NRC 2006-0079, Cycle 29 (U2C29) Core Operating Limits Report
| ML063170204 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/10/2006 |
| From: | Koehl D Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2006-0079 | |
| Download: ML063170204 (17) | |
Text
Point Beach Wear PlW Operated by Nuclear Management Company, LLC November 10,2006 NRC 2006-0079 10 CFR 50.36 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001 Point Beach Nuclear Plant Unit 2 Docket No. 50-301 License No. DPR-27 Unit 2 Cycle 29 (U2C29) Core Operating Limits Report In accordance with the requirements of Point Beach Nuclear Plant (PBNP) Technical Specification 5.6.4, "Core Operating Limits Report (COLR)", Nuclear Management Company, LLC (NMC), is submitting the Core Operating Limits Report for PBNP Unit 2 Cycle 29 (U2C29). The PBNP U2C29 COLR was issued October 27,2006. This letter contains no new commitments and no revisions to existing commitments.
/' Dennis L. Koehl 1 Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosure Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW 6590 Nuclear Road Two Rivers, Wisconsin 54241 Telephone:
920.755.2321 ENCLOSURE CORE OPERATING LIMITS REPORT POINT BEACH NUCLEAR PLANT, UNIT 2 CYCLE 29 (U2C29)
TRM 2.1 CORE OPERATING LIMITS REPORT (COLR) UNIT 2 CYCLE 29 REVISION 9
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
CORE OPERATING LIMITS REPORT (COLR)
UNIT 2 CYCLE 29 TRM 2.1 U2
Revision 9
October 27, 2006 Page 2 of 15 1.0 CORE OPERATING LIMITS REPORT
This Core Operating Limits Report (COLR) for Point Beach Nuclear Plant has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.4.
A cross-reference between the COLR sections and the PBNP Technical Specifications affected by this report is given below:
COLR Section PBNP TS Description 2.1 2.1.1 Reactor Core Safety Limits 2.2 3.1.1 3.1.4 3.1.5 3.1.6 3.1.8 Shutdown Margin Rod Group Alignment Limits Shutdown Bank Insertion Limits Control Bank Insertion Limits Physics Test Exceptions 2.3 3.1.3 Moderator Temperature Coefficient 2.4 3.1.5 Shutdown Bank Insertion Limit 2.5 3.1.6 Control Bank Insertion Limits 2.6 3.2.1 Heat Flux Hot Channel Factor (F Q(Z)) 2.7 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NH) 2.8 3.2.3 Axial Flux Difference (AFD)
2.9 3.3.1 Overtemperature T Setpoint 2.10 3.3.1 Overpower T Setpoint 2.11 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 2.12 3.9.1 Refueling Boron Concentration Figure 1 2.1.1 Reactor Core Safety Limits Figure 2 3.1.1 Required Shutdown Margin (SDM) Figure 3 3.1.6 Control Bank Insertion Limits Figure 4 3.2.1 Hot Channel Factor Normalized Operating Envelope for 422V+ Fuel Figure 5 3.2.1 RAOC Summary of W(Z) with HFP AFD Band of -8/+9% (Top 15% and Bottom 12% Excluded) Figure 6 3.2.3 Flux Difference Operating Envelope
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
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October 27, 2006 Page 3 of 15 2.0 OPERATING LIMITS 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 the NRC approved methodologies specified in Technical Specification 5.6.4.
2.1 Reactor Core Safety Limits (TS 2.1.1)
The combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in Figure 1.
Applicability: MODES 1 and 2 2.2 Shutdown Margin (TS 3.1.1 and referenced in TS 3.1.4, 3.1.5, 3.1.6, and 3.1.8) 2.2.1 SDM shall be within the limits provided in Figure 2.
Applicability: MODES 1, 2, and 3 2.2.2 SDM shall be 1% k/k. Applicability: MODES 4 and 5 2.3 Moderator Temperature Coefficient (TS 3.1.3) 2.3.1 The upper MTC limits shall be maintained within the limits.
