ML20138D309
| ML20138D309 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/31/1993 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19303F306 | List: |
| References | |
| OPPD-NA-8301-NP-R05, OPPD-NA-8301-NP-R5, NUDOCS 9302170259 | |
| Download: ML20138D309 (33) | |
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l Table 1 RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301 -NP REY. 05 Title Page Changed the revision number and date.
All Pages Updated revision number.
ii Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens Power Corp. (SPC) to reflect a corporate name change.
vi Updated Revision Sheet.
1 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens 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.
2 Moved previous reference 3-5 to 3-2 and deleted the reference to the INCA topical. The CECOR topical incorporates the INCA methods in the later topical. Adjusted the remaining references to reflect the change.
4 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens Power Corp. (SPC) to reflect a corporate name change.
5 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens Power Corp. (SPC) to reflect a corporate name change.
6 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens 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 HERMITE to derive DNBR power-to-fuel design limit. Revised peaking factors to correctly indicate tilt when appFcable.
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 bumups up to 60 MWD /kg. Updated methodology reference to the latest revision (3-1).
Deleted reference 3-2 since it ivas 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 reviston (5-1,.7-1,7-2 and 7-3).
13 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens Power Corp. (SPC) to reflect a corporate name change.
17 Changed Advanced Nuclear Fuel Corp. (ANF) to Seimens Power Corp. (SPC) to reflect a corporate name change.
9302170259 930205 PDR ADOCK 05000285 P PDR
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OMAHA PUBLIC POWER DISTRICT NUCLEAR ANALYSIS RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW-OPPD-NA-8301-NP Rev. 05 JANUARY 1993 Copy.No.
i ABSTRACT This document is a Topical Report describing Omaha Public Power District's reload core analysis .-
methodology for application to the Fort Calhoun Station Unit No.1.
The report provides an overview of the District's reload core methodology. Analyses performed by the District and its contractors are described. Details of the thermal hydraulic methodology which were previously submitted to the NRC are provided.
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PROPRIETARY DATA CLAUSE I
This document is the property of Omaha Public Power District (OPPD). Proprietary information, indicated by brackets, developed by Seimens Power Corp (SPC), ABB-Combustion Engineering (ABB-CE) and Westinghouse Electric Corporation Gy) has been removed. The SPC, ABB-CE, and W information was purchased by OPPD under proprietan/ information agreements.
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TABLE OF CONTENTS l
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_SECILOR PAGE
1.0 INTRODUCTION
1 2.0 FUEL SYSTEM DESIGN 1 3.0 NUCLEAR DESIGN 1 ,
l 4.0 THERMAL HYDRAUUC DESIGN 2 l
5.0 POSTULATED ACCIDENTS AND 6 TRANSIENTS 6.0 SETPOINT GENERATION 6 7.0 CORE OPERATING LIMITS REPORT 9
8.0 REFERENCES
11 OPPD-NA-8301 -P, Rev. 05 lii
LIST OF TABLES IbBLE TITLE PAGE 4-1 PARAMETER RANGES OF THE SOURCE DATA 13 FOR THE CE-1 CHF CORRELATION AND THE RANGE OF E ABB-CE, AND SPC 14 X 14 FOR FORT CALHOUN VALUES __
4-2 COMPARISONS BETWEEN TORC AND CETOP-D 14 OPPD-N? -8301 -P, Rev. 05 iv
l LIST OF FIGURES FIGURE TITLE PAGE 4-1 ABB-CE FUEL SPACER GRID (DELETED) 15 4-2 SPC FUEL SPACER GRID (DELETED) 16 4-3 AXIAL LOCATION OF FUEL ASSEMBLY SPACER GRIDS 17 4-4 STAGE 1 TORC CHANNEL GEOMETRY FOR FORT CALHOUN UNIT NO.1 18 4-5 STAGE 2 TORC CHANNEL GEOMETRY FOR FORT CALHOUN UNIT NO.1 19 4-6 STAGE 3 TORC CHANNEL GEOMETRY FOR FORT CALHOUN UNIT NO.1 20 4-7 AX1AL POVER DISTRIBUTIONS FORT CALHOUN CYCLE 8 21 4-8 INLET FLOW DISTRIBUTION FOR FORT CALHOUN 4-PUMP OPERATION 22 4-9 EXIT PRESSURE DISTRIBUTION FOR FORT CALHOUN 4-PUMP OPERATION 23 4-10 CETOP-D CHANNEL GEOMETRY FOR FORT CALHOUN 24 UNIT NO.1 6-1 ALGORITHM FOR DNB LCO MONITORING 25 OPPD-NA-e301 -P, Rev. 05 v
m OMAHA PUBLIC POWER DISTRICT-RELOAD CORE METHODOLOGY OVERVIEW REVISION -DATE 00 September 1983 01 June 1985 - j; 02 November 1986 -
