ML20070P372

From kanterella
Jump to navigation Jump to search
Reload Core Analysis Methodology Overview
ML20070P372
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 05/31/1994
From:
OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19304C106 List:
References
OPPD-NA-8301-NP-R06, OPPD-NA-8301-NP-R6, NUDOCS 9405110270
Download: ML20070P372 (35)
v • d • e


Text

- - - - - - - - - - - - - - - -

L

l l

l LIC-94-0092 j l

I Enclosure 4 i

l t

l l

l i

i 9405110270 940502 PDR ADOCK 050002B5 P- PDR

SUMMARY

OF CHANGES TO RELOAD CORE. ANALYSIS METHODOLOGY OVERVIEW OPPD-NA-8301-NP Rev. 06

1. Title Page Change the revision number and date.
2. All Pages Update the revision number.
3. Page iii Update Table of Contents with new page numbers.
4. Pages iv-v Update List of Tables and List of Figures with new page numbers.
5. Page vi Update Figure 4-3 to include only Westinghouse fuel assembly spacer grids,
6. Page vii Update revision sheet.
7. Page 1 - Replaced "done" with " performed."

- Replaced "the District" with "0 PPD."

8. Page 2 - Inserted discussion of ABB/CE fuel removal from core.

- Inserted "CEA" between " outer" and " guide."

9. Page 3 - Replaced "the District" with "0 PPD" where appropriate.

- Inserted " neutron" between " reduce" and " flux."

10. Page 4 - Replaced "the District" with "0 PPD" where appropriate.

- Replaced "D-TORC" with " TORC" where appropriate.

- Replaced "CE" with "ABB/CE" where appropriate.

11. Page 5 - Replaced "the District" with "0 PPD" where appropriate.

- Replaced "CE" with "ABB/CE" where appropriate.

- Inserted " nominal" between "Several" and " operating."

12. Page 6 - Added quotes on " assembly" the second time.

- Reworded the last sentence in paragraph 2.

- Updated wording to show changes to TORC and CETOP dpply to all subsequent Cycles.

- Deleted reference to mixed vendor core.

13. Page 9 - Replaced "the District" with "0 PPD" where appropriate.

- Changed all references of HERMITE to ASAS.

- Added reference to YAEC setpoint methods.

- Added References 6-3 and 6-4.

14. Page 11 - Added "LC0" acronym.

- Replaced "the District" with "0 PPD" where appropriate.

- Changed Reference 6-3 to 6-5.

15. Page 13 Updated Reference 3-1 to Revision 04, May 1994.
16. Page 14 - Updated Reference 6-2 to Revision 04, May 1994.

- Added References 6-3 and 6-4.

- Changed Reference 6-3 to 6-5.

- Updated Reference 7-2 to Revision 06, May 1994.

- Updated Reference 7-3 to Revision 04, May 1994.

17. Page 15 Deleted SPC data.
18. Page 19 Updated Figure 4-3 to make it more understandable.

I l

l l

l This document is not to be transmitted or )

reproduced without specific written approval from Omaha Public Power District. ,

I Omaha Public Power District Nuclear Analysis BfLOAD CORE ANALYSIS METHODOLQGY OVERVIEW OPPD-NA-8301-NP l Rev. 06 May 1994 Copy No.

ABSTRACI This document is a Topical Report describing 0maha Public Power District's reload core analysis methodology for application to Fort Calhoun Station Unit No. 1. l The report provides an overview of OPPD's reload core methodology. Analyses l performed by OPPD and its contractors are described. Details of the thermal hydraulic methodology which were previously submitted to the NRC are provided. ,

i l

l l

l

.i OPPD-NA-8301-NP, Rev. 06 i

TABLE OF CONTENTS Section Pace 1.0 INTRODUCTIDH ..................................................... 1 l 1

