L-2011-021, License Amendment Request for Extended Power Uprate, Attachment 8; Core Operating Limits Report (COLR) Markups

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License Amendment Request for Extended Power Uprate, Attachment 8; Core Operating Limits Report (COLR) Markups
ML110730311
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 02/25/2011
From:
Florida Power & Light Co
To:
Office of Nuclear Reactor Regulation
References
L-2011-021
Download: ML110730311 (13)
v • d • e


Text

{{#Wiki_filter:St. Lucie Unit 2 L-2011-021 Docket No. 50-389 Attachment 8 ATTACHMENT 8 LICENSE AMENDMENT REQUEST EXTENDED POWER UPRATE CORE OPERATING LIMITS REPORT MARKUPS (For Information Only) FLORIDA POWER & LIGHT ST. LUCIE UNIT 2 This coversheet plus 12 pages St. Lucie Unit 2 EPU LAR Att. 8-1 Core Operating Limits Report Markups

St. Lucie Unit 2 L-2011- Docket No. 50-389 Attachment 8 LIST OF PAGES Core Operating Limits Report 5 of 18 6 of 18 8 of 18 (replaced by Proposed COLR Figure 3.1-1a) Proposed COLR Figure 3.1-1a 12 of 18 (replaced by Proposed COLR Figure 3.2-3) Proposed COLR Figure 3.2-3 14 of 18 15 of 18 16 of 18 17 of 18 18 of 18 St. Lucie Unit 2 EPU LAR Att. 8-2 Core Operating Limits Report Markups

PC/M 08033, Rev. 0 Attachment 14, Page 5 of 18 heat rate LHRM(z). The incremental penalty factors (in excess of 2% margin decrease) are included in the W(z) function of Table 3.2-3. Incore Detector Monitoring System During operation, with the linear heat rate being monitored by the Incore Detector Monitoring System, the Local Power Density alarm setpoints shall be adjusted to less than or equal to the limits shown on Figure 3.2-1. 2.5 TOTAL INTEGRATED RADIAL PEAKING FACTOR - F,I (TS 3.2.3) 1.60 The calculated value of FrT shall be limited to < 1.70. The power dependent FrT limits are shown on Figure 3.2-3. 2.6 DNB Parameters (TS 3.2.5) The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-2: Total

a. Cold Leg Temperature
b. Pressurizer Pressure
c. Reactor Coolant System7R 7Flow rate
d. AXIAL SHAPE INDEX 2.7 Refueling Operations - Boron Concentration (TS 3.9.1)

With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: 1900

a. Either a Kef! of 0.95 or less, or
b. A boron concentration of greater than or equal to 1720 ppm.

2.8 SHUTDOWN MARGIN - Ta.J!g Greater Than 200 OF (TS 3.1.1.1) The SHUTDOWN MARGIN shall be greater than or equal to 3600 pcm. 2.9 SHUTDOWN MARGIN - T!'.l!f! Less Than or Equal To 200 OF (TS 3.1.1.2) The SHUTDOWN MARGIN shall be greater than or equal to 3000 pcm. st. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 5 of 18

PC/M08043,Rev.0 Attachment{4, Page Gof t8 Table3.2-2 DNB MARGIN LIMITS PARAME.TER FOURREACTOR COOLANT PUMPSOPERAIfNG 551 Coid Leg Temperahre(narrowRange) 53j0F+* s T s 54g0F PressurizerPressure* 2225 psia S Ppzn< 2350psia** System Total Reactor Coolant Flow Rate ) 375,000gpm AXIAL S}IAPE INDEX Within the limits specifiedin Figure 3.2-4

  • Limit not applicableduring eithera THERMAL POWERramp inmeasein excessof Syoof RATED THERMAL POWERor a TI{ERMAL POWERstepincreaseof greaterthan l1yo of RATED THERI\4AL POWER.
  • !r Applicableonly if powerlevel >70%of RATED TI{ERMAL powER.

St. Lucie Unit 2 Cycle,lSCOLR Rev 0 Page 6 of 18

PC/M 08033, Rev. 0 Attachment 14, Page 8 of 18 o 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 Time at Full Power to Realign CEA, Minutes FIGURE 3.1-1a Allowable Time to Realign CEA vs. Initial FrT St. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 8 of 18

Proposed COLR Figure 3.1-1a 1.62 (63, 1.60) 1.60 1.58 Measured Pre-Misaligned FrT 1.56 1.54 1.52 1.50 1.48 1.46 1.44 1.42 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 Time at Full Power to Realign CEA, Minutes FIGURE 3.1-1a Allowable Time to Realign CEA vs. Initial FrT St. Lucie Unit 2 - Proposed EPU COLR

PC/M 08(Jj3, Rev. 0 Attachment 14, Page 12 of 18

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                       *              = 1.70,.............
                              ................        [1 + 0.4"     (1 - P)],-l- _or ---:- --1-- -c---:-- --i' - - -, -~-'-
                                                            ,.........................      ....                                     . . . .-i'!.'r- --{--                  ..     -,--.-:---{---    .

