ANF-87-1 26
REVIStON 1
AD~MHCSDo HUCIt.EARFUSM CORPORATION
SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS
DESIGN AND SAFETY ANALYSES.
NOVEMBER 1987
87i2310i58 87i223POR ADOCK 0500058] ~
ANAFFII.IATEOF KRAFTWERK UNION
Q~ KRU
ADVANCEDNUCLEARFUELS CORPORATION
ANF-87-126Revision 1
Issue Date: 11/25/87
SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS
Design and Safety Analyses
Prepared By:J. A. White
BWR Safety AnalysisLicensing and Safety Engineering
Fuel Engineering and Technical Services
AIIAFFILIATEOF KRAFTWERK UNION
Qxsvu
CUSTOMER DISCLAIMER
IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THISDOCUMENT
PLEASE READ CAREFULLY
Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set forth in the Agreementbetween Advanced Nuclear Fuels Corporation and the Customer pursuant towhich this document is issued. Accordingly, except as otherwise expressly pro-vided In such Agreement, neither Advanced Nuclear Fuels Corporation nor anyperson acting on its behalf makes any warranty or representation, expressed orimplied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained In this document, or that the use of any information, apparatus,method or process disclosed ln this document will not infringe privately ownedrights: or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.
The information contained herein is for the sole use of Customer.
In order to avoid Impairment of rights of Advanced Nuclear Fuels Corporation inpatents or inventions which may be included in the information contained in thisdocument, the recipient, by its acceptance of this document, agrees not topublish or make public use (in the patent use of the term) of such information untilso authorized in writing by Advanced Nuclear Fuels Corporation or until after six(6) months following termination or expiration of the aforesaid Agreement and anyextension thereof, unless otherwise expressly provided in the Agreement. Norights or licenses In or to any patents are implied by the furnishing of this docu-ment.
XN NF F00.765 (1
ANF-87-126Revision 1
TABLE OF CONTENTS
Section
1.0
2.0
Pacae
INTRODUCTION. ~ ..............,....,................................ 1
FUEL MECHANICAL DESIGN ANALYSIS................................... 2
3.03.23.2.13.2.33.2.5
THERMAL HYDRAULIC DESIGN ANALYSIS.............. ~.................. 3
Hydraul i c Characteri zati on........................................ 3
Hydraul i c Compatibility........................................... 3
Fuel Centerline Temperature....................................... 3
Bypass Flowe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3
3.33.3.13.3.2
.3.34.0
MCPR Fuel Cladding Integrity Safety Limit...........Coolant Thermodynamic ConditionsDesign Basis Radial Power Distribution.Design Basis Local Power Distribution.NUCLEAR DESIGN ANALYSIS..
~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ 3
3
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4
~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5
4.1
4.24.2.14.2.24.2.45.0
Fuel Bundle Nuclear Design Analysis.......Core Nuclear Design AnalysisCore Configuration.......Core Reactivity Characteristics...,....,..Core Hydrodynamic Stability..... ~ .
ANTICIPATED OPERATIONAL OCCURRENCES......,
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5
~ ~ ~ ~ ~ ~ ~ 5
5
6
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7
5.1
5.2
5.3
5.45.55.65.76.06.1
F 1.1
itlons ~ ~ ~ ~ ~ ~ ~ ~ ~ 7
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9
10
10Loss-Of-Coolant Accident....,,......Break Location Spectrum........ 10
Analysis Of Plant Transients At Rated Cond
Analyses For Reduced Flow Operation.......Analyses For Reduced Power Operation......ASME Overpressurization Analysis..........Control Rod Withdrawal Error (CRWE)
Fuel Loading Error........Determination Of Thermal Margins..........POSTULATED ACCIDENTS...
ANF-87-1Revision
TABLE OF CONTENTS
(Continued)
Section
6.1.26.1.36.2
7.0
7.1
T.1.17.1.27.2
7.2.17.2.27.2.3737.3.17.3.28.0
Limiting Safety System Settings......HCPR Fuel Cladding Integrity Safety L
Steam Dome Pressure Safety LimitLimiting Conditions For Operation.Average Planar Linear Heat Generation
Minimum Critical Power Ratio
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
imit
Rate imits.................L'
HGR Llmlts ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~
Surveillance Requirements...... .....Scram Insertion Time Surveillance....Stability Surveillance..........METHODOLOGY REFERENCES..........
9.0 ADDITIONAL REFERENCES.......
reak Size Spectrum...............................................B
APLHGR Analyses.............................'.....................H
Control Rod Drop Accident........,...TECHNICAL SPECIFICATIONS........
