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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML17300B3071999-03-31031 March 1999 Seismic Portion of Submittal-Only Screening Review of Palo Verde Nuclear Generating Station Units Ipeee. ML17312A7631996-05-0808 May 1996 Calculation Summary of Radiological Doses for SG Tube Rupture W/Loss of Offsite Power & Stuck Open Adv. ML20117J0491996-05-0808 May 1996 At&T Round Cell Nuclear Util User'S Council Charter ML20113E1631996-05-0808 May 1996 PVNGS Unit 3 SG Eddy Current Exam Fifth Refueling Outage Nov 1995 ML17311A2031994-08-11011 August 1994 Forwards Unit 2 SG IR from Last Mid Cycle Outage.Aps Stated That It Would Present,Within Four & One Half Months of Breaker Closure,Final Regulatory Guide 1.121 Evaluation Results of Pulled Tube Analyses & Statistical Treatment ML20062M5111993-12-31031 December 1993 1993 Certification Submittal Simulation Facility Simulator Training Suite a ML20058N7161993-09-21021 September 1993 Independent Safety & Quality Engineering GL-89-01 Motor Operated Valve Programmatic Assessment ML17306A9361992-08-20020 August 1992 Rev 1 to JCO-91-02-01, Justification for Continued Operation Steam Generator Tube Rupture Analysis Concerns. ML17305B3381991-01-18018 January 1991 Justification for Continued Operation - Potential for Small Break Loss of Coolant Accident Due to Tube Rupture in Reactor Coolant Pump Seal Cooler. ML17305A9811990-07-31031 July 1990 Conceptual Design for Palo Verde Nuclear Generator Station for Diverse Auxiliary Feedwater Actuation Sys (Dafas). ML17305A3351989-10-20020 October 1989 Rev 3 to Justification for Continued Operation Control Element Assembly (CEA) Drop Events. ML17305A3211989-10-0505 October 1989 Procedure 43OP-3ZZ16,RCS Drain Operations,Not Appropriate for Circumstances. ML17305A3401989-07-18018 July 1989 Unit 3 Steam Generator Eddy Current Exam,First Refueling Outage Apr 1989, NDE Summary Rept ML17304B2941989-06-0909 June 1989 Justification for Continued Operation,Control Element Assembly Drop Events. ML17304B2321989-06-0101 June 1989 Essential & Emergency Lighting Sys Rept. ML17304B2151989-05-22022 May 1989 Revised Electrical Distribution Sys Design Assessment. ML17304B2141989-05-19019 May 1989 Unit 3 Reactor Trip Following Large Load Reject. ML20244B0961989-05-0808 May 1989 Rev 0 to Compressed Gas Sys Evaluation & Analysis for Palo Verde Units 1,2 & 3 ML20247L6921989-05-0808 May 1989 Electrical Distribution Sys Design Assessment ML20247L6851989-05-0808 May 1989 Compressed Gas Sys Evaluation & Analysis Rept ML17304B2051989-05-0606 May 1989 Steam Bypass Control Sys Overall Final Rept. ML20247L7271989-04-30030 April 1989 Atmospheric Dump Valve Engineering Analysis ML17304B0121989-01-31031 January 1989 Final Rept on Pressure-Temp Limits for Palo Verde Nuclear Generating Stations. ML17304A7551988-11-0707 November 1988 Analysis of Equipment Functionality W/Essential Chiller Sys Inoperable. ML20082D9761988-10-30030 October 1988 EE580 Field Verification Suppl to Final Rept Ref NRC Allegation RV-87-A-047 for Palo Verde Nuclear Generating Station,Oct 1988 ML17304A7601988-10-13013 October 1988 Loss of Qualified Life & Functionality of Electrical Equipment Due to Loss of Essential Cooling Sys in Unit 1. ML17304A2211988-06-0909 June 1988 Justification for Continued Operation of Auxiliary Feedwater Pumps. ML20082D9991988-05-30030 May 1988 EE580 Field Verification Final Rept Ref NRC Allegation RV-87-A-047 ML17303A8871988-02-16016 February 1988 Special Plant Event Evaluation Rept 87-02-019 & Mods to Valves Sga Uv 134 & Sga Uv 138 Render 2AFA-P01 on 871127 ML17303A6291987-09-24024 September 1987 Arizona Nuclear Power Project