ML17241A440
| ML17241A440 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/18/1999 |
| From: | Peterson S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17241A441 | List: |
| References | |
| NUDOCS 9908240122 | |
| Download: ML17241A440 (15) | |
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UNITED STATFS NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20055-mMI FLORIDA POWER 8 LIGHTCOMPANY DOCKET NO. 50-335 ST. LUCIE PLANT UNIT NO. 1 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 163 License No. DPR-67 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power & Light Company, et al., (the licensee), dated November 22, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facilitywilloperate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health'and safety of the public, and (ii) that such activities willbe conducted in compliance with the Commission's regulations; D.
The issuance of this amendment willnot be inimical to the. common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with.10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9908240i222 990818 POR AQQCK P
ll 2.
Accordingly, Facility Operating License No. DPR-67 is amended by changes to th Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.(2) to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 163, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORYCOMMISSION Sheri R. Peterson, Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 18, 1999
ATTACHMENTTO LICENSE AMENDMENTNO. 163 TO FACILITYOPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace the following pages of the Appendix "A"Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
1-3 2-2 2-4 B 2-5 3/4 2-14 5-5 6-19 6-19a n/a 1-3 2-2 2-4 B 2-5 3/4 2-14 5-5 6-19 6-19a 6-19b
DEFINITIONS DOSE EQUIVALENTI-131 1.10 DOSE EQUIVALENT1-131 shall be that concentration of I-131 (pCI/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of l-131, l-132, l-133, I-134 and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in 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)."
E - AVERAGE DISINTEGRATIONENERGY 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 REPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation.hetpoint 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.
FRE UENCY NOTATION 1.13
. The FREQUENCY NOTATIONspecified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
GASEOUS RADWASTE TREATMENTSYSTEM 1
~ 14 A GASEOUS RADWASTE TREATMENTSYSTEM 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 environment.
ST. LUCIE-UNIT1 1-3 Amendment No. &7,69, 163
'1
600 580 4
4 560 D
CL 540 ct 520 O
E E
500 4
O UNACCEPTABLE OPERATION
~,
REACTOR OPFRATION LIMITED TO I ESS THAN 560 F BY ACTUATION OF THE MAIN STEAM LINE SAFETY VALVES VESSEL FLOW LESS MEASUREMENT UNCERTAINTIES 365,000 GPM LIMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUATIONS BASED ON THE AXIAL SHAPE ON FIGURF 8 2.1-1 UNACCEPTABLE OPERATION RESSURE IN PSIA 2400 2225 2000 1750 460 ACCEPTABLE OPERATION THERMAL POWER LIMITED TO A MAXIMUM OF 112K+ OF RATED THERMAL POWER BY THE HIGH POWER TRIP LEVE
.0
.2
.6
.8 1.0 Fraction of Rated Thermal Power 1.2 1.4 1.6 (DIOPS/TS(UhllT1 2-2)RO)
FIGURE 2.1-1: REACTOR CORE THERMALMARGINSAFETY LIMIT-FOUR REACTOR COOLANT PUMPS OPERATING ST. LUCIE-UNIT 1 2-2 Amendment No. 4e, +39, 446> 163
0 t
TABLE2.2-1 REACTOR PROTECTIVE INSTRUMENTATIONTRIP SETPOINT LIMITS FUNCTIONALUNIT 1.
Power Level High (1)
Four Reactor Coolant Pumps Operating 3.
Reactor Coolant Flow Low (1)
Four Reactor Coolant Pumps Operating 4.
Pressurizer Pressure High 5.
Containment Pressure High 6.
Steam Generator Pressure Low (2) 7.
Steam Generator Water Level Low 8.
Local Power Density High (3)
TRIP SETPOINT Not Applicable
< 9.61% above THERMALPOWER, with a minimum setpoint of 15% of RATED THERMALPOWER, and a maximum of
< 107.0% of RATED THERMAL POWER.
> 95% of design reactor coolant flow with 4
um so eratin
< 2400 psia
< 3.3 pslg
> 600 psia
> 20.5% Water Level each steam generator Trip setpoint adjusted to not exceed the limitlines of Figures 2.2-1 and 2.2-2.