Pacae
10
10
11
12
12
12
12
12
12
1
1
14
14
14
15
16
APPENDICES
A. SINGLE LOOP OPERATION............. A-1
B. SEISMIC-LOCA EVALUATION....,.................,.........,.......... B-1
ANF-87-126Revision 1
LIST OF TABLES
Table Pacae
4. 1 Neutronic Design Values........................................... 23
B. 1 Comparison Of Physical And Structural CharacteristicsFor 8x8 And 9x9 Fuel Assemblies......................... .. .. . B-2
LIST OF FIGURES
Ficiur e
3.1 Susquehanna Unit 2Power vs. Flow....
Cycle 3 Hydraulic Demand Curve
Pacae
17
3.23.3
3.5
Susquehanna Unit 2
Design Basis Local
Design Basis Local
Design Basis LocalFuel.
Cycle 3 Design Basis Radial Power.............. 18
Power Distribution - ANF XN-2 9x9 Fuel......... 19
Power Distribution - ANF XN-1 9x9 Fuel......... 20
Power Distribution - GE 8x8R (Central)21
3.6
4.1
4.2
4.3
4.45.1
5.2
~ ~ ...... 22
24
~ ~ ~ ~ ~ ~ ~ ~ 25
~ ~ ~ ~ ~ ~ ~ ~ ~ 26
27
28
29
Design Basis Local Power Distribution - GE (Peripheral)8x8R Fuel.......... ~ ~ 0 ~ ~ ~ ~
Susquehanna Unit 2 Cycle 3 Enrichment Distribution ForANF92-344L-9G4 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Enrichment Distribution ForANF92-344L-10G5 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Reference Core Loading Plan...Susquehanna Unit 2 Cycle 3 - Core Power vs. Core Flow......Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal ErrorAnalysis Limiting Initial Control Rod Pattern..Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit.......
~ fj
ANF-87-126Revision 1
1.0 INTRODUCTION
This report provides the results of the analyses performed by Advanced NuclearFuels Corporation (ANF)* in support of the Cycle 3 reload for Susquehanna Unit2, which is scheduled to commence operation in the spring of 1988. This
report is intended to be used in conjunction with ANF topical report~XN-Np- -191 A, 91 4, R 11 1, Nppti 1 1 1 1 N
Company Methodology to BWR Reloads," which describes the analyses performed insupport of this reload, identifies the methodology used for those analyses,and provides a generic reference list. However, LHGR mechanical design limits(Reference 9. 1) and plant transient simulation model developments (Reference9.141 b 1 dbyANF b 4 t NRN P 1 F~F-Volume 4, Revi'sion 1. Both References 9. 1 and 9.2 have been approved by theNRC for use in referencing in license applications. Section numbers in this
9 t 1 9 dtd tt b 1 X-N- - fNJ,olume 4, Revision 1.
The Susquehanna Unit 2 Cycle 3 core will comprise a total of 764 fuelassemblies, including 236 unirradiated ANF XN-2 9x9 assemblies, 324 irradiatedANF XN-1 9x9 assemblies, 112 irradiated General Electric 8x8R fuel assemblies
(central region), and 92 irradiated GE 8x8R assemblies in the peripheralregion. The reference core configuration is described in Section 4.2.
The design and safety analyses reported in this document were based on thedesign and operational assumptions in effect for Susquehanna Unit 2 during theprevious operating cycle. Additional information and the results of designstudies covering the development of 9x9 fuel assemblies for BWR reloads arecontained in Reference 9.3.
f
ANF-87-126Revision 1
2.0 FUEL MECHANICAL DESIGN ANALYSIS
Applicable ANF Fuel Design Report: Reference 9. 1
To assure that the expected power history for the fuels to be irradiatedduring Cycle 3 of Susquehanna Unit 2 is bounded by the assumed power historyin the fuel mechanical design analysis, LHGR operating limits (Figure 3.3 ofReference 9. 1) have been specified. In addition, an LHGR transient
operating'imit
for Anticipated Operating Occurrences (Figure 3.4 of Reference 9. 1) has
been specified for ANF 9x9 fuel. Additional information on rod bow, as
requested in the NRC's safety evaluation report for Reference 9. 1, has been
transmitted in Reference 9.4.
ANF-87-126Revision 1
3.0 THERMAL HYDRAULIC DESIGN ANALYSIS
3.2 H draul ic Char aeter izat ion
3.2.1 H draulic Com atibilit
Component hydraulic resistances for the constituent fuel types in theSusquehanna Unit 2 Cycle 3 core have been determined in single phase flowtests of full scale assemblies. Figure 3. 1 shows the hydraulic demand curvesfor ANF 9x9 fuel and GE 8x8R fuel in the Susquehanna Unit 2 core. The similarhydraulic performance indicates compatibility for co-residence in 'heSusquehanna Unit 2 core.
