|
---|
Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20216G0881999-09-19019 September 1999 Proposed Tech Specs Pages,Changing TS 4.18.5.a.9 & Associated Bases for One Cycle of Operation Through End of Cycle 16 for Suppl Use of OTSG Repair Roll Technology ML20216G0721999-09-17017 September 1999 Proposed Tech Specs,Revising Once Through SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Volumetric Integranular Attack ML20212F5851999-09-17017 September 1999 Proposed Tech Specs Pages,Changing TS by Revising Curie Limits for Radioactive Gas Storage Tanks ML20211P0061999-09-0707 September 1999 Proposed TS in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria ML20211B3291999-08-19019 August 1999 Proposed TS 4.18.3.a.4 Clarification Supporting Tube End Cracking Alternative Repair Criteria ML20211C1701999-08-18018 August 1999 Proposed Tech Specs Revising ESAS Low RCS Pressure Setpoint to Ensure That TS Adequately Bounds Safety Analysis ML20211C4101999-08-18018 August 1999 Proposed Tech Specs Removing Requiremets for Tube Repair (Sleeving) & Clarifying Requirements for Preservice & ISI of Tubes to Facilitate Replacement of Original SGs ML20210K5121999-07-29029 July 1999 Proposed Tech Specs,Revising Requirements Associated with Plant Station Batteries & DC Sources to 125 Vdc Switchyard Distribution Sys ML20210K3571999-07-29029 July 1999 Proposed Tech Specs Revising SRs Pertaining to ISI Requirements for SG Tube Circumferential Cracking at top-of-tubesheet During 2P99 Planned mid-cycle Outage ML20209H5811999-07-14014 July 1999 Proposed Tech Specs Revising RB Structural Integrity Requirements Re ISI Reporting for Containment Structures, Tendons & Anchorages ML20209H8701999-07-14014 July 1999 Proposed Tech Specs 6.2.2 & 6.2.2.h,replacing Functional Description of Operations Staff Middle Manager with Ref to Functional Description Assistant Operations Manager & Revising TS Table 6.2-1 Item 5 & TS 6.2.2.f ML20209H9721999-07-14014 July 1999 Proposed Tech Specs Changes for Elimination of Programmatic Requirements Associated with PASS & Regulatory Relief from PASS Requirements of NUREG-0737 & RG 1.97 ML20195D2371999-06-0101 June 1999 Proposed Tech Specs Re Surveillance Requirements & Applicable Bases Relevant to ISI Requirements for Portions of once-through SG (OTSG) Tubes Adjacent to Primary Cladding Region of tube-to Tubesheet Roll ML20206T2651999-05-14014 May 1999 Proposed Tech Specs Revising SG Tubing SRs ML20206Q1281999-05-10010 May 1999 Proposed Tech Specs Bases,Page B 2-2,discussing Guidance to Be Used When Applying Later Editions & Addendas of ASME Code to RCS Components ML20205P4081999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Reactor Bldg Structural Integrity Requirements ML20205P6291999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Battery Requirements ML20205P3731999-04-0909 April 1999 Revised Proposed TS Page 5-1 Re Design Features Conversion ML20205P6931999-04-0909 April 1999 Proposed Tech Specs Change,Adding Specification 3.0.6 & Associated Bases to ANO-2 ML20207C0631999-02-25025 February 1999 Proposed Tech Specs Change Revising Action 2 Associated with Table 3.3-1, Reactor Protective Instrumentation. Change Adds Footnote to Action 2 Allowing Startup with Functional Units Associated with Channel D Excore Inoperable ML20207M7561999-02-24024 February 1999 Proposed Tech Specs,Allowing Startup & Continued Operation of Plant with Channel D Detector Until New Detector Is Installed During mid-cycle Outage Scheduled for Nov 1999 ML20198K1341998-12-17017 December 1998 Corrected Tech Specs Pages to 980629 Application for Amend to License NPF-6 ML20198A3661998-12-0808 December 1998 Proposed Tech Specs B3/4 6-3 & B3/4 6-4,deleting All Requirements Associated with Sodium H Addition Sys & Adding New Requirements for Trisodium Phosphate to Be Stored in Containers Located on Floor of Cb ML20196B4381998-11-24024 