03 April 1988 g 04 April 1991 05 __ January 1993 - E-i..
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t OMAHA PUBLIC POWER DISTRICT RELOAD CORE METHODOLOGY OVERVIEW
1.0 INTRODUCTION
Analyses done to license reload cores for 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 060:
however, future reload fuel may be potentially supplied by any of the four U.S. PWR nuclear fuel vendors: Seimens 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 mi.xed core, fuel -
mechanical des!gn and design methods are discussed in References 2-2 through 2-5.
in an effort to further reduce the neutron flux to the reactor vessel welds, full length Hafr'ium flux suppression rods, which are similar to the part length poison rods utilized in Cycle 10, i- will be incorporated into the fuelloading pattern. The poison rods are composed of hafnium
- metal extending the full length 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
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core location number 14 for additional neutron flux reduction to the critical welds.
3.0 NUCLEAR DESIGN The District's nuclear design methodology is discussed in Reference 3-1.
L 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 react'or pressure vessel welds.
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3.0 NUCLEAR DESIGN (Continued) 3.2 Power Distribution Measurement The District utilizes the ABB-CE methodology (Reference 3-2) to mes ure the power distributions, This methodology is discussed in the Cycles 5 and 6 reload submittals and approved in the SER's for these fuel cycles (References 3-3 and 3 -4) .
3.3 Uncertainties and Allowances The power d:stribution uncertainties which are included in the overall analysis of reload cores are:
Parameter Uncertainty 3D Peak, Fq 3-D 6.2%
Integrated Radial Peak Fr 6.0%
Planar Radial Peak, Fxy 5.3%
These values are approved for use in CENPD-153-P (Reference 3-2). A more g detailed d:scussion of the treatment of uncertainties and allowances can be found in Reference 3-5. E
! 3.4 Physics Safety Related Data The physics safety related data are produced using the methodology discussM iri Reference 3-1, i
4.0 THERMAL HYDRAULIC DESIGN 4.1 Steady State DNBR Analysis The steady state DNBR analysis is performed by the District using the TORC /CETOP/CE-1 muthodology (References 4-1,4-2,4-3,4-4 and 4-5).
This methodology was approved for use by the District in Reference 4-6.
4.1.1 Grid Spacer Loss Coefficients l 'The analysis utilizes a D-TOHC model with explicit representation of the loss coefficients associated with ABB-CE and W fue: assemblies. The nominal grid loss coefficients used in thermal hydraulic analysis are:
OPPD-NA-8301-NP Rev. 05 Page 2 of 25
4.0 - THERMAL HYDRAULIC DESIGN (Continued) 4.1 -Steady State DNBR Analvsis (Continued)- g 4.1.1 Grid Spacer Loss Coefficients (Continued)
W Soacer ABB-CE Soacer
[ } Loss Coefficient (K) .
Re = Reynolds Number
,1 These values were obtained by Westinghouse using single phase pressure '<
drop testing of a Westinghouse test assembly and a typical ABB-CE 14 x 14 i
assembly. Single phase hydraulic loss coefficients, previously transmitted in :- ~i i Reference 4-7, contained a [ ] while -
the values given above are best estimate values. Because of the sensitivity. j of DNBR calculations to the difference in spacer grid loss coefficients, the - .f District utilizes the Reynolds number expression for loss coctficients. This --
provides the most accurate representation of the pressure drop across each q spacer grid in the assembly. Thus, the cross flows between adjacent - .
. assemblies in the region of the spacer grid are accurately modeled.
