ML17146A419

Cycle 2 Reload Analysis Design & Safety Analyses.

ML17146A419
Person / Time
Site:	Susquehanna
Issue date:	05/31/1986
From:	Keheley T, Patten T, Williamson H SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17146A415	List: ML17146A414 ML17146A416 ML17146A417 ML17146A419 ML17146A421 ML17146A422 ML17146A423 ML18040B150
References
XN-NF-86-60, NUDOCS 8606240316
Download: ML17146A419 (62)

Text

XN-NF-86-60 SUSQUEHANNA UNIT 2 CYCLE 2 RELOAD ANALYSIS DESIGN AND SAFETY ANALYSES MAY 1986 RICHLAND, WA 99352 EXXON NUCLEAR COMPANY, INC.

8bOb24031b 8bObi9 PDR ADQCK 05000388 l P PDR

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XN-NF-86-60 Issue Date: 5/15/86 StjS(UEHANNA UNIT 2 CYCLE 2 RELOAD ANALYSIS Design and Safety Analyses Prepare:

T.H. Keheley, Team Lead BWR Safety Analysis Approve:

H.E. Wi amson, Manager Licensing and Safety Engineering Approve: Pi h

.W. Patten, anager Neutronics and Fuel Management Concur:

G. . War , Manager Reload Licensing Concur:

J. . Morgan, Ma ager Customer Servic's Engineering Approve: y5 Jg'er G.J. Busse man, Manager Fuel Design.

Approve:

G.L. Ri ter, Manager Fuel Engineering and Technical Services jgi/min EQON NUCLEAR COMPANY, INC.

NUCI.EAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was rlerived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub.

mitted by Exxon Nuclear to the USNRC as part of a technical contri.

bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nudear-fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstradon of compliance with the USNRC's reguladons.

Without derogating fmm the foregoing, neither Exxon Nuclear nor any person acting nn its behalf:

A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor.

mation contained in this document, or tha! 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 dan ages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN- NF. FOO, 766

XN-NF-86-60 TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

2.0 FUEL MECHANICAL DESIGN ANALYSIS.....................

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

3.2 Hydraul i c Characteri zati on...... ~ ~ 3 3.2.1 Hydraul i c Compati bi 1 i ty .. ~ ~ ~ 3 3.2.2 Thermal Margin Performance, Comparison.......... ~ ~ ~ 3 3.2.3 Fuel Centerline Temperature..................... ~ ~ 3 3.2.5 Bypass Flow...................... ~ ~ ~ ~ ~ ~ ~ 4 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........... ~ ~ ~

4.0 NUCLEAR DESIGN ANALYSIS...,.........................

4.1 Fuel Bundle Nuclear Design Analysis........

4.2 Core Nuclear Design Analysis.................... ~ ~ ~ 5 4.2.1 Core Configuration..... ~ ~ ~ 5 4.2.2 Core Reactivity Characteristics................. ~ ~ ~ 6 4.2.4 Core Hydrodynamic Stability............,........ ~ ~ ~ 6 5.0 ANTICIPATED OPERATIONAL OCCURRENCES.................

5.1 Analysis of Plant Transients at Rated Conditions....

5.2 Analyses for Reduced Flow Operation.................

5.4 ASME Overpressurization Analysis....................

5.5 Control Rod Withdrawal Error........................

5.6 Fuel Loading Error......................

5.7 Determination of Thermal Margins.............

6.0 POSTULATED ACCIDENTS ...................

6.1 Loss-of-Coolant Accident............ . 11 6.1.1 Break Location Spectrum........... ll 6.1.2 Break Size ll Analyses for ENC XN-1 9x9 Fuel Spectrum..................'APLHGR 6.1.3 11

XN-NF-86-60 TABLE OF CONTENTS (Continued)

Section Page Number Number 6.2 Control Rod Drop Accident....................... 12 7.0 TECHNICAL SPECIFICATIONS......................... 13 7.1 Limiting Safety System Settings........ 13 7.1.1 Fuel Cladding Integrity Safety Limit... 13 7.1.2 Steam Dome Pressure Safety Limit 13 7.2 Limiting Conditions for Operation-...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 13 7.2.1 Average Planar Linear Heat Generation Rate Limits for ENC XN-1 9x9 Fuel.. 13 7.2.2 Hinimum Critical Power Ratio. 14 7,2.3 LHGR Limits.......................,.... 15 7.3 Surveillance Requirements.......... 15 7.3.1 Scram Insertion Time Surveillance...... 15.

7.3.2 Stability Surveillance................. 15 8.0 METHODOLOGY REFERENCES..........,. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 16 9.0 ADDITIONAL REFERENCES.................................. 17 APPENDICES A. SINGLE LOOP OPERATION.................................. A-1 B. SEISMIC-LOCA EVALUATION................................ B-1 C. INCREASED CORE FLOW AND FINAL FEEDWATER TEMPERATURE REDUCTION PLANT TRANSIENT RESULTS.......... C-1

111 XN-NF-86-60 LIST OF TABLES Table Title Pacae 4.1 Neutronic Design Values....................... 22

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iv XN-NF-86-60 LIST OF FIGURES Ficiure Title Pacae 3.1 Hydraulic Demand Curves 'for Susquehanna Unit 2 C ycle 2 Core.............................,............. 18 3.2 Susquehanna Unit 2 Cycle 2 Safety Limit Radial Power Histrogram....................................... 19 3,3 Design Basis Local Power Distribution for ENC 9x9 Fuel .................................. .. 20 3.4 Design Basis Local Power Distribution for G.E. 8x8 Fuel......................, 21 4.1 Susquehanna Unit 2 XN-1 3.42 w/o Central Enrichment Distribution.......... ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ 23 4.2 Susquehanna Unit 2 Cycle '2 Reference Core Loading...... 24 5.1 Susquehanna Unit 2 Cycle 2 Control Rod Withdrawal Error Analysis, Initial Control Rod Pattern for 106 RBM Setting 100% Flow Case.......................... 25 5.2 Susquehanna Unit 2 Cycle 2 Control Rod Withdrawal Error Analysis, Initial Control Rod Pattern for 108 RBM Setting 100% Flow Case................... . . 26 5.3 Susquehanna Unit 2 Cycle 2 Control Rod Withdrawal Error Analysis, Initial Control Rod Pattern for 106/108 RBM Setting, 108% Flow Case....................... .. . 27 5.4 Reduced Flow MCPR Operating Limit................ . 28

