ML21090A063
| ML21090A063 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/08/2002 |
| From: | Beasley J Southern Nuclear Operating Co |
| To: | Document Control Desk |
| Lamb, J. | |
| References | |
| LCV-1617 | |
| Download: ML21090A063 (52) | |
Text
J. Barnie Beasley, Jr., P.E. Southern Nuclear Vice President Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7110 Fax 205.992.0403 SOUTHERN"\'
COMPANY Energy to Serve Your World'"
May 8, 2002 LCV-1617 Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REACTOR TRIP SYSTEM INSTRUMENTATION OVER TEMPERATURE DELTA TEMPERATURE (OT~T) AND OVERPOWER DELTA TEMPERATURE (OP~T)
REACTOR TRIP FUNCTIONS Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS) Figure 2.1.1-1, "Reactor Core Safety Limits"; Table 3.3.1-1, "Reactor Trip System Instrumentation"; and the associated Bases B2.1.1 and B 3.3 .1.
Both VEGP units have experienced steady-state aperiodic hot leg temperature fluctuations.
This is not a unique VEGP phenomenon. Similar effects have been noted at other Westinghouse plants. Although no definitive causes for the temperature fluctuations have been identified, they are believed to be caused by the upper plenum flow anomalies. Three resistance temperature detectors (RTDs) are used to account for temperature streaming in order to provide an average temperature in each hot leg. Since the average of the three RTDs is used to represent the hot leg temperature (Thot), temperature fluctuation from any RTD can adversely affect the calculation of the average Thot temperature. This, in turn, impacts the average reactor coolant system (RCS) temperature (Tavg) and RCS differential temperature
(~T) and therefore reduces the steady-state operating margins to Overtemperature Delta Temperature (OT~T) and Overpower Delta Temperature (OP~T) trip setpoints. The temperature fluctuations of interest are fluctuations in the increasing direction.
U. S. Nuclear Regulatory Commission LCV-1617 Page 2 To address the impact of the temperature fluctuations on operating margin, VEGP has implemented a number of changes to improve operating margin. These changes include:
increasing the OT~T and OP~T reference temperature to above the RCS average temperature, reducing the turbine ronback setpoint, adding a filter to the measured RCS differential temperature, and adding a filter to the temperature signal to the rod control system to reduce the frequency of spurious rod stepping.
To accommodate the effects of streaming and the associated hot leg temperature fluctuations, SNC proposes to increase the OT~T and OP~T setpoints. This program to increase the setpoints is referred to as the OT~T and OP~T Setpoint Margin Recovery Program (or MRP for short). The intent of the MRP is to revise the OT~T and OP~T setpoints to increase operating margin. This is accomplished by increasing the steady-state setpoints and by revising the dynamic compensation time constants in the setpoint equations. The setpoint allowable values and core safety limits were also revised to support the MRP.
The analyses supporting the MRP assume a revision to the Relaxed Axial Offset Control (RAOC) band and the inclusion of a limit or clamp on the compensated temperature difference term in the OT~T trip setpoint. The revision to the RAOC band and the limit or clamp on the compensated temperature difference term in the OT~T trip setpoint are currently under review by the Staff. These revisions were submitted for review in a separate amendment request (SNC letter LCV-1563 dated October 30,2001). Implementation of the MRP setpoint changes is contingent upon approval of the amendment request for the revised RAOC band and clamp. provides the basis for the proposed change. Pursuant to 10 CFR 50.92, demonstrates that the proposed change does not involve a significant hazards consideration. Enclosure 3 contains a mark-up of the affected pages from the current VEGP Technical Specifications and Bases as well as pages from the amendment request described in the previous paragraph. Enclosure 4 contains the typed version of the revised Technical Specification and Bases pages. SNC has determined that the proposed license amendment will not significantly affect the quality of the environment.
Upon approval of the amendment request in SNC letter LCV-1563, SNC will re-submit the marked-up and typed pages in Enclosures 3 and 4 to reflect what will become of the current Technical Specification and Bases pages.
SNC requests that the proposed changes be approved by September 15,2002. The changes are planned to be implemented during the next Unit 2 refueling outage in Fall 2002 and the next Unit 1 refueling outage in Fall 2003.
- u. S. Nuclear Regulatory Commission LCV-1617 Page 3 Mr. J. B. Beasley, Jr., states that he is a Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company and that, to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectfully submitted, Sworn to and subscribed before me this ~ day of 711~02
~(].~
Notary Public My commission expires: 11;1 (J~;L-
/ 7 JBBIRJF Enclosures
- 1. Basis for Proposed Change
- 2. 10 CFR 50.92 Significant Hazards Evaluation
- 3. Marked-Up Technical Specification and Bases Pages
- 4. Typed Revised Technical Specification and Bases Pages cc: Southern Nuclear Operating Company Mr. 1. T. Gasser Mr. M. Sheibani SNC Document Management
- u. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. F. Rinaldi, Project Manager, NRR Mr. John Zeiler, Senior Resident Inspector, Vogtle State of Georgia Mr. L. C. Barrett, Commissioner, Department of Natural Resources
ENCLOSURE 1 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REACTOR TRIP SYSTEM INSTRUMENTATION OVER TEMPERATURE DELTA TEMPERATURE (OTDT) AND OVERPOWER DELTA TEMPERATURE (OPDT) REACTOR TRIP FUNCTIONS BASIS FOR PROPOSED CHANGE PROPOSED CHANGE In accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS) Figure 2.l.1-1, "Reactor Core Safety Limits"; Table 3.3.1-1, "Reactor Trip System Instrumentation";
and the associated Bases B2.l.l and B 3.3.l.
Both VEGP units have experienced steady-state aperiodic hot leg temperature fluctuations. This is not a unique VEGP phenomenon. Similar effects have been noted at other Westinghouse plants. Although no definitive causes for the temperature fluctuations have been identified, they are believed to be caused by the upper plenum flow anomalies. Three resistance temperature detectors (RIDs) are used to account for temperature streaming in order to provide an average temperature in each hot leg. Since the average of the three RTDs is used to represent the hot leg temperature (Thot), temperature fluctuation from any RTD can adversely affect the calculation of the average Thot temperature. This, in turn, impacts the average reactor coolant system (RCS) temperature (Tavg) and RCS differential temperature (i1T) and therefore reduces the steady-state operating margins to Overtemperature Delta Temperature (OTi1T) and Overpower Delta Temperature (OPi1T) trip setpoints. The temperature fluctuations of interest are fluctuations in the increasing direction.
To address the impact of the temperature fluctuations on operating margin, VEGP has implemented a number of changes to improve operating margin. These changes include: increasing the OTi1T and OPi1T reference temperature to above the RCS average temperature, reducing the turbine runback setpoint, adding a filter to the measured RCS differential temperature, and adding a filter to the temperature signal to the rod control system to reduce the frequency of spurious rod stepping.
To accommodate the effects of streaming and the associated hot leg temperature fluctuations, SNC proposes to increase the OTi1T and OPi1T setpoints. This program to increase the setpoints is referred to as the OTi1T and OPi1T Setpoint Margin Recovery Program (or MRP for short). The intent of the MRP is to revise the OTi1T and OPi1T setpoints to increase operating margin. This is accomplished by increasing the steady-state setpoints and by revising the dynamic compensation time constants in the setpoint equations. The setpoint allowable values and core safety limits were also revised to support the MRP.
The analyses supporting the MRP assume a revision to the Relaxed Axial Offset Control (RAOC) band and the inclusion ofa limit or clamp on the compensated temperature difference term in the OTi1T trip setpoint. The revision to the RAOC band and the limit or clamp on the compensated temperature difference term in the OTi1T trip setpoint are currently under review by the Staff. These revisions were submitted for review in a separate amendment request (SNC letter LCV-1563 dated October 30,2001).
Implementation of the MRP setpoint changes is contingent upon approval of the amendment request for the revised RAOC band and clamp.
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BASIS FOR PROPOSED CHANGE
1.0 INTRODUCTION
The Margin Recovery Program (MRP) initially focused on the revision of the OT~T and OP~T trip setpoints to increase operating margins. The MRP also included revising the Relaxed Axial Offset Control (RAOC) band (negative side), revising core thermal limits, revising dynamic compensation terms in the OT~T and OP~T equations, increasing filter time constants, increasing RTD response times, and placing a limit or clamp on the compensated temperature difference term in the OT~T setpoint.
Technical Specifications allowable values were recalculated and reformatted.
The increase of the OT~T and OP~T setpoints has been achieved by means of revised core thermal analyses and revised core thermal limits. The revised OT~T and OP~T setpoints and dynamic compensations are provided in Section 2.0. Section 3.1 provides a detailed discussion of the revised core analyses and revised core thermal limits.
The revision to the RAOC band and the limit or clamp on the compensated temperature difference term in the OT~ T trip setpoint are currently under review by the Staff. These revisions were submitted for review in a separate amendment request (SNC letter LCV-1563 dated October 30, 2001). The analyses supporting the MRP are based on the revised RAOC band and the clamp on the OT~T trip setpoint.
Implementation of the MRP setpoint changes is contingent upon approval of the amendment request referenced above.
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2.0 REVISED OT~T AND OP~T SETPOINTS The revised OT~T and OP~T setpoint parameters for both VEGP units are shown below. The setpoint equations and terms in the setpoint equations are defined in Technical Specification Table 3.3.1-1.
OTAT SETPOINT PARAMETERS Parameter Current Revised KI 1.12 1.149 K2 0.0224/oF 0.0224/oF K3 0.00115/psi 0.00177/psi
'tl 8 sec o sec
't2 3 sec o sec
't3 2 sec 6 sec
't4 28 sec 28 sec
't5 4 sec 4 sec
't6 o sec 6 sec fl(AFD) penalty Breakpoints -32%, +10% -23%, + 10% (I)
Negative slope 3.25%/%AFD 3.3%/%AFD (I)
Positive slope 2.7%/%AFD 1.95%/%AFD (I)
Note 1 These values were used in the analyses supporting the MRP. They are being changed under a separate amendment request (SNC letter LCV-1563 dated October 30,2001) currently under review by the Staff.
