ML17285A301

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Washington Nuclear Plant-2 Cycle 5 Reload Analysis.
ML17285A301
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 01/31/1989
From: Krajicek J
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17285A298 List:
References
ANF-89-02, ANF-89-2, TAC-72251, NUDOCS 8903090451
Download: ML17285A301 (54)
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ANF-89-02'Ai@VAHC>Do HUCIt.lK>R PQZK8" CORPORATlON WNP-2 CYCLE 5 RELOAD ANALYSIS JANUARY 1989 qOSOS egOSO~noyes, 05 pgg Oui'e~OOOO~>P A Siemens Company 0

ADVANCED NUCLEAR FUELS CORPORATION

'NF-89-,02 Issue Date: 1/17/89 WNP-2 CYCLE 5 RELOAD ANALYSIS Prepared by J..Krajicek BWR safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services January 1989 NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REQARDINQ CONTENTS AND USE OF THIS 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 Commlssfon 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 and it Is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, Information, and belief.The Information con.talned 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.cerning 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 contained in this docu.ment, or that the use of any Information, apparatus, method, or process disclosed in this document will not Infringe privately owned rights, or B.Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any Information, ap.paratus, method.or process disclosed In this document.ANF 3145629A ie88 ANF-89-02 Page i TABLE OF CONTENTS Section Pacae 1.0 2.0 3.0 4.0 , 6.0

7.0 INTRODUCTION

FUEL MECHANICAL DESIGN ANALYSIS.THERMAL HYDRAULIC DESIGN ANALYSIS.3.1 Design Criteria 3.1.3 Fuel Centerline Temperature

......3.2 Hydraulic Characterization

.3.2.5 Bypass Flow..............3.3 MCPR Fuel Cladding Integrity Safety Limit 3.3.1 Coolant Thermodynamic Condition....3.3.2 Design Basis Radial Power Distribution 3.3.3 Design Basis Local Power Distribution

.NUCLEAR DESIGN ANALYSIS.4.1 Fuel Bundle Nuclear Design Analysis 4.2 Core Nuclear Design Analysis.4.2.1 Core Configuration 4.2.2 Core Reactivity Characteristics

.4.2.4 Core Hydrodynamic Stability...~~~~ANTICIPATED OPERATIONAL OCCURRENCES

.5.1 Analysis Of Plant Transients At Increased Core 5.2 Analyses For Reduced Flow Operation 5.3 Analysis For Reduced Power Operation (SLO)5.4 ASME Overpressurization Analysis......5.5 Control Rod Withdrawal Error.5.6 Loading Error for Reload Fuels..5.7 Determination Of Thermal Margins.POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident.6.1.1, Break Location Spectrum 6.1.2 Break Size Spectrum.6.1.3 MAPLHGR Analyses 6.2 Control Rod Drop Accident~~~0~~Flow Conditi~~~~imit.~~~~~~~on Rate Limits TECHNICAL SPECIFICATIONS 7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Integrity Safety L 7.1.2 Steam Dome Pressure Safety Limit 7.2 Limiting Conditions For Operation 7.2.1 Average Planar Linear Heat Generati ANF Bx8 Fuel 7.2.2 Minimum Critical Power Ratio~~~~~~~~~~~~~~~~~~ons~~~~~~~~~~~~~~~~~~~~~~~~~~~For~~~~~~2I 6 6 7 7 8 8 8 8 11 11 11 11 11 11 12 12 12 12 12 12 12 ANF-89-'02 Page ii 7.2.3 Surveillance Requirements

.7.2.3.1 Scram Insertion Time Surveillance 7.2.3.2 Stability Surveillance

.7.2.3.3 Technical Specification LHGR Surveill 9.0 ADDITIONAL REFERENCES,...

~.APPENDIX A 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLIES (LFA'S)ance 13.14 14 28 A-1 LIST OF TABLES ANF-89-02 Page iii Table 4.1 NEUTRONIC DESIGN VALUES A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES~~~~Pa e~~~~~~15~~~~~A 5 LIST OF FIGURES Ficiurd I 3.1 RADIAL POWER HISTOGRAM FOR I/4 CORE SAFETY LIMIT MODEL 3.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS ,(ANF-4 FUEL)~~3.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL)3.4 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)3.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL)4.1 WNP-2 CYCLE 5 ENRICHED ZONE ENRICHMENT DISTRIBUTION 4.2 WNP-2 CYCLE 5 REFERENCE LOADING PATTERN BY FUEL TYPE (ONE QUARTER OF SYMMETRICAL CORE LOADING)5.1 WNP-2 CYCLE 5 CONTROL ROD WITHDRAWAL ANALYSIS INITIAL CONTROL ROD PATTERN 5.2 REDUCED FLOW MCPR OPERATING LIMIT FOR NORMAL FEEDWATER 5.3 REDUCED FLOW MCPR OPERATING LIMIT FOR FFTR OPERATION.7.1 LINEAR HEAT GENERATION RATE (LHGR)LIMIT VERSUS AVERAGE EXPOSURE, ANF 8X8 FUEL.A.l XN-3 8X8 ENRICHED ZONE ENRICHMENT DISTRIBUTION