XN-NF-86-60

1. 0 INTRODUCTION This report provides the results of the analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 2 reload for Susquehanna Unit 2, which is scheduled to commence operation in September 1986. This report is intended 1 dd '

XRXNC pd 1 p ~X"-

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 11 C,R The Susquehanna Unit 2 Cycle 2 core will comprise a total of 764 fuel assemblies, including 324 unirradiated ENC XN-1 9x9 assemblies, and 440 previously irradiated Type P8x8R assemblies fabricated by General Electric.

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. Additional information and the results of design studies covering the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.1.

4 XN-NF-86-60 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ENC 9x9 Fuel Design Report: Reference 9.2 To assure that the expected power history for the 9x9 fuel to be irradiated during Cycle 2 of Susquehanna Unit 2 is bounded by the assumed power history in the fuel mechanical design analysis, an LHGR operating limit (Figure 3.3 of Reference 9.2) has been specified for ENC 9x9 fuel. In addition, an LHGR operating limit for Anticipated Operating Occurence (Figure 3.4 of Reference 9.2) has been specified for ENC 9x9 fuel.

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XN-NF-86-60 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 H draul i c Characteri zat i on 3.2.1 Hydraulic Compatibility Component hydraulic resistances for the constituent fuel types in the Susquehanna Unit 2 Cycle 2 core have been determined in single phase flow.

tests of full scale assemblies. Figure 3. 1 illustrates the hydraulic demand curves for ENC 9x9 fuel and G.E. 8x8 fuel in the Susquehanna Unit 2 core. The similar hydraulic performance 'ndicates adequate compatibility for co-residence in the Susquehanna cores.

3.2.2 Thermal Margin Performance, Comparison Core Confi uration ENC Fuel MCPR GE Fuel MCPR All GE Fuel 1.28 All ENC Fuel 1.33 Mixed Core 1.34 1.27 3.2.3 Fuel Centerline Temperature Exposure at Minimum Margin Point 5000 MWD/MT Centerline Temperature at 120% Power 4157 F Melting Point of Fuel 5080 F Margin to Centerline Melting 923 F

XN-NF-86-60 3.2.5 Bypass Flow Calculated Bypass Flow Fraction 10.3%

at 100% Power/100% Flow Calculated Bypass Flow Fraction 9.8%

at 100% Power/108% Flow 3.3 HCPR Fuel Claddin Inte rit Safet Limit Safety Limit MCPR = 1.06 3.3.1 Coolant Thermodynamic Condition Rated Thermal Power 3293 HWt Feedwater Flowrate (at SLHCPR) 16.2 Hlbm/hr Steam Dome Pressure (at SLMCPR) 1010 psia Feedwater Temperature* 420 F 3.3.2 Design Basis Radial Power Distribution See Figure 3.2 3.3.3 Design Basis Local Power Distribution See Figures 3.3 and 3.4

• Conservative assumption relative to HCPR Safety Limit Honte Carlo procedure.

XN-NF-86-60 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichment 3.31%

Radial Enrichment 'Distribution Figure 4. 1 Axial Enrichment Distribution Uniform 3.42%

with 6" natural ura'nia top bl anket Burnable Poisons Fig. 4.1 Note: Burnable poisons are distributed uniformly over the enriched length of the designated rods. The natural urania axial blanket sections do not contain burnable absorber material.

Non-Fueled Rods Fig. 4.1 Neutronic Design Parameters Table 4.1 4.2 Core Nuclear Desi n Anal sis 4.2.1 Core Confi uration Figure 4.2 Core Exposure at EOC1, HWD/MT 12220 Core Exposure at BOC2, MWD/MT 7805 Core Exposure at EOC2, HWD/HT 18305 Note: Cycle 2 safety analyses are valid for EOCl exposure from -1000 HWD/HT to +780 MWD/HT from the nominal value reported above.

XN-NF-86-60 4.2.2 Core Reactivit Characteristics BOC Cold K-effective, All Rods Out 1.10813 BOC Cold K-effective, Strongest Rod Out 0.97320 Reactivity Defect (R-Value) 0.04% rho Standby Liquid Control System Reactivity,

'Cold Conditions, 660 ppm 0.9758 4.2.4 Core H drod namic Stabilit Power Flow State Points Deca Ratio COTRAN 38.5/30 < 0.30 58/60 < 0.15 68/45 0.59 These state points bound the surveillance region. A COTRANSA2 and additional COTRAN calculations will be provided at a later date and startup tests are scheduled for demonstration of the stability performance.

XN-NF-86-60 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Report Reference 9.3 5.1 Anal sis of Plant Transients at Rated Conditions Reference 9.4 & 9.5 Limiting Transient(s): Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH)

Maximum Maximum Maximum Pressure Delta Event Power* Flow Heat Flux Power ~sia CPR Model LRWB 100% 100% 114 3% 274% 1213 0.17 COTRANSA FWCF 100% 100% 114.7% 245% 1180 0.15 COTRANSA LFWH 100% 100% N/A N/A N/A 0.08 XTGBWR Note: See Appendix A for single-loop operation.

See Appendix C for Final Feedwater Temperature Reduction (FFTR) and Increased Core Flow (ICF).

