View source for NOC-AE-02001272, Application for Amendment to Several Sections of TS & Unit 2 Operating License to Delete Information Specific to Model E Steam Generators

{{Adams
| number = ML021540264
| issue date = 05/23/2002
| title = Application for Amendment to Several Sections of TS & Unit 2 Operating License to Delete Information Specific to Model E Steam Generators
| author name = Sheppard J
| author affiliation = South Texas Project Nuclear Operating Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000498, 05000499
| license number = NPF-076, NPF-080
| contact person = 
| case reference number = NOC-AE-02001272
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Amendment
| page count = 42
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Nuclear Operating Company South Texas ProqectElectnrc Generati'n Station P.. Box 289 Wadsworth, Texas 77483
* May 23, 2002 NOC-AE-02001272 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 License Amendment Request - Proposed Amendment to Operating License and Technical Specifications Regarding Steam Generators

==Reference:==
Letter, S.M. Head to NRC Document Control Desk, "Response to NRC Regulatory Issue Summary 2001-2 1," dated January 17, 2002 (NOC-AE-02001243)
Pursuant to 10CFR50.90, STP Nuclear Operating Company (STPNOC) hereby requests an amendment to several sections of the Technical Specifications (TS) and to the Unit 2 Operating License to delete information specific to Model E steam generators (SGs). The Model E SGs in Unit I have been replaced with Delta 94 SGs and replacement in Unit 2 will be complete by the end of 2002. Therefore, differentiation between SG models will no longer be required. Specific references to Delta 94 SGs are also removed.
STPNOC has determined that the proposed amendment involves no significant hazards consideration (Attachment 1). Proposed revised TS pages are provided in Attachments 2 (mark ups) and 3 (retyped). The revised TS Bases are provided in Attachment 4 for information only.
A proposed revised page of the Unit 2 Operating License is provided in Attachments 5 (mark-up) and 6 (retyped).
The Plant Operations Review Committee and the Nuclear Safety Review Board have reviewed the proposed change. STPNOC has notified the State of Texas in accordance with 10CFR50.91(b).
This amendment request was originally planned for submittal in fiscal year 2003, rather than as part of the fourteen plant-specific submittals identified for fiscal year 2002 (see referenced letter). However, due to the administrative nature of the proposed changes and the availability of resources to prepare the amendment, STPNOC is submitting this license amendment request at this time. Approval of the proposed change is not a prerequisite for steam generator replacement scheduled for late 2002. STPNOC requests approval of the proposed amendment by January 15, 2003. Once approved, the amendment shall be implemented within 30 days, but not before steam generator replacement is complete.
STI: 31409438

NOC-AE-02001272 Page 2 of 2 If there are any questions regarding this proposed amendment, please contact Mr. Scott Head at (361) 972-7136 or me at (361) 972-8757.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on            May 23, 2002 J. J. Sheppard Vice President, Engineering & Technical Services Attachments:
: 1. Licensee's Evaluation
: 2. Proposed Technical Specification Changes (Mark-up)
: 3. Proposed Technical Specification Pages (Retyped)
: 4. Bases (For Information Only)
: 5. Proposed Unit 2 Operating License Changes(Mark-up)
: 6. Proposed Unit 2 Operating License Page (Retyped) cc:
(paper copy)                                            (electronic copy)
Ellis W. Merschoff                                      A. H. Gutterman, Esquire Regional Administrator, Region IV                       Morgan, Lewis & Bockius LLP U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400                         M. T. Hardt/W. C. Gunst Arlington, Texas 76011-8064                             City Public Service U. S. Nuclear Regulatory Commission                     Mohan C. Thadani Attention: Document Control Desk                        U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike                                    R. L. Balcom Rockville, MD 20852                                     Reliant Energy, Inc.
Richard A. Ratliff                                      A. Ramirez Bureau of Radiation Control                             City of Austin Texas Department of Health 1100 West 49th Street                                  C. A. Johnson Austin, TX 78756-3189                                   AEP - Central Power and Light Company Cornelius F. O'Keefe                                    Jon C. Wood U. S. Nuclear Regulatory Commission                     Matthews & Branscomb P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704

NOC-AE-02001272 Attachment 1 Licensee's Evaluation

NOC-AE-02001272 Attachment 1 Page 1 of 4 LICENSEE'S EVALUATION

==1.0     DESCRIPTION==

This letter is a request to amend Operating Licenses NPF-76 and NPF-80 for South Texas Project (STP) Units 1 and 2. The proposed change would revise several sections of the Technical Specifications (TS) and the Unit 2 Operating License to delete information specific to Model E steam generators (SGs). The Model E SGs in Unit 1 have been replaced with Delta 94 SGs and replacement in Unit 2 will be complete by the end of 2002. Therefore, differentiation between SG models will no longer be required. Specific references to Delta 94 SGs are also removed.
Approval of the proposed change is not a prerequisite for steam generator replacement scheduled for late 2002. STPNOC requests approval of the proposed amendment by January 15, 2003.
Once approved, the amendment shall be implemented within 30 days.