November 1998 Proposed Tech Specs Pages,Changing Administrative Controls Section to Implement Consolidated QA Plan Manual & Clarifying Responsibilities on STA Position on Shift ML20153D8491998-09-17017 September 1998 Proposed Tech Specs 2.2.1,3.3.1.1 & 3.3.2.1 Re Operating Bypass Setpoint ML20236Y5081998-08-0606 August 1998 Proposed Tech Specs Section 3.3.4.B Re Sodium Hydroxide Tank Limits ML20236Y5251998-08-0606 August 1998 Proposed Tech Specs Section 3.1.3.2,revising Action Requirements for CEA Position Indicator Channels ML20236Y5951998-08-0606 August 1998 Proposed Tech Specs,Revising Requirements Associated W/Units 1 & 2 Control Room Emergency Ventilation Sys ML20236Q2781998-07-13013 July 1998 Proposed Tech Specs Re Safety Limits & Limiting Safety Settings ML20236G6921998-06-29029 June 1998 Proposed Tech Specs,Revising TS 3.4.2 & TS 3.4.12 Re Reactor Coolant Sys Safety Valves & Low Temp Overpressure Protection ML20236G5421998-06-29029 June 1998 Proposed Tech Specs,Revising Surveillance Testing Requirements for Unit 2 Direct Current Electrical Distribution Sys ML20236G4911998-06-29029 June 1998 Proposed Tech Specs Re MSSVs & Pressurizer Safety Valves ML20236G5631998-06-29029 June 1998 Proposed Tech Specs,Revising Table 3.3-4 to Provide Range of Acceptable Values for 4160 Volt Bus Loss of Voltage Values ML20217M2621998-04-30030 April 1998 Proposed Tech Specs Re DG Load Rejection Testing ML20217N6141998-04-30030 April 1998 Proposed Tech Specs 1.6.1,revising Definition of Quadrant Power Tilt ML20217M4871998-04-28028 April 1998 Proposed Tech Specs Changing Order of Channel Rotation During Outage for Scheduling Flexibility,While Still Requiring Rotation Schedule to Be Established for Operating Cycle ML20217P8141998-04-0101 April 1998 Proposed Tech Specs Re Upper Tubesheet Volumetric Indications to Remain in Svc for Cycle 15 ML20216J7841998-03-17017 March 1998 Proposed Tech Specs Bases Change Concerning Plant Protection Sys Matrix Logic Testing ML20203D1061998-02-18018 February 1998 Proposed Tech Specs Revising Surveillance Testing Requirements for Emergency Feedwater Sys ML20202G9771998-02-0909 February 1998 Proposed Tech Specs Pages,Allowing Use of Repair Roll Technology (Reroll) for Upper Tubesheet Region of SGs ML20197A8421997-12-12012 December 1997 Proposed Tech Specs Re Establishment of Alternate Repair Criteria for Outer Diameter Intergranular Attack in Upper Tubesheet of Unit 1 SGs ML20211C6441997-09-23023 September 1997 Proposed Tech Specs,Modifying RPS & ESFAS Trip Setpoint & Allowable Values for SG Low Pressure ML20211C8911997-09-23023 September 1997 Proposed Tech Specs Reducing Min Required Reactor Coolant Sys Flow Rate Requirements Until ANO-2 SGs Are Replaced ML20151J3371997-07-28028 July 1997 Proposed Tech Specs,Modifying Actions Associated W/Ts Table 3.3-1 for Reactor Protective Instrumentation & Table 3.3-3 for Esfa Sys Instrumentation ML20149K0551997-07-21021 July 1997 Proposed Tech Specs,Revising Surveillance Testing Requirements for Facility EFW Sys as Specified in Specification 4.7.1.2 ML20141C5641997-06-18018 June 1997 Proposed Tech Specs Changes Restoring Table 2 to Table II & Revision to One of Offsite Dose Calculation Manual, Pages Previously Submitted.Also Includes Amended Pages Modified to Amends Issued Since Change Request Submitted ML20141A8001997-05-0808 May 1997 Proposed Tech Specs 3/4.3.1 & 3/4.3.2 Requesting Removal of Reference to CE NPSD-576 ML20138H8521997-05-0202 May 1997 Proposed Tech Specs,Submitting Rev to Original TS Change Request to Reduce Limit Specified in Section 3.1.6.3.b for Leakage Through SG Tubes of Any One Sg,Per SG Tube Insp Surveillance Requirements ML20138D6411997-04-24024 April 1997 Proposed Tech Specs Revising Requirements for Sample Scope Initial Size & Expansion for Inservice Insp of Steam Generator Tube Sleeves ML20141H2621997-04-18018 April 1997 Proposed Tech Specs,Permitting Delay of Scheduled three-yr Inspections of Flywheels on All Four RCPs Until Completion of 2R13 Refueling Outage,Scheduled for Winter of 1998-99 1999-09-07