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REGION 0.7

t::j:::t:::t::j:::t:::t::j:::~ . . . . ' .. - j - .. - ~ .....:- .... ~ ...... ~ .... 1..... ~ .... - ~ .. - .:- ... ~ ......
:j::: j::: t:::j::: :1::: t:::t:: j::: t: ::~:::l::: j:::t:: :j::: j:: :t:: :j::: j:::t:::I::: j:::t:::l::: j:::
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0.6 1.65 1.7 1.75 T 1.8 1.85 1.9 Measured Fr FIGURE 3.2-3 Allowable Combinations of THERMAL POWER and F,T (The expression specified in the Figure may be used for F/ at other power levels) st. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 12 of 18

Proposed COLR Figure 3.2-3 1.2 UNACCEPTABLE 1.1 OPERATION REGION Allowable Fraction of (1.60, 1.0) 1.0 0.9 RATED THERMAL POWER (P) FrT LIMIT CURVE FrT = 1.60 * [ 1 + 0.4 * (1 - P)] (1.728, 0.80) 0.8 ACCEPTABLE OPERATION REGION 0.7 0.6 1.55 1.6 1.65 T 1.7 1.75 1.8 Measured Fr FIGURE 3.2-3 Allowable Combinations of THERMAL POWER and FrT (The expression specified in the Figure may be used for FrT at other power levels) St. Lucie Unit 2 - Proposed EPU COLR

PC/M 08033, Rev. 0 Attachment 14, Page 14 of 18 3.0 LIST OF APPROVED METHODS The analytical methods used to determine the core operating limits are those previously approved by the NRC, and are listed below.

1. WCAP-11596-P-A, "Qualification ofthe PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary)
2. NF-TR-95-01, "Nuclear Physics Methodology for Reload Design ofTurkey Point &St.

Lucie Nuclear Plants," Florida Power & Light Company, January 1995 (NRC SER dated June 9, 1995), & Supplement 1, August 1997

3. CENPD-199-P, Rev. 1-P-A, "C-E Setpoint Methodology: CE Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986
4. CENPD-266-P-A, "The ROCS and DIT Computer Code for Nuclear Design," April 1983
5. CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania BumableAbsorbers," May 1988, & Revision 1-P Supplement 1-P-A, DELETED April 1999
6. CENPD-188-A, "HERMITE: A Multi-Dimensional Space - Time Kinetics Code for PWR Transients," July 1976
7. CENPD-153-P, Rev. 1-P-A, "Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System," May 1980
8. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 1: C-E Calculated Local Power Density and Thermal MarginlLow Pressure LSSS for St.

Lucie Unit 1," December 1979 .

9. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 2:

Combination of System Parameter Uncertainties in Thermal Margin Analyses for St. Lucie Unit 1," January 1980

10. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for St. Lucie Unit 1," February 1980
11. CEN-191 (B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981 St. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 14 of 18

PC/M 08033, Rev. 0 Attachme.nt 14, Page 15 of 18

12. Letter, J. W. Miller (NRC) to J. R. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No.8 to NPF-16 and SER), November 9, 1984 (Approval of CEN-123(F)-P (three parts) and CEN-191(B)-P)
13. CEN-371 (F)-P, "Extended Statistical Combination of Uncertainties," July 1989
14. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), Docket No. 50-389, "St. Lucie Unit 2 - Change to Technical Specification Bases Sections '2.1.1 Reactor Core' and
             '3/4.2.5 DNB Parameters' (TAC No. M87722)," March 14, 1994 (Approval of CEN-371 (F)-P)
15. CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986 DELETED
16. CENPD-162-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution," April 1975
17. CENPD-207-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-uniform Axial Power Distribution," December 1984
18. CENPD-206-P-A, "TORC Code, Verification and Simplified Modeling Methods," June 1981
19. CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983
20. CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," July 1974
21. CEN-161 (B)-P-A, "Improvements to Fuel Evaluation Model," August 1989
22. CEN-161 (B)-P, Supplement 1-P-A, "Improvements to Fuel Evaluation Model,"

January 1992

23. CENPD-132, Supplement 3-P-A, "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and WDesigned NSSS," June 1985
24. CENPD-133, Supplement 5-A, "CEFLASH-4A, A FORTRAN77 Digital Computer Program for Reactor Blowdown Analysis," June 1985
25. CENPD-134, Supplement 2-A, "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985
26. CENPD-135-P, Supplement 5, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 st. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 15 of 18

