ML20091E123
| ML20091E123 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/02/1992 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091E125 | List: |
| References | |
| NUDOCS 9204130347 | |
| Download: ML20091E123 (8) | |
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UNITED STATES i
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HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DQCKET NO. 50-498 SOUTH TEXAS PROJECT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 35 License No. NPI-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 August 30, 1991, as supplemented by letter dated January 24, 1992, 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 facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There a 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 Cl 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 l
construction, operation and maintenance of the facility.
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Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating Lir.ense No. NPF-76 is hereby amended to read as follows:
2.
Isthnical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
35, 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 Snesifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and to be implemented within 10 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Rc.smx.
Suzannh C. Black, Director Project Directorate IV-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 2, 1992 l-l ll I
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HOUSTON LIGHTING & _P_0ER COMPANY CITY PUBLIC_SERYlCE BOARD 1F SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-499 SOUTH TEXAS PROJECT. UNIT 2 AMENDMENT TO FA. W ITY OPERATING LICENS1 Amendment No. 26 License No. NPF-80 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 August 30, 1991, as supplemented by letter dated January 24, 1992, 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 facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and t
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.
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- Houston Lighting & Power Company is autiiorized 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 Specifi-cations 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:
Tec!1 cal Scecification.1 i
2.
The Technical Specifications contained in Appendix A, as revised through Amendment No.
26, 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 and to be implemented within 10 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h
- b. h Cb Suzanne C. Black, Director Project Directorate IV-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 2, 1992
ATTACHMENT TO LICENSE AMENDMENT NOS.35 AND 26 t
FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 DOCKET NOS. 50-498 AND 50-493 Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change, The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT.
6-20 6-20 6-20a 6-20a B 3/4 1-2 B 3/4 1-2 i
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4 ADW NISTDATIVE CONTROLS SEM] ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) in Appendix B shall be supplemented with three additional categories:
solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, class of Type A, Type B, large Quantity) and SOLIDIFICATION agent or absorbent (e cement, urea formaldehyde).
The Semiannual Radioactive Effluent Release Report to be submitted within meteorological data collected over the previous year.60 days a be either in the form of an hour-by-hour listing on magnetic tape of windThis ann or in the form of joint frequency distributions of wind sp and atmosDheric stability."
This same report shall include an assessment of the rce from the unit or station during the previous calendar year.sion co This same report shall also include an assessment of the radiation doses from radioactiv inside the Slif BOUNDARY (Figures 5.1-3 and 5.1-4) du assumptions used in making these assessments, i.e., specific activity, exposure All time, and location shall be included in these reports.
conditions concurre,nt with the time of release of radioactive materials inThe mete be used for determining the gaseous pathway doses. gaseous effluen doses shall be performed in accordance with the methodology and parameters the OFFSITE DOSE CALCULATION MANUAL (ODCM).
60 days af ter January 1 of each year shall also include an a l
releases and other nearby uranium fuel cycle sources, inclu primary ef fluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190 " Environmental Radiation P Standards for Nuclear Power Operation.", Acceptable methods for calculating i
the dose contribution from liquid and gaseous ef f'luents are given in Regulatory Guide 1.109, Rev.1, October 1977.
The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents mada during the reporting pericd.
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to Specifications 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Ractwaste Treatment Systems pursuant to Specification 6.15.
It shall also include a listing of new loca-Land Use Census pursuant to Specification 3.12.2.tions for dose cal
- In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required logical data on site in a file that shall be provided tu the NRC upon request.
meteoro-SOUTH TEXAS - UNITS 1 & 2 6-19
AQMINISTRATIVE__ CONTROLS SEMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT (Continued)
The Semiannual Radioactive Effluent Release Reports shall also include the following:
an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the
-events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.
MONTHLY OPERATING REPORTS:
6.9.1.5 Routine reports of operating statistics and shutdown experience, including do:ume'1tation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management,- U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of. each month following thd calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits snall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle, or any part of a reload cycle for the following:
1.
Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2,
Shutdown Bank _ Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion Limits for Specification 3/4.1.3.6, 4,
Axial Flux Difference limits and target pand for Specification 3/4.2.1, 5.