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217N3171999-10-19019 October 1999 Rev 1,change 024-01-0 to OP-1903.011, Emergency Response Notifications ML20217J7641999-10-0505 October 1999 Rev 1 to Unit 1 Technical Requirement Manual, Replacing Index Page ML20216G0881999-09-19019 September 1999 Proposed Tech Specs Pages,Changing TS 4.18.5.a.9 & Associated Bases for One Cycle of Operation Through End of Cycle 16 for Suppl Use of OTSG Repair Roll Technology ML20212F5851999-09-17017 September 1999 Proposed Tech Specs Pages,Changing TS by Revising Curie Limits for Radioactive Gas Storage Tanks ML20216G0721999-09-17017 September 1999 Proposed Tech Specs,Revising Once Through SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Volumetric Integranular Attack ML20211P0061999-09-0707 September 1999 Proposed TS in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria ML20211B3291999-08-19019 August 1999 Proposed TS 4.18.3.a.4 Clarification Supporting Tube End Cracking Alternative Repair Criteria ML20211C4101999-08-18018 August 1999 Proposed Tech Specs Removing Requiremets for Tube Repair (Sleeving) & Clarifying Requirements for Preservice & ISI of Tubes to Facilitate Replacement of Original SGs ML20211C1701999-08-18018 August 1999 Proposed Tech Specs Revising ESAS Low RCS Pressure Setpoint to Ensure That TS Adequately Bounds Safety Analysis ML20210N5541999-08-0404 August 1999 Rev 28,Permanent Change 4 to OP-1903.061, Communication Equipment Test ML20210N5571999-08-0404 August 1999 Rev 23,Permanent Change 5 to OP-1903.011, Emergency Response/Notifications ML20210K5121999-07-29029 July 1999 Proposed Tech Specs,Revising Requirements Associated with Plant Station Batteries & DC Sources to 125 Vdc Switchyard Distribution Sys ML20210K3571999-07-29029 July 1999 Proposed Tech Specs Revising SRs Pertaining to ISI Requirements for SG Tube Circumferential Cracking at top-of-tubesheet During 2P99 Planned mid-cycle Outage ML20209H5811999-07-14014 July 1999 Proposed Tech Specs Revising RB Structural Integrity Requirements Re ISI Reporting for Containment Structures, Tendons & Anchorages ML20209H8701999-07-14014 July 1999 Proposed Tech Specs 6.2.2 & 6.2.2.h,replacing Functional Description of Operations Staff Middle Manager with Ref to Functional Description Assistant Operations Manager & Revising TS Table 6.2-1 Item 5 & TS 6.2.2.f ML20209H9721999-07-14014 July 1999 Proposed Tech Specs Changes for Elimination of Programmatic Requirements Associated with PASS & Regulatory Relief from PASS Requirements of NUREG-0737 & RG 1.97 ML20195D2371999-06-0101 June 1999 Proposed Tech Specs Re Surveillance Requirements & Applicable Bases Relevant to ISI Requirements for Portions of once-through SG (OTSG) Tubes Adjacent to Primary Cladding Region of tube-to Tubesheet Roll ML20206T2651999-05-14014 May 1999 Proposed Tech Specs Revising SG Tubing SRs ML20206Q1281999-05-10010 May 1999 Proposed Tech Specs Bases,Page B 2-2,discussing Guidance to Be Used When Applying Later Editions & Addendas of ASME Code to RCS Components ML20206J8021999-05-0404 May 1999 PC-03 to Procedure/Work Plan 1904-002, Offsite Dose Projections Rdacs Computer Method ML20205P6931999-04-0909 April 1999 Proposed Tech Specs Change,Adding Specification 3.0.6 & Associated Bases to ANO-2 ML20205P3731999-04-0909 April 1999 Revised Proposed TS Page 5-1 Re Design Features Conversion ML20205P6431999-04-0909 April 1999 Revised Proposed Technical Requirement Manual Pages ML20205P4081999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Reactor Bldg Structural Integrity Requirements ML20205P6291999-04-0909 April 1999 Proposed Tech Specs Changes,Revising Battery Requirements ML20207C0631999-02-25025 February 