LIC-10-0014, Fort Calhoun, Unit 1, Re-transmittal of Core Operating Limits Report (Colr) Due to Typographical Error Discovered and Corrected: Difference between revisions

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#REDIRECT [[LIC-10-0014, Re-transmittal of Core Operating Limits Report (Colr) Due to Typographical Error Discovered and Corrected]]
| number = ML100620786
| issue date = 02/22/2010
| title = Fort Calhoun, Unit 1, Re-transmittal of Core Operating Limits Report (Colr) Due to Typographical Error Discovered and Corrected
| author name = Matthews T C
| author affiliation = Omaha Public Power District
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000285
| license number =
| contact person =
| case reference number = LIC-10-0014
| document type = Fuel Cycle Reload Report, Letter type:LIC
| page count = 20
}}
 
=Text=
{{#Wiki_filter:Omaha Public Power District Fort Calhoun Station 9610 Power Lane Blair, NE 68008 February 22, 2010 LIC-10-0014 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
 
==References:==
 
==SUBJECT:==
: 1. Docket No. 50-285 2. Letter from OPPD (T. C. Matthews) to NRC (Document Control Desk),"Transmittal of Fort Calhoun Station (FCS), Unit No. 1, Core Operating Limit Report (COLR), Revision 39" dated November 25, 2009 (LIC-09-0097)Re-transmittal of Fort Calhoun Station (FCS), Unit No. 1, Core Operating Limits Report (COLR)Pursuant to FCS Technical Specification 5.9.5d, an update of the FCS COLR is attached.
In Reference 2, Revision 39 of the FCS COLR was submitted.
Subsequently, a typographical error in the title was discovered and corrected.
The revision number of the COLR remains 39 as issued on November 20, 2009 for Cycle 26 operation.
No regulatory commitments are contained in this submittal.
If you have any questions, please contact Ms. Susan Baughn at (402) 533-7215.Sincerely, T. C. Matthews Manager -Nuclear Licensing TCM/mle
 
