ML20071Q087
| ML20071Q087 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/06/1982 |
| From: | Chandler J, Stout R, Williamson H SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17194B411 | List: |
| References | |
| XN-NF-82-77(NP), XN-NF-82-77(NP)-R01, XN-NF-82-77(NP)-R1, NUDOCS 8212290215 | |
| Download: ML20071Q087 (33) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
XN-NF-82-77(NP)
Revision 1 Issue Date: 12/06/82 j
DRESDEN UNIT 2 CYCLE 9 RELOAD ANALYSIS Mechanical, Thermal Hydraulic, and Nuclear Design Analyses for ENC XN-1 Reload Fuel l
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Prepared by:
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J. C. Chandler Reload Fuel Licensing l
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f.,,. ' / '/e Approve:
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- R. B. Atout, Ma#ager
/
Licensing & Safety Engineering Approve:
l H. E. Williamson, Manager Neutronics & Fuel Management O
.f./
- 8 Approve:
G. A. Sofer, Manager Fuel Engineering & Technical Services 9f ERON NUCLEAR COMPANY,Inc.
8212290215 021221 PDR ADOCK 05000237 p
XN NF 82 77 (NP)
~
l REVISION 1 DRESDEN UNIT 2 CYCLE 9 RELOAD ANALYSIS I
ROVEMBER 1982 RICHLAND, WA 99352 l
ERON NUCLEAR COMPANY,Inc.
NUCLEAR REQULATORY COMMISSION DISCLAIMER mePORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was dertwed through research and development progrens sponsored by Emmon Nuclear Company, Inc. It is being eat >
mitted by Emmon Nuclear to the USNRC as part of a technical contri-buoon to fac4tato esfety analyses by licensees of the USNRC which utdire Emmon Nuclear fabricated reload fuel or other techncat services provided by Emmon Nuclear for richt water power reactors and it is true and correct so the best of Emmon Nucteer's knowledge, information, are belief. The informaoon 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 Emmon Nucteer in their demonstreoon of comotience with the USN RC's regulations.
Without derogating from the foregoirg neither Emmon Nuclear nor any person acting an its behalf; A. Makes any warranty, empress or impfied, with respect to the acciaracy, completences, or usrfulness of the infor-mation contained in this documet, or that the use of any informatort apparatus, method, or process distfosM in this document will not infnnge privately owned rights; or B.
Assumes any liabilities with respect to the use of, or for darrages rosatting from the use of, any information, oc>
perstus, rnethod, or process disclosed in this document.
XN NF-F00,766
i XN-NF-82-77 (NP)
Revision 1 TABLE OF CONTENTS Section M
1.0 INTRODUCTION
1 2.0 FUEL MECHANICAL DESIGN ANALYSIS....................
2 3.0 THERMAL HYDRAULIC DESIGN ANALYSI S..................
2 4.0 NUCLEAR DESIGN ANALYSIS............................
3 5.0 ANTICIPATED OPERATIONAL OCCURRENCES................
4 6.0 POSTLLATED ACCIDENTS...............................
5 l
7.0 TECHNICAL SPECIFICATIONS...........................
6 9.0 ADDITIONAL REFERENCES..............................
8 APPENDIX A - Lead Test Assemblies..................
A-1 APPENDIX B - Surveillance Requirements.............
B-1 s
l l
I I
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- XN-NF-82-77(NP)
Revision 1 l
l LIST OF TABLES Table No.
Page 4.1 Dresden 2 Reload Batch XN-1 Neutronic Design Values.......................................
17
-5.1 Determination of Thermal Margins....................
19 A.1 Fuel Assembly Maximum Lattice K. Values.............
A-5 A.2 LTA MAPLHGR Limits..................................
A-6 J
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iii XN-NF-82-7XNP)
Revision 1 l
LIST OF FIGURES Figure No.
Py 3.1 Dresden 2 Cycle 9 Safety Limit Radial Power Histogram....................................
9 3.2 Dresden 2 Cycle 9 Safety Limit Local Peaking.......
10 4.1 Enrichment Distribution for Fuel Type XN-1 8x8 (Enriched Lattice 3.02 w/o U-235)..................
