ML20245E007

Cycle 4 Reload Analysis Design & Safety Analyses

ML20245E007
Person / Time
Site:	Susquehanna
Issue date:	05/31/1989
From:	Jensen S, Lotz T, Roberst C PENNSYLVANIA POWER & LIGHT CO., SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17156B218	List: ML20245D982 ML20245E007
References
ANF-89-058, ANF-89-58, NUDOCS 8906270299
Download: ML20245E007 (38)

Text

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DVANCEDNUCLEARFUELS CORPORATION 1

l SUSQUEHANNA UNIT 2 CYCLE 4 RELOAD ANALYSIS 1- DESIGN AND SAFETY ANALYSES l

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MAY 1989 I

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V ADVANCEDNUCLEARFUELS CORPORATION ANF-89-058 1

1 Issue Date: 5/11/89  !

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SUSQUEHANNA UNIT 2 CYCLE 4 RELOAD ANALYSIS Design and Safety Analyses

).

Prepared by:

f W 5 '

1 f . / S. E. Jensen' .

BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services <

)

A

+ <- _

/ T[E'tpff BWR Neutronics i

Neutronics and fuel Management Fuel Engineering and Technical Services I

C. 1. Rebirts, Jr.

BWR Safety Analysis /

Licensing and Safety Engineering c Fuel Engineering and Technical Services f-May 1989 L

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CUSTOMER DISCLAIMER I! j IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT )

PLEASE READ CAREFULLY J

Advanced Nucteer Fuels Corporaton's warranties and representatens con.

commg the subject matter of the document are those set forth in the Agreement -

)

between Advanced Nuclear Fuets Corporation and the Customer pursuant tte -

whsch the document e eeued. Accordmgfy, except as omerwee expressly pro- l

~

vided in such Agrooment. neither Advanced Nuclear Fuels Cor:craten nor any person acting on its behalf makes any warranty or representation. expressed or .

emplied, with respect to the accuracy, completeness, or usefulness of the infor- I meten contamed in this document, of that the use of any informaton, apparatus. '

method or procese dieciosed in this document wdl not intnnge pnvately owned ,

rights: or assumes any liabdites with respect to the use of any information, ap-paratus, method or procese disclosed in mis document.

The informesson antamed herem is for the solo use of Customer, in order to avoid impearment of nghts of Advanced Nuclear Fuets Corporaten in patents or mventons whkA frdy be included in the informaten contained in this documert me reccent, by its acceptance of med document, agrees not to publien or make public use (in the patent use of the term) of such informaton untd so authonzod in wnting by Advanced Nuclear Fuels Corporaten or untd after six (6) months following tennsnaten or expiration of the aforesag! Agreement and any extensen moreof, unless otherwee expressly previoed in me Agreement. No nghts or licerees in or to any patents are emphed by the fumisning of this docu-

$ ment.

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I ANF-3145 472A (12/87) 4

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ANF-89-058

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t TABLE OF CONTENTS

) Section Paae i

1.0 INTRODUCTION

............................ 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . 2 ,

3.0 THERMAL HYDRAULIC DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . 3 F 3.2 Hydraulic Characterization . . . . . . . . . . . . . . . . . . . 3 3.2.1 Hydraulic Compatibility . . . . . . . . . . . . . . . . . 3 3.2.2 Thermal Margin Performance ............... 3 3.2.3 Fuel Centerline Temperature . . . . . . . . . . . . . . . 3  :

3.2.5 Bypass Flow . . . . . . . . . . . . . . . . . . . . . . . 3 3.3 MCPR Fuel Cladding Integrity Safety Limit ........... 3 3.3.1 Coolant Thermodynamic Condition . . . . . . . . . . . . . 3 3.3.2 Design Basis Radial Power Distribution ......... 3 3.3.3 Design Basis Local Power Distribution . . . . . . . . . . 3 j 4.0 NUCLEAR DESIGN ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . 4 4.1 Fuel Bundle Nuclear Design Analysis .............. 4 4.2 Core Nuclear Design Analysis . . . . . . . . . . . . . . . . . . 4 Core Configuration 4.2.1 ................... 4 4.2.2 Core Reactivity Characteristics . . . . . . . . . . . . . 4 4.2.4 Core Reactivity Stability . . . . . . . . . . . . . . . . 5 5.0 ANTICIPATED OPERATIONAL OCCURRENCES . . . . . . . . . . . . . . . . . 6 5.1 Analysis Of Plant Transients At Rated Conditions . . . . . . . . 6 5.2 Analyses For Reduced Flow Operation .............. 6 3

5.3 Analyses For Reduced Power Operation . . . . . . . . . . . . . . 7 5.4 ASME Overpressurization Analysis . . . . . . . . . . . . . . . . 7 5.5 Control Rod Withdrawal Error (CRWE) .............. 7 5.6 Fuel Loadi ng Error . . . . . . . . . . . . . . . . . . . . . . . 7 5.7 Determination Of Thermal Margins . . . . . . . . . . . . . . . . 8 6.0 POSTULATED ACCIDENTS ........................ 9 6.1 Loss-Of-Coolant Accident . . . . . . . . . . . . . . . . . . . . 9 6.1.1 Break Location Spectrum . . . . . . . . . . . . . . . . . 9 6.1.2 Break Size Spectrum . . . . . . . . . . . . . . . . . . . 9 6.1.3 MAPLHGR Analyses .................... 9 6.2 Control Rod Drop Accident ................... 10 7.0 TECHNICAL SPECIFICATIONS ...................... 11 7.1 Limiting Safety System Settings ................ 11 7.1.1 MCPR Fuel Cladding Integrity Safety Limit . . . . . . . 11 7.1.2 Steam Dome Pressure Safety Limit ............ 11

! 7.2 Limiting Conditions For Operation ............... 11 7.2.1 Average Planar Linear Heat Generation Rate Limits . . . . 11 7.2.2 Minimum Critical Power Ratio .............. 11 7.2.3 LHGR Limits . . . . . . . . . . . . . . . . . . . . . . . 12

I ANF-89-058 Page ii 7.3 Surveillance Requirements ................... 12 7.3.1 Scram Insertion Time Surveillance . . . . . . . . . . . . 12 7.3.2 Stability Surveillance ................. 13 8.0 METHODOLOGY REFERENCES ....................... 14 9.0 ADDITIONAL REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . 15 APPENDIX A SINGLE-LOOP OPERATI0ii ................... A-1 APPENDIX B SEISMIC-LOCA EVALUATION .................. B-1 I

