L-PI-04-083, Core Operating Limits Report (COLR) for Prairie Island Unit 1 Cycle 22. Revisions 1 and 2, and Core Operating Limits Report (COLR) for Prairie Island Unit 2 Cycle 22, Revisions 1 and 2
| ML042260367 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/04/2004 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-04-083 | |
| Download: ML042260367 (37) | |
Text
C I
COMMited to r ket Prairie Island Nuclear Generafing Plant Operated by Nuclear Management Company, LLC AUG 4 2004 L-PI-04-083 TS 5.6.5.d U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Core Oneratinci Limits Report (COLR) For Prairie Island Unit 1 Cycle 22. Revisions I and 2. and Core Operating Limits Renort (COLR) For Prairie Island Unit 2 Cycle 22.
Revisions I and 2 Pursuant to the requirements of Technical Specification Section 5.6.5.d, the COLR for Prairie Island Unit I Cycle 22, Revisions I and 2, and the COLR for Prairie Island Unit 2 Cycle 22, Revisions I and 2 are attached.
The limits specified in the attached COLRs have been established using NRC approved methodologies.
The Unit I and Unit 2 COLRs Revision 1 have been revised for Cycle 22 to incorporate Westinghouse Safety Analysis Transition per License Amendments 162/153. Revision I for each COLR contains transitional values for the Overpressure and Overtemperature delta-T Trip setpoints for use while physical plant changes associated with License Amendments 162/153 were implemented.
The Unit I and Unit 2 COLRs Revision 2 have been revised for Cycle 22 to revise the Fq limit from 2.4 to 2.5 and remove the transitional values for the Overpressure and Overtemperature delta-T Trip setpoints.
Please address any comments or questions regarding this letter to Mr. Dale Vincent at 1-651-388-1121.
1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121 4&0
C Document Control Desk Page 2 Summary of Commitments In this letter NMC has not made any new or revised any Nuclear Regulatory Commission co
.ents.
Bite Vic President rairie sland Nuclear Generating Plant Noanageme Company, LLC Enclosures (4) cc:
Administrator, Region 111, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC
PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1-CYCLE 22 REVISION 1 Reviewed By:
Jon Kapitz Supervisor, Nuclear Engineering Date: 6 161oct Reviewed By: - tv fl Ed Mercied Supervisor, PWR Analysis Reviewed By /
Gabe Salamon Manager, Regulatory Affairs Reviewed Byy Scott Northard Director, Engineering Date:_(_t__/
Date: _______
Date: 6 /I 06 I Approved a a Date:
42-1 tote Joewolymossy Cl
-IeVice Presider Note: This report is not part of the Technical Specifications This report is referenced in the Technical Specifications Page 1 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1 - CYCLE 22 REVISION I This report provides the values of the limits for Unit 1 Cycle 22 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:
- 1.
2.1.1 Reactor Core SLs
- 2.
3.1.1 Shutdown Margin (SDM)
- 3.
3.1.3 Isothermal Temperature Coefficient (ITC)
- 4.
3.1.5 Shutdown Bank Insertion Limits
- 5.
3.1.6 Control Bank Insertion Limits
- 6.
3.1.8 Physics Tests Exceptions - MODE 2
- 7.
3.2.1 Heat Flux Hot Channel Factor (FQ(z))
- 8.
3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FA)
- 9.
3.2.3 Axial Flux Difference (AFD)
- 10.
3.3.1 Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1-1
- 11.
3.4.1 RCS Pressure, Temperature, and Flow -Departure from Nucleate Boiling (DNB) Limits
- 12.
3.9.1 Boron Concentration I1.
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 Reference Technical Specification section 3.1.1.
Page 2 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1
- 3.
3.1.3 Isothermal Temperature Coefficient (ITC ITC Upper limit:
- a. < 5 pcmnIF for power levels < 70% RTP; and
- a. -32.7 pcm/IF 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 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1
- 7.
3.2.1 Heat Flux Hot Channel Factor (Fp(Z The Heat Flux Hot Channel Factor shall be within the following limits:
CFQ = 2.40 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 (FN)
The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:
FAH< 1.77x [I +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 THERMAL POWER > 50% RTP.
Reference Technical Specification sections 3.2.3.
