ML18012A364
| ML18012A364 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/11/1996 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18012A363 | List: |
| References | |
| NUDOCS 9609200197 | |
| Download: ML18012A364 (30) | |
Text
Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 August 8, 1996 NFMASA File: 2495-0056 qg0qP00i+7 9609ii II PDR ADOCK 05000400',I t P PDR
Sheet 1 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 1.0 This Core Operating Limits Report (COLR) for Shearon Hams Unit 1 Cycle 7 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6.
The Technical Specifications affected by this report are listed below:
3/4.1.1.2 SHUTDOWN MARGIN - Modes 3,4, and 5 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor - Fz(Z) 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor - F~,
3/4.9.1.a Boron Concentration During Refueling Operations
Sheet 2 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 2.0 The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC-approved methodologies specified in Technical Specification 6.9.1.6 and given in Section 3.0.
2.1 (Specification 3/4.1 1.2)
~
The SHUTDOWN MARGIN versus RCS boron concentration - Modes 3, 4, and 5 is specified in Figure 1.
2.2 (Specification 3/4.1.1.3)
- 1. The Moderator Temperature Coefficient gvITC) limits are:
The Positive MTC Limit (ARO/HZP) shall be less positive than+5.0 pcm/'F for power levels up to 70% RTP with a linear ramp to 0 pcm/'F at 100% RTP.
The Negative MTC Limit (ARO/RTP) shall be less negative than P5 pcm/'F.
- 2. The MTC Surveillance limit is:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to
-37.0 pcm/'F.
where: ARO stands for All Rods Out HZP stands for Hot Zero THERMALPOWER RTP stands for RATED THERMALPOWER 2.3 'Specification 3/4.1.3.5)
Fully withdrawn for all shutdown rods shall be 231 steps.
2.4 (Specification 3/4.1.3.6)
The control rod banks shall be limited in physical insertion as specified in Figure 2. Fully withdrawn for all control rods shall be 231 steps.
2.5 (Specification 3/4.2.1)
The AXIALFLUX DIFFERENCE (AFD) target band is specified in Figure 3.
Sheet 3 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 2.6 - Fz(Z) (Specification 3/4.2.2)
Fz(Z) s Fz
- K(Z)/P for P > 0.5 Fz(Z) z F~"
- K(Z)/0.5 for P g 0.5 where: P = THERMALPOWER/RATED THERMALPOWER
- a. F<~' 2.45 for LOPAR and VANTAGE5 fuel
- b. F<"r = 2.52 for SPC fuel
- c. K(Z) is specified in Figure 4.
V(Z) Curve for PDC-3 Operation is specified in Figure 5. The V(Z) curve is sufficient to determine the PDC-3 V(Z) versus core height for Cycle 7 burnups through the end of full power reactivity plus a coastdown for a maximum cycle energy of 519 EFPDs.
2.7 - Fz, (Specification 3/4.2.3)
F~icF~ '*(1+PF~i*(1 -P))
where: P = THERMALPOWER/lMTED THERMALPOWER
- a. F~,
~ = 1.62 for LOPAR fuel
- b. F<,P' 1.65 for VANTAGE5 fuel
- c. Fz" ' 1.73 for SPC fuel
- d. PF+, = 0.3 for LOPAR fuel
- e. PFz, = 0.35 for VANTAGE5 and SPC fuel 2.8 (Specification 3/4.9.1.a)
Through the end of Cycle 7, the boron concentration required to maintain K, less than or equal to
.95 is equal to 2304 ppm. Boron concentration must be maintained greater than or equal to 2304 ppm during refueling operations.
Sheet 4 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 3.0 METHODOLOGYREFERENCES
- 1. XN-75-27(A), and Supplements 1, 2, 3, 4, and 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - Modes 3, 4, and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).
- 2. ANF-89-151(A), and Correspondence, "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 3. XN-NF-82-21(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
XN-75-32(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.2- Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
XN-NF-84-93(A), and Supplement 1, "Steamline Break Methodology for PWRs," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, and 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor).
Sheet 5 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 (Continued)
XN-NF-82-20(A), Revision 1 and Supplements 1, 2, 3, and 4, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-81-58(A), Revision 2 and Supplements 1, 2, 3, and 4, "RODEX2 Fuel Rod Thermal Response Evaluation Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-16(A), Volume 1 and Supplements 1, 2, and 3, Volume 2, Revision 1 and Supplement 1, "PWR 17x17 Fuel Cooling Test Program," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-105(A), and Supplement 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
XN-NF-7W4(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6-Control Bank Insertion Limits, and 3.2.2 -'eat Flux Hot Channel Factor).
