L-PI-08-014, Core Operating Limits Report (COLR) for Prairie Island, Unit 1 Cycle 24, Revision 2 and Unit 2 Cycle 24, Revision 3
| ML080720368 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/07/2008 |
| From: | Wadley M Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-08-014 | |
| Download: ML080720368 (58) | |
Text
N Cornmined to Nod Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant
[PINGP) Unit 1 Cvcle 24, Revision 2 and Unit 2 Cvcle 24, Revision 3 Pursuant to PINGP Technical Specification 5.6.5.d, the Nuclear Management Company LLC (NMC) is submitting the COLR for PINGP Unit 1, Cycle 24, Revision 2 and Unit 2, Cycle 24, Revision 3.
The enclosed revisions to the Unit 1 and 2, Cycle 24, COLRs correct errors between COLR Table 1 and the shutdown boron concentrations used to meet the Table 1 requirements, identified on 1/29/08. The Unit 1 Cycle 24 revision also corrects an error in Table 1 where shutdown margins (SDMs) reported were non-conservative to the boron dilution accident, identified on 2/1/08. The first error resulted in the shutdown boron concentrations (in ppm) given to site Operations by the vendor not always meeting the shutdown margin percentages given in Table 1 of both COLRs. PINGP Technical Specification 3.1 . I requires that the shutdown margins specified in the COLR be met. The modifications to the COLRs align the COLRs with the vendor supplied shutdown boron concentrations (in ppm). The second error impacts the Unit 1 Cycle 24 COLR only. Three of the shutdown margin percentages reported in Table 1 were non-conservative to the boron dilution accident 24 minute requirement. The shutdown margin percentages reported in Table 1 are for the most reactive time in the cycle. The shutdown boron concentrations (in ppm) given to site Operations by the vendor that correspond to the most reactive time in the Unit 1 Cycle 24 (COLR SDMs) were conservative to the boron dilution accident.
As a result the following changes have been made to the Unit 1 and Unit 2, Cycle 24 COLRs:
Revised Section 3.1.1 Shutdown Margin Requirements, to include a reference to Figures 6A through 6H.
Revised Section 3.1.8 Physics Test Exceptions - MODE 2, to include a reference to Figure 6B.
Revised item c) of Section 3.9.1 Refueling Boron Concentration, to include a reference to Figures 6A through 6H.
Updated Table 1 to reflect the correct Shutdown Margin Requirements.
Added Figures 6A through 6H.
1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
Document Control Desk Page 2 Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
V Michael D. Wadley Site Vice President, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota
ENCLOSURE I PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNlT 1 - CYCLE 24, REVISION 2 UNlT 2 - CYCLE 24, REVISION 3 Record of Revision (5 pages)
Unit 1 - Cycle 24, Revision 2 (25 pages)
Unit 2 - Cycle 24, Revision 3 (25 pages)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Revision Approval Unit Cvcle No. Date Remarks 3/22/90 Original Unit 2 Core Operating Limits Report, distributed with Technical Specification Revision 92.
3/22/90 Original Unit 1 Core Operating Limits Report, distributed with Technical Specification Revision 92.
7/27/90 lncorporated expanded V(z) curves.
9/27/90 Clarified rod insertion limit curve applicability.
211 1/91 lncorporated revised Fa of 2.45 as a result of NRC approval of Westinghouse Topical Report WCAP-10924-P-A, Volume 1, Addendum 4, October 1990.
- Not used.
9/27/90 Updated to Unit 2 Cycle 14, incorporated expanded V(z) curves and clarified rod insertion limit curve applicability.
211 1/91 lncorporated revised Fa of 2.45 as a result of NRC approval of Westinghouse Topical Report WCAP-10924-P-A, Volume 1, Addendum 4, October 1990.
6/25/91 Updated to Unit 1 Cycle 15.
3/9/92 Updated to Unit 2 Cycle 15 and clarified labeling of Figure 4. Clarified the actions to be taken if the nuclear enthalpy rise hot channel factor exceeds the Technical Specification limit.