Metallurgical Investigation Rept,Palo Verde Nuclear Generating Station Emergency Diesel Generator 3B Number 9L Piston Pin. ML17303A4541987-06-0404 June 1987 Unit 1 Steam Generator Eddy Current Exam,Feb 1987, NDE Summary Rept ML17303A3571987-03-12012 March 1987 Torsional Vibration Analysis KSV-20-T SN-7183-88 SO-0391. ML17303A3071987-02-28028 February 1987 Rev 0 to Rept on Steam Generator Tube Leak at Unit 1. ML20215G2981987-02-24024 February 1987 Technical Evaluation Rept for SPDS for Palo Verde Nuclear Generating Station Units 1,2 & 3 ML17300A8211987-02-0606 February 1987 Torsional Vibration Analysis of Repaired Crankshaft, KSV-20-T & SN-7187. ML17303A3551987-02-0606 February 1987 Torsional Vibration Analysis of Repaired Crankshaft KSV-20-T SN-7187 Arizona Public Svc Co,Palo Verde Nuclear Generating Station Diesel Generator III-B. ML17303A2421987-01-22022 January 1987 Rev 0 to 10CFR50,App R Safe Shutdown Evaluation.Outside Control Room Fire Spurious Actuation Study for Palo Verde Nuclear Generating Station Unit 3. ML17303A2331987-01-21021 January 1987 Loose Part Monitoring Sys,Loose Part Detection Program Rept. ML17300A6181986-10-31031 October 1986 Description of Proposed Enhancements to Palo Verde Nuclear Generating Station Control Bldg Elevation 74-Ft 0-Inches Masonry Walls for Units 1,2 & 3. ML17300A5861986-09-30030 September 1986 Evaluation of Block Masonry Walls at Palo Verde Nuclear Generating Station, Technical Rept ML17300A5301986-09-18018 September 1986 Factors Influencing Deflections in Grouted Hollow Unit Concrete Masonry Walls. ML20212A7011986-07-31031 July 1986 Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 2 (Sys 80 Nonprototype - Category 1),Evaluation of Precore Hot Functional Insp Program, Final Rept ML17303A3141986-07-29029 July 1986 Suppl to 860610 Application Re Sale & Leaseback Transactions by Public Svc Co of New Mexico ML20215E9291986-05-23023 May 1986 Observations & Comments,Initial Reactor Startup & Low Power Reactor Physics Tests,Palo Verde Nuclear Generating Station, Unit 2,NRC Region V Insp ML17299A5771985-08-30030 August 1985 Rev 0 to 10CFR50,App R Safe Shutdown Evaluation,Outside Control Room Fire Spurious Actuation Study. ML20137A6691985-06-21021 June 1985 Observations & Comments,Initial Reactor Startup & Low Power Reactor Physics Tests,Palo Verde Nuclear Generating Station,Unit 1,NRC Region V Enhanced Insp ML20116N4701985-04-30030 April 1985 Refueling Cavity Water Seal, Summary Rept in Response to IE Bulletin 84-03 ML17298B9321985-02-28028 February 1985 SPDS Sar. ML20107H5181984-12-31031 December 1984 Assessment of Bolting Integrity at Palo Verde Nuclear Generating Station Units 1,2 & 3 ML17298B6851984-12-17017 December 1984 Bechtel Study 13-ES-600, Reg Guide 1.75 Low Energy Circuit Analysis. 1999-03-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17300B3811999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pvngs,Units 1,2 & 3.With 991007 Ltr ML17300B3271999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Pvngs,Units 1,2 & 3 ML17313B0751999-08-27027 August 1999 LER 99-002-00:on 990730,test Mode Trip Bypass for EDG Output Breakers Not Surveilled.Cause Under Investigation.Operations Personnel Conservatively Invoked SR 3.0.3 for SR 3.8.1.13. with 990827 Ltr ML17313B0611999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pvngs,Units 1,2 & 3.With 990810 Ltr ML17313B0191999-07-16016 July 1999 LER 99-005-00:on 990618,RT on Low DNBR Was Noted.Caused by Hardware Induced Calculation Error.Cr Operator Was Taken to Place Reactor in Stable Condition IAW Appropriate Operating Procedure ML17300B3151999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pvngs,Units 1,2 & 3.With 