ANF-88-054(A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," Advanced Nuclear Fuels Corporation, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, and 3.2.2- Heat Flux Hot Channel Factor).
Sheet 6 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 3.0 (Continued)
- 9. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (E Proprietary).
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - Modes 3, 4, and 5, 3.2.2- Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 10. WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATIONMODEL USING THE BASH CODE," March 1987 (E Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
- 11. WCAP-11837-P-A, "EXTENSION OF METHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES," January 1990 (E Proprietary).
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 12. EMF-92-081(A), and Supplement 1, "Statistical Setpointfl'ransient Methodology for Westinghouse Type Reactors," Siemens Power Corporation, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 13. EMF-92-153(A), and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Nuclear Power Corporation, Richland, WA 99352.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 14. XN-NF-8249(A), Revision 1, and XN-NF-8249(P), Revision 1, Supplement 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
Sheet 7 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 4.0 4.1
- 1. Qgpzbjlity: The Movable Incore Detection System shall be OPERABLE with:
- a. At least 38 detector thimbles at beginning of cycle (75% of the total number),
where the beginning of cycle is defined in this instance as a flux map determination that the core is loaded consistent with design,
- b. A minimum of 38 detector thimbles for the remainder of the operating cycle,
- c. A minimum of two detector thimbles per core quadrant, and
- d. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
~pJigahility.: When the Movable Incore Detection System is used for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANTPOWER TILTRATIO, or
- c. Measurement of F>>, and Fz(Z)
- The Movable Incore Detection System shall be demonstrated OPERABLE, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to use, by irradiating each detector used and determining the acceptability of its voltage curve when required for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANTPOWER TILTRATIO, or
- c. Measurement of F<<and Fz(Z)
The OPERABILITYof the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
For the purpose of measuring F~(Z) or F<<, a full incore flux map is used.
Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.
Sheet 8 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 4.0 (Continued)
In order to change the requirements concerning the number and location of operable detectors, the NRC staff deems that a rigorous evaluation and justification is required. The following is a list of elements that must be part of a 50.59 determination and available for audit ifthe licensee wishes to change the reqiurements:
How an inadvertent loading of a fuel assembly into an improper location willbe detected, How the validity of the tilt estimates will be ensured, C. How adequate core coverage willbe maintained, How the measurement uncertainties willbe assured and why the added uncertainties are adequate to guarantee that measured nuclear heat flux hot channel factor, nuclear enthalpy rise hot channel factor, radial peaking factor and quadrant power tilt factor meet Technical Specification limits, and How the Movable Incore Detection System will be restored to full (or nearly full) service before the beginning of each cycle.
Sheet 9 of 13 Harris Unit I Cycle 7 Core Operating Limits Report - Rev. 2 Figure 1 Shutdown Margin Versus RCS Boron Concentration
~ ~ ~ ~ I ~
- - ($ 600,6+0)--
1 J' 6000 I
I U ~ I - MQGE4INIIHNORCPs -- I I I K Ig OPERATION'ND; 4000 . MODE 5 I 2! I I I
~ I ~ J I I J ~ I ~ J O
Cl I-D 8000 T.'l (5600,2600)
- - - MODE 3,-ANDklODE4-WIAAT L5hST ONE RCP INOPERAn MODES 3 &'4 (1500,1770)
I I
MODE 5 ~ ~ ~ ~
(350,1000) 0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 REQUIRED RCS BORON CONCENTRATION (ppm)
Modes 3, 4, and 5/Drained Applicable to Mode 4, with or without RCPs in operation
Sheet 10 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 Figure 2 Rod Group Insertion Limits Versus Thermal Power (Three-Loop Operation) 240 220 ~ ~ ~ ~ I ~ ~ I ~ ~ ~ ~ I ~ ~ '
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0 0 0.1 0.2 0.3 0.4 0,5 0.6 0%7 0.8 019 1 FRACTION OF RATED THERMAL POWER (Fully withdrawn shall be 231 steps)
Note: Control Banks A and B must be withdrawn from the core prior to power operation.