1999 Proposed Tech Specs Change Revising Action 2 Associated with Table 3.3-1, Reactor Protective Instrumentation. Change Adds Footnote to Action 2 Allowing Startup with Functional Units Associated with Channel D Excore Inoperable ML20207M7561999-02-24024 February 1999 Proposed Tech Specs,Allowing Startup & Continued Operation of Plant with Channel D Detector Until New Detector Is Installed During mid-cycle Outage Scheduled for Nov 1999 ML20198K1341998-12-17017 December 1998 Corrected Tech Specs Pages to 980629 Application for Amend to License NPF-6 ML20198A3661998-12-0808 December 1998 Proposed Tech Specs B3/4 6-3 & B3/4 6-4,deleting All Requirements Associated with Sodium H Addition Sys & Adding New Requirements for Trisodium Phosphate to Be Stored in Containers Located on Floor of Cb ML20196B4381998-11-24024 November 1998 Proposed Tech Specs Pages,Changing Administrative Controls Section to Implement Consolidated QA Plan Manual & Clarifying Responsibilities on STA Position on Shift ML20196C9761998-11-24024 November 1998 Rev 1 to Unit 2 Technical Requirements Manual ML20154D6681998-10-0101 October 1998 Rev 28 to Procedure/Work Plan 1904.002, Offsite Dose Projections Rdacs Computer Method ML20153D8491998-09-17017 September 1998 Proposed Tech Specs 2.2.1,3.3.1.1 & 3.3.2.1 Re Operating Bypass Setpoint ML20153H1651998-08-12012 August 1998 Rev 0 to ANO Units 1 & 2 Technical Requirements Manual ML20236Y5251998-08-0606 August 1998 Proposed Tech Specs Section 3.1.3.2,revising Action Requirements for CEA Position Indicator Channels ML20236Y5081998-08-0606 August 1998 Proposed Tech Specs Section 3.3.4.B Re Sodium Hydroxide Tank Limits ML20236Y5951998-08-0606 August 1998 Proposed Tech Specs,Revising Requirements Associated W/Units 1 & 2 Control Room Emergency Ventilation Sys ML20236Q2781998-07-13013 July 1998 Proposed Tech Specs Re Safety Limits & Limiting Safety Settings ML20236G6921998-06-29029 June 1998 Proposed Tech Specs,Revising TS 3.4.2 & TS 3.4.12 Re Reactor Coolant Sys Safety Valves & Low Temp Overpressure Protection ML20236G5421998-06-29029 June 1998 Proposed Tech Specs,Revising Surveillance Testing Requirements for Unit 2 Direct Current Electrical Distribution Sys ML20236G5631998-06-29029 June 1998 Proposed Tech Specs,Revising Table 3.3-4 to Provide Range of Acceptable Values for 4160 Volt Bus Loss of Voltage Values ML20236G4911998-06-29029 June 1998 Proposed Tech Specs Re MSSVs & Pressurizer Safety Valves ML20199F7211998-05-22022 May 1998 Rev 8 to Simulator Mod Control,Training Desk Guide 6.3 ML20217N6141998-04-30030 April 1998 Proposed Tech Specs 1.6.1,revising Definition of Quadrant Power Tilt ML20217M2621998-04-30030 April 1998 Proposed Tech Specs Re DG Load Rejection Testing ML20217M4871998-04-28028 April 1998 Proposed Tech Specs Changing Order of Channel Rotation During Outage for Scheduling Flexibility,While Still Requiring Rotation Schedule to Be Established for Operating Cycle ML20217P8141998-04-0101 April 1998 Proposed Tech Specs Re Upper Tubesheet Volumetric Indications to Remain in Svc for Cycle 15 ML20216J7841998-03-17017 March 1998 Proposed Tech Specs Bases Change Concerning Plant Protection Sys Matrix Logic Testing ML20203D1061998-02-18018 February 1998 Proposed Tech Specs Revising Surveillance Testing Requirements for Emergency Feedwater Sys ML20202G9771998-02-0909 February 1998 Proposed Tech Specs Pages,Allowing Use of Repair Roll Technology (Reroll) for Upper Tubesheet Region of SGs 1999-09-07
[Table view] |
Text
. _ _ .
The DNBR as calculated by the BAW-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.
The maximum thermal powe, for three pump operation is 88.92 percent due to a power level trip produced by the flux-flow ratic (74.7 percent flow x 1.054 = 78.73 percent power) plus the maximum calibration and instrumentation error. The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.