The spacer grid geometries for the ABB JE and W spacer grids are shown in References 2-2 and 2-1, respectively. The spacer grid envelope for '
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l both the ABB-CE and W grids is 8.115 inches by 8.115 inches. The axial-locatio'1 of the ABB-CE. W and SPC spacer grids is shown in Figure 4-3.
-a in D-TORC calculations, the spacer grid loss coefficient for a channel corresponds to the assembly type whenever a channel represents a single assembly or a portion of an assembly. The choice of loss coefficient for lumped channels in D-TORC is made such that the minimum flow is provided to the limiting fuel assembly. The CETOP model employs the spacer grid loss coefficient for the limiting assembly calculated in D-TORC..
The inlet flow fraction of the CETOP model is tuned such that the CETOP .
model produces conservative results with respect to the D-TORC model, which models allfuelassemblies.-
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p 4.0 THERMAL HYDRAUUC DESIGN (Continued) 4.1 Steady State DNBR Analysis (Continued) 4.1.2 CE-1 Correlation The District utilizes the CE-1 correlation for DNBR calculations. The range of data in the data base for the CE-1 correlation la contained in References 4-1 and 4-2, The range of parameters for the CE-1 correlation and corresponding ranges for the ABB--CE and W assemblies are shown in Table 4-1. Because the data for the W fuel assembly is within the range specified in the CE-1 data base, the use of CE-1 correlation is appropriate for the W fuel.
4.1.3 D-TORC and CETOP Models The District utilizes the D-TORC code (Reference 4-3) and the CETOP code (Reference 4-4) to perform thermal hydraulic analysis for the Fort Calhoun reload core. The fraction of inlet flow to the het assembly in the CETOP modelis adjusted such that the model yields appropriate MDNBR results when compared with results of D-TORC analysis for a given range of.
oparating conditions. The fraction of inlet flow is determined for each reload core. The use of this methodology was approved for use by the District in Reference 4-6.
The following paragraphs discuss the application of the CETOP code to the Fort Calhoun reactor, Examples are for the Cycle 8 core (which contained SPC and ABB-CE fuel types). E Thermal margin analysis utilizing the CETOP modelis supported by comparing its predictions for Fort Calhoun Station with those obtained from a detailed TORC analysis. Several operating conditions were arbitrarily -
selected for this demonstration; they are representative, but not the complete set, of conditions which would be considered for a normal DNB analysis.
O"PD-NA-8301 -NP Rev. 05 Page 4 of 25
I 4.0 THERMAL HYDRAUUC DESIGN (Continued) 4.1 Steady State DNBR Analvsis (Continued) 4.1.3 D-TORC and CETOP Models (Continued)
A thermal margin model for 1500 MWt for Fort Calhoun Unit No.1 was developed for the following operating ranges.
Inlet Temperature 450 to 600 *F S'/ stem Pressure 1750 to 2400 psia Primary System 4-Pump Flow Rate, (LCO = 196,000 gpm) 80% to 120% .
Mial Power Distribution -0.'il7 to +0.526 ASI The detailed thermal margin analyses were performed for the sr M core using the radial power distribution and detailed TORC model shown in Figures 4-4,4-5, and 4-6. The appropriate spacer grid loss coefficient was applied to each assembly" channel or partial assembly channel in each stage, in stcge 1, lumped channel 28 utilized the ABB-CE spacer grid loss coefficient because the channel was predominantly composed of ABB-CE fuel. Lumped channels 26 and 27 utilized the SPC spacer grid loss coefficient because either the channel was composed of entirely SPC fuel or contained a single ABB-CE assembly not on a boundary between channels.
The axial power distributions are given in Figure 4-7. These distributions were the most limiting ones generated for the length of the cycle and for the various power dependent insertion limits examined. The core inlet flow and exit pressure distributions used in the analyses were based on the flow model test results given in Figures 4-8 and 4-9. The results of the detailed TORC analyses are given in Table 4-2. The Cycle 14 TORC and CETOP models incorporate the appropriate changes for W fuel design parameters.