12/28/92 Updated to Unit 1 Cycle 16, removed V(z) curves and replaced them with list of bounding V(z) values for three ranges of exposures.
12/8/93 Updated to Unit 2 Cycle 16. Removed the multiple V(z) curves and replaced them with a single figure with bounding V(z) curves for four ranges of exposures. lncorporated additional discussion related to V(z) and K(z).
Page 1 of 5
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Revision Approval Unit Cycle No. Date Remarks 2 16 1 11/3/94 The table containing the bounding V(z) values and Figure 2 updated to incorporate revised bounding V(z) values for the exposure range of 14-21.5 GWDIMTU. Figures 3 through 6 re-formatted.
17 0 6117/94 Updated to Unit 1 Cycle 17. Removed the list of bounding V(z) values and replaced it with multiple V(z) curves. lncorporated additional discussion related to V(z) and K(z).
0 6/2/95 Updated to Unit 2 Cycle 17. Incorporated Table 1 and expanded Figure 2 with updated bounding V(z) values.
0 2/7/96 Updated to Unit 1 Cycle 18. Incorporated revised F, limit of 1.77. lncorporated Table 1 and updated Figure 2 with revised bounding V(z) values.
0 2127197 Updated to Unit 2 Cycle 18. Revised F , limit to 1.77. Updated Table 1 and Figures 2a through 2e with revised bounding V(z) values.
lncorporated new Figures 2f and 29 with additional bounding V(z) values.
0 9/25/97 Updated to Unit 1 Cycle 19. Updated Table 1 and Figures 2a through 2f with revised bounding V(z) values.
0 12/17/98 Updated to Unit 2 Cycle 19. Updated Table 1 and Figures 2a through 2d with revised bounding V(z) values. Deleted Figures 2e, 2f and 29.
0 5/13/99 Updated to Unit 1 Cycle 20. Updated Table 1 and Figures 2a through 2f with revised bounding V(z) values.
1 8/4/00 Technical Specification Amendment 151:
Relocate shutdown margin (SDM) requirements from Tech Specs and incorporate additional SDM requirements for Modes 3-6 from revised analysis of Uncontrolled Dilution event.
Page 2 of 5
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Revision Approval Unit Cvcle No. Date Remarks 5/31/00 Updated to Unit 2 Cycle 20. Updated Table 1 and Figures 2a through 2d with revised bounding V(z) values. Added new Table 2 and Figures 2e, 2f and 29 with additional bounding V(z) values.
Added references to Tables 1 and 2 and to Figures 2e, 2f and 2g to discussion of heat flux hot channel factor limits. Added discussion clarifying applicability of axial flux difference limits when using Tables 1 and 2 and Figures 2a through 2g. Added discussion of two tier V(z) curve presented in Table 2 and Figure 29.
8/4/00 Technical Specification Amendment 142:
Relocate shutdown margin (SDM) requirements from Tech Specs and incorporate additional SDM requirements for Modes 3-6 from revised analysis of Uncontrolled Dilution event.
9/1/00 Revised to change axial flux difference target band.
1131101 Updated to support refueling activities associated with Unit 1 Cycle 21. Revision 0 of the Unit 1 Cycle 21 COLR had to be issued prior to confirming the applicability of the LOCA analysis.
Therefore, Revision 0 of the Unit 1 Cycle 21 COLR does not contain all of the operating limits necessary to support operation of Unit 1 Cycle 21.
2119/01 Updated to Unit 1 Cycle 21. Updated Tables 1 and 2 and Figures 2a through 2f with revised bounding V(z) values.
10/02/02 Revised to support License Amendment 158 changes, including revision of all references to TS, revision of FQsymbols, addition of Table 4, ITC limits, DNB limits and refueling boron concentrations.
2/06/02 Updated to Unit 2 Cycle 21.
1 10/02/02 Revised to support License Amendment 149 Page 3 of 5
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Revision Approval Unit Cycle No. Date Remarks changes, including revision of all references to TS, revision of Fa symbols, addition of Table 4, ITC limits, DNB limits and refueling boron concentrations. Also revised to include an additional V(z) curve to give greater FQmargin between 13.0 and 16.0 GWdIMTU.