Page 4 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I
- 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, either Option 1 or Option 2:
Option 1 ATo
=
Indicated AT at RATED THERMAL POWER, %
T
=
Average temperature, IF r
=
567.3 0F P
=
Pressurizer Pressure, psig P
=
2235 psig K1 1.11 K2
=
0.009 /OF K3
=
0.000566 /psi Tj
=
30 seconds T2
=
4 seconds f(Al)
=
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)
Foi qt - qb within -12, +9 % f(AI) = 0 (b)
For each percent that the magnitude of qt - qb exceeds +9%
the AT trip setpoint shall be automatically reduced by an equivalent of 2.5% of RATED THERMAL POWER.
(c)
For each percent that the magnitude of qt - qb exceeds -12 %
the AT trip setpoint shall be automatically reduced by an equivalent of 1.5 % of RATED THERMAL POWER.
Option 2 ATo
=
Indicated AT at RATED THERMAL POWER, %
T
=
Average temperature, IF T
=
560.0 °F P
=
Pressurizer Pressure, psig P
=
2235 psig K1 1.17 K2
=
0.014 /°F K3
=
0.00100/psi
=
30 seconds Page 5 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I T2
=
4 seconds f(AI)
=
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(AI) = 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, either Option 1 or Option 2:
Option 1 AT< ATO {K4 - K5 + 3sT _ K6 (T - T) - f (A)}
ATO
=
Indicated AT at RATED THERMAL POWER, %
T
=
Average temperature, °F T'
=
560.0 OF K4 1.10 Ks
= 0.0275/°F for increasing T; 0 for decreasing T K6
= 0.002/°F for T > T'; O for T < T' T3
=
1O seconds f(AI)
=
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 -12, +9 % f(A1) = O (b)
For each percent that the magnitude of qt - qb exceeds +9%
the AT trip setpoint shall be automatically reduced by an equivalent of 2.5% of RATED THERMAL POWER.
(c)
For each percent that the magnitude of qt - qb exceeds -12 %
the AT trip setpoint shall be automatically reduced by an equivalent of 1.5% of RATED THERMAL POWER.
Page 6 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Option 2 AT
- ATo4 K K 5 r2
_s KK 6(T -T)}
ATo
=
Indicated AT at RATED THERMAL POWER, %
T
=
Average temperature, IF T'
=
560.0 OF K4 1.11 K5
=
0.0275/0F for increasing T; 0 for decreasing T K6
= 0.002/ 0F for T > T'; 0 for T T r X3
=
10 seconds
- 11.
3.4.1 RCS Pressure. Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Pressurizer pressure limit = 2205 psia RCS average temperature limit = 5640F 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) Kff
- 0.95 b) 2000 ppm c) The Shutdown Margin specified in Table 1 Reference Technical Specification section 3.9.1 Page 7 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I REFERENCES
- 1.
NSPNAD-8101-A, "Qualification of Reactor Physics Methods for Application to Prairie Island," Revision 2, October 2000.
- 2.
NSPNAD-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units," Revision 7, July 1999.
- 3.
NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology," Revision 1, October 2000.
- 4.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July, 1985.
5.a WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUIWP Code," August, 1985.
5.b WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRIJMP Code," Addendum 2 Revision 1, July 1997.
6.a WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 1,2,3, December 1988.
6.b WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 2, Volume 2 Addendum 1, December 1988.
6.c WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 4, March 1991.
- 7.
XN-NF-77-57-(A),
XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II," May 1981.
- 8.
WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report: W-COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOIm Cladding Options,"
February 1994.
- 9.
NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology," April 1993.
- 10.
NAD-P1-003, "Prairie Island Unit 1 Cycle 22 Final Reload Design Report (Reload Safety Evaluation) and USAR Update," Revision 1, October 2002.
- 11.
NAD-Pl-004, "Prairie Island Unit I Cycle 22 Startup and Operations Report,"
Revision 1, November 2002.
Page 8 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1
- 12.
WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification," February 1994.
- 13.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.
- 14.
WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989.
- 15.
WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report,"
January 1999.
- 16.
WCAP-7588 Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.
- 17.
WCAP-7908-A, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a U0 2 Fuel Rod," December 1989.
- 18.
WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
- 19.
WCAP-7979-P-A, 'TWINKLE -
A Multidimensional Neutron Kinetics Computer Code," January 1975.
- 20.
WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"
December 1985.
- 21.
WCAP-1 1394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.
22 WCAP-1 1596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
- 23.
WCAP-12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.
- 24.
WCAP-14565-P-A, 'VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
- 25.
WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.
Page 9 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Table I Minimum Required Shutdown Margin Plant Conditions Number of Chargiug Pumps Running**
0-1 Pump 2 Pumps 3 Pumps Mode 1*
Mode 2*
2.0%
2.0%
2.0%
Mode 3, Tave > 5200F 2.0%
2.0%
2.0%
Mode 3, 3500F < Tave < 5200 F 2.0%
2.0%
2.5%
Mode 4 2.0%
4.5%
6.5%
Mode 5***, T
< 200'F 2.5%
5.0%
7.0%
Mode 6, ARI***, Tave > 68OF 5.129%
5.129%
7.0%
Mode 6, ARO***, Tavc > 680F 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>1.0, 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 21 end of cycle Page 10 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1 Table 2 WNI(z) Values HeIliht BU FMWdIMTUI a.....
T -
T V
Itt' 150 4000 12000 16000 22000 AO = 2.44 AO = -0.73 AO0=-2.98 1 AO=-0.62 AO = -1.81
[BOTTOMI1 0.00 1.0000
=1.0000 1.0000 1.0000 1.0000 2
0.20 1.0000 1.0000 1.0000 1.0000 1.0000 3
0.40 1.0000 1.0000 1.0000 1.0000 1.0000 4
0.60 1.0000 1.0000 1.0000 1.0000 1.0000 5
0.80 1.0000 1.0000 1.0000 1.0000 1.0000 6
1.00 1.0000 1.0000 1.0000 1.0000 1.0000 7
1.20 1.3388 1.2555 1.2153 1.1718 1.1661 8
1.40 1.3231 1.2425 1.2029 1.1642 1.1558 9
1.60 1.3058 1.2285 1.1896 1.1565 1.1457 10 1.80 1.2873 1.2134 1.1757 1.1487 1.1359 11 2.00 1.2677 1.1977 1.1614 1.1409 1.1261 12 2.20 1.2476 1.1816 1.1468 1.1332 1.1168 13 2A0 1.2268 1.1652 1.1325 1.1254 1.1073 14 2.60 1.2057 1.1488 1.1185 1.1175 1.0974 15 2.80 1.1900 1.1380 1.1027 1.1136 1.0971 16 3.00 1.1786 1.1276 1.0984 1.1123 1.1014 17 3.20 1.1689 1.1210 1.0997 1.1121 1.1092 18 3A0 1.1618 1.1207 1.0999 1.1146 1.1230 19 3.60 1.1586 1.1225 1.0998 1.1221 1.1362 20 3.80 1.1567 1.1236 1.0994 1.1296 1.1482 21 4.00 1.1543 1.1242 1.0976 1.1360 1.1591 22 4.20 1.1513 1.1242 1.0964 1.1417 1.1685 23 4.40 1.1475 1.1235 1.0979 1.1464 1.1763 24 4.60 1.1429 1.1222 1.0997 1.1501 1.1825 25 4.80 1.1377 1.1202 1.1028 1.1527 1.1871 26 5.00 1.1317 1.1179 1.1060 1.1542 1.1900 27 5.20 1.1257 1.1145 1.1083 1.1547 1.1919 28 5A0 1.1211 1.1116 1.1109 1.1540 1.1940 29 5.60 1.1161 1.1162 1.1155 1.1527 1.1961 30 5.80 1.1096 1.1234 1.1213 1.1576 1.1975 31 6.00 1.1098 1.1319 1.1296 1.1696 1.2040 32 6.20 1.1178 1.1418 1.1404 1.1816 1.2162 33 6.40 1.1245 1.1509 1.1516 1.1922 1.2264 34 6.60 1.1305 1.1592 1.1637 1.2014 1.2354 35 6.80 1.1359 1.1666 1.1752 1.2091 1.2428 36 7.00 1.1403 1.1728 1.1854 1.2152 1.2486 37 7.20 1.1437 1.1778 1.1945 1.2196 1.2531 38 7.40 1.1459 1.1817 1.2028 1.2220 1.2567 Page 11 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1 Table 2 (Cont'd)
