Reactor Protection & Control Process Instrumentation Replacement Project at DC Cook Nuclear Plant Units 1 & 2 Reactor Protection Sys Functional Diversity Assessment.| ML17329A720 |
| Person / Time |
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| Site: |
Cook  |
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| Issue date: |
12/14/1992 |
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| From: |
AMERICAN ELECTRIC POWER SERVICE CORP. |
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| To: |
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| Shared Package |
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| ML17329A705 |
List: |
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| References |
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| 2985-VDV-01, 2985-VDV-01-R01, 2985-VDV-1, 2985-VDV-1-R1, NUDOCS 9212180095 |
| Download: ML17329A720 (6) |
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[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
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Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
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Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
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Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
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Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
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Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
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LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
REACTOR PROTECTION AND CONTROL PROCESS INSTRUMENTATION REPLACEMENT PROJECT AT DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 REACTOR PROTECTION SYSTEM FUNCTIONAL DIVERSITY ASSESSMENT REPORT NO. 2985-VDV-01, REV 0 Prepared by: Date l?.-'7- t ~
Concurred by: ()Ci~/
Approved by: ~d.l Date /A /
9212180095 921216 PDR" ADOCK 050003i5
'P PDR
UALITATIVE FUNCTIONAL DIVERSITY ASSESSMENT EXECUTIVE
SUMMARY
On April 21, 1992, AEPSC representatives had a meeting with the NRC on the replacement of existing analog reactor protection process instrumentation with digital Foxboro Spec 200/Spec 200 Micro Electronics instrumentation. During this meeting, AEPSC was asked to assume a common mode failure (CMF) of the software of the new digital equipment during a postulated accident and then provide details as to whether operators could mitigate the consequences of the accident.
In response to this request, a functional diversity assessment of each updated FSAR (UFSAR) event assuming a common mode failure of the software has been performed. In this assessment, all the events for both Units 1 and 2 of the Cook Nuclear Plant given in the UFSAR were considered. A review was performed to divide events into potentially affected and not affected. Table-1 lists these events and indicates whether they would be potentially affected or not affected if a CMF were to occur. The potentially affected transients were then individually evaluated qualitatively in light of the FSAR analysis.
Each event evaluation was recorded on a form of the type shown in Appendix A.
This form outlines the thought process employed. The first column in Appendix A contains the UFSAR transient number listed in Table-1. The second column includes the name of the transient. The third column depicts the trip/safeguard function for reactor trip. This information was obtained from the UFSAR. The fourth column includes the information on the impact of common mode failure on the reactor trip function. If the trip function is processed outside of the new digital reactor protection system, then the trip is available, e.g., trip on nuclear instrumentation system high flux. If the trip is processed by a function that is a part of the new digital equipment, then the trip/ESF function is assumed to be lost. However, for some functions, alternate indications and/or diverse alarms are available. The alarm/alternate indications that are available to the operator to mitigate the transient are given in the next column. The sixth column lists pertinent diagram numbers. The seventh column summarizes the consequences of the unavailability of diverse alarm. The last column provides the evaluation of the event. In this column, we have discussed the consequences of the operator's response on reactor safety.
Based on this evaluation, we have concluded that the CMF of the new digital equipment has no significant adverse impact on the public safety. Some reactor trips are not affected by the installation of the new digital equipment. Among these trips are neutron high flux and high rate trips, undervoltage and underfrequency trips and reactor trip on turbine trip. However, for events protected by trip actuations affected by the CMF, the operator will be alerted to the event by an alarm. He will then provide the appropriate actuations manually and enter the emergency operating procedures. For some accidents, such as locked rotor, the consequences could be more severe than currently analyzed due to the longer response time for the required actuation. However, our evaluation indicates that the affected unit can be brought to a safe condition and the current LOCA offsite dose evaluation will remain bounding. From these results, it is believed that a CMF of the new digital system would have no adverse effect on the health and safety of the public.
Table-1 POTENTIALLY TRP2lSIENT 0 TRANSIENT AFFECTED (A)/
NOT AFFECTED(NA) 14.1.1 ncontrolled RCCA Withdrawal from a Subcritical Condition A 14.1.2 ncontrolled RCCA Withdrawal at Power A 14.1.3 od Cluster Control Assembly Misalignment A 14.1.4 CCA Drop A 14.1.5 Chemical Volume and Control System Malfunction A 14.1.6 Loss of Reactor Coolant Flow A 14.1.7 Startup of an Inactive Reactor Coolant Loop A 14.1.8 Loss of External Electrical Load A 14.1.9 Loss of Normal Feedwater Flow A 14.1.10 Excessive Heat Removal due to Feedwater System Malfunction A 14.1.11 Excessive Load Increase Incident A 14.1.12 Loss of All A.C. Power to the Plant Auxiliaries A 14.1.13 Turbine-Generator Safety Analysis A 14.2.1 Fuel Handling Accident A 14.2.2 ccidental Release of Radioactive Liquids A 14.2.3 ccidental Waste Gases Release A 14.2.4 Steam Generator Tube Rupture A 14.2.5 upture of a Steam Pipe A 14.2.6 upture of a Control Rod Drive Mechanism Housing (RCCA A Ejection) 14.2.7 Secondary System Accidents Dose Consequences A 14.2.8 ajor Rupture of a Main Feedwater Pipe A 14.3.1 Large Break LOCA Analysis A 14.3.2 Loss of Reactor Coolant from Small Ruptured Pipes or from A Cracks in Large Pipes which Actuates the Emergency Core Cooling System 14.3.3 Core and Internals Integrity Analysis NA 14 ' ' Containment Integrity Analysis A 14.3.5 Environmental Consequences of a Loss of Coolant Accident A 14.3.6 Hydrogen in the Containment After a Loss of Coolant A ccident 14.3.7 Long Term Cooling NA 14.3.8 Nitrogen Blanketing NA 14.4.2 Postulated Pipe Failure Analysis Outside Containment NA 14.4.3 nalysis of Emergency Conditions NA 14.4.4 Stress Calculations NA 14.4.5 Description of Pipe Whip Analysis NA 14.4.6 Compartment Pressures and Temperatures NA 14.4.7 Description of Jet Impingement Load Analysis NA 14.4.8 Containment Integrity NA 14.4. 9 Plant Modifications NA 14.4.10 Environment NA 14.4.11 Electrical Equipment Environmental Qualification A
APPENDIX A UNIT 2 fSAR TRANSIENT TRIP/SAF ECUARO FUNCTION IMPACT OF COMMON W)E FOR LARN/ALTERNATE IMDICATIOM DIAGRAM 0 COMSEOUEMCES OF FVALUA'IION Of EVENf TRANSIENT RX TRIP fSAR FAILURE (CNF) ON TRIP TSTEN AVAILABLE UMAVAILABILITT OF FUNCTION DIVERSE ALARM
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