104% power used in analysis as design bases.

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

XN-NF-86-60 5.2 Anal ses for Reduced Flow 0 eration Reference 9.4 Limiting Transient(s): Recirculation Flow Increase Transient (RFIT) 5,4 SHE Over ressurization Anal sis Reference 9.4 Limiting Event Full HSIV Isolation Worst Single -Failure Direct Scram Maximum Pressure 1315 psig Maximum Steam Dome Pressure 1301 psig 5.5 Control Rod Withdrawal Error CRWE Starting Control Rod Pattern for Analysis For 100% Flow 9 106 RBH setting Figure 5.1 For 100% Flow 9 108 RBM setting Figure 5.2 For 108% Flow 9 106/108 RBM setting Figure 5.3 100% Flow 108% Flow Distance Distance Block Settio ~ft Withdrawn Delta CPR ~ft Withdrawn Delta'od CPR 105 4.0 0.15 4.5 0.22 106* 4.5 0.17 5.0 0.23 107 5.0 0.18 6.0 0.24 108* 6.0 0.21 8.0 0.26

• Rod Block Monitor settings selected for Cycle 2 operation 5.6 Fuel Loadin Error Maximum Delta-CPR 0.15

XN-NF-86-60 5.7 Determination of Thermal Mar ins Summary of Thermal Margin Requirements Event Power Flow Delta-CPR MCPR Limit LRWB 100% 100% 0.17 1.23 LFWH 100% 100% 0.08 1.14 FWCF 100% 100% 0.15 1.21 CRWE ,

100% 100% 0. 17 at 106% RBH 1.23 CRWE 100% 100% 0.21 at 108% RBH 1.27 FWCF* 100% 100% 0.17 .1.23 LRWB 100% 100% 0.15 1.21 LRWB 100% 108% 0.17 1.23 LFWH 100% 108% 0.08 1.14 FWCF 100% 108% 0.16 1.22 CRWE 100% 108% 0.23 at 106% RBM 1.29 CRWE 100% 108% 0.26 at 108% RBH 1.32 FWCF* 100% 108% 0.17 1.23 LRWB 100% 108% 0.15 1.21 Note  : Events are results of bounding analyses (1.06 safety limit used).

HCPR Operating Limits at Rated Conditions Rod Block Settin MCPR 0 eratin Limit 106% 1.23 108% 1.27

• At Final Feedwater Temperature Reduction of 65'F.

10 XN-NF-86-60 NCPR Operating Limits at Off-Rated Conditions.

At 100% Power/108% Flow Rod Block Settin MCPR 0 eratin Limit 106% 1.29 108% 1.32 At Reduced Flow Figure 5.4 Power Dependent HCPR Operating Limit Nominal Feedwater Tem FFTR GE PSxSR ENC 9x9 GE PSxSR ENC 9x9 100/100 1.20 1.21 1.22 1.23 80/100 1.28 1.30 1.26 1.28 65/100 1.29 1.31 1.30 1.32 40/100 1.32 1.35 1.32 1.35 100/108 1.21 1.22 1.22 1.23 80/108 1.26 1.28 1.26 1.28 65/108 1.29 1.31 1.30 1.32 40/108 1.33 1.36, 1.32 1.36

• 65'F reduction in Feedwater Temperature.

XN-NF-86-60 6.0 POSTULATED ACCIDENTS

6. 1 Loss-of-Coolant Accident

6. 1. 1 Break Location S ectrum Reference 9.6

6. 1.2 Break Size S ectrum Reference 9.6

6. 1.3 MAPLHGR Anal ses for XN-1 9x9 Fuel Reference 9.7 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 Discharge Coefficient Bundle Average Peak Clad Peak Local Exposure MAPLHGR Temperature MWR GWD MT ~Percent 0 10.2 2060 3.9 5 10.2 2069 3.7 10 10.2 2121 3.7 15 10.2 2140 4.8 20 10.2 2147 5.2 25 '9.6 2016 2.7 30 '8.9 1839 1.0 35 8.2 1752 0.7 40 7.5 1675 0.5

12 XN-NF-86-60 6.2 Control Rod Oro Accident Reference 8. 1 Dropped Control Rod Worth, mk 8.3 Doppler Coefficient, 1/k dk/dT -9.5 "x(10) -6 Effective Delayed Neutron Fraction 0.0045 Four-Bundle Local Peaking Factor 1.35 Haximum Deposited Fuel Rod Enthalpy, cal/gm 147

13 XN-NF-86-60

7. 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

7. 1.2 Steam Dome Pressure Safet Limit Pressure Safety Limit (as measured in steam dome) 1325 psig Analysis shows that a steam dome pressure of 1359 can be allowed but the 1325 psig value used in Cycle 1 is to be conservatively 'retained.

7.2 Limitin Conditions for 0 eration 7.2. 1 Avera e Planar Linear Heat Generation Rate Limits for XN-1 9x9 Fuel Bundle Average Exposure HAPLHGR GWD MT ~kw ft 0 10.2 5 10.2 10 10.2 15 10.2 20 10.2 25 9.6 30 8.9 35 8.2 40 7.5

14 XN-NF-86-60 7.2.2 Minimum Critical Power Ratio MCPR Operating Limits at Rated Conditions Rod Block Settin MCPR 0 eratin Limit 106% 1.23 108% 1.27 MCPR Operating Limits at Off-Rated Conditions At 100% Power/108% Flow Rod Block Settin MCPR 0 eratin Limit 106% 1.29 108% 1.32 At Reduced Flow Figure 5.4 Power Dependent MCPR Operating Limit Nominal Feedwater Tem FFTR GE PSxSR ENC 9x9 GE PSx8R ENC 9x9 100/100 1.20 1.21 1.22 1 ..23 80/100 1.28 1 ~ 30 1,26 1.28 65/100 1.29 1.31 1.30 1.32 40/100 1.32 1.35 1.32 1.35 100/108 1.21 1.22 1.22 1.23 80/108 1.26 1.28 1.26 1.28 65/108 1.29 1.31 1.30 1.32 40/108 1.33 1.36 1.32 1.36

15 XN-NF-86-60 7.2.3 LHGR Limits LHGR Limits Figure 3.3 and 3.4 of Reference 9.2 7.3 Surveillance Re uirements 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.

