ML17228B484
| ML17228B484 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/03/1996 |
| From: | Wiens L NRC (Affiliation Not Assigned) |
| To: | Plunkett T FLORIDA POWER & LIGHT CO. |
| References | |
| NUDOCS 9605060317 | |
| Download: ML17228B484 (31) | |
Text
Iait ItbCIId
+
0 ss s
0 i
va O
~O
+sI*W+
Nr. Thomas F. Plunkett President, Nuclear Division Florida Power and Light Company Post Office Box 14000 Juno
- Beach, Florida 33408-0420 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 Hay 3, 1996
Dear Nr. Plunkett:
SUBJECT:
DRAFT 1982-83 PRECURSOR REPORT Enclosed for your information are excerpts from the draft Accident Sequence Precursor (ASP) Report for 1982-83.
This report documents the Accident Sequence Precursor (ASP)
Program analyses of operational events which occurred during the period 1982-83.
We are providing the appropriate section of this draft report to each licensee with a plant which had an event in 1982 or 1983 that has been identified as a precursor.
Three of these precursors occurred at St. Lucie, Units 1 and 2.
Also enclosed for your information are copies of Section 2.0 and Appendix A from the 1982-83 ASP Report.
Section 2.0 discusses the ASP Program event selection criteria and the precursor quantification process; Appendix A describes the models used in the analyses.
We emphasize that you are under no licensing obligation to review and comment on the enclosures.
The analyses documented in the draft ASP Report for 1982-83 were performed primarily for historical purposes to obtain the two years of precursor data for the NRC's ASP Program which had previously been missing.
We realize that any review of the precursor analyses of 1982-83 events by affected licensees would necessarily be limited in scope due to: (1) the extent of the licensee's corporate memory about specific details of an event which occurred 13-14 years
- ago, (2) the desire to avoid competition for internal licensee staff resources with other, higher priority work, and (3) extensive changes in plant design, procedures, or operating practices implemented since the time period 1982-83, which may have resulted in significant reductions in the probability of (or, in some cases, even precluded) the occurrence of events such as those documented in this report.
The draft report contains detailed documentation for all precursors with conditional core damage probabilities
> 1.0 x 10 However, the relatively large number of precursors identified for the period 1982-83 necessitated that only summaries be provided for grecursors wit) conditional core damage probabilities between 1.0 x 10 and 1.0 x 10 We will begin revising the report about Hay 31, 1996, to put it in final form for publication.
We will respond to any comments on the precursor analyses which we receive from licensees.
The responses will be placed in a separate mI: I:RjE II:EIIIT>M>>
9605060327 960503 PDR ADOCK 05000335 P
A y ~
1 I
Mr. T. F. Plunkett section of the final report.
Florida Power and Light Company is on distribution for the final report.
Please contact me at (301) 415-1495 if you have any questions regarding this letter.
Any response to this letter on your part is entirely voluntary and does not constitute a licensing requirement.
Sincerely, Original signed by Leonard A.'iens;, Senior Project Manager Project Directorate II-3 Division of Reactor Projects-I/II Office 'of Nuclear Reactor Regulation Docket No. 50-335,:,
and 50-389,,
f f
Enclosures:
1.
Section B.41, "Precursor Analy'sis of 9/2/82 Reactor Trip and Loss of Grid Synchronization Due to shorting of Generator Relay During Testing;"
2.
Section C.43, "Summary of Precursor Analysis of ll/26/82 Inadvertent Safety Injection and Loss of Vital Power Supplies."
3.
Section C.53,"Summary of Precursor Analysis of 7/28/83 Trip with Emergency Diesel B and AFW Pump C Inoperable."
4.
Section 2, "Selection Criteria and quantification.u 5.
Appendix A, nASP Models.n cc w/enclosures:
See next page Distribution Docket File PUBLIC St. Lucie Reading S.
Varga J. Zwolinski ACRS E. Merschoff,, RII OGC DOCUMENT NAME:
G:iASP8283.STD To receive e copy of thle document, indicate ln the box: "C
~ Copy without ettechment/enclosure E
a Copy with ettechment/enclosure "N" ~ No copy OFFICE PD II-3/LA PDII-3/PH P,
I-HAHE BC I ayton LWicns
,EH bd DATE 05/
/96 05/
/96 05/3 /96 OFFICIAL RECORD COPY
V r
ll 1
[
i' 1'e
)
I I
v" Y, k
f,
Hr. T. F. Plunkett Florida Power and Light Company St.
Lucie Plant CC:
Jack Shreve, Public Counsel Office of the Public Counsel c/o The Florida Legislature 111 West Madison Avenue, Room 812 Tallahassee, Florida 32399-1400 Senior Resident Inspector St. Lucie Plant U.S. Nuclear Regulatory Commission 7585 S.
