ML20092C671
| ML20092C671 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/05/1995 |
| From: | Alexion T NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20092C674 | List: |
| References | |
| NUDOCS 9509130089 | |
| Download: ML20092C671 (15) | |
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p ara gi UNITED STATES NUCLEAR REGULATORY COMMISSION l
If WASHINGTON, D.C. 20565-0001
%*****j HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-498 SOUTH TEXAS PROJECT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 79 License No. NPF-76 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Houston Lighting & Power Company *
(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL),
and City of Austin, Texas (C0A) (the licensees), dated May 31, 1995, j
as supplemented by letter dated August 2, 1995, complies with the i
standards and requirements of the Atomic Energy Act of 1954, as i
amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; 4
C.
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 will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not 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.
- Houston Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
9509130089 950905 PDR ADOCK 05000498 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-76 is hereby amended to read as follows:
2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 79, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION vfuns<'/l1 J
Thomas W. Alexion, P oject Manager Project Directorate IV-I Division of Reactor Projects III/IV 1
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995 l
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[p no k UNITED STATES uq NUCLEAR REGULATORY COMMISSION o
f WASHINGTON, D.C. 20666-4001
(..... j/
HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-499 l
SOUTH TEXAS PROJECT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 License No. NPF-80 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Houston Lighting & Power Company *
(HL&P) acting on behalf of itself and for the City Public Service 4
Board of San Antonio (CPS), Central Power and Light Company (CPL),
and City of Austin, Texas (C0A) (the licensees), dated May 31, 1995, as supplemented by letter dated August 2, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as anended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not 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 Comission's regulations and all applicable requirements have been satisfied.
- Houston Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-80 is hereby amended to read as follows:
2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 68, and the Environmental Protection Plan 4
contained in Appendix B, are hereby incorporated in the license.
1 t
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION V
(h1f@'
l Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV i
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 5, 1995 i
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ATTACHMENT TO LICENSE NMENDMENT NOS. 79 AND 68 FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-89 DOCKET NOS. 50-498 AND 50-499 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 1-9 3/4 1-9 3/4 1-10 3/4 1-11 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 4-7 3/4 4-7 8 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 8 3/4 4-1 B 3/4 4-1 8 3/4 4-2 B 3/4 4-2 i
i Pages 3/41-9 through 3/41-15 have been deleted.
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SOUTH TEXAS - UNITS 1 & 2 3/4 1-9 Unit 1 - Amendment No. E,79 (Next page is 3/41-16) Unit 2 - Amendment No. H,68 1
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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES i
GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1. 3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.
APPLICABILITY: MODES 1* and 2*.
ACTION:
With one or more full-length rods inoperable due to being immovable a.
as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With one full-length rod trippable but inoperable due to causes i
other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than i 12 steps i
(indicated position), POWER OPERATION may continue provided that within 1 hour:
1.
The rod is restored to OPERABLE status within the above alignment requirements, or i
2.
The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits as specified in the Core Operating Limits Report (COLR).
The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.
The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.
POWER OPERATION may then continue provided that; a)
A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall con-firm that the previously analyzed results of these acci-dents remain valid for the duration of operation under these conditions; b)
The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; QSce Special Test Exceptions Specifications 3.10.2 and 3.10.3.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-16 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17 SEP S 1991
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i Page 3/4 4-7 has been deleted.
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4 SOUTH. TEXAS - UNITS 1 & 2 3/4 4-7 Unit 1 - Amendment No. 79 l
Unit 2 - Amendment No. 68 L
REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Co setting of 2485 psig i 2%,de safety valves shall be OPERABLE with a lift l
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification t
4.0.5.
'The lift setting pressure shall correspond to ambient conditions of the valve l at nominal operating temperature and pressure.
'The as left lift setting shall be within 11% following valve testing.
l SOUTH TEXAS'- UNITS 1 & 2 3/4 4-8 Unit 1 - Amendment No. 69,78 Unit 2 - Amendment No. 47,67 325Y
3/4.1 REACTIVITY CONTROL SYSTEMS BASES
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3/4.1.1 BORATION CONTROL i
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3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN I
A sufficient SHUTDOWN MARGIN ensures that:
(1) the reactor can be made j
subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
l SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T t no load operating In MODES I and 2, the 4
mostrestrictiveconditionoccursatEOL,withT,7teamlinebreakaccidentand a
3 temperature, and is associated with a postulated i
resulting uncontrolled RCS cooldown.
In the analysis of this accident, a i
minimum SHUTDOWN MARGIN OF 1.3% Ak/k is required to control the reactivity transient. The 1.3% Ak/k SHUTDOWN MARGIN is the design basis minimum for the 14-foot fuel using silver-indium-cadmium and/or Hafnium control rods (Ref.
FSAR Table 4.3-3).
Accordingly, the SHUTDOWN MARGIN requirement for MODES 1
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and 2 is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.
In MODES 3, 4, and 5, the most restrictive condition j
occurs at BOL, when the boron concentration is the greatest.