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OPAT SETPOINT PARAMETERS Parameter Current Revised Comment Kt 1.095 1.10 Ks 0.02/°P 0.02/°P for increasing Tavg 0.0 0.0 for decreasing Tavg
~ 0.002/°P 0.00244/°P 1>T" 0.0 0.0 T<T"
"[7 10 sec 10 sec
"[) 8 sec Osee
"[2 3 sec Osee
"[3 2 sec 6 sec
"[6 Osee 6 sec f2(APD) penalty 0.0% 0.0%
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3.0 ANALYSES AND EVALUATIONS 3.1 CORE THERMAL-HYDRAULIC, NUCLEAR, AND FUEL ROD DESIGN ANALYSES 3.1.1 Introduction and Background This section describes the core thermal-hydraulic, nuclear, and fuel rod design analyses inputs provided to support the Margin Recovery Program (MRP) for Vogtle Electric Generating Plant (VEGP) Units 1 and 2.
3.1.2 Core Thermal-Hydraulic Design The main purpose of the VEGP MRP was to increase the OTilT and OPilT reactor trip setpoints and the associated dynamic compensation. To support the increase of the OTilT and OPilT setpoints, revised thermal-hydraulic design analyses were performed which included the elimination of overly conservative assumptions in the analyses as well as the reallocation of margins built into the current analyses.
The revised analysis parameters for the VEGP MRP are more representative of current VEGP conditions, i.e., the minimum measured flow for the departure from nucleate boiling (DNB) analyses is consistent with the current limit in the Technical Specifications, the revised bypass flow fraction is consistent with the use of thimble plugs in both VEGP cores, and the reduced LOPAR F6H limit will be bounding for VEGP reload core designs which have only used a limited number of reinserted LOPAR assemblies located in lower power core positions. The DNB analyses for the VEGP MRP assume that future VEGP core designs are primarily VANTAGE+ fuel. The DNB analyses to support the continued reinsertion of the LOPAR fuel into low power core positions is addressed in Section 3.1.2.2.3.
The core inlet temperature used in the DNB analyses for the MRP is based on the upper bound of the RCS temperature range which was analyzed for the VEGP Rerate Program (Amendment 60 for Unit 1 and Amendment 39 for Unit 2 dated March 22, 1993). Use of the upper bound temperature conservatively addresses the temperature range for the MRP DNB analyses.
3.1.2.1 Description of Thermal-Hydraulic Calculational Methods The thermal-hydraulic design criteria and methods for the VEGP MRP remain the same as those presented in the VEGP Updated Final Safety Analysis Report (UFSAR). As discussed in the UFSAR, the primary DNB correlation used in the analysis of the VANTAGE+ fuel is the WRB-2 DNB correlation.
The WRB-l DNB correlation is used for the analysis of the LOPAR fuel. The W-3 DNB correlation is used where the primary DNB correlations are not applicable. The improved THINC-IV PWR design modeling method (WCAP-12330-A, "Improved THINC-IV Modeling for PWR Core Design," September 1991) continues to be used in the DNB ratio (DNBR) analyses for the Vogtle units.
As discussed in the VEGP UFSAR, the design method employed to meet the DNB design basis for the VANTAGE+ fuel and LOPAR fuel is the Revised Thermal Design Procedure (RTDP) (WCAP-l1397-P-A, "Revised Thermal Design Procedure," April 1989) for treating uncertainties in reactor power, primary coolant temperature, pressurizer pressure, and RCS flow. The RTDP design limit DNBR values specified in the VEGP UFSAR for VEGP Units 1 and 2 are unchanged for the MRP (1.24 and 1.23 for the typical and thimble cells, respectively, for VANTAGE+ fuel, and 1.23 and 1.22 for the typical and thimble cells, respectively, for LOPAR fuel).
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In addition to the above considerations for uncertainties, additional DNBR margin was maintained by performing the safety analyses to DNBR limits higher than the design limit DNBR values. Sufficient DNBR margin was maintained in the safety analysis DNBR limits to offset the known DNBR penalties.
The net remaining DNBR margin, after consideration of the DNBR penalties, is available for operating and design flexibility (e.g., VANTAGE+ transition core DNBR penalty associated with the limited insertion of LOPAR fuel) as discussed in VEGP UFSAR Section 4.4.1.
3.1.2.2 DNB Performance The DNB analyses which were performed to support the VEGP MRP are addressed below.
3.1.2.2.1 Revised Core Thermal Limits To support the revised OT~T and OP~T trip setpoints for the VEGP MRP, the DNB core thermal limit lines for VEGP were revised from the current limits. For the current VEGP DNB core thermal limits, the LOPAR fuel is the most limiting fuel design. The revised DNB core thermal limits for the VEGP MRP are based on V ANTAGE+ as the most limiting fuel design. The Fllli limit for the VANTAGE+ fuel remains at the current value of 1.65. The VANTAGE+ safety analysis DNBR limits for the MRP analysis were reduced by decreasing the DNBR margin which was retained in the current safety analysis DNBR limits. The DNBR margin (approximately 17%) which was previously retained was necessary to offset the transition core DNBR penalty associated with the first LOPAR-to-VANTAGE-5 transition cycle.
Since the current VEGP core designs are primarily VANTAGE+ (or all VANTAGE+), the amount of retained DNBR margin needed to address any limited reinsertion of LOPAR fuel assemblies is significantly reduced. The new reduced VANTAGE+ safety analysis DNBR limits resulted in less restrictive DNB core thermal limits to support the improved OThT trip setpoints.
The DNB analyses to support the continued reinsertion of the LOPAR fuel into low power core positions is addressed in Section 3.1.2.2.3.
3.1.2.2.2 Revised RAOC Band The RAOC band used in the MRP analyses is + 10% and -15% AFD at 100% power and +26% and -30%
AFD at 50% power. As discussed in Section 1.0, this is the same RAOC band used to support the amendment request currently under review by the Staff(SNC letter LCV-1563 dated October 30,2001).
3.1.2.2.3 LOPAR Fuel The DNBR analyses for the revised core limits assumed that the VEGP core designs are primarily VANTAGE+ fuel. To ensure that the LOPAR fuel is not limiting with respect to DNB, the LOPAR Fllli limit was reduced from the current limit of 1.53 in the Core Operating Limits Report (COLR) to a value of 1.30. Only a limited number of LOPAR fuel assemblies may be reinserted in low power core locations in future VEGP core designs. Operation with a mixed core ofVANTAGE+ and LOPAR fuel is still addressed using the transition core DNB methodology described in VEGP UFSAR Section 4.4.2.
The LOP AR safety analysis DNBR limits are unchanged from the limits in the current safety analyses.
The mixed core DNBR effect of the LOPAR fuel on the VANTAGE+ DNBR analyses will continue to be addressed by the application of available DNBR margin. The maximum number of LOPAR fuel assemblies that can be reinserted in a mixed core design is limited by the VANTAGE+ DNBR margin which is available for the specific cycle to offset the transition core DNBR penalty. This will be evaluated on a cycle-specific basis during the reload core design.
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3.1.2.3 Hydraulic Evaluation The VEGP MRP does not impact the hydraulic analyses. The fraction of flow that bypasses the core through the thimble guide tubes was changed for the DNB analyses to reflect operation with thimble plugs in place. However, the VEGP fuel assembly lift forces that were evaluated in support of the power uprate (Amendment 60 for Unit I and Amendment 39 for Unit 2 dated March 22, 1993) bounded operation with or without the use of thimble plugging devices. Therefore, the fuel assembly lift force analysis is not affected by the VEGP MRP.
3.1.2.4 Fuel Temperatures The fuel temperatures currently used in the safety analyses are applicable to the VEGP MRP.
3.1.3 Nuclear Analysis Input 3.1.3.1 RAOC Band For the VEGP MRP The RAOC band used in the MRP analyses is + 10% and -15% AFD at 100% power and +26% and -30%
AFD at 50% power. As discussed in Section 1.0, this is the same RAOC band used to support the amendment request currently under review by the Staff (SNC letter LCV-1563 dated October 30, 2001).
3.1.4 Fuel Rod Design Considerations Fuel rod design considerations can typically impact the negative side of the RAOC band. In fuel rod design analysis using the PAD code (WCAP-I0851-P-A, "Improved Fuel Performance Models for Westinghouse Fuel Rod design and Safety Evaluations," August 1988), individual fuel rod power histories are evaluated to determine the best-estimate, steady-state fuel rod performance parameters. At certain times during the depletion, the rod powers are increased to the appropriate overpower (transient) limits to determine the cladding stress and strain, fuel rod internal pressure, and cladding temperatures.
These conservative transient evaluations are performed to verify adherence to the design criteria of no rod failures considering the effects of transient power increases during Condition I and II events.
The following two sections describe the fuel rod design considerations which, in part, determined the final Vogtle MRP RAOC band.
3.1.4.1 PAD Yield Stress Limit Confirmation The stress analysis, using the transient limits for the MRP, exhibits sufficient margin to the 0.2 percent offset yield stress and, therefore, sufficient margin to the cladding stress and strain design limits. The use of the clamp on the compensated temperature difference term in the OTI':lT trip setpoint has generated sufficient margin that it is unlikely the cladding yield stress limit will be challenged in future VEGP cycles.
The clamp on the compensated temperature difference term in the OTI':lT trip setpoint is currently under review by the Staff in a separate amendment request (SNC letter LCV-1563 dated October 30, 2001).
3.1.4.2 Other Fuel Rod Design Criteria Related to Transient Limits The rod internal pressure analyses must consider the additional fission gas released during a Condition I or Condition II event during each cycle of fuel irradiation. Also, the cladding corrosion criteria include a limit on the cladding temperature during Condition II events. The transient limits generated during the EI-7
MRP are significantly less limiting than the current transient limits. Since the rod internal pressure design limit and cladding temperature limit are met with the current transient limits, these criteria will also be met with the MRP transient limits.