2.0 FU E L S Y ST EM D E S I G N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2  ;

1 I

3.0 NUCLEAR DESIGN ................................................... 3 ,

3.1 Fu e l Ma n a g eme n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.2 Power Di st ri buti on Mea s u remen t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.3 Uncertainties and Allowances ............................... 3 3.4 Physics Safety Related Data .............. ................. 3 4.0 IHE RM A L H YD R AU L I C D ESlHN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.1 St eady S tate DNBR Anal ysi s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.1.1 Grid Spacer Loss Coefficients ...................... 4 4.1.2 C E- 1 C o r r e l a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.1.3 D-TORC an d C ETOP Mod el s . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 5.0 POSTUL AT ED ACCIDENIS AND TRANSI ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 6.0 S E T PO I NT GE N E R A T I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 6.1 I n co re LCO Mon i t o ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 6.2 Uncertainties .............................................. 11 7.0 LORE OffRATING LIMITS REPORT ..................................... 12

8.0 REFERENCES

....................................................... 13 OPPD-NA-8301-NP, Rev. 06 ii

LIST OF TABLES Tabla Title Page 4-1 Parameter Ranges of the Source Data for the CE-1 CHF CorrelationandtheRangeofWestinghouseandABB/CE14X14 Fuel for Fort Calhoun Values .................................. 15 4-2 Comparisons Between TORC and CETOP-D .......................... 16 l

1 I

l l

1 l

l l

l l

1 l

l l

OPPD-NA-8301-NP, Rev. 06 iii

1 I

LIST OF FIGU_Bil Ficure li.tle P_ age 4-1 ABB/CE Fuel Spacer Gri d (DELETED) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 i l l i 4-2 SPC Fuel Spacer Gri d (DELETED) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 )

4-3 Axial Location of Fuel Assembly Spacer Grids . . . . . . . . . . . . . . . . . . 19 4-4 Stage 1 TORC Channel Geometry for Fort Calhoun Unit No. 1 ..... 20 l 4-5 STAGE 2 TORC Channel Geometry for Fort Calhoun Unit No.1. . . . . 21 l

l l

4-6 Stage 3 TORC Channel Geometry for Fort Calhoun Unit No.1.. . . 22 4-7 Axial Power Distributions for Fort Calhoun Cycle 8 . . . . . . . . . . . . 2J 4-8 Inlet Flow Distribution for Fort Calhoun 4-Pump Operation .... 24 4-9 Exit Pressure Distribution for Fort Calhoun 4-Pump Operation .. 25 4-10 CETOP-D Channel Geometry for Fort Calhoun Unit No.1. . . . . . . . . 26 6-1 Al gori thm For DNB LC0 Moni tori ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27  ;

l 1 OPPD-NA-8301-NP, Rev. 06 1

iv

OMAHA PUBLIC POWER DISTRICT RELOAD CORE METHODOLOGY OVERVIEW Revision Dslta 00 September 1983 01 June 1985 02 November 1986 03 April 1988 04 April 1991 05 January 1993 06 May 1994 l l

DPPD-NA-8301-NP, Rev. 06 v

OMAHA PUBLIC POWER DISTRICT RELOAD CORE ANALYSIS METHODOLOGY OVERVIEW 1.0 INTRODRCTION Analyses performed to license reload cores for fort Calhoun Station consist of the analysis performed by the Omaha Public Power District (OPPD) 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), Westinghouse, or Babcock and Wilcox. The following sections discuss the reload analyses and consolidate information about the OPPD methodology previously submitted. E OPPD-NA-8301-NP, Rev. 06 Page 1 of 27

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. TheABB/CEfuel (Reference 2-2) is no longer co-resident with the W fuel since the entire core is comprised of W fabricated fuel assemblies. However,theABB/CEvendordata will be retained for historical purposes.

In an effort to further reduce the neutron flux to the reactor vessel welds, full length Hafnium flux suppression rods, which are similar to the part length poison rods utilized in Cycle 10, will be incorporated into the fuel loading 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. These rods will reside in the outer CEA guide tubes of quarter core g 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 critical welds.