7.3.2 Stabilit Surveillance Stability surveillance established to provide assurance of stable operation during Cycle 1 shall be continued during Cycle 2.

• 65'F reduction in Feedwater Temperature.

16 XN-NF-86-60 8.0 METHODOLOGY REFERENCES See XN-NF-80-19, Volume 4 for complete bibliography.

17 XN-NF-86-60 9.0 ADDITIONAL REFERENCES

9. 1 "Demonstration of 9x9 Assemblies for BWRs", EPRI NP-1580-5, Electric Power Research Institute, Palo Alto, California (May 1984).

9.2 "Generic Mechanical Design for (Apri 1 1986) .

R . 1, E Exxon Nuclear R 1 p, Ill Jet Pump BWR hl Reload I, W Fuel",

hip 9.3

~X- ->>,

"Exxon Nuclear Plant Transient (November 1981) .

R 11 I., E Methodology 1

for Boiling p, 1 Water Reactors",

hl d, W hl 9.4 "Susquehanna Unit 2 Cycle 2 Plant Transient Analysis", XN-NF-86-55, Exxon Nuclear Company, Richland, Washington (HAY 1986).

9.5 "A Generic phil W (February 1986).

R ",X~,E Analysis of the Loss of Feedwater 1 p,Transient Heating 1hl d,ll For 9.6 "Generic LOCA Break Spectrum Analysis BWR 3 & 4 with Modified Low Injection Logic Using the EXEH Evaluation Model",

Pressure 1984).

Coolant E R 1 1, lll hl d, hl E P 9.7 "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for ENC 9x9 Fuel",

XN-NF-86-65, Exxon Nuclear Company, Richland, Washington (May 1986).

I POWER VERSUS F'LOW 1.2 ENC 9x9 GE 8x8R 4

I 0.8 o.e 4

V) 0.0 0.2

>C 5 7 ASSEMBLY PONER (%) I CO C)

Figure 3.1 Hydraulic Demand Curves For Susauehanna Unit 2 Cycle 2 Core

DESIGN BRSIS RRDIRL POWER DISTRIBUTION 90 80 70 60 C3 50 4

40 30 20 10 0.2 0.0 0.6 0.8 1 1.2 1.6 1.8 RRDIRL PERKING FRCTOR Figure 3.2 Susquehanna Unit 2 Cycl Safety Limit Radial Power Histogram

20 XN-NF-86-60 0.88  : 0.91  : 0.97  : 1.04  : 1.03  : 1.04  : 0.97  : 1.01  : 0.97 0.91  : 0.94  : 0.98  : 0.94  : 1.05  : 0.93  : 0.99  : 0.94  : 1.01 0

0.97  : 0.98  : 0.90  : 1.04  : 1.03  : 1.05  : 1.04  : 0.97  : 0.97 1.04  : 0.94  : 1.04 : 1.00  : 1.00  : 1.01 : 1.05  : 1.07 : 1.04 1.03  : 1.05  : 1.03 : 1.00  : 0.00  : 0.97 : 1.05 : 1.06 : 1.04 1.04  : 0.93  : 1.05 : 1.01  : 0.97  : 0.00 : 1.02 : 0.95 : 1.05

\

0.97  : 0.99  : 1.04 : 1.05  : 1.05  : 1.02 : 1.06 : 1.00 : 0.97 0

1.01  : 0.94  : 0.97 : 1.07  : 1.06  : 0.95 : 1.00 : 0.95 : 1.02 0

0.97  : 1.01  : 0.97 : 1.04  : 1.04  : 1.05 : 0 '7  : 1.02 : 0.97 FIGURE 3.3 DESIGN BASIS LOCAL POHER DISTRIBUTION ENC XN-1 9X9 fUEL

• Rod adjacent to control blade location.

21 XN-NF-86-60

~ ~

1.03  : 1.00  : 0.99  : 0.99  : 0.99  : 0.99  : 1.00 : 1.03 1.00  : 0.97 0.99 1.02  : 1.03 : 1.03  : 0,99  : 1.00 0.99  : 0.99  : 1.02  : 1.01  : 1.02 : 0.91  : 1.03 : 0.99 0.99  : 1.02  : 1.01  : 0.91  : 0.00 : 1.02  : 1.02 0 99 0.99  : 1.03  : 1.02  : 0.00  : 1.02 : 1.01  : 0.99  : 0.99 0

0.99  : 1.03 0.91  : 1.02  : 1.01 0.98 : 0.99  : 0.99

\

1.00  : 0'.99 1'.03 : 1.02  : 0.99 : 0.99 : 0.97  : 1.00 1.03 1.00  : 0.99  : 0.99  : 0.99 : 0.99 : 1.00 1.03 FIGURE 3.4 DESIGN BASIS LOCAL POWER DISTRIBUTION G.E, 8XBR FUEL

• Rod adjacent to control blade location.

