ML20212N021

Rev 2 to Reload Core Analysis Methodology Overview

ML20212N021
Person / Time
Site:	Fort Calhoun
Issue date:	11/30/1986
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19292G906	List: ML20212N015 ML20212N021 ML20212N027 ML20212N032
References
OPPD-NA-8301-NP-R02, OPPD-NA-8301-NP-R2, NUDOCS 8703120315
Download: ML20212N021 (34)

Text

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i OMAHA PUBLIC POWER DISTRICT ,

NUCLEAR ANALYSIS RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW t

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O ABSTRACT This document is a Topical Report describing.0maha 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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OPPD-NA-8301-NP Rev. 02 i ..c c-risucec, . c.. cqcdSus]k

O PROPRIETARY DATA CLAUSE

, This document is the property of Omaha Public Power District (OPPD) and contains proprietary information, indicated by brackets, developed by Combustion Engi-neering (CE) and Exxon Nuclear Company, Inc. (ENC). The CE and ENC information was purchased by OPPD under proprietary information agreements.

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TABLE OF CONTENTS SECTION PAGE 1

1.0 INTRODUCTION

2.0 FUEL SYSTEM DESIGN I 1

3.0 NUCLEAR DESIGN 3

4.0 THERMAL HYDRAULIC DESIGN 11 5.0 POSTULATED ACCIDENTS AND TRANSIENTS 11 j 6.0 SETPOINT GENERATION 16

7.0 REFERENCES

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RELOAD CORE METHODOLOGY OVERVIEW REVISION DATE 4

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1-i OMAHA PUBLIC POWER DISTRICT RELOAD CORE METHODOLOGY OVERVIEW f

1.0 INTRODUCTION

Analyses done to license reload cores for Fort Calhoun Station consists 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  ;

Combustion Engineering (CE); however, future reload fuel may be suppifed by any of the four PWR nuclear fuel vendors: Exxon Nuclear Company, Inc.

(ENC), Combustion Engineering (CE), Westinghouse, or Babcock and Wilcox.

-The following sections discuss the reload analyses and consolidate infor-mation 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 Combustion Engineering are described in Reference 1.a. Exxon Nuclear Company, Inc. fuel mechanical design and I design methods are discussed in References I b . and 1.c.

In an effort to further reduce the flux to the reactor vessel welds, part length poison rods may be incorporated into the fuel loading pattern. The poison rods are composed of 84 C over the middle 50% of the length. Inert I material comprises the balance of the rod. When used they will reside in the guide tubes of the selected assemblies. For example, in Cycle 10 the poison rods were in the four outside guide tubes of the assembifes in quarter core locations I, 2, 3, and 8.

3.0 NUCLEAR DESIGN The District's nuclear design methodology is discussed in Reference 2.

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3.0 NUCLEAR DESIGN (Continued) 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.

i 3.2 Power Distribution Measurement The District utilizes the CE methodology (Reference 3) to measure the power distributions. This methodology is discussed in the Cycles 5 and 6 reload submittals:end approved in the SER's for these fuel cycles (References 4 andc5).

3.3 Uncertainties and Allowances The power distribution uncertainties which are included in the overall analysis of reload cores are:

Parameter Uncertainty 3D Peak, F 3-D 6.2%

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Integrated Radial Peak, FR 6.0%

Planar Radial Peak, F 5.3%

xy These values are approved for use in CENPD-153-P (Reference 6). A more detailed discussion c.f the treatment of uncertainties and allowances can be found in Reference 7.

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'3.0 NUCLEAR DESIGN (Continued)

. 3.4 Physics Safety Related Data The physics safety _related data are produced using the methodology

. discussed in Reference 2.

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 methodology (References 8, 9, 10, 11 and 12). This methodology was approved for use by the District in Reference 13.

