ML18012A523

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Cycle 7 Colr,Rev 4.
ML18012A523
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 02/19/1997
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18012A522 List:
References
NUDOCS 9702260347
Download: ML18012A523 (15)
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Text

Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 Attachment 9 to HNP Procedure PLP-106, "Technical Specification Equipment List Program and Core Operating Limits Report," Rev. 12 9702260347 970219 fI f .

PDR P

ADQCK 05000400 PDRQ

Attachment 9 Sheet 1 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 IMPORTANT NOTE ASSOCIATED WITH H1C7 COLR, REV. 4:

The requirements listed in this revision and located in the following sections are valid for the remainder of Cycle 7 only:

Peaking factors limits (Sections 2.6 and 2.7)

Minimum number of detector thimbles (Section 4.1.1.b)

Minimum number of detector thimbles per quadrant (Section 4.1.1.c)

Regular requirements associated with the above sections shall be reinstated in the H1C8 COLR.

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 Enthalpy Rise Hot Channel Factor F<<

3/4.9.1.a Boron Concentration During Refueling Operations PLP-106 Rev. 12 Page 67 of 82

Attachment 9 Sheet 2 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 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 W (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. l. 3)

1. 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.

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 THERMAL POWER RTP stands for RATED THERMAL POWER 2.3 w (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 AXIAL FLUX DIFFERENCE (AFD) target band is specified in Figure 3.

PLP-106 Rev. 12 Page 68 of 82

Attachment 9 Sheet 3 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4

- F~(Z) (Specification 3/4.2.2)

FQ(Z) < F ""

  • K(Z) /P for P > 0. 5 FQ(Z) < F"
  • K(Z) /0 5 for

~ P < 0 5 ~

where: P = THERMAL POWER/RATED THERMAL POWER

a. F~" '.40 for LOPAR and VANTAGE 5 fuel
b. F" 2.47 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 F<< (Specification 3/4.2.3)

F.. < F."" * (1 + PF<<* (1 - P) )

where: P = THERMAL POWER/RATED THERMAL POWER a F, "" = 1.58 for LOPAR fuel F<<" '.61 for VANTAGE 5 fuel F " 1. 69 for SPC fuel PF<< 0.3 for LOPAR fuel PF<< ~ 0.35 for VANTAGE 5 and SPC fuel 2.8 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.

PLP-106 Rev. 12 Page 69 of 82

Attachment 9 Sheet 4 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4 3.0 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 F 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 3.2.1 Axial Flux =Difference, 3.2.2 - Heat Flux Hot Insertion'imits, 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).

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).

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).

PLP-106 Rev. 12 Page 70 of 82

Attachment 9 Sheet 5 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 3.0 R E N (continued)

EXEM PWR Large Break LOCA Evaluation Model as defined by:

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-78-44(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).

t 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).

PLP-106 Rev. 12 Page 71 of 82

p Attachment 9 Sheet 6 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 3.0 (continued)

9. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary).

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - Modes 3, 4, and 5, 3.2 ' - Heat Flux Hot Channel Factor, and 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor).

E

10. WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS .

EVALUATION MODEL USING THE BASH CODE," March 1987 (W 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 (g Proprietary).

(Methodology for Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor).

12. EMF-92-081(A), and Supplement 1, "Statistical Setpoint/Transient 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-82-49(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).

PLP-106 Rev. '12 Page 72 of 82

Attachment 9 Sheet 7 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4, 4.0 4.1 v 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,

b. A minimum of 25 detector thimbles for the remainder of the

.operating cycle, C. A minimum of three detector thimbles per core qu'adrant, and

d. Sufficient movable detectors, drive, and readout equipment to map these thimbles.

When the Movable Incore Detection System is used for:

a ~ Recalibration of the Excore Neutron 'Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or
c. Measurement of F, and F~(Z)

V The Movable Encore 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 TILT RATIO, or
c. Measurement of F, and F~(Z)

The OPERABILITY of 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.

PLP-106 Rev. 12 Page 73 of 82

Attachment 9

, Sheet 8 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4 N (continued)

R 5.

In order to change the requirements concerning the number and location of operable detectors, the NRC staff deems that a rigorous evaluation and justification's required. The following is a list of elements that must be part of a 50.59 determination and available for audit requirements:

if the licensee wishes to change the How an inadvertent loading of a fuel assembly into an improper location will be detected,

b. How the validity of the tilt estimates will be ensured, How adequate core coverage will be maintained, How the measurement uncertainties will be 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 limits, and tilt factor meet Technical Specification
e. How the Movable Incore Detection System will be restored to full (or nearly full) service before the beginning of each cycle.

PLP"106 Rev. 12 Page 74 of 82

Attachment 9 Sheet 9 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4 Figure 1 Shutdown Margin Versus RCS Boron Concentration Modes 3, 4, and 5/Drained

  • Applicable to Mode 4, with or without RCPs in operation 8000 II t

7000 >

(2600,6400) 6000

'I E

MM~

4 WITH NO RCPs EL IN OPERATION, AND 4m MODE 5, I

&4'ODE I

V) sm) ll'2600,2600)

MODE3,AND MODE4 WITH AT LEAST ONE RCP IN OPERATION MODES 3 (1500,1770)

MODE 5 (350,1000) 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 REQUIRED RCS BORON CONCENTRATION (ppm)