22 XN-NF-86-60 Table 4.1 Neutronic Design Values Core Data Number of fuel assemblies 764 Rated thermal power, NW 3293 Rated core flow, Hlbm/hr 100.0 Core inlet subcooling, BTU/ibm 24.0 Hoderator temperature, F 549 Channel thickness, inch 0.080 Fuel assembly pitch, inch 6.0 Water gap thickness, inch 0.562 Narrow water gap thickness, inch 0.562 Control Rod Data Absorber material B4C Total blade span, inch 9.75 Total blade support span, inch 1.58 Blade thickness, inch 0:26 Blade face-to-face internal dimension, inch 0.20 Absorber rods per blade 76 Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70.0

23 XN-NF-86-60

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

• 4 L: HL H: H: MH H* M: ML HL

• : HL  : H : H*  : H  : H H  : HH  : H* HL

• 4 H: H H: H: H H: H: MH M M: MH H: H W: MH: H HH: H H*: H: H MH W: HH M* o o

~

ML H: HH H: H: MH HH H 'L HL HL : H* MH  : MH  : H* HL  : ML HL 'L H HL': HL LL Rods ( 1) 1.45 w/o U235 L Rods.( 5) 1.95 w/o U235 HL Rods (18) 2.58 w/o U235 M Rods (20) 3.27 w/o U235 MH'Rods (13) 4.18 w/o U235 H Rods (15) 4.68 w/o U235 H* Rods ( 7) 3.27 w/o U235 + 4.00 w/o Gd203 W Rods ( 2) Inert Water Rods Figure 4. 1 Susquehanna Unit 2 XN-1 3.42 w/o Central Enrichment Distribution

24 X N- NF 60

A1 : A1  : CO  : A1  : CO : A1  : CO : A1 ". CO : A1  : CO  : A1  : CO  : A1  : A1

A1  : CO  : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : AI  : CO  : A1

CO  : A1  : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : A1

A1  : CO  : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1

CO  : A1  : CO  : A1  : CO  : A1  : CO : A1 : CO : A1 : CO : A1 : CO : A1 : A1

A1: CO: A1: CO: A1 'O: A1: CO: A1: CO: A1: CO: A1: CO: AI

~ P

CO  : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : Al

A1  : CO*. A1  : CO  : A1  : CO  : A1  : CO : A1 : CO : A1 : CO : A1 : A1

CO  : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 : CO : A1 A1

A1: CO: A1: CO: A1: CO: A1: CO: A1: CO: A1: A1: B1

CO  : A1  : CO  : A1 : CO : A1 : CO : A1 CO  : A1 : A1 0

Al CO  : A1: CO: A1: CO: A1: CO: Al: Al

CO:A1:CO:A1:CO:A1: CO:A1:A1:B1

A1: CO  : A1: CO: A1: CO  : A1: A1

A1  : A1  : A1  : A1  : A1  : A1  : A1 : XY = Fuel Type X Burned Y Cycles

~Fuel T e No. of Bundles Descri tion A 432 G 8x8 Type III 2.19 w/o U-235 B 8 GE 8x8 Type II 1.76 w/o U-235 C 324 ENC 9x9 3.31 w/o 7Gd4.0 Figure 4,2 Susquehanna Unit 2 Cycle 2 Reference Core Loading

25 XN-NF-86-60 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 55 14 -- 28 14 r55 51 28 5) 47 14 -- 06 -- 12 06 14 43 28 28 -- 43 39-- 14 06 . -- 16 -- 28 -- 16 06 -- 39 35-- -- 35 3.1 28 -- 12 -- 28 -- 00* -- 28 12 -- 28 -- 31 27 -- 77 23 14 06 -- 16 -- 28 -- 16 06 19-- 28 28 ~ - l9 15 06 12 -- 06 14 15 11 28 lI 07 P8 -- 14 07 03 03 02 06 .10 14 18 22 26 30 34 38 46 .0 54 58 Control Rod Being withdrawn Rod Position in Notches Withdrawn lull in = 00 Full out =--

Figure 5. 1 Susquehanna Unit 2 Cycle 2 Control Rod Withdrawal Error Analysis Initial Control Rod Pattern For 106 RBH Setting 100% Flow Case

26 XN-NF-86-60 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 55 08 ,-- 04 -- 08 55 51 51 47 08 -- 10 06 10 08 47 43-- 40 36 36 40--- -- 43 39-- 08 10 -- 16 08 16 10 08 -- 39 35-- 36 34 36 -- 35 31 04 06 -- 08 20 08 06 04 -- 31 27-- 36 36 -- 27 23-- 08 10 -- 16 00~ -- 16 10 08 -- 23 19-- 40 40 19 15 08 -- 10 06 -- 10 08 15 11 11 07 08 -- 04 -- 08 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58

• Control Rod Being withdrawn Rod Position in Notches lli thdrawn Full in =- 00 Full out =--

Figure 5.2 Susquehanna Unit 2 Cycle 2 Control Rod llithdrawal I'.rror Analysis Initial Control Rod Pattern For 108 RBII Setting 100% Flow Case.

27 XN-NF-86-60 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59

.55 00 -- 16. -- 00 55 51 16 16 51 47 00 06 -- 08 -- 06 00,-- 47 43 16 16 -- 43 39-- 00 06 -- 08 -- 10 -- 08 06 00 -- 39 35-- -- 35 31 16 -- 08 -- 10 -- 00* -- 10 08 -- 16 -- 31 27-- -- 27 23-- 00 06 -- 08 -- 10 -- 08 06 00 23 19-- 16 16 19 15 00 06 -- 08 -- 06 00 I5 ll

~

ll 16 00 -- 16 -- no 16 07 07 03 03 02 06 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 out =--

Figure 5.3 Susquehanna Unit 2 Cycle 2 Control Rod 'llithdrawal Frror Analysis Initial Control Rod Pattern 108% Flow Case

NOTE: The MCPR operating limit shall be the maximum of this curve and the full flow NCPR operating limit or power dependent HCPR operating limit.