The same methods apply to the mixed ABB-CE and W core that were applied to the ABB-CE and SPC mixed core, y The CETOP design model was a total of four thermal hydraulio channels to model the open-core fluid phenomena. Figure 4-10 shows the layout of '
these channels. Channel 2 is a quadrant of the hottest assembly which represents OPPD-NA-8301 -NP, Rev. 05 Page 5 of 25 I
4.0 THERMAL HYDRAUUC DESIGN (Contiriued)-
4.1 S.teady State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued) the average coolant conditions for the remaining portion of the core. The boundary between channels 1 and 2 is open for crossflow; the remaining outer boundaries of channel 2 are assumed to be impermeable and adiabatic. Channel 2 includes channels 3 and 4. Channel 3 lumps the subchannels adjacent to the MDNBR hot channel 4. The " hot" assembly determined frce D-TORC analysis was an SPC assembly. Since CETOD modeis a quadrant of the " hot" assembly, the SPC spacer grid loss coefficient was used in the analysis.
The CETOP model described above was applied to the same cases as the detailed TORC analyses. The results from the CETOP model analyses are compared with those from the detailed analyses in Table 4-2. It was found that a constant inlet flow split providing hot assembly inlet mass velocity of
[ ] of the core average value is appropriate for 4-pump operation so that MDNBR results predicted by the CETOP model are either conservative or accurate for the v%Ie 8 core. The uncertainties associated with the thermal hydraulic antlysis are combined statistically (Reference 4-8). In this method, the impact of component uncertainties on DNBR is assessed and the SAFDL is increased to include the effects of the uncertainties.
5.0 POSTULATED ACCIDENTS AND TRANSIENTS The postulated accidents and transients are analyzed using the methodology discussed in Reference 5-1.
6.0 SETPOINT GENERATION The District utilizes the methodology discussed in CENPD-199-P (Reference 6-1) to generate setroints for Fort Calhoun Station. TM District's reactor physics methodology is discussed in Reference 6-2.
l OPPD-NA-8301 -NR Rev. 05 Page 6 of 25 i
6.0 SETPOINT GENERATION (Continued)
The axial shape analysis is performed using the HERMITE code. The power-to-fuel design limit on centerline melt is derived using the HERMITE code with the appropriate comb %ations of planar radial peaking factor, F, and axial power distribution. The power-to-fuel decign limit on DNBR is derived using the HERMITE code with the appropriate combinations of integrated radial peaking factor, Fn and axial power distribution.
The thermal margin analysis is dons using the CETOP code with the appropriate combinations of the integrated radial peaking factor, Fr axial power distribution, RCS inlet g temperature, and RCS pressure.
The Fort Cc!houn Reactor Protection System utilizes the " standard" local power density trip and TM/LP trip. The sequential CEA withdrawal is analyzed using the methods described in Reference 6-1 and not included in the TM/LP tnp considerations. The RCS depressurization event provides the transient analysis input into the TM/LP trip.
6.1 innofe LCO Monitoring The Better Axial Shape Selection System (BASSS) monitors the Limiting Conditions for Operation on peak linear heat rate and departure from nucleate boil:ng using as input the data available from the MINI-CECOR code and the plant computer. This arrangement is similar to the one used by Baltimore Gas and Electric at their Calvert Cliffs Units, and described in the ABB-Combustion Engineering Setpoint Topical (Ref. 6-1).
MIN!-CECOR is a plant computer version of ABB-Combustion Engineering's CECOR code, it is used in this application to synthesize the following parameters from readings of the fixed in-core detectors:
T
- 1. The three-dimensional power peaking factor (Fq)
- 2. The core average axial shape index (I)
- 3. The total planar radial peaking factor (FxyT)
- 4. The total integrated radial peaking factor ( FrT)
These inputs to BASSS are descnptive of the existing core power distribution.
The inputs to BASSS obtained from the plant computer are the following:
- 1. Measured core power level
OPPD-NA-8301-NP Rev. 05 Page 7 of 25
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6.0 l SETPOINT GENERATION (Continued) 6.1 Incore LCO Monitorina (Continued)
BASSS consists of two algorithms: one for peak linear heat rate monitoring 1 and another for DNB monitoring. The peak linear heat rate algorithm uses the 3-D power peaking factor and the measured core power level to calculate the core peak linear heat rate. The algorithm applies appropriate uncertainties and allowances (per the Technical Specifications) to the 3-D peaking factor. The measured peak linear heat rate is compared to the monitoring limit, which is based on both LOCA and AOO transient analysis considerations, and an alarm is activated when the monitonng limit is exceeded. The power operating limit on linear heat rate is also calculated and displayed as an indication of the available operating margin. The DNB algorithm is an improvement over the excore ASI monitoring system in that it uses in-core axial shape index, CEA group position and the radial peaking .
factors to establish the plant's power operating limit. An alarm is activated -
when the power operating limit is exceeded. A gain in operating margin results from the following:
- 1. A reduction in AS1 uncertainty due to the use of in-core ASI versus ex-core ASI.