W(z) Values Heiaht BU [MWdIMTUJ Iftl 150 4000 12000 16000 20000 AO = 2.44 AO = -0.73 AO
-2.98 AO = -0.62 AO = -1.81 39 7.60 1.1480 1.1852 12104 1.2224 1.2579 40 7.80 1.1492 1.1873 1.2163 1.2208 1.2566 41 8.00 1.1491 1.1877 1.2204 1.2169 1.2527 42 8.20 1.1475 1.1862 1.2224 1.2109 1.2461 43 8.40 1.1444 1.1827 1.2224 1.2028 1.2368 44 8.60 1.1399 1.1779 1.2201 1.1914 1.2257 45 8.80 1.1334 1.1688 1.2156 1.1844 1.2137 46 9.00 1.1347 1.1672
-1.2139 1.1803 1.2007 47 9.20 1.1467 1.1798 1.2157 1.1763 1.1902 48 9.40 1.1556 1.1914 1.2153 1.1771 1.1872 49 9.60 1.1664 1.2019 1.2225 1.1768 1.1822 50 9.80 1.1773 1.2105 1.2361 1.1775 1.1798 51 10.00 1.1869 1.2191 1.2473 1.1805 1.1802 52 10.20 1.1968 1.2290 1.2555 1.1842 1.1837 53 10.40 1.2066 1.2372 1.2657 1.1890 1.1871 54 10.60 1.2171 1.2466 1.2710 1.1934 1.1906 55 10.80 1.2271 1.2538 1.2753 1.1980 1.1942 56 11.00 1.0000 1.0000 1.0000 1.0000 1.0000 57 11.20 1.0000 1.0000 1.0000 1.0000 1.0000 58 11.40 1.0000 1.0000 1.0000 1.0000 1.0000 59 11.60 1.0000 1.0000 1.0000 1.0000 1.0000 60 11.80 1.0000 1.0000 1.0000 1.0000 1.0000
[TOP] 61 12.00 1.0000 1.0000 1.0000 1.0000 1.0000 Page 12 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1 Table 3 FWQ(Z) Penalty Factor I
Exposure Range l
Fwo(Z) Penalty Factor BOC-EOC l
1.02 Page 13 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Figure 1 Reactor Core Safety Limits 680 660 640 IL 6 620 V,
0, I-
> 600 a:
580 560 540 0
0.2 0.4 0.6 0.8 1
1.2 Fraction of Rated Thermal Power 1.4 Page 14 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision 1 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 200 ul 150 0a4
.o
.2 en0 no 100 C
50 0
0 20 40 60 80 Power Level, % of Rated Thermal Power 100 Bank Positions Given By:
- BankD=(150/63)*(P -100)+
185
- BankC=(150/63)*(P-100)+185+128
- BankB=(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 15 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4) 224 200 150 In 0.Q co C',
02 100 I'r 0a-xto M
50 0
0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:
- BankD=(150/63)*(P-90)+224
- BankC=(150/63)*(P-90)+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 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4) 224 200 150 ISO W
0:
100 0.a-M 50 0
0 20 40 60 I
80 100 Power Level, % of Rated Thermal Power Bank Positions Given By.
- BankD=(15063)*(P -70)+224
- BankC=(150/63)*(P-70)+224+128 NOTE: The top of the active fuel height corresponds to 224 steps. The ARO parking position maybe any position above 224 steps.
Page 17 of 18
Core Operating Limits Report Unit 1, Cycle 22 Revision I Figure 5 Flux Difference Operating Envelope 110 100 90 80 70 I..