1.4

~ p E

1.3 5-CD 1.2 1.1 OC Total Core Recirculating Flow (X Rated) I ll Figure 5.4 Reduced Flow HCPR Operating Limit I CO CXl I

Ch CD

A-I XN-NF-86-60 APPENDIX A SUS UEHANNA UNIT 2 SINGLE-LOOP OPERATION WITH ENC 9X9 FUEL Analyses have been performed for Susquehanna Units I and 2 for normal two pump operation both by the NSSS vendor and Exxon Nuclear Company (ENC). Generally, both analyses showed similar'esults and yielded comparable allowed operating limits. Since the ENC and vendor 8x8 fuel designs are very similar, this result is to be expected. ENC analysis for 8x8 and 9x9 fuel in the Susquehanna reactors justifies modification of 8x8 calculated HAPLHGR limits to be appropriate for 9x9 fuel on an equal planar power basis.

The ability to operate the Susquehanna reactors with only one recirculation pump running is highly desirable in the event that a recirculation pump or other component maintenance renders one loop inoperative. In order to justify single-loop operation, the NSSS vendor has performed additional accident and transient analyses for single-loop operating conditions (Reference A. 1). The single-loop operation. analysis generally showed that operation within the full-power two pump operating limits will assure that the safety limit is not violated and that substantial margin to the safety limit exists for single-loop operation due to the reduced power. For these cases, ENC fuel will likewise experience the benefit of the power reduction and application of two pump full-power limits for the ENC fuel designs i.s conservative and appropriate. This Appendix discusses appropriate limits For Susquehanna Unit 2 Cycle 2 operation with ENC 9x9 fuel and their bases.

A.I ROD WITHDRAWAL ERROR The rod block system is designed to stop rod withdrawal at a minimum critical power ratio (HCPR) higher than the fuel cladding safety limit. For

A-2 XN-NF-86-60 single-loop operation, a procedure has been established for correcting the APRH rod block equation to account for the discrepancy between actual flow and indicated flow in the active loop. This procedure preserves the original relationship between APRH rod block and actual effective drive flow when operating with a single-loop. The APRH scram trip settings are flow biased in the same manner as the APRM rod block setting. Modification to the rod block equation and lower power assures the HCPR safety limit is not violated. This applies for both 8x8 and 9x9 ENC fuel designs.

A.2 TRANSIENT MCPR LIMITS Operating with one recirculation loop results in a maximum power output which is about 25% below that which is attainable for two pump operation.

Therefore, the NSSS vendor single-loop analysis showed that the consequences of abnormal operation transients will be considerably less severe than those analyzed from a two-loop operational mode. These results are shown in Table 15.c.3-3 of Reference A.l. The limiting transients from an allowed HCPR operating limit of 1.38 gave transient MCPRs of 1.20-1.21 which are well above the GE safety limit of 1.07 with a 0 13 0 14 margin in CPR. For pressurization, flow increase, flow decrease, and cold water injection transients, results for two-loop operation bound both the thermal and overpressure consequences of one-loop operation. It was concluded that the HCPR operating limits established for two-pump operation are also applicable to single-loop operation conditions. This is true even for the increased safety limit associated with single-loop operation (see A.4).

The increased HCPR margin for single-loop operation at reduced power is also applicable to ENC fuel designs. Therefore, the operating HCPR limits established for two-pump operation with ENC fuel will be conservative when applied to single-loop operation for the same reasons as for the vendor fuel, This applies for both 8x8 and 9x9 fuel designs. Applicability of two-pump limits for single pump operation is discussed phenomenologically in the following section.

A-3 XN-NF-86-60 A.3 ABNORMAL OPERATING TRANSIENTS HCPR limits established for full flow two loop operation are conservative for single loop operation because of the physical phenomena related to part-power part-flow operation, not because of features in reactor analysis models or compatible fuel designs. A review of the most limiting delta CPR transients for single-loop operation was conducted. Under single-loop conditions, steady state operation cannot exceed approximately 75% power and 60% core flow because of the capability of the recirculation loop pump. Thus, the HCPR limit at maximum power is higher than the two-pump operating HCPR limit due to the flow dependent HCPR function. This flow dependence is based on a flow increase transient from runup of two pumps. Flow runups from a single recirculation pump would be much less severe but the conservative two pump limit is retained.

A.3. I Load Re'ection Without 8 ass The limiting system transient for the Susquehanna Units is the Load Rejection Without Bypass (LRWB) pressurization transient. In this transient, the primary phenomena is the pressurization caused by abruptly stopping the steam flow through rapid closure of the turbine control valve. When the rapid pressurization reaches the core it causes a power excursion due to void collapse.

At reduced power and flow there is a corresponding reduction in steam flow.

With lower steam flow the maximum pressurization of the core is reduced in comparison to rated conditions when the control valve is closed. The resulting power excursion and associated delta CPR are reduced below those of the full power full flow case.

Thus the HCPR limits based on LRWB analyses at full power are conservatively applicable to the lower powers associated with single loop conditions based on the physics of the transient. Furthermore, LRWB analyses by GE 'nd

A-4 XN-NF-86-60 preliminary ENC analyses at reduced power and flow conditions with two-loop operation confirm this trend, and GE analyses 'nder single-loop conditions also confirm this trend.

A,3.2 Feedwater Controller Failure The second most limiting transient for Susquehanna is the Feedwater Controller Failure (FWCF). This transient is also less severe at the reduced power and

'flow conditions associated with single-loop operation.

This transient assumes the feedwater controller fails to maximum demand and allows the maximum amount of subcooled feedwater into the downcomer. When this cooler water reaches the core, the power rises. The core power rise is terminated through a turbine trip scram initiated 'by a high water level trip in the downcomer due to the additional amount of feedwater being injected.

At the reduced recirculation flows the subcooling in the downcomer due to high feedwater injection takes longer to transverse to the core such that a high level trip occurs before the core power rises as much as in the full power case. In the subsequent pressurization transient, the result of turbine trip is less severe for the reduced powers in transients from single-loop conditions because of the reasons discussed in the LRWB transient.