2.3.2 The maximum upper MTC limits shall be:
5 pcm/°F for power levels 70% RTP 0 pcm/°F for power levels >70% RTP Applicability: MODE 1 and MODE 2 with keff 1.0. 2.4 Shutdown Bank Insertion Limit (TS 3.1.5)
NOTE: This limit is not applicable while performing SR 3.1.4.2.
2.4.1 Each shutdown bank shall be fully withdrawn.
2.4.2 Fully withdrawn is defined as 225 steps.
Applicability: MODES 1 and 2
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October 27, 2006 Page 4 of 15 2.5 Control Bank Insertion Limits (TS 3.1.6)
NOTE: This limit is not applicable while performing SR 3.1.4.2.
The control banks shall be within the insertion, sequence and overlap limits specified in Figure 3.
Applicability: MODE 1 and MODE 2 with keff 1.0 2.6 Nuclear Heat Flux Hot Channel Factor (F Q(Z)) (TS 3.2.1)
The Heat Flux Hot Channel Factor shall be within the following limits:
F Q (Z) CF Q
- K(Z) / P for P > 0.5 F Q (Z) CF Q
- K(Z) / 0.5 for P 0.5 Where P is the fraction of Rated Power at which the core is operating.
F Q (Z) is both:
- Steady State F Q C(Z) = F Q (Z)
- 1.08
- Transient F Q W(Z) = F Q C(Z)
- W(Z)
CF Q = 2.60 (422V+ Fuel)
K(Z) is the function in Figure 4 W(Z) is the function in Figure 5 Applicability: MODE 1 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F NH) (TS 3.2.2)
The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:
2.7.1 F NH <1.77 x [1 + 0.3(1-P)] (422V+ Fuel) where: P is the fraction of Rated Power at which the core is operating.
Applicability: MODE 1
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
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October 27, 2006 Page 5 of 15 2.8 Axial Flux Difference (AFD) (TS 3.2.3)
NOTE: The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.
The indicated axial flux difference in % flux difference units shall be maintained within the allowed operational space defined by Figure 6.
Applicability: MODE 1 with THERMAL POWER 50% RTP 2.9 Overtemperature T Setpoint (TS 3.3.1, Table 3.3.1-1 note 1)
Overtemperature T setpoint parameter values:
T O = indicated T at Rated Power, °F T = average temperature, °F T 569.0°F P = 2235 psig K 1 1.16 K 2 = 0.0149 K 3 = 0.00072 1 = 25 sec 2 = 3 sec 3 = 2 sec for Rosemont or equivalent RTD
= 0 sec for Sostman or equivalent RTD 4 = 2 sec for Rosemont or equivalent RTD
= 0 sec for Sostman or equivalent RTD f(I) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where q t and q b are the percent power in the top and bottom halves of the core respectively, and q t + q b is total core power in percent of Rated Power, such that:
2.9.1 For q t - q b within -12, +5 percent, f(I) = 0. 2.9.2 For each percent that the magnitude of q t - q b exceeds +5 percent, the T trip setpoint shall be automatically reduced by an equivalent of 2.12 percent of Rated Power.
2.9.3 For each percent that the magnitude of q t - q b exceeds -12 percent, the T trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of Rated Power.
Applicability: MODES 1 and 2
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
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October 27, 2006 Page 6 of 15 2.10 Overpower T Setpoint (TS 3.3.1, Table 3.3.1-1 note 2)
Overpower T setpoint parameter values:
T O = indicated T at Rated Power, °F T = average temperature, °F T 569.0°F K 4 1.10 of Rated Power K 5 = 0.0262 for increasing T K 5 = 0.0 for decreasing T K 6 = 0.00103 for T T K 6 = 0.0 for T < T 5 = 10 sec 3 = 2 sec for Rosemont or equivalent RTD = 0 sec for Sostman or equivalent RTD 4 = 2 sec for Rosemont or equivalent RTD = 0 sec for Sostman or equivalent RTD Applicability: MODES 1 and 2
2.11 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (TS 3.4.1) 2.11.1 Tavg shall be 574°F. 2.11.2 Pressurizer pressure shall be maintained 2205 psig during operation.