.A.2 9X9-IX ENRICHED ZONE ENRICHMENT DISTRIBUTION

.A.3 9X9-9X ENRICHED ZONE ENRICHMENT DISTRIBUTION

.A.4 LHGR LIMIT FOR 9X9-IX FUEL.A.5 LHGR LIMIT FOR 9X9-9X FUEL.A.6 ANF 9X9-IX AND,9X9-9X MAPLHGR LIMITS.TEMPERATURE PLANAR Pa<ac I 17 18'9 20 21 22 23 24 25 26 27 A-7 A-8 A-9 A-10 A-11 A-12 ANF-89-02 Page

1.0 INTRODUCTION

This 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).WNP-2 is scheduled to commence Cy'cle 5 operation in June 1989.This report is intended to be usedtt tttANFtpt 1 p X~X-NF--N, F1 t, R"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'generic reference list.Section numbers in tlt p<<1 p dt p t.P 1~X-1-Volume 4, Revision'.1.

Final feedwater temperature reduction (FFTR)analysis to support, cycle extension was 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 when reactor operation~~~t j N N's being extended by reduction of the feedwater temperature.

The WNP-2 Cy'cle 5 core will comprise a total of 764 fuel assemblies:

including 140 ANF 8x8 unirradiated assemblies; 2 ANF 9x9-IX unirradiated Lead Fuel Assemblies (LFA);2 ANF 9x9-9X unirradiated LFA's;152 once irradiated, ANF 8x8 assemblies,'48 twice irradiated ANF 8x8 assemblies; 128 thrice irradiated ANF 8x8 assemblies; and 192 irradiated P8x8R assembiles 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.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 operating cycle which encompass core flow up to 106%of the design basis value.

N 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 9.8 ANF-89-02 Page 2 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.

ANF-89-02 Page 3 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS" 3.1,Desi n Criteria I 3.1.3 Fuel Centerline Tem erature'he LHGR curve in Figure 3~4 of, Reference 9.8 shows that the ANF Bx8 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 Bx8 exposures within th'e bounds of the referenced LHGR curves.I I I I 3.2 H draulic Characterization I.I..I~FF Calculated Bypass Flow Fraction (100/power/106/

flow)10.7/.3 MCPR Fuel Claddin Inte rit Safet Limit 3.3.1 Coolant Thermod namic Condition Core Power Core Inlet Enthalpy Steam Dome Pressure Feedwater Temperature 3950 Mwt 525.6 Btu/ibm 1021 psia 414'F 3.3.2 Desi n Basis Radial Power Distribution See Figur e 3.1.3.3.3 Desi n Basis Local Power Distribution See Figures 3.2, 3.3, 3.4 and 3.5.

CII t I~

I 4.0 NUCLEAR DESIGN ANALYSIS ANF-89-02.-Page 4.4.1 Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichment Radial Enrichment Distribution Axial Enrichment Distribution 2.62 w/o U-235 Figure 4: 1 Uniform 2.79 w/o U-235 with 6-inch top and bottom natural uranium blankets Burnable Poisons Figure 4.1 lNon-Fueled Rods Figure 4.1 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 Reload Analysis Report ANF-88-02 and 4 9x9 LFA's of the 2.53/2.59 w/o U-235 design described in Appendix A.4.2'ore Nuclear Desi n Anal sis 4.2.1 Core Confi uration Core Exposure at EOC4 (HWd/MTU)Core Exposure at BOC5 (HWd/HTU)Core Exposure at EOC5 (HWd/HTU)Figure 4.2 16,700 12,300 18,100 4.2.2 Core Reactivit Characteristics BOC Cold k-eff, All Rods Out BOC Cold k-eff, Strongest Rod Out Reactivity Defect (R-Value)Standby Liquid Control System (SBLC)660 ppm Boron, Cold k-eff 1.1133 0.9868 0.0 0.9633 ANF-89-02 Page 5 4.2.4 Core'drod namic Stabilit%Power%Flow State Points 65/45*47/27.6**42/23 8***Deca Ratio COTRAN 0.49 I 0.89 0.82*45 percent flow-APRM Rod Block intercept point.**Two pump minimum'flow

-46 percent power.***Natural circulation flow-APRM Rod Block intercept point.

ANF-89-02 Page 6.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Transient Analysis Report Reference 9.3 5.1 Anal sis Of Plant Transients At Increased'ore 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.