Thus, because of the slower transport phenomena caused by the lower flow in the downcomer'nd because of the lower steam line flow in the pressurization portion of the transient, and the higher full-power MCPR limit, the FWCF has larg'er margin to the operating limit in single-loop operation than in full-power two-loop operation.

A-5 XN-NF-86-60 A.3.4 ~Summar It is very conservative to use the reduced flow two-loop operating HCPR limit for single-loop operations. The reduced flow HCPR limit .is to protect against boiling transition during flow excursions to maximum two-pump flow; excursions to such high flows are not possible during single-loop one-pump operation.

Thus, conservatively maintaining this two-loop limit assures that there is even more thermal margin under single-loop conditions than under two-loop full power full flow conditions.

A.4 SAF TY LIMIT MCPR For single-loop operation, the NSSS vendor found that an increase of 0.01 in the HCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation. ENC has evaluated the effects of the increased flow measurement uncertainties on the safety limit HCPR and found that the NSSS vendor determined increase in the allowed safety limit HCPR is also applicable to ENC fuel during single-loop operation. Thus, increasing the safety limit HCPR by 0.01 for single-loop operation (1.07) with ENC fuel is sufficiently conservative to also bound the increased flow measurement uncertainties for single-loop operation.

A.5 MAPLHGR LIMITS The NSSS vendor has also evaluated the changes in the two-loop MAPLHGR limits required to permit single-loop operation. A multiplier of 0.81 is to be applied to the appropriate two-loop MAPLHGR limit to obtain the MAPLHGR limit for single-loop operation. The need to reduce the allowed HAPLHGR arises because of the conservative assumption of early boiling transition (at 0. I sec) in the LOCA-ECCS analysis applied for single-loop operation at reduced core flow.

A-6 XN-NF-86-60

'To support operation of Susquehanna Unit 2 with Exxon Nuclear Company (ENC) 9x9 fuel with a single recirculating pump operating, the GE MAPLHGR limits for the highest enriched GE 8x8R fuel design with a multiplier of 0.81 are to be applied on an equal planar power basis to ENC 9x9 fuel for single-loop operation. The basis for this is two-fold:

1) The phenomena which require the reduction in MAPLHGR limits are a result of operation of the Susquehanna Unit 2 system with single active recirculation loop, and are therefore, equally applicable to both GE and ENC fuel design's, and

2) For the expected exposures during Cycle 2 operation the analysis methods used by GE have yielded conservative MAPLHGR limits relative to the MAPLHGR limits obtained using the ENC approved analysis models. Therefore, applying the more conservative GE MAPLHGR limit to ENC fuel provides a limit which assures conformance to NRC 10 CFR 50.46 criteria.

The major difference between operation with both recirculation pumps running and operating with only one active recirculation pump are reduced operating core flow, reduced core power, and reverse flow through the inactive loop jet pumps. Flow dependent MCPR limits assure reduced maximum assembly power during single-loop operation. The primary system coolant inventory and LOCA break conditions're essentially unchanged from the two-loop operation. Thus, the uncovery of the jet pump suction, recirculation suction line uncovery, and system depressurization rate would be expected to change little between one and two-loop operation. The phenomena associated with these key parameters largely determine LOCA analysis results for both ENC and GE analyses. The analyses performed by GE confirm this system behavior in that the limiting pipe break LOCA is essentially unchanged from the two-loop analysis, as are the break size and core uncovery and reflood times. Although ENC LOCA analysis methods differ from those of GE, similar results would be expected

A-7 XN-NF-86-60 from an ENC analysis because the phenomena are governed by the system parameters.

The principal LOCA concern associated with single-loop operation is the possibility of the LOCA break occurring in the operating loop, in which case there is no coastdown of an intact loop recirculation pump to sustain jet pump and core flow during the early portion of the system blowdown. An early boil.ing transition (CHF) may result "from this early loss of flow capability.

To account for this possibility, GE derived a single-loop operation MAPLHGR multiplier of 0.81 to be used with calculated two-loop MAPLHGR limits during single-loop operation. The analyses which determined this- multiplier assumed a near instantaneous boiling transition (0.1 sec) even though a longer boiling transition time may have been calculated using approved models. This assumption is very conservative when applied to the GE fuel and would be even more conservative when applied to ENC 9x9 fuel because of lower stored energy in 9x9 fuel.

The major difference between the ENC and GE methodologies that would effect analysis differences between single and two-loop operation is in the blowdown heat transfer. ENC's more mechanistic model calculates boiling transition times that are equivalent to or later than those reported from the GE model, and the ENC model explicitly calculates the blowdown heat transfer throughout the blowdown period while the GE model assumes an adiabatic heatup period.

Thus, the conservative approach taken in the GE analysis of assuming an early boiling transition (O. 1 sec) for single-loop operation would yield a greater penalty using ENC methodology than for the more conservative GE methods. For this reason, limits based on the more conservative GE analysis are recommended. ENC's more mechanistic heat transfer during the GE adiabatic heatup period would partially offset this effect, thus, risking the recommended limits conservative.

A-8 XN-NF-86-60 Application of 'GE calculated 8x8 MAPLHGR limits modified on equal planar power basis for ENC 9x9 fuel for single-loop operation will conservatively assure that the NRC criteria of 10 CFR 50.46 will be met for the following reasons:

1) Since ENC has performed LOCA analyses for a number of BWRs under two-loop operation and MAPLHGR limits for ENC 8x8 fuel are higher than the equivalent GE 8x8 fuel limits in all cases for bundle exposures less than 19,000 MWD/Mt, an ENC analysis for the similar single-loop operating conditions would be expected to also yield MAPLHGR limits equal to or higher than those obtained by GE.

