ML17333A695
| ML17333A695 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/04/1996 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. |
| References | |
| NUDOCS 9612120438 | |
| Download: ML17333A695 (16) | |
Text
December 4,
1996 Hr.
E.
E. Fitzpatrick, Vice President Indiana Michigan Power Nuclear. Generation Group 500 Circle Drive
- Buchanan, HI 49107
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT DONALD C.
COOK NUCLEAR PLANT, UNIT 1
Dear Hr. Fitzpatrick:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational event at the Donald C.
Cook Nuclear Plant, Unit 1, reported in Licensee Event Report (LER) No. 315/95-011.
~
This final analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evaluation of comments on the preliminary analysis and comments received from the NRC staff and from our independe'nt contractor, Sandia Natio'nal Laboratories (SNL).
We were informed that you had no comments on, the preliminary analysis.
Resolution of the comments we did receive resulted in minor revisions to the analysis and associated documentation.
The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-3017 if you have any questions regarding the enclosures.
We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, Docket No. 50-315
Enclosure:
ASP Analysis cc w/encl:
See Next Page Original signed by:
John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DISTRIBUTION:
JRoe WBeckner Docket File PUBLIC OGC ACRS PO'Reilly, AEOD GMarcus PD8 3-3 Reading JCaldwell, RIII DOCUHENT NAME:
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December 4,
1996 Hr. E.
E. Fitzpatrick, Vice President Indiana Hichigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, HI 49107
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT DONALD C.
COOK NUCLEAR PLANT, UNIT 1
Dear Hr. Fitzpatrick:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational event at the Donald C.
Cook Nuclear Plant, Unit 1, reported in Licensee Event Report (LER) No. 315/95-011.
This final analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evaluation of comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL).
We were informed that you had no comments on the preliminary analysis.
Resolution of the comments we did receive resulted in minor revisions to the analysis and associated documentation.
The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-3017 if you have any questions regarding the enclosures.
We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, Docket No. 50-315
Enclosure:
ASP Analysis cc w/encl:
See Next Page Original signed by:
John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Bl IBIBUII B:
U I I PIP PUBLIC JRoe OGC ACRS WBeckner PO'Rei1 ly, AEOD GHarcus PD¹ 3-3 Reading JCaldwell, RIII DOCUMENT NAME:
G:iDCCOOKiASP.LTR To teoebe ~ copy ot thte ootenent. Inckete In the bottt 'c' copy withottt ettechmentlendoetee v copy wtth ettechment/endoetee 'It'to copy OFFICE PM:PD33 E
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 4, 1996 Mr. E.
E. Fitzpatrick, Vice President Indiana Michigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, MI 49107
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT DONALD C.
COOK NUCLEAR PLANT, UNIT 1
Dear Mr. Fitzpatrick:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational event at the Donald C.
Cook Nuclear Plant, Unit 1, reported in Licensee Event Report (LER) No. 315/95-011.
This final analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evaluation of comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL).
We were informed that you had no comments on the preliminary analysis.
Resolution of the coIImIents we did receive resulted in minor revisions to the analysis and associated documentation.
The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-3017 if you have any questions regarding the enclosures.
We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, Docket No. 50-315
Enclosure:
ASP Analysis cc w/encl:
See next page John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Hr. E.
E. Fitzpatrick Indiana Hichigan Power Company CC ~
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Marrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 Mest Ottawa Street Lansing, Hichigan 48913 Township Supervisor Lake Township Hall P.O.
Box 818 Bridgman, Hichigan 49106 Al Blind, Site Vice President Donald C. Cook Nuclear Plant 1 Cook Place Bridgman, Hichigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, Hichigan 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
M.
Mashington, DC 20037 Hayor, City of Bridgman Post Office Box 366.
Bridgman, Hichigan 49106 Special Assistant to the Governor Room 1 State Capitol Lansing, Hichigan 48909 Drinking Mater and Radiological Protection Division Hichigan Department of Environmental guality 3423 N. Hartin Luther King Jr Blvd P. 0.
Box 30630 CPH Hailroom Lansing, Hichigan 48909-8130 Donald C. Cook Nuclear Plant Hr. Steve J.
Brewer Indiana Hichigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, Hichigan 49107
LER No. 315/95-011 LER No. 315/95-011 Event
Description:
One safety injection pump unavailable for 6 months Date ofEvent:
September 12, 1995 Plant:
D. C. Cook, Unit 1 Event Summary As the result ofa surveillance test performed while the unit was shut down in Mode 6, personnel determined that the West Centrifugal Charging Pump (CCP) had been inoperable for about 6 months.