==2.0     PROPOSED CHANGE==
S Specifically, the proposed changes would:
    "* Delete the rated thermal power associated with Model E SGs in definition 1.27.
    "* Delete the SG low-low water level trip setpoint and allowable value associated with Model E SGs in Table 2.2-1.
    "* Delete the thermal design reactor coolant system flow value associated with Model E SGs in TS 3.2.5.
    "* Delete the SG low-low water level trip setpoint and allowable value for auxiliary feedwater initiation associated with Model E SGs in Table 3.3-4.
    "* Delete references to voltage-based repair criteria and to repairing Model E SG tubes in TS 4.4.5.2.
    "* Delete descriptions of and references to tube sleeving, tube repairing, repair limit, and tube support plate plugging limit in TS 4.4.5.4.
    "* Delete references to tube repairing, and the notification and reporting requirements for Model E SGs specified in TS 4.4.5.5.
    "* Delete references to tube repairing in Table 4.4-2.
    "* Delete Table 4.4-3 entirely.
    "* Delete inoperable safety valve data and neutron flux setpoint data associated with Model E SGs in Table 3.7-1.
    "* Delete all references to Delta 94 SGs in the TS.
    "* Delete differentiation in the definition of 100% power level based on SG type in the Unit 2 Operating License

NOC-AE-02001272 Attachment 1 Page 2 of 4

==3.0     BACKGROUND==

The Model E SGs in Unit 1 were replaced with Delta 94 SGs in mid-1998. Operation with the Delta 94 SGs has increased reactor coolant system flow due to lower flow resistance. Thermal design flow increased from 95,400 gpm per loop to 98,000 gpm per loop and mechanical design flow increased from 106,600 gpm per loop to 110,000 gpm per loop. To reduce the resultant increased lift forces on the fuel assemblies, the thimble plugging devices were eliminated, resulting in a 2% increase in the design core bypass flow fraction. The design core bypass flow was also increased by 2% to account for upper head TCOLD conversion. The SG normal water level was increased and the SG low-low water level trip setpoint was reduced to provide increased operating margin. Other differences in the Delta 94 SGs include:
* 39% increase in heat transfer area
* 24% increase in primary side volume
* 4.5% increase in normal full load water level
* 6% reduction secondary side volume o   60% reduction in moisture carryover Beginning in 1999, as a result of the differences between the SG models, the TS were revised to reflect both models while Unit 1 was operating with Delta 94 SGs and Unit 2 was operating with Model E SGs (References 1 through 4). The Model E SGs in Unit 2 will be replaced with Delta 94 SGs by the end of 2002.

==4.0     TECHNICAL ANALYSIS==

The proposed amendment is administrative in nature. Operational limitations and surveillances for both Model E and Delta 94 SGs have been approved by the NRC (References 1 through 5).
With the replacement of the Unit 2 Model E SGs in late 2002, the TS and Operating License should reflect operational limitations for the Delta 94 SGs only. Therefore, STPNOC proposes to remove all information associated with the Model E SGs and all specific references to Delta 94 SGs.
The proposed TS amendment does not impact any accident analysis described in the Updated Final Safety Analysis Report.
5.0     REGULATORY SAFETY ANALYSIS 5.1     No Significant Hazards Consideration STPNOC has determined whether a significant hazards consideration is involved with the proposed amendment by focusing on the three criteria set forth in 10CFR50.92 as discussed below:

NOC-AE-02001272 Attachment I Page 3 of 4
: 1.      Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The Operating Licenses currently reflect plant operation with both Delta 94 and Model E SGs, but all Model E SGs will be replaced with Delta 94 SGs by the end of 2002. The proposed administrative change deletes information associated with the Model E SGs and deletes references to Delta 94 SGs. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2.      Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The Operating Licenses currently reflect plant operation with both Delta 94 and Model E SGs, but all Model E SGs will be replaced with Delta 94 SGs by the end of 2002. The proposed administrative change deletes information associated with the Model E SGs and deletes references to Delta 94 SGs. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3.      Does the proposed change involve a significant reduction in a margin of safety?
Response: No The Operating Licenses currently reflect plant operation with both Delta 94 and Model E SGs, but all Model E SGs will be replaced with Delta 94 SGs by the end of 2002. The proposed administrative change deletes information associated with the Model E SGs and deletes references to Delta 94 SGs. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, STPNOC concludes that the proposed amendment involves no significant hazards consideration under the criteria set forth in 10CFR50.92 and, accordingly, a finding of "no significant hazards consideration" is justified.
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