Hwy AlA Jensen
- Beach, Florida 34957 Joe Hyers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 J.
R.
Newman
- Morgan, Lewis
& Bockius 1800 H Street, N.W.
Washington, DC 20036 John T. Butler, Esquire
- Steel, Hector and Davis 4000 Southeast Financial Center Hiami, Florida 33131-2398 Hr. Thomas R.L. Kindred County Administrator St. Lucie County 2300 Virginia Avenue Fort Pierce, Florida 34982 Hr. Charles
- Brinkman, Manager Washington Nuclear Operations ABB Combustion Engineering, Nuclear Power 12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Bill Passetti Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323 H. N. Paduano, Manager Licensing
& Special
- Programs, Florida Power and Light Company P.O.
Box 14000 Juno
- Beach, Florida 33408-0420 W. H. Bohlke, Site Vice President St. Lucie Nuclear Plant P. 0.
Box 128 Ft. Pierce, Florida 34954-0128 J. Scarola Plant General Hanager St.
Lucie Nuclear Plant P.O.
Box 128 Ft. Pierce, Florida 34954-0128 Hr. Kerry Landis U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323-0199 E. J.
Weinkam Licensing Manager St.
Lucie Nuclear Plant P.O.
Box 128 Fort Pierce, Florida 34954-0128
B.41 LKRNo. 335/S2-040 Event
Description:
Reactor Trip and Loss ofGrid Synchronization Due to Shorting OfGenerator Relay During Testing Date ofEvent:
Plant:
September 2, 1982 St. Lucie 1 B.41.1 Summary On September 2, 1982, personnel conducting a generator relay test short circuited it, causing the generator breakers to open and a reactor/turbine trip. Transfer ofthe vital busses to startup power did not occur and the emergency power system was actuated.
The conditional core damage probability estimated for this event is 3.1 x 10~.
B.41.2 Event Description During full power operation, a generator trip relay was briefly shorted while being tested.
This caused the generator breakers to open and a synchronizing inhibittimer to start. By the time the reactor tripped due to a turbine overspeed trip, the timer had cycled so transfer ofthe vital busses to startup power did not occur. The diesel generators started automatically and loaded properly. Offsite power and normal plant status was restored about 28 minutes after the short circuit occurred.
B.41.3 Additional Event-Related Information A similar bus loss was reported in LER 335/79-028.
B.41.4 Modelling Assumptions Since this event, in effect, isolated the plant from offsite power, itwas modeled as a plant centered losswf-offsite power (LOOP). However, this is probably conservative since the event involved a failure to transfer the vital buses.
The analysis was carried out using the BLACKOUTand EVENTEVLcodes.
The followingprobabilities, required as inputs to solve the LOOP event tree in the EVENTEVL code, were also obtained from the BLACKOUTcode:
LER No. 335/82-040 ENCLOSURE 1
B.41-2 Branch SEAL.LOCA Description Probability that an RCP seal LOCAwilloccur.
Probability 4.0 x 10'2 OFFSITE.PWR.REC/
-EP.AND-AFW OFFSITE.P WR.REC/
-SEAL.LOCA Probability offailing to recover offsite power within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> given that EP and AFW are successful.
1.4 x 10'robability offailingto recover offsite power given the occurrence ofa RCP seal LOCA.
Probability offailing to recover offsite power given
'hat there is no RCP seal LOCA.
4.8 x 10'.2x10~
Probability offailing to recover offsite power within6, 9.9 x 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> given that EP is successful but AFW fails.
EP - emergency power AFW - auxiliary feedwater RCP - reactor coolant pump The values in the above table are shown on page B.41.4 as changes to the corresponding branch probabilities.
8.41.5 Analysis Results The conditional core damage probability (CCDP) estimated for this event is 3.1 x 10'. The dominant core damage sequence, shown in Figure. B.41.1, involves the LOOP, successful reactor trip, failure ofemergency power (EP), success ofAFW, a challenge to the pressure operated reliefvalves (PORVs), and failure of the PORVs to reseat.
LER No. 335/82-040
EP PORV CHALL PORV RESEAT OFFSITE RCP SEAL POWER I.OCA RECOV (LONG)
FEED 8
BLEED RECOV SEC SIDE COOUNG RCS COOt:
DOWN END STATE OK OK OK CD OK CD CD OK CD CD OK OK OK CD CD CD OK OK CD CD CD OK OK CD OK CD CD CD OK CD CD OK OK CD OK CD CD CD OK CD CD CD SEO.