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ANF-89-053 i Page iii 1

LIST OF TABLES Table Pace i

4.1 NEUTRONIC DESIGN VALUES ...................... 22 B.1 COMPARIS0N OF PHYSICAL AND STRUCTURAL CHARACTERISTICS OF FUEL ASSEMBLIES . . . . . . . . . . . . . . . . B-2 i

LIST OF FIGURES Fiaure Paae 3.1 SUSQUEHANNA UNIT 2 CYCLE 4 DESIGN BASIS RADIAL POWER . . . . . . . . 16

) 3.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-344L-9G5 ANF-3 FUEL . . . . . . . . . . 17 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION l

ADVANCED NUCLEAR FUELS ANF92-327L-9G4 ANF-3 FUEL . . . . . . . . . . 18 3.4 DESIGN BASIS LOCAL POWER DISTRIBUTION

) ADVANCED NUCLEAR FUELS XN-2 FUEL . . . . . . . . . . . . . . . . . . 19 3.5 DESIGN BASIS LOCAL POWER DISTRIBUTION

) ADVANCED NUCLEAR FUELS XN-1 CENTRAL FUEL . . . . . . . . . . . . . . 20 3.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-1 PERIPHERAL FUEL ............ 21 4.1 SUSQUEHANNA UNIT 2 CYCLE 4 ENRICHMENT DISTRIBUTION FOR THE ANF92-327L-9G4 ANF-3 FUEL LATTICE ............. 23 4.2 SUSQUEHANNA UNIT 2 CYCLE 4 ENRICHMENT DISTRIBUTION FOR THE ANF92-344L-9G5 ANF-3 FUEL LATTICF ............. 24 L

4.3 SUSQUEHANNA UNIT 2 CYCLE 4 REFERFKF WC L MDING PLAN . . . . . . . 25 5.1 SUSQUEHANNA UNIT 2 CYCLE 4 CONTROL RUD WITHDRAWAL

} ERROR ANALYSIS CONTROL R0D PATTERN . . . . . . . . . . . . . . . . . 26 L 5.2 SUSQUEHANNA UNIT 2 CYCLE 4 FLOW MCPR OPERATING LIMIT . . . . . . . . 27 h

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F ANF-89-058 Page 1 L

1.0 INTRODUCTION

This report provides the results of the analyses performed by Advanced j Nuclear Fuels Corporation (ANF)* in support of the Cycle 4 reload for Susquehanna Unit 2, which is scheduled to commence operation in the Fall of 1989. This report is intended to be used in conjunction with ANF topical report XN-NF-80-19(P)( A), Volume 4, Revision 1, " Application of the Exxon Nuclear Company 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 8.0). However, LHGR

, mechanical design limits (Reference 9.1) and plant transient simulation model developments (Reference 9.2) have been revised by ANF subsequent to NRC approval of XN-NF-80-19(P)(A), Volume 4, Revision 1. Both References 9.1 and 9.2 have been approved by the NRC for use in referencing in license applica-tions. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1.

All applicable neutronic calculations used the XTGBWR reactor simulator code described in XN-NF-80-19(P)(A), Volume 1 and Supplements I and 2, as i modified by Reference 9.11. This modification improved the numerical fitting of cross-section data associated with Sid history accounting in the simulator.

The Susquehanna Unit 2 Cycle 4 core will comprise a total of 764 9x9 fuel assemblies, including 204 unirradiated ANF-3 assemblies, 236 irradiated ANF i XN-2 assemblies, and 324 irradiated ANF XN-1 fuel assemblies. 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 Susquehanna Unit 2 during the previous operating cycle except for use of a two-second main steam isolation valve (MSIV) closure time for Cycle 4 (previously three se ands).

Additional information and the results of design studies detailii.g the I

development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.3.

• Formerly Exxon Nuclear Company (ENC).

- - - - _ . - _ _ - - - _ - ___.-___.._-_____-___-_-_-____m_____s_. , _ _ _ _ _ _ . _ _ _ , _ _ _ _ _ , . _

L ANF-89-058 4 Page 2 I f

'2.0 l

f FUEL MECHANICAL DESIGN ANALYSIS

, Applicable ANF Fuel Design Report: Reference 9.1 i To ensure that the expected power history for the fuel to be irradiated l

during Cycle 4 of Susquehanna Unit 2 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits (Figure 3.3 of Reference 9.1) have been specified. In addition, an LHGR operating limit to prevent fuel damage during 'eticipated operating occurrences (Figure 3.4 of Reference 9.1) has been spm,ified for ANF 9x9 fuel. The NRC has approved the ANF 9x9 fuel design for assembly exposures up to 40,000 mwd /MTV (Reference 9.4).

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3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2' Hydraulic Characterization 3.2.1. Hydraulic Compatibility p

The Susquehanna Unit 2 Cycle 4 core consists of a full core loading of ANF 9x9 fuel; thus, compatibility is assured.

[ .

j- 3.2.2 Thermal Marain Performance i The thermal margin performance of ANF 9x9 fuel is given by Reference 9.10.

3'2.3 Fuel Centerline Temperature L Applicable Generic Report Reference 9.1 i

3.2.5 ' Byoass Flow I

Calculated Bypass Flow Fraction at 10.0%

104% Power /100% Flow 3.3 MCPR Fuel Claddina Intearity Safety limit MCPR Safety Limit 1.06 3.3.1 Coolant Thermodynamic Condition I Rated Thermal Power 3293 MWt

Feedwater Flow Rate (at SLMCPR) 16.2 Mlbm/hr

> Core Pressure (at SLMCPR) 1043 psia Feedwater Temperature 383*F 3.3.2 Desian Qasis Radial Power Distribution Figure 3.1

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3.3.3 Desian Basis tccal Power Distribution Figures 3.2-3.6

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• ANF-89-058 Page 4 4.0 NUCLEAR DESIGN ANALYSIS

! 4.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment 3.17% and 3.53% '

Radial Enrichment Distribution Figures 4.1 and 4.2

< Axial Enrichment Distribution Uniform 3.27% and 3.44% with a 6" natural uranium top blanket Burnable Poisons

• Figures 4.1 and 4.2 Non-Fueled Rods Figures 4.1 and .