11/25/02 Updated to Unit 1 Cycle 22. Updated Tables 1 and 2 and Figures 2a through 2f with revised bounding V(z) values. Incorporated new Figure 29 with additional bounding V(z) values.
Updated Table 3 with revised minimum shutdown margin limits. Deleted and revised text to eliminate duplication with the Technical Specifications and the Bases.
0 9119/03 Updated to Unit 2 Cycle 22. Updated Tables 1 and 2. A reduced number of exposure ranges were calculated in Table 1, therefore new Figures 2a through 2e with revised bounding V(z) values replaced Figures 2a through 2f. New Figure 2f replaced Figure 2g for the 2 tier band bounding V(z) values. Updated Table 3 with revised minimum shutdown margin limits. Deleted and revised text to eliminate duplication with the Technical Specifications and the Bases.
1 7/6/04 Revision to incorporate Westinghouse Safety Analysis Transition per LA 1621153. Revision 1 contains transitional values for the OPIOT AT Trip setpoints that will be used while the physical changes are implemented.
1 7/6/04 Revision to incorporate Westinghouse Safety Analysis transition per LA 1621153. Revision 1 contains transitional values for the OPIOT AT Trip setpoints that will be used while the physical changes are implemented.
22 2 7/12/04 Revised Fq limit from 2.4 to 2.5. Removed OP and OT delta-T setpoints based on NMC methodology and replaced with Westinghouse developed setpoints.
22 2 7/16/04 Revised Fq limit form 2.4 to 2.5. Removed OP and OT delta-T setpoints based on NMC Page 4 of 5
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Core Operating Limits Report Record of Revision Revision Approval Unit Cvcle No. Date Remarks methodology and replaced with Westinghouse developed setpoints.
10/20/04 Updated to Unit 1 Cycle 23.
Not used due to core redesign.
5/19/05 Updated to Unit 2 Cycle 23 and to support redesign of Unit 2 Cycle 23 core.
7111105 Revised ITC upper limit from < 0 pcmIoFfor power levels > 70% RTP to less than a line that slopes linearly from 0 pcm/"F at 70% RTP to -2.9 pcm/"F at 100% RTP. Revised the title of Figure 3 to reference T.S. 3.1.4 Condition B and revised the title of Figure 4 to reference T.S. 3.1.4 Condition A. Added references 24 and 25 to include the 50.59 screenings written to issue revision 1.
5/10/06 Updated to Unit 1 Cycle 24.
8/7/06 Updated Table 3 to reflect the correct FqW(z) penalty factors.
11/26/06 Updated to Unit 2 Cycle 24 Modes 5 and 6.
12/6/06 Updated to Unit 2 Cycle 24 for Modes 1-6.
914107 Revised to support LA-1791169. Revised reference 24 to include the revision number (revision 0) and the correct date of the report (January 2005). Revised references 6a, 6b, 6c, and 8 to say 'Deleted.' These references referred to the old LBLOCA methodology and model.
211 1/08 Updated Table 1 to reflect correct Shutdown Margin Requirements and added Figures 6A through 6H.
211 1/08 Updated Table 1 to reflect correct Shutdown Margin Requirements and added Figures 6A through 6H.
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Core Operating Limits Report Unit 1, Cycle 24 Revision 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 - CYCLE 24 REVISION 2 Reviewed By: Date: a-/*/d
/-
Supervisor, NSSS Date:
Supervisor, PWR Analyses Approved By: i &CbgL 1
Mike q'arlson Date:
Director, Site Engineering Note: This report is not part of the Technical Specifications This report is referenced in the Technical Specifications Page 1 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 - CYCLE 24 REVISION 2 This report provides the values of the limits for Unit 1 Cycle 24 as required by Technical Specification Section 5.6.5. These values have been established using NRC approved methodology and are established such that all applicable limits of the plant safety analysis are met. The Teclmical Specifications affected by this report are listed below:
Reactor Core SLs Shutdown Margin (SDM)
Isothermal Temperature Coefficient (ITC)
Shutdown Bank Insertion Limits Control Bank Insertion Limits Physics Tests Exceptions - MODE 2 Heat Flux Hot Channel Factor (FQ(z))
Nuclear Enthalpy Rise Hot Channel Factor (F,~)
Axial Flux Difference (AFD)
Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1- 1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Boron Concentration 2.1.1 Reactor Core Safety Limits Reactor Core Safety Limits are shown in Figure 1.