3.1.5 Core Thermal-Hydraulic, Nuclear, and Fuel Rod Design Conclusions As discussed in Sections 3.1.2,3.1.3, and 3.1.4, all DNB, peaking factor, centerline melt, and fuel rod design related limits are met with sufficient margin so as to provide confidence that the MRP RAOC band can be supported for future VEGP cycles. It should be noted that standard methods (WCAP-I0216-P-A, Revision lA, "Relaxation of Constant Axial Offset Control- FQ Surveillance Technical Specification,"
February 1994) will be used on a cycle-specific basis to confirm this.
The above conclusion regarding the validity of the MRP RAOC band assumes the clamp on the compensated temperature difference term in the OT~T trip function, currently under review by the Staff in a separate amendment request (SNC letter LCV-1563 dated October 30, 2001), is approved.
3.2 NON-LOSS OF COOLANT ACCIDENT (NON-LOCA) TRANSIENTS The VEGP MRP incorporates changes that affect the VEGP UFSAR Chapter 15 non-LOCA analyses.
The changes are primarily related to the revised OT~T and OP~T reactor trip functions. Specifically, the setpoints are revised to account for the new core limits and a revised RAOC band. The dynamic compensation terms are modified to include 6 second filters on measured Tavg and measured ~T, a 5.5 second R TD response time, and the time constants on the lead/lag function on measured~T changed from 8/3 to 0/0. A summary of the current and revised OT~T and OP~T setpoint constants and dynamic compensation terms was provided in Section 2.0.
Each of the MRP features as related to the VEGP UFSAR Chapter 15 safety analyses are discussed in Section 3.2.1. The UFSAR Chapter 15 non-LOCA event analyses and evaluations performed to support the MRP are listed in Section 3.2.2.
3.2.1 Margin Recovery Program Analysis Changes 3.2.1.1 Revised Core Thermal Limits Revised core thermal limits were calculated for the VEGP MRP. The core limits are revised to account for a higher Minimum Measured Flow to be consistent with the current reactor coolant system flow measurement uncertainty, a decrease in the core bypass flow (assumes thimble plugs installed), the removal of the fuel transition core penalty (core limits are applicable to VANTAGE-5 and VANTAGE+
fuel), and an overpower limit increase from 118% to 120%.
The revised core limits continue to be based on the Revised Thermal Design Procedure (RTDP) (WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989). The RTDP methodology statistically convolutes the random uncertainties on the plant operating parameters (power, temperature, pressure, and flow) into the design limit DNBR value. The random uncertainties are unchanged from those used in the Vogtle V ANTAGE-5 and Rerating Programs (Amendments 43 and 44 for Unit 1 and Amendments 23 and 24 for Unit 2, dated September 19, 1991, for the use ofVANTAGE-5 fuel; Amendment 60 for Unit 1 and Amendment 39 for Unit 2, dated March 22, 1993, for the rerating).
Additional details concerning the development of the core thermal limits are included in Section 3.1.2.2.1.
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3.2.1.2 Reactor Trip Setpoints The core limits based on the revised analysis assumptions, described above, were used to calculate the safety analysis limit OT~T and OP~T setpoints. The safety analysis limit setpoints are calculated using the approved methodology described in WCAP-8745-P-A (WCAP-874S-P-A, "Design Bases for the Thermal Overpower ~ T and Thermal Overtemperature ~ T Trip Functions," September 1986).
Included in the calculation ofK 1 and~, the fundamental OT~T and OP~T setpoints, respectively, is an allowance for 6.1 OF of uncertainty on the reactor coolant system average temperature (Tavg) measurement.
The allowance accounts for burndown, loop asymmetry, mismatch between protection and control system reference temperatures, and upper plenum anomaly effects.
3.2.1.3 Reactor Coolant System Average Temperature (Tav,J Clamping in OT~T Trip Function The non-LOCA safety analyses specifically credit the OT~T reactor trip function to provide primary protection for several events. The non-LOCA events which credit the OT~T reactor trip function as the primary reactor trip function include the Uncontrolled Bank Withdrawal at Power event (FSAR Section 15.4.2), the RCS Depressurization event (FSAR 15.6.1), the Loss of Load/Turbine Trip event (FSAR IS.2.3), and the Uncontrolled Boron Dilution event (FSAR Section 15.4.6). In general, the thermal hydraulic conditions that occur for these transients tend to reduce the OT~T setpoint, and eventually result in a reactor trip. That is, the events result in an increase in the RCS Tavg (heatup events) or a reduction in pressurizer pressure (RCS depressurization event) which tends to reduce the OT~T setpoint.
The proposed clamp on the compensated temperature difference term in the OT~T trip function will limit how much the OT~T setpoint can increase as a result of decreases in the RCS Tavg. Since the non-LOCA analyses, in general, result in a reduction in the OT~ T setpoint, and the clamp limits the amount that the OT~T setpoint can increase, the non-LOCA safety analyses will not be adversely affected by the proposed clamp.
The clamp on the compensated temperature difference term in the OT~T trip setpoint is currently under review by the Staff in a separate amendment request (SNC letter LCV-1563 dated October 30, 2001).
3.2.1.4 OT~T and OP~T Dynamic Compensation and Trip Response Time The Margin Recovery Program incorporates significant changes to the dynamic compensation of the OT~T and OP~T trip functions. The time constants, "tl and "t2, for the lead/lag function on measured ~T are changed from 8/3 to 0/0. The filter time constant on measured ~T, "t3, is increased from the current value of 2 seconds to 6 seconds. A filter time constant on measured Tavg, "t6, of 6 seconds is added. Also the RTD response time assumed in the analyses is increased from 4 to 5.S seconds.
3.2.2 Non-LOCA Event Analyses and Evaluations The Margin Recovery Program assumptions as related to the non-LOCA analyses were described in Section 3.2.1. The changes are primarily related to the revised OT~T and OP~T reactor trip setpoints and dynamic compensation. With the exception of the minimum measured flow, other key analysis parameters are unchanged from the current licensing basis.
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The following VEGP Units 1 and 2 UFSAR Chapter 15 safety analyses rely on OT~T and OP~T for primary protection and were reanalyzed or evaluated to incorporate the MRP features.
- Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (UFSAR Section 15.4.2, credits OT~T trip function)
- Loss of Extema I Electrical Load and/or Turbine Trip (UFSAR Section 15.2.2 and 15.2.3, credits OT~T trip function)
- Accidental Depressurization of the Reactor Coolant System (UFSAR Section 15.6.1, credits OT~T trip function)
- Steamline Break Core Response at Power (UFSAR Section 15.1.5, credits OP~T trip function) 3.2.3 Non-LOCA Computer Codes Consistent with the current licensing basis analyses presented in the VEGP UFSAR, the non-LOCA events analyzed for the Margin Recovery Program utilized the LOFTRAN computer code (WCAP-7907-P-A, "LOFTRAN Code Description," April 1984).
3.2.4 Conclusions The results of all of the analyses and evaluations demonstrate that applicable safety analysis acceptance criteria have been satisfied at the MRP conditions described in Section 3.2.1. The analysis results for the non-LOCA transients are summarized in Table 1.
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Table 1 Summary of Results of Non-LOCA Event Analyses and Evaluations Event Acceptance Calculated Results Comments Uncontrolled RCCA 1. DNBR> safety analysis limits 1. > limit value DNBR for the limiting case calculated Withdrawal at Power with THINC.
- 3. Secondary pressure < 1318.5 psia 3. < 1318.5psia Uncontrolled Boron Dilution Time available to prevent a complete Calculated minimum operator Indication of the event is in progress
- Mode 1 - manual rod loss of shutdown margin. For the event action time of 30.5 minutes. is an OTDT signal.
control case (UFSAR in Mode 1, the acceptance criterion is Section 15.4.6) > 15 minutes from time of event indication until complete loss of shutdown margin.
Loss of External Electric 1. DNBR > safety analysis limits 1. > limit value LOFTRAN calculated DNBR based Load and/or Turbine Trip 2. RCS pressure < 2748.5 psia 2. < 2748.5 psia on limiting Thimble cell DNBR limit (UFSAR Section 15.2.2 and only.
1.5.2.3) 3. Secondary pressure < 1318.5 psia 3. < 1318.5 psia Accidental Depressurization 1. DNBR> safety analysis limits 1. > limit value LOFTRAN calculated DNBR based of the Reactor Coolant on limiting Thimble cell DNBR limit System (UFSAR Section only.
15.6.1)
Steam System Piping Failure 1. DNBR> safety analysis limits 1. >limit values Analysis acceptance criteria are (UFSAR Section 15.1.5) 2. No fuel melting 2. Maximum kW/ft < value at conservatively based on ANS I
which fuel melting is Condition II acceptance criteria.
predicted.
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3.3 OT8T AND OP8T UNCERTAINTY EVALUATIONS Westinghouse has perfonned instrument uncertainty calculations for the OThT and OP~T reactor trips. To perfonn these calculations, Westinghouse was provided with the following infonnation: RCS process temperatures, rack calibration and drift data, calibration procedures, pressurizer pressure transmitter calibration and drift data, and the effects of steam line break elevated temperature environment on cable insulation resistance.
From this infonnation, Westinghouse evaluated the following parameters: ~T and Tavg variation with burnup; ~T and Tavg variation due to upper plenum anomaly (UPA) transients; loop-to-Ioop temperature asymmetries; process rack drifts for the~T, Tavg, pressurizer pressure, and axial flux difference (AFD) portions of the channels; and pressurizer pressure transmitter drift.
Using the above as well as the secondary-side power calorimetric uncertainty and a maximum mismatch between the protection and control system reference temperatures, the uncertainties were detennined for the OT~T and OP~T trip functions.