NO 201 202 203 204 205 206 207 208 209 210 211 212 213 214 215 218 217 218 219 220 221 222 223 224 225 226 227 228 229 230 231 232 233 234 235 238 237 238 239 240 241 242
) ~
~ i B.41-4 CONDITIONALCORE DAMAGEPROBABILITYCALCULATIONS Event Identifier:
335/82-040 Event
Description:
Reactor trip due to shorted generator relay Event Date:
9/2/82 Plant:
St. Lucie 1
INITIATINGEVENT HON-RECOVERABLE INITIATINGEVENT PROBABILITIES LOOP 2.1E-01 SEQUENCE CONDITIONAL PROBABILITY SUNS End State/Initiator Probability LOOP Total 3.1E-OS 3.1E-05 SEQUENCE CONDITIONAL PROBABILITIES (PROBABILITY ORDER)
Sequence 231 loop -rt(loop) ep -afw/ep porv.chal l/sbo porv.reseat/ep 228 loop -rt(loop) ep -afv/ep porv.chall/sbo -porv.reseat/ep SEAL
.LOCA OFFSITE.PWR.REC/SEAL. LOCA 241 loop -rt(loop) ep afM/ep 216 loop -rt(loop) -ep afe -OFFSITE.PWR.REC/-EP.AHD.AFll feed. bleed
- non-recovery credit for edited case SEQUENCE CONDITIONAL PROBABILITIES (SEQUENCE ORDER)
Sequence End State Prob CD CD CD CD 1.0E-05 9.9E-06 9.1E-06 1.1E-06 End State Prob N Rec**
1.8E-01 1.8E-01 6.4E-02 9.4E-02 H Rec**
216 228 231 24'I loop -rt(loop) -ep afw -OFFSITE.PWR.REC/-EP.AHD.AFlJ feed. bleed loop -rt(loop) ep -afv/ep porv.chall/sbo -porv.reseat/ep SEAL
.LOCA OFFSITE.PAIR.REC/SEAL. LOCA loop -rt(loop) ep -afN/ep porv.chall/sbo porv.reseat/ep loop -rt( loop) ep afM/ep CD CD 1 'E-06 9.9E:06 1 'E-05 9.1E-06 9.4E-02 1.8E-01 1.8E-01 6.4E-02
- non-recovery credit for edited case SEQUENCE NODEL:
BRANCH NODEL:
PROBABILITY FILE:
No Recovery Limit c:iaspimodeislpwrg8283.cmp c: Xaspimodelsisluci e1.82 c: iaspimodels lpwr8283.pro LER No. 335/82-040
B.41-5 BRANCH FREQUENCIES/PROBABILITIES Branch trans loop loca sgtI't rt( loop) afw afw/atws afw/ep mfw porv.chall porv.chall/afw porv.chall/loop porv.chall/sbo porv.reseat porv.reseat/ep srv.reseat(atws) hpi feed. bleed emrg.boration recov.sec.cool recov.sec.cool/offsite.pwr rcs.cooldown rhr CSI'pr ep SEAL. LOCA Branch Model:
1)OF.1 Train 1
Cond Prob:
OFFSITE.PNR.REC/-EP.AND.-AFil Branch Model:
1.OF.1 Train 1
Cond Prob:
OFFSITE.PMR.REC/-EP.AND.AFM Branch Model:
1.OF.1 Train 1
Cond Prob:
OFFSITE.PIIR.REC/SEAL. LOCA
'Branch Model:
1.0F.1 Train 1
Cord Prob:
OFFSITE.PIIR.REC/-SEAL. LOCA Branch Model:
1.OF.1 Train 1
Cond Prob:
sg.iso.and.rcs.cooldown rcs.cool. below.rhr prim.press. limited branch model file
- forced System 7.2E-04 6.7E-05 2.4E-06 1.6E-06 2.8E-04 O.DE+00 3.8E-04 4.3E-03 S.OE-02 1.9E-01 4.0E-02 1.DE+00 1.0E-01 1.DE+00 2.0E-02 2.0E-02 1.0E-01 3.0E-04 2.0E-02 O.OE+00 2.0E-01 3.4E-01 3.0E-03 B.OE-03 4.0E-03 1.5E-04 2.9E-03 4.8E-02
> 4.0E-02 4.8E-02
> 4.0E-02 2.5E-01
> 1.4E-01I 2.5E-01
> 1.4E-01 5.7E-02
> 9.9E-04 5.7E-02
> 9.9E-04 6.0E-01
> 4.8E-01 6.0E-01
> 4.8E-01 1.1E-02
> 2.2E-OS 1.1E-02
> 2.2E-05 1.0E-02 3.0E-03 8.8E-03 Non-Recov 1 'E+00 2.1E-01 5.4E-01 1.DE+00 1.0E-01 1.0E+00 4.5E-01 1.DE+00 3.4E-01 3.4E-01 1.BE+00 1.DE+00 1.DE+00 1.DE+00 1 ~ 1E.02 1.DE+00 1.0E+00 8.9E-01 1.DE+00 1.DE+00 1.DE+00 1.DE+00 1.0E+00 7.0E-02 1.DE+00 1.0E+00 8.9E-01'.DE+00 1.BE+00 1.DE+00 1.DE+00 1.DE+00 1.0E-01 1.DE+00 1.0E+00 Opr Fail 1.0E-02 1.0E-02 1.0E-03 1.0E-03 1.0E-03 3 'E-03 LER No. 335/82-040
C-43 C.42 LER Number 335/82-062 Event
Description:
Inadvertent Safety Injection and Loss ofVital Power Supplies Date ofEvent:
November 26, 1982 Plant:
St. Lucie 1 Summary On November 26, 1982, during fullpower operation, Safety Injection Actuation Signals (SIAS) for channels A and B ofthe Emergency Safety features Actuation System (ESFAS) were actuated due to a trip test switch being incorrectly positioned by maintenance personnel performing a monthly preventative maintenance test.
