ML20073H252
| ML20073H252 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/30/1991 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19298E441 | List: |
| References | |
| OPPD-NA-8301-NP-R04, OPPD-NA-8301-NP-R4, NUDOCS 9105070034 | |
| Download: ML20073H252 (34) | |
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l OMAHA PUBLIC POWER DISTRICT I 1
l NUCLEAR ANALYSIS RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301-NP Rev. 04 APRIL 1991 Copy No, 9105070034 910419 POR ADOCK 05000285 P PDR
s ABSTRACT This document is a lopical Report describing Omana 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.
OPPD-NA-8301-NP, Rev. 04 i
s PROPRIETARY DATA CLAUSE This document is the property of Omaha Public Power District (OPPD) and contains the nonproprietary information, indlcated by brackets, developed by Advanced Nuclear Fuels Corp (ANF), ABB Combustion Engineering (CE) and Westinghouse Electric Corporation (W). The ANF, CE, and Westinghouse information was purchased by OPPD under proprietary information agreements.
OPPD NA-8301-NP. Rev. 04 ,
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s TABLE OF CONTENTS SECTION PAGE
1.0 INTRODUCTION
1 2.0 FUEL SYSTEM DESIGN 1 3.0 NUCLEAR DESIGN 1 4.0 THERMAL HYDRAULIC DESIGN 2 5.0 POSTULATED ACCIDENTS AND 6 TRANSIENTS 6.0 SETPOINT GENERATION 6 7.0 CORE OPERATING LIMITS REPORT 9 g
80 REFERENCES 11 R
OPPD-NA-8301-NP, Rev. 04 lii
1 4 4 LIST OF TABLES TABLE T!TLE PAGE 4-1 PARAMETER RANGES OF THE SOURCE DATA 13 FOR THE CE-1 CHF CORRELATION AND THE RANGE OF ANF. CE, AND W 14 X 14 FOR FORT g
CALHOUN VALUES 4-2 COMPARISONS BETWEEN TORC AND CETOP-D 14 OPPD-NA-8301-NP, Rev. 04 iv
LIST OF FIGURES EJEuBE TITLE P. AGE 4-1 CE FUEL SPACER GRID (DELETED) 15 g 4-2 ANF FUEL SPACER GRID (DELETED) 16 l
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 AXIAL POWER 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-8301-NR Rev. 04 v
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OMAHA PUBLIC POWER DISTRICT l RELOAD CORE METHODOLOGY OVERVIEW REVISION DATE 1
00 September 1983 L
01 June !985 02 November 1986 03- April 1988 04 April 1991 - g i
i OPPD-NA-8301-NP, Rev. 04 vi
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OMAHA PUBLIC POWER DISTRICT RELOAD CORE METHODOLOGY OVERVIEW
1.0 INTRODUCTION
Analyses done to license reload cores for Fort Calhoun Station consist of the analys: i performed by the Omaha Public Power Olstrict and the analysis performed by the nu lear fuel vendor The current nuclear fuel vendor is Westinghouse Electric Corporation (W); h iwever, future reload fuel may be potentially supplied by any of the four U.S. PWR nuclear fu l vendors:
Advanced Nuclear Fuels Corp. gANF), ABB Combustion Engineering (CE), Westinghc ase, or Babcock and Wilcox. The following sections discuss the reload analyses and consc :idate 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 fue vendor.
The fuel mechanical design and design methods utilized for Fort Calhoun Station by Westinghouse are describea in Reference 2-1. Combust!on Engineering, the co-res dent fuel in the m!xed core, fuel mechanical design and design methods are discuased in Ref rences 2-2 and 2-3.
In an effort to furtner reduce the neutron flux to the reactor vessel welds, full length H. fnium flux suppression rods, which are similar to the part length poison rods utilized in Cycle 1f , will be incorporated into the fuel loading pattern. The poison rods are composed of hafnion metal over the full length of the active fuel. Inert material comprises tne balance of the rod They will reside in the uuter guide tubes of quarter core assembly numbers 1,2 and 8 The f el system design will also incorporate four natural uranium fuel assemblies in quarter core loca ion 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.