The Allowable Value magnitude and approach has been detennined for the OT~ T and OP ~ T reactor trips after an evaluation of the channel calibration procedures and statistical evaluations of process rack calibration data and drift data. A review of the plant calibration procedures notes that the OT~T and OP~T channels are calibrated based on the appropriate input parameters. The uncertainty calculations perfonned for these two functions explicitly reflect these plant procedures. The evaluation of process rack calibration data concluded that loop calibration of the appropriate instrument strings for OT~T and OP~T is acceptable. The evaluation of process rack drift data concluded that the perfonnance of the process racks was within expectations and the recommendations of Westinghouse Technical Bulletin ESBU-TB-97-02, "Analog Process Rack Operability Detennination Criteria." The Allowable Values provided reflect the calibration accuracies of the channel parameter inputs. Thus, the Allowable Values are directly linked to the plant calibration and verification procedures, are consistent with Westinghouse operability recommendations, and reflect parameter magnitudes explicitly modeled in the instrument uncertainty calculations for these two protection functions.
3.4 OT8T and OP8T SETPOINTS OPERATING MARGINS EVALUATION In order to show the effectiveness ofthe revised OT~ T and OP~ T setpoints, a plant data analysis was perfonned. The limiting Condition I transient, a 50% load rejection from 100% power, was analyzed with the revised OT~T and OP~T setpoints.
3.4.1 Plant Data Analysis and Results Since the plant data may not capture the worst temperature fluctuations, Westinghouse typically uses a "hypothetical" spike superimposed on the plant data for the data analysis. A review of the recent plant data showed no significant temperature fluctuations as it may not have captured the worst fluctuations. As such, a "hypothetical" hot leg temperature spike was superimposed on the plant data to result in a conservative bounding temperature fluctuation. This "hypothetical" data was used to demonstrate the effectiveness of the revised setpoints in recovering the steady-state operating margins.
El - 12
By replaying the RTD data through a simulation of the OT.:1T and OP.:1T trip systems, improvement in the steady-state OT.:1 T and OP.:1 T operating margins for the revised setpoints was evaluated.
The minimum margins to OT.:1T and OP.:1T reactor trips with the current setpoints are about 0.1 %
and 3.1 %, respectively, for the "hypothetical" data. With the revised setpoints and time constants, the minimum margins to OT~T and OP.:1T reactor trips have increased to about 13%
and 7.5%, respectively, for the "hypothetical" data. Therefore, with the revised setpoints, the net gains in the OT.:1T and OP.:1T setpoint margins are about 13% and 4.5%, respectively. Note that the margin is defined as the difference between the setpoint and the compensated measured .:1 T expressed in % of full power .:1T.
3.4.2 50% Load Rejection From 100% Power The 50% load rejection transient is the most severe design basis Condition I transient that the plant is expected to sustain without an automatic reactor trip. For both VEGP units, this is equivalent to a 50% turbine load change. The RCS average temperature and the secondary side steam temperature and pressure increase rapidly following this transient. Steam dump to the condenser is required to prevent both reactor trip and steam generator safety valve actuation.
This transient was analyzed using the LOFTRAN computer code (WCAP-7907-P-A, "LOFTRAN Code Description," April 1984).
With the revised setpoints, the minimum OT.:1T and OP.:1T margins to the trip setpoints were 12%
and 8% of full power .:1T, respectively. For current setpoints, the minimum OT.:1T and OP.:1T margins were 3% and 7% of full power .:1T, respectively.
3.4.3 Conclusions The revised OT.:1T and OP.:1T setpoints will provide sufficient steady-state operating margin to the OP.:1 T and OT.:1T trips. If a given RTD measured temperature spike is no worse than the hypothetical data, the expected net gains in the steady-state operating margins are about 13% for OT~T and 4.5% for OP.:1T with the revised setpoints and time constants. These increased operating margins will allow the OT.:1T and OP.:1T turbine runback and alarm setpoints to be returned to 3% below the OT.:1T and OP.:1T trip setpoints.
The revised OT.:1T setpoint will provide significantly higher transient margin to trip setpoint for a 50% load rejection transient (limiting operational transient). The transient margin to OP.:1T trip setpoint has improved slightly.
3.5 MAIN STEAMLINE BREAK OUTSIDE CONTAINMENT 3.5.1 Identification of Causes and Accident Description Steamline ruptures occurring outside the reactor containment structure may result in significant releases of high-energy fluid to the equipment surrounding the steam systems. Superheated steam blowdowns following the steamline break have the potential to raise compartment temperatures outside containment. The impact of the steam releases on this equipment depends on the mass (flow rate) and energy (enthalpy) of the steam which is determined by the plant configuration at the time of the break, the plant response to the break, as well as the size and location of the break.
El - 13
Because of the interrelationship among many of the factors (including the OT~T and OP~T setpoints) which influence steamline break mass and energy releases, an appropriate determination of a single limiting case with respect to mass and energy releases cannot be made.
Therefore, it is necessary to analyze the steam line break event outside containment for a range of conditions.
3.5.2 Input Parameters and Assumptions A spectrum of mass and energy releases was calculated for a range of initial power levels ranging from 102% to 0% power and break sizes ranging and from 1.0 ff to 0.1 ft2.
The break flows and enthalpies of the steam release through the steam line break outside containment are analyzed with the LOFTRAN computer code (WCAP-7907-P-A, "LOFTRAN Code Description," April 1984). Blowdown mass and energy releases determined using LOFTRAN include the effects of core power generation, main and auxiliary feedwater additions, engineered safeguards systems, reactor coolant system thick-metal heat storage, and reverse steam generator heat transfer.
The Vogtle NSSS is analyzed using LOFTRAN to determine the transient steam mass and energy releases outside containment following a steamline break event. The mass and energy releases are used as input conditions to the environmental evaluation of safety-related electrical equipment in the main steam isolation valve compartment. The environmental conditions were evaluated using the GOTHIC code. The GOTHIC code has been benchmarked against the COMPACT code. The previous main steam isolation valve compartment analyses were performed using the COMPACT code as discussed in the VEGP UFSAR.
A series of mass and energy releases was calculated minimizing the secondary side inventory by minimizing the steam generator fluid mass and assuming main feedwater isolation at the time of reactor trip, which leads to earlier steam generator tube bundle uncovery. This maximizes the enthalpy of the steam released through the faulted loop.
A series of mass and energy releases was calculated maximizing the secondary side inventory by maximizing the steam generator fluid mass and assuming continued main feedwater addition with steam flow increase prior to reactor trip with a flow coastdown after trip or receipt of a feedwater isolation signal. These cases maximize the amount of steam released out the break (though steam enthalpy may be reduced) while still allowing for steam generator tube bundle uncovery and the production of superheated steam.
Because AFW flow can have a significant affect on the time of steam generator tube uncovery, several cases were run to assess the effect of AFW flow on the overall analysis results. Cases were run with the following AFW flow assumptions: two AFW pumps operating at minimum flow, three AFW pumps operating at minimum flow, and three AFW pumps operating at maximum flow. The results of these cases are included in the spectrum of cases that were considered.
The reactor trip functions modeled were: low-low steam generator water level, low pressurizer pressure, power range high neutron flux, OT~T, OP~T, and safety injection. The proposed revised OT~T and OP~T setpoints were used in these analyses.
El - 14
3.5.3 Results and Conclusion The mass and energy releases were used to evaluate the qualification of equipment in the main steam valve room area outside containment. The required equipment was determined to be qualified for operation in the calculated environmental conditions with the proposed OT~ T and OP~T setpoints.
CONCLUSION Based on the above, the proposed change can be implemented without adverse impact to the safety analyses and plant systems. Implementation of the revised VEGP OT~T and OP~T reactor trip setpoints will continue to ensure that fuel melt and DNB criteria are met. In addition, the setpoint changes will improve operating margin to the OT~T and OP~T reactor trip setpoints.
El - 15
ENCLOSURE 2 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REACTOR TRIP SYSTEM INSTRUMENTATION OVER TEMPERATURE DELTA TEMPERATURE (OTDT) AND OVERPOWER DELTA TEMPERATURE (OPDT) REACTOR TRIP FUNCTIONS 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUTION PROPOSED CHANGE In accordance with the requirements of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) proposes to revise Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2 Technical Specifications (TS) Figure 2.1.1-1, "Reactor Core Safety Limits"; Table 3.3 .1-1, "Reactor Trip System Instrumentation"; and the associated Bases B2.1.1 and B 3.3.1.
Both VEGP units have experienced steady-state aperiodic hot leg temperature fluctuations. This is not a unique VEGP phenomenon. Similar effects have been noted at other Westinghouse plants. Although no definitive causes for the temperature fluctuations have been identified, they are believed to be caused by the upper plenum flow anomalies. Three resistance temperature detectors (RIDs) are used to account for temperature streaming in order to provide an average temperature in each hot leg. Since the average of the three RTDs is used to represent the hot leg temperature (Thot ), temperature fluctuation from any RID can adversely affect the calculation of the average T hot temperature. This, in turn, impacts the average reactor coolant system (RCS) temperature (Tavg) and RCS differential temperature (~T) and therefore reduces the steady-state operating margins to Overtemperature Delta Temperature (OT~T) and Overpower Delta Temperature (OP~T) trip setpoints. The temperature fluctuations of interest are fluctuations in the increasing direction.
To address the impact of the temperature fluctuations on operating margin, VEGP has implemented a number of changes to improve operating margin. These changes include:
increasing the OT~T and OP~T reference temperature to above the RCS average temperature, reducing the turbine run back setpoint, adding a filter to the measured RCS differential temperature, and adding a filter to the temperature signal to the rod control system to reduce the frequency of spurious rod stepping.
To accommodate the effects of streaming and the associated hot leg temperature fluctuations, SNC proposes to increase the OT~T and OP~T setpoints. This program to increase the setpoints is referred to as the OT~T and OP~T Setpoint Margin Recovery Program (or MRP for short).
The intent of the MRP is to revise the OT~T and OP~T setpoints to increase operating margin.
This is accomplished by increasing the steady-state setpoints and by revising the dynamic compensation time constants in the setpoint equations. The setpoint allowable values and core safety limits were also revised to support the MRP.