4.2 Neutronic Design Parameters Table 4.1 i ,

4.2 Core Nuclear Desian Analysis 4.2.1 Core Configuration Figure 4.3 Core Exposure at EOC3, mwd /MTU 20,799 Core Exposure at BOC4, mwd /MTU 13,719 Core Exposure at E0C4, mwd /MTU 24,564

Maximum Cycle 4 Licensing Exposure Limit, mwd /MTU 25,253

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) 4.2.2 Core Reactivity Characteristics 3250 mwd /MTU Cold k-eff, All Rods Out 1.11213 3250 mwd /MTU Cold k-eff, Strongest Rod Out 0.98367 -

Reactivity Defect (R-Value) 0.18% rho Standby Liquid Control System Core k-eff, Cold Conditions, 660 ppm Boron 0.97926 L

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• Burnable poisons are distributed uniformly over the enriched length of the -

designated rods. The natural urania axial D1anket sections do not contain ..

burnable absorber material.

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4.2.4 fore Reactivity Stability Power / Flow State Points Decay Ratio (C0TRAN) 66%/45% 0.75 70%/47% 0.75 75%/50%* 0.73 g Linear extrapolation along a constant decay ratio line to a higher power / flow l

!. condition is conservative compared to COTRAN calculations.

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• Intercept of APRM rod block.

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ANF-89-058 Page 6 1

5.0 ANTICIPATED OPERATIONAL OCCURRENCES i Applicable Generic Transient. References 9.5 I

Analysis Methodology Report and 9.7 l

5.1 Analysis ~0f Plant Transients At Rated Conditions Reference 9.6 i Limiting Transients: Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH) l

% Rated  % Rated Maximum Maximum Maximum Pressure Delta Event Power

• flgy Heat Flux Power (osia) CPR** Model

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LRWB 100% 100%' 121.7% 367% 1198 0.27 COTRANSA/ J XCOBRA-T FWCF 100% 100% 116.7% 210% 1178 0.22 COTRANSA/ )

XCOBRA-T

{

-LFWH '100%' 100% 121.4% 123% 1080 0.16 PTSBWR3/

XCOBRA Single-Loop Operation Appendix A i

1 5.2 Analyses For Reduced Flow Operation Reference 9.6 Limiting Transient: Recirculation Flow Increase Transient (RFIT) l

)

) Total Core Reduced Flow Recirculation Flow MCPR

(% Rated) Operatina limit l l

100 1.10 I 90 1.14 1

, 80 1.18 l 70 1.22 l 60 1.29 i- 50 1.42 40 1.60

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• 104% power used in analysis as design bases. .)

• • Uelta-CPR results for most limiting fuel type.

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I ANF-89-058 Page 7 5.3 Analyses For Reduced Power Ooeration Reference 9.6 Limiting Transient: Feedwater Controller Failure (FWCF)

% Power Transient Delta CPR 104 FWCF 0.22 80 FWCF 0.31 65 FWCF M3 3 40 FWCF 0.36 3 5.4 ASME Overoressurization Analysis Reference 9.6 Limiting Event Full MSIV Isolation g Worst Single Failure Direct Scram 5 Maximum Pressure 1,311 psig Maximum Steam Dome Pressure 1,295 psig 5.5 Control Rod Withdrawal Error (CRWE)

Limiting Control Rod Pattern for 108% Figure 5.1 Rod Block Setting 100% F1ow Distance Rod Block Setting Withdrawn Delta

(%) (ft) CPR 105 3.0 0.22 <

106 3.0 0.22 107 3.5 0.26 ,

108* 3.5 0.26 5.6 Fuel loadina Error Maximum Delta CPR 0.14 I! 1 I

• Rod Block Monitor setting recor 1..ded for Cycle 4 operation. h ll

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ANF-89-058

.Page 8 5.7 - Determination Of Thermal Marains

-Summary of Thermal Margin Requirements:' i Event Ep.WE flgg Delta CPR* MCPR Limit

.LRWB: 100%** 100% 0.27~ 1.33 FWCF 100%** 100% 0.22 1.28 LFWH 100%** 100% 0.16 1.22 CRWE 100% 100% 0.26 at 108% RBM 1.32

..MCPR Operating Limits at Rated Conditions:

MCPR Operatina Limit

'1.33 from LRWB' Reduced Flow MCPR ' Limits Figure 5.2 Power. Dependent MCPR Operating Limit Results for. Cycle 4:

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. Limiting Power. Dependent

' % Power /% Flow Transient MCPR Limit 100**/100 LRWB 1.33 80/100 FWCF 1.37 65/100 FWCF 1.39

.40/100 FWCF 1.42 l

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'* Delta- CPR'results for most limiting fuel type.

• • 104% power used in m lysis as design bases.

s ANF-89-058 Page 9 6.0 POSTULATED ACCIDENTS h 6.1 Loss-Of-Coolant Accident Seismic-LOCA Appendix B 6.1.1 Break location Soectrum Reference 9.8 s

6.1.2 Break Size Soectrum Reference 9.8 I

l 6.1.3 MAPLHGR Analyses Reference 9.9 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 discharge coefficient l

Bundle Average Peak Clad Peak Local

. Exposure MAPLHGR Temperature

• MWR*

l (GWd/MTV) (kw/ft) (Dearee F) (Perg'nt) 0 10.2 2060 3.9 5 10.2 2069 3.7 10 13.2 2121 3.7 15 10.2 2140 4.8 .

20 10.2 2147 5.2 L

25 9.6 2016 2.7

'30 8.9 1839 1.0 r 35 8.2 1752 0.7

) 40 7.5 1676 0.5 l

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• Peak clad' temperatures and metal water reaction (MWR) for XN-1 and XN-2 fuel are bounded by these results.

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ANF-89-058 Page 10 Bundle Average.