Heat Flux Hot Channel Factor, K(Z), Power Factor Multiplier, and (FRTP) for Specification 3/4.2.2, and x
6.
Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3.
The CORE OPERATING LIMITS REPORT shall be maintained available in
-the Control Room.
6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1.
WCAP 9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July, 1985 (W Proprietary).
(Methodology for Specification 3.1.1.3 - hoderator Temperature Coefficient, 3.1.3.5 - Shutdown Rod Insertion Limit, 3.1.3.6 -
Control Bank-Insertion' Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
1.A. WCAP 12942-P-A, " SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SOUTH TEXAS FROJECT ELECTRIC GENERATING STATION UNITS 1 AND 2."
SOUTH TEXAS - UNITS 1 & 2 6-20 Unit 1 - Amendment No.
9, 27, 3E Unit 2 - Amendement No. I, 17, 26
ADMR111RATIVE IDRIROLS_
4 CORE OPERATING LIMITS REPORT (Continued)
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient) 2.
WCAP 8385, " POWER DISTRIBUTION AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT", September, 1974 (W Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)
3.
Westinghouse letter NS-TMA-2198, T.M. Anderson (Westinghouse) to K. Kniel (Chief of Core Performance Branch, NRC) January 31, 1980 -
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).
Approved by NRC Supplement No. 4 to NUREG-0422 January,1981 Docket Nos. 50-369 and 50-370. )
4.
NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July, 1981.
Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC),
Rev. 2, July 1981.
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)
5.
WCAP 9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION", February 1982 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
6.
WCAP 95El-P-A, ADD. 3, Rev.1, "BART A-1:
A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS - SPECIAL REPORT:
THIMBLE MODE!.ING W ECCS EVALUATION MODEL", July,1986 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector, SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
SOUTH TEXAS - UNITS 1 & 2 6-20a Unit 1 - Amendment No.
27, 35 Unit 2 - Amendement No. 17, 26
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:
(1) the reactor can be made subtritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.
In MODES I and 2, the most restrictive condition occurs at EOL, with T,yg at no~ load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.75% ak/k is required to control the re;;ctivity transient.
The 1.75% 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 l
Table 4.3-3).
Accordingly, the SHUTDOWN MARGIN requirement for MODES 1 and 2 is based upon this limiting condition and is consistent with FSAR safety anal-ysis assumptions.
In MODES 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concentration is the greatest.
In these modes, the required SHUTDOWN MARGIN is composed of a constant requirement and a variable requirement, which is a function of the RCS boron concentration.
The constant SHUTDOWN MARGIN requirement of 1.75% Ak/k is based on an uncontrolled RCS cool-down from a steamline break acciden..
The variable SHUTDOWN MARGIN requirement is based on the results of a boron dilution accident analysis, where the SHUT-DOWN HARGIN is varied as a function of RCS boron concentration, to guarantee a minimum of 15 minutes for operator action after a boron dilution alarm, prior to a loss of all SHUTDOWN MARGIN.
The baron 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 limited 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 SHUTDOWN MARGIN requirements for MODE 5 shall apply.
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (KTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.
The MTC values of this specification are applicable to a specific set of 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.10 Unit 2 - Amendment No. 2
. REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)
The most negative MTC value, equ'ivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC 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 condi'.. ions, and (2) subtracting from this a
value the largest differences in MTC observed at EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skesing, 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 EOL 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.
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 remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron 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 P.eactor Coolant System average temperature less than 551*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 rar.ge, (3) the pressurizer is capable of being in an GPERABLE status with a steam bubble, and (4) the reactor vessel-is above its minimum RT temperature.
NDT 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.
The components required to perform this function include:
(1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 350 F, a minimum of two boron injec-tion flow paths are required to ensure single functional capability in the event an assumed failure renders one of the tiow paths inoperable.
The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions.of 1.75% Ak/k after xenon decay and cooldown to 200 F.
The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 27,000 gallons of 7000 ppm borated water from the boric acid storage system or 458,000 gallons of 2500 ppm borated water from the refueling water storage tank (RWST).
The RWST volume is an ECCS requirement and is more than adequate for the required boration capability.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 1-2 Unit 1 - Amendment No.
27, 35 Unit 2 - Amendment No.
17, 26 I
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