Reference Technical Specification section 2.1.1.
- 2. 3.1.1 Shutdown Margin Requirements Minimum Shutdown Margin requirements are shown in Table 1 and Figures 6A through 6H.
Reference Technical Specification section 3.1.1.
Page 2 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2
- 3. 3.1.3 Isothermal Temperature Coefficient (ITC')
ITC Upper limit:
- a. < 5 p c d ° F for power levels < 70% RTP; and
- b. less than a line which slopes linearly from i) 0 pcmI0F at power level = 70% RTP to ii) -1.5pcrnPF at power level = 100% RTP ITC Lower limit:
- a. -32.7 pcmI0F Reference Technical Specification section 3.1.3.
- 4. 3.1.5 Shutdown Bank Insertion Limits The shutdown rods shall be fully withdrawn.
Reference Technical Specification section 3.1.5.
- 5. 3.1.6 Control Bank Insertion Limits The control rod banks shall be limited in physical insertion as shown in Figures 2, 3, and 4.
The control rod banks withdrawal sequence shall be Bank A, Bank B, Bank C, and finally Bank D.
The control rod banks shall be withdrawn maintaining 128 step tip-to-tip distance.
Reference Technical Specification section 3.1.6.
- 6. 3.1.8 Physics Tests Exceptions - MODE 2 Minimum Shutdown Margin requirements during physics testing are shown in Table 1 and Figure 6B.
Reference Technical Specification section 3.1.8.
Page 3 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2
- 7. 3.2.1 Heat Flux Hot Channel Factor (Fa.(Z21 The Heat Flux Hot Channel Factor shall be within the following limits:
CFQ = 2.50 K(Z) is a constant value = 1.0 at all elevations.
W(Z) values are provided in Table 2.
F ~ ~ ( Penalty z) Factors are provided in Table 3.
Applicability: MODE 1.
Reference Technical Specification section 3.2.1
- 8. 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F& -
The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:
FAHI 1.77 x [ l + 0.3(1- P)]
where: P is the fraction of RATED THERMAL POWER at which the core is operating.
Applicability: MODE 1.
Reference Technical Specification section 3.2.2
- 9. 3.2.3 Axial Flux Difference (AFD)
The indicated axial flux difference, in % flux difference units, shall be maintained within the allowed operational space defined by Figure 5.
Applicability: MODE 1 with RATED THERMAL POWER 2 50% RTP Reference Technical Specification sections 3.2.3.
Page 4 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2
- 10. 3.3.1 Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1- 1; Overtemperature AT Setpoint Overtemperature AT setpoint parameter values:
Indicated AT at RATED THERMAL POWER, %
Average temperature, OF 560.0 OF Pressurizer Pressure, psig 2235 psig 1.17 0.014 1°F 0.00100 /psi 30 seconds T2 = 4 seconds
= A function of the indicated difference between top and bottom f(A1) detectors of the power range nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where qt and qt, are the percent power in the top and bottom halves of the core respectively, and q, + q b is total core power in percent of RATED THERMAL POWER, such that (a) For qt - qb within -13, +8 % f(A1) = 0 (b) For each percent that the magnitude of qt - qb exceeds +8%
the AT trip setpoint shall be automatically reduced by an equivalent of 1.73 % of RATED THERMAL POWER.
(c) For each percent that the magnitude of qt - qb exceeds -13 %
the AT trip setpoint shall be automatically reduced by an equivalent of 3.846 % of RATED THERMAL POWER.