NOC-AE-02001272 Attachment 1 Page 4 of 4

==6.0      ENVIRONMENTAL CONSIDERATION==

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==7.0      REFERENCES==
: 1. NRC letter, T. W. Alexion to W. T. Cottle, "South Texas Project, Units I and 2 Issuance of Amendments re: Replacement Steam Generator Reactor Coolant Flow Differences (TAC Nos. MA1911 and MA1912)," dated November 8, 1999
: 2. NRC letter, T. W. Alexion to W. T. Cottle, "South Texas Project, Units 1 and 2 Issuance of Amendments re: Replacement Steam Generator Water Level Trip Setpoint Differences (TAC Nos. MA2500 and MA2501)," dated December 29, 1999
: 3. NRC letter, T. W. Alexion to W. T. Cottle, "South Texas Project, Units 1 and 2 Amendment Nos. 107 and 94 to Facility Operating License Nos. NPF-76 and NPF-80 (TAC Nos. MA3761 and MA3762)," dated April 19, 1999
: 4. NRC letter, M. C. Thadani to W. T. Cottle, "South Texas Project (STP), Unit 2 - Issuance of Amendment Revising the Technical Specifications to Implement 3-Volt Alternate Repair Criteria for Steam Generator Tube Repair (TAC No. MA8271)," dated March 8, 2001
: 5. NRC letter, M. C. Thadani to W. T. Cottle, "South Texas Project, Units 1 and 2 Issuance of Amendments Approving Uprated Core Thermal Power and Revising the Associated Technical Specifications (TAC Nos. MB2899 and MB2903)," dated April 12, 2002

NOC-AE-02001272 Attachment 2 Proposed Technical Specification Changes (Mark-ups)

1.0 DEFINITIONS PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.25 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.27     RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3,853 Mwt. (Model A94 steam generAtorc in~tafled) or 3,800MvA (Model F steam REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components or methodology for verification have been previously reviewed and approved by the NRC.
REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SOUTH TEXAS - UNITS 1 & 2                        1-5                    Unit 1 - Amendment No. 439 Unit 2 - Amendment No. 427

CD, 0
C
  -I
  -I M                                                    TABLE 2.2-1 (Continued) m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS z
FUNCTIONAL UNIT                      TRIP SETPOINT                    ALLOWABLE VALUE CN 90
: 13. Steam Generator Water Level          &#x17d;:20.0% of narrow range          >18.0% of narrow range Low-Low                               instrument span                 instrument span fMedel Eq                            &#x17d;!330% of narrow range            _ 30.7% of narrow ratge nstument &pan                  instrument spaR
                                                >20.0-    of narrow range           18. 0, of_narrow range instrument sWa                  intrmentsa
: 14. Undervoltage                          &#x17d;! 10,014 volts                 &#x17d;: 9339 volts Reactor Coolant Pumps
: 15. Underfrequency
                                                > 57.2 Hz                       > 57.1 Hz Reactor Coolant Pumps
: 16. Turbine Trip
: a. Low Emergency Trip Fluid         > 1245.8 psig                   &#x17d; 1114.5 psig Pressure CC
: b. Turbine Stop Valve Closure       < Fully closed                   Fully closed 3 3
: 17. Safety Injection Input from ESFAS     N.A.                             N.A.
CDCD 3 3 zz PZ

POWER DISTRIBUTION LIMITS 3/4.2.5  DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5      The following DNB-related parameters shall be maintained within the limits following:
: a. Reactor Coolant System Tavg, -<the limit as specified in the Core Operating Limits Report
: b. Pressurizer Pressure, > the limit as specified in the Core Operating Limits Report
: c. Thermal Design Reactor Coolant System Flow fG           , &#x17d;392,000 gpm
: 1. Moitdol E Steam Genoratoem,             2370,000 gpmn 2., Modol A94 Steam Gonorators             > 392,000 gp~m APPLICABILITY:       MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.5.1   Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. The provisions of Specification 4.0.4 are not applicable for verification that RCS flow is within its limit.
4.2.5.2   The RCS flow rate indicators shall be subjected to a channel calibration at least once per 18 months.
NOTE SR  4.2.5.3 is required   at beginning-of-cycle with reactor power > 90% RTP.
4.2.5.3   The RCS total flow rate shall be determined by precision heat balance or elbow tap AP measurements at least once per 18 months. The provisions of Specification 4.0.4 are not applicable.
SOUTH TEXAS - UNITS 1 & 2                      3/4 2-11            Unit 1 - Amendment No. 14 5,1 Unit 2 - Amendment No. 103,  ...