NOTE: Pressurizer pressure limit does not apply during:
- 1) THERMAL POWER ramp
>5% RTP per minute; or
- 2) THERMAL POWER step
>10% RTP. 2.11.3 Reactor Coolant System raw measured Total Flow Rate shall be maintained 182,400 gpm.
Applicability: MODE 1 2.12 Refueling Boron Concentration (TS 3.9.1)
Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained 2300 ppm.
Applicability: MODE 6 POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT (COLR)
I UNIT 2 CYCLE 29 TRM 2.1 U2 Revision 9 October 27,2006 FIGURE 1 REACTOR CORE SAFETY LIMITS CURVE (Cores containing 422V+ fuel) Page 7 of 15
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October 27, 2006 Page 9 of 15 FIGURE 3 CONTROL BANK INSERTION LIMITS NOTE: The "fully withdrawn" parking position range 225 steps can be used without violating this Figure.
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
CORE OPERATING LIMITS REPORT (COLR) UNIT 2 CYCLE 29 TRM 2.1 U2 Revision 9
October 27, 2006 Page 10 of 15 FIGURE 4 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE (K(Z)) FOR 422V+ FUEL
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
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October 27, 2006 Page 11 of 15 FIGURE 5 RAOC Summary of W(Z) with HFP AFD Band of -8/+9 % (Top 15% and Bottom 12% Excluded)
W(Z) Height (feet) 150 MWD/MTU 6000 MWD/MTU 10000 MWD/MTU 14000 MWD/MTU 0.0 1.0000 1.0000 1.0000 1.0000 0.2 1.0000 1.0000 1.0000 1.0000 0.4 1.0000 1.0000 1.0000 1.0000 0.6 1.0000 1.0000 1.0000 1.0000 0.8 1.0000 1.0000 1.0000 1.0000 1.0 1.0000 1.0000 1.0000 1.0000 1.2 1.0000 1.0000 1.0000 1.0000 1.4 1.0000 1.0000 1.0000 1.0000 1.6 1.2615 1.1888 1.1643 1.1692 1.8 1.2431 1.1749 1.1528 1.1580 2.0 1.2247 1.1600 1.1405 1.1462 2.2 1.2065 1.1445 1.1277 1.1339 2.4 1.1880 1.1291 1.1152 1.1218 2.6 1.1695 1.1137 1.1050 1.1096 2.8 1.1515 1.0976 1.1005 1.0966 3.0 1.1333 1.0928 1.0979 1.0932 3.2 1.1201 1.0923 1.0961 1.0927 3.4 1.1162 1.0930 1.0945 1.0920 3.6 1.1135 1.0937 1.0927 1.0961 3.8 1.1127 1.0942 1.0904 1.1001 4.0 1.1122 1.0946 1.0883 1.1037 4.2 1.1112 1.0951 1.0892 1.1068 4.4 1.1102 1.0965 1.0934 1.1092 4.6 1.1089 1.0984 1.0975 1.1111 4.8 1.1073 1.0999 1.1012 1.1142 5.0 1.1053 1.1009 1.1042 1.1172 5.2 1.1028 1.1017 1.1070 1.1196 5.4 1.1005 1.1033 1.1088 1.1210 5.6 1.1006 1.1070 1.1101 1.1219 5.8 1.1053 1.1129 1.1193 1.1310 6.0 1.1156 1.1191 1.1336 1.1437 6.2 1.1279 1.1284 1.1482 1.1550 6.4 1.1391 1.1393 1.1615 1.1653 6.6 1.1496 1.1491 1.1741 1.1748 6.8 1.1589 1.1579 1.1855 1.1830 POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
CORE OPERATING LIMITS REPORT (COLR)
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October 27, 2006 Page 12 of 15 FIGURE 5 (con't) RAOC Summary of W(Z) with HFP AFD Band of -8/+9 % (Top 15% and Bottom 12% Excluded)
W(Z) Height (feet) 150 MWD/MTU 6000 MWD/MTU 10000 MWD/MTU 14000 MWD/MTU 7.0 1.1670 1.1654 1.1956 1.1901 7.2 1.1738 1.1722 1.2043 1.1967 7.4 1.1791 1.1789 1.2114 1.2028 7.6 1.1830 1.1848 1.2169 1.2073 7.8 1.1852 1.1892 1.2206 1.2101 8.0 1.1857 1.1922 1.2223 1.2111 8.2 1.1843 1.1937 1.2220 1.2103 8.4 1.1811 1.1935 1.2196 1.2076 8.6 1.1760 1.1918 1.2148 1.2030 8.8 1.1671 1.1872 1.2081 1.1962 9.0 1.1657 1.1860 1.1977 1.1958 9.2 1.1722 1.1904 1.1894 1.2002 9.4 1.1774 1.1967 1.1883 1.2016 9.6 1.1855 1.2058 1.2003 1.2073 9.8 1.1993 1.2176 1.2131 1.2152 10.0 1.2140 1.2304 1.2240 1.2240 10.2 1.2267 1.2418 1.2339 1.2322 10.4 1.0000 1.0000 1.0000 1.0000 10.6 1.0000 1.0000 1.0000 1.0000 10.8 1.0000 1.0000 1.0000 1.0000 11.0 1.0000 1.0000 1.0000 1.0000 11.2 1.0000 1.0000 1.0000 1.0000 11.4 1.0000 1.0000 1.0000 1.0000 11.6 1.0000 1.0000 1.0000 1.0000 11.8 1.0000 1.0000 1.0000 1.0000 12.0 1.0000 1.0000 1.0000 1.0000 POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