I I

i 1

l l

l l

l OPPD-NA-8301-NP, Rev. 06 Page 2 of 27

l l

l l

3.0 EQCLEAR_ DESIGN l l

OPPD's nuclear design methodology is discussed in Reference 3-1. l 3.1 EMEL MANAGEMEtlI i 1

l The reload core fuel management is performed by OPPD. Current fuel management schemes are selected to reduce neutron flux to the reactor pressure vessel welds.

3.2 POWER DLSTRIBUTLQ_N MEASUREMENT OPPD utilizes the ABB/CE methodology (Reference 3-2) to measure the power g l

distributions. This methodology is discussed in the Cycles 5 and 6 reload submittals and approved in the SERs for these fuel cycles (References 3-3 and 3-4).

3.3 UNLERTAINTIES AND ALLQWAUCES l

The power distribution uncertainties which are included in the overall analysis of reload cores are:

1 Parameter Untertainty 30 Peak, Fg 3-D 6.2% l Integrated Radial Peak, Fa 6.0%

Planar Radial Peak, Fxy 5.3%

These values are approved for use in CENPD-153-P (Reference 3-2). A more detailed discussion of the treatment of uncertainties and allowances can be found in Reference 3-5.

3.4 PHYSICS SAFETY RELATED DATA The physics safety related data are produced using the methodology discussed in Reference 3-1.

OPPD-NA-8301-NP, Rev. 06 Page 3 of 27

4.0 J11ERMAL H1DRAULIC DESISH 4.1 STEADY STATE DM R ANALYSIS ThesteadystateDNBRanalysisisperformedbyOPPDusingtheTORC/CETOP/CE-1 g.

methodology (References 4-1, 4-2, 4-3, 4-4 and 4-5). This methodology was approved for use by 0 PPD in Reference 4-6. l 4.1.1 GRID SPACER LOSS COEFFICIENTS The analysis utilizes a TORC model with explicit representation of the g loss coefficients associated with W fuel assemblies. The nominal grid loss coefficients used in thermal hydraulic analysis are:

W Spacer ABB/CE Spacer

[ ] Loss Coefficient (K)

Re = Reynolds Number These values were obtained by W using single phase pressure drop testing ofaWassemblyandatypicalABB/CE14x14 assembly. Single phase g hydraulic loss coefficients, previously transmitted in Reference 4-7, contained a [ ] while the values given above are best estimate values. Because of the sensitivity of DNBR calculations to the dif ference in spacer grid loss coefficients, OPPD utilizes the Reynolds Number expression for loss coefficients.

This provides the most accurate representation of the pressure drop l

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 W and ABB/CE spacer grids are shown in References 2-1 and 2-2 respectively. The spacer grid envelope for both the W and ABB/CE grids is 8.115 inches by 8.115 inches. The axial location of the W and ABB/CE spacer grids is shown in Figure 4-3.

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. De choice of loss coefficient

! for lumped channels in D-TGRC is made such that the minimum flow is OPPD-NA-8301-NP, Rev. 06 Page 4 of 27

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 all fuel assemblies.

4.1.2 CE-1 CORRELATION OPPD utilizes the CE-1 correlation for DNBR calculations. The range of g data in the data base for the CE-1 correlation is contained in References 4-1 and 4-2. The range of parameters for the CE-1 correlationandcorrespondingrangesfortheABB/CEandWassembliesareg 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 OPPD utilizes the D-TORC code (Reference 4-3) and the CETOP code g (Reference 4-4) to perform thermal hydraulic analysis for the Fort Calhoun Station reload core. The fraction of inlet flow to the hot assembly in the CETOP model is adjusted such that the model yields appropriate MDNBR results when 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 methodology was approved for use by OPPD in Reference 4-6. l The following paragraphs discuss the application of the CETOP code to the Fort Calhoun Station reactor. Examples are for the Cycle 8 core (which contained SPC and ABB/CE fuel types).