The analyses supporting the MRP assume a revision to the Relaxed Axial Offset Control (RAOC) band and the inclusion of the limit or clamp on the compensated temperature difference term in the OT~T trip setpoint. The revision to the RAOC band and the limit or clamp on the compensated temperature difference term in the OT~T trip setpoint are currently under review by the Staff. These revisions were submitted for review in a separate amendment request (SNC letter E2-1
LCV-1563 dated October 30, 2001). Implementation of the MRP setpoint changes is contingent upon approval of the amendment request for the revised RAOC band and clamp.
Pursuant to 10 CFR 50.92, Southern Nuclear Operating Company (SNC) has reviewed the proposed change to determine if a significant hazards consideration is involved. The proposed change, as defined below, has been reviewed and deemed not to involve any significant hazards considerations as defined in 10 CFR 50.92. The basis for this determination follows.
EVALUATION
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change can be implemented without adverse impact to the safety analyses and plant systems. Implementation of the revised VEGP OTf). T and OPf). T reactor trip setpoints will continue to ensure that fuel melt and departure from nucleate boiling (DNB) criteria are met. In addition, the setpoint changes will improve operating margin to the OTf). T and OPf). T reactor trip setpoints. The setpoints provide reactor protection and are not event initiators and therefore do not affect the probability of occurrence of an accident previously evaluated.
There is no change in the radiological consequences of any accident since the fuel clad, the reactor coolant system pressure boundary, and the containment are not changed, nor will the integrity of these physical barriers be challenged. In addition, the proposed change will not change, degrade, or prevent any reactor trip system actuations.
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?
The proposed change can be implemented without adverse impact to the safety analyses and plant systems. Implementation of the revised VEGP OTf).T and OPf).T reactor trip setpoints will continue to ensure that fuel melt and departure from nucleate boiling (ONB) criteria are met. In addition, the setpoint changes will improve operating margin to the OTf).T and OPf).T reactor trip setpoints. The revised OTf).T and OPf).T reactor trip setpoints would not create any new transients nor would they invalidate the OTf).T and OPf).T design bases.
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 the margin of safety?
The proposed change can be implemented without adverse impact to the safety analyses and plant systems. Implementation of the revised VEGP OTf).T and OPf).T reactor trip setpoints will continue to ensure that fuel melt and departure from nucleate boiling (ONB) criteria are met. In addition, the setpoint changes will improve operating margin to the OTf).T and OPf).T reactor trip setpoints. The margin of safety provided by the Technical Specifications is not significantly affected because the proposed changes are based on the same accident acceptance limits, i.e., the OTf).T and OPf).T design bases continue to be met.
E2-2
Therefore, the proposed change does not involve a significant reduction in the margin of safety.
CONCLUSION Based on the preceding evaluation, Southern Nuclear has determined that the proposed change meets the requirements of 10 CFR SO.92(c) and does not involve a significant hazards consideration.
ENVIRONMENTAL EVALUATION Southern Nuclear has evaluated the proposed changes and determined they do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Sl.22(c X9). Therefore, pursuant to 10 CFR Sl.22(b), an environmental assessment of the proposed changes is not required.
E2-3
ENCLOSURE 3 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REACTOR TRIP SYSTEM INSTRUMENTATION OVER TEMPERATURE DELTA TEMPERA TURE (OTDT) AND OVERPOWER DELTA TEMPERATURE (OPDT) REACTOR TRIP FUNCTIONS MARKED-UP TECHNICAL SPECIFICATION AND BASES PAGES Marked-up Note Pages Figure 2.1.1-1 1 Page 3.3.1-20 2 Page 3.3.1-21 2 Figure B2.1.1-1 1 Page B3.3.1-19 2 Page B3.3.l-21 1 Page B3.3.1-65 2 NOTES
- 1. Mark-up of current Technical Specification or Bases pages.
- 2. Mark-up of Technical Specification or Bases pages submitted for review in a separate amendment request (SNC letter LCV-1563 dated October 30,2001). This amendment request is currently under review by the Staff.
SLs 2.0 .
Figure 2.1.1-1 Reactor Core Safety limits VogUe Units 1 and 2 2.0-2 Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
/NSEf<r A 670
to--
-1 r--- r-.:..r--.-,.
1 I T DO NOT OPERATE IN THIS AREA 650
--- ~PIiII ~
r--
~
r---..
r=:::: r---.. r"-.
~
0-1MPIiII .
r-- r--.
r--.. r--...
r-..... --
-....... r--..
~
r\.
r--.. ~
11i2O PIiII r-.......... ~ """-
r--..
t---. r--- -
r---... r--.-. r--..
~
~
"" ~
ACCEPTABlE OPERATION 560 550 o 10 20 30 40 50 eo 70- 80 80 100 110 120 PERCENT OF RATED THERMAL POWER Agure 2.1.1-1 Reactor Core Safety Limits
RTS Instrumentation 3.3.1 r--- INSlP:r i Table 3.3.1-1 (page 7 of 8)
Reactor Trip System Instnmentation Note 1: Overtemoerature OeJta-T The Otertemperetllfe gella T I=IIAaieA Alle'lla~le Vallie shall Aet. exeeell the NeFfliAsl TRp S.slAt lIefi.. eEl tJ)' the felleni"l1 Where: AT measured loop specific RCS differential temperature, degrees F ATo indicated loop specific RCS differential at RTP, degrees F ll.IJl lead-lag compensator on measured differential temperature 1+'1" ,c- -0
'fl, 'f2 time constants utilized in lead-lag compensator for differential temperature: 'f'~
'f2JO)seconds '
~=-O
-L 1+t,S . lag compensator on measured differential temperature. ". -- L .
.......~g time constant utilized in lag compensator for differential temperature, ~aeconda .
~ ~
'fl t::I K, fundamental setpoint, ~ RTP. '\: 1/'1-*1 .
Kz modifier for temperature,
- 2.2.." RTP per degreeF 1'!:l4 .
1+'1" lead-lag compensator on dynamic temperature compensation
't4, '(I time constants utilized in lead-lag compensator for temperature compensation: 't4 ~ 28 seconds.,
'f. S .. seconds T measured loop specific RCS average temperature, degrees F
-L 1+teS lag compensator on measured ave.rage temperatUre y ~ 6
'tI time constant utilized In lag compensator for average temperature,~
T' Indicated loop specific RCS average temperature at RTP, S 588.4 degrees F Kz* modifier for pressure,~ ~ per psi; . =0./'17 P measured Res pressurizer pressure, psi; po reference pressure, ~ 2235 paig s Laplace transform variable, inverse seConds f , (AFD) modifier for Axial Flux Difference (AFe):
- 1. for AFD between -23% and +10%," 0% RTP
- 2. for each % AFD is below -23%, the trip setpoint shall be reduced by 3.3% RTP .
- 3. for each % AFD is above +10%, the trip setpoint shan be reduced by 1.95% RTP Vogtle Units 1 and 2 3.3.1-20 Amendment No. (Unit 1)
Amendment No" (Unit 2)
Insert 1 The Allowable Value of each input to the Overtemperature Delta-T Function as defined by the equation below shall not exceed its as-left value by more
-than the following:
(1) 0.5 % AT span for-the AT channel (2) 0.5 0" AT span for the T ewt channel (3) 0.5 % AT span for the pressurizer pressure channel (4) 0.5 % AT span for the f1(AFD) channel _
RTS Instrumentation 3.3.1 Iable 3.3.1-1 (page 8 of 8)
Reactor Trip System Instnmentation NOte 1; Oyertemperatur'e petta-I (conthIed)
(0) The compensated temperature difference
{I
{I
+ 't .. I}
+ 'CS 1}
[1 T--- *
{l +ql}
r ] shall be no more negative than S degrees F.
Note 2: Oyerpower [)ella-I 1fte 9'.e",8'0'I8r g81&8 T FWAG'bn AJ.L,OItJ.AiLoi '.J.~1i 8~all "lilt IIMII8II" 1111 tJIIMiMI Trip &11"111", "ellAlI"~' It..... """ .~...."
~ AlOJe ~.A a 8~[100- gf ~ {1 + Til} . 1 ] [
S K4- Ks
[ (1' Z I) :1 ]..[' 1 I - T
]
-r - f2(AFD)
.]
LAIO' (1 + 1'2 I} (1+f3I). (1 +T I) {1+1', I} , (1 + TI ) .
z Where: AI measured loop specific RCS differenIIaI temperat\.n, degrees F Alo Indicated loop specific ReS differential at~, degrees F l:I::W lead-lag compensator on measured dlfferantial temperature. .
1...r.e . . ' . =0 time constants utilized In lead-lag compensator for dHferential temperat\.n:
'r:a.~
"'12T"""'1da. . .
"t,,'tt
'rt~ .
~~~.;o
..L 1...r.e lag compensator on m~ ~Irt"..,ual temperature Y ~ 6
'fa time constant utilized In lag compensator for differential temperature,~seconda . .
~ fLRiamental seIpOInt,~ ~ -- ~ / /0 '.
Ke modifier for temperature change: :it 2% RIP per degree F for increasing temperature, tt 0% R1P per degree F for decreasing temperature .
..lzL 1+'t18 rate-lag compensator on dynamic temperat\.n compensation
'tz time constant utilized In iate-lag compensator for temperabn ccmpensatlon,.tt 10 aecondI I measured loop specific RCS average temperat\.n, degrees F
--L.
1+'ttS lag compensator on measured a~ge temperature' .' ~ .~ {,.
'fI time constant utilized In I8g compensator for average temperature, ~seconda
~--------------~~------------
Ke modifier for temperature: ~ RIP per degree F for I > T",
- 0% RIP tor I $ r.