ALLOWABLEVALUES Not Applicable
< 9.61% above THERMALPOWER, and a minimum setpoint of 15% of RATED THERMALPOWER and a maximum of
< 107.0% of RATED THERMAL POWER.
> 95% of design reactor coolant flowwith 4
um so eratin
< 2400 psia
< 3.3 pslg
> 600 psia
> 19.5% Water Level each steam generator Trip set point adjusted to not exceed the limitlines of Figures 2.2-1 and 2.2-2.
- Design reactor coolant flowwith 4 pumps operating is 365,000 gpm.
ST. LUCIE-UNIT 1 2-4 Amendment No.e, &7,8&,45,~,+R,+45, 163
2.2 LIMITINGSAFETY SYSTEM SETTINGS BASES
- Reactor Coolant Flow'-Low (continued) relationship between steam generator ciifferential pressure, core inlet temperature, instrument errors and response times. When the calculated RCS flowfalls below the trip setpoint an automatic reactor trip signal is initiated. The trip setpoint and allowable values ensure that for a degradation of RCS flow resulting from expected transients, a reactor trip occurs to prevent violation of local power density or DNBR safety limits.
Pressurizer Pressure-Hi h
The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal liftsetting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.
Containment Pressure-Hi h
The Containment Pressure High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.
Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.
The setting of 600 psia is sufficiently below the full-load operating point so as not
'I ST. LUCIE - UNIT 1 B 2-5 Amendment No. 88, 45, 48, 88,
~ 163
TABLE3.2-'I DNB MARGIN LIMITS Parameter Cold Leg Temperature Pressurizer Pressure Reactor Coolant Flow Rate AXIALSHAPE INDEX Four Reactor Coolant Pumps Operating
< 549'F
> 2225 psia *
> 365,000 gpm COLR Figure 3.2-4 Limitnot 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. W, 48, ~,
~,~,
163
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:
a.
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 VOLUME 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~
5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
" 5.6 FUEL STORAGE CRITICALITY 5.6.1.a The spent fuel storage racks are designed and shall be maintained with:
1.
k, less than or equal to 0.95 iffullyflooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.
ST. LUCIE - UNIT 1 5-5 Amendrpggt No. 22, W, W, e+,
ADMINISTRATIVECONTROLS ANNUALRADIOLOGICALENVIRONMENTALOPERATING REPORT (continued) 6.9.1.9 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shall be prepared and submitted to the NRC.
6.9.1.10 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 LIMITSREPORT COLR 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 3.1.1.4 Moderator Temperature Coefficient Specification 3.1.3.1 Full Length CEA Position Misalignment > 15 inches Specification 3.1.3.6 Regulating CEA Insertion Limits Specification 3.2.1 Linear Heat Rate Specification 3.2.3 Total Integrated Radial Peaking Factor-F Specification 3.2.5 DNB Parameters I'pecification 3.9.1
~ 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:
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 & St. Lucie Nuclear Plants," Florida Power & Light Company, January 1995.
3.
XN-75-27(A) and Supplements 1 through 5, [also issued as XN-NF-75-27(A)], "Exxon Nuclear Neutronic(s) Desig'n 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 Septembet'981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P) 4.
ANF-84-73(P)(A) Revision 5, Appendix B, & Supplements 1 and 2, "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, October 1990 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 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 ST. LUCIE - UNIT 1 6-19 Amendment No. 59, 69, 66, ~,
~,163
ADMINISTRATIVECONTROLS CORE OPERATING LIMITSREPORT (continued) 6.
b)
EMF-84-093(P)(A) Revision 1, "Steam Line Break Methodology for PWRs," Siemens Power Corporation, February 1999 (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) b)
XN-NF-82-49(P)(A) Revision 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Advanced Nuclear Fuels Corporation, April 1989 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-78-44(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 2.
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 ST. LUCIE - UNIT 1 6-19a Amendment No.~, 163
ADMINISTRATIVECONTROLS CORE OPERATING LIMITSREPORT (continued) 11.
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)(A) Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCAApplications," Siemens Power Corporation, June 1999.
12.
XN-NF-82-06(P)(A) Revision 1, and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," 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 Burnups 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 C.
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 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.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.
ST. LUCIE - UNIT 1 6-19b Amendment No," pe