Sheet 11 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 Figure 3 Axial Flux Difference Limits as a Function of Rated Thermal Power 120
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(DEVIATION FROM TARGET AFD)
Note: At power levels less than HFP, the deviation is applied to the target AFD appropriate to that power level. The target AFD varies linearly between the HFP target and zero at zero power.
Sheet 12 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 Figure 4 K(Z) - Local Axial Penalty Function for FQ(Z) 1.2
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Sheet 13 of 13 Hams Unit 1 Cycle 7 Core Operating Limits Report - Rev. 2 Figure 5 V(Z) Versus Core Height 1.250 I ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ I ~ ~ I ~ ~ I I ~ ~
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- 11 400 1.000 411.600 1.000
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- 12.000 1.000 Note: Top and bottom 15% excluded as per Technical specification 4.2.2.4.
Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 August 14, 1996 NFM8cSA File: 2495.0056
Sheet 1 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 1.0 This Core Operating Limits Report (COLR) for Shearon Harris Unit 1 Cycle 7 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6.
The Technical Specifications affected by this report are listed below:
3/4.1.1.2 SHUTDOWN MARGIN- Modes 3, 4, and 5 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor - F<(Z) 3/4.2.3 Nuclear Enthaipy Rise Hot Channel Factor - Fzu 3/4.9.1.a Boron Concentration During Refueling Operations
Sheet 2 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 2.0 The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC-approved methodologies specified in Technical Specification 6.9.1.6 and given in Section 3.0.
2.1 (Specification 3/4.1.1.2)
The SHUTDOWN MARGIN versus RCS boron concentration - Modes 3, 4, and 5 is specified in Figure 1.
2.2 (Specification 3/4.1.1.3)
The Moderator Temperature Coefficient (MTC) limits are:
The Positive MTC Limit (ARO/HZP) shall be less positive than+5.0 pcm/'F for power levels up to 70% RTP with a linear ramp to 0 pcm/'F at 100% RTP.
The Negative MTC Limit (ARO/RTP) shall be less negative than -45 pcm/'F.
The MTC Surveillance limitis:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to
-37.0 pcm/'F.
where: ARO stands for AllRods Out HZP stands for Hot Zero THERMALPOWER RTP stands for RATED THERMALPOWER 2.3 (Specification 3/4.1.3.5)
Fully withdrawn for all shutdown rods shall be 231 steps.
2.4 (Specification 3/4.1.3.6)
The control rod banks shall be limited in physical insertion as specified in Figure 2. Fully withdrawn for all control rods shall be 231 steps.
2.5 (Specification 3/4.2.1)
The AXIALFLUX DIFFERENCE (AFD) target band is specified in Figure 3.
Sheet 3 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 2.6 - Fz(Z) (Specification 3/4.2.2)
F<(Z) Z F< ~* K(Z)/P for P > 0.5 F<(Z) z F<"~
- K(Z)/0.5 for P g 0.5 where: P = THERMALPOWER/RATED THERMALPOWER
- a. Fz" =2.45forLOPARandVANTAGE5 fuel
- b. Fz~ = 2.52 for SPC fuel
- c. K(Z) is specified in Figure 4.
V(Z) Curve for PDC-3 Operation is specified in Figure 5. The V(Z) curve is sufficient to determine the PDC-3 V(Z) versus core height for Cycle 7 burnups through the end of full power reactivity plus a coastdown for a maximum cycle energy of 519 EFPDs.
2.7 - Fz, (Specification 3/4.2.3)
Fm~Fau *(1+PFnu" (1-P))
where: P = THERMALPOWER/RATED THERMALPOWER Fz~ = 1.62 for LOPAR fuel Fq~ = 1.65 for VANTAGE5 fuel Fz" = 1.73 for SPC fuel PF>, 0.3 for LOPAR fuel PF~, = 0.35 for VANTAGE5 and SPC fuel 2.8 (Specification 3/4.9.1.a)
Through the end of Cycle 7, the boron concentration required to maintain ~ less than or equal to
.95 is equal to 2304 ppm. Boron concentration must be maintained greater than or equal to 2304 ppm during refueling operations.
Sheet 4 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 3.0 METHODOLOGY REFERENCES XN-75-27(A), and Supplements 1, 2, 3, 4, and 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.1.2- SHUTDOWN MARGIN- Modes 3, 4, and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).