11 4.2 Dresden Unit 2 Cycle 9 Reference Loading Pattern (One Quarter of Symmetrical Core Loading)..........
12 4.3 Decay R ati o vs Reactor Power.......................
13 5.1 Starting Control Rod Pattern for Control Rod Withdrawal Analysis................................
14 5.3a MCPR for Automatic Flow Control (AFC)..............
15 5.3b MCPR for All Conditions............................
16 1
l I
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1 XN-NF-82-77(NP )
Revision 1
1.0 INTRODUCTION
This report presents the results of analyses performed by Exxon Nuclear Company (ENC) in support of the Cycle 9 (XN-1) reload for Dresden Unit 2, which is scheduled to commence operation in the Spring of 1983. Dresden 2 is the second BWR/3 to be licensed on the basis of ENC analyses and is a sister plant to Dresden 3, which began operation with ENC fuel in April 1982. All limits and analyses reported herein are consistent with their counterparts in the Dresden 3 documentation.
The Cycle 9 core will comprise 224 unirradiated reload fuel assemblies fabricated by ENC, 384 once-and twice-irradiated Type 8x8R assemblies f abricated by General Electric Company (G.E.), and 116 G.E. Type 8x8 fuel assemblies irradiated from three to five cycles each. Except as noted below, the ENC-fabricated assemblies are as described in XN-NF-81-21 (Reference 9.1).
The core configuration is described in Section 4.0 of this report.
Cycle 9 operation will involve the use of four Lead Test Assemblies (LTAs) placed in symmetrical, non-limiting locations in the core. The LTAs, are described in Appendix A.
Operating limits for the LTAs are also given in Appendix A.
- Brackets identify ENC proprietary information.
2 XN-NF-82-77(NP)
Revision 1 This report is intended to be used in conjunction with XN-NF-80-19 Volume 4, " Application of the ENC Methodology to BWR Reloads," which describes the analyses which were performed in generation of the results reported in this document.
2.0 FUEL MECHANICAL DESIGN ANALYSIS
~
Applicable Fuel Design Report Reference 9.1 The power history depicted in Figure 5.10 of Reference 9.1 bounds the expected power 4
history for the Dresden 2 Type XN-1 fuel Fuel Centerline Temperature Exposure at Minimum Margin Point 21,200 MWD /MT
)
Centerline Temperature at 120% Overpower 46070F
]
Melting Point of Fuel 49000F 1
~~
Margin to Centerline Melting 2930F 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization Reference 9.7 3.2.5 Calculated Bypass Flow Fraction 10.8%
3.3 MCPR Fuel Cladding Integrity Safety Limit Reference 9.3 3.3.1 Coolant Thermodynamic Condition Core Rated Thermal Power 2527 MWt
}
Core Inlet Flow Rate 98 x 106 lbm/hr Steam Dome Pressure 1020 psia Feedwater Temperature 3200F S
3 XN-NF-82-77(NP)
Revision 1 3.3.2 Design Basis Radial Power Distribution Figure 3.1 3.3.3 Design Basis Local Power Distribution Figure 3.2 4.0 NUCLEAR DESIGN ANALYSIS w
4.1 Fuel Bundle Nuclear Design Analysis for Fuel Type XN-1 8x8 Assembly Average Enrichment 2.83%
Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform 3.02% with 6" Natural Uranium Ends Burnable Poisons Figure 4.1 II Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 N
Maximum Lattice K.
1.224 4.2 Core Nuclear Design Analysis f_-
4.2.1 Core Configuration Figure 4.2 Core Exposure at E0C8(1), MWD /MT 21,649/21,158 4F Core Exposure at B0C9, MWD /MT 13,096 Core Exposure at E0C9, MWD /MT 20,826 4.2.2 Core Reactivity Characteristics BOC9 Cold K-effective, All Rods Out 1.111 BOC9 Cold K-effective, All Rods In
.958 B0C9 Cold K-effective, Strongest Rod Out
.989 Technical Specification R-Value
.04%(2)
SBLC Reactivity, 700F, 600 ppm
.950 (1) Nominal Value/Value Used in Shutdown Reactivity (Calculations.