Overpower AT Setpoint Overpower AT setpoint parameter values:
ATo = Indicated AT at RATED THERMAL POWER, %
T = Average temperature, OF T' = 560.0 OF I(4 51.11 ,
K5 = 0.0275I0F for increasing T; 0 for decreasing T Kg = 0.002I0F for T > T' ; 0 for T I T' x3 = 10 seconds Page 5 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 1 1. 3.4.1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Pressurizer pressure limit = 2205 psia RCS average temperature limit = 564OF RCS total flow rate limit = 178,000 gpm Reference Technical Specification section 3.4.1.
- 12. 3.9.1 Refueling Boron Concentration.
The boron concentration of the reactor coolant system and the refueling cavity shall be sufficient to ensure that the more restrictive of the following conditions is met:
a) Gff5 0.95 b) 2000 ppm c) The Shutdown Margin specified in Table 1 and Figures 6A through 6H Reference Technical Specification section 3.9.1.
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Core Operating Limits Report Unit 1, Cycle 24 Revision 2 REFERENCES NSPNAD-8 101-A, "Qualification of Reactor Physics Methods for Application to Prairie Island," Revision 2, October 2000.
NSPNAD-8 102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units," Revision 7, July 1999.
NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology," Revision 1, October 2000.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July, 1985.
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August, 1985.
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," Addendum 2 Revision 1, July 1997.
WCAP- 10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 1,2,3, December 1988.
WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 2 , Volume 2 Addendum 1, December 1988.
WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 4, March 1991.
XN-NF-77-57-(A), XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase 11," May 1981.
WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report: W-COBRAITRAC 2-Loop Upper Plenum Injection Model Update to Support zIRL,oTM Cladding Options,"
February 1994.
NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology," Revision 0, April 1993.
WCAP- 10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification," February 1994.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.
WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989.
Page 7 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report,"
January 1999.
WCAP-7588 Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.
WCAP-7908-A, "FACTRAN -A FORTRAN IV Code for Thermal Transients in a UOz Fuel Rod," December 1989.
WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
WCAP-7979-P-A, TWINKLE - A Multidimensional Neutron Kinetics Computer Code," January 1975.
WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"
December 1985.
WCAP-11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.
WCAP- 11596-P-A, "Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
WCAP- 12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.
WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.
50.59 Evaluation 1054, "Unit 1 Cycle 24 Core Reload."
Page 8 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Table 1 Minimum Required Shutdown Margin, %Ap 0 MWdIMTU 2.5 4.5 7.0 12000 MWdIMTU 2.0 3.5 5.0 23000 MWdIMTU 2.0 2.0 2.5 Operational Mode Definitions, as per TS Table 1.I-1.
For Mode 1 and Mode 2 with Keff2 1.O, the minimum shutdown margin requirements are provided by the Rod Insertion Limits.
Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service column.
Page 9 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Table 1, Continued Minimum Required Shutdown Margin, %Ap 12000 MWdIMTU 5.129 5.5 8.5 23000 MWdlMTU 5.129 5.129 5.5 Operational Mode Definitions, as per TS Table 1.?-I.
- Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service column.
These values are also applicable for the Unit 1 Cycle 23 end of cycle.
Page 10 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Page 11 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Table 3 F ~ ~ ( Penalty z) Factor Cycle Burnup (MWDIMTU) F ~ ~ ( Penalty z) Factor 0 1.020 13281 1.020 13432 1.023 13583 1.027 13734 1.027 13885 1.027 14036 1.025 F ~ ~ (=z1.020
) for all burnups except those listed above. Linear interpolation is adequate for intermediate cycle burnups.