All appropriate automatic actions occurred, however, the Static Uninterruptablc Power Supply (SUPS) was lost due to an incorrectly set time delay for the 480 volt Emergency Bus Undervoltage relay. The reactor was manually tripped. Thc DGs automatically loaded to provide AC power to thc plant vital loads. Vital power and normal plant status was restored in approximately 45 minutes.
The event was modeled using three diFerent scenarios.
In the baseline case it was assumed that both diesels operated as they did during the actual event. In the second case it was assumed that one diesel failed to start or was unavailable, i.e. only one train of emergency power (EP) was available.
Finally, the case of both diesels failing(both trains ofEP) was considered.
For the second two cases the appropriate numbers oftrains of auxiliaiy feedwater (AFW), feed and bleed, residual heat removal (RHR), containment spray recirculation (CSR), and high pressure recirculation (HPR) werc made unavailable in the event tree branches.
The conditional core damage probabilities (CCDPs) were then calculated for each case using a transient as the potential initiator.
The results of the three cases werc then weighted by multiplying the CCDP by the probability of the corresponding number offailed dicsels.
Summing these values provided a final weighted average of 5.6 x 10~.
Based on the weighted probabilities, the dominant accident sequence consisted of a successful reactor trip, loss ofAFW, loss ofmain feedwater, and failure offeed and bleed.
Summarized Precursors ENCLOSURE 2
C-53 C.52 LER Nos. 389/S3-037 and -039 Event
Description:
Trip with Emergency Diesel B and AFW Pump C Inoperable Date ofEvent:
July 28, 1983 Plant:
St. Lucie 2 Summary On July 28, 1983, with the unit at 0% power, diesel 2B failed to load onto its 4160V bus during a loss-of-offsite (LOOP) power test.
The cause ofthe failure was traced to a broken electrical lug which prevented the output breaker in the diesel generator circuit from closing. The 2A diesel started and loaded normally. Both offsite power sources were available.
The 2B diesel was returned to service within five hours.
This event was reported in LER 389/83-037.
On the same date, after the LOOP test, the 2C auxiliary fecdwater (AFW) turbine driven pump tripped three times during attempts to start it manually. No cause for the pump failure to start could be determined and it was returned to service.
The unit tripped twice on July 26, 1983, two days prior to the diesel generator and AFW pump failures, as well as on July 28.
This event was modeled as a trip on July 28 with one diesel generator (DG) and thc turbine'driven AFW pump unavailable.
These same equipment unavailabilities may have existed during the trips which occurred on July 26.
The conditional core damage probability (CCDP) estimated for this event is 3.7 x 10 The dominant core damage sequence consists of a transient, successful reactor trip, failure of AFW, failure of main feedwater, and failure offeed and bleed.
Summarized Precursors ENCLOSURE 3
2.0 Selection Criteria and Quantification 2.1 Accident Sequence Precursor Selection Criteria The Accident Sequence Precursor (ASP) Program identifies and documents potentially important operational events that have involved portions of core damage sequences and quantifies the core damage probability associated with those sequences.
Identification ofprecursors requires the review ofoperational events for.instances in which plant functions that provide protection against core damage have been chaHenged or compromised. Based on previous experience with reactor plant operational events, it is known that most operational events can be directly or indirectly associated with four initiators: trip [which includes loss of main feedwater (LOFW) within its sequences],
loss'-offsite power (LOOP), small-break loss-of~oolant accident (LOCA),and steam generator tube ruptures (SGTR) (PWRs only). These four initiators are primarily associated with loss ofcore cooling. ASP Program staff members examine licensee event reports (LERs) and other event documentation to determine the impact that operational events have on potential core damage sequences.
2.1.1 Precursors This section describes the steps used to identify events for quantification. Figure 2.1 illustrates this process.
A computerized search ofthe SCSS data base at the Nuclear Operations Analysis Center (NOAC) of the Oak Ridge National Laboratory was conducted to identifyLERs that met minimum selection criteria forprecursois.