The pump was inoperable because a relay calibration had been performed incorrectly 6 months earlier.
The unavailability of the West CCP primarily affects the unit's response to a small-break loss-of-coolant accident (SLOCA) event.
The estimated increase in core damage probability (CDP) for this event (i.e., the importance) is 7.7 x 10 above a base probability ofcore damage (the CDP) for the same period of2.9 x 10'.
Event Description On September 12, 1995, the plant was shut down in Mode 6 when the West CCP was started to perform the emergency core cooling system (ECCS) full flow test surveillance.
The West CCP provides injection flow on the receipt of a safety injection (Sl) signal. After operating at full flow for 7 min, the pump tripped on motor overcurrent.
Personnel determined that the pump tripped because the 1-51-TA8 time over current relay was set incorrectly. It was determined that this relay was last calibrated on March 15, 1995, 180 days before the full-fiowtest. The West CCP was rendered inoperable for the preceding 6 months.
During the event review, the Instrumentation and Control (1&C) technicians involved in calibrating the relays demonstrated the way they typically determine the relay pick-up current.
Because their technique was incorrect, the relays were miscalibrated.
Both I&Ctechnicians involved in the relay calibration were trained and qualified in the D. C. Cook Nuclear Plant relay training program.
However, a significant amount oftime had elapsed between the end ofthe training program and the time the 1-51-TA8 time overcurrent relay on the West CCP breaker was calibrated incorrectly.
Additional Event-Related Information During normal plant operation, both charging pumps (East and West) are configured for their charging function.
One charging pump is sufficient to supply full charging flow and reactor coolant pump seal injection during normal leakage and normal letdown conditions.
A third positive displacement charging pump is available but is not normally used.
On receipt of a valid SI signal, the CCPs operate in the high pressure injection (HPI) mode.
ENCLOSURE
LER No. 315/95-011 D. C. Cook also has a separate SI system.
The system, with two pumps operating in parallel, runs in an intermediate pressure injection mode. The two SI pumps deliver flowfrom the Refueling Water Storage Tank (RWST) at a maximum injection pressure of approximately 1100 psig.
The residual heat removal (RHR) pumps can be aligned for recirculation &om the containment sump to the suction of either the SI pumps or the CCPs.
The licensee indicated that the East CCP had been inoperable for less than 18 h during the 6-month period that the West CCP was unavailable.
Additionally, the emergency diesel generator (EDG) supporting the East CCP was unavailable for less than 50 h during the 6-month period that the West CCP was unavailable.
Modeling Assumptions This event was modeled as a long term (4320 hours0.05 days <br />1.2 hours <br />0.00714 weeks <br />0.00164 months <br />, 180 days x 24 h/day) unavailability of the West CCP.
The event model was broken into three cases based on reported equipment availability.
The first case modeled only the West CCP as being unavailable for 4252 hours0.0492 days <br />1.181 hours <br />0.00703 weeks <br />0.00162 months <br />.
The second case took into account that the opposite train EDG was periodically unavailable for time periods totaling 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> while the West CCP was unavailable.
Finally, the third case accounted for the report that both CCPs were simultaneously unavailable for various maintenance periods totaling 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
Loss of off-site power (LOOP) sequences are prominent in the second case when only one EDG was available.
LOOP probabilities for short-term and long-term off-site power recovery and the probability of a reactor coolant pump (RCP) seal LOCA following a postulated station blackout were developed based on data distributions contained in NUREG 1032, Evaluation ofStation Blackout Accidents at ¹clear Power Plants.
The RCP seal LOCA models were developed as part of the NUREG 1150 PRA efforts.
These probabilities and models are described in Revised LOOP Recovery and PWR Seal LOCA Models, ORNL/NRC/LTR-89/11, August 1989.
The CCPs were subject to common-cause failure during this 6-month period resulting from incorrect maintenance practices.
Because the success criterion in the Integrated Reliability and Risk Analysis System (IRRAS) model assume both CCPs are required for success of the CCP portion of the HPI function in response to either an SLOCA or a steam generator tube rupture (SGTR), no changes were required to model the increased potential for common-cause failure.
Success of one ofthe bvo SI pumps also ensures success of the HPI function in the IRRAS model, independent ofthe success of the CCPs.