Applicable Generic Report
3.2 ' Fuel Centerline Tem erature~ ~
Reference 9. 1
.2.2 ~21Calculated Bypass Flow Fractionat 104% Power/100% Flow
10.1%
3.3 MCPR Fuel Claddin Inte rit Safet Limit
Safety Limit MCPR = 1.06
3.3.1 Coolant Thermod namic Condition
Rated Thermal Power
Feedwater Flowrate (at SLMCPR)
Core Pressure (at SLMCPR)
Feedwater Temperature
3293 Mwt
16. 1 Mlbm/hr
1042.9 psia383'F
ANF-87-1Revision
3.3.2 Desi n Basis Radial Power Distribution
See Figure 3.2
3.3.3 Desi n Basis Local Power Distribution
See Figures 3.3 through 3.6
ANF-87-126Revision 1
4.0 NUCLEAR DESIGN ANALYSIS
4.1 Fue Bund e Nuclea Desi n Anal sis
Assembly Average Enrichment
Radial Enrichment DistributionAxial Enrichment Distribution
Burnable Poisons
Note: Burnabl e poi sons aredistributed uniformly over
the enriched length of thedesignated rods. The
natural urania axial blanketsections do not containburnable absorber material.
Non-Fueled Rods
Neutronic Design Parameters
3. 33%
Figure 4. 1 and 4.2
Uniform 3.44%
with 6" naturaluranium topblanketFigure 4. 1 and 4.2
Figure 4.1 and 4.2
Table 4. 1
4.2 Core Nuclear Desi n Analsis';2.
1 Core Confi oration Figure 4.3
Core Exposure at EOC2, HWd/HTU
Core Exposure at BOC3, MWd/HTU
Core Exposure at EOC3, HWd/MTU
Maximum Cycle 3 Licensing ExposureLimit, HWd/MTU
18350.7
10911.2
21740.8
22076
ANF-87-12Revision
4.2.2 ore Reactiv't Characteris ics
BOC Cold K-effective, All Rods Out
BOC Cold K-effective, Strongest Rod Out
1.11353
0.98524
Reactivity Defect (R-Value) 0.00% rho
4.2.4
Standby Liquid Control System Reactivity,Cold Conditions, 660 ppm
I
Core H drod namic Stabilit
0.98348
Power/flow Map Figure 4.4
Power Flow State Points
64/42*
69/47**
66/45**
Deca Ratio COTRA
0.82
0.75
0.75
*Two pump minimum flow - APRN Rod Block intercept point. Extended operationat lower flow is not allowed by Technical Specifications.
**Operation at less than 45% flow requires APRH/LPRN surveillance. Inaddition, operation at power/flow,combinations above and to the left of theline connecting these two points requires APRH/LPRtl surveillance. See Figbre4.4.
ANF-87-126Revision 1
5.0 ANTICIPATED OPERATIONAL OCCURRENCES
Applicable Generic TransientAnalysis Methodology Report References 9.5 5 9.7
5.1 Anal sis Of Plant Transients AtRated Conditions Reference 9.6
Limiting Transient(s): Load Rejection Without Bypass (LRWB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LFWH)
Event
LRWB
Power*
100%
FWCF 100% 100% 1 16. 8%
LFWH 100% 100% 121 '%
233%
123%
1179
1078
% Rated % Rated MaximumMaximum Maximum Pressure
Flow Heat Flux Power , ~aiaI
100% 116.2% 267% 1194
Del taCPR**
0.24
0.23
0.16
Model
COTRANSA/XCOBRA-T
COTRANSA/XCOBRA-T
PTSBWR3/XCOBRA
Single Loop Operation: Appendix A
5.2 Anal ses For Reduced Flow 0 eration Reference 9.6
Limiting Transient(s): Recirculation Flow Increase Transient (RFIT)
*104% power used in analysis as design bases.
**Delta-CPR results for most limiting fuel type.
ANF-87-1Revision
5.3 Anal ses For Reduced Power 0 eration Reference 9.6
Limiting Transient(s): Feedwater Controller Failure (FWCF)
% Power TransientDelta CPR
ANF 9x9 GE 8x8R
104
80
65
40
FWCF
FWCF
FWCF
FWCF
0.23
0.25
0.280.31
0.20
0.23
0.26
0.28
5.4 ASME Over ressurization Anal sis Reference 9.6
Limiting Event
Worst Single FailureMaximum Pressure
Maximum Steam Dome Pressure
Full MSIV
IsolationDirect Sera
1297 psig1281 psig
5.5 Control Rod Withdrawal Error CRWE
Starting Control Rod Pattern for Analysis Figure 5.1
Rod Block Settin
105
106*
107
108*
100% FlowDistanceWithdrawn~ft
4.04.55.0
5.0
DeltaCPR
0.22
0.24
0.26
0.26
*Rod Block Monitor settings recommended for Cycle 3 operation.
