ML20045E054
| ML20045E054 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/25/1993 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19303F662 | List: |
| References | |
| OPPD-NA-8301-NP-R05, OPPD-NA-8301-NP-R5, NUDOCS 9307010037 | |
| Download: ML20045E054 (46) | |
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Table 1 RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301 -NP REV. 05 Title Page Changed the revision number and date.
All Pages Updated revision number.
ii Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change.
vi Updated Revision Sheet.
1 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change. Changed Combustion Engineering to ABB Combustion Engineering. Also added references 2-4 and 2-5 to allow ABB-CE fuel to achieve 60 MWD /kg burnup. l 2 Moved previous reference 3-5 to 3-2 and deleted the reference to the INCA topical. The CECOR topicalincorporates the INCA methods in the later topical. Adjusted the remaining references to reflect the change.
4 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change.
5 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change.
6 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change.
7 Updated the reference to use of the QUlX code to reflect the use of the HERMITE code. Corrected production of scram reactivity codes to axial shape analysis performance. Added use of HERMITEto derive DNBR power-to-fuel design limit. Revised peaking factors to correctly indicate tilt when applicable.
9-10 Reflect use of COLR rather than plan to use per Technical Specification Amendment No.141. Deleted concurrent NRC distribution.
11 Added References 2-4 and 2-5 to allow fuel rod average burnups up to 60 MWD /kg. Updated methodology reference to the latest revision (3-1).
Deleted reference 3-2 since it was replaced with reference 3-5, then moved reference 3-5 to 3-2 location and moved 3-6 to 3-5 to accurately reflect discussion in sections 3.2 and 3.3.
12 Updated methodology reference to the latest revision (5-1,7-1,7-2 and 7-3).
13 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to reflect a corporate name change. Added proprietary brackets to information in the table per ABB-CE.
17 Changed Advanced Nuclear Fuel Corp. (ANF) to Siemens Power Corp. (SPC) to ry%ct a corporate name change.
9307010037 930625 f4 DR ADDCK0500g5
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I I PROPRIETARY DATA CLAUSE 1
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l This document is the property of Omaha Public Power District (OPPD). Proprietary information, l indicated by brackets, developed by Siemens Power Corp (SPC), ABB-Combustion Engineenng i (ABB-CE) and Westinghouse Electric Corporation (W) has been removed. The SPC, ABB-CE, and W information was purchased by OPPD under proprietary information agreements.
OPPD-NA-8301-NP Rev. 05 ii
OMAHA PUBLIC POWER DISTRICT RELOAD CORE METHODOLOGY OVERVIEW
1.0 INTRODUCTION
Analyses done to license reload cores Tor Fort Calhoun Station consist of the analysis performed by the Omaha Public Power District and the analysis performed by the nuclear fuel vendor. The current nuclear fuel vendor is Westinghouse Electric Corporation (W); .
however, future reload fuel may be potentially supplied by any of the four U.S. PWR nuclear fuel vendors: Siemens Power Corp. (SPC), ABB Combustion Engineering (ABB-CE), g Westinghouse, or Babcock and Wilcox. The following sections discuss the reload analyses and consolidate information about the District's methodology previously submitted.
2.0 FUEL SYSTEM DESIGN The fuel assembly mechanical design and analysis are performed by the nuclear fuel vendor.
The fuel mechanical design and design methods utilized for Fort Calhoun Station by W are described in Reference 2-1. ABB-CE, the co-resident fuelin the mixed core, fuel mechanical design and design methods are discussed in References 2-2 thr.ough 2-5.
i in an effort to further reduce the neutron flux to the reactor vessel welds, full length Hafnium flux suppression rods, wh;ch are similar to the part length poison rods utilized in Cycle 10, k
will be incorporated into the fuelloading pattern. The poison rods are composed of hafnium metal extending the fulllength of the active fuel. Inert material comprises the balance of the >
rod. They will reside in the outer guide tubes of quarter core assembly numbers 1,2 and 8.
The fuel system design will also incorporate four natural uranium fuel assemblies in quarter core location number 14 for additional neutron flux reduction to the cntical welds.