4.1.1 Grid Spacer Loss Coefficients The analysis utilizes a D-TORC model with explicit repre-sentation of the loss coefficients associated with ENC and CE fuel assemblies. The nominal grid loss coefficients used in thermal hydraulic analysis are:

ENC Spacer CE Spacer l Loss Coefficient (K) ~

Re I eynolds R Number l

These values were obtained by ENC using single phase pressure drop testing of an ENC test assembly and a typical CE 14 x 14

- assembly. Single phase hydraulic loss coefficients, previously O

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'4.0 THERMAL HYDRAULIC DESIGN (Continued) 4.1 Steady State DNBR Analysis (Continued) 4.1.1 Grid Spacer Loss Coefficients (Continued) transmitted in Reference 14, contained a [

] while the values given above are best-estimate values. Because of the sensitivity of DNBR calcu-lations to the difference in spacer grid loss coefficients, the District utilizes the Reynold's number expression for loss coefficients. This provides the most accurate repre-sentation of the pressure drop across each 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 CE and EllC spacer grids are shown in the attached Figures 1 and 2. The spacer grid envelope for both the CE and EflC grids is 8.115 inches; by 8.115 inches. The axial location of the CE and ENC spacer grids is shown in Figure 3.

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 chan-nels in D-TORC is made such that the minimum flow is pro-vided 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 all fuel assemblies.

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.\ j 4.0 THERMAL HYDRAULIC DESIGN (Continued) 4.1 Steady State DNBR Analysis (Continued) 4.1.2 CE-1 Corralation The District utilizes the CE-1 correlation for DNBR calcu-lations. The range of data in the data base for the CE-1 correlation is contained in References 8 and 9. The range of parameters for the CE-1 correlation and corresponding ranges for the'CE and ENC assemblies are shown in Table 1.

Because the data for the ENC fuel assembly is within those specified in CE-1 data base, the use of CE-1 correlation is appropriate for the ENC fuel.

A 4.1.3 D-TORC and CETOP Models b

The District utilizes the D-TORC code (Reference 10) and the CETOP code (Reference 11) to perform thermal hydraulic analysis f.or the Fort Calhoun reload core. The fraction of inlet flow to the hot assembly in the CETOP model is adjust-edsuchthatfhemodelyieldsappropriateMDNBRresultswhen compared with results of D-TORC analysis for a given range of operating conditions. The fraction of inlet flow is determined for each reload core. The use of this nethodo-logy was. approved for use by the Di trict in Reference 13.

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. TABLE 1 4 PARAMETER RANGES OF THE SOURCE DATA FOR THE CE-1 CHF CORRELATION AND THE RANGE OF EXXON AND CE 14 x 14 FOR FORT CALHOUN VALUES

. CORRELATION CE EXXON PARAMETER RANGE RANGE RANGE Pressure (psia) 1785 to 2415 N/A N/A Local Coolant Quality .16 to .20 N/A N/A Local Mass Velocity (lbm/hr-ft2) 0.87x106 to 3.21x106 N/A N/A Subchannel Wetted Equiv. .3588 to .4043 to .4010 to Diameter (in) .5447 .5449 .5402 Subchannel Heated Equiv. .4713 to .5334 to .5270 to Diameter (in) .7837 .7840 .7760 Heated Length (in) 84,150 128 128

! Grid Spacing 14.2" to 18.25" 16.8 16.8 I

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4.0 THERMALHYDRAULICDESIGN(Continued) 4 4.1 -Steady State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued)

The following paragraphs discuss the application of the CETOP code to the Fort Calhoun reactor. Examples are for I

the Cycle 8 core.

Thermal margin ' analysis utilizing the CETOP model is sup-ported by comparing its predictions for Fort Calhoun Station witii 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.

, A thermal margin model for 1500 MWT for Fort Calhoun Unit No. I was developed for the following operating ranges.

- Inlet Temperature 450 to 600 F i

! - System Pressure 1750 to 2400 psia

- Primary System 4-Pump I

Flow Rate, (LC0 = 197,000 gpm) 80% to 120%

i - Axial Power Distribution -0:517 to +0.526 ASI

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%Y 4.0 THERMAL HYDRAULIC DESIGN (Continued) 4.1 Steady State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued)

The detailed thermal margin analyses were performed for the -

sample core using the radial power distribution and detailed TORC model shown in Figures 4, 5, and 6. The appropriate spacer grid loss coefficient was applied to each " assembly" channel or' partial assembly channel in each stage. In stage 1, lumped channel 28 utilized the CE spacer grid loss coefficient because the channel was predominantly composed of CE fuel. Lumped channels 26 and 27 utilized the ENC spacer grid loss coefficient because either the channel was composed of entirely ENC fuel or contained a single CE assembly not on a boundary between channels. The axial power distributions are given in Figure 7. These distri-butions 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 c, the flow model test results given in Figures 8 and 9. The results of the detailed TORC analyses are given in Table 2.