ANF-87-126Revision 1
5.6 Fuel Loadin Error
Maximum Delta CPR 0.16
5.7 Determination Of Thermal Har ins
Summary of Thermal Margin Requirements
Event
LRWB
FWCF
LFWH
CRWE
Power
1P0%**
1PP%**
1PP%9c*
100%
Flow
100%
100%
100%
100%
Delta CPR*
0.24
0.23
0.16
0.24 at 106% RBH0.26 at 108% RBM
MCPR Limit
1.30
1.29
1.22
1.301.32
HCPR Operating Limits at Rated Conditions
MCPR 0 eratin Limit
1.30 at 106% RBM
1.32 at 108% RBM
Reduced Flow MCPR Limits Figure 5.2
Power Dependent HCPR Operating Limit Results for Cycle 3:
100*+/100
80/100
65/100
40/100
LimitingTransient
LRWB
FWCF
FWCF
FWCF
ANF 9x9
1.30
1.31
1.34
1.37
GE 8x8R
1.27
1.29
1.32
1.34
,i
*Delta CPR results for most limiting fuel type.
**104% power used in analysis as design bases.
10 ANF-87-126Revi si on 1
6.0 POSTULATED ACCIDENTS
6.1 Loss-Of-Coolant Accident
Sei smi c- LOCA: Appendix B
6. 1. 1 Break Location S ectrum Reference 9.8
6. 1.2 Break Size S ectrum Reference 9.8
6. 1.3 MAPLHGR Anal ses
ANF 9x9 Fuel Reference 9.9
Limiting Break: Double-ended guillotine pipe break
Recirculation pump discharge line0.4 Discharge Coefficient
Bundle AverageExposureGWD MTU
05
10152025303540
MAPLHGR~kw ft
10.210.210.210.210.29.68.98.27.5
Peak CladTemperature*~F
206020692121214021472016183917521676
Peak LocalMWR**
~Percent
3.93.73.74.85.22.71.00.70.5
*Peak clad temperatures for XN-1 and XN-2 fuel are bounded by these results.
**Metal Water Reaction.
ll ANF-87-1Revision
6.2 Control Rod Dro Accident Section 8.0
Dropped Control Rod Worth, mk
Doppler Coefficient, 1/k dk/dT
Effective Delayed Neutron FractionFour-Bundle Local Peaking Factormaximum Deposited Fuel Rod Enthalpy, cal/gmNumber of Rods Exceeding 170 cal/gm
13.5
-10.6 x(10) 6
0.0058
1.34
205
(250
12 ANF-87-126Revision 1
7.0 TECHNICAL SPECIFICATIONS
7.1 Limitin Safet S stem Settin s
7.1.1 MCPR Fuel Claddin Inte rit Safet Limit
MCPR 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 safety limit of 1358 psigis allowed but the 1325 psig value used in Cycle 2 is to be
conservatively retained.
7.2 Limitin Conditions For 0 er ation
7.2. 1 Avera e Planar Linear Heat Generation Rate Limits
Bundle AverageExposure
GWD MT
0
5
10
15
20
25
30
35
40
MAPLHGR Limits kw ftANF 9x9 Fuel
10.2
10.2
10.2
10.2
10.2
9.68.98.27.5
13 ANF-87-1Revision
7.2.2 Minimum Critical Power Ratio
MCPR Operating Limits at Rated Conditions:
MCPR 0 eratin Limit
1.30 at 106% RBM
1.32 at 108% RBM
MCPR Operating Limits at Off-Rated Conditions:
At Reduced Flow Figure 5.2
Total CoreRecirculation Flow
% Rated
100
96
92
83
76
60
50
40
Reduced FlowMCPR
0 eratin Limit
1.12
1.14
1.16
1,20
1.23
1.31
1.44
1.61
At Reduced Power
Power Level% Rated
100*
80
65
40
Reduced PowerMCPR
0 eratin Limit
1.30
1.31
1.34
1.37
*104% power used in analysis as design bases.
ANF-87-126Revision I
7.2.3 LHGR Limits
LHGR Limits Figures 3.3 and 3.4 ofReference 9. 1
7.3 Surveillance Re uirements
7.3.1 Scram Insertion Time Surveillance
Thermal limits established in Section 5.0 are based on minimum acceptablescram insertion performance as defined in the Technical Specifications. No
additional surveillance for scram insertion is required for validation ofthermal limits.
.3.2 Stabilit Surveillance~ ~
Power/Flow Map Figure 4.4
The Unit 2 Cycle 2 Technical Specifications require APRM/LPRM surveillance tothe left of the 45% Constant Flow line and above the 80% Rod Block line.Based on core hydrodynamic stability analyses, operation at power/flowcombinations above and to the left of the line connecting the 66% Power/45%
Flow and 69% Power/47% Flow points but below the APRM Rod Block line needs tobe added to the APRM/LPRM surveillance requirement (see Section 4.2.4).
15 ANF-87-126Revision 1
8.0 METHODOLOGY REFERENCES
See XN-NF-80-19(P)(A), Volume 4, Revision 1 for complete bibliography.
\q, „"~r
0 ~ )
6 44-I
16 ANF-87-126Revision 1
9.0 ADDITIONAL REFERENCES
9. 1 "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"~X---F,R.X,Addll 1 FXCF l,lhhl d,Washington, September 4, 1986.