Peak Clad Temper .e Peak Local MWR I  ;

Exposure (GWd/MTU)

MAPLHGR (kw/ft)

(Dearee P _

ANF-3A* ANF 's6*

(Percent)-

ANF-3A* ANF-38*

)m 'l u 10.2 1998 1990 2.6 2.5 E 5 10.2 1937 1928 1.4 1.4 g 10 10.2 2079 2045 3.1 2.5 15 10.2 2124 2126 4.4 4.4 20 10.2 2161 2150 5.0 5.0 25 9.6 1992 1996 2.4  ?. 5 l 30- 8.9 1829 1831 1.0 1.0 j 35 8.2 1740 1744 0.7 0.7 31 40 7.5 1665 1670 0.5 0.5 gI

.6.2 [ontrol Rod Droo Accident '

Dropped Control Rod Worth, mk 15.4 Doppler Coefficient,1/k dk/dT -10.33 x (10)-6 .

Effective Delayed Neutron Fraction 0.0055 Four-Bundle Local Peaking Factor 1.38 Maximum Deposited Fuel Rod Enthalpy, cal /gm 249**

Number of Rods Exceeding 170 cal /gm <650 1

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• ANF-3A denotes ANF92-317B-9G4, and ANF-3B denotes ANF92-333B-9G5.

• • Deposited enthalpy was determined from parametric analysis where all ,

parameters, except Dropped Control Rod Worth, were enveloped. The well- 5 3j behaved relationship between rod worth and deposited enthalpy indicates that an extrapolation based on rod worth results in a conservative value of 1 deposited enthalpy.

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ANF-89-058 Page 11 7.0 TECHNICAL SPECIFICATIONS L 7.1 Limitina Safety System Settinas 7.1.1 MCPR Fuel Claddina Intearity Safety Limit MCPR Safety Limit 1.06 I

7.1.2 Steam Dome Pressure Safety limit Pressure Safety Limit (as measured in steam dome) 1325 psig Analysis shows that a steam dome pressure safety limit of 1358 psig is L allowed but the 1325 psig value used in Cycle 3 is to be conservatively retained.

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7.2 Limitina Conditions For Operation 7.2.1 Averaae Planar Linear Heat Generation Rate limits Byndle Average Exposure

> (GWd/MTV) MAPLHGR Limits (kw/ft) 0 10.2

{ 5 10.2 10 10.2 15 10.2 l

20 10.2 25 9.6 30 8.9 35 8.2 l 40 7.5 7.2.2 hinimum Critical Power Ratio ,

MCPR Operating Limits at Rated Conditions:

MCPR Operatina limit 1.33 from LRWB l

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ANF-89-058 Page 12 MCPR Operating Limits at Off-Rated Conditions:

E Figure 5.2 I,

At Reduced Flow Total Core Reduced Flow l

Recirculation Flow MCPR

(% Rated) Ooeratina limit 100 1.33 90 1. ?,3 80 70 1.33 1.33 3

5 (Interpolated) 56.9 1.33 50 1.42 40 1.60 At Reduced Power Reduced Power 4 Power Level MCPR

(% Rated) Ooeratina limit 100* 1.33 80 1.37 '

65 1.39 40 1.42 7.2.3 LHGR Limits LHGR Limits Figures 3.3 and 3.4 of Reference 9.1 7.3 Surveillance Requirements 7.3.1 Scram Insertion Time Surveillance Thermal limits established in Section 5.0 are based on minimum acceptable scram insertion performance as defined in the Technical Specifications. No additional surveillance for scram insertion is required for validation of thermal limits.

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'" ANF-89-058 Page '3 7.3.2' Stability Surveillance PP&L Will establish stability surveillance requirements for Susquehanna.-

Unit 2 ~ Cycle 4 in conformance with the interim operating. guidelines presented in NRC . Bulletin 88-07. Supplement 1 based on the calculation results prepared -

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ANF-89-058 Page 14 8.0 METH000LCGY REFERENCES A complete bibliography of applicable methodology references is provided

! in the following document: " Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads,"

XN-NF-80-19fP)(A), Volume 4, Revision 1, Advanced Nuclear Fuels Corporation *,

Richland, Washington 99352, March 1985.

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j. *Formerly Exxon Nuclear Company (ENC).

h ANF-89-058 Page 15 9.0 ADDITIONAL REFERENCES 9.1 " Generic Mechanical Design for Exxo' Nuclear Jet Pump BWR Reload Fuel,"

! XN-NF-85-67(P)(A), Revision 1, Advanced Nuclear Fuel s Corporation *,

Richland, Washington, September 4, 1986.

9.2 " Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," XN-NF-80-19f M(A1, Volume 3, Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, l January 1987.

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9.3 " Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power

- Research Institute, Palo Alto, California, May 1,1984.

9.4 " Qualification of Exxon Nuclear Fuel for Extended Burnup - Supplement 1 Extended Burnup Qualification of ENC 9x9 BWR Fuel," XN-NF-82-06(P)(A),

Supplement 1, Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1988.

9.5 " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

l XN-NF-79-71(P), Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, November 16, 1981.

> 9.6 "Susquehanna Unit 2 Cycle 4 Plant Transient Analysis," ANF-89-057, Advanced Nuclear Fuels Corporation, Richland, Washington, April 1989.

9.7 "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core ,

Analysis," XN-NF-84-105(P)(A), Volume 1 and Volume 1, Supplements 1 and 2, Advanced Nuclear Fuels Corporation, Richland, Washington, February 1987.

9.8 " Generic LOCA Break Spectrum Analysis BWR 3 &4 with Modified Low Pressure Coolant Injection Logic Using the EXEM Evaluation Model , "

XN-NF-84-117(P), Advanced Nuclear Fuels Corporation, -

Richland, Washington, December 1984.

9.9 "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for ENC 9x9 Fuel,"

XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, -

May 1986.

l 9.10 "Susquehanna Unit a Cycle 2 Reload Analysis Design and Safety Analyses,"

XN-NF-86-60, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.

9.11 Letter, R. A. Copeland (ANF) to M. W. Hodges (NRC), " Void History Correlation," RAC:058:88, September 13, 1988.

• Formerly Exxon Nuclear Company (ENC).