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 ve sel welds.
OPPD-NA-8301-NP, Rev. 04 Page 1 of 25
3.0 NUCLEAR DESIGN (Continued) 3.2 Ecwe, D;stribution Menurement The Distnet utikzes the CE methodology (Reference 3-2) to measure the po) er distobutions This methodo!ogy is disctissad in the Cycles 5 and 6 reload s bmittals and approved in the SER's for these fuel cycles (References 3 3 and 3-4) 2.3 Uncettair.ltitL0ttd Ailowances The power distrihyton uncertenties which are included !n the overall analysi of reload cores are:
Entamglel Uncertajry -
3D Peak. Fq 3-D 0.2%
unegrated Radial Peak, F, 60%
Planar Radial Peak. F,y 5.3 %
These values are approved foi ute in CENPD-153-P (Reference 3-5). A mc e detailed discussion of the treatment of uncertainties and allowances can be found in teference 3- 0.
3.4 Physics Safety RejalEQ.Qala The physics tafety related datu are produced using the methodology discus ed in Reference 3-1.
4.0 THERMAL HYDRAUL!C DESIGN 4.1 Sicady State DNBR Analysis Tne steady state DNBR analysis is performed by the District using the TORC /CETOP/CE-1 methodology (References 4-1,4-2,4-3,4-4 and 4-5). This methodology was approved for use by the District in Reference 4-0.
4.1.1 Grid Spacer Loss Coefficients The analysis utihtes a D-TORC model with wpicit representation o the loss coefficients assoated with CE and W Uel assembhes The nomini gnd loss g coeffic.ents used in tr.3rmal hydrauhc analysis are:
OPPD-NA-8301-NP, Rev. 04 Page 2 of 25
4.0 THERMAt. HYDRAULIC DESIGN (Continued) 4.1 SteadLStalPLQNBELA!1a!m (Continued) 4.1.1 Grid Spacer Loss Coefficients (Continued)
__W_Spret CE_Saftter l
[ ] Loss Coefficient (K)
Re a Reynolds Number These values were obtained by Westinghouse using single phase pi esstre drop testing of a Westinghouse test assembly and a typical CE 14 x 14 assembly. Single phase hydraulle loss coefficients, previously trant nitted in Reference 4-7, contained 0 ( ) vhile the values given above are best estimate values. Because of the nonst Aty of DNBR calculations to the difference in spacer grid loss coefficients, he District utillzes the Reynolds number expression for loss coefficients This 6 ovices the most accurate representation of ihe pressure drop across each ! Jacer gnd in the assembly. Tnus, the cross flows between adjacent assemblit iin the region of the spacer grid are accurately modeled.
The spacer gnd geometries for the CE and W spacer gods are show in References 2-2 and 2-1, respectively. The spacer grid envelope fc both the CE and W gods is 8115 inches by 8.115 inches. The axial location 't the CE, W and ANF spacer grids is shown in Figure 4-3.
In D-TORC calcu;ations, tne spacer grid loss coefficient for a chann I corresponds to the assembly type whenever a channel represents a ningle assembly or :, portion of an assembly. The choice of loss coefficier for lumped channels M D-TORC is made such that the minimum flow is arovided to the limiting fue: assembly. The CETOP model employs the spact grid loss coefficient for the limiting assembly calculated in D-TORC. The snit flow fraction of the CETOP model is tuned such that the CETOP model p: iduces conservative results with respect to the D-TORC model which mod s all fuei assemblies.