9.2 "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal111«N hd1 d P
Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington,January, 1987.
9.3 "Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric PowerResearch Institute, Palo Alto, California, Hay 1, 1984.
9.4 Letter, G. N. Ward (ANF) to G. C. Lainas (NRC), "Additional Informationon Rod Bow," serial no. GNW:021:87, dated March 11, 1987.
9.5 "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"~h-p--,h 11 X,AddN1F1C l,ltlhl d,Washington, November 16, 1981.
~
~
~
9.6 "Susquehanna Unit 2 Cycle 3 Plant Transient Analysis," ANF-87- 125,Rev. 2, Advanced Nuclear Fuels Corporation, Richland, Washington,November 1987.
9.7
9 ~ 8
"XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic CoreA 1 i,"X~, 1 »d2, Advanced Nuclear Fuels Corporation, Richland, Washington, February1987.
"Generic LOCA Break Spectrum Analysis BWR 3 8 4 with Hodified LowPressure Coolant Injection Logic Using the EXEH Evaluation Model," XN-NF-~84-117 P, Advanced Nuclear Fuels Corporation, Richland, Washington,December 1984.
9.9 "Susquehanna LOCA-ECCS Analysis HAPLHGR Results for ENC 9x9 Fuel," XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, May1986.
9. 10 "Principal Reload Fuel Design Parameters, Fuel Design, Susquehanna Unit 2Reload XN-2," XN-NF-1058, Advanced Nuclear Fuels Corporation, Richland,Washington, March 1987,
Formerly Exxon Nuclear Company.
Advanced Nuclear Fuels9x9
0~)
O~I
General Electric8x8R
~op coo
Cp0OI
Q~oLIp0 q>vCi
I.>oooIK
100.00 . 105.00 TIO.OO 115.M 120.00 125.00 QO.M Q5.M 140.00Assembly Flow Rate, KLB/HR
0
CO
%5.00 150.00
Figure 3. 1 Susquehanna Unit 2 Cycle 3 Hydraulic Oemand CurvePower vs. Flow
80
70
60
50
00C)
CL
So
20
10
00 0.2 0.0 0.6 0.8 1 1.2
RRDIFIL POHER PERKING
Figure 3.2 Susquehanna 2 Cycle 3 Oesign Basis Radial Power
19 ANF-87-126Revision 1
* ~
: 0.88 : 0.91 : 0.96 : 1.04 : 1.02 : 1.04 : 0.96 : 1.00 : 0.96 :* ~
** ~
: 0.91 : 0.93 : 0.98 : 1.07 : 0.91 : 1.07 : 0.97 : 1.04 : 1.01* ~
** ~
0.96 : 0.98 : 0.90 : 1.04 : 1.03 : 1.04 : 1.04 : 0.99 : 0.96 :
** ~
1.04 : 1.07 : 1.04 : 1.00 : 0.99 : 1,00 : 1.05 : 0.94 : 1.04
** ~
* : 1.02 :* ~
0.91 : 1.03 0.99 : 0.00 : 0.98 : 1.05 : 1.07 : 1.04
* ~
* ~
1.04 : 1.07 : 1.04 : 1.00 : 0.98 : 0.00 : 1.03 : 0.94 : 1.05
: 0.96 : 0 '7 : 1 '4 : 1.05 : 1.05 : 1.03 : 1.06 : 1.00 : 0,97
1.00 : 1.04 : 0.99 : 0,94 : 1.07 : 0.94 : 1.00 : 0.94 1.01
0.96 : 1.01 : 0.96 : 1.04 : 1.04 : 1.05 : 0.97 : 1.01 0.97
Figure 3.3Design Basis Local Power DistributionAdvanced Nuclear Fuels XN-2 9X9 Fuel
20 ANF-87-1Revisio
* ~
* ~ 0.91 : 0.92 : 0.95 : 1.01 : 1.01 : 1.01 : 0.96 : 0.98 : 0.95
* ~
* ~
* ~
0.92 : 0.94 : 0.98 : 0.97 ; 1.05 : 0.95 : 0.99 : 0. 95 : 0.98
* ~
* ~
* ~
*
0.95 : 0.98 : 0,93 : 1.06 : 1.05 : 1.06 : 1.05 : 0. 97 : 0.96
* ~
**
1.01 : 0.97 : 1,06 : 1.03 : 1,03 : 1.04 : 1.07 : 1. 06 : 1.02
* ~
* ~ 1.01 : F 05 : 1.05 : 1.03 : 0.00 : 1.01 : 1.07 : 1.06 : 1.01
1 01 ' 95 1.06 : 1.04 : 1.01 : 0 F 00 : 1.04 : 0.96 : 1.02
0.96 : 0.99 1.05 : 1.07 : 1.07 : 1.04 : 1.06 : 1. 00 : 0.96
0.98 : 0.95 : 0.97 : 1.06 : 1.06 : 0.96 1.00 : 0.95 : 0.98
0.95 : 0.98 : 0.96 : 1.02 : 1.01 : 1.02 : 0.96 : 0.98 : 0.96