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October 27, 2006 Page 13 of 15 FIGURE 6 FLUX DIFFERENCE OPERATING ENVELOPE 50 55 60 65 70 75 80 85 90 95 100 105-35.00-30.00-25.00-20.00-15.00-10.00-5.000.005.0010.0015.0020.0025.0030.0035.00Delta I (%)Power (%)(9, 100)(-8, 100)(9, 90)(-8, 90)(-28, 50)(27, 50) NOTE: This figure represents the Relaxed Axial Offset Control (RAOC) band used in safety analyses, it may be administratively tightened depending on in-core flux map results. Refer to Figure 2 of ROD 1.2 for the administrative limit.
POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
CORE OPERATING LIMITS REPORT (COLR)
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October 27, 2006 Page 14 of 15 TABLE 1 NRC APPROVED METHODOLOGIES FOR COLR PARAMETERS COLR Section Parameter NRC Approved Methodology 2.1 Reactor Core Safety Limits WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.2 Shutdown Margin WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.3 Moderator Temperature Coefficient WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.4 Shutdown Bank Insertion Limit WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.5 Control Bank Insertion Limits WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.6 Height Dependent Heat Flux Hot Channel Factor (F Q) WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control F Q Surveillance Technical Specification," February 1994 WCAP-14449-P-A, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," Revision 1, October 1999 (cores containing 422V + fuel)
WCAP-10924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addenda, December 1988 (cores not containing 422V + fuel)
WCAP-10924-P-A, "LBLOCA Best Estimate Methodology: Model Desc ription and Validation: Model Revisions," Volume 1, Addendum 4, August 1990 (cores not containing 422V + fuel)
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," July 1997 POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL
CORE OPERATING LIMITS REPORT (COLR)
UNIT 2 CYCLE 29 TRM 2.1 U2
Revision 9
October 27, 2006 TABLE 1 NRC APPROVED METHODOLOGIES FOR COLR PARAMETERS Page 15 of 15 COLR Section Parameter NRC Approved Methodology 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F NH) WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 2.8 Axial Flux Difference (AFD) WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control F Q Surveillance Technical Specification," February 1994 2.9 Overtemperature T Setpoint WCAP-8745-P-A, "Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions," September 1986 2.10 Overpower T Setpoint WCAP-8745-P-A, "Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions," September 1986 2.11 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989, for those events analyzed using RTDP
WCAP-14787-P, Rev. 2, "Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 MWt-NSSS Power with Feedwater Venturis, or 1679 MWt-NSSS Power with LEFM on Feedwater Header)", October 2002.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 for those events not utilizing RTDP 2.12 Refueling Boron Concentration WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985