(n 0
C                                          TABLE 3.3-4 (Continued) m-I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT                           TRIP SETPOINT                    ALLOWABLE VALUE
   --I 5. Turbine Trip and Feedwater Isolation (Continued)
   =1\
co
: d. Deleted
: e. Safety Injection                    See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
: f. Tavg - Low Coincident with          > 5740 F                         >571.7 0 F Reactor Trip (P-4)
(Feedwater Isolation Only)
C,, 6. Auxiliary Feedwater C,,
: a. Manual Initiation                   N.A.                             N.A.
: b. Automatic Actuation Logic           N.A.                             N.A.
: c. Actuation Relays                    N.A.                             N.A.
: d. Steam Generator Water Level-         .. 20.0% of narrow range        >  18.0% of narrow range Low-Low                             instrument span                  instrument span CC
                                                 &#x17d;&#xfd; 33.01%of narroW range            80.7,0 94nrrwrag inStF~ment span                               6~-ep~a 3  B                                               &#x17d;20.01% of narrow rango           18.0% of narrow rangp
:3 3                                             h;RetFUnent-epaA                      umeaR 3 3       e. Safety Injection                    See Item 1. above for all Safety Injection Trip Setpoints and
=00                                              Allowable Values zz 00O

REACTOR COOLANT SYSTEM 3/4.4.5     STEAM GENERATOR LIMITING CONDITION FOR OPERATION 3.4.5     Each steam generator shall be OPERABLE.
APPLICABILITY:           MODES 1,2, 3, and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing Tavg above 200 0 F.
SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. and Table 4.4 3. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Whon applying the ex            ef 4.4.5.2.a veptifis              r      5.2., provious defectS or imperfections in the area repaired by sleeving are net Gonsidorod an Area requiring rspection. The tubes selected for each inservice inspection shall include at least 3% of the total number of nonplugged          wepaied tubes in all 1 I=m g                a             E steam  generat      9oy) 2003 of th  total number of rop tubes in all steam generators; the tubes selected for   these inspections  shall  be selected on a random basis except:
: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
: b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
: 1)     All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
: 2)     Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2                 3/4 4-12              Unit 1 - Amendment No. 90 , 1 0 Unit 2 - Amendment No. 77, 94

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
: 3)   A tube Inspection (pursuant to Specification 4.4.5.4a.9) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
: c. The tubes selected as the second and third samples (if required by Table 4.4-2 G Table-44-3) during each inservice inspection may be subjected to a partial tube inspection provided:
: 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
: 2)  The inspections include those portions of the tubes where imperfections were previously found.
d --
: 2)   All ntrcioeWith large mixed rocidua!G that could potentially mask flaw responeos at or above the voltage repair limits shall be inpepoted by rotating pancake col (or equivalent).
SOUTH TEXAS - UNITS 1 & 2                  3/4 4-13       Unit 1 - Amendment No. 82,83,90,96,110 Unit 2 - Amendment No.Th83,94r144

RE-AC-TO-R COOL-A.N1T SYSTEM RIPAU
 .   .   . N.PAMr.P.
SURV2EIL'NCU-E REQUI-1R1-4EMENTS (ConlonueE4 The results of each sample inspection shall be classified into one of the following three categories.
Category                        Inspection Results C-1              Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2             One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3              More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
iSOUT-1H TEXAS -UNITS 1 &2                      14 4 1-3a            Unit 1 Amendmont No.3 Unit 2 -AmGndmont No. 1444

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
: a. As used in this specification:
: 1)     b        Tube means that portion of the tube or-eleev which forms the primary system to secondary system pressure boundary;
: 2)  Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
: 3)   Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube;
: 4)   Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
: 5)   % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
: 6)   Defect means an imperfection of such severity that it exceeds the plugging oF tepa' limit. A tube containing a defect is defective;
: 7)   Plugging Limit Gr ResaiLi           means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thic~kness; by plugging or (for Model I=steam generators only) ropairod by cleoving in the affectod area bocauso it may b0coMe unsor~iceablo prior to the noxt finSpoction.&#xfd; The plugging or repair limit imperfoction depths are SPocifiod in porcontago of the nomna wall thicknose as follows, aoigia tube wall                                  4"0
: b. Wostinghouse laser woldod sleeve wall                40%
For Model E steam; genorators, this dofinition does not apply to tube support pato intorsections for which the voltage based repair criteria aro befin~g applied. Ro4Fr to 4.4.5.4.a.1 1 for the repair limnit applicable to theco intersectiens.
: 8)     Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affct its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above;
: 9) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; SOUTH TEXAS - UNITS 1 & 2                   3/4 4-15              Unit 1 - Amendment No. 83,90, 96, 107 Unit 2 - Amendment No. 77, 83, 94

