M*2 RODS ( 1) ---

2.51 W/0 U235 + 4.50 W/0 G0203 W RODS ( 9) ---

INERT WATER R00 Il FIGURE A.2 9X9-IX ENRICHED ZONE ENRICHMENT DISTRIBUTION 1 I

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ANF-89-02 Revision 1

                                                          • Page A.8  !
L  : M  : H  : H  : H  : H  : H  : M
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L RODS ( 4) ---

1.92 W/0 U235 M RODS (12) ---

2.51 W/0 U235 H RODS (50) ---

2.90 W/0 U235 M*1 RODS ( 5) ---

2.51 W/0 U235 + 1.80 W/0 GD203 M*2 RODS ( 1) ---

2.51 W/0 UP35 + 4.50 W/0 GD203 W RODS ( 9) ---

INERT WATER R0D FIGURE A.3 9X9-9X ENRICHED ZONE ENRICHMENT DISTRIBUTION I -

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I-ANF-89-02 Revision 1 Issue Date: 3/10/89 I l

W WNP-2 CYCLE 5 RELOAD ANALYSIS I

Distribution:

0. C. Brown R. E. Collingham I- R. A. Copeland W. S. Dunnivant L. J. Federico I S. J. Haynes J. G. Ingham S. E. Jensen I D. C. Kilian/R. B. Stout J. E. Krajicek S. L. Leonard I J. L. Maryott L. A. Nielsen A. Reparaz R. S. Reynolds I G. L. Ritter H. E. Williamson I Y. U. Fresk/WPPSS (51)

Document Control (5)

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