ML20217M809
| ML20217M809 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 08/19/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217M793 | List: |
| References | |
| NUDOCS 9708250309 | |
| Download: ML20217M809 (13) | |
Text
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UNITE 3 STATES NUCLEAR REGULATORY COMMISSION WAaHINGToN, D.C. 30MH001 HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-499 SOUTH TEXAS PROJECT. UNIT 2 AMENDMENT TO FAtlllTY OPERATING LICENSE Amendment No. 76 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 Board of San Antonio (CPS), Central Power and Light Company (CPL),
and City of Austin, Texas (COA) (the licensees), dated April 22, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this license amendment will not be inimical to the comon 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.
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. 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)llows:of Facility Operating License No. NPF-80 is hereby amended to read as fo 2.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 76, and the Environmental Protection Plan conta<ned in A)pendix B are hereby incorporated in the license, The licensee s1all opera,te the facility in accordance with the Technical Specifir.ations and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Y
1^2bLf lW Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects !!!/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 19, 1997 4
i I
ATTACHMENT TO LICENSE AMENDMENT NOS. 89 AND 76 FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 DOCKET NOS. 50-498 AND 50-499 Replace the following anges of the Appendix A Technical Sptcifications with the attached pages. Tie 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 5-6 5-6 6-22 6-22 6-23 6-23
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DEllGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies.
Each fuel assembly-l shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material.
Limited substitutions of zirconium alloy, ZlRLO or stainless steel filler rods l for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assens11es shall be limited to those fuel designs that have been analyzed with appilcable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
LONTROL R0D ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.g inches of absorber material. The absorber material within each assembly shall be silver-indium-cadmium or hafnium. Mixtures of hafnium and silver-indium-cadmium are not permitted within a bank. All control rods shall be clad with I
stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERAiURE 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 50'F.
XWE 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,814 1 100 cubic feet at a nominal T, of 561'F.
55 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.
5.6 FUEL STORI E 5.6.1 CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
Unit 1 - Amendment No. 2,10,15.tt,.
SOUTH TEXAS - UNITS 1 & 2 5-6 6h45,89 Unit 2 - Amendment No. 2,5,32,50,54,76
ADMINISTRATIVE CONTROLS MONTHLY _0PERATING_ REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience, including documentation 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,ional Office of the NRC. no Washington, D.C.
20555, with a copy to the Regional Administrator of the Reg later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 4.9.1.6.a Core operating limits shasl 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 E0L limits, and 300 ppa 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 band for Specification 3/4.2.1, 5.
Heat Fgx Hot Channel Factor, K(Z), Power Factor Multiplier, and(F
) for Specification 3/4.2.2, and 6.
Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3.
T 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 - Moderator 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 E0L MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNITS 1 AND 2.'
SOUTH TEXAS - UNITS 1 & 2 6-21 Unit 1 - Amendment No. 2,27,35,t7,89 Unit 2 - Amendment No. 1,17,2C,3C,76
ADMINISTRATIVE CONTROLS 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. Kntel (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 nxial 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), pov. 2 July 1981.
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)
Sa. WCAP-10266-P-A, Rev. 2, WCAP-ll524-NP-A, Rev. 2, 'The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", Kabadi, J.N., et al., March 1987; including Addendum 1-A,"PowershapesensitivityStudies," December 1987 and Addendum 2-A, ' BASH eiethodology Improvements and Reliability Enhancements' May 1988.
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
5b. WCAP-It610-P-A, ' VANTAGE + Fuel Assembly Reference Core Report,"
April,1995 (W Proprietary)O clad fuel for rod heatup for Loss of Coolant Accident (LOCA)
Evaliation models with ZlRL calculation.
(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 a >plicable limits (e.g., fuel thermal-mechanical limits, core tiermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
No. f ',ff',ff,ff',k SOUTH TEXAS - UNITS 1 & 2 n
2 - Amen men
ADNINISTRATIVE CONTROLS CORE DPERATING LIMITS REPORT (Continued) 6.9.1.6.d The CORE OPERATING LINITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance,ies to the for each reload cycle to the NRC Document Control Desk, with cop 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.
6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations the following records shall be retained for atleasttheminimumperiodindicated.
6.10.2 The following records shall be retained for at least 5 years l
a.
Records and logs of unit operation covering time interval at each power level; b.
Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety; c.
All REPORTABLE EVENTS; d.
Records of surveillance activities inspections, and calibrations required by these Technical Specifications; e.
Records of changes made,to'the procedures required by Specification
- 6.8.1; f.