This assumption is not as stringent that of the plant Individual Plant Examination, which is that one of two CCPs and one of two SI pumps are required in response to an SLOCA.
The IRRAS response to an SGTR was modified. Previously, a loss of the HPI function lead directly to core damage.
The possibility of lowering RCS prcssure below the steam generator safety valve set point within 30 min was allowed followingthe loss ofHPI capability by adding a basic event PCS-XHE-DEPRES.
Based on the operator burden under a short time constraint, a failure probability of0.1 was assigned to the new basic event, PCS-XHE-DEPRES.
LER No. 315/95-011 Analysis Results Determining the overall increase in the CDP required determining the increase in the CDP for the three different cases and then summing the cases.
The three cases are:
Case 1
Case 2 Case 3 the increase in the CDP due to the long-term unavailability ofthe West CCP (4252 h).
the increase in the CDP from the opposite train EDG being unavailable periodically while the West CCP was unavailable (50 h).
the increase in the CDP due to the time that the CCPs were simultaneously unavailable because ofvarious maintenance activities (18 h).
Combining the probability estimates for the three cases result in an overall increase of 7.7 x 10 in the CDP for the 180-day period. This is above a base probability for core damage (the CDP) for the same period of 2.9 x 10
. Most ofthe increase (56%) is driven by the long-term unavailability of the West CCP (Case 1).
An additional 44% of the increase in CDP is added by Case 2.
The dominant core damage
- sequence, highlighted as sequence number 6 on the event tree in Fig. I, contributes approximately 44% to the combined increase in the CDP estimate for all three modeled cases.
Sequence number 6 involves:
an SLOCA, the successful trip ofthe reactor, the successful operation ofthc AuxiliaryFeedwater (AFW) system, and the failure ofthe HPI system to provide sufficient cooling flow.
The next most dominant sequence involves a LOOP and contributes approximately 13% to the combined increase in the CDP estimate for all three modeled cases.
The nominal CDP over a 6-month period estimated using the Accident Sequence Precursor (ASP) models for D. C. Cook is approximately 2.9 x 10s.
The failed West CCP increased this probability by 28% to 3.7 x 10'. This latter value (3.7 x 10 ') is the conditional core damage probability (CCDP) for the 6-month period in which the West CCP was inoperable.
For most ASP analyses of conditions (equipment failures over a period of time during which postulated initiating events could have occurred), sequences and cut sets associated with the observed failure dominate the CCDP (i.e., the probability of core damage over the unavailability period, given the observed failures).
The increase in CDP because of the failures is, therefore, essentially the same as the CCDP, and the CCDP can be considered a reasonable measure ofthe significance ofthe observed failures. However, for this event, sequences unrelated to the failure of the West CCP dominated the CCDP estimate.
The increase in CDP given the West CCP inoperability, 7.7 x 10, is, therefore, a better measure of the significance of the failure ofthe West CCP.
LER No. 315/95-011 Definitions and probabilities for selected basic events are shown in Table 1.
The conditional probabilities associated with the highest probability sequences for the condition assessment are shown in Table 2.
The sequence logic associated with the sequences listed in Table 2 are given in Table 3. Table 4 lists the system names associated with the dominant sequences for the condition assessment.
Minimal cut sets associated with the dominant sequences for the condition assessment are shown in Table 5.
Acronyms AFW ASP CCDP CCP CDP ECCS EDG HPI HPR INC IRRAS LOCA LOOP PRA RCP RCS RHR RT RWST SGTR SI SLOCA auxiliary feedwater accident sequence precursor conditional core damage probability centrifugal charging pump core damage probability emergency core cooling system emergency diesel generator high pressure injection high pressure recirculation instrumentation and control integrated reliabilityand risk analysis system loss-of-coolant accident loss ofoff-site power probabilistic risk assessment reactor coolant pump reactor coolant system residual heat removal reactor transient refueling water storage tank steam generator tube rupture safety injection small-break loss-of-coolant accident References LER 315/95-011, Rev 0, "West Centrifugal Charging Pump Inoperable due to Inability to Meet Design Basis Requirements for Six Months as a Result ofPersonnel Error During Relay Calibration,"
November 20, 1995.
2.
Indiana Michigan Power Company, Donald C. Cook Nuclear Plant Individual Plant Examination Summary Report.
3.
Indiana Michigan Power Company, Donald C. CookNuclear Plant Final Safety Analysis Report.
LER No. 315/95-011 4.
Evaluation ofStation Blackout Accidents at Nuclear Power Plants, NUREG 1032.