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• 0.86 : 0.89 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 1.00 : 0.95 :

• 0.89 : 0.92 : 0.98 : 1.07 : 0.92 : 1.08 : 0.98 : 1.04 : 1.00 :

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• 0.95 : 0.98 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.98 : 0.95 :

• 1.04 : 1.07 : 1.04 : 1.00 : 1.00 : 1.01 : 1.05 : 0.94 : 1.05 :

• 1.03 : 0.92 : 1.03 : 1.00 : 0.00 : 0.99 : 1.06 : 1.08 : 1.04 :

• 1 04 : 1.08 : 1.04 : 1.01 : 0.99 : 0.00 : 1.04 : 0.95 : 1.05 :

i  : 0.95 : 0.98 : 1.04 : 1.05 : 1.06 : 1.04 : 1.07 : 0.99 : 0.96 :

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j  : 1.00 : 1.04 : 0.98 : 0.94 : 1.08 : 0.95 : 0.99 : 0.93 : 1.00 :

0.95 : 1.00 : 0.95 : 1.05 : 1,04 : 1.05 : 0.96 : 1.00 : 0.95 :

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FIGURE 3.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-344L-9G5 ANF-3 FUEL ..

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• 0.87 : 0.91 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 0.99 : 0.97 : a

• :  :  :  :  : : : :  : g

• 0.91 : 0.92 : 0.97 : 1.07 : 0.92 : 1.07 : 0.97 : 1.04 : 1.00 : E 5'

• 0.95 : 0.97 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.98 : 0.95 :

• 1.04 : 1.07 : 1.04 : 1.00 : 1.00 : 1.01 : 1.05 : 0.94 : 1.04 :

• 1.03 : 0.92 : 1.03 : 1.00 : 0.00 : 0.99 : 1.06 : 1.08 : 1.04 :

• 1.04 : 1.07 : 1.04 : 1.01 : 0.99 : 0.00 : 1.04 : 0.95 : 1.05 :

0.95 : 0.97 : 1.04 : 1.05 : 1.06 : 1.04 : 1.07 : 0.99 : 0.96 :

:  :  :  :  : : : :  : E

0.99 : 1.04 : 0.98 : 0.94 : 1.08 : 0.95 : f 49 : 0.93 : 1.00 :

g

0.97 : 1.00 : 0.95 : 1.04 : 1.04 : 1.05 : 0.96 : 1.00 : 0.97 :

]

FIGURE 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-327L 9G4 ANF-3 FUEL I

E I

ANF-89-058 Page 15

-9.0 ADDITIONAL REFERENCES g 9.1 " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

g XN-NF-85-67(P)(A), Revision 1, Advanced Nuclear Fuels Corporation *,

Richland, Washington, September 4,1986.

9.2 " Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," XN-NF-80-19(P)(A), Volume 3, Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, January 1987.

9.3 " Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power Research Institute, Palo Alto, California, May 1,1984.

9.4 " Qualification of Exxon Nuclear Fuel for Extended Burnup - Supplement 1 Extended Burnup Qualification of ENC 9x9 BWR Fuel," XN-NF-82-06(P)(A),

I Supplement 1, Revision 2, Richland, Washington, May 1988.

Advanced Nuclear Fuels Corporation, I

• 9.5 " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"

XN-NF-79-71(P), Revision 2, Advanced Richland, Washington, November 16, 1981.

Nuclear Fuels Corporation, 9.6 "Scsquehanna Unit 2 Cycle 4 Plant Transient Analysis," ANF-89-057, Aovanced Nuclear Fuels Corporation, Richland, Washington, April 1989.

I 9.7 "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," XN-NF-84-105(P)(A), Volume 1 and Volume 1, Supplements 1 and 2, Advanced Nuclear Fuels Corporation, Richland, Washington, February 1987.

I. 9.8 " Generic LOCA Break Spectrum Analysis BWR 3 & 4 with Modified Low Pressure Coolant Injection Logic Using the EXEM Evaluation Model , "

I XN-NF-84-117(P), Advanced Nuclear Fuels Richland, Washington, December 1984.

Corporation, 9.9 "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for ENC 9x9 Fu el , " -

I. XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.

9.10 "Susquehanna Unit 2 Cycle 2 Reload Analysis Design and Safety Analyses,"

h XN-NF-86-60, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.

9.11 Letter, R. A. Copeland (ANF) to M. W. Hodges (NRC), " Void History Correlation," RAC:058:88, September 13, 1988.

I I

• Formerly Exxon Nuclear Company (ENC).

2 5

7 1

0 5

1

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g ANF-89-058 Page 17 m

E, am um

• 0.36 : 0.89 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 1.00 : 0.95 :

g . . . . . .

• 0.89 : 0.92 : 0.98 : 1.07 : 0.92 : 1.08 : 0.98 : 1.04 : 1.00 :

I. .. . .

I *: 0.95 : 0.98 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.98 : 0.95 :

I .

• 1.04 : 1.07 : 1.04 : 1.00 : 1.00 : 1.01 : 1.05 : 0.94 : 1.05 :

I * :.................__....................____...__ .............:

• 1.03 : 0.92 : 1.03 : 1.00 : 0.00 : 0.99 : 1.06 : 1.08 : 1.04 :

5- * :......___..........__............._____ ..____...__...___...._:

• 1.04 : 1.08 : 1.04 : 1.01 : 0.99 : 0.00 : 1.04 : 0.95 : 1.05 :

I *. . . . . . .

0.95 : 0.98 : 1.04 : 1.05 : 1.06 : 1.04 : 1.07 : 0.99 : 0.96 :

I .

I  : 1.00 : 1.04 : 0.98 : 0.94 : 1.08 : 0.95 : 0.99 : 0.93 : 1.00 :

I  :  :  :  :  :  :  :

0.95 : 1.00 : 0.95 : 1.05 : 1.04 : 1.05 : 0.96 : 1.00 : 0.95 :

FIGURE 3.2 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-344L-9G5 ANF-3 FUEL I

I I

1 I

ANF-89-058 Page 18 8

• : 0.87 : 0.91 : 0.95 : 1.04 : 1.03 : 1.04 : 0.95 : 0.99 : 0.97 :

e . . e e .

