ML17229A925

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Proposed Tech Specs Revising Thermal Margin SL Lines of TS Figure 2.1-1 to Reflect Increase in Value of Design Min RCS Flow from 345,000 Gpm to 365,000 Gpm & Change Flow Rates Stated in Tables 2.2-1 & 3.2-1
ML17229A925
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 11/22/1998
From:
FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17229A923 List:
References
NUDOCS 9812010026
Download: ML17229A925 (23)


Text

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment RCS

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e 83 ST. LUCIE UNIT 1 MARKED-UPTECHNICALSPECIFICATION PAGES Page 1-3 Page 2-2 INSERT - A (Revised Figure 2.1-1)

Page 2-4 Page B 2-5 Page 3/4 2-14 Page 5-5 Page 6-19 Page 6-19a INSERT - B (Revised List of Analytical Methods for TS 6.9.1.11.b) ia~aoaaoai vszxiz >I, Pan AaOCX OSOOOSSS'm+

ICRP-30, Supplement to Part 1, pages 192-212, Tables entitled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of unit Activity(Sv/Bq)."

DEFINITIONS DOSE E UIVALENT I-131 1.lO DOSE EQUIVALENT I-131 shall be that concentration of I-131 {pCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.

The thyroid dose conversion fa tors use for this alculat on shall be those. listed in n--

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- AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration {in MEV) for

isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95>> of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES

RESPONSE

TIME 1.12 The ENGINEERED SAFETY FEATURES

RESPONSE

TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function {i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADMASTE TREATMENT SYSTEM 1.,14 A GASEOUS RADMASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of -reducing the total radioactivity prior to release to the environ-ment.

ST.

LUCIE - UNIT 1 1-3 Amendment No.Z7,

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575 550 525 UNACCEPTABLE OPERATION REACTOR OPERATIO IMITED TO LESS THAN 580 Deg F BY A UATION OF TIIE MAIN STEAM LINE SAFETY ALVES VESSEL FLOW LESS MEASUREME T UNCERTAINTIES 346,000 GPM I.IMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUAT S

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500 BASED ON TIIE AXIAL SHAP ON FIGURE B 2.l-l 475 450 PRESSURE IN PSIA 1750 2

225 2400 ACCEPTABLE OPERATION 0.2 0.4 0.6 0.8 l.

1.2 F~ract,ion of Rat.ed Thermal Power FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT FOUR REACTOR COOLING PUMPS OPERATING

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~4'ABLE 2.2-1 REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT I

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Hanual Reactor Trip 2.

Power Level - High (1)

Four Reactor Coolant Pumps Operating 3.

.Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps Operating TRIP SETPOINT Not Applicable

<< 9.61K above THEfNAL POWER, with a miniaxm setpoint of 15K of RATED THERNL POWER, and a maxilla of << 107.0X of RATED THERNL POWER.

of design reactor coolant Plow with l puwps operating*

ALLOWABLE VALUES Iht Applicable

<<9.61'X above THERMAL POWER, and a minimum setpoint of 15K of RAT THENAL POWER and a Naxism of

+ 107.0$ of RATEO THERNAL POWER.

. of design reactor coolant I

7)ow with 4 pueps operating~

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Pressurizer Pressure - High

<< 2400 psia

< 2i00 psia 5.

Contaireent Pressure - High

<<3+3 psig 6.

Steam Generator Pressure - Low (2)

> 600 psia 7.

Steaa Generator Mater Level -Low

> 20.5$ Mater Level - each steam generator

< 3.3 psig

> 600 psia

> 19.N Mater Level - each steaw generator rt C7 8.

Local Power Density - High (3)

Trip setpoint a+usted to not exceed the limit lines of Figures 2.2-1 and 2.2-2 Trip set point ad)usted to not exceed the 1&it lines of Figures 2.2-1 and 2;2-2.

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TABLE 3.2-1 DNB MARGIN LlMITS Parameter Cold Leg Temperature Pressurizer Pressure Reactor Coolant Flow Rate AXIALSHAPE INDEX Four Reactor Coolant Pumps Operating S 549'F 2 2225 psia gpm COLR Figure 3.2-4 Limit not applicable during either a THERMALPOWER ramp increase in excess of 5% of RATED THERMALPOWER or a THERMALPOWER step increase of greater than 10% of RATED THERMALPOWER.