Thermal margin analysis utilizing the CETOP model is supported by comparing its predictions for Fort Calhoun Station with those obtained from a detailed TORC analysis. Several nominal operating conditions were selected for this demonstration; they are representative, but not l

the complete set, of conditions which would be considered for a normal DNB analysis.

1 OPPD-NA-8301-NP, Rev. 06 Page 5 of 27

A thermal margin model at 1500 MWt for Fort Calhoun Station was developed for the following operating ranges:

Inlet Temperature 450 F to 600 F  ;

System Pressure 1750 psia to 2400 psia I Primary System 4-Pump Flow Rate 80% to 120%

(LC0 = 196,000 gpm)

Axial Power Distribution -0.517 to +0.526 ASI The detailed thermal margin analyses were performed for the sample core using the radial power distribution and detailed TORC roodel 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 g each stage. In stage 1, lumped channel 28 utilized the ABB/CE spacer grid loss coefficient because the channel was predeminantly composed of ABB/CE fuel. Lumped channels 26 and 27 utilized the SPC spacer grid loss coefficient because the channel was either composed entirely of SPC g 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 TORC and CETOP models for Cycle 14 and subsequent cycles incorporate the appropriate changes for W fuel design parameters.

The CETOP design model was a total of four thermal hydraulic 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 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. )

l Channel 3 lumps the subchannels adjacent to the MDNBR hot channel 4.

The " hot" assembly determined from D-TORC analysis was a SPC assembly.

Since CETOP models a quadrant of the " hot" assembly, the SPC spacer grid j loss coefficient was used in the analysis.

1 OPPD-NA-8301-NP, Rev. 06 Page 6 of 27 l

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 Cycle 8 core. The uncertainties associated with the thermal hydraulic analysis 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.

l l

l l

OPPD-NA-8301-NP, Rev. 06 Page 7 of 27

,-.4 . i .. __ , . -,...a., . . . . . - , , . , . ..

t:

5.0 POSTULATED ACCIDENTS AND TRANSIENTS l The postulated accidents and transients are analyzed using the methodology' discussed in Reference 5-1.

i i

I i

l l

q l

l

'l l

l i

OPPD-NA-8301-NP, Rev. 06 Page 8 of 27 l

i

l 6.0 SETPOINT GENERATION j OPPD utilizes the methodology discussed in CENPD-199-P (Reference 6-1) to generate setpoints for the Fort Calhoun Station. OPPD's reactor physics I methodology is discussed in Reference 6-2.

l The axial shape analysis is performed using the ASAS code which is described in Reference 6-3. The power-to-fuel design limit on centerline melt is l j derived using the ASAS code with the appropriate combinations of planar radial peaking factor (FXY) and axial power distribution as described in Reference 6-4. The power-to-fuel design limit on DNBR is derived using the ASAS code with the appropriate combinations of integrated radial peaking factor (F R) and i axial power distribution as described in Reference 6-4. l The thermal margin analysis is done using the CETOP code with the appropriate combinations of the integrated radial peaking factor (FR ), axial power distribution, RCS inlet temperature, and RCS pressure.

The FCS 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 trip considerations. The RCS depressurization event provides the transient analysis input into the TM/LP trip.

6.1 IE0_Rl_LCD30RITORING The Better Axial Shape Selection System (BASSS) monitors the Limiting Conditions for Operation (LCO) 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 the Calvert Cliffs Units, and described in the ABB/CE Setpoint Topical (Reference 6-1).