- 2. K'iowledge of the actual CEA group position versus the ex-core system's assumption that the CEAs are inserted to the PDll's transient ,
insertion limit.
- 3. Knowledge of the actual radial peaking factors versus the ex-core system's assumptions that radial peaks are at the Technical Specification limits.
OPPD-NA-8301-NP Rev. 05 Page 8 of 25
6.0 SETPOINT GENERATION (Continued) 6.1 Incore LCO Monitorina (Continued)
BASSS is also provided with the capability to monitor the Limiting Conditions for Operation on Fy T and Fr T. If the Technical Specification for Operation on FyT or FrTare exceeded during normal plant operation, BASSS will activate an alarm and calculate the proper trade-off with maximum allowed power that ensures that the Axial Power Distnbution and Thermal Margin / Low Pressure Trips remain conservative. An alarm is activated if the measured power levelis higher than the allowed power level 6.2 UncPJ1ainties The uncertainties are treated statistically in the District's setpoint analysis as identified in references 6-1 and 6-3.
7.0 CORE OPERATING UMITS REPORT The core operating limits report (COLR) was incorporated into the plant Technical Specifications as part of the Cycle 14 reload license application. The COLR utilizes the guidance provided in Reference 7-1. The reload analysis for Fort Calhoun for generating the COLR values utikzes the methods described in References 7-2,7-3 and 7-4. The use of these NRC approved methodologies does not permit substantial discretion on the part of OPPD and does not require substantial engineering judgement to be utilized to derive the cycle specific parameter limits included in the COLR.
OPPD-NA-8301-NP Rev. 05 Page 9 of 25
7.0 CORE OPERATING LIMITS REPORT (Continued)
The COLR consists of the following items: g Thermal Margin / Low Pressure for 4 Pump Operation Refueling Boron Concentr ation Limiting Conditions for Operation for Excore Monitoring of LHR Limiting Conditions for Operation for DNB Monitoring Power Dependent insertion Limit (PDIL)
Unrodded Integrated Radial Peaking Factor ( F,)
Unrodded Planar Radial Peaking Factor (Fn)
Allowable Peak Linear Heat Rate vs. Burnup FnT,FrT and Core Power Limitations Core inlet Temperature in accordance with Reference 7-1 requirements, updates to the COLR during the operating cycle will be issued to the NRC (NRR, Region IV and Senior Resident inspector). E l
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8.0 REFERENCES
Section 2 References 2-1 WCAP-12977 " Westinghouse Reload Fuel Mechanical Design Evaluation for the Fort Calhoun Station Unit 1". June 1991.
2-2 " Omaha Batch M Reload Fuel Design Report", CEN-347(0)-P, Rev. 01, January 1987.
2-3 " Fuel Rod Maximum Allowable Gas Pressure", CEN-372-P-A, March 1990.
2-4 Letter, LD-92-118 from S. A. Toelle (ABS-CE) to R. C. Jones (NRC), " Request for Generic Approval of the Acceptability of 1-Pin Burnup Umit of 60 MWD /kg for C-E i 14x1' PWR Fuel" dated December 18,1992.
2-5 "Verifiwion of the Acceptab!!!ty of a 1 -Pin Burnup Limit of 60 MWD /kg for Calvert Cliffs Units 1 and 2", CEN-382(B)-P, January,1989.
Section 3 References 3-1 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification",
OPPD-NA-8302-R Rev. 03, January 1993.
3-2 " INCA /CECOR Power Peak Uncertainty", CENPD-153-8 Revision 1-P-A, May 1980.
3-3 Letter from R.W. Reid (NRC) to T.E.Short (OPPD), December 5,1978.