,.. Indicated loop specific RCS average temperatln at RTP, $ 588.4 degr8ea F s Laplace transfonn variable, Inverse seconds fz(AFD) modifier for AxIal Flux DHference (AFD),
- 0% RIP for an AFD Vogtle Units 1 and 2 . 3.3.1-21 Amendment No. (Unit 1)
Amendment No. (Unit 2):
Inserl2 The Allowable Value of each input to the Overpower Delta-T Fundion as defined by the equation below shall not exceed its as-left value by more than the following:
(1) 0.5 %.bT span for the AT channel (2) 0.5 % !lT span for the Tng chamel
Reactor Core SLs B 2.1.1 Figure B 2.1.1-1 (page 1 of 1)
REACTOR CORE SAFETY LIMITS VS. BOUNDABV OF PROTECTION Vogtle Units 1 and 2 B 2.1.1-7 Revision No. 0
~~-------'----~~~A=~~~~~~~~~=E=AA~~~~~IN~~~~~~==~*r--=~~~~~~A~T~~i.:.~~~~~I~i' ABOVE AND TO ~E RIGHT OF LINES (2235 PIlI!) . -. 'Ii I
650 -I-..:::...-.......;;;---r-CORE SAFETY UMIT (2235 PIIg) -+-------+....:.-.------Hk-Il
&W~.~~~-+------_F~~--1-----~r_--~--r_1 i
630 IL
~~1-------~------~----~~------_r--~~_1--~r_~
r t-o
~ 610 6OOL-------h-----~--~~~------_t------~~~1r_t 5~l-------~--~~~------_1--~----t_------_r----~~
~L---~~~----~--~---+------_t------_t----~ri ThII Ilgure lor ~ cn,.
Do not use lor apIOlIan.
~o~-------L------~--------~------~--------~----~~
o "20 40 60 80 100 120 Percent 01 R.ted Therrnel Power ('JIo)
Figure B 2.1.1-1 (page 1 of 1)
REACTOR CORE SAFETY LIMITS VS. BOUNDARY OF PR9TECTION
RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES,
- 6. . Overtemperalure I1T (CQrrtiooe<l) C 1dS1{t"3 .
LeO, and' as close as possible to 588.4° F. 11=18 'Ja1W8 9t T' fer \l:te APPLlCABLlTY ,F9FRsiRiR6 RCS lee~s will l3e set eJl~r9~riately less tNt..
S88.4QF I3sseEi eA the eett:lelleeJ' 9J'eeifie iAElieeted TiUi.
In the case of decreasing temperature, the ,compensated temperature difference shall be no more negative than'3 OF to limit 1he increase in the setpoint during cooldown transientS.
The engineering scaling calculations use each of the ,
referenced parameters as an exact gain or reference value. '
Tolerances are not ,applied to the individual gain or reference parameters. Tolerances are applied to each calibration. ",
module and the overall string calibratior:l.ln order to ensure that the Overtemperature 6T eetJ'si'" is'consistent with the assumptions of the safety analyses, It is necessary to verify during the CHANNEL OPERATIONAL TEST that the , '
-~>~ OveFlefftJ:)ers1tJFe ATeetJ'eiAt Ie within the appropriate" '
calibration tolerances for the defined calibration conditions (Ref. 9);
The Leo requires all four channels of the Overtemperature ~T
'trip Function to be OPERABLE. Note that the Overtemperahn
~T Function receives input from channels shared with other RTS '
Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affect~ Functions. '
,In MODE 1 or 2, the Overtemperature ~T trip m'ust be OPERABLE to prevent DNB. In MOD~ 3, 4, 5,' or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DNB. '
(continued) , ,
vogt\e UnHs 1 and 2 B 3.3.1-19
Insert 3 The instrument uncertainty calculations and safety analyses, in combination, have accounted for loop variation in loop specific, full power, indicated AT and T"II' Withresped to Thi' a value for r common to all four loops is permis~ible within the limits identified in the uncertainty calculations. Outside of those limits, the value of r will be set appropriately to reflect indicated, loop specific, full power values.
Insert 4 instrument channel is performing in 8 manner Insert 5 magnitude of instrument drift from the as left condition is within limits, and ~
the input parameters to the trip function are
RTS Instrumentation B 3.3.1 BASES APPLICABLE 7. Overpower AT (continued)'
SAFETY ANALYSES.
LCO. and Delta-To. as used in the overtemperature and overpower APPLICABILITY IIT trips. represents the 100% RTP value as measured 'for each loop. This normalizes each loop's AT trips to the actual operating conditions eXisting at the time of measurement. thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in RCS loop IIT can be due to several factors. e.g ** difference in RCS loop flows and slightly asymmetric power distributions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values.
Therefore, loop specific ATo values are measured as needed to ensure they represent actual core conditions.
The value for r is Ii key reference parameter corresponding directly to plant safety analyses initial conditions assumptions for the Overpower AT function. For the purposes of performing a CHANNEL CALIBRATION, the values for~, Ke. Ke, and ,. are utilized in the safety analyses without explicit tolerances. but .
should be considered as nominal values for instrument settings.
That is, while an exact setting is not expected, a setting as dose as reasonably possible is desired. Note that for ,.. the value for INSI!"IJ:r b the hottest RCS loop will be set as dose as possible to 588.4 0 F ."¥The vslHe af r fer the reMahti". RQS lee,e will
'be eet eJ!lJ!lreJ!lrietely Ieee the" 588.4 9 1= l!taeel!i eft 'he settlel .
leeJ!l eJ!leeifie i,uUeeteEf ThO, The engineering scaling. .
calculations use each of the referenced parameters as an exact gain 'or reference value. Tolerances are not applied to the individual gain or reference parameters. Tolerances are applied to each calibration module anc;l the overall string calibration. *In order to ensure that the Overpower AT
--.....;>~ eetJ!lei"t ie' consistent with the' assumptions of the safety
. analyses. it is necessary to verify during the CHANNEL .
OPERATIONAL TEST that the .()verfJewerAT eet,ei,,' il within the appropriate calibration tolerances for defined .
calibration conditions (Ref. 9). Note that for *the parameter K I *
(continued)
VogUe Units 1 and 2 B 3.3.1-21 Rev. 1-6198
- Insert 6 The instrument uncertainty calculations and safety analyses, in combination, have accounted for loop variation in loop specific, full power, indicated AT and T ftIIo With respect to T"I' a value for-r' common-*to-all four-loops is pennissibte within the limits identified in the uncertainty calculations. Outside of those limits, the value of T-' will be set appropriately to retied indicated, loop specific, full power values.
RTS Instrumentation B 3.3.1 BASES REFERENCES 2. FSAR, Chapter 6~
(continued)
- 3. FSAR, Chapter 15.
- 4. IEEE-279-1971.
- 5. 10 CFR 50.49.
- 6. WCAP-11269, Westinghouse Setpoint Methodology for
. Protection Systems; as supplement~ by: ..
e Amendments 34 (Unit 1) and 14 (Unit 2), RTS Steam .
Generator Water Level - Low Low, ESFAS Turbine Trip .
and i=eedwater Isolation SG Water Level - High High, and ESFAS AFW SG Water Level- Low LOw. .
e Amendme'nts 48' and 49 (Unit 1) and Arilendmen1S 2i..
and 28 (Unit 2), deletion of RTS Po\yer Range Neutron Aux High Negative Rate Trip. '. .
- e Amendments 60 (Unit 1) and 39 (Unit 2), RTS
. Overternperature ~T setpoint revision.
e Amendments 57 (Unit 1) and 36 (Unit 2), RTS.
- Overtemperature and Overpower~T time constants and Overtemperature ~T setpolnt.* .
e Amendments 43 and 44 (Unit 1) and 23 and 24. (Unit 2),
revised Overtemperature and Overpower ~T trip setpolnts and allowable values.
.e Amendments 104 (Unit 1) and 82 (Unit 2), revised RTS Intermediate Range Ne~on. Rwe, Source R : . .
Neutron Rux,* a.nd P-6 trap setpoints and allo .
values.' ' . .
e Amendments (Unit 1) anci (Unit 2). reVised Overtemperature ~T trip setpolnt to limit value of the compensated temperature difference an~ revised the ItJS~r'7
. ~odifi.er for axial flux d~erence. I
- 7. WCAP-1 0271-P-A. Supplement 1, May 1986.
- 8. FSAR. Chapter 16.
- 9. Westinghouse Letter GP-16696. November 5. 1997..'
- 10. WCAP-13632-P-A Revision 2. -Elimination of Periodic .
Sensor Response Time Testing Requirements,* January 1996. .
(continued)
VogUe Units 1 and 2 B 3.3.1-65
INSERT 7
- Amendments (Unit 1) and (Unit 2), revised Overtemperature ll.T and Overpower ll.T trip setpoints to increase the fundamental setpoints Kl and ~, and to modifiy coefficients and dynamic compensation terms.
ENCLOSURE 4 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REACTOR TRIP SYSTEM INSTRUMENTATION OVER TEMPERATURE DELTA TEMPERA TURE (OTDT) AND OVERPOWER DELTA TEMPERATURE (OPDT) REACTOR TRIP FUNCTIONS TYPED REVISED TECHNICAL SPECIFICATION AND BASES PAGES
SLs 2.0 670
--.. ......... I I I
- --- 2U5L.
660 DO NOT OPERATE IN THIS AREA
--. -..... -""""'-'r----.. ~
650
~35P81g -- .............. r---....
E w
a: 640 r----.. ...... .............
r--....... ..............
~ 630
..--.... r---- r---..... -..... r---.... ~
w
-~r-=:::~1885"'"
A.
- E I!! 620
~ ............... ~
~ '\
~ r---....
1920.,.111 w
............. r---.... .......
IA.
§ 610 600
--..... ~ ~ ..............
~
~
tiw 590
- c I 580 I ACCEPTABLE 570, . - OPERATION 560 550 o 10 20 30 40 50 60 70 60 90 100 110 120 PERCENT OF RATED THERMAL POWER Figure 2.1.1-1 Reactor Core Safety Limits VogtJe Units 1 and 2 2.0-2 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 9)
Reactor Trip System Instrunentatlon APPUCABLE MODES OR NOMINAL OTHER SPECIFIED REQUIRED SURVSUANCE ALLo\yABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VAWE SETPOINr<n)
- 2. Power Range Neutron Flux
- b. Low 4 E SR 3.3.1.1 S 27.3% RTP. 25% RTP SR3.3.1.8 SR 3.3.1.11 SR3.3.1.15
- 3. Power Range 1,2 4 E SR3.3.1.7 S 6.3% RTP 5%RTP Neutron Flux H~ SR 3.3.1.11 with time with time Positive Rate constant constant
~2seC. ~2sec
- 4. Intennedlate 1(b), 2(c) 2 F,G SR3.3.1.1 S41.9%RTP 25%RTP Range Neutron SR3.3.1.8 Flux SR3.3.1.11 2<d) 2 H SR 3.3.1.1 S41.9% RTP 25%RTP SR3.3.1.8 SR 3.3.1.11 (continued)
(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.