Records of sealed source and fission detector leak tests and results; and g.
Records of annual physical inventory of all sealed source material of record.
6.10.3 The following records shall be retained for the de ion of the unit Operating License:
a.
Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; b.
Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; c.
Records of doses received by all individuals for whom monitoring was required; d.
Records of gaseous and liquid radioactive material released to the environs; SOUTH TEXAS - UNITS 1 & 2 6-23 Unit 1 - Amendment No. 4hW,89 Unit 2 - Amendment No. M,76
ADMINISTRATIVE CONTROLS 8.10 RECORD RETENT10N fContinued)
Records of transient or okerational cy les for those.u' nit e.
components identified in he UFSAR; f.
Records of reactor tests and experiments; g.
Records o'f training and qualification for current members of the unit staff; h.
Records of inservice inspections performed pursuant to these Technical Specifications; 1.
Records of quality assurance activities required by the Opera-tional Quality Assurance Plan; j.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.5g; k.
Records of meetings of the PORC and the NSRB;.
1.
Records of the service lives of all hydraulic and mechanical i
snubbers required by Specification 3.7.g including the date at which the service life commences and associated installation and maintenance records; m.
Records of secondary water sampling and water quality; and ~
n.
Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis a'; a later date. This should include procedures effective at speelfied times and QA records showing that these procedures were followed.
c.
Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL AND THE PROCESS CONTROL PROGRAM.
p.
Records of radioactive sMpments.
l 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Proced6res for personnel radiation protection-shall be prepared consistent with the requirements of 10 CFR Part to and shall be approved, maintained, and adhered to for all operatians involving personnel radiation exposure.
6.12 HIGH RADIATION AREA 6.12.1 Pursuanttogaragraph 20.1601(c) of 10 CFR Part 20, in lieu of the l
' control device
- or alarm signal" required by paragraph 20.1601(a), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of.
SOUTH TEXAS - UNITS 1 & 2 6-24 Unit 1 - Amendment No. 46,47,57 Unit Amendment No. 44,46,46 DEC 3 01995 m
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 89 AND 76 TO fA(1LITY OPERATING LICENSE NOS. NPF-76 AND NPF-BQ HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NOS. 50-498 AND 50-499 SOUTH TEXAS PROJECT. UNITS 1 AND 2 l
1.0 INTRODUCTION
By application dated April 22, 1997, Houston Lighting & Power Company, et.al.,
(the licensee) requested changes to the Technical Specifications (TSs) he (Appendix A to Facility Operating License Nos. NPF-76 and NPF-80) for t South Texas Project, Units 1 and 2 (STP).
The proposed changes would revise TS 5.3.1, Fuel Assemblies, and 6.9.1.6, Core Operating Limits Report, to allow use of an alternate zirconium-based fuel cladding, ZIRLO, and limited substitution of fuel rods by ZlRLO filler rods.
2.0 BACKGROUND
The use of ZlRLO clad material in Westinghouse fuel was described in Topical Report WCAP-12610. " VANTAGE + Fuel Assembly Reference Core Report," and was approved by the Nuclear Regulatory Commission (NRC) staff for irradiation up to 60,000 MWD /MTU rod average burnup.
Extensive testing of this fuel has been conducted by Westinghouse through lead test assembly programs and it has been selected as reload fuel by other utilities.
3.0 EVALUATION TS 5.3.1 requires, in part, that each fuel assembly shall consist of a matrix of zircaloy clad fuel rods.
Zircaloy or stainless steel filler rods may be substituted in place of fuel rods in accordance with approved applications of the fuel rod configurations and by cycle-specific reload analysis.
The pro)osed amendment would modify TS 5.3.1 to allow fuel rods to be constructed witi ZlRLO and allow fuel assembly reconstitution with ZlRLO filler rods. The use of other zirconium alloys would require an exemption from D CFR 50.46 in that only zircaloy and ZlRLO are identified in that regulation, l
w
. TS 6.1.9.6 lists the analytical methods, which have been reviewed and approved by the NRC, that are used to determine the core operating limits.
The staff approved the ZlRLO fuel design in a safety evaluation (SE) dated July 1,1991, of Westinghouse Topical Report WCAP-12610. The NRC staff also approved loss of coolant accident (LOCA) methodologies in another SE, dated October 9,1991, of Westinghouse Topical Reports WCAP-12610 Appendix F, "LOCA NOTRUMP Evaluation Model:
ZlRLO Modifications " and Appendix G "LOCA Plant-Specific Accident Evaluation." The July 1, 1991, SE concluded that:
a.