Page 12 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 1 Reactor Core Safety Limits Fraction of Rated Thermal Power Page 13 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:
Bank C = (150 / 63) * (P - 100) + 185 + 128 NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page 14 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4 Condition B) 0 20 40 60 80 100 Power Level. % of Rated Thermal Power Bank Positions Given By:
Bank C = (150 / 63) * (P - 90) + 224 + 128 NOTE: The top o f the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page 15 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4 Condition A)
I I
I I
I Operation I Forbidden in I t h ~ sReg~on I
I 1
I I
I I
I I
I I
I I
I I
I I
I I
0 0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:
Bank C = (150163) * (P-70) + 224 + 128 NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page 16 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 5 Flux Difference Operating Envelope Page 17 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6A Minimum Required Shutdown Margin Mode 2 I0000 15000 Cycle Burnup (MWdlMTU)
Page 18 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6B Minimum Required Shutdown Margin Physics Testing in Mode 2 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
Page 19 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6C Minimum Required Shutdown Margin Mode 3, Tave 2 520°F (N-I) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except for the most reactive rod is out.
Page 20 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6D Minimum Required Shutdown Margin Mode 3,350°F ITave q 520°F (N-I) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except for the most reactive rod is out.
Page 21 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6E Minimum Required Shutdown Margin Mode 4,200°F < Tave < 350°F (N-I) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except for the most reactive rod is out.
Page 22 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6F Minimum Required Shutdown Margin Mode 5, 68'F I Tave I200°F (N-I)
I 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except for the most reactive rod is out.
Page 23 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6G Minimum Required Shutdown Margin Mode 6,68"F I Tave < 200°F (ARI) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
Page 24 of 25
Core Operating Limits Report Unit 1, Cycle 24 Revision 2 Figure 6H Minimum Required Shutdown Margin Mode 6,68"F 2 Tave < 200°F (ARO) 0 5000 I0000 15000 20000 25000 Cycle Bumup (MWdlMTU)
Page 25 of 25
Core Operating Limits Report Unit 2 , Cycle 24 Revision 3 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 2 - CYCLE 24 REVISION 3 Reviewed By: c3zk.A 5clbL&d Date: ~ / g / n %
Lynn a h s o n Supervisor, NSSS Keviewed By:
Hoelscher Date: 2b(/0r Supervisor, PWR Analyses Approved By: Date:
Director, Site Engineering Note: This report is not part of the Technical Specifications This report is referenced in the Technical Specifications Page 1 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 2 - CYCLE 24 REVISION 3 This report provides the values of the limits for Unit 2 Cycle 24 as required by Technical Specification Section 5.6.5. These values have been established using NRC approved methodology and are established such that all applicable limits of the plant safety analysis are met. The Technical Specifications affected by this report are listed below:
Reactor Core SLs Shutdown Margin (SDM)
Isothermal Temperature Coefficient (ITC)
Shutdown Bank Insertion Limits Control Bank Insertion Limits Physics Tests Exceptions - MODE 2 Heat Flux Hot Channel Factor (FQ(z))
Nuclear Enthalpy Rise Hot Channel Factor (F:~)
Axial Flux Difference (AFD)
Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1- 1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Boron Concentration
- 1. 2.1.1 Reactor Core Safety Limits Reactor Core Safety Limits are shown in Figure 1.
Reference Technical Specification section 2.1.1.
- 2. 3.1.1 Shutdown Margin Requirements Minimum Shutdown Margin requirements are shown in Table 1 and Figures 6A through 6H.
Reference Technical Specification section 3.1.1.
Page 2 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3
- 3. 3.1.3 Isothermal Temperature Coefficient (ITC) 1TC Upper limit:
- a. < 5 pcm/"F for power levels < 70% RTP; and
- b. less than a line which slopes linearly from i) 0 p c d ° F at power level = 70% RTP to ii) - 1.5pcd°F at power level = 100% RTP ITC Lower limit:
- a. -32.7 pcmJ0F Reference Technical Specification section 3.1.3.
- 4. 3.1.5 Shutdown Bank Insertion Limits The shutdown rods shall be fully withdrawn.
Reference Technical Specification section 3.1.5.
- 5. 3.1.6 Control Bank Insertion Limits The control rod banks shall be limited in physical insertion as shown in Figures 2, 3, and 4.
The control rod banks withdrawal sequence shall be Bank A, Bank B, Bank C, and finally Bank D.