3-4 Letter from R.W. Reid (NRC) to W.C. Jones (CPPD), April 1,1980.
3-5 " Statistical Combination of Uncertainties", CEN-257(0)-P-A, Parts 1 and 3, November 1983, including Supplement 1-R August 1985.
Section 4 References 4-1 "CE Cntical Heat Flux", CENPD-162-P-A, Part 1, Combustion Enqineering, September 1976.
4-2 "CE Catical Heat Flux", Part 2, CENPD-207-P-A, Combustion Engineering June 197o.
4-3 "A Computer Code for Determining the Thermal Margin of a Reactor Core,"
CENPD-161-P-A, TORC Code, Combustion Engineering April 1986. R 4-4 *CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 & 2."
CEN- 191(B)-P, Combustion Engineering, December 1981.
4-5 Letter from Cecil Thomas (NRC) to Mr. A. E. Scherer (CE), November 2,1984.
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8.0 REFERENCES
(Continued)
Section 4 References (Continued) i 4-6 Letter frorri R. A. Clark (NRC) to W. C. Jones (OPPD), March 15,1983.
l 4-7 Letter from W. C. Jones (OPPD) to R. W. Reid (NRC). December 4,1979.
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- Statistical Combination of Uncertainties," CEN-257(0)-P- A Part 2, November g 1983, including Supplement 1 -P, August 1985. 3 i
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 l
Section 6 References 6-1 "CE Setpoint Methodolocy." CENPD-199-P, Revision 1-P- A, January 1986.
l 6-2 " Reload Core Analysis Methodology, Neutronics Design Methods and Venfication,"
OPPD-NA-8302-P- A, Rev. 03, JanJary 1993.
, 6-3 " Statistical Combination of Uncertainties," CEN-257(0)-P- A. Parts 1,2, and 3, l November 1983, including Supplement 1-R August 1985. E I
Section 7 References j 7-1 " Removal of Cycle-Specific Parameter Limits from Technical Specifications", NRC Generic Letter 88-16. October 4,1988.
7-2 "Retoad 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-P, 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 Page 12 of 25
TABLE 4-1 PARAMETER RANGES OF THE SOURCE DATA FOR THE CE-1 CHF CORRELATION AND THE RANGE OF WESTINGHOUSE, ABB-CE AND SPC 14 x 14 FOR FORT CALIIOUN VALUES CORRELATION ABB-CE SPC WESTINGHOUSE g PARAMETER RANGE RANGE R.MGE RANGE l Pressure (psia) 1785 to 2415 N/A N/A N/A l
Local Coolant Quality .16 to .20 N/A N/A N/A 1
Local Mass Velocity 0.87x!(f to 3.21x106 N/A N/A N/A 2
(Ibm /hr-ft )
i Subchannel Wetted Equiv. .3583 to .4043 to .4010 to .4043 to ,
Diameter (in) .5447 .5449 .5402 .5449 I Subchannel Heated 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 OPPD-NA-8301 -NP, Rev. 05 Page 13 of 25
TABLE 4-2 COMPARISONS BETWEEN TORC AND CETOP-D L-Axial Elev.
Operating Parameters MDNBR Ouality at MDNBR of MQNHR.{tn)
Detailed CETOP Detailed CETOP TORC Inlet TORC Inlet inlet Avg Mass Core Avg. Shape Relative Flow Relative Flow Pres Temp Velocity Heat Flux Index Flow in Factor Flowin Factor Detailed (psia) ( F) (1) (2) (ASI) IAc5 [ ] Loc 5 [ ] TORC CETOP L _ _
l 1750 450 1.7432 242409 .517 l 2100 450 1.7432 257008 .517 2250 450 1.7432 261195 .517 l 2400 450 1.7432 264118 .517 l 2100 545 2.1790 216494 .517 1750 600 1.7432 149283 .206 l
2l00 600 '1.7432 168727 .206 2250 600 1.7432 176398 .206 l 2400 600 1.7432 184260 206 2100 545 2.1790 257118 .206 i 2100 545 2.1790 '282778 .004 2100 545 2.1790 298644 .203 1750 450 1.7432 295262 .527 1750 545 1.'432 227063 .527 1750 600 1.7432 147319 .527 2100 545 2.1790 255014 .527 _ _
6 (1) (10 lbm/hr-ft2)
! (2) (BTU /hr-ft2) l OPPD-NA-8301-NP Rev. 05
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DELETEl' ABB-CE Omaha Public Power District Figure FUEL SPACER GRID Fort Calhoun Station - Unit No.1 4-1 OPPD-NA-8301-NR Rev. 05 Page 15 of 25
e DELETED '
.I SPC Omaha Public Power District Figure .