3.0 NUCLEAR DESIGN The District's nuclear design methodology is discussed in Reference 3-1.
3.1 Fuel Management The reload core fuel management is performed by the District. Current fuel management schemes are selected to reduce flux to the reactor pressure vessel welds.
OPPD-NA-8301-NP Rev. 05 Page 1 of 25
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8.0 REFERENCES
(Continued) l Sp.ction 4 References (Continued) l 4-6 Letter from R. A. Clark (NRC) to W. C. Jones (OPPD), March 15,1983.
4-7 Letter from W. C. Jones (OPPD) to R. W. Reid (NRC), December 4,1979.
4-8 " Statistical Combination of Uncertainties," CEN-257(0)-P-A, Part 2, November 1983, including Supplement 1 -P, August 1985.
Section 5 References 5-1 " Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," OPPD-NA-8303-P, Rev. 04, January 1993. E Section 6 References 6-1 "CE Setpoint Methodology," CENPD-199-P, Revision 1-P-A, January 1986.
l 6-2 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification,"
( OPPD-NA-8302-R Rev. 03, January 1993. E 6-3 " Statistical Combination of Uncertainties," CEN-257(0)-P- A, Parts 1,2, and 3, November 1983, including Supplement 1-R August 1985. E Section 7 References 7-1 " Removal of Cycle-Specific Parameter Limits from Technical Specifications", NRC Generic Letter 88-16. October 4,1988.
7-2 " Reload Core Analysis Methodology Overview," OPPD-NA-8301-P, Rev. 05, January 1993.
7-3 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification,"
OPPD-NA-8302-R Rev. 03, January 1993.
7-4 " Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," OPPD-NA-8303-P, Rev. 04, January 1993.
OPPD-NA-8301 -NR Rev. 05 1 I
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TABLE 4-1 l
PARAMETER RANGES OFTIIE SOURCE DATA FOR T11E CE-1 CIIF CORRELATION ANDTHE RANGE OF WESTINGliOUSE, ABB-CE AND SPC 14 x 14 FOR FORT CALHOUN VALUES I
CORRELATION ABB-CE SPC WESTINGHOUSE g PARAMETER RANGE RANGE RANGE RANGE Pressure (psia) [ ] N/A N/A N/A Local Coolant Quality [ ] N/A N/A N/A l Local Mass Velocity [ ] N/A N/A N/A (Ib 'hr-ft )
m 2
Subchannel Wetted Equiv. .3588 to .4043 to .4010 to .4043 to Diameter (in) .5447 .5449 .5402 .5449 Subchannel lieated Equiv. .4713 to .5334 to .5270 to .5334 to Diameter (in) .7837 .7840 .7760 .7840 Heated Length (in) 84 to 150 128 128 128 Grid Spacing (in) 14.2 to 18.25 16.8 16.8 16.8 i
OPPD-NA-8301 -NR Rev. 05 Page 13 of 25 l
STAGE 1 TORC CHANNEL Omaha Public Power District Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station - Unit No.1 4-4 OPPD-NA-8301-NR Rev. 05 Page 18 of 25
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L STAGE 2 CHANNEL Omaha Public Power District Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station - Unit No.1 4-5 OPPD-NA-8301-NR Rev. 05 Page 19 of 25
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i l STAGE 3 CHANNEL - Omaha Public Power District - Figure l GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station - Unit No.1 4-6 l l
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, i INLET FLOW DISTRIBUTION Omaha Public Power District Figure FOR FCS 4-PUMP OPERATION Fort Calhoun Station - Unit No.1 4 OPPD-NA-8301 -NP, Rev. 05 j Page 22 of 25 l
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EXIT PRESSURE DISTRIBUTION Omaha Public Power District Figure FOR FCS 4-PUMP OPERATION Fort Calhoun Station - Unit No.1 4-9 OPPD-NA-8301-NR Rev. 05 i Page 23 of 25 l
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i CETOP-D CHANNEL GEOMETRY Omaha Public Power District - Figure (CHANNEL 1 NOT SHOWN) - Fort Calhoun Station - Unit No.1 4-10 OPPD-NA-8301-NR Rev. 05 Page 24 of 25
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l ALGORITHM FOR DNB Omaha Public Power District Figure LCO MONITORING Fort Calhoun Station - Unit No.1 6-1 OPPD-NA-8301 -NR Rev. 05 Page 25 of 25
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i Table 3 TRANSIENT AND ACCIDENT METHODS AND VERIFICATION
]
OPPD-NA-8303-NP l Rev. 04 Eage Section Cha022 Title Page Changed revision number and date.