The control rod banks shall be withdrawn maintaining 128 step tip-to-tip distance.
Reference Technical Specification section 3.1.6.
- 6. 3.1.8 Physics Tests Exceptions - MODE 2 Minimum Shutdown Margin requirements during physics testing are shown in Table 1 and Figure 6B.
Reference Technical Specification section 3.1.8.
Page 3 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 3.2.1 Heat Flux Hot Channel Factor ( F Q 0 The Heat Flux Hot Channel Factor shall be within the following limits:
CFQ = 2.50 K(Z) is a constant value = 1.0 at all elevations.
W(Z) values are provided in Table 2.
FWQ(z)Penalty Factors are provided in Table 3.
Applicability: MODE 1.
Reference Technical Specification section 3.2.1
- 8. 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor ( ~-~ H f f i The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:
FAH< 1.77 x [ l + 0.3(1- P)]
where: P is the fraction of RATED THERMAL POWER at which the core is operating.
Applicability: MODE 1 .
Reference Technical Specification section 3.2.2
- 9. 3.2.3 Axial Flux Difference (AFD]
The indicated axial flux difference, in % flux difference units, shall be maintained within the allowed operational space defined by Figure 5.
Applicability: MODE 1 with RATED THERMAL POWER 2 50% RTP.
Reference Technical Specification sections 3.2.3.
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Core Operating Limits Report Unit 2, Cycle 24 Revision 3
- 10. 3.3.1 Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1- 1; Overtemperature AT Setpoint Overtemperature AT setpoint parameter values:
Indicated AT at RATED THERMAL POWER, %
Average temperature, OF 560.0 OF Pressurizer Pressure, psig 2235 psig 1.17 0.014 /OF 0.00100 /psi 30 seconds 4 seconds A function of the indicated difference between top and bottom detectors of the power range nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and q, + qb is total core power in percent of RATED THERMAL POWER, such that (a) For qt - qb within -13, +8 % f(A1) = 0 (b) For each percent that the magnitude of qt - q b exceeds +8%
the AT trip setpoint shall be automatically reduced by an equivalent of 1.73 % of RATED THERMAL POWER.
(c) For each percent that the magnitude of q, - qb exceeds -13 %
the AT trip setpoint shall be automatically reduced by an equivalent of 3.846 % of RATED THERMAL POWER.
Ovel-power AT Setpoint Overpower AT setpoint parameter values:
ATO = Indicated AT at RATED THERMAL POWER, %
T = Average temperature, OF T' = 560.0 OF K, 5 1.11 K5 = 0.0275I0Ffor increasing T; 0 for decreasing T I(6 = 0.002/OF for T > T' ; 0 for T I T' 73 = 10 seconds Page 5 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3
- 11. 3.4.1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Pressurizer pressure limit = 2205 psia RCS average temperature limit = 564OF RCS total flow rate limit = 178,000 gpm Reference Technical Specification section 3.4.1.
- 12. 3.9.1 Refueling Boron Concentration.
The boron concentration of the reactor coolant system and the refueling cavity shall be sufficient to ensure that the more restrictive of the following conditions is met:
a) I&ff 10.95 b) 2000ppm c) The Shutdown Margin specified in Table 1 and Figures 6A through 6H Reference Technical Specification section 3.9.1 Page 6 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 REFERENCES NSPNAD-8 101-A, "Qualification of Reactor Physics Methods for Application to Prairie Island," Revision 2, October 2000.
NSPNAD-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety EvaIuation Methods for Application to PI Units," Revision 7, July 1999.
NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology," Revision 1, October 2000.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July, 1985.
WCAP- 10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August, 1985.
WCAP- 10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," Addendum 2 Revision 1, July 1997.
Deleted.
Deleted.
Deleted.
XN-NF-77-57-(A), XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase 11," May 1981.
Deleted.
NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology," Revision 0, April 1993.
WCAP-10216-P-A, Revision lA, "Relaxation of Constant Axial Offset ControW FQ Surveillance Technical Specification," February 1994.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.