0 a.-
60 50 40 30 -
20 10 _
0-
-25
-20
-15
-10
-5 0
AI%
5 10 15 20 25 Page 18 of 18
PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1-CYCLE 22 REVISION 2 Reviewed By:
Jo,,
Jon Kapitz Supervisor, Nuclear Engineering Reviewed By:
Ed Mercier Supervisor, PWR Analysis Reviewed By:__
Gabe Salamon Manager, Regulatory Affairs Reviewed By:
Scott Northard Director, Engineering Approved By:
V-Joe lymossy Vice President Date:
/
Date: 6_/_/0 Date:
(O /I Date:
Oqp--2/7{
Note: This report is not part of the Technical Specifications This report is referenced in the Technical Specifications Page 1 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 PRAIRIE ISLAND NUCLEAR GENERATING PLANT CORE OPERATING LIMITS REPORT UNIT 1-CYCLE 22 REVISION 2 This report provides the values of the limits for Unit 1 Cycle 22 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:
- 1.
2.1.1 Reactor Core SLs
- 2.
3.1.1 Shutdown Margin (SDM)
- 3.
3.1.3 Isothermal Temperature Coefficient (ITC)
- 4.
3.1.5 Shutdown Bank Insertion Limits
- 5.
3.1.6 Control Bank Insertion Limits
- 6.
3.1.8 Physics Tests Exceptions - MODE 2
- 7.
3.2.1 Heat Flux Hot Channel Factor (FQ(z))
- 8.
3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Fk)
- 9.
3.2.3 Axial Flux Difference (AFD)
- 10.
3.3.1 Reactor Trip System (RTS) Instrumentation Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1-1
- 11.
3.4.1 RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits
- 12.
3.9.1 Boron Concentration I1.
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 I Reference Technical Specification section 3.1.1.
Page 2 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2
- 3.
3.1.3 Isothermal Temperature Coefficient (ITC)
ITC Upper limit:
- a. < 5 pcm/0F for power levels < 70% RTP; and
- a. -32.7 pcm/0F 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 1, Cycle 22 Revision 2
- 7.
3.2.1 Heat Flux Hot Channel Factor (FQ(Z)
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 (FM)
The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:
FmS < 1.77 x [I + 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 I with THERMAL POWER > 50% RTP.
Reference Technical Specification sections 3.2.3.
Page 4 of 17
Core Operating Limits Report Unit 1, Cycle 22 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:
ATo
=
Indicated AT at RATED THERMAL POWER, %
T
=
Average temperature, 0F r
=
560.0 OF P
=
Pressurizer Pressure, psig P
=
2235 psig K1 1.17 K2 0.014 rF K3
=
0.00100 /psi
=
30 seconds 2
=
4 seconds f(al)
=
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(AI) = 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 q. - 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, 0F Te
=
560.0 OF K4 1.11 K5
=
0.0275/ 0F for increasing T; 0 for decreasing T K6
=
0.002/OF for T > T'; 0 for T 5 T' T3
=
1O seconds Page 5 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2
- 11.
3.4.1 RCS Pressure. Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits Pressurizer pressure limit = 2205 psia RCS average temperature limit = 5640F 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) Kef < 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 1, Cycle 22 Revision 2 REFERENCES
- 1.
NSPNAP-8101-A, "Qualification of Reactor Physics Methods for Application to Prairie Island," Revision 2, October 2002.
- 2.
NSPNAP-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units," Revision 7, July 1999.
- 3.
NSPNAD-97002-P, "Northern States Power Company's "Steam Line Break Methodology," Revision 1, October 1998.
- 4.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July, 1985.
5.a WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August, 1985.
5.b WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," Addendum 2 Revision 1, July 1997.
6.a WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 1,2,3, December 1988.
6.b WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 2, Volume 2 Addendum 1, December 1988.
6.c WCAP-10924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology,"
Revision 1, Volume 1 Addendum 4, March 1991.
- 7.
XN-NF-77-57-(A),
XN-NF-77-57, Supplement 1 (A),
"Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II," Revision, May 1981.
- 8.
WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report: W-COBRAITRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOd Cladding Options,"
February 1994.
- 9.
NSPNAD-93003-A, "Prairie Island Units 1 and 2 Transient Power Distribution Methodology," Revision 0, April 1993.
- 10.
NAD-PI-003, "Prairie Island Unit 1 Cycle 22 Final Reload Design Report (Reload Safety Evaluation) and USAR Update," Revision 1, October 2002.
- 11.
NAD-PI-004, "Prairie Island Unit I Cycle 22 Startup and Operations Report,"
Revision 1, November 2002.
Page 7 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2
- 12.
WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification," February 1994.
- 13.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.
- 14.
WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.
- 15.
WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report,"
January 1999.
- 16.
WCAP-7588 Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.
- 17.
WCAP-7908-A, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod," December 1989.
- 18.
WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.
- 19.
WCAP-7979-P-A, "TWINKLE -
A Multidimensional Neutron Kinetics Computer Code," January 1975.
- 20.
WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"
December 1985.
- 21.
WCAP-I 1394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.
22 WCAP-I1596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.
- 23.
WCAP-12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.
- 24.
WCAP-14565-P-A, "VIPRE-1 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
- 25.
WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.
Page 8 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Table 1 Minimum Required Shutdown Margin Plant Conditions Number of ng Pump Running**
0-1 Pump 2 Pumps 3 Pumps Mode I* *_
Mode 2*
2.0%
2.0%
2.0%
Mode 3, T., > 520'F 2.0%
2.0%
2.0%
Mode 3, 350'F < T.v < 5200F 2.0%
2.0%
2.5%
Mode 4 2.0%
4.5%
6.5%
Mode 5***, Tave < 2000F 2.5%
5.0%
7.0%
Mode 6, ARI***, Tave > 680F 5.129%
5.129%
7.0%
Mode 6, ARO***, Tave > 680F 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>1.0, 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 I Cycle 21 end of cycle Page 9 of 17
Core Operating Limits Report Unit 1, Cyde 22 Revision 2 Table 2 W(z) Values Height I BU [MWd/MTU_
[ft]
150 4000 12000 16000 22000
+
4 4
4-AO = 2.44 AO = -0.73 AO = -2.98 l AO = -0.62 AO = -1.81
[BOTTOM] 1 0.00 1.0000 1.0000 1.0000 1.0000 1.0000 2
0.20 1.0000 1.0000 1.0000 1.0000 1.0000 3
0.40 1.0000 1.0000 1.0000 1.0000 1.0000 4
0.60 1.0000 1.0000 1.0000 1.0000 1.0000 5
0.80 1.0000 1.0000 1.0000 1.0000 1.0000 6
1.00 1.0000 1.0000 1.0000 1.0000 1.0000 7
1.20 1.3388 1.2555 1.2153 1.1718 1.1661 8
1.40 1.3231 1.2425 1.2029 1.1642 1.1558 9
1.60 1.3058 1.2285 1.1896 1.1565 1.1457 10 1.80 1.2873 1.2134 1.1757 1.1487 1.1359 11 2.00 1.2677 1.1977 1.1614 1.1409 1.1261 12 2.20 1.2476 1.1816 1.1468 1.1332 1.1168 13 2.40 1.2268 1.1652 1.1325 1.1254 1.1073 14 2.60 1.2057 1.1488 1.1185 1.1175 1.0974 15 2.80 1.1900 1.1380 1.1027 1.1136 1.0971 16 3.00 1.1786 1.1276 1.0984 1.1123 1.1014 17 3.20 1.1689 1.1210 1.0997 1.1121 1.1092 18 3.40 1.1618 1.1207 1.0999 1.1146 1.1230 19 3.60 1.1586 1.1225 1.0998 1.1221 1.1362 20 3.80 1.1567 1.1236 1.0994 1.1296 1.1482 21 4.00 1.1543 1.1242 1.0976 1.1360 1.1591 22 4.20 1.1513 1.1242 1.0964 1.1417 1.1685 23 4.40 1.1475 1.1235 1.0979 1.1464 1.1763 24 4.60 1.1429 1.1222 1.0997 1.1501 1.1825 25 4.80 1.1377 1.1202 1.1028 1.1527 1.1871 26 5.00 1.1317 1.1179 1.1060 1.1542 1.1900 27 5.20 1.1257 1.1145 1.1083 1.1547 1.1919 28 5.40 1.1211 1.1116 1.1109 1.1540 1.1940 29 5.60 1.1161 1.1162 1.1155 1.1527 1.1961 30 5.80 1.1096 1.1234 1.1213 1.1576 1.1975 31 6.00 1.1098 1.1319 1.1296 1.1696 1.2040 32 6.20 1.1178 1.1418 1.1404 1.1816 1.2162 33 6.40 1.1245 1.1509 1.1516 1.1922 1.2264 34 6.60 1.1305 1.1592 1.1637 1.2014 1.2354 35 6.80 1.1359 1.1666 1.1752 1.2091 1.2428 36 7.00 1.1403 1.1728 1.1854 1.2152 1.2486 37 7.20 1.1437 1.1778 1.1945 1.2196 1.2531 38 7.40 1.1459 1.1817 1.2028 1.2220 1.2567 Page 10 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Table 2 (Cont'd)