ANF-89-151(A), and Correspondence, "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
P XN-NF-82-21(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
XN-75-32(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod l3owing," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 5. XN-NF-84-93(A), and Supplement 1, "Steamline Break Methodology for PWRs," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, and 3.2.3-Nuclear Enthalpy Rise Hot Channel Factor).
Sheet 5 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 3.0 (Continued)
XN-NF-82-20(A), Revision 1 and Supplements 1, 2, 3, and 4, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-81-58(A), Revision 2 and Supplements 1, 2, 3, and 4, "RODEX2 Fuel Rod Thermal Response Evaluation Model," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-16(A), Volume 1 and Supplements 1, 2, and 3, Volume 2, Revision 1 and Supplement 1, "PWR 17x17 Fuel Cooling Test Program," Exxon Nuclear Company, Richland, WA 99352.
XN-NF-85-105(A), and Supplement 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 7. XN-NF-78%(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6-Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).
ANF-88-054(A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," Advanced Nuclear Fuels Corporation, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, and 3.2.2 - Heat Flux Hot Channel Factor).
Sheet 6 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 3.0 (Continued)
- 9. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"July 1985 (E Proprietary).
(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - Modes 3, 4, and 5, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 10. WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATIONMODEL USING THE BASH CODE," March 1987 Qf. Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
- 11. WCAP-11837-P-A, "EXTENSION OF METHODOLOGYFOR CALCULATING TRANSITION CORE DNBR PENALTIES," January 1990 (E Proprietary).
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 12. EMF-92-081(A), and Supplement 1, "Statistical SetpoinbTransient Methodology for Westinghouse Type Reactors," Siemens Power Corporation, Richland, WA 99352.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5-Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 13. EMF-92-153(A), and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Nuclear Power Corporation, Richland, WA 99352.
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 14. XN-NF-8249(A), Revision 1, and XN-NF-82-49(P), Revision 1, Supplement 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Exxon Nuclear Company, Richland, WA 99352.
(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
"J Sheet 7 of 13 Hams Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 4.0 4.1
- 1. Q~~~i': The Movable Incore Detection System shall be OPERABLE with:
At least 38 detector thimbles at beginning of cycle (75% of the total number),
where the beginning of cycle is defined in this instance as a flux map determination that the core is loaded consistent with design, A minimum of 38 detector thimbles for the remainder of the operating cycle, A minimum of two detector thimbles per core, quadrant, and Sufficient movable detectors, drive, and readout equipment to map these tliimbles.
- 2. 3gg~tl~: When the Movable Incore Detection System is used for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANT POWER TILTRATIO, or
- c. Measurement of F, and F<(Z)
- 3. : The Movable Incore Detection System shall be demonstrated OPERABLE, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to use, by irradiating each detector used and determining the acceptability of its voltage curve when required for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANT POWER TILTRATIO, or
- c. Measurement of F, and Fz(Z)
The OPERABILITYof the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining tlie acceptability of its voltage curve.
For the purpose of measuring Fz(Z) or F<<, a full incore flux map is used.
Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring QUADRANTPOWER TILT RATIO when one Power Range channel is inoperable.
Sheet 8 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 4.0 (Continued)
In order to change the requirements concerning the number and location of operable detectors, the NRC staff deems that a rigorous evaluation and justification is required. The following is a list of elements that must be part of a 50.59 determination and available for audit ifthe licensee wishes to change the requirements:
How an inadvertent loading of a fuel assembly into an improper location willbe detected, How the validity of the tiltestimates will be ensured, How adequate core coverage willbe maintained, How the measurement uncertainties willbe assured and why the added uncertainties are adequate to guarantee that measured nuclear heat flux hot channel factor, nuclear enthalpy rise hot channel factor, radial peaking factor and quadrant power tilt factor meet Technical Specification limits, and How the Movable Incore Detection System willbe restored to full (or nearly full) service before the beginning of each cycle.
ro Sheet 9 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 Figure 1 Shutdown Margin Versus RCS Boron Concentration I I I I I (
I I I 7000 ~ ~ I I I 4 I I I I I I
- -((600,@00)--
I I I I
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I I I I 6000 I I z - MQBE 4 Q I I I ~ WITHAIORCPs I ~ ~
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I 1ooo MODE 5 I ~ I I I 4 t I I I (350, 1000)
I 0
0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 REQUIRED RCS BORON CONCENTRATION (ppm)
Modes 3, 4, and 5/Drained Applicable to Mode 4, with or without RCPs in operation
Sheet 10 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 Figure 2 Rod Group Insertion Limits Versus Thermal Power (Three-Loop Operation) 240 r
iii 220 ~ I I ~ 41
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0 0 0.1 0.2 043 044 0.5 0.6 0.7 0.8 049 1 FRACTlON OF RATED THERMAL POWER (Fully withdrawn shall be 231 steps)
Note: Control Banks A and B must be withdrawn from the core prior to power operation.