990714 Ltr ML17313A9921999-06-21021 June 1999 Special Rept:On 990525,RMS mini-computer Was Removed from Service to Implement Yr 2000 Mod & Was OOS Longer than 72 H Allowed.Caused by Planned Y2K Mods.Preplanned Alternate Sampling Program Was Initiated ML17313A9911999-06-18018 June 1999 Special Rept:On 990510,loose-part Detection Sys Channel 2 Was Declared Inoperable.Caused by Malfunction of Mineral Cable Connector to Accelerometer.Licensee Will Implement Modifications Which Will Enhance loose-part Detection Sys ML17313A9731999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pvngs,Units 1,2 & 3.With 990608 Ltr ML17313A9281999-05-0707 May 1999 LER 99-004-00:on 990408,PSV Lift Pressures Were Outside of TS Limits.Caused by Lift Pressure Setpoint Drift.Psvs Have Been Tested,Disassembled,Inspected,Reassembled & Certified at Wyle Labs ML17313A9201999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pvngs,Units 1,2 & 3.With 990512 Ltr ML17313A8951999-04-14014 April 1999 LER 99-003-00:on 990317,required Surveillance Requirement Not Completed Due to Deficient Procedure,Was Determined. Caused by Cognitive Personnel Error.St Procedures Revised to Require Chiller to Be Operating & Oil Temperature Checked ML17313A8921999-04-13013 April 1999 LER 98-003-01:on 980902,discovered That MSSV as-found Lift Pressures Were Outside TS Limits.Caused by Bonding of Valve Disc to Nozzle Seat.Affected Valves Were Adjusted,Retested & Returned to Svc ML17313A8891999-04-0909 April 1999 LER 99-001-00:on 990310,RT on High Pressurizer Pressure Was Noted.Caused by Loss of Heat Removal.Cr Supervisor Was Removed from Shift Duties for Diagnostics Skills Training. with 990409 Ltr ML17300B3071999-03-31031 March 1999 Seismic Portion of Submittal-Only Screening Review of Palo Verde Nuclear Generating Station Units Ipeee. ML17313A8801999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pvngs,Units 1,2 & 3.With 990412 Ltr ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207H7471999-03-10010 March 1999 1999 Emergency Preparedness Exercise 99-E-AEV-03003 ML17313A8361999-03-0101 March 1999 LER 99-001-00:on 990103,TS Violation for Power Dependent Insertion Limit Alarm Being Inoperable.Caused by Personnel Error.Revised Procedure to Clarify How Computer Point Is to Be Returned to Scan Mode.With 990302 Ltr ML17313A8501999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Palo Verde Nuclear Generating Station.With 990311 Ltr ML17313A7791999-02-0505 February 1999 Safety Evaluation Accepting Licensee Rev to Emergency Plan That Would Result in Two Less Radiation Protection Positions Immediatelu Available During Emergencies ML17313A8061999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Pvngs,Units 1,2 & 3.With 990218 Ltr ML17313A7701999-01-15015 January 1999 LER 96-008-00:on 960507,inadequate Procedure Results in Nuclear Power Channels Not Calibrated During Power Ascension Tests Occurred.Caused by Deficient Procedure.Procedure Revised ML17313A7381998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Palo Verde Nuclear Generating Station,Units 1,2 & 3.With 990113 Ltr ML20206H2101998-12-31031 December 1998 SCE 1998 Annual Rept ML17313A7031998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Pvngs,Unit 1,2 & 3. with 981209 Ltr ML17313A6701998-11-0404 November 1998 Rev 2 to PVNGS Unit 2 Colr. ML17313A6741998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Pvngs,Units 1,2 & 3.With 981109 Ltr ML17313A6611998-10-24024 October 1998 LER 98-008-00:on 980729,EQ of Electrical Connectors Were Not Adequately Demonstrated.Caused Because Test Was Conducted with Only Single Lv Connector & Without Fully Ranged Inputs. Revised EQ Requirements