PLP-106 Rev. 12 Page 75 of 82

l t~ ~

Attachment 9 Sheet 10 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 Figure 2 Rod Group Insertion Limits Versus Thermal Power (Three-Loop Operation) 240 220 200 ~ t ~ tt ~ t ~ I~ ~ I t~ I ~ ~ I ~ ~ ~ I t~ ~ ~ ~ ~ ~ I~ I ~ ~ ~ ~ t~ ~ ~

BANK C I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

160 ~ 'I I I ~

(i,-:>86)

I~

140 0)

(O,:.128)

~ I ~ ~ ~ ~

120 ~ I~ I O

M C) 100 CL BANK D hC 80 ~ t t ~ ~ ~ ~ ~ I ttt ~ I I~ ~ I ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ \t ~ ~ ~ ~ t ~ ~ t~~ \~ ~ ~ ~ ~

CQ

~ ~ ~ ~ ~

O, 60 I ~ ~ ~ ~ I ~ t ~

40 20

~ ~ I~ ~ ~ ~ ~ ~

0 0 0.1 0.2 0.3 0.4 0.5 0.6'.7 0.8 0.9 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.

PLP-106 Rev. 12 Page 76 of 82

Attachment 9 Sheet 11 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4 Figure 3 Axial Flux Difference Limits, as a Function of Rated Thermal Power 120 r

110 ~ ~ II ~ 1 ~ ~ ~ r ~ ~ ~ ~ ~ ~ ~ I\ ~ r~ ~

(-10, 1O0): (7, 100) 100

.'UNA(ACCEPTABLE::

90 UNACC/PTABLE r ~ r ~ ~ ~ ~ ~ ~

~

~ ~ ~ ~ ~ rr ~ ~ ~ ~ ~ ~ '

O Oo

~ r ~ ~ ~ ~ 'r ~

r 70 r ~ ~ ~ r ~ r > ~ r ~ ~ r ~ ~ ~ ~ r r ~ II ~ r ~ ~ I rr rrr ~ r ~ r ~ rr \ ~ r ~ ~

LLI x ACCEPTABLE 50):.:.:;::::.:.:

50 CL (-54. (20,.:50):.

50 O

40 I ~ ( ~ ~ rr ~ 1 ~ ~ ~ ~ r ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ r ~ ~ ~ q I ~ I~ ~ ~ ~ ~ ~ ~ rr ~ ~ r ~ ~ I ~ ~ ~ ~ ~ ~ r ~ I ~ I r ~ r ~ ~ rrr ~ > ~ ~ r ~ ~

30 20 ~ r ~ ~ ~ r ~

~ ~ ~ ~ ~

10 ~ ~ ~ p r ~

~ ~ ~

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.

PLP-106 Rev. 12 Page 77 of 82

Attachment 9 Sheet 12 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report Rev. 4 Figure 4 K(Z) - Local Axial Penalty Function for FQ(Z)

- 1.2

~ I rP ~ I~ ~

(6.0, 1.0)

'I (12.0, Q.925)

CL 0.9 O

O \\ ~

0.0 K ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

0.7 CL O I~ ~ ~ \~

LIJ 0.8

~ ~ ~ ~ ~

CL O

Z 0.5 I ~ ~ ~ ~ \ ~ 0 ~ I ~ 'I ~

0.4 . . ~;

'~ ~,.

~ ~

0.3 0.2 0 1 2 3 4 5 6 7 0 9 10 ll 12 CORE HEIGHT (Feet)

PLP-106 Rev. 12 Page 78 of 82

Attachment 9 Sheet 13 of 13 Harris Unit 1 Cycle 7 Core Operating Limits Report - Rev. 4 Figure 5 V(Z) Versus Core Height H eloh1 1.250 (10 e I) v(z) 0.000 1.000 0.200 1.000 0.400 1.000 0.600 1.000 I.225 0,800 1.000 1.000 1.000 1,200 1.000 1.400 1.000 1.600 1.000 1.800 1.172 I.200 2.000 1.167 2.200 1.182 2.400 1.155 2,600 1.149 2.800 1.142 I.I I 5 3,000 1.134 3.200 1.126 3,400 1.119

~ a e 3.800 1.123 3.800 1.130 4.000 1.135 1.150 4.200 1.139 4.400 1.143 4.800 1,150 4.800 1.158 5.000 IrIBI 5.200 1,185 1.12$

  • r 5.400 1.168 5.6oo 1.169 5.800 1,188 8.000 1.185 6.200 1,181 1.100 6.400 1.157 6.6oo 1.180 8.800 1.159 "r' " ~ - ' r \-. r - ~ r r ~ r" I r 7.000 1.158 7,200 1.158 7.400 1,153 I.07$ 7.6OO 1,149 7.800 1.143 8.000 1.137 8.200 1.131 8.400 1.126 6.6oo 1.122 I,050 r r ~ ~'
  • 8,800 1.115 9.000 1,109 9.200 1.103 9.400 1.100 9.800 1,100 I.025 ~ " ~ ' 4 r ~

9.800 1.103 h

10.000 '1,108 10.200 1.112 10.400 1.000 10.800 1.000 10.800 1.000 1.000 I 1.000 1.000 0 I 2 2 4 5 ~ 7 4 4 10 II 12 11.200 1.000 11.400 1.000 CORE HEIGHT (foot) 11.6OO 1.000 11.800 1.000 12.000 1.000 Note: Top and bottom 15% excluded as per Technical Specification 4.2.2.2.g Rev. 12 Page 79 of 82

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