W(z) Values Height BU [MWd/MTUJ Ift1 150 4000 12000 16000 20000 AO = 2.44 AO = -0.73 AO = -2.98 AO = -0.62 AO = -1.81 39 7.60 1.1480 1.1852 1.2104 1.2224 1.2579 40 7.80 1.1492 1.1873 1.2163 1.2208 1.2566 41 8.00 1.1491 1.1877 1.2204 1.2169 1.2527 42 8.20 1.1475 1.1862 1.2224 1.2109 1.2461 43 8.40 1.1444 1.1827 1.2224 1.2028 1.2368 44 8.60 1.1399 1.1779 1.2201 1.1914 1.2257 45 8.80 1.1334 1.1688 1.2156 1.1844 1.2137 46 9.00 1.1347 1.1672 1.2139 1.1803 1.2007 47 9.20 1.1467 1.1798 1.2157 1.1763 1.1902 48 9.40 1.1556 1.1914 12153 1.1771 1.1872 49 9.60 1.1664 1.2019 1.2225 1.1768 1.1822 50 9.80 1.1773 1.2105 1.2361 1.1775 1.1798 51 10.00 1.1869 1.2191 1.2473 1.1805 1.1802 52 10.20 1.1968 1.2290 1.2555 1.1842 1.1837 53 10.40 1.2066 1.2372 1.2657 1.1890 1.1871 54 10.60 1.2171 1.2466 1.2710 1.1934 1.1906 55 10.80 1.2271 1.2538 1.2753 1.1980 1.1942 56 11.00 1.0000 1.0000 1.0000 1.0000 1.0000 57 11.20 1.0000 1.0000 1.0000 1.0000 1.0000 58 11.40 1.0000 1.0000 1.0000 1.0000 1.0000 59 11.60 1.0000 1.0000 1.0000 1.0000 1.0000 60 11.80 1.0000 1.0000 1.0000 1.0000 1.0000
[rOPI 61 12.00 1.0000 1.0000 1.0000 1.0000 1.0000 Page 1 1 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Table 3 FWQ(Z) Penalty Factor Exposure Range Fwp(Z) Penalty Factor BOC-EOC 1.02 Page 12 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Figure I Reactor Core Safety Limits 680 0) 0)
CD) co, 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 1, Cycle 22 Revision 2 Figure 2 Rod Insertion Limit, 128 Step Tip-to-Tip 200
,, 150 a-0o Ct) 0 00 100 le 50 0
0 20 40 60 80 Power Level, % of Rated Thermal Power 100 Bank Positions Given By:
- BankD=(150/63)*(P-100)+185
- BankC=(150/63)*(P-100)+185+128
- BankB=(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 14 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Figure 3 Rod Insertion Limit, 128 Step Tip-to-Tip, One Bottomed Rod (Technical Specification 3.1.4) 224 200 150 U) 1 0)
U, CO 0 *~100 U) 0 C5 50 o
L 0
Power Level, % of Rated Thermal Power Bank Positions Given By:
- BankD=(150/63)*(P-90)+224
- BankC=(150/63)*(P-90)+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 15 of 17
Core Operating Limits Report Unit 1, Cycle 22 Revision 2 Figure 4 Rod Insertion Limit, 128 Step Tip-to-Tip, One Inoperable Rod (Technical Specification 3.1.4) 224 200 150 00.
co 0a-Ae C-100 50 0
0 20 40 60 80 100 Power Level, % of Rated Thermal Power Bank Positions Given By:
- BankD=(150/63)*(P-70)+224
- BankC=(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 22 Revision 2 Figure 5 Flux Difference Operating Envelope 110 100 90 80 70 L.
0a-60 50 40 30 20 10 0
-25
-20
-15
-10
-5 0
5 10 15 20 25 Al%
Page 17 of 17