• : 0.91 : 0.92 : 0.97 : 1.07 : 0.92 : 1.07 : 0.97 : 1.04 : 1.00 :

• : 0.95 : 0.97 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.98 : 0.95 : I

• : 1.04 : 1.07 : 1.04 : 1.00 : 1.00 : 1.01 : 1.05 : 0.94 : 1.04 :

• : 1.03 : 0.92 : 1.03 : 1.00 : 0.00 : 0.99 : 1.06 : 1.08 : 1.04 :

• : 1.04 : 1.07 : 1.04 : 1.01 : 0.99 : 0.00 : 1.04 : 0.95 : 1.05 :

0.95 : 0.97 : 1.04 : 1.05 : 1.06 : 1.04 : 1.07 : 0.99 : 0.96 : g

:  :  :  :  : : : :  : 5

:  :  :  :  : : : :  : E

0.99 : 1.04 : 0.98 : 0.94 : 1.08 : 0.95 : 0.99 : 0.93 : 1.00 :

g

0.97 : 1.00 : 0.95 : 1.04 : 1.04 : 1.05 : 0.96 : 1.00 : 0.97 :

......................................_.._es................as....

FIGURE 3.3 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS ANF92-327L-9G4 ANF-3 FUEL I

I; ,

I L_._________ _ . _ _

m ylf- '

G

' lp

_~

r .A.;.

j7 1, hh -

o+ ANF-89-058-T Page 19 e

c b *I..............a..................................................  ;

• 0.91.: LO.92 :'0.95;: 1.01 : 1.01 : 1.01 :'0.95 : 0.97 : 0.95.: 1 m o '*::, -:  :  :  :-  :  :  :-  :  :

• : . .. .... . . . . .... .. . . .. .. .. . ... .. .. .. . . . . . .. . ... . .. . ... . . . . ... .._: 'i

"., * ~... - -.

1 i' *  : 0.92 : 0.94 : 0.98 : 1.05 : 0.94 : 1.06 : 0.97 : 1.01 : 0.97 : I L

- . *;.1 ..

1 4

i. *. .

p'.

• 0.95 : 0.98 : 0.93 : 1.05 : 1.05 : 1.06 : 1.05 : 0.99 : 0.96 :

~

o.

,. . *l:..............................................................:

.l ,

.. . . . . e .  ;

• 1.01 :-1.05 : 1.05 :'1.03 : 1.03 : 1.03 : 1.06 : 0.96 : 1.01 :

• 1:..............................................................:

. i i

1*  : 1.01 : 0.94 i 1.05 : 1.03 : 0.00 : 1.01 :-1.06 : 1.06 : 1.01 : .. . .

i

,;. i*: . .

1 1

J

• :................................................................: j

. 'I m

'* : 1.01 :'1.06 : 1.06 : 1.03 : 1.01 : 0.00 : 1.04 : 0.98 :.1.02 : 1 a

p . . .. ..

q .*  :..............................................................:

L .: 0.95 :'0.97.: 1.05 : 1.06 : 1.06 : 1.04 : 1.06 : 0.98 : 0.96 :

l k' .

.  :.  : ~: .:  :  :  :  :-  : I j

~ :'  :  :  :  :  :

': 0.97 : 1.01 : 0.99 :'0.96 : 1.06 : 0.98 : 0.98 : 0.95 : 0.98 :  ;

f'  : -:-  :  :  :-  :

L  : 0.95 :.0.97 : 0.96 : 1.01 : 1.01' : 1.02 : 0.96 : 0.98 : 0.96 :

v i

!! . . . FIGURE 3.4 DESIGN BASIS LOCAL POWER DISTRIBUTION

. 1

$ ADVANCED NUCLEAR FUELS XN-2 FUEL l'

1)

[.

7 L

L 1 1

s

%. aI l ) y'

s l

l.

ANF-89-058 Page 20 I,

• * ********************* g

• ..._......................................__.__ ..............__ W '

• : 0.91 : 0.91 : 0.93 : 0.99 : 0.99 : 1.00 : 0.94 : 0.95 : 0.94 :

• : 0.91 : 0.92 : 0.98 : 0.97 : 1.06 : 0.96 : 0.99 : 0.93 : 0.95 :

• : 0.93 : 0.98 : 0.95 : 1.08 : 1.08 : 1.09 : 1.06 : 0.98 : 0.94 :

• : 0.99 : 0.97 : 1.08 : 1.06 : 1.06 : 1.07 : 1.09 : 1.07 : 1.00 :

• : 0.99 : 1.06 : 1.08 : 1.06 : 0.00 : 1.03 : 1.09 : 1.07 : 1.00 :

• : 1.00 : 0.96 : 1.09 : 1.07 : 1.03 : 0.00 : 1.06 : 0.97 : 1.00 :

0.S4 : 0.99 : 1.06 : 1.09 : 1.09 : 1.06 : 1.06 : 0.99 : 0.95 :

:  :  :  :  :  :  : :  : E

0.95 : 0.93 : 0.98 : 1.07 : 1.07 : 0.97 : 0.99 : 0.94 : 0.96 :

3

0.94 : 0.95 : 0.94 : 1.00 : 1.00 : 1.00 : 0.95 : 0.96 : 0.94 :

FIG'JRE 3.5 DESIGN BASIS LOCAL POWER DISTRIBUTION

' ADVANCED NUCLEAR FUELS XN-1 CENTRAL FUEL I

I

.i ;.

n, .-

^

m

^

ANF-89-058-Page 21-i 5

In

}

M - *:....................................................___ ........

. '.* :.l.00':

.. 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :

-*: 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 'i

[ *. .

l1 *. . .

j I. *: '1.00 : l'.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : -

. i

• :........................................................_____.: l

• :.1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1

.* :..............................................................: i

• . .. . . . . . . . . 'l

• $11.00$'1.00$1.00$1.00$0.00$1.00$1.00$1.00$1.00$ l

[1.. .. . . . .. . . . . .

• $1.00$1.00$.1;00'$1.00$'1.00$0.00$1.00$1.00$1.00$ j i

c f * :..............................................................:

:-  :  :  :  :  : :  :  : j L  ::1.00': l'.00 :.1.00 : 1.00 : 1.00 : 1.00 :.1.00 : 1.00 : 1.00 : j

[

: .:  :  :  :  : :  :  : i p; . .-- . . -

'l.00 :-l.00 : 1.00': 1.00 : 1.00 : 1.00 : 1.00 : 1.00-:.1.00 :

1

................__............................................: i l'.:

l  : 1.00 : 1.00 : 1.00 : 1.00 : ' 00 : 1.00 : 1.00 : 1.00 : 1.00 : . l

't l.-','.

I- FIGURE 3.6 DESIGN BASIS LOCAL POWER DISTRIBUTION ADVANCED NUCLEAR FUELS XN-1 PERIPHERAL FUEL li l

p .