(2) Accounts for B C Settling in Control Rod Tubes Maximum K-effective 4
with strongest rod withdrawn occurs at BOC9).
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4 XN-NF-82-77 (NP)
)
Revision 1 4.2.4 Stability Analysis Reactor Core Stability Figure 4.3 Maximum Decay Ratio Value 0.46 e
Channel Hydrodynamic Stability XN-1 8x8 Fuel Decay Ratio Value 0.30 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Report Reference 9.2
]
5.1 Analysis of Plant Transients at Rated Conditions Reference 9.3 i
limiting Transients:
i Generator Load Rejection Without Bypass (LRWB)
Loss of Feedwater Heating (LFWH)
Feedwater Controller Failure - Maximum Demand (FWCF) 5.2 Analyses for Reduced Flow Operation Reference 9.4 Limiting Transient: Recirculation Flow Increase i
5.3 ASME Overpressurization Analysis Reference 9.3 Event MS!v Closure Single Failure MS!V Position Sc-am Tr Maximum Pressure
/31'7.3 13M6-psig
_f Maximum Sensed Pressure
/3 DY, 7
-T337iFf psig 5.4 Control Rod Withdrawal Error (CRWE) i Starting Control Rod Pattern for Analysis Figure 5.1 m
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5 XN-NF-82-77 Revision 1 (NP)
Rod Distance ACPR Block Setting Withdrawn ENC 8x8 GE 8x8,8x8R 106 4.5 ft.
0.10 0.08 107 5.0 0.11 0.09 108 5.5 0.12 0.10 6.0 0.13 0.11 109(1) 110 6.0 0.13 0.11 5.5 Fuel Loading Error ENC 8x8 GE 8x8,8x8R 7
f ACPR 0.14 0.14 5.6 Determination of Thermal Margins Table 5.1 MCPR Operating Limits at Rated Conditions
[
Fuel Type MCPR Operating Limit ENC XN-1 8x8 1.31 GE 8x8, 8x8R 1.31 MCPR Operating Limits at Off-Rated Conditions Automatic Flow Control Figure 5.3a u
All Conditions Figure 5.3b r
6.0 POSTULATED ACCIDENTS 6.1 LOSS OF COOLANT ACCIDENT 6.1.1 Break location Spectrum Reference 9.5
=
6.1.2 Break Size Spectrum Reference 9.5 6.1.3 MAPLHGR Analyses Reference 9.6 P"
Limiting Break: Double-Ended Guillotine Break Recirculation Pump Suction Line 1.0 Break Coefficient 3s E
==
(1) Rod Block setting of 110% selected for Cycle 9 operation.
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6 XN-NF-82-77(NP)
Revision 1 Bundle Average Peak Clad Peak Local
=
Burnup (MWD /MT)
MAPLHGR Temperature (OF)
MWR (%)
0 13.0 kw/ft 1900 0.8 12,000 13.0 1856 0.7 6.2 CONTROL R0D OROP ACCIDENT See XN-NF-80-19, Vol.
Dropped Control Rod Warth 7.8 mk Doppler Coefficient (773CF)
-9.8x10-6f.,(opy-1 Effective Delayed Neutron Fraction 0.0055 Four Bundle Local Peaking Factor 1.19 Maximum Deposited Fuel Rod Eithalpy 111 cal /gm 7.0 TECHNICAL SPECIFICATIONS 7.1 LIMITING SAFETY SYSTEM SETTINGS 7.1.1 MCPR Fuel Cladding Integrity Safety Limit
=
All Fuel Types 1.05 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 1345 psig 7.2 LIMITING CONDITIONS FOR OPERATION 7.2.1 Average Planar Linear Heat Generation Rate (Fuel Type XN-1 8x8)
Bundle Average
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0 MWD /MT 13.0 kw/ft 12,000 13.0 A
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XN-NF-82-77(NP)
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7.2.2 Minimum Critical Power Ratio
- l. - Q 1 Fuel Type MCPR u
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Reduced Flow MCPR Limits Automatic Flow Control Figure 5.3a s
All Conditions Figure 5.3b
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Revision 1 9.0 ADDITIONAL REFERENCES 9.1 S. F. Gaines, " Generic Mechanical Design for Exxon Nuclear det Pump BWR Reload Fuel," XN-NF-81-21(A), Revision 1 (January 1982).