This computerized search identified LERs potentially involving failures in plant systems that provide protective functions forthe plant and those potentially involving core damage-related initiating events. Based on a review ofthe 1984-1987 precursor evaluations and all 1990 LERs, this computerized search successfully identifies almost all precursors and the resulting'ubset is approximately one-third to one-half of the total LERs. It should be noted, however, that the computerized search scheme has not been tested on the LER database for the years prior to 1984. Since the LER reporting requirements for 1982-83 were different than for 1984 and later, the possibility exists that some 1982-83 precursor events were not included in the selected subset. Events described in NUREG -0900~ and in issues ofNuclear Safety that potentially impacted core damage sequences were also selected for review.
Those events selected for review by the computerized search ofthe SCSS data base underwent at least two independent reviews by different staff members. The independent reviews of each LER were performed to determine ifthe reported event should be examined in greater detail. This initial review was a bounding review, meant to capture events that in any way appeared to deserve detailed review and to eliminate events that were clearly unimportant. This process involved eliminating events that satisfied predefined criteria for rejection and accepting all others as either potentially significant and requiring analysis, or potentially significant but impractical to analyze. Allevents identified as impractical to analyze at any point in the stuay are documented in Appendix E. Events were also eliminated fmm further review ifthey had littleimpact on core damage sequences or provided littlenew information on the risk impacts ofplant operationforexample, short-term single failures in redundant systems, uncomplicated reactor trips, and LOFW events.
Selection Criteria and Quantification ENCLOSURE 4
~ ~
I
~
g r
If'
~
~
2-2 LERs requiring review Does the event only involve:
~ componcot failure (no toss of redundancy)
~ loss of redundaocy (single system)
~ seismic qualification/design error
~ environmental qualification/design error
~ pre-critical cvcnt
. structural degradation
~ design error discovered by re.analysis
~ bounded by trip or LOFW
~ no appreciable safety system impact
~ shutdown-related event
~ post-core damage Impacts only Reject No an event bc reasonably analyzed by PRA-based models?
No Idendfy as potentially significant but Impraedcal to analyze Pcrfonn detailed review, analysis, and quantification Define impact ofevent In terms of initiator observed and trains of systems unavailable.
ASP models lant drawings.
system descriptions, FSARs. etc.
Modifybranch probabilpdes to reflect cvcnt.
Calculate conditional probability associated with event using modilied event trees.
Does operational event involve:
~ a core damage initiator
.a total loss of a system
~ a loss of redundancy in two or more systems
~ a reactor trip with a dcgradcd midgatiog system No Reject ls conditional probability 2 I&
No Reject based on low probability Yes Document as a precursor Figure 2.1 ASP Analysis Process Selection Criteria and Quantification
2-3 LERs were eliminated from further consideration as precursors ifthey involved, at most, only one of the following:
a component failure with no loss of redundancy, a short-term loss of redundancy in only one system, a seismic design or qualification error, an environmental design or qualification error, a structural degradation, an event that occurred prior to initialcriticality, a design error discovered by reanalysis, an event bounded by a reactor trip or LOFW, an event with no appreciable impact on safety systems, or an event involving only post core-damage impacts.
Events identified for further consideration typically included the following:
unexpected core damage initiators (LOOP, SGTR, and small-break LOCA);
all events in which a reactor trip was demanded and a safety-related component failed; all support system failures, including failures in cooling water systems, instrument air, instrumentation and control, and electric power systems; any event in which two or more failures occurred; any event or operating condition that was not predicted or that proceeded differently from the plant design basis; and any event that, based on the reviewers'xperience, could have resulted in or significantly affected a chain ofevents leading to potential severe core damage.
Events determined to be potentially significant as a result of this initial review were then subjected to a thorough, detailed analysis. This extensive analysis was intended to identify those events considered to be precursors to potential severe core damage accidents, either because of an initiating event, or because of failures that could have aftected the course ofpostulated off-normal events or accidents. These detailed reviews were not limited to the LERs; they also used final safety analysis reports (FSARs) and their amendments, individual plant examinations (IPEs), and other information related to the event ofinterest.
The detailed review of each event considered the immediate impact of an initiating event or the potential impact of the equipment failures or operator errors on readiness of systems in the plant for mitigation of off-normal and accident conditions. In the review of each selected event, three general scenarios (involving both the actual event and postulated additional failures) were considered.
Ifthe event or failure was immediately detectable and occurred while the plant was at power, then the event was evaluated according to the likelihood that itand the ensuing plant response could lead to severe core damage.
Ifthe event or failure had no immediate effect on plant operation (i.e., ifno initiating event occurred), then the review considered whether the plant would require the failed items for mitigation of potential severe core damage sequences should a postulated initiating event occur during the failure period.