The mechanical design bases and limits for ZIRLO clad fuel assembly design are the same as those for the previously licensed Zircaloy-4 clad fuel assembly design, except those specified for clad corrosion.
b.
The neutronic evaluations have shown that ZlRLO clad fuel nuclear design bases are satisfied and that key safety parameter limits are applicable. The nuclear design models and methods accurately describe the behavior of ZlRLO clad fuel, c.
The thermal and hydraulic design basis for ZIRLO clad fuel is
(
unchanged.
d.
The methods and computer codes used in the analysis of the non-LOCA licensing-basis events are valid for ZlRLO clad fuel, and all licensing-basis criteria will be met, e.
The large-break LOCA evaluation model was modified (without l
effecting model parameters as specified in Appendix K) to reflect the behavior of the ZIRLO clad material during a LOCA, Consequently, the revised evaluation model satisfies 10 CFR 50.46 f
and Apptndix K of 10 CFR Part 50.
In the October 9, 1991 SE for WCAP-12610, Appendices F and G, the NRC concluded that the LOCA analyses and methods used demonstrated conformance with the criteria given in 10 CFR 50.46 and 10 CFR Part 50, Appendix K.
The SE stated that its conclusions were based upon the c. lose similarity between the material properties of the ZIRLO alloy of zirconium to those of other zirconium materials that have been previously licensed for use as cladding material. Based on this simih rity, the NRC staff found that it is appropriately conservative to apply the criteria of 10 CFR 50.46 and 10 CFR Part 50, Appendix K, when reviewing VANTAGE + (ZlRLO) fuel applications, including WCAP-12610, Appendices F and G.
Use of ZIRLO is intended to remediate the phenomenon of incomplete rod insertion, which has been experienced at STP.
In-vessel compressive loading and irradiation growth of the fuel assembly guide tubes have been determined
4.
to be the cause of incomplete partial insertion.
The material of the guide tubes is being changed to ZlRLO for better dimensional stability and corrosion resistance, as well as compatibility with the fuel assembly skeleton.
Changing to ZIRLO cladding will also inhibit in-core fuel rod corrosion, which studies have shown to be of concern relative to high burnup fuel and longer cycles.
The change from Zircaloy-4 to ZIRLO is consistent with 10 CFR 50.44, 10 CFR 50.46 and NUREG-1431, " Standard Technical Specifications for Westinghouse Plants," which specifically includes ZIRLO as an acceptable cladding material. The licensee proposes to add the Westinghouse report WCAP-12610-P-A, " VANTAGE + Fuel Assembly Reference Core Report," to TS 6.1.9.6 to reflect the methodology used for the rod heatus calculation in the LOCA evaluation models with 71RLO clad fuel. This met 1odology adequately models core performance with ZIRLO and is acceptable. Thus, in light of the l
similarities in hydraulic, mechanical, and thermal characteristics of the fuel, the NRC staff concludes that the use of ZIRLO clad fuel at STP is acceptable.
4.0 TS CHANGES REACTOR CORE. FUEL ASSEMBLIES. TS 5.3.1 The licensee has proposed to add "ZIRLO" in the fuel rod design in TS 5.3.1 in j
the following sentences:
Each fuel assembly shall consist of a mt.trix of Zircaloy or ZIRLO fuel rods....
Limited substitutions of zirconium alloy, or ZlRLO....
I On the basis of its evaluation, the NRC staff concludes that the proposed change to add ZIRLO to TS 5.3.1 is acceptable.
CORE OPERATING LIMITS REPORT (COLR). TS 6.9.1.6.b The licensee also proposed to add a reference to NRC-approved Topical Report WCAP-12610-P-A, April 1995 (Westinghouse Proprietary), for a LOCA evaluation model with ZlRLO clad fuel for the rod heatu) calculation (Methodology for Specification 3.2.2 - Heat Flux Hot Channel factor).
Use of this approved methodology will ensure that values for cycle-specific parameters are determined such 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. Therefore, the staff considers that the modification to Section 6.1.9.6.b is acceptable.
. 5.0
SUMMARY
The NRC staff has reviewed the licensee's submittal regarding the use of ZIRLO clad fuel and the associated TS changes. On the basis of its evaluation, the staff concludes that use of ZIRLO is consistent with the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K and the proposed changes to the IS and the COLR are acceptable.
6.0 STATE CONSULTATION
In accordance with the Comission's regulations, the Texas State official was notified of the proposed issuance of the amendments. The State official had no coments.
l
7.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public coment on such finding (62 FR 27795). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
8.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the pt.blic will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributor:
A. Attard Date: August 19. 1997