For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.
REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressitrized Water, BAW-10000A, May 1976.
(2) FSAR, Section 3.2.3.1.1.c.
i 9
8205050320 820429
~
PDR ADOCK 05000313 P PDR
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective Ta provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.
Bases-The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a preselected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in Table 2.3-1. The safety analysis has been based on these protection ,
system instrumentation trip setpoints plus calibration and instrumentation errors.
Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis.
A. Overpower Trip Based on Flow and Imbalance The power level trip setpoint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant-flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any electrical malfunction.
11
The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip setpoint '
produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:
- 1. Trip would occur when four reactor coolant pumps are operating if power is 105.4 percent and reactor flow rate t 100 percent or flow rate is 94.88 percent and power level is 100 percent.
- 2. Trip would occur when three reactor coolant pumps are operating ir power is 78.73 percent and reactor flow rate is 74.7 percent or flow rate is 71.16 percent and power level is 75 percent.
- 3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.85 percent and reactor flow rate is 49.2 percent or flow rate is 46.49 percent and the power level is 49.0 percent.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core tent valve because of the core vent valve surveillance program during each refueling outage. For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in top half of core minus power in bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces ths power level trip associated with reactor power-to-reactor power imbalance boundaries by 1.054 percent for a 1 percent flow reduction.
B. Pump Monitors In conjunction with the power imbalance / flow trip, the pump l
monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant 12
pump (s). The pump monitors also restrict the power level for the number of pumps in operation.
C. RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint.
The trip setting limit shown in Figure 2.3-1 for high RCS pressure (2300 psig) has been established to maintain.the system pressure below the safety limit (2750 psig) for any design transient.(2)
The low pressure (1800 psig) and variable low pressure (11.75 T - 5103) trip setpoints shown in Figure 2.3-1 have beenestNkishedtomaintaintheDNBratiogreaterthanor equal to 1.3 for those design accidents that result in a pressure reduction.(2,3)
Due to the calibration and instrumentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11.75 T out - 5143).
D. Coolant Outlet Temperature Thehighreactorcoolantoutlettemperaturetripsettinglimitl (618F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range.
Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620F.
E. Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
F. Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.
The reactor protection system segments that can be bypassed are shown in Table 2.3-1. Two conditions are imposed when the bypass is used:
- 1. A nuclear overpourtrip setpoint of 55.0 percent of rated power is automatically imposed during reactor shutdown.
l 2. A high reactor coolant system pressure trip setpoint of l 1720 psig is automatically imposed.
13
- O The purpose of the 1720 psig high pressure trip setpoint is to prevent normal operation with part of the reactor protection system bypassed. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The overpower trip setpoint of 55.0 prevents any significant reactor power from being produced when performing the physics tests.
Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
References (1) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4) FSAR, Section 14.1.2.8 (5) FSAR, Section 14.1.2.6 14 l
Tablo 2.3-1 Reactor Protection System Trip Setting Units
.Four RC Pumps Three.RC Pumps One RC Pump Operating Operating (Nominal) Operating (Nominal in each loop (Nominal Operating Power Operating Power Operating Power. . Shutdown 100%) 75%) -
49%) Bypass
' Nuclear power, %'of 104.9 104.9 104.9 5.0" rated, max ,
Nuclearbp wer based 1.054 times flow minus 1.054 times' flow minus 1.054 times flow minus Bypassed on flow and imbal- reduction due to imbal- reduction due to imbal- reduction due to imbal- '
ance, % of rated, :alance(s). ance(s). ance(s).
. max.
Nuclear power based NA NA 55 Bypassed onpumpmonitgrs,%
of rated; max a
High RC system 2300 2300 2300 1720 pressure, psig, max.
Low RC system ,
1800 1800 1800 Bypassed pressure, psig, min.
11.75 T out -5103 d d d M Variable low RC 11.75 Tout -5103 11.75 T out -5103 Bypassed system pressure, '
psig, min.
RC temp, F, max 618 618 618 '618 High reactor bldg. 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) pressure, psig,. max.
a Automatically set when other segments of the:RPS (as specified) are bypassed.
b Reactor coolant system flow.
c The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two pump operation.
d T
out is given indegrees Fahrenheit (F).
V i
t
. _ . . . - _ . _ ~_ ~ .._ . . _ - . . _ , . - . .- . _ _ _ _ _ _ _ _