Cycle 5 specific analyses of transient events were performed for two recirculation pump'operation conditions, with the recirculation pump trip (RPT)in service and out of service, and for two scram conditions which are normal scram speed (NSS)and technical specification scram speed (TSSS).Analyses were performed at end-of-cycle exposures which produced the results shown in following table.Generic analyses were performed for FFTR to extend cycle operation (Reference 9.11).The loss of feedwater heatirfg event was analyzed on a plant specific bounding value basis and the.-delta CPR results are bounding values for WNP-2.

ANF-89-02 Page 7 Transient*

LRNB, NSS RPT Operable Maximum Delta CPR%Power/Maximum Maximum Pressure GE ANF%Flow Heat Flux%Power%~si Fuel Fuel 104/106 121, 403 1169 0.28 0.25 LRNB, NSS RPT Inoperable LRNB, TSSS RPT Operable LRNB, TSSS RPT Inoperable FWCF, NSS RPT Operable LOFH 104/106 104/106 104/106 47/106 N/A 127 127 132 N/A 501 594 163 N/A 1181 1174 1189 1026 N/A 0.35 0.31 0.35 0.31 0.41 0.35/0.23 0.20 0.09 0.09 5.2 Anal ses For Reduced Flow 0 eration Limiting Transient:

Recirculation Flow Increase References 9.3 and 9.11 5.3 Anal sis For Reduced Power 0 eration 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 transient events, the pump seizure accident and the loss-of-coolant-accident (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 measured plant scram insertion data, se Section 7.2.'3.1.

ANF-89-02 Page 8 I SLO MCPR Operating Limit for ANF and GE fuel 1.35 Two-loop HAPLHGR limits which are shown in Section 6.1.3 for ANF.fuel apply during 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 ASHE Over ressurization Anal sis'imiting Event ,Worst Single Failure Maximum Pressure Haximum Steam'ome Pressure References 9,3 and 9.11 MSIV Closure MSIV Position Scram Trip 1315 psig, 1286 psig 5.5 Control Rod Withdrawal Error Initial Control Rod Pattern for CRWE Analysis Figure 5.1 Rod Block onitor Settin 106%107%*108%Distance" Withdrawn (ft)5.5 6.0 7.0 ANF Fuel Delta-CPR 0.17 0.19'.21 GE Fuel Delta-CPR 0.17 0.19 0.21 5.6 Loadin Error for Reload Fuels Maximum LHGR, kW/ft Minimum MCPR With Loadin Error 16.6 1.26 Correctly Loaded Core 13.3 1.44 5.7 Determination Of Thermal Har ins Summary of Thermal Margin Requirements Rod Block Monitor Setting (RBH)of 107%.

ANF-89-02 Page 9 All system transient results were analyzed at the more limitin 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%).Event r Equipment 0 erational Status GE ANF Fuel Fuel GE.ANF Fuel Fuel Delta CPR MCPR Limit Model LRNB LRNB~"'RNB LRNB FWCF LOFH RPT Openable, NSS RPT Inoperable, NSS RPT Operable, TSSS RPT Inoperable, TSSS RPT Operable, NSS N/A 0.35 0.31 1.41 1.37 It II 0;35 0.31 1.41 1.37 0.41 0.35 1.47 1.41 0.23 0.20 1.29 1.26 0,09 0.09 1.15 1.15 II II XTGBWR 0.28 0~25 1.34 1.31 COTRANSA/XCOBRA-T 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 EOC-2000 MWd/MTU are based on the CRWE (100%To 106%Flow)4,~Fuel T e ANF GE MCPR Limit 107%RBM 1.25 1.25 ANF-89-02 Page 10 MCPR Operating Limits At Rated Condition From EOC-2000 MWd/MTU To EOC (100%To 106%Flow)With, Normal Feedwater Temperature

~Fuel T e ANF GE MCPR Limit 1.31 1.34 e MCPR Operating Limits At Rated Condition Beyond All Rods Out With Reduced Feedwater Temperature (100%To 106%Flow And Thermal Coastdown)

Point (EOC5)~Fuel T e ANF GE MCPR'Limit 1.33 1.36 MCPR Limits at Off-Rated Conditions Figures 5.2 and 5.3 Reduced Flow MCPR Limit References 9.3 and 9.11 0

d ANF-89-02 Page 11.0 POSTULATED ACCIDENTS d 6.1 Loss-Of-Coolant Accident 6.1.1'Break Location S ectrum Reference 9.4 6.1.2 Break Size S ectrum Reference 9.4 6.1.3 MAPLHGR Anal ses (ANF Fuel-Two-Loop Operation and SLO)References 9.5, 9.13 and 9.14 Limiting Break: Split'reak in the Recirculation Suction Piping With an Area Equal to Sixty Percent of the Double-Ended Cross-Sectional Pipe Area Bundle Average Exposure~d 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 MAPLHGR~kW ft 13.0 13.0,, 13.0 13.0'3.0 11.3 9.4 7.9 Peak'Clad*