MINI-CECOR is a plant computer version of ABB/CE's CECOR code. It is used in this application to synthesize the following parameters from readings of the fixed in-core detectors:

1. The three-dimensional power peaking factor (FnT )
2. The core average axial shape index (T)
3. The total planar radial peaking factor (Fyy )

T

4. The total integrated radial peaking f actor ( Fa )

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

OPPD-NA-8301-NP, Rev. 06 Page 9 of 27

l l

l l

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

1. Measured core power level i
2. Percent -insertion of the lead CEA regulating group.

BASSS consists of two algorithms: one for peak linear heat rate monitoring and another for DNB monitoring. The peak linear heat rate algorithm uses the 3-D power peaking factor hnd 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 A00 transient analysis considerations, and an alarm is activated when the monitoring 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 ASI uncertainty due to the use of in-core ASI versus  ;

ex-core ASI. t

2. Knowledge of the actual CEA group position 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.

1 i

J i

i OPPD-NA-8301-NP, Rev. 06 Page 10 of 27

BASSS is also provided with the capability to monitor the Limiting Conditions for Operation (LCO) on FxyT and FnT . If the LCO for Fxyi or Fni is exceeded g 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 Distribution and Thermal Margin / Low Pressure Trips remain conservative. 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 OPPD setpoint analysis as identified in References 6-1, 6-2 and 6-5. I 1

l l

l l

I I

OPPD-NA-8301-NP, Rev. 06 Page 11 of 27

7.0 CDRE OPERATING LIBITS 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.

Tne reload analysis for Fort Calhoun Station for generating the COLR values utilizes 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.

The COLR consists of the following items:

Thermal Margin / Low Pressure for 4 Pump Operation Refueling Boron Concentration Limiting Conditions for Operation for Excore Monitoring of LHR e

Limiting Conditions for Operation for DNB Monitoring Power Dependent Insertion Limit (PDIL) e Unrodded Integrated Radial Peaking Factor (FR )

e Unrodded Planar Radial Peaking Factor (Fxy)

Allowable Peak Linear Heat Rate vs. Burnup

  • T Fxy , pig 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) .

I OPPD-NA-8301-NP, Rev. 06 Page 12 of 27

7.0 CORE OPERATIriG LIMITS 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 Station for generating the COLR values utilizes 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.

The COLR consists of the following items:

. Thermal Margin / Low Pressure for 4 Pump Operation e Refaeling Boron Concentration e Limiting Conditions for Operation for Excore Monitoring of LHR e Limiting Conditions for Operation for DNB Monitoring

  • Power Dependent Insertion Limit (PDIL)
  • Unrodded Integrated Radial Peaking Factor (FR) e Unrodded Planar Radial Peaking Factor (FXY)

. Allowable Peak Linear Heat Rate vs. Burnup e FxyT , pig and Core Power Limitations e 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) .

l OPPD-NA-8301-NP,.Rev. 06 Page 12 of 27 1

i

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.

Section 3 References 3-1 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification," OPPD-NA-8302-P, Rev. 04, May 1994. E 3-2 " INCA /CECOR Power Peak Uncertainty," CENPD-153-P, Revision 1-P-A, May 1980.

3-3 Letter from R. W. Reid (NRC) to . E. Short (OPPD), Decemt'er 5,1978.

3-4 Letter from R. W. Reid (NRC) to W. C. Jones (OPPD), April 1,1980. ,

I 3-5 " Statistical Combination of Uncertainties," CEN-257(0)-P-A, Parts 1 and 3, November 1983, including Supplement 1-P, August 1985. ,

1 1

Section 4 Refer _ances 4-1 "CE Critical Heat Flux," CENPD-162-P-A, Part 1, Combustion Engineering, J September 1976. I l

4-2 "CE Critical Heat Flux," Part 2, CENPD-207-P-A, Combustion Engineering, June 1976.

1 4-3 "A Computer Code for Determining the Thermal Margin of a Reactor Core,"  !

CENPD-161-P-A, TORC Code, Combustion Engineering, April 1986.

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 f rom Cecil Thomas (NRC) to Mr. A. E. Scherer (CE), dated November 2, 1984.

4-6 Letter from R. A. Clark (NRC) to W. C. Jones (OPPD), dated March 15, 1983.