ML17313A6561998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for PVNGS Units 1,2 & 3.With 981007 Ltr ML17313A5961998-09-14014 September 1998 LER 98-002-00:on 980814,B Train H Recombiner Was Noted Inoperable Due to cross-wired Power Receptacle.Cause of Event Is Under investigation.Cross-wired Power Supply Receptacle for B Train H Recombiner Was re-wired ML17313A5761998-09-0808 September 1998 LER 98-003-01:on 980113,discovered That One Channel of RWT Level Sys Had Failed High.Caused by Water Intrusion Into Electrical Termination Pull Box.Weep Holes Were Drilled Into Bottoms of Pull Boxes Nearest Level Transmitters ML17313A5591998-08-28028 August 1998 LER 98-001-00:on 980730,entered TS 3.0.3 Due to Safety Injection Flow Instruments Being Removed from Svc.Caused by Personnel Error.Transmitters Were Unisolated & Returned to svc.W/980828 Ltr ML20151S0941998-08-21021 August 1998 Rev 6 to COLR for PVNGS Unit 3 ML20151S0861998-08-21021 August 1998 Rev 4 to COLR for PVNGS Unit 1 ML20151S0901998-08-21021 August 1998 Rev 1 to COLR for PVNGS Unit 2 ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML17313A5401998-08-13013 August 1998 Special Rept:On 980715,declared PASS Inoperable.Caused by Failure of Offgas Flush/Purge Control Handswitch HS0101. Handswitch Replaced & Post Maintenance Retesting Was Initiated ML17313A5301998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Pvgns,Units 1,2 & 3.W/980812 Ltr ML17313A5201998-07-30030 July 1998 LER 98-004-00:on 980630,personnel Discovered That Pressure Safety Valve Had Not Received Periodic Set Pressure Test for ASME Class 1 Pressure Safety Valve.Caused by Personnel Error.Pressure Safety Valve reviewed.W/980730 Ltr ML17313A5791998-07-0707 July 1998 to PVNGS SG Tube ISI Results for Seventh Refueling Outage Mar & Apr 1998. ML17313A5001998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Palo Verde Nuclear Generating Station,Units 1,2 & 3.W/980710 Ltr ML17313A4671998-06-19019 June 1998 LER 98-007-00:on 980520,CR Personnel Observed Flow & Pressure Perturbations on Chemical & Vol Control Sys Letdown Sys.Caused by Cyclic Fatigue Due to Dynamic Pressure Transients.Unit Letdown Piping Replaced ML17313A4521998-06-19019 June 1998 Rev 5 to COLR for Pvngs,Unit 3. ML17313A4501998-06-19019 June 1998 Rev 4 to COLR for Pvngs,Unit 3. ML17313A4131998-06-0505 June 1998 LER 98-006-00:on 980507,determined That Plant Was Outside Design Basis Due to SI Discharge Check Valve Reverse Flow. Check Valve Was Disassembled,Examined & Reassembled, Whereupon Valve Met Acceptance Criteria ML17313A4211998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Pvngs,Units 1,2 & 3.W/980609 Ltr ML17313A3951998-05-26026 May 1998 LER 98-005-00:on 980428,noted That Required Response Time Testing Had Not Been Performed.Caused by Personnel Error. Coached I&C Personnel Responsible for Reviewing Work Authorization Documentation ML17313A3691998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for PVNGS.W/980412 Ltr ML17313A3251998-04-0101 April 1998 LER 98-004-00:on 980304,safety Valves as-found Pressures Out of Tolerance.Cause of Event Is Under Investigation.Three Mssv'S & Psv Will Be Replaced W/Refurbished & Recertified Valves During Refueling Outage U1R7 1999-09-30
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Enclosure to ST-96%287 Page 1of9 Enclosure to ST-96-0287 Calculation Summary of Radiological Doses for Steam Generator Tube Rupture with Loss of Offsite Power and Stuck Open ADV Prepared By ehran lbabai NamelSignaturtJD ate VERIFICATIONSTATUS'OMPLETE The Safety-Related design information contained in this document has been verified to bc correct by means of Design Rcvicw using the Other Design Document Checklist of QPM 101.