5.
Revised LOOP Recovery and PFR Seal LOC4 Models, ORNL/NRC/LTR-89/11, August 1989.
7 6
0 10 11 12 13 14 15 16 17 18 19 20 21 OK CD OK CD CD CD OK CD CD OK OK CD OK CD OK CD CD CD COOK 1, ASP PWR B SMALLLOCAEVENTTREE
LER No. 315/95-011 Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 315/95-011 Event name CVC-MDP-FC-IA HPI-MDPCF-ALL HPI-MDP-FC-IA HPI-MDP-FC-IB HPI-MOV4C-SUC HPI-MOV~RWST HPI-XHE-NOREC HPR-XHE-NOREC PCS-XHE-DEPRES RHR-MDP4F-ALL RHR-MDP-FC-IA RHR-MDP-FC-IB RHR-MORC-SUC I RHR-MORC-SUC2 RHR-MOV~RWST RHR-XHE-NOREC Description Failure ofCharging Pump A High Prcssure Injection (HPI)
Pump Comme~ause Failures HPI Motor-Driven Pump A Fails HPI Motor-Driven Pump B Fails HPI Serial Component Failures Failure to Isolate the Refueling Water Storage Tank (RWST)
From the HPI System Operator Fails to Rccovcr thc HPI System Operator Fails to Rccovcr the High Prcssure Recirculation (HPR) System Failure to Dcprcssurize the Reactor Coolant System (RCS)
Within30 Minutes Residual Heat Removal (RHR)
Pump Common<ause Failures RHR Motor-Driven Pump IA Fails RHR Motor-Driven Pump IB Fails Failure ofRHR Hot Leg Suction Motor~rated Valve (MOV)A Failure ofRHR Hot Lcg Suction MOVB Failure to Isolate thc RWST During RHR Operator Fails to Rccovcr thc RHR System Base probability 9.0 E404 7.8 E404 3.9 E403 3.9 E403 1.4 E404 3.0 E403 8.4 E401 1.0 Et000 1.0 E401 4.5 E404 4.1 E403 4.1 E403 3.0 E403 3.0 E403 3.0 E403 1.0 E+000 Current probability 1.0 E&00 7.8 E404 3.9 E403 3.9 E403 1.4 E404 3.0 E403 8.4 E401 1.0 E&00 1.0 E-001 4.5 E404 4.1 E403 4.1 E403 3.0 E403 3.0 E403 3.0 E403 1.0 E+000 Type TRUE NEW Modified for this event Ycs No No No No No No No No No No No No No No No
LER No. 315/95-011 Table 2. Sequence Conditional Probabilities for LER No. 315/95-011 Event tree name SLOCA SGTR SLOCA Sequence name 06 08 03 Conditional core damage probability (CCDP) 3.3 E-006 5.4 E-007 2.2 E-008 Core damage probability (CDP) 2.9 E-008 4.8 E-009 2.0 E-006 Importance (CCDP-CDP) 3.3 E-006 5.4 E-007 1.5 E-007 Percent contributiona 77.2 12.5 3.7 Subtotal Case 1
(shown)'ubtotal Case 2 Subtotal Case3'.3 E-005 3.7 E-006 1.4 E-007 2.9 E-005 3.4 E-007 1.2 E-007 4.3 E-006 3.4 E-006 1.8 E-008 Total (all sequences) 3.7 E-005 2.9 E-005 7.7 E-006 Case t represents the increase in the core damage probability due to the long-term unavailability ofthc West CCP (4252 h).
Case 2 rcprcsents the increase in the CDP from thc opposite train EDG being unavailable periodically while thc West CCP was unavailable (50 h).
'ase 3 rcprcscnts the incrcasc in the core damage probability due to the time that the CCPs werc simultaneously unavailable because ofvarious maintenance activities (18 h).
~ Pcrccnt contribution to the total importance.
LER No. 315/95-011 Table 3. Sequence Logic for Dominant Sequences for LER No. 315/95-011 (Case 1 only)
Event tree name Sequence name Logic SLOCA SGTR SLOCA 06 08 03
/RT, /AFW, HPI
/RT, /AFW, HPI, RCS-SG-H
/RT, /AFW,/HPI,
/COOLDOWN, RHR, HPR Table 4. System Names for LER No. 315/95-011 (Case 1 only)
System name AFW COOLDOWN HPI HPR RCS-SG-H RT Logic No or Insufficient AFWFlow RCS Cooldown to RHR Pressure Using Turbine-Bypass Valves, etc.