OPPD-NA-8301-NP, Rev. 06 Page 13 of 27

4-7 Letter from W. C. Jones (OPPD) to R. W. Reid (NRC), dated 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, Revision 04, January 1993.

Section 6 References 6-1 "CE Setpoint Methodology," CENPD-199-P, Revision 1-P-A, January 1986.

6-2 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification," OPPD-NA-8302-P, Revision 04, May 1994.

6-3 Theoretical Manual and User's Manual for Axial Shape Analyzer for SIMULATE-3 (ASAS), April 19, 1994.

6-4 Maine Yankee SER on YAEC-Ill0, " Maine Yankee Reactor Protection System Setpoint Methodology," dated May 27, 1977.

6-5 " Statistical Combination of Uncertainties," CEN-257(0)-P-A, Parts 1, 2, and 3, November 1983, including Supplement 1-P, August 1985.

r Section 7 Refgrancn 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, Revision 06, May 1994. l 7-3 " Reload Core Analysis Methodology, Neutronics Design Methods and Verification," 0 PPD-NA-8302-P, Revision 04, May 1994. E 7-4 " Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," OPPD-NA-8303-P, Revision 04, January 1993.

OPPD-NA-8301-NP, Rev. 06 Page 14 of 27

i i

TABLE 4-1 PARAMETER RANGES OF THE SOURCE DATA FOR THE CE-1 CHFCORRELATIONANDTHERANGEOFWESTINGHOUSEANDABB/CE 14 x 14 FUEL FOR FORT CALHOUN VALUES CORRELATION ABB/CE W PARAMETER ILA_ NEE Rat 1E ILAllE Pressure (psia) [ ] N/A N/A Local Coolant Quality [ ] N/A N/A Local Mass Velocity [ ] N/A N/A (lbn /hr-ft 2)

Subchannel Wetted Equiv. 0.3588 0.4043 0.4043 Diameter (in.) to to to 0.5447 0.5449 0.5449 Subchannel Heated Equiv. 0.4713 0.5334 0.5334 Diameter (in.) to to to 0.7837 0.7840 0.7840 Heated Length (in.) 84 to 150 128 128 Grid Spacing (in.) 14.2 to 18.25 16.8 16.8 1

OPPD-NA-8301-NP, Rev. 06 Page 15 of 27 .

. _ . - . ~ . _ . . _ .. .. _- _ . . . _ _ . _ _ . . _ . __ _ __ . _

TABLE 4-2 COMPARISONS BETWEEN TORC AND CETOP-D Axial Elev.

_.__0neratina Parammers MDNSB Qual 11y_aLMDNBR gLED.hB_Riin 1 Core Det. CETOP Det. CETOP Avg. Avg. TORC Inlet TORC Inlet Inlet Mass Heat Shape Rel. Flow Rel. Flow Pres Temp Velocity Flux Index Flow in Factor Flow in Factor Detailed (psia) (*F) (1) (2) (ASI) Loc 5 [0.84, 0.76] Loc 5 [0.76] TORC CETOP

~ ~

1750 450 1.7432 242409 -0.517 2100 450 1.7432 257008 -0.517 2250- 450 .1.7432 261195 -0.517 2400 450 1.7432 264118 -0.517 2100 545 2.1790 216494 -0.517 1750 600 1.7432 149283 -0.206 2100 600 1.7432 168727 -0.206 2250 600 1.7432 176398 -0.206 2400 600 1.7432 184260 -0.206 2100 545 2.1790 257118 -0.206 2100 545 2.1790 282778 0.004 2100 545 2.1790 298644 0.203 1750. 450 1.7432 295262 0.527 1750 545 1.7432 227063 0.527 1750 600 1.7432 147319 0.527 2100 545 2.1790 255014 0.527 __ ._.

(1) (1061bm/hr-f t2) 2 (2) (BTU /hr-f t )  ;

I t

OPPD-NA-8301-NP, Rev. 06 Page 16 of 27

<rc - , --

. - . - _~ .. - ._. .. . . - . . . . . . .. - ..