Name < ~ ~~<<<>><KiSignaturc~ ~~ Date~~ii Independent Reviewer 9g0SZ<oasa 9 o PDR ADOCK 05000528 II PDR
Enclosure to ST-96-0287 Page 2 of9
Subject:
Calculation of the Radiological Doses for the Steam Generator Tube Rupture with LOSS OF OFFSITE POWER and Stuck Open ADV event at 2% Stretch Power
SUMMARY
The following discussion summarizes selected methods and assumptions used in calculation of the radiological releases for the Steam Generator Tube Rupture with LOSS OF OFFSITE POWER and Stuck open ADV at 2% stretch power. In summary:
- 1) The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> primary to secondary leakages and secondary steam releases remain the same as the original calculation. The primary to secondary leakage is a function of the RCS and steam generators pressures which are mostly controlled by the operator during the event and are not impacted by the 2% power increase or lower AFAS setpoint.
For the erst two hours, the secondary steam releases are driven by the stuck open ADV on the affected steam generator which removes much more heat than the core generates as decay heat. For the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the total integrated decay heat generated by the stretch power core using the 1979 ANS decay heat curve is less than that assumed in the original analysis.
- 2) The additional radiological doses are due to the increased period of tube uncovery during the event. The assumed period of uncovery was increased &om 887 seconds in the original calculation to 1230 seconds (effectively assuming a tube uncovery period &om 460 seconds to 1690 seconds) for the stretch power analysis.
- 3) A copy of the chronology of the event was marked up and is attached. The only markup is the time that the level in the faulted steam generator rises above the top of the U-tubes. This value has been increased &om 1385 seconds (in the copy) to 1690 seconds. Even though the changes considered in the stretch power calculation would slightly impact the timing of the event, the increase in time of level recovery above the U-tubes is the only change needed to represent the conservative approach used for radiological dose calculation.
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Enclosure to ST-96-0287 Page 3 of 9 DISCUSSION The radiological doses for this event are primarily due to the releases &om the stuck open ADV on the affected steam generator. Most of the releases occur between the time that the affected steam generator ADV becomes stuck and the time that the level in the affected steam generator rises above the top of the tubes. During this period, the entire primary to secondary leakage is assumed to flash immediately and be released to atmosphere with a DF of 1.0. The original calculation of this event established this time to be &om 460 seconds to 1347 seconds, for a total of 887 seconds (14.8 minutes). The 2%
stretch power calculation increased this time period to 1230 seconds (20.5 minutes),
effectively assuming a tube uncovery period from 460 seconds to 1690 seconds. The additional 343 seconds are due to the following considerations:
- 1) 212 seconds resulting &om reduction of the AFW rate &om 750 GPM to 650 GPM.
- 2) 60 seconds to account for lower steam generator masses at initiation of the event and the time of AFAS initiation.
- 3) 71 seconds resulting &om the reduction of AFAS initiation analytical setpoint &om 25% to 21% of wide range span.
The 2% higher power does not affect the two hour or 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> steam releases and radiological doses. For the first two hours, the secondary steam releases are driven by the stuck open ADV on the affected steam generator which removes much more heat than the core generates as decay heat. For the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, the total integrated decay heat generated by the stretch power core using the 1979 ANS decay heat curve is less than that assumed in the original analysis.
Based on the above discussion, the two hour and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PIS and GIS radiological doses were calculated as follows:
2 Hour PIS From the original calculation, the radiological doses during the tube uncovery period (460 to 1347 seconds) were 185.2 REM. The total two hour radiological doses were 206.6 REM.
The two hour PIS doses for stretch power are calculated by ratioing the original dose release in proportion to the increased time of tube uncov'ery, as follows:
K20.5/14.8)
- 185.2] + (206.6- 185.2) = 278 REM.
Enclosure to ST-96-0287 Page 4 of 9 The radiological doses were conservatively increased by another 5% to account for the expanded MSSV setpoint tolerances &om 1% to 3%. This brings the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> PIS dose to 292 REM (rounded up).
2 Hour GIS From the original calculation, the radiological doses between 1337 and 1690 seconds (the additional period of tube uncoveiy for stretch power) were recalculated, assuming that the entire primary to secondary leakage Qashes immediately and becomes airborne with a DF of 1.0, The additional doses during this period were calculated to be 30.6 REM. The additional doses during this period were added to the 40.4 REM, the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses &om the original calculation. This resulted in a two hour GIS dose of 71 REM.
The increased MSSV tolerances do not significantly impact the overall GIS doses since the GIS spiking is small during the MSSV opening this period.
