ML15134A014

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TDB-VI, Technical Data Book, Change No. EC 63878, Core Operating Limits Report, Cycle 28
ML15134A014
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 04/30/2015
From: Bessey K
Omaha Public Power District
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EC 63878 TDB-VI
Download: ML15134A014 (21)
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PASSPORT DOCUMENT To HORN KRISTI Facility FC Department : TRANSMITTAL Address DIRECTOR, OFFICE OF NRC - DOCUMENT From Address  :

CONTROL DESK, U.S. NRC WASHINGTON, DC 20555 DOCCON FC-2-3 Attention: Doc Management Distribution 11IIII Page IIIIII

+IIIII11 III 1llll City Fort Calhour State: P rovince:

Postal Country: UNITED STATES Email Contact Kristi Horn 402-533-6714 6714 Date/Time 04/30/2015 10:14 Transmittal Group Id: 043015-1 Trans No. 000153237

Title:

04/30/15 - ISSUE 1 Total Items: 00001 See Notes and Comments below.

Item Facility Type Sub Document Number Sheet Doc Status Revision Doc Date Copy # Media Cpyl 0001 FC PROC TDB TDB-VI ACTIVE 042 51 P G 01 Security : D Destroy Documents Date:

Form of Destruction Signature of Destroyer Signature of Witness Notes and Comments GENERAL MM-ST-HSS-0004, REV 18, FIELD COPY 04/28/15@1115 PE-RR-VX-0414S, REV 12, FIELD COPY 04/28/15@1100 MS-CP-01-DG2, REV 13, FIELD COPY 04/28/15@0850 EM-ST-EE-0005, REV 27, FIELD COPY 04/28/l5@1720 MS-CP-01-MG/22KV, REV 27, FIELD COPY 04/28/15@1340

T-2 PASSPORT DOCUMENT To HORN KRISTI Facility FC Department : TRANSMITTAL Address DIRECTOR, OFFICE OF NRC - DOCUMENT CONTROL DESK, U.S. NRC WASHINGTON, DC 20555 Page: 2 From Address DOCCON FC-2-3 Attention: Doc Management Distribution lNlNllllI[UllIlNl City Fort Calhoun State: P rovince:

Postal Country: UNITED STATES Email Contact Kristi Horn 402-533-6714 6714 Date/Time 04/30/2015 10:14 Transmittal Group Id: 043015-1 Trans No. 000153237

Title:

04/30/15 - ISSUE 1 Total Items: 00001 If a document was not received or is no longer required check the response below and return to sender.

Documents noted above not received (identify those not received).

I no longer require distribution of these documents (identify those no longer required).

Date: Signature:

PAGE 1 OF 19 Fort Calhoun Station Unit 1 TDB-VI TECHNICAL DATA BOOK CORE OPERATING LIMITS REPORT Change No. EC 63878 Reason for Change Change Cycle 27 to Cycle 28. Add revision numbers and issue dates to references.

Requestor L. Lees Preparer K. Bessey Issue Date 04-30-15 3:00pm R42

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 2 OF 19 Fort Calhoun Station, Unit 1 Core Operating Limits Report 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.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 3 OF 19 TABLE OF CONTENTS Item Description Page

1. INT R O D UC T IO N............................................................................................. 6
2. CO RE O PERATING LIM ITS ........................................................................... 6
3. T M/LP LIMIT ................................................................................................ .. 8
4. MAXIMUM CORE INLET TEMPERATURE .................................................... 8
5. POWER DEPENDENT INSERTION LIMIT .................................................... 8
6. LIN EA R HEAT RA TE ...................................................................................... 9
7. EXCORE MONITORING OF LHR ................................................................... 9
8. PEAKING FACTO R LIM ITS ........................................................................... 9
9. D NB MO NITO R ING ........................................................................................ 9
10. FRT AND CORE POWER LIMITATIONS ........................................................ 9
11. REFUELING BORON CONCENTRATION ...................................................... 9
12. AXIAL POWER DISTRIBUTION ................................................................... 10
13. SHUTDOWN MARGIN WITH Tcold > 210°F .................................................. 10
14. MOST NEGATIVE MODERATOR TEMPERATURE COEFFICIENT ............ 10
15. STEAM GENERATOR DIFFERENTIAL PRESSURE ................................... 10 R42