DELETED I

.,1 l

l l

ABB/CE Omaha Public Power District Figure FUEL SPACER GRID Fort Calhoun Station Unit No. 1 4-1 OPPD-NA-8301-NP, Rev. 06 Page 17 of 27

.- . . - . . . ._ . = . . . - . - .

3 DELETED l

i 1-

~, SPC Omaha Public Power District. Figure FUEL SPACER GRID Fort Calhoun Station Unit No. 1 4-2 OPPD-NA-8301-NP, Rev. 06 Page 18 of 27

W FUEL ABB/CE FUEL o 134.759 o 134.937 16.089 16.15 ll 118.67 ll 118.687 16.81 16.813 ll 101.860 ll 101.875 16.81 16.813 L 85.050 ll 85.063 16.81 16.813 68.240 l 68.25 16.81 16.813 L 51.43 , 51.437 16.81 16.813 ll 34.62 < l 34.625 16.81 16.813 17.81 17.812 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. 06 I Page 19 of 27  :

. . . - _ = ._- . . -. - . - _- . ..

l' l

1 l

l l

l i

6 STAGE 1 TORC CHANNEL Omaha Public Power District Figure -

GE0 METRY FOR FCS UNIT NO. 1 Fort Calhoun Station Unit No. 1 4-4 OPPD-NA-8301-NP, Rev. 06 Page 20 of 27  :

l i

l i

i t

e t

I s.

-1 1

'l l

)

STAGE 2. TORC CHANNEL Omaha Public Power District Figure GE0 METRY FOR FCS UNIT N0. 1 Fort Calhoun Station Unit No. 1 4-5 <

i OPPD-NA-8301-NP, Rev. 06 )

Page 21 of 27 l

T 2

h STAGE 3 TORC CHANNEL Omaha Public Power District Figure GE0 METRY FOR FCS UNIT NO. 1 Fort Calhoun Station Unit No.-1 4-6 ,

OPPD-NA-8301-NP, Rev. 06 Page 22 of-27

f *

~

~,

/

/

' .,.tl'f I,j' -

7 ./ N. \.

C.  !

. ../

u i

i

/  :

l '/

/

~

1 -

I I l

t o

/ -

=

\j  ! , _ g l\ ,

',k -

/

'/ \

/ ,. \ \ -

/

/ '

/ g a / '

\.

\

8

( l \ _

l

~

\, \

l , ,

srw 7'%

i i

\'N,N, i .

l

~. ~- eu 9 9 5 at 9 = m * =

u m - -- -. -- -- o a o

~. .

1 l.

s a s s a  :

i 4 4 s R l a

?  ? = = . d l  ! .

l l

AXIAL POWER DISTRIBUTIONS Omaha Public Power District Figure  !

FORT CALHOUN CYCLE 8 Fort Calhoun Station Unit No. 1 4-7 OPPD-NA-8301-NP, Rev. 06 -l Page 23 of 27

i i

INLET FLOW DISTRIBUTION Omaha Public Power District Figure FOR FCS 4-PUMP OPERATION Fort Calhoun Station Unit No. 1 4-8 OPPD-NA-8301-NP,.Rev. 06 Page 24 of 27

t-1 i

EXIT PRESSURE DISTRIBUTION Omaha Public Power District Figure

.FOR FCS 4-PUMP OPERATION Fort Calhoun Station Unit No. 1 4-9 OPPD-NA-8301-NP, Rev. 06 Page 25~of 27

l 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. 06-Page 26 of 27

i l

l l

l b

j l

i l

i 1

.1 l

ALGORITHM FOR DNB Omaha 'Public Power District Figure LC0 MONITORING Fort Calhoun Station Unit No. 1 6-1 OPPD-NA-8301-NP, Rev. 06 Page 27 of 27

. - .

Retrieved from "https://kanterella.com/w/index.php?title=ML20070P372&oldid=2824440"