Tem erature'F 1779 1755 1761 1765 1771 1659 1513 1385 Peak Local*MWR.0.50 0.4'5 0.46 0.46 0,51 0.32 0.16 0.09 Heatup analysis shows insignificant changes in PCT's and local MWR, but no change in MAPLHGR limits, from the MAPLHGR analysis for the earlier ANF Bx8 fuel desig'n which is shown in Reference 9.5.6.2 Control Rod Dro Accident Dropped Control Rod Worth, mK Doppler Coefficient dk/kdT, 1/'F Effective Delayed Neutron Fraction Four-Bundle Local Peaking Factor Maximum Deposited Fuel Rod Enthalpy (cal/gm)Reference 9.7 8.1-10.0 x 10 6 0.0050 1.17 121 For the ANF-4(6Gd2) fuel design PCT's and MWR's, I

ANF-89-02 Page 12.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit HCPR Safety Limit 1.06'.1.2 Steam Oome Pressure Safet Limit Pressure Safety Limit 1346 psig 7.2 Limitin Conditions For'eration 7.2.1 Avera e Planar Linear Heat Generation Rate Limits For ANF 8x8 FuelBundle Average Exposure MWd MTU 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 HAPLHGR~kM ft 13.0 13.0 13.0 13.0 13.0 11.3 9.4 7.9 For single-loop operation these limits also apply to ANF Fuel when using a SLO HCPR limit of at least 1.35.7.2.2 Minimum Critical Power Ratio Rated Condition MCPR Operating Limit Up To EOC-2000 MWd/HTU Exposure (100%To 106%Flow)~Fuel T e ANF GE Limit 107%RBM 1.25 1.25 ANF-89-02 Page 13 Rated Conditions HCPR Operating Limits From EOC-2000 HWd/HTU To EOC (100%To 106%Flow)~Fuel T e ANF'E, Limit 1;31 1.34 Thermal Coastdown and FFTR Rated Condition MCPR Operating Limit Beyond All Rods Out Point With Reduced Feedwater Temperature (100%to 106%Flow)~Fuel T e ANF GE I~Limit 1.33 1.36 Reduced Flow HCPR Limit (all cycle exposures)

Figures 5.2 and 5.3~Single-Loop Op'eration (SLO)HCPR Limit (all cycle'exposur'es)

~Fuel T e Limit ANF 1.35 GE 1.357.2.3 Surveillance Re uirements 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 values which correspond to the technical specification (TSSS)control rod scram times (see Section 5.7).

.Position Inserted From Full Withdrawn Average Rod Time In Seconds As Defined In Footnote*ANF-89-02'Page 14 Notch 45 Notch 39 Notch 25 Notch 5 0.404 0.660 1.504 2.624 7.2.3.2 Stabilit Surveillance Core hydrodynamic stability analyses require.slight modification to 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).1 Surveillance requirements remain unchanged for Cycle 5, e.g., surveil-lance is required when operating in a power flow region above the 80%rod line and less than 45%core flow.~~~~.2.3.3 Technical S ecification 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.'his figure was developed from information contained in Reference 9.1, and the region of permissible Operation is shown.S lowest measured average control rod insertion time to specified notches for ach group of four control rods arranged in a 2x2 array.

TABLE 4.1 NEUTRONIC DESIGN VALUES ANF-89-02 Page 15 4 Fuel Pellet Fuel Material Density, g/cc%of T.D.Diameter, inch Enriched Fuel Natural Fuel U02 Sintered Pellets 10.36 94.5.0.4055 0.4045 Fuel Rod Fuel Length, inch Cladding Material Clad, I.D., inch Clad, O.D., inch 150 Zircaloy-2 0.414 0.484 Fuel Assembl Number of Fuel Rods Number of Inert Water Rods Fuel Rod Enrichments Fuel Rod Pitch, inch Fuel Assembly Loading, kgU 62 Figure 4.1 0.641 176.0 TABLE 4.1 NEUTRONIC DESIGN VALUES (Continued)