9.2 R. H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71, Revision 2 (November 1981).
9.3 R. H. Kelley, " Plant Transient Analysis for Dresden Unit 2 Cycle 9,"
XN-NF-82-84, Revision 1 (November 1982).
9.4 R. H. Kelley, "Dresden Unit 3 Analyses for Reduced Flow Operation," XN-NF-81-84 (December 1981).
9.5 J. E. Krajicek, " Generic Jet Pump BWR3 LOCA Analysis Using the ENC EXEM Evaluation Model," XN-NF-81-71(A) (October 1981).
9.6 D. J. Braun and P. J. Valentine, "Dresden Unit 2 LOCA Analysis Using the ENC EXEM/BWR Evaluation Model; MAPLHGR Results," XN-NF-82-88, Revision 1 (November 1982).
9.7 J. C. Chandler, "Dresden Unit 3 Cycle 8 Reload Analysis," XN-NF-81-76, Revision 1 (December 1981).
9.8
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Fuel Number of Type Assemblies Description A
220 XN-1 8x8 2.83 w/o U-235 E
4 LTA Figure 4.2 Dresden Unit 2 Cycle 9 Reference Loading Pattern (One Quarter of Symnetrical Core Loading) k
--mm
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13 XN-NF-82-77(NP)
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Revision 1 Table 4.1 Dresden 2 Reload Batch XN-1 Neutronic Design Values Fuel Pellet Reference 9.1 Fuel Rod Reference 9.1 Fuel Assembly Reference 9.1 Fuel Assembly Loading, Kgu02 197.2 Fuel Assembly Loading, KgU 173.8 Core Data Number of fuel assemblies 724 Rated thermal power, MW 2527 Rated core flow, 106 lbm/hr 98.0 Core inlet subcooling, BTU /lbm 24.6 Moderator temperature, OF 546 Channel thickness, inch 0.080 Channel inside face-to-face dimension, inch 5.278 Fuel assembly pitch, inch 6.0 Wide water gap thickness, inch 0.750 Narrow water gap thickness, inch 0.374 Control Rod Data Absorber material 8C 4
Total blade span, inch 9.750 Totalcen.tralsupport span, inch 1.562 Blade thickness, inch 0.3120 m..
18 XN-NF-82-77(NP)
Revision 1 Table 4.1 Dresden 2 Reload Batch XN-1 Neutronic Design Values (Cont.)
Blade face-to-f ace internal dimension, inch 0.200 Absorber rods per blade 84 Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70
Table 5.1 Determination of Thermal Margins Indicated Maximum Maximum Maximum MCPR Event Model Exposure Power Flow Heat Flux lower Pressure Limit (2)
LRWB COTRANSA E0C9 100%
100%
114.5%
350.3%
1281.1 psig 1.31/1.31 FWCF COTRANSA EOC9 100%
100%
116.9%
198.4%
1207.2 1.26/1.26 LFWH PTSBWR3 E0C9 100%
100%
<120.0%
1.u.1%
1039.9 1.21/1.21 CRWE(1)
XTGBWR B0C9 100%
100%
1.18/1.16 G
e (1) Rod Block setting of 110% selected for Cycle 9 operation.
(2) Indicated limits for ENC 8x8 fuel /G.E. 8x8 fuel.
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LEAD TEST ASSEMBLIES
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Revision 1 A.1 PROGRAM DESCRIPTION Cycle 9 operation of Dresden 2 will include a lead test assembly program of fuel assemblies designed and fabricated by Exxon Nuclear Company. The locations of the four lead test assemblies (LTAs) were selected on the basis of allowing adequate exposure accumulation and not to result in additional operating limitations. The locations of the LTAs are shown in Figure 4.2 of this report.
A.2 FUEL STORAGE CRITICALITY A comparison of the maximum lattice K. values for the ENC 8x8, ENC LTA, and GE 8x8 fuel designs with and without gadolinia is given in Table A.2.