Figure 3.4Design Basis Local Power DistributionAdvanced Nuclear Fuels XN-1 9X9 Fuel
21 ANF-87-126Revision 1
** ~
* ~
1.03 : 1.00 : 1.00 : 1.00 : F 00 : 1.00 : 1.01 : 1.03
* ~
* : 1.00 : 0.98 : 1.00* ~
*
1.02 : 1.02 : 1.03 : 1.00 : 1.01
* ~
* : 1.00* ~
** ~
1.00 : 1.01 : 1.01 : 1.01 : 0.90 : 1.03 : 1.00 :
* : 1.00 : 1.02* ~
*
1.01 : 0.89 : 0.00 : 1.01 : 1.02 : 1.00
* ~
* : 1.00* ~
1.02 1.01 : 0.00 : 0.89 : 1.01 : 0.99 : 1.00
* ~
* : 1.00* ~
*
1.03 : 0.90 1.01 : 1.01 : 0.98 : 1.00 : 1.00
1.01 : 1.00 : 1.03 : 1.02 0.99 : 1.00 : 0.98 : 1.00
1.03 : 1.01 : 1.00 : 1.00 : 1.00 1.00 : 1.00 : 1.03
Figure 3.5Design Basis Local Power Distribution
General Electric (Central) SXSR Fuel
22 ANF-87-'evisio
* ~
1.00 : 1,00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :* ~
0
* ~
1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :* ~
** ~
* 100* ~
** ~
1.00 ; 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 :
* : 1.00 : 1.00
** ~
1.00 : 1.00 : 0.00 : 1.00 : 1.00 1.00 :
1.00 : 1.00 : 1.00 : 0.00 : 1.00 : 1.00 : 1.00 1.00
** 100* ~
*
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 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00 : 1.00
Figure 3.6Design Basis Local Power Distribution
General Electric (Peripheral) SXSR Fuel
23 ANF-87-126Revision 1
TABLE 4. 1 NEUTRONIC DESIGN VALUES
Fuel Pellet Reference 9.10
Fuel Rod Reference 9.10
Fuel Assembl Reference 9.10
Core Data
Number of fuel assembliesRated thermal power, HW
Rated core flow, Hlbm/hrCore inlet subcooling, Btu/ibmHoderator temperature, F
Channel thickness, inchFuel assembly pitch, inchWide water gap thickness, inchNarrow water gap thickness, inch
'64329310024.0548.8.0806.000.5620.562
Control Rod Data
Absorber materialTotal blade span, inchTotal blade support span, inchBlade thickness, inchBlade face,-to-face internal dimension, inchAbsorber rods per bladeAbsorber rod outside diameter, inchAbsorber rod inside diameter, inchAbsorber density, % of theoretical
B4C9.751.580.2600.200760.1880.13870.0
24 ANF-87-126Revision
** : LL : L : HL : M : N* H : HL : HL
*
~ 4
HL M : MH : N* HH N* ' HL
* ~*: HL*
M ". H* H: H O': HH H: HL
**: H: HH: H: H: H* \
H: H M
H : N* : H : H : W : HH : H : HH
*
'A'~
.*
N: MH: H H NH W: MH H*
HL : H* : MH : H H MH : MH HL
HL : H : H : M~ MH : H* : M ML ML
L : HL : HL : H M: HL: HL-: L
LL RODS ( 1)L RODS ( 5)
HL RODS (16)H RODS (20)
MH RODS (13)H RODS (15)
H* RODS ( 9)W RODS ( 2)
1.45 W/0 U2351.95 W/0 U2352.55 W/0 U2353.27 W/0 U2354.23 W/0 U2354.66 W/0 U2353.27 W/0 U235 + 4.00 W/0 GD203INERT WATER ROD
Figure 4. 1 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for theANF92-344L-9G4 Xi4-2 Fuel Lattice
25*************9: * 0 ********** ANF-87-126Revision 1
*\* ~ LL o L: ML M ML NL L
* ~
* ~
* ~ N
*ML NH : M* : MH : M* ML
'
* . ML . N* M**
H: H MH M: NL
* ~
* o
* ~
MH H: H H- H: H N*: M
~ J
N* ~H W: "MH H: MH
t*'* ~
* ~
M: NH: H H MH W : NH : M* : M
ML : N~ : MH : H : H : NH : NH M: ML
ML : M : N : M* : MH : M* : M ML ML „:
ML : NL : M M: ML ML
LL RODS ( 1)L RODS ( 5)
ML RODS (16)M RODS (19)
MH RODS (13)H RODS (15)
M* RODS (10)W RODS ( 2)