I ANF-89-058 Page 22 TABLE 4.1 NEUTRONIC DESIGN val.UES Il Furl Pellet Reference 9.1 Fuel Rod Reference 9.1 Fuel Assembly Reference 9.1 I'

Core Data Number of fuel assemblies 764 Rated thermal power, MW 3293 Rated core flow, Mlbm/hr 100 Core inlet subcooling, Btu /lbm 24.0 Moderator temperature, 'F 548.8 Channel thickness, inch 0.080 Fuel assembly pitch, inch 6.00 Wide water gap thickness, inch 0.562 Narrow water gap thickness, inch 0.562 Control Rod Data Absorber material' BC 4 g

' Total blade ~ span, inch 9.75 5 Total blade support span, inch 1.58 Blade thickness, inch 0.260 '

Blade face-to-face internal dimension, inch 0.200 '

Absorber rods per blade 76 Absorber rod outside diameter, inch 0.188 4 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70 I

I I

I ,

II

ANF-89-058 Page 23

• i 1

• _......................................._._.............................. i

• :  :  :  :  :  :  :  :  : l

• LL : L  : ML  : M  : M  : M : ML : ML : L :

• L : ML : M  : MH  : M*  : MH  : M*  : M : ML :  ;

• :  :  :  :  :  :  :  :  : 1

• ML : M : M* : H : H  : H : MH  : M : ML  :

• M : MH : H : H : H : H : H : M*  : M :

• M : M*  : H : H : W  : MH : H : MH : M :

• M : MH : H : H : MH : W : MH : M*  : M :

• ML : M* : MH : H : H : MH : MH : M : ML :

ML : M : M : MH : M : ML :

I

M*  : M*  : ML  :

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

LL RODS ( 1) ---

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

1.95 W/0 U235 ML RODS (16) ---

2.42 W/0 U235 M RODS (20) ---

3.10 W/0 U235 MH RODS (13) ---

4.01 W/0 U235 H RODS (15) ---

4.42 W/0 U235 M* RODS ( 9) ---

3.10 W/0 U235 + 5.00 W/0 GD203 W RODS ( 2) ---

INCP.T WATER R0D FIGURE 4.1 S'JSQUEHANNA UNIT 2 CYCLE 4 ENRICHMENT DISTRIBUTION FOR THE ANF92-327L-9G4 ANF-3 FUEL LATTICE I

q 1

ANF-89-058 Page 24 i

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • i 3

• : LL : L : ML : M : M : M : ML : ML : L :

• :  :  :  :  :  :  :  :  :  : an Ed '

• : L : ML : M : MH : M*  : MH : M*  : M : ML :

• : ML : M : M*  : H : H : H : MH : M : ML : E

:  :  : 3

• : M : MH : H : H : H : H : H : M*  : M :

MH :

m

• : M : M*  : H : H : W  : MH : H : M :

• : M : MH : H : H : MH : W  : MH : M*  : M :

• : ML : M*  : MH : H : H : MH : MH : M : ML  :

ML : M : M : M*  : MH : M*  : M : ML  : ML  :

L : ML : ML : M : M : M : ML : ML :

L :

LL RODS ( 1) ---

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

1.95 W/0 U235 ML RODS (16) ---

2.55 W/0 U235 _

M RODS (20) ---

3.27 W/0 U235 MH RODS (13) ---

4.23 W/0 U235 H RODS (15) ---

4.66 W/0 U235 M* RODS ( 9) ---

3.27 W/0 U235 + 4.00 W/0 GD203 3 ,,

W RODS ( 2) ---

INERT WATER R00 5' FIGURE 4.2 SUSQUEHANNA UNIT 2 CYCLE 4 ENRICHMENT DISTRIBUTION l

FOR THE ANF92-344L-9G5 ANF-3 FUEL LATTICE I _ _ _ -

l ANF-89-058 Page 25 l

'I

.l

'l 2 3 4 5 6 7 8- 9 10 11 12 13 14 15

.1 A2 C1 A2 Cl A2 C1 A2 C1 A2 C1 A2 C1 A2 C1 A2 2 C1 D0 A2 A2 00 A2 DO A2 DO A2 E0 A2 E0 B1 A2 i-

3 A2 A2 B1 E0' C1 EO B1 E0 C1 E0 81 EO B1 B1 A2 q L.

4 C1 A2 EO A2 C1 A2 DO A2 00 A2 EO A2 E0 81 A2 F

5. A2 DO C1 C1 A2. D0' A2 B1 A2 E0 B1 EO C1 81 A2 ]

-6 C1 A2 EO A2 DO A2 DO B1 EO A2 DO A2 A2 B1 A2 ,

1 7 A2 00 81 DO A2 DO. A2 EO A2 B1 A2 B1 DO B1 A2 8 Cl- A2 E0' A2. B1 B1 EO A2 DO A2 B1 DO B1 A2 l

9 ~A2 DO C1 00 A2 EO A2 DO A2 C1 00 B1 A2 k '10 ~ C1 'A2 EO A2 EO A2- B1 A2 C1 Al A2 A2 A2 11 A2' E0 B1 EO . B1 DO A2 B1 DO A2 A2 12 C1 A2 EO A2 EO A2 B1 D0' B1 A2 XY - Fuel Type X Burned Y Cycles 1 13 A2 E0. 81 EO C1. A2 D0 B1 A2 A2

'14 C1 B1 B1 B1 B1 B1 B1 A2

'15 'A2 'A2 A2. A2 A2 A2. A2 o

l Fuel Type No. of Bundles Description i 'A' 324 XN-1 ANF92-331B-7G4 B 140 XN-2 ANF92-3338-9G4 C 96 XN-2 ANF92-3338-10G5 l- D 100 ANF-3 ANF92-3178-9G4 l E 104 ANF-3 ANF92-333C-9G5 FIGURE 4.3 SUSQUEHANNA UNIT 2 CYCLE 4 REFERENCE CORE LOADING PLAN

[

L c '

ANF-89-058 I Page 26 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 I

55 -- --

08 --

16 --

08 -- -- 55 51- -- -- 36 --

38 --

38 --

36 -- --

51 47 -- --

02 --

02 --

04 --

02 -- 02 -- --

47 I 43 -- --. 36 -- 38 --

42 --

42 -- 38 --

36 -- --

43  ;