The calculations were performed with the ENC XFYRE code. As shown, the ENC 8x8 fuel with no gadolinia has the highest reactivity.
The high density spent fuel storage racks were analyzed by NSC and were found to meet the fuel storage Keff requirements of 0.95 for the limiting ENC 8x8 3.02 w/o U-235 fuel design with no gadolinia(1}.
Based on the bundle reactivity cunparisons as shown in Table A.1, the high density storage racks are acceptable for storage of the ENC LTA fuel with considerable reactivity margin when the gadolinia is considered.
Previous an slyses have demonstrated that the dry storage racks and the spent fuel storage racks are acceptable for storage of the GE 8x8 (2.82 w/o U-235) Cycle 8 reload fuel.
Since the ENC fuel with gadolinia has a lower maximum K,than the GE fuel with gadolinia, the ENC 8x8 and LTA fuel may be (1) Wong, Kin W. (testimony), Atomic Safety and Licensing Board, January 21, 1982.
A-3 XN-NF-82-77(NP)
Revision 1 stored in the dry storage and spent fuel storage racks without compromising the technical specification fuel storage requirements.
A.3 STABILITY ANALYSIS a
The inclusion of four LTAs does not impact the core stability analysis a
reported in Section 4.2 of the body of this report. The LTA fuel was analyzed for channel hydrodynamic stability with a resulting maximum decay ratio of 0.29.
A.4 OPERATING LIMITS Operating limits for the LTAs are established based on consistence with a
the limits calculated for ENC production fuel.
Extrapolation of the production fuel bundle power limits as determined using ENC's plant transient methodology (Ref. 8.8) results in an operating limit MCPR of 1.35 for the LTA fuel.
f Operating limit APLHGR values were determined by extrapolating the 8x8 APLHGR limits from Section 6.1.3 of the body of this report 1
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Revision 1 Table A.1 Fuel Assembly Maximum Lattice K. Values 4
Maximum K. -
Maximum K. -
m No Gadolinia/
With Gadolinia/
Fuel Design (Exposure, MWD /MTU)
(Exposure, MWD /MTU)
ENC 8x8 3.02 w/o U-235 Five Gd Rods 3.5 w/o Gd 023 1.339 (0) 1.224 (8000)
ENC LTA 1.334 (0) 1.220 (8000)
GE 8x8 2.82 w/o U-235 (80RL282L) 1.321(0) 1.231 (6000) i J
b
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Revision 1 Table A.2 LTA MAPLHGR Limits Bundle Average
- MAPLHGR, Burnup, MWD /MT kW/ft 0
10.2 12,000 10.2 4
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APPENDIX B J
SURVEILLANCE REQUIREMENTS
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Revision 1 APPENDIX B - SURVEILLANCE REQUIREMENTS The thermal margin (MCPR) requirements associated with the generator load rejection transient without bypass to the condenser (LRWB) are based on a statistical combination of uncertainties in calculated parameters and measured plant performance in the area of control rod drive performance. The Plant Technical Specifications require that control rod drive performance be monitored on an individual rod basis at regular intervals.
This Appendix provides for modification of MCPR operating limits if the measured control rod drive performance f alls outside the statistical basis used in the thermal margin calculation.
For a mean control rod insertion time to 90% insertion of 2.74 seconds or less, the MCPR operating limits established by the statistical evaluation of the LRWB transient are valid. For a mean 90% insertion time corresponding to the Technical Specification limit of 3.50 seconds, an additional thermal margin conservatism of 0.07 is required. Between those two values, the MCPR operating limit should be determined by the following formula:
MCPRS = MCPRa + 0.092T - 0.252 where:
Operating Limit MCPR adjusted for observed scram time statistical behavior; MCPRa=
Operating Limit MCPR obtained from cycle analysis; and T
Statistical mean of observed scram insertion times to
=
the 90% insertion point.
XN-NF-82-77(NP)
Revision 1 Issue Date: 12/06/82 DRESDEN JXIT 2 CYCLE 9 RELOAD ANALYSIS l
Distribution JC chandler GF Owsley FA Shallo CECO /LC O'Malley (60)
Document Control (5) 1
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