OPPD-NA-8301-NP, Rev. 04 Page 3 of 25
A 4.0 THERMAL HYDRAULIC DESIGN (Continued) 4.1 Sleady_ State DNBR Analysis (Continued) 4.1.2 CE-1 Correlation The District utilizes the CE-1 correlation for DNBR calculations. The ange of data in the data base for the CE-1 correfetion is contained in Refere ces 4-1 and 4-2. The range of parameters for the CE-1 correlation and cori isponding ranges for the CE and W assemblies are shown in Table 4-1. Becat 3e the data for the W fuel assembly is within the range specified in the CE. drita I 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 CET( P code (Reference 4-4) to perform thermal hydraulic analysis for the Fort Ce noun reload core. The fraction of inlet flaw to the hot assembly in the CE' DP model is adjusted such that the model yields appropriate MDNBR results v ten compared with results of D TORC analysis for a given range of opet iting conditions. The traction of inlet flow is determined for each reload c ire. The use of this methodology was approved for use by the District in Reft ence 4-0 The following paragraphs discuss the application of the CETOP cod to the fort Calhoun reactor. Examples are for the Cycle 8 core (which cont inned ANF g and CE fuel types). 5 Thermal margin analysis utilizing the CETOP model is supported by :omparing its predictions for Fort Calhoun Station with those obtained from a di talled TORC analysis. Several operating condttions were arbitrarily select' d for this demonstration, they are representative, but not the complete set, of onditions which would be considered for a normal DNB analysis.
OPPD-NA-8301-NP, Rev. 04 Page 4 of 25
4.0 THERMAL HYDRAUUC DESIGN (Continued) 4.1 Steady _ State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued)
A thermal marg;n inodel for 15',10 MWt for Fort Calhoun Unit No 1 wt 5 developed for the following operating rangs inlet Temperature 45 to 600 'F System Pressure 1750 to .'400 psia Primary System 4-Pump Flow Rate, (LCO = 193.000 ppm) BO' to 120%
Axtal Power Distribution -0 517 to 4 520 ASI The detailed thermal margin analyses were performed for the sampi. core using the radial power distribution and detailed TORC model shown n Figures 4-4. 4-5. and 4-0 The appropriate spacer grid loss coeff:clent wat applied to each " assembly" channel or partial assembly channel in each stag ( in stage 1, lumped channel 28 utilized the CE spacer pid iosa coefficient be ause the channel was predominantly composed of CE fuel Lumped channel 26 and 27 utilized the ANF spacer grid loss coefficient because either the c: annel was composed of entirely ANF fuel or contained a single CE assembly n t on a boundary between channels. The axial power distr;butions are givet in Figure 4-7. These distributions were the most limiting ones generated for 1 e length of the cycle and for the various power dependent insertion limits ext nined.
The core inlet flow and exit pressure distributions used in Ine analys s were based on the flow model test results given in Figures 4-8 and 4-9. he results of the detailed TORC analyses are given in Table 4 2. The Cycle 14 TORC and CETOP models incorporate the appropdate changes for W fuel . esign parameters. The same methods apply to the mixed CE and W core iat were applied to the CE and ANF mixed core.
The CETOP design model was a total of four thermal hydraulle chan els to model the open-core fluid phenomena Figure 4-10 shows the lay .it of these channels. Channel 2 is a quadrant of tne hottest assemely wt :h represents OPPD-NA-8301-NP, Rev. 04 Page 5 of 25
9 4 4
-O 4.0 THERMAL HYDRAUUC DES!GN (Continued) 4.1 Steadv State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued) the average coolant conditions for the remalising portion of the core. The houndary between channels 1 and 2 la open for crossflow; the rema ling outer boundaries of channel 2 are assumed to be impermeable and adiab itic.
Channel 2 includes channels 3 and 4. Channel 3 lumps the subchat ,els adjacent to the MDNBR hot channel 4. The
- hot" assembly determli ed from D-TOHC analysis was an ANF assembly. Since CETOP models a c iadrant of the ' hot
- assembly, the ANF spacer grid loss costficient was used it the analysis.