Selection Criteria and Quantification
~
~
~
Ifthe event or failure occurred while the plant was not at power, then the event was first assessed to determine whether it impacted at-power or hot shutdown operation. Ifthe event could only occur at cold shutdown or refueling shutdown, or the conditions clearly did not impact at-power operation, then its impact on continued decay heat removal during shutdown was assessed; otherwise it was analyzed as ifthe plant were'at power. (Although no cold shutdown events were analyzed in the present study, some potentially significant shutdown-related events are described in Appendix D).
For each-actual occurrence or postulated initiating event associated with an operational event reported in an LER or multiple LERs, the sequence of operation of various mitigating systems required to prevent core damage was considered. Events were selected and documented as precursors to potential severe core damage accidents (accident sequence precursors) ifthe conditional probability ofsubsequent core damage was at least 1.0 X 10 (see section 2.2). Events of low significance are thus excluded, allowing attention to be focused on the more important events.
This approach is consistent with the approach used to define 1988-1993 precursors, but differs from that of earlier ASP reports, which addressed all events meeting the precursor selection criteria regardless ofconditional core damage probability.
As noted above, 115 operational events with conditional probabilities of subsequent severe core damage a
1.0 X 10~ were identified as accident sequence precursors.
2.1.2 Potentially SigniTicant Shutdown-Related Events No cold shutdown events were analyzed in this study because the lack ofinformation concerning plant status at the time of the event (e.g., systems unavailable, decay heat loads, RCS heat-up rates, etc.) prevented development of models for such events. However, cold shutdown events such as a prolonged loss ofRHR cooling during conditions ofhigh decay heat can be risk significant. Sixteen shutdown-related events which may have potential risk significance are described in Appendix D.
2.1.3 Potentially SigniTicant Events Considered Impractical to Analyze In some cases, events are impractical to analyze due to lack ofinformation or inability to reasonably model within a probabilistic risk assessment (PRA) framework, considering ge level ofdetail typically available in PRA models and the resources available to the ASP Program.
Forty-three events (some involving more than a single LER) identified as potentially significant were considered impractical to analyze. It is thought that such events are capable of impacting core damage sequences.
However, the events usually involve component degradations in which the extent ofthe degradation could not be determined or the impact of the degradation on plant response could not be ascertained.
1 For many events classified as impractical to analyze, an assumption that the affected component or function was unavailable over a 1-year period (as would be done using a bounding analysis) would result in the conclusion that a very significant condition existed. This conclusion would not be supported by the specifics ofthe event as reported in the LER(s) or by the limited engineering evaluation performed in the ASP Pngram.
Descriptions ofevents considered impractical to analyze are provided in Appendix E.
Selection Criteria and Quantification
2-5 2.1.4 Containment-Related Events In addition to accident sequence precursors, events involving loss of containment functions, such as containment cooling, containment spray, containment isolation (direct paths to the environment only), or hydrogen control, identified in the reviews of 1982-83 LERs are documented in Appendix F. It should be noted that the SCSS search algorithm does not specifically search forcontainment related events. These events, ifidentified for other reasons during the search, are then examined and documented.
2.1$ Interesting" Events Other events that provided insight into unusual failure modes with the potential to compromise continued core cooling but that were determined not to be precursors were also identified. These are documented as "interesting".events in'Appendix G.
2.2 Precursor Quantification Quantification of accident sequence precursor significance involves determination of a conditional probability ofsubsequent severe core damage, given the failures observed during an operational event. This is estimated by mapping failures observed during the event onto the ASP models, which depict potential paths to severe core damage, and calculating a conditional probability of core damage through the use of event trees and system models modified to refiect the event. The effect of a precursor on event tree branches is assessed by reviewing the operational event specifics against system design information. Quantification results in a revised probability ofcore damage failure, given the operational event. The conditional probability estimated for each precursor is useful in ranking because itprovides an estimate ofthe measure ofprotection against core damage that remains once the observed failures have occurred. Details ofthe event modeling process and calculational results can be found in Appendix A ofthis report.
The fiequencies and failure probabilities used in the calculations are derived in part from data obtained across the light-water reactor (LWR) population for the 1982-86 time period, even though they are applied to sequences that ate plant-specific in nature.
Because of this, the conditional probabilities determined for each precursor cannot be rigorously associated with the probability ofsevere core damage resulting from the actual event at the specific reactor plant at which it occurred. Appendix Adocuments the accident sequence models used in the 1982-83 precursor analyses, and provides examples of the probability values used in the calculations.
The evaluation ofpiecursors in this report considered equipment and recovery pmcedures believed to have been available at the various plants in the 1982-83 time frame. This includes features addressed in the current (1994) ASP models that were not considered in the analysis of 1984-91 events, and only partially in the analysis of 1992-93 events.