The CETOP design model was a total of four thermal hydraulic channels to model the open-core fluid phenomena. Figure 10 shows the layout of these channels. Channel 2 is a quadrant C OPPD-NA-8301-NP Rev. 02 1 Page 8 of 28  ;

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TABLE 2 COMPARISONS BETWEEN TORC AND CETOP-D 2

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Operating Parameters MDNBR Quality at MDNBR of MONBR (in) l Detailed CETOP-D Detailed CETOP-D TORC Inlet TORC Inlet Inlet Avg Mass Core Avg. Shape Relative Flow Relative Flow Pr:ssure Temperature Velocity Heat Flux Index Flow in Factor Flow in Factor Detailed (psia) *F 6 (BTU /hr-ft2) (ASI) Location 5 [ ] Location 5 [ _ ] TORC CETOPl (10 lbm/hr-ftr) 1750 450 1.7432 242409 .517 1.177 2100 450 1.7432 257008 .517 1.166

'2250 450 1.7432 261195 .517 1.168 2400 450 1.7432 264118 .517 1.169 ,"

2100 545 2.1790 216494 .517 1.200 1750 600 1.7432 149283 206 1.245 2100 600 1./432 168727 .206 1.240

2250 600 1.7432 176398 .206 1.240 i 2400 600 1.7432 184260 .206 1.224 i

' 2100 545 2.1790 257118 .206 1.236

' 2100 545 2.1790 282778 .004 1.284 2100 545 2.1790 298644 .203 1.310

!1750 450 1.7432 295262 .527 1.276 i

1750 545 1.7432 227063 .527 1.261 1750 600 1.7432 147319 .527 1.230 l

, 2100 545 2.1790 255014 .527 1.294 1

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4.0 THERMAL HYDRAULIC DESIGN (Continued) 4.1 Steady State DNBR Analysis (Continued) 4.1.3 D-TORC and CETOP Models (Continued) of the hottest assembly which represents 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 b'oundaries 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 from p D-TORC analysis was an ENC assembly. Since CETOP models a A quadrant-of the " hot" assembly, the ENC 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 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 Cycle 8 core.

The uncertainties associated with the thermal hydraulic 4 analysis are combined statistically (Reference 15). In this method, the impact of component uncertainties on DNBR is assessed and the SAFDL is increased to include the effects of the uncertainties.

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5.0 POSTULATED ACCIDENTS AND TRANSIENTS The postulated accidents and transients are analyzed using the methodology.

discussed in Reference 16.

6.0 SETPOINT GENERATION The District utilizes the methodology discussed in CENPD-199-P (Reference

17) to generate setpoints for, Fort Calhoun Station. The District's reactor physics methodology is discussed in Reference 2.

The scram reactivity curves are produced using the QUIX code. The power-to-fuel design limit on centerline melt is derived using the QUIX code with the appropriate combinations of planar radial peaking factor, Ffy, and O axial power distribution.

The thermal margin analysis is done using the CETOP code with the appro-priate combinations of the integrated radial peaking factor, Ff, axial power distribution, RCS inlet temperature, and RCS pressure.

The Fort Calhoun RPS utilizes the " standard" local power density trip and TM/LP trip. The sequential CEA withdrawal is analyzed using the methods described in Reference 17 and not included in the TM/LP trip considera-tions. The RCS depressurization and excess load events provide the tran-sient analysis input into the TM/LP trip.

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tJ 6.0 SETPOINT GENERATION (Continued) 6.1 Incore LCO Monitoring The Better Axial Shape Selection System (BASSS) monitors the Limiting Conditions for Operation on peak linear heat rate and departure from nucleate boiling 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 Cl-iffs Units, and described in the Combustion Engineering Setpoint Topical (Ref. 17).

MINI-CECOR is a mini-computer version of Combustion Engineering's CECOR code. It is used in this application to synthesize the follow-ing parameters from readings of the fixed in-core detectors:

1. The three-dimensional power peaking factor (Fq )

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2. The core average axial shape index (T)

3. The total planar radial peaking factor (Ffy)

4. Thetotalintegratedradialpeakingfactor(Ff)

These inputs to BASSS are descriptive of the existing core power distribution.