ST. LUCIE - UNIT 1 3/4 2-14 Amendment No. &, 48, 480, 446

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DESIGN FEATURES

,CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies.

The control element assemblies shall be designed and maintained in accordance with the original design provisions contained in Section

4. 2. 3.2 of the FSAR with allowance for normal, degradation pursuant to the applicable Surveillance Requirements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

In accordance with the code. requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and C.

For a temperature of 650'F, except for the pressurizer which is 700'F.

VOLUISE 5.4.2 The total water and steam volume of the reactor coolant system is 11,100 180 cubic feet at a nominal T of 567'F, when not accounting for steam generator tube plugging.

avg 5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency. core cooling systems are designed and shal'l be main-tained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the appl,icable Surveillance Requirements.

5. 6 FUEL STORAGE CRITICALITY

+ less than or equal to 0.95 iffullyflooded with unborated water, which I includes an allowance for uncertainties as described in Section 9.1 of the

,. Updated Final Safety Analysis Report 5.6.l.a The spent fuel storage racks are designed and shall be maintained with:

1.

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l.UCIE - UNIT 1

5-5 Amendment Nn. 22.27 7%, PP,'~1

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment RCS u

lo and SU date SERT-B 2

a es 3.

XN-75-27(A) and Supplements 1 through 5, (also issued as XN-NF-75-27(A)], "Exxon Nuclear Neutronic(s) Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Inc.

/Advanced Nuclear Fuels Corporation, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P) 4.

ANF-84-73(P)(A) Revision 5, Appendix B, 8 Supplements 1 and 2, "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, October 1990 5.

XN-NF-82-21(P)(A) Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Inc., September 1983 6.

a)

ANF-84-93(P)(A) and Supplement 1, [also issued as XN-NF-84-93(P)(A)], "Steamline Break Methodology for PWRs," Advanced Nuclear Fuels Corporation, March 1989 b)

EMF-84-093(P) Revision 1, "Steam Line Break Methodology for PWRs," Siemens Power Corporation, June 1998 (This document is a Revision to ANF-84-93) 7.

XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, Inc., October 1983 8.

Siemens Power Corporation Small Break LOCA methodology as defined by:

a)

XN-NF-8249(P)(A) Revision 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Advanced Nuclear Fuels Corporation, April 1989 b)

XN-NF-82-49(P)(A) Revision 1 Supplement 1, "Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model," Siemens Power Corporation, December 1994 9.

XN-NF-78%(NP)(A), "AGeneric Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Inc., October 1983 10.

XN-NF-621(P)(A) Revision 1, "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Inc., September 1983 11.

EXEM PWR Large Break LOCA Evaluation Model as defined by:

a) 1.

XN-NF-82-20(P)(A) Revision 1

Supplement 2, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Inc., February 1985 (Cog f P.

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. ADMINISTRATIVECO LS NtI AL RADIOLOGICALENVIRONMENTALOPERATIN REP RT (continued)

. 6.9.1.9 6.9.1.10 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shalt be prepared and submitted to the NRC.

At least once every 10 years, an estimate of the actual population within 50 miles of the plant shall be prepared and submitted to the NRC.

6.9.1.11 CORE OPERATING LIMITS REPORT OLR a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

Specification Specification Specification Specification Specification Specification Specification 3.1

~1.4 3.1.3.1 3.1.3.6 3.2.1 3.2.3 3.2.5 3.9.1 Moderator Temperature Coefficient Full Length CEA Position - Misalignment ) 15 inches Regulating CEA Insertion Limits Linear Heat Rate Total Integrated Radial Peaking Factor - F T r

DNB Parameters Refueling Operations - Boron Concentration b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in the following documents or any approved Revisions and Supplements thereto:

1.

WCAP-11596-P-A, "Qualification of the 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 of Turkey Point 8 St. Lucie Nuclear Plants," Florida Power 8 Light Company, January 1995.

3.

N-75-27(A), Rev. 0 and Supplement 1 through 5, "Exxon Nuclear

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Ne onics Design Methods for Pressurized Water Reactors," ~rf Nuclea ompany, Rev. 0 dated June 1975, Supplement 1 dyte8 September 76, Supplement 2 dated December 1980, gupplement 3 dated Septemb r~1981, Supplement 4 dated Decem r 1986, Supplement 5 dated ebruary 1987.

4.