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
: 10)  Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
At the flow distribution baffle inoscinat the cold log support plato intersec-tions, and at the hot leg support plato intersections with suppoed plates L through R (as identified in Figure 5.1 of WA.I/\ 15163, Reviio 1), the plugging (repair) limit is; based on maintaining steamn generato tube SerViceability as described in a), b), G)and d) below:
a)     Steam generator tubes, whose degradation is,atkibuted to outside diamete~r sotress o~reroion cracking within the boGunds of the tube support plate with bobbin voltage les than orF equal to the lower voltage repair limit (Note 1), will be allowed to emi in sor'e.
b)     Steamn generator tubes, whose degradation is.afttributod to outside diameter strs corrosion crackinq within the boundsG of the tube isupport plate with a bobbin voltage gF"e% ~ateF than the lower voltage repair limit (Note 1), will be repaired E) ~Gplugede-RGe      sted in 4.4.5.4.a. 1 1.Gbelow.
SOUTH TEXAS - UNITS 1 & 2              3/4 4-16         Unit 1 - Amendment No. 82,83,90, 96, 10:7 Unit 2 - Amendment No. 77, 83, 94,14

RRE.AC-TO-R Q- 0L ANT- SYST-E.M STEAM GENERTR 21OUr~h I ANrm~RFQI IIRF=!NA!rR'TS (Gontindonwe 4)-
Note 1,;      The lower voltage repair limit is 1.0 volt for 3/4 inch diamneter tubiRg.
watc, 2.      The upper voltage repair limit (YuRJ i clu&#xfd;                                                td o ec isec according to the methodology in Generic Letter 95 05 as supplomentod.                                                I shall be the larger of the a':erage growth ratoc eXpe9Riened intheto' rc ccebut not les6 than 30% Per effoctiVo full powerj','ar-.
For Unit 2 Cycle 9 only, at the hot leg support plate intersections with suppot plates C, F, and J (as identified in Figure 5.1 of WCAR 15163, RGVisiGR 1), the plugging (repair) limit is based on maintaining steamn generator t, -h sor.'ceaoiuty as aescr.ioca in 9;) f), a.ruy below; e) Steam generato r tubes, Whose deg radation i; attri bute toaillrented outside diameter stre corsincacking Within the bounds of the tuibe support plate witha bobbin voltage less than, Or equal to 3.0 volts may renain in se-i f) ,Steam generator tubes;, whore degradation is attributed to axially oriented Outside diameter strescorin cacking within the hounds of the tube support plate with a bobbin voltage greater than 3.0 volts,6hall be plugged or repaired rogardipes of, vghethe-r or not a rotating pancake coil inspection detects degradation-.
g) ifone or mornniatosi the tube support plate intersections are confirmned by non destructlve oxamlnO ion to extend oegYGRG tnc cage OfTH                                                wIUU       P)F ~I, woo uplate, 3 volt alterna;.te repair criteria shall not be used in any steam. generator. Exceptions to this requirement may be al!Gwed for these indications that are determined by the NRC staff to be physically insignificant for the purposes of6afoFy and rs asses6sment. Approval for the use ofthe 3 volt altoRnate repair crieria may be granted by the staff iwriting on a onetime basis, following the staf review an consian                     t            ct              re             t dstererneval of plugs~ Mat were oreviousv minsallea a
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                                      *L *
* Z** T* * *T*Z*                         ***
* klil iiN&#xb8; Tube repairen a cGEreGtive or prE~ventive) measure. Atube qn                                              spection per 4.4.5.4 .a.9 i FeW            Ing PFGVIPL&#xfd;ly pluqg(13&#xfd; tul "S " ov SOUTH TEXAS - UNITS 1-":                                               3/44-46a                               Unit 1 Amendment No. 83,909,0 Unit 2Amendment No. 77,83,94, 114

REACTOR GOOLANT SYSTEMK QTrAILA    KIDATCDQ SREILLIANICE REQUI!REMENITS (Ontinued
: b. The steam generator shall be determined OPERABLE after plugging Gempleting th..corropondin g tOn...[plug or (for Model E          g..eratore
                                                                       ...        only) repair all
                                                                                 .eam.
tubes exceeding the plugging 6r epai limit and all tubes containing through-wall cracks] required by Table 4.4-2 and Table 1.4-3.
4.4.5.5 Reports
: a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged e repai*ed in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
: b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
: 1) Number and extent of tubes inspected,
: 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
: 3) Identification of tubes plugged
: c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
9 SOUITH TE=XAR -JUNTS     I X2                  W4 4-46b      Unit 1 -Amendmnent No. 83,90, 96, 107 Unit 27-Amendment No., 77, 83, 94,,114

REACTOR COOLANT SYSTEF-M STEAm GENERATORS RUH'.IRVIL'NI IIUIIL' ANoniuo                 I
                                                                                             #44G4
: 2) IfGiFumfe~e rial cFAracK i!K' indiaicauos aro aetoceed at to tube support plate
: 3) ifindications are identifiled that extend boy E)Rdthe confines of the tubee puppo4~t We-.
: 4) if indications are identified at the tube support plate ole':ations; that are attributable to primary water stros corsoGracking.
K-S) a) The methodology of Generic Letter 95-05 for intersections at the flo distrbution baffles, at the applicable cold leg ssupport plates, and at the hot leg support plates l through R; and b) A total main steam line break tube burst probability of 1 x 1OV4for:hot leg intersections at support plates G, F, and j.
: 6) ifcracking is observod in the tube rsupport plates.
emt SOU0 ITH TEXAS - UNITS 1 & 2                    3/44-446G              Unit 1 -,Amendment No; I 'nit 2) - .4mtnrlmpnt NA 114