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 4 OF 19 LIST OF TABLES Table No. Title Paqe 1 TM/LP Coefficients .......................................................................................... 8 2 Refueling Boron Concentrations .................................................................... 9 R42

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 5 OF 19 LIST OF FIGURES Figure No. Title Paqe 1 Thermal Margin/Low Pressure for 4 Pump Operation .................................... 11 2 Power Dependent Insertion Limit .................................................................... 12 3 Allowable Peak Linear Heat Rate vs. Burnup ................................................. 13 4 Excore Monitoring of LHR ............................................................................... 14 5 D NB Mo nito ring .......................................................................................... . . 15 6 FRT and Core Power Limitations ................................................................... 16 7 Axial Power Distribution LSSS for 4 Pump Operation .................................... 17 8 Axial Power Distribution Limits for 4 Pump Operation with Incores Inoperable.. 18 9 Minimum Boric Acid Storage Tank Level vs. Stored Boric Acid Storage Tank C o nce ntratio n ............................................................................................... . . 19 R42

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 28 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
  • Maximum Radial Peaking Factor (FRT)
  • 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 28 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.

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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," Revision 0, September 1986.

" ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Revison 0, 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-1 961 (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.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 8 OF 19

3. TM/LP LIMIT The TM/LP coefficients are shown below:

Table 1 -TM/LP Coefficients Coefficient Value a 29.6 13 20.63 P

-12372 The TM/LP setpoint is calculated by the PVAR equation, shown below and in Figure 1:

PVAR = 29.6 PF(B) AI(YI)B + 20.63TIN - 12372 PF(B) = 1.0 for B > 100%

= -0.008(B)+1.8 for 50% < B < 100%

= 1.4 for B < 50%

Al (YI) = -0.6666(YI) + 1.000 for Y,

  • 0.00

= +0.3333(Yi) + 1.000 for Y, > 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°F. This value includes instrumentation uncertainty of +/-2 0F (Ref: FCS Calculation FC06292, Revision 2).

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.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK 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 maximum 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. FR*T 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 27 operation and is valid until the beginning of core reload for Cycle 28.

Listed below in Table 2 are the refueling boron concentration values for Cycle 28 operations:

Table 2 - Refueling Boron Concentrations Cycle Average Burnup Refueling Boron (MWD/MTU) Concentration (ppm)

BOC 2,160

_2,000 2,016

>4,000 1,900 R42

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 28 corresponds to 22.776 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-F Whenever the reactor is in hot shutdown, hot standby or power operation conditions, the shutdown margin shall be _Ž3.6% 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 -3.30 x 10-4 Ap/°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 28 corresponds to 22.776 kw/ft.

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 11 OF 19 590 ASI = 0.00 U- 580 L

570 24010 psia I- 560 550 IL w

540 z

530 0

0 520 4. 4 .1 4 4.

w 0

0 510 4. 4 4 4 4.

500 60 70 80 90 100 110 120 CORE POWER (% OF RATED)

PVAR = 29.6 PF(B)A1(YI)B + 20.63TIN - 12372 PF(B) = 1.0 B > 100% A1(YI) = -0.6666Y, + 1.000 Y, _*0.00

= -.008B+1.8 50% < B < 100% = +0.3333Y, + 1.000 Y, > 0.00

= 1.4 B _<50%

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FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 12 OF 19 R42

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 13 OF 19 18 UNACCEPTABLE OPERATION 16 15.5 KW/FT 14 U.. ACCEPTABLE OPERATION

' 12 LUJ I.-

10 I--

z 6

4 2

0 5000 10000 15000 20000 BURNUP (MWD/MTU)

COLR j

ALLOWABLE PEAK LINEAR HEAT RATE LOWBEVS. BURNUP1 FIGURE 3

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