2) The MAPLHGR reduction factor to protect against early boiling transition determined by GE is based on a conservative early boiling transition assumption which is even more conservative when applied to 9x9 fuel.

3) ENC analysis for two-loop operation at expected exposures in Cycle 2 of Susquehanna Unit 2 with 9x9 fuel justifies MAPLHGR limits equal to or greater than the GE 8x8 design on an equivalent planar power basis. That is, 9x9 MAPLHGR limits are equal or greater than the GE 8x8 limits times the ratio of heated rods in the 8x8 assembly to heated rods in the 9x9 assembly. On this basis 8x8 MAPLHGR limits can be conservatively modified for application to 9x9 fuel.

for Cycle 2 of Susquehanna Unit 2 single-loop operation with ENC 9x9 fuel, a MAPLHGR limit corresponding to 0.81 times the MAPLHGR limits for the highest enriched Cycle 1 GE fuel type can be conservatively used. These 8x8 MAPLHGR limits are to be adjusted by the ratio (62/79) to be on an equal planar power basis for 9x9 fuel.

A-9 XN-NF-86-60 A.6 STABILITY Susquehanna Units I and 2 have adopted a detect and suppress approach to avoid unstable reactor operation. This is consistent with single-loop operation requirements stated in NRC Generic Letter 886-09 (Reference A.3). The detect and suppress criteria will be conservatively applicable to ENC 9x9 fuel in the Susquehanna Unit 2 reactor.

A-10 XN-NF-86-60 A.7 REFERENCES A. 1 General Electric Co., "Susquehanna Single-Loop Operation Analysis",

, GP84-142, General Electric Co., June 1984.

A.2 "Extended Load Line Limit Analyses for Susquehanna Steam Electric Station Unit 1", NED022128, General Electric Co., Hay 1982.

A.3 "Technical Resolution of Generic Issue No. B-59-(N-I) Loop Operation in BWRs and PWRs", (Generic Letter No. 86-09), Harch 31, 1986.

B-I XN-NF-86-60 APPENDIX B SEISMIC- LOCA EVALUATION The structural response of Exxon Nuclear's 9x9 fuel is the same as the structural response of the Bx8 fuel it replaces in the Susquehanna Unit 2 core. Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.

The physical and structural properties of the 9x9 and the 8x8 fuel types which are important to the dynamic response of the fuel are summarized in Table Bl.

The close agreement between the important parameters for the two fuel types indicates that the structural response would be very similar for both fuel types.

Similarity in the natural frequencies of the two fuel types is further assured by the stiffness of the fuel assembly channel box. Both fuel types use the same fuel assembly channel box, and the channel'ox dominates the overall dynamic response of the incore fuel. ENC 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. Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

B-2 XN-NF-86-60 TABLE B1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8x8 AND 9x9 FUEL ASSEMBLIES

~Pru ert ~Fuel T es ENC 9x9 ENC 8x8 GE Bx8R Assembly-Weight, lbs 580 596 600 Number of Spacers Overall Assembly Length, in 171.29 171.29 171.40 Assembly Frequencies, cps Mode 1 1.9 1.7 2 3.7 3.5 3 6.5 6.5 4 10.4 10.8 5 15.5 16.6 6 21.9 24.2 7 29.1 33.9

• GE proprietary.

C-I XN-NF-86-60 t

APPENDIX C INCREASED CORE FLOW (ICF) AND FINAL FEEDMATER TEHPERATURE REDUCTION (FFTR) PLANT TRANSIENTS RESULTS Load rejection without bypass, feedwater controller failure, and HSIV closure were evaluated at increased core flow and final feedwater temperature reduction combinations. The delta-CPR's are given in'ables C. 1 and C.2.

C-2 XN-NF-86-60 Table C.l RESULTS OF SYSTEM PLANT TRANSIENT ANALYSIS AT INCREASED CORE FLOW AND AT REDUCED FEEDWATER TEMPERATURE Load Re 'ection Without B ass

% Power/% Flow Maximum Maximum Maximum System Delta Neutronic Flux Core Average Pressure (psia) CPR

(% Rated) Heat Flux

(% Rated) 100/100 (FFTR) 253 112.9 1191 0.15 100/108 (NFT) 241 112.1 1210 0.17 100/108 (FFTR) 222 110.8 1187 0.15 ASME Over ressure MSIV Closure si Vessel Lower Vessel Dome Plenum Steam Line 100/100 (FFTR) 1264 1279 1265 100/108 (NFT) 1290 1307 1296 100/108 (FFTR) 1257 1274 1259

• Final Feedwater Temperature Reduction (65'F).

• • Nominal Feedwater Temperature.

C-3 XN-NF-86-60 Table C.2 ~

FEEDWATER CONTROLLER FAILURE DELTA CPR OF ICF AND FFTR ANALYSIS Nominal Feedwater Tem . FFTR

% Power % Flow GE P8xSR ENC 9x9 GE PSxSR ENC 9x9 100 / 100 0.14 0.15 0.16 0.17 80 / 100 0.22 0.24 0.20 0.22 65 / 100 0.23 0.25 0.24 0.26 40 / 100 0.26 0.29 0.26 0.29 100 / 108 0.15 0.16 0.16 0.17 80 / 108 0.20 0.22 0.20 0 '2 65 / 108 0.23 0.25 0.24 0.26 40 / 108 0.27 0.30 0.26 0.30

• 65'F reduction in Feedwater Temperature.

I I

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XN-NF-86-60 Issue Date: 5yI5g86 SUSQUEHANNA UNIT 2 CYCLE 2 RELOAD ANALYSIS Design and Safety Analyses Distribution D.J. Braun J.C. Chandler R.E. Collingham S.F. Gaines K.D.'Hartley S.E. Jensen T.H. Keheley J.E. Krajicek T.L. Krysinski J.N. Morgan L.A. Nielson

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