In these modes, l
the required SHUTDOWN MARGIN is composed of a constant requirement and a variable requirement, which is a function of the RCS boron concentration. The i
constant SHUTDOWN MARGIN requirement of 1.3% Ak/k is based on an uncontrolled l
RCS cooldown from a steamline break accident. The variable SHUTDOWN MARGIN requirement is based on the results of a boron dilution accident analysis, i
where the SHUTDOWN MARGIN is varied as a function of RCS boron concentration, to guarantee a minimum of 15 minutes for operator action after a boron j
dilution alarm, prior to a loss of all SHUTDOWN MARGIN.
l The boron dilution analysis assumed a common RCS volume, and maximum i
dilution flow rate for MODES 3 and 4, and a different volume and flow rate for l
MODE 5.
The MODE 5 conditions assumed limited mixing in the RCS and cooling i
with the RHR system only. The MODE 5 SHUTDOWN MARGIN curve (Figure 3.1-2) can be used to provide the required C in the rapid refueling condition (MODE 5 with ARO). The cycle-specific reload safety analysis verifies this curve to i
be bounding in this condition.
i 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.
I The MTC values of this specification are applicable to a specific set of
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plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 1-1 Unit 1 - Amendment No. 40 61 Unit 2-AmendmentNo.E,$0 L
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REACTIVITY CONTROL SYSTEMS l.
j-BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)
The most negative MTC value, equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MOC used in the FSAR analysis to nominal operating conditions. These corrections involved:
(1) a conversion of the MDC used in the FSAR analysis to its equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this i
value the largest differences in MTC observed at E0L, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in nominal operation and lead to a significantly more negative E0L MTC at RATED THERMAL POWER. These corrections transformed the MDC values used in the FSAR analysis into the limiting E0L MTC value specified in the CORE OPERATING LIMITS REPORT (COLR). The 300 ppm surveillance MTC value specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration, and is obtained by making these corrections to the limiting MTC value.
i The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC i
remains within its limits since this coefficient changes slowly due principally to the reduction in RCS baron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 561*F. This limitation is required to ensure:
(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.
m 3/4.1.2 DELETED SOUTH TEXAS -UNITS 1 & 2 B 3/4 1-2 Unit 1 - Amendment No. 27,35,51,54,51,79 Unit 2 - Amendment No. 17,25,40,43,50,68 j
REACTIVITY CONTROL SYSTEMS BASES I
3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that:
(1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and-thereby ensure compliance with the control rod alignment and insertion limits.
Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 259 steps withdrawn for the Control Banks and 18, 234, and 259 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.
Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 234 steps, only points in the indicated ranges are picked for verification of 4
agreement with demanded position.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation.
In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or y
equal to 561*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.
These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
i SOUTH TEXAS - UNITS 1 & 2 8 3/4 1-4 Unit 1 - Amendment No. 62 Unit 2 - Amendment No. 51
Page B 3/41-3 has been deleted.
i SOUTH-TEXAS - UNITS 1 & 2 8 3/4 1-3 Unit 1 - Amendment No. 51,54,52,79 Unit 2 - Amendment No. 40,03,51,68
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal operations and anticipated transients.
In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, and in MODE 5 witn reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least twa loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
The boron dilution analysis assumed a common RCS volume, and maximum dilution flow rate for MODES 3 and 4, and a different volume and flow rate for MODE 5.
The MODE 5 conditions assumed limiting mixing in the RCS and cooling with the RHR system only.
In MODES 3 and 4, it was assumed that at least one reactor coolant pump was operating.
If at least one reactor coolant pump is not operating in MODE 3 or 4, then the maximum possibledilution flow rate must be limited to the value assumed for MODE 5.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.
3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 504,950 lbs per hour of saturated steam at the valve setpoint of 2500 psia.
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e, innn SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-1 Unit l '2'kmen'dmentN'.Y9'^~~~
o Unit 2 - Amendment No. 68 i
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BASES a
SAFETY VALVES (Continued) l During Modes 1, 2, and 3, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 4
2735 psig. The combined relief capacity of all of these valves is greater j
than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e...'no credit is taken for a direct Reactor trip on the turbine trip j
resulting from loss-of-load) and also assuming no operation of the power-3 1
operated relief valves or steam dump valves.
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Demonstration of the safety valves' lift ' settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
l 3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the i
parameter is restored to within its limit following expected transient i
operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement i
that a minimum number of pressurizer heaters be OPERABLE enhances the r
capability of the plant to control Reactor Coolant System pressure and establish natural circulation.
J 3/4.4.4.
RELIEF VALVES i
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The power-operated relief valves (PORVs) and steam bubble function to a
j relief RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.
Each PORY has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
i The OPERABILITY of the PORVs and block valves is determined on the basis j
of their being capable of performing the following functions:
j A.
Manual control of PORVs is used to control reactor coolant system pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown. Manual control of PORVs is a j
safety related function.
i B.
Maintaining the integrity of the reactor coolant pressure boundary. This i
is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure j
boundary leakage.
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SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2 Unit 1 - Amendment No. M,
Unit 2 - Amendment No. 44, m
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