ML20211G228
| ML20211G228 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 08/24/1999 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20211G208 | List: |
| References | |
| NUDOCS 9908310207 | |
| Download: ML20211G228 (24) | |
Text
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' NOC-AE-030601 Attachment 2 Page 1 of 6 ATTACHMENT 2 MARKED UP PAGES TO THE TECHNICAL SPECIFICATIONS l
9908310207 990824 DR ADOCK 0 4 ,9 8
> > NOC-AE-CC0601 Attachment 2 Page 2 of 6 PROPOSED CH ANGES TO TECHNICAL SPECIFICATIONS
Reference:
- 1) I2tter from T. H. Cloninger, STP Nuclear Operating Company, to the Nuclear Regulatory Commission Document Control Desk dated June 7,'1999 (NOC-AE-000471) )
- 2) 12tter from T. H. Cloninger, STP Nuclear Operating Company, to the Nuclear Regulatory Commission Document Control Desk dated June 24,1999 (NOC-AE-000569)
The Technical Specification pages listed below are marked-up to identify the proposed changes !
associated with this submittal and are enclosed in this attachment. The attached pages (including inserts) supercede and replace the marked-up pages submitted in the referenced letters above.
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SOUTH TEXAS -UNITS 1 & 2 2-8 Unit 1 - Amendment No. {
Unit 2 - Amendment . . No. &{r'9
ADMINISTRATIVE CONTROLS MONTHLY OPERATING 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 Manage-ment, 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 the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle, or any part of a reload cycle for the following:
gq a)". Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3/4.1.1.3, M.J. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 5 ,L Control Bank Insertion Limits for Specification 3/4.1.3.6,
(,,g. Axial Flux Difference limits and target band for Specification 3/4.2.1,
- .J.
Heat and (FFg) Hot Channel Factor, for Specification K(Z),end 3/4.2.2, Power *- Factor Multiplier, y
8.g. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor IMtrY g .--> Multiplier for Specification 3/4.2.3xy oM 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 (H 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, jnQ 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor 4 osed 3 2. r- ON8 Parame.tces.}
- 2. h f. M CAP 12942-P-A,
- SAFETY EVALUATION SUPPORTING A MORE
, i;EGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT TECHNICAL
,d'!PECIFICATION FOR THE SOUTH TEXAS PROJECT ELECTRIC g-GENERATING STATION UNITS 1 AND 2."
t SOUTH TEXAS - UNITS 1 & 2 6-21 Unit 1 - Amendment No. A-GF,46, 44 Unit 2 - Amendment No h 4+, 84h 46;-
m ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT Runtinuedi (Methodology for Specification 3.1.1.3 - Hoderator Temperature gg9 Coefficient) y * ,E' WCAP 8385, " POWER DISTRIBUTION AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT", September,1974 (l! Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).) I f -k 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.)
G. 4T 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. J (Methodology for Specification 3.2.1 - Axial flux Difference (Constant Axial Offset Control).) ,
l Ga. WCAP-10266-P-A, Rev. 2, WCAP-11524-NP-A, Rev. 2, "The 1981 !
7- l '
Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", Kabadi, J.N., et al., March 1987; including Addendum 1-A, " Power Shape Sensitivity Studies," December 1987 1 and Addendum 2-A, " BASH Methodology Improvements and Reliability Enhancements" May 1988.
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel l Factor.)
8, 4b. WCAP-12610-P-A, " VANTAGE + Fuel Assembly Reference Core Report,"
April,1995 (H Proprietary) for loss of Coolant Accident (LOCA)
Evaluation models with ZlRLO clad fuel for rod heatup 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 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.
main T rv As um 1 s i r, 2 6-?? Unit 1 Awndent No. FI-M . 5 7, 7 ? . g.
t
, NoC-AE-000(OI Attachment 2 Page 6 of 6 Insert 4
- 1. Safety limits for thermal power, pressurizer pressure, and the highest operating loop coolant temperature (T,vg) for Specification 2.1,
- 2. Limiting Safety System Settings for Reactor Coolant Flow-Low Loop design flow, Overtemperature AT, and Overpower AT setpoint parameter values for Specification 2.2, Insert 5
- 9. DNB related parameters for Reactor Coolant System T,vg, Pressurizer Pressure, and the Minimum Measured Reactor Coolant System Flow for Specification 3/4.2.5.