The CETOP model describid above was applied to the same car,et os the i detailed TORC analyses. The results from the CETOP model analyt is are .
compared with those from 'he detailed analyses in Table 4-2. It 'Not found that a constant inlet flow split providing hot assembly inlet mass velocity of [ ] of the core average value is oppropriate for 4-r Jmp operation so that i IDNDR results predicted by the CETOP model are either conservative or actarate for the Cycle 8 core, The uncer1ainties associated with the thermal hyd tulle analysis are combined statistically (Reference 4-8). In this mothod, he impact of component uncertainties on DNBR is assessed and the SAFDL is nereased to include the effects of the uncertaintles.
5,0 POSTULATED ACCIDirNS AND TRANSIENTS The postulated accidents and transients are analyzed using the methodology discus ed in Reference 5-1, 6,0 SETPOINT GENERATION The District uti!!zes the methodology discussed in CENPD-199-P (Reference 6-1) to generate setpoints for Fort Calhoun Station. The District's reactor physics methodology is dis assed in Reference 6-2.
OPPD-NA-8301-NP, Rev. 04 Page 6 of 25
- - - - - ,- v vg -,*gmp. 3 iwp g .,w.s- c.-.eyy m--c.. r-em- e- i.--e bg.e* m.,, , . _____s- .-.P--*---u + ma- _m.-m-- -------h*__-
6.0 SETPOINT GENERATION (Continued)
The scram reactivity r.urves are produced using the QUlX code. The power-to-fuel ( ! sign limt!
on centerline melt is derived using the OUlX code with the appropriate combination *, af planar radial peaking factor, F/, and axial power distnbution.
The thermal margin analysis is done using the CETOP code with the appropriate cor o! nations of the integrated racial peaking f actor, FnT , axial power distribution. RCS inlet temoe iture, and RCS pressure.
The fort Calhoun RPS utilizes the " standard" local power density trip and TM/LP trip The 1
sequential CEA withdrawalis analyzed using the methodt described in Reference O- I and not l included in the TM/LP trip considerations. The RCS depressurization event provides he g l transient analysis input into the TM!LP trip 3 l 0.1 incore LCO McIntoring The Better Axial Shapo Selection System (BASSS) monitors the Limi ng I Conaitions for Operation on peak linear heat rate and departure fron nucleate bolling using as input the data available from the MINI-CECOR codi and the plant computer. This arrangement is similar to the one used by Balt: note Gas and Electric at their Calvert Cliffs Units, and described in the Combt .tlon Engineenng Sotpoint Topical (Ref. 6-1),
MINI-CECOR is a plant computer version of Combuetion Engineerin 's CECOR code. It is used in this application to synthesize the followir a parameters from readings of the fixed in core detectors:
- 1. The three-dimensional power peaking factor (Fq)
- 2. The core average axial shape index (1)
- 3. The total planar radial peaking factor (FxyT )
- 4. The total integrated radial peaking factor ( fnT)
These inputs to BASSS are descriptive of the existing core power di irlbution, The inputs to BASSS obtained from the plant computer are the folios ing
- 1. Measured core power level
, , I 1
i
)
6.0 SETPOINT GENERATION (Continued) i i
6.1 incore LCO Monitoring (Continued) 1 l
1 BASSS consists of two algorithmm: one for peak linear heat rate mc storing and another for DNB monitoring The peak linear neat rate algorithn uses the 3-D power peaking factor and the measured core power level to coi ulate the core peak linear heat rate. The algorithm applies appropriate uncer tinties and allowances (per the Technical Speelfications) to the 3 D peaking fat or. The measured peak linear heat rate is compared to the monitoring limit, rhich is based on both LOCA and AOO transient analysis considerations ar 1 an alarm is activated when the monitoring limit is exceeded. The power oper ting limit on linear heat rate is also calculated and displayed as an indication af the available operating margin. The DNB algorithm is an improvement ( ser the excore ASI monitoring system in that it uses in-core axial shape ind< x, CEA group position and the radial peaking factors to establish the plant's power operating limit An alarm is activated when the power operating limi is exceeded. A gain in operating margin results from the following:
- 1. A reduction in ASI uncertainty due to the use of in-core ASI vers' s ex-core ASI.