No or Insufficient Flow From HPI System No or Insufficient HPR Flow No or Insufficient Flow From RHR System Failure to Depressurizc thc RCS Below the Steam Generator Safety Valve Setpoint Without HPI Reactor Fails to Trip During Transient
I ~
~
LER No. 315/95-011 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 315/95-011 Cut set no.
Percent contribution CCDP importance)'ut sets'LOCA Sequence 06 83.1 14.9 1.6 SGTR Sequence 08 83.1 14.9 1.6 SLOCA Sequence 03 86.4 3.2 1.7 1.7 1.7 Subtotal Case 1'shown above)
Subtotal Case 2'ubtotal Case 3'otal (all sequences) 3.3 E-006 2.8 E-006 5.0 E-007 5.4 E-008 5.4 E-007 4.5 E-007 8.0 E-008 8.6 E-009 1.6 E-007 1.5 E-007 6.2 E-009 3.6 E-009 3.6 E-009 3.6 E-009 49 E-006 3.4 E-006 1.8 E-008 7.7 E-006 CVC-MDP-FC-IA,HPI-MDP<F-ALL,HPI-XHF NOREC CVC-MDP-FC-IA,HPI-MOV4C-SUC, HPI-XHE-NOREC CVC-MDP-FC-IA,HPI-MDP-FC-IA,HPI-MDP-FC-IB, HPI-XHF NOREC CVC-MDP-FC-IA,HPI-MDPCF-ALL,HPI-XHF NOREC, PCS-XHE-DEPRES CVC-MDP-FC-IA,HPI-MOV4C-SUC, HPI-XHF NOREC, PCS-XHE-DE PRES CVC-MDP-FC-IA,HPI-MDP-FC-IA,HPI-MDP-FC-IB, HPI-XHE-NOREC, PCS-XHE-DEPRES RHR-MDPCF-ALL,RHR-XHE-NOREC, HPR-XHE-NOREC RHR-MDP-FC-IA, RHR-MDP-FC-IB, RHR-XHE-NOREC, HPR-XHE-NOREC CVC-MDP-FC-IA,HPI-MOV4O-RWST, RHR-MOVE-SUC2, RHR-XHE-NOREC, HPR-XHE-NOREC CVC-MDP-FC-IA,HPI-MOV4O-RWST, RHR-MOV4O-RWST, RHR-XHE-NOREC, HPR-XHE-NOREC CVC-MDP-FC-IA,HPI-MOV4O-RWST, RHR-MOV-CC-SUCI, RHR-XHE-NOREC, HPR-XHE-NOREC 10
LER No. 315/95-011
'hc change in conditional probability (importance) is dctcrmincd by calculating the conditional probability for thc period in which the condition cxistcd, and subtracting thc conditional probability for the same period but with plant cquipmcnt assumed to bc operating nominally. Thc conditional probability for each cut sct within a sequence is dctermincd by multiplying thc probability that the portion of thc sequence that makes the precursor visible (e.g., the system with a failure is dcmandcd) willoccur during the duration ofthe cvcnt by the probabilities ofthe remaining basic events in the minimal cut sct. This can bc approximated by 1 - e~, where p is dctcrmincd by multiplying the expected number of initiators that occur during the duration ofthe event by the probabilities ofthe basic events in that minimal cut sct. The expected number of initiators is given by At, where A is thc frequency of thc initiating event (given on a pcr-hour basis), and t is the duration time of the cvcnt.
This approximation is conservative for prccursors made visible by thc initiating cvcnt.
The frcqucncies ofinterest for this event are:
Ara
~ 5 3 x 10"/h, Atoner 3 8 x 10~/h, AsLec 1.0 x 10~/h, and A sera 1.6 x 10~/h.
b Case 1 represents thc incrcasc in the core damage probability duc to the long term unavailability ofthc West CCP (4252 h).
Case 2 rcprescnts the increase in thc CDP from thc opposite train EDG being unavailable periodically while the West CCP was unavailable (50 h).
Case 3 rcprcsents thc increase in the core damage probability due to thc time that thc CCPs werc simultaneously unavailable because ofvarious maintcnancc activities (18 h).
'asic event, CVC-MDP-FC-IA, is a TRUE type event which is not normally included in the output of fault tree reduction programs.
This cvcnt has bccn added to aid in understanding thc scqucnccs to potential core damage associated with the event.
11