ANF-89-02 Page 16 Core Data Number of Fuel Assemblies Rated Thermal Power, MW Rated Core Flow, Mlbm/hr Core In)et Subcooling, Btu/ibm Reactor Pressure, psia Channel Thickness, inch Fuel Assembly Pitch, inch'I Water Gap Thickness (symmetric), inch 764 3323 108.5 19,0 008.0 0.100 6.00 0.522 ontrol Rod Data Absorber Material Total Blade Span, inch Total Blade Support Span, inch Blade Thickness, inch Blade Face-To-Face Internal Dimension, inch Absorber Rods Per Blade Absorber Rod Outside Diameter, inch Absorber Rod Inside Diameter, inch Absorber Density,%of Theoretical B4C 9.75 1.58 0.260 0.200 76 0.188 0.138 70 100'~~60 CQ<0 20 0 0 0.25 0.50 0.75 1 1.25 1.50 BUNDLE POWER FACTOR FIGURE 3.1 RADIAL POWER HIS AN FOR I/O CORE SAFETY LINIT NODEL ANF-89-02 Page 18*~.936:.977: 1.023: 1.015: 1;011: 1.041: 1.076: 1.052*~4*:.977: 1.011:.907: 1.042: I*~1.035:.932:.962: 1.075*~*: 1.023:.907: 1.017:.988:.974:.996:.931: 1.040-%l*~,~'*+*~*: 1.015: 1.042:.988:.000:.850:.972: 1.033: 1.009 1.011: 1.035:.,974:.850:.000:.985: 1.038: 1.011*~4*~1.041:.932:.996:.972:.985: 1.012:.901: 1.043 1.076:.962:.931: 1.033: 1.038:.901:~976: 1.078: 1.052: 1.075: 1.040: 1.009: 1.011: 1.043: 1.078: 1.054 FIGURE 3.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)

ANF-89-02.t****************************************************

  • ~*:.944:.962: 1.011: 1.044: 1.043: 1.010:.960:.943*~~r*~I*:.962:.980: 1.064:.893: 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*~~~*~*: 1.010: 1.059: 1.002:.980:.988;1.004: 1.060: 1.065:*~:-.960: 1.034:.915: 1.032:.952: 1.060:.966: 1.053.943:.961: 1.010: 1.042: 1.041: 1.065: 1.053: 1.019: FIGURE 3.3 WNP-2 CYCLE 5 SAFETY LIHIT LOCAL PEAKING FACTORS (ANF XN-3 fUEL)

ANF-89-02 Page 20****************************************************

    • ~*:.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:.000:.935:.996: 1.033: 1.027*~**~*~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:.950:.963: 1.000: 1.027: 1.027: 1.042: 1.029: 1.003 FIGURE 3.4 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)

ANF-89-02 0*~*:.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:'~*: 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: 1.012: 1.000: 1.006: 1.017: 1.047: 1.032.968: 1.013:.944: 1.030:.976: 1.047:.975: 1.020:.966:.968:.997: 1.019: 1.020: 1.032: 1.020: 1.003 FIGURE 3.5 WNP-2 CYCLE 5 SAFETY LIHIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL)0