Table 4.4-2 STEAM GENERATOR TUBE INSPECTION 1st SAMPLE INSPECTION                              2nd SAMPLE INSPECTION                3rd SAMPLE INSPECTION Sample Size          Result       Action Required              Result        Action Required         Result     Action Required C-1               None                    N.A.               N.A.               N.A.             N.A.
A minimum of S Tubes per SG.         C-2     Plug eF4epak defective            C-1                None               N.A.             N.A.
tubes and inspect additional 2S tubes in            C-2     Plug or-repai defective       C-1             None this SG.                                  tubes and inspect additional 4S tubes in        C-2    Plug eF Fepa this SG                              defective tubes C-3    Perform action for C 3 result of first sample C-3     Perform action for            N.A.             N.A.
C-3 result of first sample.____
C-3     Inspect all tubes in this    All other                                  N.A.             N.A.
SG, plug 'e *epai            SGs are C-1             None defective tubes and inspect 2S tubes in          Some SGs     Perform action for C-2        N.A.             N.A.
each other SG.               C-2 but no   result of second sample additional Notify NRC pursuant to       SG are C-3 10CFR50.72(b)(3)(ii)
Additional   Inspect all tubes in each     N.A.             N.A.
SG is C-3    SG and plug 9. ,epai, defective tubes.
Notify NRC pursuant to 1 OCFR50.72(b)(3)(ii)
S=3N%
where N is the number of steam generators in the unit, and n is the number of steam generators inspected n
during an inspection.
SOUTH TEXAS - UNITS 1 & 2                                3/44-18                               Unit 1 - Amendment No. 90 Unit 2 - Amendment No. 77,114

n D
n ko 1)

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION MODEL A   S       GNERATORS ONLY MAXIMUM NUMBER OF INOPERABLE               MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY                    NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR                (PERCENT OF RATED THERMAL POWER) 1                                              61 2                                              43 3                                              26 MOEi STEAM GENRTR 6ONLY MAXIMUM NUMBER OF INOPERABLE               MAXIMUM ALLOWABLE POWER RANGE SAFETY VAL'.'E ONAN                      NEUTRON FLUX HIGH SET-POINT OPERA.TING STEAM GENERATOR                 ID &#xfd;fNT OF RATED THERMIAL D(QA1ID&#xfd; 4
2                                              46 a
SOUTH TEXAS - UNITS 1 & 2          3/4 7-2           Unit 1 - Amendment No. 48 Unit 2 - Amendment No. 427

NOC-AE-02001272 Attachment 3 Proposed Technical Specification Pages (Retyped)

1.0     DEFINITIONS PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.25 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3,853 Mwt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components or methodology for verification have been previously reviewed and approved by the NRC.
REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SOUTH TEXAS - UNITS 1 & 2                        1-5                    Unit 1 - Amendment No.
Unit 2 - Amendment No.

TABLE 2.2-1 (Continued) 0 C                                      REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS I-4i m
x       FUNCTIONAL UNIT                      TRIP SETPOINT                       ALLOWABLE VALUE z   13. Steam Generator Water Level          &#x17d; 20.0% of narrow range             > 18.0% of narrow range Cn      Low-Low                               instrument span                    instrument span C'
90
: 14. Undervoltage                         >_10,014 volts                      >_9339 volts Reactor Coolant Pumps
: 15. Underfrequency                        &#x17d; 57.2 Hz                          >_57.1 Hz Reactor Coolant Pumps
: 16. Turbine Trip
: a. Low Emergency Trip Fluid        > 1245.8 psig                       _>1114.5 psig Pressure 01
: b. Turbine Stop Valve Closure       < Fully closed                     Fully closed
: 17. Safety Injection Input from ESFAS     N.A.                               N.A.
CC
>33 CDCD 3
*- - 3:
zz P.P

POWER DISTRIBUTION LIMITS 3/4.2.5   DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5      The following DNB-related parameters shall be maintained within the limits following:
: a. Reactor Coolant System Tavg, < the limit as specified in the Core Operating Limits Report
: b. Pressurizer Pressure, > the limit as specified in the Core Operating Limits Report
: c. Thermal Design Reactor Coolant System Flow, > 392,000 gpm APPLICABILITY:        MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.5.1    Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. The provisions of Specification 4.0.4 are not applicable for verification that RCS flow is within its limit.
4.2.5.2    The RCS flow rate indicators shall be subjected to a channel calibration at least once per 18 months.
NOTE SR 4.2.5.3 is required at beginning-of-cycle with reactor power > 90% RTP.
4.2.5.3    The RCS total flow rate shall be determined by precision heat balance or elbow tap AP measurements at least once per 18 months. The provisions of Specification 4.0.4 are not applicable.
SOUTH TEXAS - UNITS 1 & 2                      3/4 2-11                   Unit 1 - Amendment No.
Unit 2 - Amendment No.