These features include the potential use of the residual heat removal system for long-term decay heat removal followinga small-break LOCA in PWRs, the potential use ofthe reactor core isolation cooling system to supply makeup following a smaH-break LOCA in BWRs, and core damage sequences associated with failure to trip the reactor (this condition was previously designated "ATWS,"and not developed). In addition, the potential long-term recovery ofthe power conversion system forBWR decay heat removal has been addressed in the models.
Selection Criteria and QuantiTication
2-6 Because of these differences in the models, and the need to assume in the analysis of 1982-83 events that equipment reported as failed near the time ofa reactor trip could have impacted post-trip response (equipment response following a reactor trip was required to be reported beginning in 1984), the evaluations for these years may not be directly comparable to the results for other years.
~ Another difference between earlier and the most recent (1994) precursor analyses involves the documentation ofthe significance ofpiecursors involvingunavailable equipment without initiating events. These events are termed unavailabilities in this report, but are also referred to as condition assessments.
The 1994 analyses distinguish a precursor conditional core damage probability (CCDP), which addresses the risk impact of the failed equipment as well as all other nominally functioning equipment during the unavailability period, and an importance measure defined as the difference between the CCDP and the nominal core damage probability (CDP) over the same time period. This importance measure, which estimates the increase in core damage probability because of the failures, was referred to as the CCDP in pre-1994 reports, and was used to rank unavailabilities.
For most unavailabilities that meet the ASP selection criteria, observed failures significantly impact the core damage model. In these cases, there is littledifference between the CCDP and the importance measure.
For some events, however, nominal plant response dominates the risk.
In these cases, the CCDP can be considerably higher than the importance measure.
For 1994 unavailabilities, the CCDP, CDP, and importance are all provided to better characterize the significance of an event. This is facilitated by the computer code used to evaluate 1994 events (the GEM module in SAPHIRE), which reports these three values.
The analyses of 1982-83 events, however, were performed using the event evaluation code (EVENTEVL) used in the assessment of 1984-93 piecursors.
Because this code only reports the importance measure for unavailabilities, that value was used as a measure ofevent significance in this report. In the documentation ofeach unavailability, the importance measure value is referred to as the increase in core damage probability over the period of the unavailability, which is what it represents.
An example of the difference between a conditional probability calculation and an importance calculation is provided in Appendix A.
2.3 Review ofPrecursor Documentation With completion of the initial analyses of the precursors and reviews by team members, this draft report containing the analyses is being transmitted to an NRC contractor, Oak Ridge National Laboratories (ORNL),
for an independent review. The review is intended to (1) provide an independent quality check ofthe analyses, (2) ensure consistency with the ASP analysis guidelines and with other ASP analyses for the same event type,,
and (3) verify the adequacy of the modeling approach and appropriateness of the assumptions used in the analyses. In addition, the draft report is being sent to the pertinent nuclear plant licensees forreview and to the NRC staff for review. Comments received from the licensees within 30 days will be considered during resolution ofcomments received from ORNL and NRC staff.
2.4 Precursor Documentation Format The 1982-83 precursors are documented in Appendices B and C. The at-power events with conditional core damage probabilities (CCDPs) > LO x 10're contained in Appendix B and those with CCDPs between 1.0 x 10'~ and 1.0 x 10~ are summarized in Appendix C. For the events in Appendix B, a description ofthe event Selection Criteria and QuantiTication
~
~
r II
~
c 2-7 is provided with additional information relevant to the assessment ofthe event, the ASP modeling assumptions and approach used in the analysis, and analysis results. The conditional core damage probability calculations are documented and the documentation includes probability summaries for end states, the conditional probabilities for the more important sequences and the branch probabilities used.
A figure indicating the dominant core damage sequence postulated for each event willbe included in the final report. Copies ofthe
~ LERs are not provided with this draft report.
2.5 Potential Sources ofError As with any analytic procedure, the availability ofinformation and modeling assumptions can bias results. In this section, several of these potential sources oferror are addressed.
Evaluation ofonly a subset of 1982-83 LERs. For 1969-1981 and 1984-1987, all LERs reported during the year were evaluated for precursors. For 1988-1994 and for the present ASP study of 1982-83 events, only a subset ofthe LERs were evaluated after a computerized search ofthe SCSS data base. While this subset is thought to include most serious operational events, it is possible that some events that would normally be selected as precursors were missed because they were not included in the subset that resulted from the screening process.
Reports to Congress on Abnormal Occurrences'NUREG-0900 series) and operating experience articles in Nuclear Safety were also reviewed for events that may have been missed by the SCSS computerized screening.
Inherent biases in the selection process.
Although the criteria for identification of an operational event as a precursor are fairly well-defined, the selection of an LER for initial review can be somewhat judgmental. Events selected in the study were more serious than most, so the majority of the LERs selected for detailed review would probably have been selected by other reviewers with experience in LWR systems and their operation. However, some differences would be expected to exist; thus, the selected set ofprecursors should not be considered unique.