==Attachment:==
 
Core Operating Limits Report, Revision 39 c: E. E. Collins, NRC Regional Administrator, Region IV (w/o Attachment)
L. E. Wilkins, NRC Project Manager (w/o Attachment)
J. C. Kirkland, NRC Senior Resident Inspector (w/o Attachment)
Lk (M(Employment with Equal Opportunity 4171 PAGE 1 OF 19 Fort Calhoun Station Unit 1 TDB-Vl TECHNICAL DATA BOOK CORE OPERATING LIMITS REPORT I Change No. EC 44759 Reason for Change Updated for Cycle 26 operation.
Requestor D. Williamson Preparer .L. Hautzinger Editorial Correction (a) Pages 1, 2 and 6 (02-18-10)
EC 48315 Issue Date 11-20-09 3:00 pm R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 2 OF 19 Fort Calhoun Station, Unit I Core Operatinq Limits Report I Due to the critical aspects of the safety analysis inputs contained in this report, changes may not be made to this report without concurrence of the Nuclear Engineering Department.
R39 FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 3 OF 19 TABLE OF CONTENTS Item Description Paqe 1. IN T R O D U C T IO N .............................................................................................
6 2. CO RE O PERATING LIM ITS ...........................................................................
6 3 .T M /L P L IM IT .................................................................................................
..8 4. MAXIMUM CORE INLET TEMPERATURE
....................................................
8 5. POWER DEPENDENT INSERTION LIMIT ....................................................
8 6. LIN EA R H EA T RA T E ......................................................................................
9 7. EXCORE MONITORING OF LHR ...................................................................
9 8. PEA KING FACTO R LIM ITS ............................................................................
9 9. D N B M O N IT O R IN G ........................................................................................
9 T 10. F FT AND CORE POWER LIMITATIONS
.............................
9 11. REFUELING BORON CONCENTRATION
......................................................
9 12. AXIAL POW ER DISTRIBUTION
...................................................................
10 13. SHUTDOWN MARGIN WITH Tcold> 210°F ..................................................
10 14. MOST NEGATIVE MODERATOR TEMPERATURE COEFFICIENT
.............
10 15. STEAM GENERATOR DIFFERENTIAL PRESSURE ...................................
10 R39 FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 4 OF 19 LIST OF TABLES Table No. Title Pacqe 1 T M /LP C oefficients
..........................................................................................
8 2 R efueling Boron C oncentrations
.........................................................................
9 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 5 OF 19 LIST OF FIGURES Figure No.1 2 3 4 5 6 7 8 9 Title Page Thermal Margin/Low Pressure for 4 Pump Operation
.....................................
11 Power Dependent Insertion Limit ...................................................................
12 Allowable Peak Linear Heat Rate vs. Burnup ................................................
13 Excore M onitoring of LH R ................................................................................
14 D N B M o nito ring ............................................................................................
..15 FRT and Core Power Limitations
.....................................................................
16 Axial Power Distribution LSSS for 4 Pump Operation
.....................................
17 Axial Power Distribution Limits for 4 Pump Operation with Incores Inoperable..
18 Minimum Boric Acid Storage Tank Level vs. Stored Boric Acid Storage Tank C o nce ntratio n ..............................................................................................
..19 R39 FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 6 OF 19 CORE OPERATING LIMITS REPORT 1. INTRODUCTION This report provides the cycle-specific limits for operation of the Fort Calhoun Station Unit 1 for Cycle 26 operation.
It includes limits for: " TM/LP LSSS for 4 Pump Operation (PVAR)* Core Inlet Temperature (TIN)" Power Dependent Insertion Limit (PDIL)" Allowable Peak Linear Heat Rate* Excore Monitoring of LHR* Integrated Radial Peaking Factor (FRT)* DNB Monitoring
* FRT versus Power Trade-off Curve" Refueling Boron Concentration
* Axial Power Distribution (APD)* Shutdown Margin with TCOLD > 210°F* Most Negative Moderator Temperature Coefficient These limits are applicable for the duration of the cycle. For subsequent cycles the limits will be reviewed and revised as necessary.
In addition, this report includes a number of cycle-specific coefficients used in the generation of certain reactor protective system trip setpoints or allowable increases in radial peaking factors.2. CORE OPERATING LIMITS All values and limits in this TDB section apply to Cycle 26 operation.
This cycle must be operated within the bounds of these limits and all others specified in the Technical Specifications.
This report has been prepared in accordance with the requirements of Technical Specification 5.9.5. The list of references below are complete citations of topical reports and include the report number, title, revision, date, and any supplements in accordance with the basis for NRC approval of License Amendment No. 196 which eliminated these specific entries from Technical Specification 5.9.5. NRC approval of Amendment No. 196 is consistent with the requirements of the Technical Specification Task Force, Improved Standard Technical Specification Change Traveler, "Revise Topical Report References in ITS 5.6.5 COLR" (TSTF-363-A, Rev. 0). In accordance with this Traveler and Amendment No. 196, this information must be maintained within this TDB section.R39 FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 7 OF 19 The values and limits presented within this TDB section have been derived using the NRC approved methodologies listed below:* OPPD-NA-8301, "Reload Core Analysis Methodology Overview," Revision 8, dated August 2004. (TAC No. MC4304)* OPPD-NA-8302, "Reload Core Analysis Methodology, Neutronics Design Methods and Verification," Revision 6, dated August 2004. (TAC No. MC4304)* OPPD-NA-8303, "Reload Core Analysis Methodology, Transient and Accident Methods and Verification," Revision 7, dated August 2005. (TAC No. MC4304)* XN-75-32(P)(A)