39 --

08 --

02 --

08 -- 16 --

08 --

02 --

08 --

39 35 --

-- 38 --

42 --

30 --

30 --

42 --

38 -- --

35 j i

31 --

16 --

04 --

16 -- ** --

16 --

04 --

16 --

31 27 -- --

38 -- 42 --

30 --

30 --

42 --

38 -- --

35 23 --

08 --

02 --

08 --

16 --

08 --

02 --

08 --

39 19 -- --

36 --

38 --

42 -- 42 -- 38 --

36 -- --

43 l

15 -- --

02 --

02 -- 04 -- 02 --

02 -- --

47 11 -- -- 36 -- 38 -- 38 -- 36 -- --

51 7- -- -- 08 --

16 --

08 -- --

55 3 -- -- -- -- -- -- --

3 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 l Ii 1

• • Control Rod 8eing Withdrawn Rod Position in Notches Withdrawn Full in = 00 'l )~

Full out --

I )

FIGURE 5.1 SUSQUFHANNA UNIT 2 CYCLE 4 CCNTROL R0D WITHDRAWAL ERROR ANALYSIS CONTROL R0D PATTERN I

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W ANF-89-058 Page A-1

~ APPENDIX A

[

SINGLE-LOOP OPERATION' I

i l

This appendix provides limits and justification of those limits for j single-loop operation. .

1 A.1 ANTICIPATED OPERATIONAL OCCURRENCES Reference A.1 )

L The NSSS supplicr has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an

(- extended period of time. These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed when both recirculation systems are in operation. The physical i interdependence between core power and recirculation flow rate inherently ]

limits'the core to less than rated power.

ANF fuel was designed to be compatible with the co-resident fuel in 1

' thermal hydraulic, nuclear, and mechanical design performance. The ANF methodology has given. results which are consistent wiu' those of previous

(  : analyses for normal two-loop operation. Many analyses performed by the NSSS supplier for single-loop ' operation are also applicable to single-loop cperation with fuel provided by ANF. ,

[ For single-loop operation, the NSSS vendor found that an increase of 0.01 L in the MCPR safety limit was needed to account for the increased flow measurement uncertainties and increased TIP uncertainties associated with l- single-pump operation. ANF has evaluated the effects of the increased flow measurement uncertainties on the MCPR safety limit and found that the NSSS vendor determined increase in the allowed MCPR safety limit is also applicable to ANF fuel during single-loop operation. Thus, increasing the l MCPR safety limit by 0.01 for single-loop operation (1.07) with ANF fuel is sufficiently conservative to also bound the increased flow measurement uncertainties'for single-loop operation.

l' o

I ANF-89-058 Page A-2 The full power fu; low two-loop MCPR operating limit plus .01, together with the MCPRf curve for two-loop operation plus 0.01 and the MCPRp curve for gi '

two-loop operation plus 0.01, conservatively bound all transient::. E s The stability surveillance requirements addressed in Section 7.3.2 are h, applicable to single-loop operation.

A.2 POSTULATED ACCIDENTS References A.1 & A.2 A pump seizure accident event was analyzed for Susquehanna Unit 2 Cycle 4 to confirm the insignificance of this event relative to the design basis LOCA.

Any fuel rods which experience boiling transition would be expected to be in the film boiling mode for a short period. In addition, the film boiling would be limited to small localized areas in the af fected fuel assemblies.

Because of this short duration, fuel failurer due to overheating or clad strain would not be expected as a result of this accident. Thus, the consequences of this event are bounded by the LOCA where fuel failures are assumed to be extensive.

ANF performed LOCA analyses for singic toop conditions and has determined that the MAPLHGR limit curves (Section 7.2) for two-loop operation are also applicable to single-loop operation.

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ANF-89-058 Page A-3 REFERENCES I'

A.1 "Susquehanna Unit 2 Cycle 2 Single Loop Operation Analysi s , "

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XN-NF-86-146, 'Avanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.

A.2 "Susquehanna LOCA Analysis for Single Loop Operation." XN-NF-86-125, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.

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ANF-89-058 Page B-1 APPENDIX B SEISMIC-LOCA EVALUATION 1

1 The structural response of Advanced Nuclear Fuels Corporation's (ANF's)

[

9x0 fuel is similar to the structural response of the GE 8x8R fuel it replaces I in the Susquehanna Unit 2 core. Inerefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains f-r applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.

l The physical and structural properties of the 9x9 and the 8x8 fuel types 1

which are important to the dynmic response of the fuel are summarized in Table B.I. The close agreement between the important parameters for the ANF 9x9 and GE 8x8R fuel types indicates that the structural response would be very similar for both fuel types.

L Similarity of the natural frequencies of the two fuel types mentioned above. is further assured by the stiffness of the fuel assembly channel box.

f Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel. ANF calculations show that approximately 97% of the stiffness of a fuel assembly is attributable to the stiffness of the channel box. For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable, f~ Caformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

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ANF-89-058

j. Page B-2 TABLE B.1 COMPARIS0N OF PHYSICAL AND STRUCTURAL CHARACTERISTICS OF FUEL ASSEMBLIES Fuel Tvoes Property ANF 9x9 GE 8x8R I

Assembly Weight, lbs 580 600 g Number of Spacers 7 7 Overall Assembly Length, in 171.29 171.40  !

Assembly Frequencies, cps )

1.9

• Mode 1 3.7 2

3 4

6.5 10.4 3

5 ,

5 15.5 6 21.9 .-

7 29.1  ;

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• GE proprietary.

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I ANF-89-058 Issue Date: 5/11/89 SUSQUEHANNA UNIT 2 CYCLE 4 RELOAD ANALYSIS Design and Safety Analyses Distribution:

D. J. Braun

0. C.' Brown

'R. E. Collingham R. A. Copeland L. J. Federico

'R. G. Grummer K. D. Hartley M. J. Hibbard A. L. B..Ho M. L. Hymas S. E. Jensen T. H. Keheley

'T. L. Lotz T. E. Millsaps L. A. Nielsen-C. C. Robe'rts Jr.

R. B. Stout C. J. Volmer

'H. E. Williamson

... H. G. Shaw/PP&L-(20)-

F. Document Control (5) f-L L

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