8 Hour PIS and GIS The additional dose increases determined above for two hour PIS and GIS, resulting &om the increased tube uncovery period for stretch power were subsequently recalculated using the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to two hour dispersion factor ratio and then added to the respective 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PIS and GIS doses of the original calculation.
t Enclosure to ST-96-0287 Page 5 of9 a valve Vruasru I'Oars DECREASE IN REACTOR COOLANT INVENTORY Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 1 of 5)
Time Setpoint Success Path (sec) Event or Value or Comment 0.0 Tube rupture occurs 40 Third charging pump -0.75 Primary system started, feet'below integrity program level 40 Letdown control valve -0.75 Primary system throttled back to integrity minimum flow, feet below pxogram level 47 CPC hot leg saturation Reactivity trip signal generated control 47.15 Trip breakers open Reactivity control Turbine/generator trip Secondary system integrity 51 Loss of offsite power 52 LH main steam safety 1) 265 Secondary system valves open, psia integrity 52 RH main steam safety 1,265 Secondary system valves open, psia integrity 56 Maximum steam generator 1,330 pressures both steam generators, psia (1)
March 1990 15.6-40 Revision 2
PVNGS UPDATED FSAR I ~
DECREASE IN REACTOR Enclosure to ST-96-0287 COOLANT INVENTORY Page6of9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULIY STUCK OPEN ADV (Sh'eet 2 of 5)
Time Event Setpoint Success Path (sec) or Value or Comment 121 Steam generator water 25+ Secondary system level reaches auxiliary integrity feedwater actuation signal (AFAS) analysis setpoint .in unaffected generator, percent wide range level 122 AFAS generated 131 Steam generator water 25+ Primary system level reaches AFAS integrity analysis setpoint in the affected generator, percent wide range level 132 AFAS generated 167. 0 Auxiliary feedwater Secondary system initiated to unaffected integrity steam generator 177. 0 Auxiliary feedwater Secondary system initiated to affected integrity steam generator
- The analysis used a setpoint of 21%. Even though this change would slightly impact the timing of the event, the only change needed to represent the conservative approach of the analysis is the time of level recovery above the u-tubes.
L (1)
Revision March 1990 15. 6-41 2
CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-0287 Page7of9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FUI LY STUCK OPEN ADV (Sheet 3 of 5)
Time Event Setpoint Success Path.
(sec) or Value or Comment 460 Operator initiates Reactor heat plant cooldown by removal opening one ADV on each SG ADV of the affected SG instantane-ously opens fully 484 Pressurizer empties 513 MSIS actuation 919 Secondary system secondary pressure. integrity psia 535 Automated isolation of 185 Secondary system AFW to affected SG, integrity hP SGs, psi 581 Pressurizer pressure 1, 578 Reactivity control reaches safety injection actuation signal (SIAS) analysis setpoint, psia 581 Safety injection actuation signal generated (1)
March 1990 15.6-42 Revision 2
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4 CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-02&7 Page 8 of 9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 4 of 5)
Time Event Setpoint Success Path (sec) or Value or Comment 581 Safety injection flow Reactivity control initiated 655 Operator overrides the AFW isolation signal and starts feeding the affected SG with AFW 775 Operator takes manual control of the AFW system, feeds affected SG with both AFW pumps 895 Operator shuts the ADV of the unaffected steam generator 1015 Operator initiates auxiliary spray to the pressurizer Level in the affected 71. 5 16 fg SG above the top of U-tubes, percent wide range (1)
March 1990 15.6-43 Revision 2
CREASE IN REACTOR COOLANT INVENTORY Enclosure to ST-96-0287 Page 9 of 9 Table 15.6.3-6 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER AND FULLY STUCK OPEN ADV (Sheet 5 of 5)
Time Event Setpoint Success Path (sec) or Value or Comment 2040 Pressurizer level, 50 percent 2400 Operator controls HPSI 20 flow, backup pressur-izer heater output, and auxiliary spray flow to control RCS pressure and subcooling, F 28.800 Shutdown cooling entry 400/350 conditions are reached; RCS pressure, psia/
temp. F 28,800 Operator activates shutdown cooling system (1)
March 1990 15.6-44 Revision 2