ML052200386
| ML052200386 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/08/2005 |
| From: | Thomas J. Palmisano Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-Pl-05-071 | |
| Download: ML052200386 (20) | |
Text
Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC AUG 0 8 2005 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie lsland Nuclear Generating Plant Unit 1 Docket 50-282 License No. DPR-42 Core Operatina Limits Report (COLR) for Prairie lsland Unit 1 Cycle 23, Revision 1 Pursuant to the requirements of Technical Specification 5.6.5.d, the COLR for Prairie lsland Unit 1 Cycle 23, Revision 1 is attached. The limits specified in the attached COLR have been established using Nuclear Regulatory Commission (NRC) approved methodologies.
The Unit 1 COLR has been revised for Cycle 23 to incorporate the following changes:
Revised Isothermal Temperature Coefficient (ITC) upper limit from <O pcmI0Ffor power levels >70% Rated Thermal Power (RTP) to less than a line that slopes linearly from 0 pcmI0Fat 70% RTP to -2.9 pcmI0Fat 100% RTP.
Revised the title of Figure 3 to reference Technical Specification 3.1.4 Condition B.
Revised the title of Figure 4 to reference Technical Specification 3.1.4 Condition A.
Added references 24 and 25 to include the 50.59 screenings written to issue Revision I.
Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.
Thomas J. Palmisano Site Vice President, Prairie lsland Nuclear Generating Plant Nuclear Management Company, LLC Enclosure (1) 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
Document Control Desk Page 2 cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota, Commerce Department
ENCLOSURE 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 CYCLE 23 REVISION 1 17 pages follow
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 - CYCLE 23 REVISION 1 ReviewedBy: ' Date:
Jon Kapitz Supervisor, Nuclear ngineering Reviewed By: Date:
6P-/o s Ed Mercier Supervisor, PWR Analysis Director, Engineering Note: This report is not part of the Technical Specifications This report is referenced in the Technical Specifications Page Iof 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 - CYCLE 23 REVISION 1 This report provides the values of the limits for Unit 1 Cycle 23 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 Safetv 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.
Reference Technical Specification section 3.1.1.
Page 2 of 17
Core Operating Limits Report
.. Unit 1, Cycle 23 Revision 1
- 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
- 1. 0 pcm/"F at power level = 70% RTP to
- a. -32.7 p c d ° F 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.
Reference Technical Specification section 3.1.8.
Page 3 of 17
Core Operating Limits Report Unit I,Cycle 23 Revision 1
- 7. 3.2.1 Heat Flux Hot Channel Factor (FQm 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 (FAA -
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 (@D)
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 > 50% RTP.
Reference Technical Specification sections 3.2.3.
Core Operating Limits Report Unit 1, Cycle 23 Revision 1
- 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 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 qt + 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 q, - 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 O F K', 5 1.11 K5 = 0.0275/"F for increasing T; 0 for decreasing T K6 = 0.002/"F for T > T' ;O for T 5 T' 73 = 10 seconds Page 5 of 17
Core Operating Limits Report
,. Unit 1, Cycle 23 Revision 1
- 11. 3.4.1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling IDNB) 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. 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 Reference Technical Specification section 3.9.1 Page 6 of 17
Core Operating Limits Report Unit ? , Cycle 23 Revision 1 REFERENCES
- 1. NSPNAD-8101-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 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 I, July 1997.
WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision I, 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, "1 0 CFR 50.46 Evaluation Model Report: W-COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support zIRLoTMCladding Options,"
February 1994.
- 9. NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology," Revision 0, April 1993.
- 10. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification," February 1994.
- 11. WCAP-8745-P-A, "Design Bases for the Thermal Overpower A T and Thermal Overtemperature AT Trip Functions," September 1986.
Page 7 of 17
Core Operating Limits Report Unit 1 , Cycle 23
.- Revision 1 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.
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-P/ANC 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 Screening 2347, Rev. 0, "Unit 1 Cycle 23 COLR Revision to Figures 3 and 4 Titles."
50.59 Screening 2429, Rev. 0; "Unit 1 Cycle 23 COLR Revision to Section 3.1.3 Isothermal Temperature Coefficient."
Page 8 of 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 Table 1 Minimum Required Shutdown Margin Number of Charging Pumps Running*
- Plant Conditions 0-1 Pump 2 Pumps 3 Pumps Mode 1* - - -
I Mode 2* 1 2.0% 1 2.0% ( 2.0% 1
[ Mode 3, Twe> 520°F I 2.0% 2.0% 2.0% I Mode 3,350°F 5 Tave< 520°F 2 .O% 2.0% 2.5%
Mode 4 2.5% 4.5% 7.0%
Mode 5***. T,,, < 200°F 2.5% 5.O% 7.5%
Mode 6, AM***, TaVe> 68°F 5.129% 5.129% 7.0%
Mode 6, ARO***, T,, 2 68°F 5.129% 6.0% 9-0%
Physics Testing in Mode 2 0.5% 0.5% 0.5%
Operational Mode Definitions, as per TS Table 1.1-1.
- For Mode 1 and Mode 2 with Keff >_ 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.
- These values are also applicable for the Unit 1 Cycle 22 end of cycle Page 9 of 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 Table 2 - W(z) Values(Top 10% and Bottom 8% excluded)
Page 10 of 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 I
60 11.8000 1 1.0000 1.0000 1.0000 1.0000 1.0000
[Top] 61 12.0000 I 1.0000 1.0000 1.0000 1.0000 1.0000 Page 11 of 17
Core Operating Limits Report Unit ? , Cycle 23 Revision 1 Table 3 F ~ ~ ( Penalty z) Factor Exposure Range F ~ ~ ( Penalty z) Factor BOC - EOC 1.02 Page 12 of 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 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 17
Core Operating Limits Report Unit I,Cycle 23 Revision 1 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 0 20 40 60 80 I00 Power Level, % of Rated Thermal Power Bank Positions Given By:
BankD=(150/63) * (P- 100)f 185 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 17
Core Operating Limits Report
.. Unit 1, Cycle 23 Revision 1 Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4 Condition B)
Bank Positions Given By:
Bank D = (150 / 63) * (P - 90) + 224 Bank C = (150 / 63) * (P - 90) +-224 + 128 NOTE: The top of the active &el height corresponds to 224 steps. The ARO parking position may be any position above 224 steps.
Page 15 of 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4 Condition A)
Bank Positions Given By:
Bank D=(150/63) * (P-70)+224 Bank C=(150/63) * (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 17
Core Operating Limits Report Unit 1, Cycle 23 Revision 1 Figure 5 Flux Difference Operating Envelope Page 17 of 17