  • LL*~**~*L*~~k**~L M ANF-89-02 Page 22 ML*H H: M: ML*: t<L ML: M: t I: M: ML: L*~ML ML**~*H: H: H H: M M*~*: M*t*H': W: M H: H: M*~*~*~*H H M: W H'~*o*~*M H: H: H H: ML+;MML: ML*M H H ML*: ML: ML ML M M M tlL: 'LL RODS (1)L RODS (5)ML RODS (9)M RODS (21)H RODS (20)ML*RODS (6)W RODS (2)1.50 W/0 U235 2',00 W/0 U235 2.50 W/0 U235 2.64 W/0 U235 3,43 W/0 U235 2.50 W/0 U235+2.00 W/0 GD203 INERT WATER RODFIGURE 4.1 WNP-2 CYCLE 5 ENRICHED ZONE ENRICHMENT DISTRIBUTION ANF-89-02 Page 23 1 2'4 5 6 7 8 9 10 11'2 13 14 15'2 H, I C f, C C H B , B B, B C 10 1 H B'2 0 H" 0 13 14 Fuel Type Type Number of Assemblies Description 56 136 128 148 24 128 136 4 GE SxS Type II 1.76 w/o U-235 (Cycle 1)GE SxS Type II I 2.19 w/o U-235 (Cycle 1)ANF SxS 2.72 w/o U-235 (Cycle 2)ANF SxS 2.72 w/o U-235 (Cycle 3)ANF SxS 2.72 w/o U-235 (Cycle 4)ANF SxS 2.64 w/o U-235 (Cycle 4)ANF Sx8 2.64 w/o U-235 (Cycle 5)ANF SxS 2.62 w/o U-235 (Cycle 5)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 (ONE QUARTER OF SYMMETRICAL CORE LOADING) 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 ANF-89-02 Page 24 I 59 51 00--40 00 55 51 47 43 36--16--00--16--36 47 43 39--'0 35 31--40 16 00 00--36 36--00*--00--16--00--39 35 36--00--40--31 27 23--00 19>>16 36 00--36 16--00 00-'-,16--00 16--36 27 23 19 15~----00--40--00 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 Full OutFIGURE 5.1 WNP-2 CYCLE 5 CONTROL ROD WITHDRAWAL ANALYSIS INITIAL CONTROL ROD PATTERN 1.6 1.5 NOTE: The MCPR operating limit shall be the greater of the rated condition MCPR operating limit or the value for reduced flow'rom this curve 1.4-M 1.3 C4 O C4 1.2 20 30 40 50 60'0 80 90 100 110'TOTAL CORE RECIRCULA G FLOW (%%uRATED) fIGURE 5.2 REDUCED FLOW HCPR OPERATING FOR NORHAL FEEDWATER TEHPERATURE ll Ql I CO fD I 1.6 NOTE: The HCPR operating limit shall be the maximum of this curve or the rated condition MCPR operating limit.30 40 50 60 70 80 90=100 TQTFIL CQRE BEC I BCULRT ING t LQH-(/BATEO)FIGURE 5.3 REDUCED FLOW NCPR OPERATING LIMIT FOR FFTR OPERATION

~~~~~~~'I~~\: PERMISSIBLE REGION OF OPERAIION 10000 20000 30000.40000 Average Planar Exposure (MWD/MT)60000 JJgaR 0 16 62 610 lb.62 2,680 1'.lO 6.230 14.71 7,940 10.19.10,470 14.13 13,220 14.06 16,990 14.06 18.780 14.OO 21,690 13.93 24,420 13.93 27.280 13.08 30.160 12.24, 33.060 11.40 3b,960 10.47 38.900'.{){)41.830 8.6{)44.760 7.77 FIGURE 7.1 LINEAR HEAT GENERATION RATE (LHGR)LIMIT VERSUS AVERAGE PLANAR SURE, ANF 8X8 FUEL ADDITIONAL REFERENCES ANF-89-02 Page 28 9.1 9.2 9.3 9.4 9.5 9.6S.F.Gaines,"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Rl d F I," X~IPNP-I-IA, R 11 I, E N I C p y, Richland, WA 99352, January 1982.R.H.Kelley,"Exxon Nuclear Plant Transient Methodology for Boiling N,"X~F->>->>,R I I 2,E N I C p Richland, WA 99352, November 1981.J.E.Krajicek,"WNP-2 Cycle 5 Plant Transient Analysis," ANF-89-01, Advanced Nuclear Fuels Corporation, Richland, WA 99352, January 1989.d.t.2'I k,"LOCA R k I'IIR d,"~X-F-->>P, I Nuclear Company, Inc., Richland, WA 99352, December 1985, D.J.Braun,"WNP-2 LOCA-ECCS.

Analysis, HAPLHGR Results," XN-NF-85-139, Exxon Nuclear Company, Inc., Richland, WA 99352, December 1984.H.H.Smith,"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Rl dF I,"X~,R II I,Xppl tl,E II Company, Inc., Richland, WA 99352, March 1985.V"Exxon Nuclear Methodology for Boiling Water Reactors-Neutronics Methods f P lg dA Iyi,"~XN-Ny-

-,Vl I dX pl'.,E Nuclear Company, Inc., Richland, WA 99352, Hay 1980.9.8"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," R I I I, E N I p , I., Itl 11 d, N 99352, September 1986.9.9"Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods f R lg A lyl,"~X--->>, Vl I, Xppl<<I d 2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.9.10 J.B.Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.9.11 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 1988.9'2 J.E.Krajicek,"WNP-2 Single Loop Operation Analysis," ANF-87-119, Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1987.9.13 J.E.Krajicek and T.Tahvili,"WNP-2 LOCA Analysis For Single Loop Operation," ANF-87-118, Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1987.