r
- 2. Knowledge of the actual CEA group position versue, the ex-core system's assumption that the CEAs are inserted to the PDIL's transient insertir a limit. 'I
- 3. Knowledge of the actual radial peding factors versus the ex-co e system's assumptions that radial peaks are at the Technical Speelfu 3 tion limits.
P
-t
~
OPPD-NA-8301-NP, Rev. 04 Page 8 of 25
_ _ _ _ _ ~ _ _ . _ . . _ . _ . _ _ . _ _ _ ..._ _ __..:_ _ _ _ _ _ _ _ . _ _ _ -
i 6.0 SETPOINT GENERATION (Continued) 6.1 Incore LCO Monitoring (Continued) l I
l 1
1 i
BASSS is also provided with the capability to rnonitor the Lim! ting Condition for Operation on F nT and F,T. If the Technical Specification for Operation on ,y i or F,T are exceeded during normal plant operation, BASSS will activate an alarm a d calculate the proper trade-off with maximum allowed oower that ensures tha the Axlal Power Distribution and Thermal Margin / Low Pressure Trips remain conservat /e. An alarm is activated if the measured power level is higher than the allowed pot er level.
6.2 Uncertainties The uncertainties are treated statistically in the District's setpoint analycis at identified in references 6-1 and 6-3.
7.0 CORE OPERATING LIMITS REPORT g The core operating limits report (COLR) will be incorporated into the plant Technicai Specifications as part of the Cycle 14 reload license application. The COLR will utili e the guidance provided in Reference 7-1. The reload analysis for Fort Calhoun to gener te the COLR values will utilize the methods described in References 7-2. 7-3 and 7-4. Tt i use of these NRC approved methodologies does not permit substantial discretion on the p; 1 of OPPD and does not require substantial engineering judgement to be utilizeu to deriv the cycle specific parameter limits included in the COLR.
OPPD-NA-8301-NP, Rev. 04 Page 9 of 25
7.0 CORE OPERATING LIMITS REPORT (Continued) g The COLR will consist of the following items-Thermal Margin / Low Pressure LSSS for 4 Pump Operation Refueling Boron Concentration Limiting Conditions for Operation for Excore Monitoring of LHR Power Dependent insertion Limit (PDIL)
Unrodded integrated Radial Peaking Factor ( Fr)
Unrodded Planar Radial Peak 6ng Factor (F,y)
Allowable Peak Linear Heat Rate vs. Burnup F,yT, F,7 and Core Power Limitations in accordance with Reference 7-1 requirements, updates to the COLR auring the opi ating '
cycle will be issued to the NRC (NRR, Region IV and Senior Resident inspector) cor :urrent with internal OPPD distribution.
OPPD-NA-8301-NP, Rev. 04 Page 10 of 25
l
8.0 REFERENCES
Section 2 References 2-1 " Westinghouse Reload Fuel Mechanical Design Evaluation for the Fort Calhc in Station Unit 1". March 1991, 2-2 " Omaha Batch M Reload Fuel Design Report", CEN-347(0)-P, Rev. 01, Janu ry 1987.
2-3 ' Fuel Rod Maximum Allowable Gas Pressure", CEN-372-P-A, March 1990.
Section 3 References 3-1 " Reload Core Analysis Methodology, Neutronics Design Methods and Verifl( ttion",
OPPD-NA-8302-P- A, Rev. 02, April 1988. E 3-2 " INCA, Method of Analyzing incore Detector Data in Pressurized Water Reat ars",
CENPD-145-P, April 1,1975 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 (OPPD). April 1,1980.