cj 0
7                                           TABLE 3.3-4 (Continued) m        ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT                          TRIP SETPOINT                    ALLOWABLE VALUE C
z   5. Turbine Trip and Feedwater Isolation
     .-I (Continued) 90
: d. Deleted
: e. Safety Injection                   See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
: f. Tavg - Low Coincident with         > 5740 F                         > 571.7 0 F Reactor Trip (P-4)
(Feedwater Isolation Only)
: 6. Auxiliary Feedwater 4DO
: a. Manual Initiation                  N.A.                             N.A.
: b. Automatic Actuation Logic          N.A.                             N.A.
: c. Actuation Relays                   N.A.                             N.A.
: d. Steam Generator Water Level-        __20.0% of narrow range         >_18.0% of narrow range Low-Low                            instrument span                  instrument span cc          e. Safety Injection                   See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values
  ) 33 (D0 2    S) zzt Pp 0    0

REACTOR COOLANT SYSTEM 3/4.4.5      STEAM GENERATOR LIMITING CONDITION FOR OPERATION 3.4.5      Each steam generator shall be OPERABLE.
APPLICABILITY:          MODES 1,2, 3, and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing Tavg above 2000 F.
SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
4.4.5.3 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of nonplugged tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
: a. Where experience in similar plants with similar water chemistry indicates critical I
areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
: b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
: 1)     All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
: 2)     Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2                 3/4 4-12                 Unit 1 - Amendment No.
Unit 2 - Amendment No.

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
: 3)    A tube Inspection (pursuant to Specification 4.4.5.4a.9) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
: c. The tubes selected as the second and third samples (if required by Table 4.4-2 during each inservice inspection may be subjected to a partial tube inspection provided:
: 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
: 2)   The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories.
Categorv                        Inspection Results C-1              Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2              One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3               More than 10% of the total tubes inspected are degraded tubes or more than 1 % of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
SOUTH TEXAS - UNITS 1 & 2                    3/4 4-13                  Unit 1 - Amendment No.
Unit 2 - Amendment No.

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
: a. As used in this specification:
: 1)  Tube means that portion of the tube which forms the primary system to secondary system pressure boundary;
: 2)  Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
: 3)   Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube;
: 4)   Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
: 5)   % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
: 6)   Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;
: 7)   Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness;
: 8)  Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above;
: 9)  Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg;
: 10)  Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
SOUTH TEXAS - UNITS 1 & 2                3/4 4-15                      Unit 1 - Amendment No.
Unit 2 - Amendment No.

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
: b. The steam generator shall be determined OPERABLE after plugging all tubes exceeding the plugging limit and all tubes containing through-wall cracks required by Table 4.4-2.
4.4.5.5   Reports
: a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; I
: b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
: 1) Number and extent of tubes inspected,
: 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
: 3) Identification of tubes plugged.
: c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
SOUTH TEXAS - UNITS 1 & 2                   3/4 4-16              Unit 1 -Amendment No.
Unit 2 - Amendment No.

Table 4.4-2 STEAM GENERATOR TUBE INSPECTION 1st SAMPLE INSPECTION                              2nd SAMPLE INSPECTION              3rd SAMPLE INSPECTION Sample Size          Result          Action Required           Result          Action Required     Result     Action Required A minimum f S         C-1                      None             N.A.                  N.A.          N.A.             N.A.
tubes per SG.         C-2      Plug defective tubes             C-1                  None           N.A.             N.A.
and inspect additional 2S tubes in this SG.             C-2        Plug defective tubes     C-1             None and inspect additional 4S tubes in this SG      C-2    Plug defective tubes C-3    Perform action for C 3 result of first sample C-3        Perform action for       N.A.             N.A.
C-3 result of first
________sample.
C-3      Inspect all tubes in this   All other                                N.A.             N.A.
SG, plug defective          SGs are C-1               None tubes and inspect 2S tubes in each other         Some SGs        Perform action for C-2   N.A.             N.A.
SG.                         C-2 but no      result of second additional      sample Notify NRC pursuant to      SG are C-3 10CFR50.72(b)(3)(ii)
Additional      Inspect all tubes in     N.A.             N.A.
SG is C-3       each SG and plug defective tubes.
Notify NRC pursuant to 10CFR50.72(b)(3)(ii)
S=3N%        where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an n
inspection.
SOUTH TEXAS - UNITS 1 & 2                                         3/4 4-18                                  Unit 1 - Amendment No.
Unit 2 - Amendment No.