ANF-89-02 Page" 29 9.14 Letter, R.'.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.

ANF-89-02 Page A-1 APPENDIX A 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLIES (LFA'S)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 LFA's unless stated otherwise.

The insertion of only four ANF 9x9 LFA's in the Cycle 5 core will have~~~~~~~egligible effects upon core wide transient performance.

However, some 9x9 FA specific analyses have been performed to assure that the Cycle 5 operating limits are also applicable to the LFA's.Fuel specific LHGR and MAPLHGR limits have been developed for these LFA's and are presented in this appendix.A.2 FUEL MECHANICAL DESIGN A mechanical design analysis showing that the 9x9-IX and 9x9-9X fuel meet approved criteria will be documented in ANF-89-014(P).

P~Op The dynamic response of the LFA's i's 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 LFA's is very close to that of the 8x8's.It thus follows that the dynamic response should be the same.A.3 THERMAL HYDRAULIC DESIGN The 9x9 LFA's are hydraulically compatible with the co-resident ANF 8x8 uel assemblies based on a comparison of fuel component hydraulic resistances.

ANF-89-02 Page A-2 Steady state thermal hydraulic analysis has shown that even though the ANF 9x~LFA design has a somewhat smaller flow area than the ANF SxS design, no reduction in thermal margin is experienced in the Cycle 5 core.This is due to the increased critical power performance of the ANF 9x9 LFA design relative to the ANF Sx8 design at WNP-2 Cycle 5 conditions.

A.4 NUCLEAR DESIGN The average enrichment and enrichment distribution for the 9x9-IX and 9x9-9X fuel assemblies have been selected to match, as closely as possible, the neutronic performance of the four SxS XN-3 2.64 w/o U-235 reload assemblies included in the Cycle 5 reload.The fuel assembly'verage 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 Gd203 blended with 2.51 w/o U-235.The 9x9 fuel assembly contains 72 fueled rods and one central water channe displacing nine rod positions.

The key neutronic design parameters for th ANF 9x9 LFA designs are presented in Table A.1 along with the corresponding values for the ANF XN-3 SxS reload fuel design.The nuclear characteristics of the 9x9 LFA's are similar to the characteristics of the ANF SxS fuel,.The effect of replacing four ANF Sx8 assemblies with the four ANF 9x9 LFA's on the Cycle 5 core neutronics is negligible.

The maximum cold uncontrolled non-voided km of the.9x9 fuel is 1.215 compared to the maximum k~of 1.229 for the XN-3 8x8 fuel;thus the 9x9 fuel is compatible with the SxS fuel for fuel storage.The LFA's 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;no detectable difference was noted in stability performance relative to the co-resident SxS fuel.

ANF-89-02 Page A-3.5 ANTICIPATED OPERATIONAL OCCURRENCES Analyses of the WNP-2 Cycle 5 limiting transients have been performed for ANF 8x8, ANF 9x9 LFA's, and" GE P8x8R fuel.It has,been shown that using the XN-3 ANF CHF correlation, the bundle power required to produce transition boiling in an ANF 9x9 LFA is higher than that for an ANF 8x8 bundle.That is, when an'NF 9x9 LFA bundle*is modeled as an 8x8 bundle with equivalent condition's, there is margin to the HCPR safety limit during all AOO's.The Cycle 5 Safety Limit Analysis considered the LFA's such that the HCPR safety limit of 1.06 is also applicable to the LFA's.Therefore, the ANF 9x9 LFA's can be monitored to the ANF 8x8 fuel'limits.A.6 POSTULATED ACCIDENTS Since heatup.is primarily a planar and not an axial phenomena, the I appropriate bundle power limit is, derived from a LOCA analysis is the peak bundle planar power.The ANF 9x9 LFA's have'better cooling during LOCA conditions relative to an ANF 8x8 fuel a'ssembly due to the lower stored energy~~~~~'n the fuel rods, a greater surface area provided by the larger number of uel rods, and more inert surface from the central water channel.Thus, a LOCA analysis for the ANF 9x9 LFA's would yield lower Peak Cladding Temperatures (PCT's)and metal-water reactions than an ANF 8x8 assembly at the same bundle peak planar power.The HAPLHGR limits for the ANF 9x9 LFA's 