The inputs to BASSS obtained from the plant computer are the following:

1. Measured core power level

2. Percent insertion of the lead CEA regulating group.

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,I 6.0 )iETPOINTGENERATION(Continued) 6.1 Incore LCO Monitoring (Continued)

BASSS consists of two algorithms: one for peak linear heat rate moni-s toring and another for DNB monitoring. The peak linear heat rate algoritnm uses the 3-D power peaking factor and the measured core power level to calculate the core peek linear heat rate. The algo-rithm ' applies appropriate. uncertainties and allowances (per the Technical Specifications) to the 3-D peaking factor. The measured ,

l peak linear heat rate is compared to the monitoring limit, which is j l

based on both LOCA and A00 transient analysis considerations, and an alarm is activated when the monitoring limit is exceeded. The power p operating limit on linear heat rate is also calculated and displayed as an iridication 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 ASI uncertainty due to the use of in-core ASI versus ex-core ASI.

2. Knowledge of the actual CE> t ug 4sition versus the ex-core system's assumption that the CEAs are inserted to the PDIL'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.

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2 LO 6.0.. SETPOINT GENERATION (Continued) 6.1 Incore LC0 Monitoring (Continued)

The algorithm for DNB LCO monitoring is shown in Figure 11.

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6.0 SETPOINT GENERATION (Continued)

6. I' Incore LCO Monitoring (Continued) 4 BASSS is also provided with the capability to monitor the Limiting andFf.IftheTechnicalSpecifi-Conditions cation for Operationfor Operation on F xy or F on Ffy f are exceeded '

operation, BASSS will activate an alann'and calculate the proper tradeoff with maximum allowed power that ensures that the Axial Power Distribution and Thermal Margin / Low Pressure Trips remain conserva- ,

tive. An alarm is activated if the measured power level is higher

. than the allowed power level.

6.2 Uncertainties The uncertainties are treated statistically in the District's setpoint

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7.0 REFERENCES

1.a. "0maha Batch M Reload Fuel Design Report", CEN-347(0)-P, Rev. 01, January 1987.

1.b. " Generic Mechanical Design Report for Exxon-Nuclear Fort Calhoun 14 x 14 Reload Fuel Assembly", XN-NF-79-70-P, September 1979.

1.c. "Ft. Calhoun Design Report Extended Burnup Analysis", XF-NF-82-61, October 1982.

2. " Reload Core Analysis Methodology, Neutronics Design Methods and Veri-fication", OPPD-NA-8302, Rev.01, November 1986.

( 3. " INCA, Method of Analyzing Incore Detector Data in Pressurized Water Reactors", CENPD-145-P, April 1, 1975.

4. Letter, R.W. Reid (NRC) to T.E.Short (OPPD), December 5, 1978.

5. Letter, R.W. Reid (NRC) to W.C. Jones (0 PPD), April 1,1980.

6. " INCA /CECOR Power Peak Uncertainty", CENPD-153-P, Revision 1-P-A, May 1980.

7. " Statistical Combination of Uncertainties", CEN-257(0)-P, Parts 1 and 3, November 1983.

8. "CE Critical Heat Flux", CENPD-162-P-A, Part 1, Combustion Engineer-ing, September 1976.

9. "CE Critical Heat Flux", Part 2, CENPD-207-P-A, Combustion Engineer-

frs l (- ing, June 1976.

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7.0 REFERENCES

(Continued)

10. "A Computer Code for Determining the Thermal Margin of a Reactor Core," CENPD-161-P, TORC Cod,e, Combustion Engineering, July 1975.

11. "CETOP-D Code Structure and Mcdeling Methods for Calvert Cliffs Units 1 & 2," CEN-191(B)-P, Combustion Engineering, December 1981,

12. Letter from Cecil Thomas .(NRC) to Mr. A, E. Scherer (CE), November 2, 1984.

13. Letter from R. A. Clark (NRC) to W. C. Jones (0 PPD), March 15, 1983.

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14. Letter from W. C. Jones (OPPD) to R. W. Reid (NRC), December 4,'1979.

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15. . " Statistical Combination of Uncertainties," CEN-257(0)-P, Part 2, L

November 1983.

16. " Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," 0 PPD-NA-8303, Rev.01, November 1986.

17. "CE Setpoint Methodology," CENPD-199-P, Revision 1-P, April 1982.

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