Sheet 11 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 Figure 3 Axial Flux Difference Limits as a Function of Rated Thermal Power 120
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \~~~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~
- I 110 II \ I ~ ~ ~ ~ II ~ ~ \~ % ~ ~ '
~ I ' ~ I ~ ~ ~ ~ ~
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100
(-10, 1QO) (7, iso): ~ I ~ ~
I
~ ~ I ~ ~ ~ ~ ~ ~ ~ ~
I 'I UNACCfPTABLg I UNACCEPTABLE.
90 I~ I ~1 ~ ~ ~ ~ ~ ~ I ~ ~ 1%
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
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70 I~ 'I ~ ~ I ~ ~ ~ ~ I~ ~ ~ ~ 'I ~ ~ % I\ ~ I ~ I ~ ~ ~ ~ \ ~ I I~ ~ I I ~ \ ~ ~ ~ I ~ ~ ~ ~ ~ ~ I~
IJJ x ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
ACCEPTABLE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~
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60 I~ ~% \ ~ ~ I ~ I ~ I~ I I~ I I~ ~ I
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LL 50 ~
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(28..:50)
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- a. 30 I ~ I~ 1% 0 I I~ ~I ~ I I~ ~ ~ \ I ~ ~ ~ ~ I~ ~ ( I ~ ~ II~ % ~ ~ ~ I ~ ~ ~ ~ ~ ~ I ~ ~ ~ I~ ~~ ~ I~ ~ ~ ~ 'I ~~ ~ I' ~ 8%
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0
-50 -40 -30 -20 -10 0 10 20 30 40 50 AXIAL FLUX DIFFERENCE (% DELTA I)
(DEVIATION FROM TARGET AFD)
Note: At power levels less than HFP, the deviation is applied to the target AFD appropriate to that power level. The target AFD varies linearly between the HFP target and zero at zero power.
Sheet 12 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 Figure 4 K(Z) - Local Axial Penalty Function for FQ(Z) 1.2
'I ~ ~
I~ ~ I~ ~ I~ 4 ' ~ I ~ ~ ~ I ~ I~ ~ I I~ ~ I~ J I ~ ~ I~ ~ ~ ~ ~ I~ ~ I~ ~ ~ ~ ~ ~
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~ ~ ~ ~ ~ ~ I
- (0:0,3.0) (6.0; 1.0)
(12,0, 04925)
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K hC 0.7 ~ I I ~ 1 ~ ~ ~ I~ ~ ~ '1 tl \ ~ ~ ~ ~ ~ 1 ~ ~ I I I~ ~ 4%\ ~ ~ 4% ~ I ~ I ~ I~ ~ ~ ~ I ~ ~ 1 ~ \~I I~ I ~ ~ ~ ~ \ I I ~ ~ I 0 I O ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'I ~ ~ ~ ~ '
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LIJ 0.6 I~ ~ I~ ~ ~ ~ ~ II ~ JI ~ ~ ~ I ~ ~ I ~ ~ ~ ~ ~ ~ I ~ I ~ ~ ~ ~ ~ >\ ~ ~ ~ ~ I ~ I ~ ~~ ~ ~ I~ ~ ~ ~ ~ ~ I ~ JI ~ ~ I ~ ~ 14 ~ I I
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0.2 0 1 2 3 4 5 6 7 8 9 10 11 12 CORE HEIGHT (Feet)
Sheet 13 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 3 Figure 5 V(Z) Versus Core Height 1.260 Height (feet) V(Z>
I I I I I ~ I I I I I I I I I I I ~ I I I ~ ~
0.000 1.000 I ~ I I I ~ I I I I ~ I I I ~ ~ I ~ ~ ~ ~ ~ ~
0.200 1.000
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~ I ~ ~ I I I ~ I ~ I ~ ~ ~ ~ ~ ~ I ~ I I I I 10.200 1.112
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