Insert 6
- 3. WCAP-8745-P-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary Class 2)
(Methodology for Specification 2.1 - Safety Limits, and 2.2 - Limiting Safety System Settings) l l
l
o NOC- AE-0C0601 Attachment 3 Page 1 of 18 l
l ATTACHMENT 3 l l
i 1
Reconstituted Technical Specification Pages 1
l l
l l
1
INDEX 2.0' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 R E ACT O R C O R E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.2 REACTO R COOLANT SYSTEM PRESSU R E. . .. .. .. . . ... .. ....... .. . .. ... .. . . . .. ... .... . . 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS.......................... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS........ 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS l
2.1.1 R EA CTO R CO R E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-1 1 2.1.2 REACTOR COOLANT SYSTEM PRESSU R E.. ... ...... ... ....... ..... ..... . ..... ........... B 2-2 l
2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ........................ B 2-3 l
l l
l SOUTH TEXAS - UNITS 1 & 2 iii Unit 1 - Amendment No. 97, Unit 2 - Amendment No. 84,
o 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,va) shall not exceed the limits shown the Core Operating Limits Report.
-2.1.1.1 in MODES 1 and 2, the departure from nucleate boiling ratio (DNBR) shall be maintained 2 171 for the WRB-1 DNB correlation.
2.1.1.2 in MODES 1 and 2, the peak fuel centerline temperature shall be maintained < 5080 F, decreasing by 58 'F per 10,000 MWD /MTU of burnup.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
I REACTOR COOLANT SYSTEM PRESSURE l
l 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
1 APPLICABILITY: MODES 1,2,3,4, AND 5.
l l ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3,4 and 5:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.
I l
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I SOUTH TEXAS - UNITS 1 & 2 2-1 UNIT 1 - Amendment No. 97, UNIT 2 - Amendment No. 84, i
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l SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Cafety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during i
operation and therefore THERMAL POWEn and reactor coolant temperature and pressure have been l related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been developed to predict the DNB nux and the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.
The DNB design basis is as follows: uncertainties in the WRB-1 correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with a 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and 11 events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties, in addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.
The reactor core Safety Limits are established to preclude violation of the following fuel design riteria:
- a. There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and
- b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.
The reactor core Safety Limits are used to define the various Reactor Protection System (RPS) functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower AT I reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less ,
than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the i DNBR correlation. Appropriate functioning of the RPS ensures that, for variations in the Thermal Power, RCS Pressure, RCS average temperature, RCS flow rate, and Al, the reactor core Safety Limits will be satisfied during steady state operation, normal operational transients, and AOOs.
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SOUTH TEXAS - UNITS 1 & 2 B 2-1 Unit 1 - Amendment No. 64, Unit 2 - Amendment No. 60, 48246-07--
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de LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Ranoe. Neutron Flux l
l The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about .
5 10 counts per second unless manually blocked when P-6 becomes active. The Source Range l l channels are automatically blocked above P-10. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
Overtemperature AT l l
The Overtemperature AT trip provides core protection to prevent DNB for all combinations of I pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow l with respect to piping transit delays from the core to the temperature detectors, and pressure is within '
l the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power l distribution, this Reactor trip limit is always below the core Safety Limit as shown in the Core Operating Limits Report. If axial peaks are greater than design, as indicated by the difference between top and I bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1 and as specified in the Core Operating Limits Report.
l Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all pos5ible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/tt) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, ' Reactor Core Response to Excessive-Secondary Steam Releases."
SOUTH TEXAS - UNITS 1 & 2 B 2-5 Unit 1 - Amendment No.
Unit 2 - Amendment No.
e LIMITING SAFETY SYSTEM SETTINGS BASES 1
Pressurizer Pressure '
In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure. 1 On decreasing power, the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL 4 POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent), an '
automatic Reactor trip will occur if the flow in more than one loop drops below approximately 92% of nominal full loop flow. Above P-8 (a power level of approximately 40% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below approximately 92% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7, an automatic Reactor trip will occur on low reactor coolant flow in more than one loop, and below P-7 the trip function is automatically blocked. The value for loop design flow is the analytical value consistent with the thermal design flow assumed in the DNB analysis.
)
Steam Generator Water Level l The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in i the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The .
I specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System. l i
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l SOUTH TEXAS - UNITS 1 & 2 B 2-6 Unit 1 - Amendment No. 1 Unit 2 - Amendment No.
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, l 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 T.yg, s 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,2 3'0,000 gpm APPLICABILITY: MODE 1. l ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 2 90% RTP.
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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, j SOUTH TEXAS - UNITS 1 & 2 3/4 2-11 Unit 1 - Amendment No. 61 97,108, Unit 2 - Amendment No. 60r84r96,
F POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
! (Continued)
When an Fa measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor, Fo(Z), remains within its limit. The Fy limit for RATED THERMAL POWER (F,y") as i provided in the Core Operating Limits Reports (COLR) per Specification 6.9.1.6 was determined from l expected power control manuevers over the full range of burnup conditions in the core.
3/4.2.4 OUADRANT POWER TILT RATIO l
1 The OUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies l the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
! For purposes of monitoring OUADRANT POWER TILT RATIO when one excore detector is I
inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the OUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H 3, H-13, L-5, L-11, N-8.