3-5 " INCA!CECOR Power Peak Uncertainty". CENPD-153-P, Revision 1-P- A, M y 1980.
3-6 " Statistical Combination of Uncertainties", CEN-257(0)-P- A, Parts 1 and 3, ! ovember g 1983.
Section 4 References 4-t "CE Critical Heat Flux", CENPD-162-P- A, Part 1, Combustion Engineering, eptember 1976.
4-2 "CE Critical Heat Flux", Part 2, CENPD-207-P- A, Combustion Engineering, , une 1976.
4-3 "A Computer Code for Determining the Thermal Margin of a Reactor Core "
CENPD-161-P, TORC Code, Combustion Engineering, July 1975.
4-4 "CETOP-D Code Structure and Mode:Ing 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 2,198e OPPD-NA-8301-NP, Rev. 04 Page 11 of 25
8.0 REFERENCES
(Continued)
Section 4 References (Continued) 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, Novemt tr 1983. E Section 5 Referenqes 5-1
- Reload Core Analysis Methodology, Transient and Accident Analysis Methe is and Verification," OPPD-NA-8303-P, Rev. 03, April 1991. 3 SEllon 6 References 6-1 *CE Setpoint Methodology." CENPD-199-P. Revision 1-P- A. January 1986 3 6-2 *Rebad Core Analysis Methodology, Neutronics Design Methods and Verific ition,"
OPPD-NA-8302 P- A. Rev. 02, April 1988.
6-3
- Statistical Combination of Uncertainties
- CEN-257(0)-P- A, Parts 1,2, and November 1983.
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. 04 April 1991.
7-3 " Reload Core Analysis Methodology, Neutronics Design Methods and Verific ition,"
OPPD-NA-8302-P-A, Rev. 02, April 1988.
7-4 " Reload Core Analysis Methodology, Transient and Accident Analysis Mothc is and Vertrication," OPPD-NA-8303-P, Rev. 03, April 1991.
OPPD-NA-8301-NP, Rev. 04 Page 12 of 25
4 TABLE 4-1 PARAMETER RANGES OF THE SOURCE DATA FOR THE CE-1 CHF CORRELATION AND THE RANGE OF WESTINGHOUSE, CE AND ANF 14 x 14 FOR FORT CALHOUN VALUES CORRELATION CE ANF WESTit 3 HOUSE EABAfdETER RAffGE rat 4GE RANGE RAIGE Pressure (psia) 1785 to 2415 N/A N/A NA Local Coolant Quality .16 to .20 N/A N/A N4 Local Mass Velocity 0.87x106t o
?
(Ibm /ht-ft ) 3.21x106 N/A N/A N4 Subchannel Wetted Equiv. .3588 to .4043 to 4010 to .40 I to Diameter (in) .5447 .5449 .5402 .5e 19 Subchannel Heated Equiv. .4713 to .5334 to .5270 to .53: I to Diameter (in) .7837 .7840 .7780 .7f10 Heated Length (in) 84 to 150 128 128 1. 3 Grid Spacing (in) 14.2 to 18.25 18.8 16.8 11 8 OPPD-NA-8301-NP, Rev. 04 Page 13 of 25
TABLE 4-2 COMPARISONS BETWEEN TORC AND CETOP-D As il Elev.
(ineratine l'arameters MDNUR Ouality at MDNilR of M JNilR l'ini Detailed CETOP 1)etailed CETOP TORC Inlet TOl(C Inlet inlet Avg Mass Core AvF. Shape Itriative Flow Relative Flow 1 Pres Temp Yelocity IIcat Flux Index Flowin Factor flow in Factor !Ntaile !