(b) Below the P-10 (Power Range Neutron Flux) Interlocks.
(c) Above the P-6 (Intennedlate Range Neutron Flux) Intel1ocks.
(d) Below the P-6 (Intennedlate Range Neutron Flux)lnterloclcs.
(n) A channel Is OPERABLE with an actual Trip Setpoint value outside Its calibration tolerance band provided the Trip Setpoint value Is conservative with respect to Its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoinl A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary In response to plant conditions.
Vogtle Units 1 and 2 3.3.1-14 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1*1 (page 2 of 9)
Reactor Trip System Instn.mentation APPUCABLE MODES OR NOMINAL OTHER TRIP SPECIRED REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REOUIREMENTS VAlUE SETPOI~n)
- 5. Source Range 2 I,J SR 3.3.1.1 S1.7ES 1.0ES Neutron Rux SR3.3.1.8 cps cps SR 3.3.1.11 2 J,K SR3.3.1.1 S 1.7ES 1.0ES SR3.3.1.7 cps cps SR 3.3.1.11 L SR3.3.1.1 SR 3.3.1.11 NA NA
- 6. Overtemperaua.:\T 1,2 4 E SA 3.3.1.1 Refer to Note 1 Refer to Note 1 SR 3.3.1.3 (Page 3.3.1-20) (Page 3.3.1-20)
SR3.3.1.6 SR3.3.1.7 SR 3.3.1.10 SR3.3.1.15
- 7. Overpower .:\T 1,2 4 E SR 3.3.1.1 Refer to Note 2 Refer to Note 2 SR3.3.1.7 (Page 3.3.1-21) (Page 3.3.1-21)
SR3.3.1.10 SR3.3.1.15 (continued)
(a) With RTBs closed and Rod Control System capable of rod withdrawal.
(d) Below the P-6 (Intennediate Range Neutron Rux) Intertocks.
(e) WIth the RTBs open. In this condition, source range Function does not provide reactor trip but does provide Input to the High Rux at Shutdown Alann System (LCO 3.3.8) and indication.
(n) A channel Is OPERABLE with an actual Trip Setpolnt value outside its calibration tolerance band provided the Trip SetpoInt value Is conservative with respect to its associated Allowable Value and the channel Is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary In response to plant conditions. .
Vogtle Units 1 and 2 3.3.1-15 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 30f 9)
Reactor Trip System Instrunentation APPUCABLE MODES OR NOMINAL OTHER TRIP SPECIAED REQUIRED SURVElUANCE AUO'NABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT<n)
- 8. PresslXlzer Pressure
- 9. Pressurizer Water 3 M SR 3.3.1.1 S93.9% .
leYeI- HIgl SR3.3.1.7 SR3.3.1.10
- 10. Reactor Coolant FIow- Low
- a. Single Loop 1(h) 3 per loop N SR 3.3.1.1 ~89.4""
SR3.3.1.7 SR3.3.1.10 SR3.3.1.15
- b. Two Loops 3 per loop M SR3.3.1.1 ~89.4""
SR3.3.1.7 SR3.3.1.10 SR3.3.1.15 (continued)
(f) Above the P-7 (Low Power Reactor Trips Block) Interlock.
(g) TIme constants utilized In the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead 8nd 1 second for lag.
(h) Above the P-8 (Power Range Neutron Aux) Interlock.
(0 Above the P-7 (Low Power Reactor Trips Block) Interlock and below the P-8 (Power Range Neutron Aux) Interlock.
(n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpolnt value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary In response to plant conditions.
Vogtle Units 1 and 2 3.3.1-16 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 9)
Reactor Trip System Instnmentatlon APPUCABLE MODES OR OTHER NOMINAL SPECIRED REQUIRED SURVElu.ANCE TRIP AU-OWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT<n)
- 11. Undervoltage 1(f) 2 per bus M SR 3.3.1.9 ~9481 V 9600 V RCPs SR 3.3.1.10 SR 3.3.1.15
- 12. Underfrequency 1 (f) 2 per bus M SR 3.3.1.9 ~57.1 Hz 57.3 Hz RCPs SR 3.3.1.10 SR 3.3.1.15
- 13. Steam Generator 1,2 4perSG E SR 3.3.1.1 ~35.9% 37.8%
(SG) Water Level - SR3.3.1.7 Low Low SR 3.3.1.10 SR3.3.1.15 (continued)
(f) Above the P-7 (Low Power Reactor Trips Block) Interlock.
(n) A channel is OPERABLE with an actual Trip Setpoint value outside Its calibration tolerance band provided the Trip Setpoint value Is conservative with respect to Its associated Allowable Value and the channel Is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary In response to plant conditions.
VogtJe Units 1 and 2 3.3.1-17 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 9)
Aeactor Trip System Instnmentation APPUCABLE MODESOA NOMINAL OTHEA TRIP SPECIFIED REOUIRED SUAVElu.ANCE AU.OYVABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT<n)
- 14. Turbine Trip
- 15. Safety Injection (51) 1,2 2 trains Q SA 3.3.1.13 NA NA Input from EngIneered Safety FeatL1'8 Actuation System (ESFAS)
- 16. Reactor Trip System Interlocks
P-6
- b. Low Power 1 1 per train 5 SA 3.3.1.5 NA NA Reactor Trips Block, P-7
- e. Power Range 1,2 4 A SA 3.3.1.11 (I,m) (I,m)
Neutron Flux. SA 3.3.1.12 P-10 and Input toP-7
- f. Turbine 2 5 SA 3.3.1.10 S 12.3% Impulse 10% Impulse Impulse SA 3.3.1.12 PressL1'8 Presslnl Pressure, Equivalent Equivalent P-13 turbine turbine (continued)
(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.
- 0) Above the P-9 (Power Aange Neutron Aux) Interlock.
(Q For the P-10 input to P-7, the Allowable Value is S 12.3% RTP and the Nominal Trip Setpolnt is 10% RTP.
(m) For the Power Range Neutron Aux, P-10, the Allowable Value is ~ 7.7"/0 ATP and the Nominal Trip Setpoint is 10% ATP.
(n) A channel is OPERABLE with an actual Trip Setpolnt value outside its calibration tolerance band provided the Trip Setpolnt value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.
Vogtle Units 1 and 2 3.3.1-18 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 9)
Reactor Trip System Instnmentation APPUCABLE MODES OR NOMINAL OTHER SPECIRED REQUIRED SURVEILLANCE AU.OWABLE TRIP VALUE SETPOINr<n)
FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS
- 17. Reactor Trip 1,2 2 trains T,V SR3.3.1.4 NA NA Breakers(k) 3(a), 4(a), 5(a) 2 trains C SR3.3.1.4 ,NA NA
- 18. Reactor Trip 1,2 1 each per U,V SR3.3.1.4 NA NA Breaker RTB Undervoltage and Shunt Trip '3(a),'4(a),5(a) 1 each per C SR3.3.1.4 NA NA Mechanisms RTB
- 19. Automatic Trip 1,2 2 trains a,v SR3.3.1.5 NA NA logic 3(~), 4(a), 5(a) 2 trains C SR3.3.1.5 NA NA (a) With RTBs closed and Rod Control System capable of rod withdrawal.
(k) Including any reactor trip bypass breakers that are rackeclln and closed for bypassing an AlB.
(n) A channel Is OPERABLE with an actual Trip Setpolnt value outside Its calibration tolerance band provided the Trip Setpolnt value Is conservative with respect to Its associated Allowable Value and the channel Is readjusted to within the established calibration tolerance band of the Nominal Trip Setpolnt. A Trip Setpolnt may be set more conservative than the Nominal Trip Setpolnt as necessary In response to plant conditions.
Vogtle Units 1 and 2 3.3.1-19 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 9)
Reactor Trip System Instnmentation Note 1: Overtemoerature DeIta-T The Allowable Value of each input to the Overtemperature Delta-T function as defined by the equation below shall not exceed Its as-left value by more than the following:
(1) 0.5%.1T span for the.1T channel (2) 0.5% .1T span for the T8WV channel (3) 0.5% .1T span for the pressurizer pressure channel (4) 0.5% .1T span for the f, (AFD) chamel Where: .1T measured loop specifIC RCS differential temperature, degrees F
.1To indicated loop specifIC RCS differential at RTP, degrees F 1.+/-hI lead-lag compensator on measured differential temperature 1+'t2S t" t2 time constants utilized In lead-lag compensator for differential temperature: 'tl = 0 seconds,
. 't2 = 0 seconds
_1_
1+'t3S lag compensator on measured differential temperatln
.t3 time constant utilized in lag compensator for differential temperature, :S 6 seconds K, fundamental setpolnt, :S 114.90/0 RTP
~ modifier for temperature, =2.24% RTP per degree F l+/-W 1+'t5S lead-lag compensator on dynamic temperature compensation
't4,'t6 time constants utilized In lead-lag compensator for temperature compensation: 't4 ~ 28 seconds,
't6 :S 4 seconds T measured loop specifIC RCS average temperature, degrees F
-L 1+tsS lag compensator on measured average temperature
't6 time constant utilized In lag compensator for average temperature, S 6 seconds T' indicated loop specific RCS average temperature at RTP, S 588.4 degrees F K:, -modifier for pressure, = 0.177% RTP per pslg P measured RCS pressurizer pressure, psig P' reference pressure, ~ 2235 psig s Laplace transform variable, Inverse seconds Vogtle Units 1 and 2 3.3.1-20 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 9)
Reactor Trip System Instrunentation Note 1: Overtemoerature Delta-T (continued) f,(AFD) modifier for Axial Rux Difference (AFD):
- 1. for AFD between -23% and +10%, = 0% RTP
- 2. for each % AFD is below -23%, the trip setPoint shall be redUced by 3.3% RTP
- 3. for each % AFD is above +10%, the trip setpoint shall be reduced by 1.95% RTP (0) The compensated temperature difference
{I + 4 [1 1 S}
T - 1"'
] shall be no more negative than 3 degrees F.