WCAP- 11397-P-A, "Revised Thermal Design Procedure," April 1989.
WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report,"
January 1999.
WCAP-7588 Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.
Page 7 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 WCAP-7908-A, "FACTRAN -A FORTRAN N Code for Thermal Transients in a U02 Fuel Rod," December 1989.
WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
WCAP-7979-P-A, "TWINKLE - A Multidimensional Neutron Kinetics Computer Code," January 1975.
WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"
December 1985.
WCAP- 11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.
WCAP- 11596-P-A, "Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
WCAP- 12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.
WCAP- 14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.
WCAP- 16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Revision 0, January 2005.
50.59 Evaluation 1055, "Unit 2 Cycle 24 Core Reload."
Page 8 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Table 1 Minimum Required Shutdown Margin, %Ap 0 MWdIMTU 2.0 4.5 6.5 12000 MWdlMTU 2.0 3.5 5.5 1 24700 MWdIMTU I 2.0 2.0 2.5 Operational Mode Definitions, as per TS Table 1.l-1.
- For Mode 1 and Mode 2 with kff r 1.O, the minimum shutdown margin requirements are provided by the Rod Insertion Limits.
- Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service column.
Page 9 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Table 1, Continued Minimum Required Shutdown Margin, %Ap 0 MWdlMTU*** 5.129 6.5 10.0 12000 MWdlMTU 5.129 5.5 8.5 24700 MWdlMTU 5.129 5.129 5.5 Operational Mode Definitions, as per TS Table 1.l-1.
- Charging pump(s) in service only pertains to steady state operations. It does not include transitory operations. For example, operations such as starting a second charging pump in order to secure the operating pump would fall under the one pump in service column.
- These values are also applicable for the Unit 2 Cycle 23 end of cycle.
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Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Page 11 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Table 3 F ~ ~ ( Penalty z) Factor Cycle Burnup (MWDIMTU) F ~ ~ ( Penalty z) Factor 0 1.020 F ~ ~ (=z1.020
) for all burnups except those listed above. Linear interpolation is adequate for intermediate cycle burnups.
Page 12 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 1 Reactor Core Safety Limits 0 0.2 0.4 0.6 0.8 1 1.2 1.4 Fraction of Rated Thermal Power Page 13 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:
Bank C = (1501 63) * (P- loo)+ 185 + 128 Bank B = (150 / 63) * (P - 100) + 185 + 128 + 128 NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page I 4 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4 Condition B)
Page 15 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4 Condition A)
Power Level, % of Rated Thermal Power Bank Positions Given By:
NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page 16 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 5 Flux Difference Operating Envelope Page 17 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6A Minimum Required Shutdown Margin Mode 2 0 5000 I0000 15000 20000 25000 Cycle Burnup (MWdlMTU)
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Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6B Minimum Required Shutdown Margin Physics Testing in Mode 2 10000 15000 Cycle Burnup (MWdlMTU)
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Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6C Minimum Required Shutdown Margin Mode 3, Tave I520°F (N-I)
I 0 5000 I0000 15000 20000 25000 I
Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except the most reactive rod is out.
Page 20 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6D Minimum Required Shutdown Margin Mode 3, 350°F 5 Tave < 520°F (N-I) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except the most reactive rod is out.
Page 21 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6E Minimum Required Shutdown Margin Mode 4,200°F< Tave < 350°F (N-I) 0 5000 I0000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except the most reactive rod is out.
Page 22 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6F Minimum Required Shutdown Margin Mode 5, 68'F ITave I200°F (N-I) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
N-1 means all the rods are in except the most reactive rod is out.
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Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6G Minimum Required Shutdown Margin Mode 6,68"FI Tave < 200°F (ARI) 10000 15000 Cycle Burnup (MWdlMTU)
Page 24 of 25
Core Operating Limits Report Unit 2, Cycle 24 Revision 3 Figure 6H Minimum Required Shutdown Margin Mode 6,68"FI Tave < 200°F (ARO) 0 5000 10000 15000 20000 25000 Cycle Burnup (MWdlMTU)
Page 25 of 25