(psial ( F) (1) (2) ( ASI) Loc 5 l } Loc 5 [ l TORC CETOP
_ i 1750 4$0 1.7432 242404 .517 l 21(N) 450 1.7432 257(X)N .517 l
2250 450 1.7432 261195 .517 '
24txl 450 1.7432 264!!8 .517 21(K) 545 2.1740 216444 .517 1750 NN) 1.7432 1442N3 .206 21(X) N K) 1.7432 16X727 .206
)
2250 NX1 1.7432 176398 .2(ki 24(N) N N) 1.7432 184260 .206 21(K) 545 2.1740 257118 .206 21(N) 54 5 2.1790 282778 .004 21(X) 545 2.1790 248644 .203 1750 450 1.7132 245262 .527 1750 545 1.7432 227063 .527 1750 NN) 1.7432 147319 .527 2l00 545 2.1740 25N)l4 .527 (i) (1(f Ib ,,/hr-fg2) i (2) tilTU hr-ft 2)
OPPD-NA-8301-NP. Rev. 04 Page 14 of 25
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l DELETED CE Omaha Public Power District Figure FUEL SPACER GRID Fort Calhoun Station Unit No.1 41 OPPD-NA-8301-NP, Rev. 04 Page 15 of 25
1 del.ETED 1
ANF Omaha Public Power District Figure FUEL SPACER GRID Fort Calhoun Station Unit No.1 42 OPPD-NA-8301-NP, Rw 04 Page 16 of 25
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l SE W .J11 134.937 134.75g 134 g3 16.25 16.089 16.15 118.687 116.67 .. 118.800 16.813 16.81 16.81 101.875 tot.800 igj gg$
16.813 16.81 16.81 65.063 85.0*C .. 85.182 16.813 16.81 16.81 68.25 68.240 60.369 16.813 16.81 16.81 51.437 51.43 51.S$6 10813 16.81 16.61 34.625 34 62 34.743 16.813 10.81 16.81 17.812 17.81 17.93 Assemory Lower Tie Plate AXIAL LOCATION OF FUEL Omaha Public Power Distnet Figure ASSEMBLY SPACER GRIDS Fort Calhoun Station Unit No.1 4-3 OPPD-NA-8301-NP, Rev. 04 Page 17 of 25 1
CL CHANNEL NUM$tR A!![M!LY AVERAGE M 01At. 01 02
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.6963 .4648 03 04 05 06 01
.4264 .9976 1.1990 .9554 1.!!!8 I OS of to 11 12 13
.42C3 1.1311 1.0$25 1.0985 1.2621 1.0793 14 16 15 17 18 19
.9964 1.0536 1.2207 1.0669 1.2697 1.0845 l
20 21 22 13 24 25 g 1.1979 1.1009 1.C632 1.2024
.9162 1.2342
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STAGE 1 TORC CHANNEL Omaha Public Power District Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station Unit No.1 44 OPPD-N A-8301-NP. Rev. 04 Page 18 of 25
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12 i a 3 s is 14 15 STAGE 2 TORC CHANNEL Omaha Public Power District Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station Unit No.1 45 OPPD-NA-8301-NP, Rev. 04 Pa9:19 of 25 J
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STAGE 3 TORC CHANNEL Omaha Public Power District Figure GEOMETRY FOR FCS UNIT NO.1 Fort Calhoun Station Unit No.1 46 I
OPPD-NA-8301-NP, Rev, 04 Page 20 of 25
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AXIAL POWER DISTRIBUTIONS Omaha Public Power District Figure FORT CALHOUN CYCLE 8 Fort Calhoun Station Unit No.1 47 OPPD-N A-8301-NP, Rev. 04 Page 21 of 25
VALUt OtNott! THE RAf to Cr THE EU'10Lt01 02 INLif YtLC:liY PA55 FA55 vtLC:liY to THE CCRC AVERAGE 0.88 0.94 03 04 05 06 0/
0.90 0.94 0.94 0.98 0.99 08 09 10
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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 to 21 22 13 24 25 1.02 1.02 1.07 1.04 0.97 1.05 o,91 27 28 29 30 31 32 33 1.05 1.00 1.05 1.05 1.04 1.06 0*I' 34 35 16 )? 38 39 1.05 1.01 1.09 1.08 1.01 1.10 INLET FLOW DISTRIBUTION Omaha Public Power Distnct Figure FOR FCS,4 PUMP OPERATION Fort Calhoun Station Unit No.1 48 OPPD-NA-8301-NP, Rev 04 Page 22 of 25