Supplements 1, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.* XN-NF-79-56(P)(A), Revision 1, Supplement 1, "Gadolinia Fuel Properties of LWR Fuel Safety Evaluation," November 1981.* XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Revision 1, October 1986.* XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," August 1985.* ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," December 1991.* EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," Revision 0, February 1999.* XN-NF-78-,44(P)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," October 1983.* XN-NF-82-21 (P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Revision 1, September 1983.* EMF-1961(P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Revision 0, July 2000." ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Revision 0, May 1992." EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Revision 1, January 2005." EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," Revision 0, April 2003." EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," Revision 0, March 2001.* EMF-96-029(P)(A)
Volume 1, EMF-96-029(P)(A)
Volume 2, EMF-96-029(P)(A)
Attachment, "Reactor Analysis System for PWRs, Volume 1 -Methodology Description, Volume 2 -Benchmarking Results," January 1997.* EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Revision 1, May 2004.* BAW-10240(P)(A), "Incorporation of M5TM Properties in Framatome ANP Approved Methods," Revision 0, May 2004.R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 8 OF 19 3. TM/LP LIMIT The TM/LP coefficients are shown below: Table 1 -TM/LP Coefficients Coefficient a 13 Value 29.6 20.63-12372 The TM/LP setpoint is calculated by the PVAR equation, shown below and in Figure 1: PVAR = 29.6 PF(B) AI(Y)B + 20.63T1N -12372 PF(B) = 1.0= -0.008(B)+1.8
= 1.4 AI(Y) = -0.6666(Y 1) + 1.000= +0.3333(Y 1) + 1.000 for B > 100%for 50% < B < 100%for B < 50%for Yj < 0.00 for Y 1> 0.00 Where: B = High Auctioneered thermal (AT) or Nuclear Power, % of rated power Y = Axial Shape Index, asiu TIN = Core Inlet Temperature, 'F PVAR = Reactor Coolant System Pressure, psia 4. MAXIMUM CORE INLET TEMPERATURE The maximum core inlet temperature (TIN) shall not exceed 545&deg;F. This value includes instrumentation uncertainty of +/-2 0 F (Ref: FCS Calculation FC06292, 6/9/95).This limit is not applicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step greater than 10% of rated thermal power.5. POWER DEPENDENT INSERTION LIMIT The power dependent insertion limit is defined in Figure 2.R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 9 OF 19 6. LINEAR HEAT RATE The allowable peak linear heat rate is shown in Figure 3.7. EXCORE MONITORING OF LHR The allowable operation for power versus axial shape index for monitoring of LHR with excore detectors is shown in Figure 4.8. PEAKING FACTOR LIMITS The maximum full power value for the integrated radial peaking factor (FRT) is 1.732.9. DNB MONITORING The core operating limits for monitoring of DNB are provided in Figure 5. This figure provides the allowable power versus axial shape index for the cycle.10. FRRT AND CORE POWER LIMITATIONS Core power limitations versus FRT are shown in Figure 6.11. REFUELING BORON CONCENTRATION The refueling boron concentration is required to ensure a shutdown margin of not less than 5% with all CEAs withdrawn.
The refueling boron concentration must be at least 1,900 ppm through the end of Cycle 25 operation and is valid until the beginning of core reload for Cycle 26.Listed below in Table 2 are the refueling boron concentration values for Cycle 26 operations:
Table 2 -Refueling Boron Concentrations Cycle Average Burnup (MWD/MTU)BOC> 2,000> 4,000 Refueling Boron Concentration (ppm)2,160 2,016 1,900 R39 FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 10 OF 19 12. AXIAL POWER DISTRIBUTION The axial power trip is provided to ensure that excessive axial peaking will not cause fuel damage. The Axial Shape Index is determined from the axially split excore detectors.
The setpoint functions, shown in Figure 7 ensure that neither a DNBR of less than the minimum DNBR safety limit nor a fuel centerline temperature greater than the associated safety limit (that which would result in fuel melting) will exist as a consequence of axial power maldistributions.
The calculated cycle-specific FCM temperature for Cycle 26 corresponds to 22.845 kw/ft. Allowances have been made for instrumentation inaccuracies and uncertainties associated with the excore symmetric offset -incore axial peaking relationship.
Figure 8 combines the LHR LCO tent from Figure 4, the DNB LCO tent from Figure 5, and the APD LSSS tent from Figure 7 into one figure for a visual comparison of the different limits.13. SHUTDOWN MARGIN WITH Tcold > 210&deg;F Whenever the reactor is in hot shutdown, hot standby or power operation conditions, the shutdown margin shall be _3.6% Ak/k. With the shutdown margin <3.6% Ak/k, initiate and continue boration until the required shutdown margin is achieved.14. MOST NEGATIVE MODERATOR TEMPERATURE COEFFICIENT The moderator temperature coefficient (MTC) shall be more positive than-3.25 x 10-4 Ap/0 F, including uncertainties, at rated power.15. STEAM GENERATOR DIFFERENTIAL PRESSURE The steam generator differential pressure trip of Technical Specification Table 2-11, Item 9 at 135 psid ensures that neither a DNBR of less than the minimum DNBR safety limit nor a fuel centerline temperature greater than the associated safety limit (that which would result in fuel melting) will exist as a consequence of axial power maldistributions resulting from asymmetric steam generator transients.
The calculated cycle-specific FCM temperature for Cycle 26 corresponds to 22.845 kw/ft.)R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 11 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 12 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 13 OF 19 18 16 14 LL S12 Lu 10 U.I z 6 4 2 UNACCEPTABLE OPERATION 15.5 KW/FT ACCEPTABLE OPERATION II I 4 4 4.4 I. 4 I h I 0 5000 10000 15000 20000 BURNUP (MWD/MTU)COLR ALLOWABLE PEAK LINEAR HEAT RATE FIGURE IZ E LLWALEVS.
BURNUP 1 3 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 14 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 15 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 16 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 17 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 18 OF 19 R39 FORT CALHOUN STATION TECHNICAL DATA BOOK TDB-VI PAGE 19 OF 19 R39}}

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