1.45 W/0 U2351.95 W/0 U2352.55 W/0 U2353.27 W/0 U2354,23 W/0 U2354.66 W/0 U2353.27'W/0 U235 + 5.00 W/0 GD203INERT WATER ROD
igure 4.2 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for theANF92-344L-IOG5 XN-2 Fuel Lattice
26 ANF-87-12Revi sio
A2 C1 A2 C1 A2 C1 A2 C1 DO C1 A2 Ci EO C1 A2
C1 DO C1 DO C1 A2 C1 DO C1 A2 C1 FO C1 C1 A2
A2 C1 DO A2 DO Ci DO A2 00 C1 ~ DO C1 EO C1 A2
C1 00 A2 DO Ci 00 00 A2 00 C1 EO C1 C1 A2
A2 C1 DO C1 A2 C1 DO C1 DO A2 DO C1 EO C1 A2
C1 A2 C1 DO O'I C1 A2 DO A2 EO EO Ci C1 A2
A2 C1 00 C1 00 A2 C1 C1 00 C1 EO C1 EO C1 A2
C1 00 A2 DO Ci DO C1 EO C1 EO C1 00 C1 A2
DO C1 00 A2 00 A2 00 C1 EO C1 EO A2 82
C1 A2 DO A2 EO C1 EO C1 C1 C1 A2 A2
A2 Ci 00 DO EO C1 EO C1 A2
C1 EO 80 Ci EO C1 DO A2 A2
EO C1 EO Ci EO C1 EO C1 82 A2
C1 C1 C1 Ci C1 C1 C1 A2
A2 A2 A2 A2 A2 A2 A2 XY = Fuel Type X
Burned Y Cycles
~Fuel T e Ho. of Buouieo Descri tion
A
8C'
E
1968
32414096
GE BX8 Type III 2.19 w/o U ~ 235GE BX8 Type II 1.76 M/o U.235XN. 1 ENC92.3318.7G4XN.2 ANF92.333B.904XN.2 ANF92.3338. 10G5
Figure 4.3 Susquehanna Unit 2 Cycle 3 Reference Core I.eading
27 ANF-87-126Revision 1
120
110 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ h ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
100
90
80
N70
~ ~ ~ ~
~ ~
APPM
ROD BLOCK:
e
APRNSCRAM
LIN)~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~
~~ ~ ~ ~ ~ ~ ~
rr~/
r ~
~ r~ /e' 'r/ ~/
e /~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ) ~
66/45)
err
~ ( ~ ~
/e100/v XeR00.' IN
~4
~ ~ ~ ~ ~ ~ ~ ~ ~ $ ~ ~ ~
ROO BiOCK~ MONITOR
80
WE 50 ~ ~
ee ~
I ~
I
~ el'
45K
80K
e
CORE PLOW
R00 LINE
40
30
20
e ~ ~ ~ ~ e
~ e ~ ~ ~ ~
~ ~
~ ~ ~ 4'
P ~
~ ~ ~ ~ ~ ~
e
~ ~ ~ ~ ~ 'I ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~
~ ~ ~ ~
~ ~ 4 ~
~ ~ ~
~ ~ ~ 4 ~
~ ~ ~ ~ ~ 4 ~
~ ~ e
10
NTCIRC
~ ~ ~M
'-PUMP
)N FLOW;
00 10 20 30 40 50 60 70 80 90
CORE FLOW, % RATED100
Figure 4.4 Susquehanna Unit. 2 Cycle 3 - Core Power vs. Core Flow
28 ANF-87-12''Revision
59
55
51
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58
12 -- 00 -- 12
20 -- 26 -- 26 -- 20
59
55
51
47
43 -- -- 20 -- 20 20
00 -- 12 -- 08 -- 12 -- 00
20
47
43
39 -- 12 -- 08 -- 08 -,- 00 -- 08 08 -- 12 39
35 -- -- 26 44 44 26 35
31 -- 00 -- 04 -- 00 -- 00 -- 00
27 -- -- 26 44 44
23 -- 12 -- 08 -- 08 -- 00* -- '819 -- -- 20 -- 20 20
04 -- 00
26
08 -- 12
20
31
27
19
15 00 -- 12 -- 08 -- 12 -- 00
20 -- 26 -- 26 -- 20
12 -- 00 -- 12
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58
Cycle ExposureControl Rod Density
0.0 HHD/HTU23.3 %
Control Rod Being Withdrawn = 00*Rod Fully Inserted =, 00Rod Fully Withdrawn =--
Figure 5. 1 Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal ErrorAnalysis Limiting Initial Control Rod Pattern
1.60
1.60
1.40f4
1.30O
Note: The MCPR operatinglimit shall be the maximum ofthis curve, the full flowMCPR operating limit or the poorerdependent MCPR operating limit.
A1.80
O
A
1.10
40 50 60 70 80 90 100
TOTAL CORE RECIRCULATION FLOW (% RATED)
figure 5.2 Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit
Ah
+C"
A-1 ANF-87-126Revision 1
APPENDIX A
U
SINGLE LOOP OPERATION
This Appendix provides limits and justification of those limits for SingleLoop Operation (SLO).