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1 VALUt OENOTt! Tat etVIAf!C.'l FrsCM THE AVERACE CCRE [11T FRtsst;tt Mi TSF 03 04 05 06 of 27.60 16.99 27.60 44.59 $7.32 08 09 10 11 12 13 II* 2 '3.18 1.43 11.7 3.185 6.37 14 15 16 17 ~ 18 19 10.61 23.4 .t?.6 10.6 12.7 7.43 20 21 22 23 24 25 25 4.246 24,4
.tl.8 25.8 10.6 14.8 21 28 19.11 29 30 31 g 33 I 431 23.4 39.3 22.3 1.43 4.25 36.C9 34 15 36 ); 33 33 3I'84 21.2 .t0.7 20.2 1.43 8.49 EXIT PRESSURE DISTRIBUTION Omaha Public Power District Figure FOR FCS,4. PUMP OPERATION Fort Calhoun Station. Unit No.1 49 1
OPPD+N A-8301-NP, Rev. 04 Page 23 of 25 '
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CETOP D CHANNEL GEOMETRY Omaha Public Power District Figure (CHANNEL 1 NOT SHOWN) Fort Calhoun Station Unit No.- 1 4 10 OPPD-NA-8301-NP, Rev. 04 Page 24 of 25
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CORE AVERAct A51 (T)
ALGORITHM FOR DNB Omaha Public Power District Figure LCO MONITORING Fort Calhoun Station :
Unit No.1 61 OPPD-NA-8301-NP, Rev. 04 Page 25 of 25
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t' Table 1 RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301-NP REV 04 Title Page Changed the revision number and dato.
All Pages Updated rovision number.
All Pages Change all references of Combustion Engincoring to ADB Combustion Engincoring.
All Pages Replaced referonces to ANF with references to Westinghouse (also donoted by W) where appropriato.
II Added Westinghouse Electric Corporation to Proprietary Data Clause.
ill Updated Tabio of Contents to includo Chapter 7 on Core Operating Limits Report lv Added reference to W (Westinghouso) to Tablo 4-1 on List of Tables v Deleted Figures 4-1 and 4-2 from list ,
v1 Updated Revision Shoot.
1 Added reforenco to Wostinghouse Electric Corporation as current fuel vendor. Added words *potentially" and "U.S,"
Changed paragraph 3 to oiscuss curront PTS flux reduction efforts.
3 Changed reference to Figures 4-1 and 4-2 to References 2-2 and 2-1, respect!voly. Replaced ANF with W grid spacer loss coefficients and used test data for CE fuel.
4 Added historical detall on core loading during Cyclo 8.
5 Added sentonce regarding use of TORC with Westinghouse fuel assemblies.
7 Deleted "and excess load". Changed "ovents provido" to " event providos" 9.10 Added Section 7.0 for Coro Operating Limits Report, w
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Tablo 1 (Continued)
RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301-NP REV 04 )
I 11 Added references 2-1 and 2-3 to Section 2 references. Deleted references for ANF fuel from Section 2 roferenco list.
Renumborod references, Added - A to NRC approved I topleal reports S-1 and 3-6.
12 Updated reference 6-1 to the latest revision. Added refstences for Soctinn 7. Added -A to NRC approved topical reports for referenco 6-3.
13 Addod parameter ranges for Westingnouse fuel.
17 Added dimensions for Westinghouse fuel grids, normallzod distanco to contor of the grid.