FUEL SPACER GRID - Fort Calhoun Station - Unit No. L1 4-2 OPPD-NA-8301 -NP, Rev. 05 Page 16 of 25
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. - _ _ _ _ _ - _ _ _ _ _ _ _ _ . . _ - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ - _ _ _ ----------s
I ADB.=CE W SP_C I
134.937 134.759 134.95 16.25 16.089 16.15 l
118.687 118.67 118.808 16.813 16.81 16.81 101 875 101.860 101.995 16.813 16.81 16.81 85.063 85.050 85.182 16.813 16.81 16.81 68.25 68.240 68.369 16.813 16.81 16.81 51.437 ._ 51.43 51.556 16.813 16.81 16.81 34.625 _ 34.62 34.743 16.813 16.81 16.81 17.812 17.81 17.93 Assembly Lower Tie Plate AXIAL LOCATION OF FUEL Omaha Public Power District Figure ASSEMBLY SPACER GRIDS Fort Calhoun Station - Unit No.1 4-3 OPPD-NA-8301 -NP Rev. 05 Page 17 of 25
,-4 . .-. _ _ . _ . _ _ _ _ _
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CL UtAhhEL NUMBER A55EP.!LY AVERAGE M0!AL 01 02 POWEA FACT 02
.6963- .4648 03 04 05 06 - 07- "
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J STAGE 1 TORC CHANNEL Omaha Public Power District - Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station-Unit No.'1 ~4-4
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GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station Unit No.'1 :4 5:
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AX!AL POWER DISTRIBUTIONS Omaha Pub!!c Power District Figure FORT CALHOUN CYCLE 8 Fort Calhoun Station Unit No.1 4-7 OPPD-NA-8301 -NP, Rev. 05 Page 21 of 25
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1 VALUE ;! NOTE! THE MT10 or THE St# IDLE 01 02 INtET PA55 VELC0!!Y 70 fME C;RE AvtMGE PA55 Vtt001TT 0.88 0.94 03 04 05 04 01 0.93 0.94 0.94 0.98 0.99 05 09 10 11 12 13 0.94 0.95 1.03 1.03 0.97 1.00 14 15 16 !? 18 19 -
1.02 0.99 1.05 1.04 1.00 1.07 20 21 22 23 24 25 1.02 1.02 1.07 1.04 0.97 1.05 0.91 27 28 29 30- 31 32 1.05 1.00 1.05 33 1.05 1.04 1.06 0.94 34 35 35 17 38 39 1.05 1.01 .l.09 1.08 1.01 1.10 INLET FLOW DISTRIBUTION Omaha Public Power District Figure FOR FCS. 4-PUMP OPERATION Fort Calhoun Station Unit No.1 4-8
- i. OPPD-NA-8301 -NP, Rw. 05 Page 22 of 25
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AVERACE CCAC C117 ;Rti!URE !!1 P1F gg,g). 7,); .
03 04 05 06 01 27.60 16.99 27.60 44.59 57.32 : -
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7.431 23.4 39.3 33 22.3 7.43 4.25-16.09 34 35 36 37 - 38. 39 31.84 21.2 28.1 20.2 ' . 7.43 - 8.49 i v L
EXIT PRESSURE DISTRIBUTION - s Omaha Public Power District Figur_e
-FOR FCS. 4 PUMP OPERATION- Fort Calhoun Station-Unit No.1 c 4 -
- OPPD-NA-8301 -NP Rev,05 - .
Ii ' Page 23 of 25.-
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ALGORITHM FOR DNB. Omaha Public Power District- Figure LCO MONITORING Fort Calhoun Station-Unit No.1 61.
OPPD-NA-8301 -NP, Rev. 05 Page -25 of 25 I