All Pages Updated the revision number, iii-iv Table of Contents Updated Table of Contents.
v List of Tables Updated page numbers, added Table 5.14.4-1.
vi-vii Table of Figures Update page numbers.
viii Revisions Added Revision 4.
19 5.1.5 Reference to CETOP changed to TORC since TORC is now used for calculating ROPMs.
21 5.2.4 Revised RCS volumes used in Boron Dilution calculations to reflect a more accurate RCS volume.
Added clarification of initial conditions in Mode 3.
Changed "most reactive rod" to " rod with greatest worth". Added "(Mode 5)" designation to refueling.
22 5.2.5 Editorial correction.
Included pressure condition of makeup water.
22 5.2.6 Updated results section to reflect the most current use of the methodology of Rev. 03 of OPPD-NA-8303-P-A.
25 5.3.6 Revised results section to remove cycle specific results.
l 26 5.3.4 - 1 Added proprietary brackets to information in the table per ABB-CE.
27 5.4.1 Replaced specific trip setpoint with " Low Flow Trip setpoint" 28 5.4.7 Removed reference to CETOP conservatisms since CETOP has been replaced by TORC in this analysis.
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36-37 5.6.5 Removed first paragraph.
Updated MTC calculational methods.
Added paragraph detailing assumptions on RTD response times.
38 5.6 Changed initial Core Power to 1500 MWt (core rated power) for consistency with the application of statistical combination of uncertainties. Adde proprietary brackets to information in the table per ABB-CE.
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) l Table 5.3.4-1 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Parameter 1).DitS Value l Initial Core Power MWt 1500*
Initial Core inlet 'F . Maximum allowed
- Temperature by Tech. Specs.
Initial RCS Pressure psia Minimum allowed
- by Tech. Specs.
Initial Core Mass Flow Rate x106lbm/hr Minimum allowed
- by Tech. Specs. ,
Moderator Temperature x10 'Ap/*F Most negative Coefficient allowed by Tech.
Specs.
CEAInsertion % insertion Maximum allowed by.
Tech. Specs.
Radial Peaking Distortion Maximum value Factor predicted during core life Dropped CEA Worth %Ap [ ] value predicted g during core life for the CEA producing the
[ ] distortion g factor Core Average Hgap BTU /ht-ft2 *F Maximum value ,
predicted I during core life Fuel Temperature x10 'Ap/*F Most negative value Coefficient predicted during core life
- For DNBR calculations, the effects of uncertainties on these parameters are combined statistically.
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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)
I Table 5.6.4-1 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS Parameter Units .. . Value Initial Core Power MWt 1500* 5 l Initial Core Inlet *F Maximum alicwed* ,
Temperature by Tech. Specs.
initial Reactor Coolant psia Minimum allowed * -
System Pressure by Tech. Specs. '
I initial Core Mass x10*lbm/hr Minimum allowed
- Flow Rate by Tech, Specs.
i l Axial Shape index - asiu Most Negative allowed by DNB LCO Tent. ,
RTD C ayTime sec [ ] Hot Leg
[ ] Cold Leg Moderator Temperature x10 dAp/'F ' Negative values up Coefficient to the most negative value allowed by Tech. Specs.
Radial Peaks Maximum Allowed byTech Spec. for a Given Initial Power Level Scram Reactivity % Minimum Predicted During a Cycle High Power Trip Analysis Setpoint % of 1500 MWt 112.0 Variable High Power Trip Setpoint % Above Initial 10.0 Power Level Temperature Shadowing Factor % Power /'F [ ]'
- For DNBR calculations, effects of uncertainties on these parameters were ,
combined statistically.
OPPD-NA-8303-NP, Rev. 04 38 of 133 ;
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