A.l ANTICIPATED OPERATIONAL OCCURRENCES Reference A. 1
The NSSS supplier has provided analyses which demonstrate the safety of plantoperation with a single recirculation loop out of service for an extended
V
period of time. These analyses restrict the overall operation of the plant tolower bundle power levels and lower nodal power levels than are allowed. when
oth recirculation systems are in oper ation. The physical interdependencebetween core power and recirculation flow rate inherently limits the core toless than rated power. ANF fuel was designed to be compatible with the co-
resident fuel in thermal hydraulic, nuclear, and mechanical designperformance. The ANF methodology has given results which are consistent withthose of previous analyses for normal two-loop operation. Many analysesperformed by the NSSS supplier for single loop operation are also applicableto single loop operation with fuel and analyses provided by ANF.
For single loop operation, the NSSS vendor found that an increase of 0.01 inthe HCPR safety limit was needed to account for the increased flow measurement
uncertainties and increased tip uncertainties associated with single pump
operation. ANF has evaluated the effects of the increased flow measurement
uncertainties on the safety limit HCPR and found that the NSSS vendordetermined increase in the allowed safety limit MCPR is also applicable to ANF
fuel during single loop operation. Thus, increasing the safety limit HCPR by
0.01 for single loop operation (1.07) with ANF fuel is sufficientlyonservative to also bound the increased flow measurement uncertainties for
single loop operation.
A-2 ANF-87-Revisio
The limiting MCPR operating limit for single loop operation is conservativelyset using the limiting pump seizure accident delta CPR plus the single loop
operation HCPR safety limit. This limit together with the. HCPRf curve for two
loop operation plus .Ol and the MCPRp curve for two loop operation plus .Ol
conservatively bound all transients.
The Technical Specifications require APRH/LPRH surveillance to the left of the
45% Constant Flow line and above the 80% Rod Block line. Based on core
hydrodynamic stability analyses for Cycle 3, operation at power/flowcombinations above and to the left of the line connecting the 66% Power/45%
Flow and 69% Power/47% Flow points needs to be added to the APRM/LPRM
surveillance requirements. Figure 4.4 shows the core power versus core flowestablished for Cycle 3.
A-3 ANF-87-126Revision 1
A.2 POSTULATED ACCIDENTS Reference A.2
ANF performed LOCA analyses for single loop conditions and has determined thatthe MAPLHGR limit curve (Section 7.2) for two-loop operation is also
applicable to single loop operation for ANF 9x9 fuels.
A-4 ANF-87-Revisio
REFERENCES
A. 1 "Susquehanna Unit 2 Cycle 2 Single Loop Operation Analysis," XN-NF-86-146, Advanced Nuclear Fuels Corporation; Richland, WA 99352, November1986.
A.2 "Susquehanna LOCA Analysis for Single Loop Operation," XN-NF-86-125,Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.
B-1 ANF-87-126Revision I
APPENDIX B
SEISMIC- LOCA EVALUATION
The structural response of Advanced Nuclear Fuels Corporation's (ANF's) 9x9
fuel is similar to the structural response of the GE BxBR fuel it replaces inthe 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 Bx8 fuel types which
are important to the dynamic response of the fuel are summarized in Table B. l.he close agreement between the important parameters for the ANF 9x9 and GE
x8R fuel types indicates that the structural response would be very similarfor both fuel types.
Similarity in the natural frequencies of the two fuel types mentioned above isfurther assured by the stiffness of the fuel assembly channel box. Both fueltypes use the same fuel assembly channel box, and the channel box dominates
the overall dynamic response of the incore fuel. ANF calculations show thatapproximately 97% of the stiffness of a fuel assembly is attributable to thestiffness of the channel box. For this reason, the dynamic structuralresponse 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 tooccur.
B-2 ANF-87-Revisio
TABLE B. 1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICSFOR 8X8 AND 9X9 FUEL ASSEMBLIES
~Pro ert
Assembly Weight, lbs
Number of Spacers
Overall Assembly Length, in
Assembly Frequencies, cps
Mode 1
23
567
ANF 9x9
580
171.29
1.93.76.5
10.415.521.929.1
Fuel T esGE 8x8R
600
171.40
*GE proprietary
ANF-87-126Revision 1
Issue Date: 11/25/87
SUS(UEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS
Design and Safety Analyses
Distribution:
D.D.R.L.S.R.K.H.S.T.J.L.D.G.C.J.H.
A. AdkissonJ. BraunE. CollinghamJ. FedericoF. GainesG. GrummerD. HartleyJ. HibbardE. JensenH. KeheleyN. MorganA. NielsenF. RicheyL. RitterJ. VolmerAD WhiteE. Williamson
H. G. Shaw/PP8L (20)Document Control (5)