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION I
MAXIMUM NUMBER OF INOPERABLE            MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY                 NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR              (PERCENT OF RATED THERMAL POWER) 1                                           61 2                                           43 3                                           26 SOUTH TEXAS - UNITS 1 & 2        3/4 7-2          Unit 1 - Amendment No.
Unit 2 - Amendment No.

NOC-AE-02001272 Attachment 4 Bases (For Information Only)

NOC-AE-02001272 Attachment 4 Page 1 of 3 POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS In Mode 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limiting transient. In all other Modes, the power level is low enough that the DNB is not a concern.
The values presented in the COLR are indicated values and include measurement uncertainties. The value for pressurizer pressure is averaged using plant computer/QDPS readings from a minimum of at least 3 channels. The value for RCS coolant average temperature is averaged using control board readings from a minimum of at least 3 channels.
The value for RCS flow rate is the average from a minimum of at least 2 flow transmitters per RCS loop using plant computer/QDPS points.
The value for thermal design RCS flow rate presented in Technical Specification 3.2.5 is an analytical limit. The minimum thermal design RCS flow rate is 392,000 gpm. To provide additional operating margin, a higher value for thermal design flow rate may be used if supported by cycle specific analysis. The minimum measured flow in the Core Operating Limits Report is the thermal design flow rate assumed for a particular cycle plus RCS flow measurement uncertainties. The RCS flow measurement uncertainty is 2.8% using the precision heat balance method or 2.1% using the elbow tap methods described in WCAP 15287, "RCS Flow Measurement for the South Texas Projects Using Elbow Tap Methodology",
dated August, 1999. The elbow tap Dp measurement uncertainty presumes that elbow tap Dp measurements are obtained from either QDPS or the plant process computer. Based on instrument uncertainty assumptions, RCS flow measurements using either the precision heat balance or the elbow tap Dp measurement methods are to be performed at greater than or equal to 90% RTP at the beginning of a new fuel cycle.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

NOC-AE-02001272 Attachment 4 Page 2 of 3 REACTOR COOLANT SYSTEM BASES STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to minimize corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the 3.4.6.2.c limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System. Cracks having a primary-to secondary leakage less than this limit during operation have a reasonably high likelihood of achieving "leak-before-break" conditions. Operating plants have demonstrated that primary-to secondary leakage as low as 150 gallons per day per steam generator can readily be detected.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

NOC-AE-02001272 Attachment 4 Page 3 of 3 REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE Pressure boundary leakage of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any pressure boundary leakage requires the unit to be promptly placed in cold shutdown.
Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
The primary-to-secondary accident-induced leakage rate for the limiting design basis accident other than the steam generator tube rupture shall not exceed the leakage rate assumed in the safety analysis in terms of the total leakage rate for all steam generators, and the leakage rate for an individual steam generator. The total leakage shall not exceed 1 gpm. The steam generator tube leakage limit of 150 gpd for each steam generator not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 150 gpd limit per steam generator is conservative compared to the assumptions used in the analysis of these accidents.
The 10 gpm identified leakage limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of unidentified leakage by the leakage detection systems.
The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

NOC-AE-02001272 Attachment 5 Proposed Unit 2 Operating License Changes (Mark-up)

(1)     STPNOC pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the designated location in Matagorda County, Texas, in accordance with the procedures and limitations set forth in this license; (2)     Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and the City of Austin, Texas (COA), pursuant to the Act and 10 CFR Part 50, to possess the facility at the designated location in Matagorda County, Texas, in accordance with the procedures and limitations set forth in this license; (3)     STPNOC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)    Maximum Power Level STPNOC is authorized to operate the facility at reactor core power levels not in excess of 3,853 megawatts thermal (100% power) (Mod. i A.. steam *en..ator.
m*stalled) or aonM 3,800       th       (10d%
heal     ptonr) (Moedel E steamgnrarein 4nstalleo) in accordance with the conditions specified herein.
Amendment No. 427

NOC-AE-02001272 Attachment 6 Proposed Unit 2 Operating License Page (Retyped)

(1)   STPNOC pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the designated location in Matagorda County, Texas, in accordance with the procedures and limitations set forth in this license; (2)   Houston Lighting & Power Company (HL&P), the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and the City of Austin, Texas (COA), pursuant to the Act and 10 CFR Part 50, to possess the facility at the designated location in Matagorda County, Texas, in accordance with the procedures and limitations set forth in this license; (3)    STPNOC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)    STPNOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)    Maximum Power Level STPNOC is authorized to operate the facility at reactor core power levels not in excess of 3,853 megawatts thermal (100% power) in accordance with the conditions specified herein.
Amendment No.}}