PC/M 08033, Rev. 0 Attachment 14, Page 16 of 18

27. Letter, R. L. Baer (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-135, Supplement #5," September 6, 1978
28. CENPD-137, Supplement 1-P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977
29. CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977
30. Letter, K. Kniel (NRC) to A. E. Scherer (CE), "Evaluation ofTopical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P," September 27,1977
31. CENPD-138, Supplement 2-P, "PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977
32. Letter, C. Aniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-138, Supplement 2-P," April 10, 1978
33. Letter, W. H. Bohlke (FPL) to Document Control Desk (NRC), "St. Lucie Unit 2, Docket No. 50-389, Proposed License Amendment, MTC Change from -27 pcm to-30 pcm," L-91-325, December 17,1991
34. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), "St. Lucie Unit 2 - Issuance of Amendment Re: Moderator Temperature Coefficient (TAC No. M82517)," July 15, 1992
35. Letter, J. W. Williams, Jr. (FPL) to D. G. Eisenhut (NRC), "St. Lucie Unit No.2, Docket No. 50-389, Proposed License Amendment, Cycle 2 Reload," L-84-148, June 4, 1984
36. Letter, J. R. Miller (NRC) to J. W. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No.8 to NPF-16 and SER), November 9, 1984 (Approval of Methodology contained in L-84-148)
37. Letter, A. E. Scherer Enclosure 1-P to LD-82-0D1, "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981 DELETED 38. Safety Evaluation Report, "CESEC Digital Simulation of a Combustion Engineering Steam Supply System (TAC No.: 01142)," October 27, 1983
39. CENPD-282-P-A, Volumes 1,2, and 3, and Supplement 1, "Technical Manual for the CENTS Code," February 1991, February 1991, October 1991, and June 1993, respectively st. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 16 of 18

PC/M 08033, Rev. 0 Attachment 14, Page 17 of 18

40. CEN-121 (B)-P, "CEAW, Method of Analyzing Sequential Control Element Assembly Group Withdrawal Event for Analog Protected Systems," November 1979 (NRC SER dated December 21,1999, Letter K. N. Jabbour (NRC) to T. F. Plunkett (FPL), TAC No. MA4523)
41. CEN-133(B), "FIESTA, A One Dimensional, Two Group Space-Time Kinetics Code for Calculating PWR Scram Reactivities," November 1979 (NRC SER dated December 21,1999, Letter K. N. Jabbour (NRC) to T. F. Plunkett (FPL), TAC No.

MA4523)

42. CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1991
43. CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990
44. CENPD-183-A, "C-E Methods for Loss of Flow Analysis," June 1984
45. CENPD-190-A, "C-E Method for Control Element Assembly Ejection Analysis," July 1976
46. CENPD-199-P, Rev. 1-P-A, Supplement 2-P-A, "CE Setpoint Methodology", June 1998
47. CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Bumable Absorbers," August 1993 DELETED
48. CEN-396(L)-P, "Verification of the Acceptability of a i-Pin Bumup Limit of 60 MWD/KG for St. Lucie Unit 2," November 1989 (NRC SER dated October 18, 1991, Letter J. A. Norris (NRC) to J. H. Goldberg (FPL), TAC No. 75947)
49. CENPD-269-P, Rev. 1-P, "Extended Bumup Operation of Combustion Engineering PWR Fuel," July 1984
50. CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2,"

December 1984 (NRC SER dated December 21, 1999, Letter K. N. Jabbour (NRC) to T. F. Plunkett (FPL), TAC No. MA4523)

51. CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998
52. CENPD-140-A, "Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," June 1976
53. CEN-365(L), "Boric Acid Concentration Reduction Effort, Technical Bases and Operational Analysis," June 1988 (NRC SER dated March 13, 1989, Letter J. A.

Norris (NRC) to W. F. Conway (FPL), TAC No. 69325) St. Lucie Unit 2 Cycle 18 COLR Rev 0 Page 17 of 18

PC/M 08033, Rev. 0 ~ Attachment 14, Page 18 of 18

54. DP 468, F. M. Stern (CE) to E. Case (NRC), dated August 19, 1074, Appendb( 68 to CESSl\R System 80 PSAR (NRC SER, NUREG 76/112, Doel<:et No. STN 60 470, "NRC SER SmndaFd RefeFOnee System, GESSAR System 80/' December 1076)
55. CENPD-387-P-A, Revision 000, "ABB Critical Heat Flux Correlations for PWR Fuel,"

May 2000

56. CENPD-132, ~upplement 4-P-A, "Calculative Methods for the CE Nuclear Power large Break LOCA Evaluation Model," March 2001
57. CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998
58. CENPD-404-P-A, Rev. 0, "Implementation of ZIRLO' Cladding Material in CE Nuclear Power Fuel Ass~mbly Designs," November 2001
59. WCAP-9272-P-A, *Westinghouse Reload Safety Evaluation Methodology July 1985
60. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control; FQ Surveillance Technical Specification," February 1994
61. WCAP-11397-P-A, (Proprietary), "Revised Thermal Design Procedure," April 1989 62~ WCAP-14565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, n October .

1999

63. WCAP-14565-P-A, Addendum 1-A, Revision 0, "Addendum 1 to WCAP-14565-P-A Qualification of ABS Critical Heat Flux Correlations with VIPRE-01 Code,') August 2004
64. Letter, W. Jefferson, Jr. (FPL) to Document Control Desk (USNRC)t uSt. Lucie Unit 2 bocket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L-2003-276, December, 2003 (NRC SER dated January 31,2005, Letter B. T. Moroney (NRC) to J. A. Stall (FPL), TAC No. MC1566)
65. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.
66. WCAP-7908-A, "FACTRAN-A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod,"

December, 1989.

67. WCAP-7979-P-A, "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," January 1975.
68. WCAP-7588, Rev 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods," January 1975.}}
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