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 LFA's in Cycle 5.The fuel loading error was analyzed for the ANF 9x9 LFA's.Results show that if the loading error went undetected, the offsite consequences would remain well within the guidelines specified in 10 CFR Part 100.A.7 TECHNICAL SPECIFICATIONS All operational limits used for ANF 8x8 fuel are applicable to the ANF 9x9 LFA's except for fuel type specific HAPLHGR limits and the 9x9-IX and 9x9-9X LHGR limits.The LHGR limits for the 9x9-IX and 9x9-9X LFA's are shown in Figures A.4 and A.5 respectively, and the HAPLHGR limits for the LFA's are~~hown in Figure A.6.The numerical values of Figure A.6 are 0.861 (62/72)

ANF-89-02 Page A-4 times the HAPLHGR values of Section 7.2.1.The LFA single-loop operation (SLO)limits are bounded by the two-loop operation limits..

ANF-89-02 Page A-5 TABLE A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES Fuel Pellets Fuel Material Reload XN-3 8x8 U02 Sintered Pellets 9x9-IX Except 9x9-9X Gd Rods and IX Gd Rods U02 Sintered UOp Sintered Pellets Pellets Density g/cc%of TD Diameter, inch Enriched Fuel Natural Fuel Fuel Rods Fuel Length, inch 10.36 94.5 0.4055 0;4045 150 10.55'6.26 0.3740 0.3740 150 10.36 94.5 0.3665 0.3665 150 Cladding Material~~ladding Liner Material Clad I.D., inch Clad O.D., inch Zircaloy-2 N/A 0.414 0.484 Zirconium 0.3807 0.431 N/A 0.373 0.431 Zircaloy-2 Zircaloy-2 ANF-89-02 Page A-6"TABLE A.1 ANF 9X9-IX AND 9X9-9X LEAD FUEL ASSEMBLY NEUTRONIC DESIGN VALUES (CONTINUED)

Water Rods Number Cladding I.D., inch Cladding O.D., inch Central Water Channel Outsi'de Width, inch Thickness, inch Fuel Assembl Data Reload XN-3 8x8 2 0.414 0.484 N/A N/A 9x9-IX Except Gd Rods N/A N/A N/A 1.65 0.0285 9x9-9X and IX Gd Rods N/A N/A N/A , 0.0285 Number of Fuel Rods Fuel Rod Enrichment Fuel Rod Pitch, inch Fuel Assembly Loading, KgU 62 Figure A.I 0.641 176.0 72 72 Figure A.2 Figure 0.569 0.569 176.8 167.7 ANF-89-02 Page A-7*: LL: L: t1L**M: M: ML: L LL*L: ML: H: ML*H';H: M ML H: H: H: H H: ML*: flL**M ML*H W tl H: H~M M H.;H: M WML~~H'H: H: H H: M~~~t1: ML*:.H: M: H: thL*: tlL LL L: t1L: M: t1: t3: t1L LL RODS L RODS ML RODS tl RODS H RODS ML*RODS W RODS (3)(7)(9)(I6)(22)(5)(2)1,50 W/0 U235 1.94 W/0 U235 2.50 W/0 U235 2.86 W/0 U235 3.43 W/0-U235 2,5Q W/0 0235+2.QQ W/0 GD203 INERT WATER ROD FIGURE A.l XN-3 8X8 ENRICHED ZONE ENRICHMENT DISTRIBUTION

  • ~'*~M: H: H H': H ANF-89-02 Page A-8.~M~L~*~*~~~'k~*M H: H: M*1 H: H: H: H: M*~*~*H: H: M*1: H~~t1: H: M*1: H~I H*~*~*~**~*~*H th*1: H H: M W W W: W: H: H W: W: t1: H H H*~*~*~*H: H: H: W W: W~~H H~*4*~H H,: M*1: H M: H M*2: H M H: H;H H: H: H H'M L M: H: H H: H: H: M L RODS (4)M RODS (12)H RODS (50)M*1 RODS (5)M*2 RODS (1)W RODS 9)1.92 W/0 U235 2.51 W/0 U235 2.82 W/0 U235 2.51 W/0 U235+1.80 W/0 GD203 2 51 W/0 U235+4 50 W/0 GD203 INERT WATER ROD FIGURE A.2 9X9-IX ENRICHED ZONE ENRICHMENT DISTRIBUTION
            • +***********+*********ANF-89-02 Page A-9~~L: M: H: H: H: H: H M: L H: H:, M*1: H H: H: H: M H: H:, M*1: H M: H: M*1: H H H: M*1: H: W~I~~W: W: H: H~I H H: H: M: W~~W: W: M: H H r~,H: H H': W W: W H: H H.H H: M*1: H: M H: M*2 H tl: H: H: H H H: H: H M L: M: H: H: H: H: H L L RODS (4)M RODS (12)H RODS (50)I'l*l RODS (5)M*2 RODS (1}W RODS (9)1.92 W/0 U235 2.51 W/0 U235 2.90 W/0 U235 2.51 W/0 U235+1.80 W/0 GD203 2,51 W/0 U235+4.50 W/0 GD203 INERT WATER ROD FIGURE A.3 9X9-9X ENRICHED ZONE ENRICHMENT DISTRIBUTION 13.7 15.0, 13.7 CU I-Q QO CC C9 Ul CO 70, 5.5io 50 20 30 40 PLANAR EXP E.GHD/MTU FIGURE A.4 LHGR LIHIT FOR 9X9-IX FUEL 60 UW fV R EA ll (D CO I I 70 ore 13.I Q Gl I-Q hC CD IX CQ Ol CO 15.5, 13.1 0 10 20 30 40-50 PLANAR EXPOSVRE, GHD/HTV FIGURE A.5 LHGR LIHIT FOR 9X9-9X FUEL 5 0 I 5 10 15 20 ASSEMBLY AVERAGE BURNUP, GWD/NTt'l 30 35 FIGURE A.6 AHF 9X9-ND 9X9-9X HAPLHGR LINITS ANF-89-02 Issue Date: 1/17/89 WNP-2 CYCLE 5 RELOAD ANALYSIS Distribution:

0.C.R.E.R.A.W.S.L.J.S.J.J.G.S.E.D.C.J.E.S.L.J.L, L.A.A.Re R.S.G.L.R.B.H.E.Brown Collingham Copeland Dunnivant Federico Haynes Ingham Jensen Kil i an Krajicek Leonard Haryott Nielsen paraz Reynolds Ritter Stout Williamson Y.U.Fresk/WPPSS (50)Document Control (5)

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