3/4.2.5 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 /ODPS points.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 Unit 1 - Amendment No. 27, Unit 2 - Amendment No. 47,
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j ie l POWER DISTRIBUTION LIMITS 1
BASES 3/4.2.5 DNB PARAMETERS (Continued)
The valuo for RCS flow rate presented in Technical Specification 3.2.5 is the thermal design reactor coolant system flow rate used in the analysis approved by the Nuclear Regulatory Commission in Amendments 97 and 84 on September 29,1998. This flow rate is an analytical limit consistent with 10% plugging of the steam generator tubes and Departure from Nucleate Boiling requirements.
l The RCS flow measurement uncertainty of 2.8% bounds the precision heat balance and the elbow tap Ap measurement methods. The elbow tap Ap measurement uncertainty presumes that elbow tap Ap measurements are obtained from either ODPS or the plant process computer. Based on instrument uncertainty assumptions, RCS flow measurements using either the precision heat balance or the elbow tap Ap measurement methods are to be performed at greater than or equal to 90% RTP at the beginning of a new fuel cycle. The elbow tap Ap RCS flow measurement methodology is described-in ST-HL-AE-5707, " Proposed Amendment to Technical Specification Table 2.2-1 and 3/4.2.5 for Reactoi Coolant System Flow Monitoring - Revised," dated August 6,1997, and in ST-HL-AE-5752,
" Amended Response to Request for Additional Information on the Proposed Elbow Tap Technical Specification Change (Table 2.2-1 and Section 3/4.2.5)," dated September 18,1997.
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.
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SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-6 Unit 1 - Amendment No. S1,97,108, l
Unit 2 - Amendment No. 50, Bd.,95, 3429-98 7730-40 L
ADMINISTRATIVE CONTROLS MONTHLY OPERATING 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, 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 the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall 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. Safety limits for thermal power, pressurizer pressure, and the highest operating loop coolant temperature (T.,,) for Specification 2.1, l
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- 2. Limiting Safety System Settings for Reactor Coolant Flow-Low Loop design flow, 1 Overtemperature AT, and Overpower AT setpoint paiemeter values for Specification 2.2,
- 3. Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit l for Specification 3/4.1.1.3, J
- 4. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, l
- 5. Control Bank Insertion Limits for Specification 3/4.1.3.6,
- 6. Axial Flux Difference limits and target band for Specification 3/4.2.1, l
- 7. Heat Flux Hot Channel Fa'c tor, K(Z), Power Factor Multiplier, and (F/TP g Specification 3/4.2.2, l
- 8. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification j 3/4.2.3, and
- 9. DNB related parameters for Reactor Coolant System T.ya, Pressurizer Pressure, and the Minimum Measured Reactor Coolant System Flow for Specification 3/4.2.5.
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 l Difference,3.2.2 - Heat Flux Hot Channel Factor,3.2.3 - Nuclear Enthalpy Rise Hot Channel j Factor, and 3.2.5 - DNB Parameters.)
SOUTH TEXAS - UN!TS 1 & 2 6-21 Unit 1 - Amendment No. 9,27,35,47, Unit 2 - Amendment No.1,17,26r-36, '
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<o ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
- 2. WCAP 12942-P-A," SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL !
MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNITS 1 AND 2."
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)
- 3. WCAP-8745-P-A," Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary Class 2)
(Methodology for Specification 2.1 - Safety Limits, and 2.2 - Limiting Safety System Settings)
- 4. 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).)
l l 5. Westinghouse letter NS-TMA-2198, T.M. Anderson (Westinghouse) to K. Kniel (Chief of l Core Performance Branch, NRC) January 31,1980 -
Attachment:
Operation and Safety Analysis Aspects of an improved Load Follow Package.
l l (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.)
- 6. NUREG-0800, Standard Review Plan, U. S. Nuclear Regulatory Commission, Section 4.3, l 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).)
- 7. WCAP-10266-P-A, Rev. 2, WCAP-11524-NP-A, Rev. 2, "The 1981 Version of the l !
Westinghouse ECCS Evaluation Model Using the BASH Code", Kabadi, J.N., et al., March l 1987; including Addendum 1-A, " Power Shape Sensitivity Studies," December 1987 and !
Addendum 2-A, " BASH Methodology improvements and Reliability Enhancements" May 1988. !
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) i
- 8. WCAP-12610-P-A," VANTAGE + Fuel Assembly Reference Core Report," April,1995 (W l ,
Proprietary) for Loss of Coolant Accident (LOCA) Evaluation models with ZlRLO clad fuel for rod heatup 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 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 safsty analysis are met.
SOUTH TEXAS - UNITS 1 & 2 6-22 Unit 1 - Amendment No. 27,35, 47,72, S9, Unit 2 - Amendment No.17,2S, SS, S1,76,