{I + 15s} {I + 't(is}
Note 2: Overpower Delta-T The Allowable Value of each input to the Overpower Delta-T Function as defined by the equation below shall not exceed its as-left value by more than the following:
J-+11 : .J -+'2 IAFDJ ]
(1) 0.5% ~T span for the AT channel (2) 0.5% AT span for the TIVO channel
[100 ::. 1;': ~~}{1' ~ J[ <4-[" 11 ~'::S}{l':. oj T r Where: AT measured loop specific RCS differential temperature, degrees F
~To indicated loop specific RCS differential at RTP, degrees F
~ lead-lag compensator on measured differential temperature 1+'t2S 1" 't2 time constants utilized In lead-lag compensator for differential temperature: 1:, =0 seconds,
't2 = 0 seconds
_1_
1+'t3S lag compensator on measured differential temperature
't3 time constant utilized In lag compensator for differential temperature, S 6 seconds I
K. fundamental setpolnt, S 110% RTP I
~ modifier for temperature change: ~ 2% RTP per degree F for increasing temperature, ~ 0% RTP per degree F for decreasing temperature
..l1L 1+trS rate-lag compensator on dynamic temperature compensation t7 time constant utilized in rate-lag compensator for temperature compensation, ~ 10 seconds T measured loop specific ReS average temperature, degrees F
-L-1+'t66 lag compensator on measured average temperature Vogtle Units 1 and 2 3.3.1-21 Amendment No. (Unit 1)
Amendment No. (Unit 2)
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 9 of 9)
Reactor Trip System Instnmentation Note 2: Overtemoerature Delta-T (continued) time constant utilized in lag compensator for average temperature, ~ 6 seconds modifier for temperature: ~ 0.244% RTP per degree F for T > T"', = 0% RTP for T s T'"
TN indicated loop specific ReS average temperature at RTP, S 588.4 degrees F s Laplace transform variable, inverse seconds modifier for AxIal Flux Difference (AFD), = 0% RTP for all AFD Vogtle Units 1 and 2 3.3.1-22 Amendment No. (Unit 1)
Amendment No. (Unit 2)
Reactor Core SLs B 2.1.1 6ror---------r---------~------~--------_r--------_nr_--==~,
UNACCEPTABLE OPERATION IN THIS AREA OP DELTA T ~_2.1~Z~
ABOVE AND TO THE RIGHT Of LINES (2235 psIg) --. Iii m
~~~
650-I--.lI........:::::::---t-CORE SAFETY ILIMIT (2235 psig) - + - _ - - + - _ . . . . . . . . . - . -.....~~~ .
~O-l-~~~--t---------r-~----~---------+---------r~~----~
i 6~-I---------~--~~~~------~------~~~-------r--~~~~
E6~-I-----------~---------~------~~-----------+---~~~---~~~
r t-o
~ 610~--------r-------~~------~----~~~~-------+----~~~
~
~-I---------r-------~~--~ __~---------+---------+~~~~~
ACCEPTABLE OPERATION 5~-I---------~----~~~------~---------+---------+-----"--~~
~~----~~r-------~~---"-----"-1---"---"-----"-+-------"---+---"---"---"~~
ThIs figure for lustraUon triy.
OonotuseforopemJon.
~O~.---"---"--~~---"---"--~---"---"---"---"~---"---"---"---+---"---"---"---"-+---"---"---"-&-I o 20 40 eo 80 100 120 Percent of Rated Thennal Power (%)
Figure B 2.1.1-1 (page 1 of 1)
REACTOR CORE SAFETY LIMITS VS. BOUNDARY OF PROTECTION Vogtle Units 1 and 2 B 2.1.1-7
RTS Instrumentation 83.3.1 BASES APPLICABLE 6. Overtemperature ~T (continued)
SAFETY ANALYSES, LCO, and as close as possible to 588.4 of. The instrument uncertainty APPLlCABLlTY calculations and safety analyses, in combination, have accounted for loop variation in lQ9p specific, full power, indicated ~T and T avg. With respect to T avg, a value for T' common to all four loops is permissible within the limits identified in the uncertainty calculations. Outside of those limits, the value of T' will be set appropriately to reflect indicated, loop specific, full power values. In the case of decreasing temperature, the compensated temperature difference shall be no more negative than 3 OF to limit the increase in the setpoint during cooldown transients. The engineering scaling calculations use each of the referenced parameters as an exact gain or reference value. Tolerances are not applied to the individual gain or reference parameters.
Tolerances are applied to each calibration module and the overall string calibration. In order to ensure that the Overtemperature ~T instrument channel is performing in a manner consistent with the assumptions of the safety analyses, it is necessary to verify during the CHANNEL .
OPERATIONAL TEST that the magnitude of instrument drift from the as-left condition is within limits, and that the input parameters to the trip function are within the appropriate calibration tolerances for the defined calibration conditions (Ref. 9).
The LCO requires all four channels of the Overtemperature ~T trip Function to be OPERABLE. Note that the Overtemperature
.~T Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.
In MODE 1 or 2, the Overtemperature ~T trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DN8.
(continued)
VogtJe Units 1 and 2 83.3.1-19
RTS Instrumentation B3.3.1 BASES APPLICABLE 7. Overpower ~T (continued)
SAFETY ANALYSES,
-LCO, and Delta-To, as used in theovertemperature and overpower APPLICABILITY ~T trips, represents the 100% RTP value as measured for each loop. This normalizes each loop's ~T trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in RCS loop ~T can be due to several factors, e.g., difference in RCS loop flows and slightly asymmetric power distributions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific ~T values.
Therefore, loop specific ~To values are measured as needed to ensure they represent actual core conditions.
The value for'" is a key referencQ parameter corresponding directly to plant safety analyses initial conditions assumptions for the Overpower ~T function. For th~ purposes of performing a CHANNEL CALIBRATION, the values for 1<., K&, Ke, and,.. are utilized in the safety analyses without explicit tolerances, but should be considered as nominal values for instrument settings.
That is, while an exact setting is not expected, a setting as close as reasonably possible is desired. Note that for"', the value for the hottest RCS loop will be set as close as possible to 588.4 OF. The instrument uncertainty calculations and safety analyses, in combination, have accounted for loop variation in loop specific, full power, indicated ~T and T.v!J' With respect to T avg, a value for TN common to all four loops is permiSSible within the limits identified in the uncertainty calculations.
Outside of those limits, the value of TN will be set appropriately.to reflect indicated, loop specific, full power values. The engineering scaling calculations use each of the referenced parameters as an exact gain or reference value.
Tolerances are not applied to the individual gain or reference parameters. Tolerances are applied to each calibration module and the overall string calibration. In order to ensure that the Overpower AT instrument channel is performing in a manner consistent with the assumptions of the safety analyses, it is necessary to verify during the CHANNEL OPERATIONAL TEST that the magnitude of instrument drift from the as-left condition is within limits, and that the input parameters to the trip function are within the appropriate calibration tolerances for defined calibration conditions (Ref. 9). Note that for the parameter K 5, (continued)
VogUe Units 1 and 2 B 3.3.1-21
RTS Instrumentation 83.3.1 BASES REFERENCES 2. FSAR, Chapter 6.
(continued)
- 3. FSAR, Chapter 15.
- 4. IEEE-279-1971.
- 5. 10 CFR 50.49.
- 6. WCAP-11269, Westinghouse Setpoint Methodology for Protection Systems; as supplemented by:
- Amendments 34 (Unit 1) and 14 (Unit 2), RTS Steam Generator Water Level - Low Low, ESFAS Turbine Trip and Feedwater Isolation SG Water Level- High High, and ESFAS AFW SG Water Level- Low Low.
- Amendments 48 and 49 (Unit 1) and Amendments 27 and 28 (Unit 2), deletion of RTS Power Range Neutron Flux High Negative Rate Trip.
- Amendments 60 (Unit 1) and 39 (Unit 2), RTS Overtemperature ~T setpoint revision.
- Amendments 57 (Unit" 1) and 36 (Unit 2), RTS Overtemperature and Overpower ~T time constants and Overtemperature ~T setpoint.
- Amendments 43 and 44 (Unit 1) and 23 and 24 (Unit 2),
revised Overtemperature and Overpower ~T trip setpoints and allowable values.
- Amendments 104 (Unit 1) and 82 (Unit 2), revised RTS Intermediate Range Neutron Flux, Source Range Neutron Flux, and P-6 trip setpoints and allowable values.
- Amendments (Unit 1) and (Unit 2), revised Overtemperature ~T trip setpoint to limit value of the compensated temperature difference and revised the modifier for axial flux difference.
- Amendments (Unit 1) and (Unit 2), revised Overtemperature ~T and Overpower ~T trip setpoints to increase the fundamental setpoints K1 and Kt. and to modify coefficients and dynamic compensation terms.
- 7. WCAP-1 0271-P-A, Supplement 1, May 1986.
- 8. FSAR, Chapter 16.
- 9. Westinghouse Letter GP-16696, November 5, 1997.
(continued)
Vogtle Units 1 and 2 83.3.1-65
RTS Instrumentation 83.3.1 8ASES REFERENCES 10. WCAP-13632-P-A Revision 2, "Elimination of Periodic (continued) Sensor Response Time Testing Requirements," January 1996.
- 11. WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.
- 12. WCAP-14333-P-A, Rev. 1, October 1998.
- 13. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.
Vogtle Units 1 and 2 83.3.1-66