ANF-84-73(P), Rev. 3, "Adv nqad Nucle uels Methodology for Pressurized Water Reactors: Anal i

Chapter15 Events,"Advanced Nuclear Fuel Corporation, dated y

88.

5.

XN-NF-82-21(A), Rev. 1g pplication of Exxo uclear Company PWR Thermal Margin Megto8ology to Mixed Core Con rations," Exxon Nuclear Compa y, dated September 1983.

6.

ANF A), Rev. 0 and Supplement 1, "Steamline Break Mept ology for PWR's," Advanced Nuclear Fuels Corporation, Re QCPed March 1 989, Supplement 1 dated March 1989.

ST. LUCIE-UNIT 1 6-19 Amendment No. 69, 69, 86,

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- CORE OPERATING LIMITS REPORT (continued)

XN-75-32(A), Supplements 1, 2, 3, and 4, "Computational Procedure f Evaluating Fuel Rod Bowing,'xxon Nuclear Company, dated Octo r

1983.

8.

X F-82<9(A), Rev.

1 and Supplement'1, "Exxon Nuclear mpany Eval qtion Model EXEM PWR Small Break Model," Advan d Nuclear Fuels oqrporation, Rev.

1 dated April 1989, Supplement dated December 1994.

9.

XN-NF.-78

), 'A Generic Analysis of the Contro od Ejection Transient for P qssurized Water Reactors,'xxo uclear Company, dated October 10 3.

10. XN-NF-621(A), Rev.

'Exxon Nuclear Dg Correlation of PWR Fuel Design," Exxon Nuclea Company, date September 1983.

11.

EXEM PWR Large Break L CA Ev I ation Model as defined by:

a)

XN-NF-82-20(A), Rev.1 an Supplements1 through 4, "Exxon Nuclear Company Evaluation odel EXEM/PWR ECCS Model Updates," Exxon Nucle fComp ny, all dated January 1990.

b)

XN-NF-82-07(A), ge

. 1, 'Exxon N lear Company ECCS Cladding Swelling and Ru ture Model,'xxon uclear Company, dated November 198 c)

XN-NF 8(A), Rev. 2 and Supplements through 4, 'RODEX2 Fuel Rog ermal-Mechanical Response Eva ation Model,'xxon Nucle I'Company, Rev. 2 and Supplement 1 ah 2 dated March 19, Supplements 3 and 4 dated June 1990.

d)

-NF-85-16(A), Volume 1 through Supplement 3; Volume 2, Rev.

1 and Supplement 1, 'PWR 17x17 Fuel Cooling Thats Program," Exxon Nuclear Company, all dated February 10 0.

e)

XN-NF-85-105(A), Rev. 0 and Supplement 1, "Scaling of FCT Based Reflood Heat Transfer Correlation for Other Bundle Desi qs,'xxon Nuclear Company, all dated January 1990.

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

'PECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.

ST. LUCIE-UNIT1 6-19a Amendment No.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment CS a

d SUdae NSER

- B Co t ued XN-NF-82-20(P)(A) Revision 1 and Supplements 1, 3 and 4, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Advanced Nuclear Fuels Corporation, January 1990 b)

XN-NF-82-07(P)(A) Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., November 1982 c) 1.

XN-NF-81-58(P)(A) Revision 2, and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical

Response

Evaluation Model," Exxon Nuclear Company, Inc., March 1984 ANF-81-58(P)(A) Revision 2 Supplement 3, and Supplement 4, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," Advanced Nuclear Fuels Corporation, June 1990 d)

XN-NF-85-16(P)(A) Volume 1, and Supplements 1, 2 and 3; Volume 2, Revision 1 and Supplement 1, "PWR 17x17 Fuel Cooling Test Program," Advanced Nuclear Fuels Corporation, February 1990 e)

XN-NF-85-105(P)(A) and Supplement 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Advanced Nuclear Fuels Corporation, January 1990 f)

EMF-2087(P) Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation, August 1998 12.

XN-NF-82-06(P)(A) Revision 1, and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Bumup," Exxon Nuclear Company, Inc., October 1986 13.

ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels'WR Design Methodology for Rod Bumups of 62 GWd/MTU,"Advanced Nuclear Fuels Corporation, December 1991 14.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, Inc., November 1986 15.

ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, May 1992 16.

XN-NF-507(P)(A), Supplements 1 and 2, "ENC Setpoint Methodology for C. E. Reactors: Statistical Setpoint Methodology," Exxon Nuclear Company, Inc., September 1986

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