Lack ofappropriate event information. The accuracy and completeness of the LERs and other event-related documentation in reflecting pertinent operational information for the 1982-83 events are questionable in some cases. Requitements associated with LER reporting at the time, plus the approach to event reporting practiced at particular plants, could have resulted in variation in the extent of events reported and report details among plants. In addition, only details of the sequence (or partial sequences for failures discovered during testing) that actually occurred are usually provided; details concerning potential alternate sequences ofinterest in this study must often be inferred. Finally, the lack of a requirement at the time to linkplant trip information to reportable events required that certain assumptions be made in the analysis ofcertain kinds of 1982-83 events. Specifically, through use of the "Grey Books" (Licensed Operating Reactors Status Report, NUREG4200)'t was possible to determine that system unavailabilities reported in LERs could have overlapped with plant trips if it was assumed that the component could have been out-of-service for Vi the test/surveillance period associated with that component. However, with the linkbetween trips and events not being described in the LERs, it was often impossible to determine whether or not the component was actually unavailable during the trip or whether it was demanded Selection Criteria and Quantification
2-8 during the trip. Nevertheless, in order to avoid missing any important precursors for the time period, any reported component unavailability which overlapped a plant trip within i/z of the component's test/surveillance period, and which was believed not to have been demanded during the trip, was assumed to be unavailable concur'rent with the trip. (Ifthe component had been demanded and failed, the failure would have been reported; ifithad been demanded and worked successfully, then the failure would have occurred after the trip). Since such assumptions may be conservative, these events are distinguished from the other precursors listed in Tables 3.1 - 3.6. As noted above, these events are termed "windowed" events to indicate that they were analyzed because the potential time window for their unavailability was assumed to have overlapped a plant trip.
Accuracy ofthe ASP models and probability data.
The event trees used in the analysis are plant~lass specific and reflect differences between plants in the eight plant classes that have been defined. The system models are structured to reflect the plant-specific systems, at least to the train level. While major differences between plants are represented in this way, the plant models utilized in the analysis may not adequately reflect all important differerices.
Modeling improvements that address these problems are being pursued in the ASP Program.'ecause ofthe sparseness of system failure events, data from many plants must be combined to estimate the failure probability of a multitrain system or the frequency of low-and moderate-frequency events (such as LOOPs and small-beak LOCAs). Because ofthis, the modeled response for each event willtend toward an average response for the plant class. If systems at the plant at which the event occurred ate better or worse than average (difficultto ascertain without extensive operating experience), the actual conditional probability for an event could be higher or lower than that calculated in the analysis.
5.
Known plant-specific equipment and procedures that can provide additional protection against core damage beyond the plant-class features included in the ASP event tree models were addressed in the 1982-83 precursor analysis for some plants. This information was not uniformlyavailable; much ofit was based on FSAR and IPE documentation available at the time this report was prepaid. As a result, consideration of additional features may not be consistent in precursor analyses ofevents at different plants. However, analyses of multiple events. that occurred at an individual plant or at similar units at the same site have been consistently analyzed.
Dtjftcultyin determining the potential for recovery offailed equipment.
Assignment of recovery credit foran event can have a significant impact on the assessment ofthe event. The approach used to assign recovery credit is described in detail in Appendix A. The actual likelihood offailing to recover from an event at a particular plant during 1982-83 is difficult to assess and may vary substantially from the values currently used in the ASP analyses. This difficultyis demonstrated in the genuine differences in opinion among analysts, operations and maintenance personnel, and others, concerning the likelihood ofrecovering from specific failures (typically observed during testing) within a time period that would prevent core damage followingan actual initiating event.
Assumption ofa 1-month test interval. The core damage probability for piecursors involving Selection Criteria and Quantification
2-9 unavailabilities is calculated on the basis ofthe exposure time associated with the event. For failures discovered during testing, the time period is related to the test interval. A test interval of 1 month was assumed unless another interval was specified in the LER. See reference'1 for a more comprehensive discussion of test interval assumptions.
Selection Criteria and Quantification
Appendix A:
ASP MODELS ASP MODELS fNCLOSURf 5
A-2 A.O ASP Models This appendix describes the methods and models used to estimate the significance of 1982-83 precursors.
The modeling approach is similar to that used to evaluate 1984-91 operational events.
Simplified train-based models are used, in conjunction with a simplified recovery model, to estimate system failure probabilities specific to an operational event.
These probabilities are then used in event tree models that describe core damage sequences relevant to the event. The event trees have been expanded beyond those used in the analysis of 1984-91 events to address features ofthe ASP models used to assess 1994 operational events (Ref. 1) known to have existed in the 1982-83 time period.
A.l Precursor Significance Estimation The ASP program performs retrospective analyses ofoperating experience.
These analyses require that certain methodological assumptions be made in order to estimate the risk significance of an event. Ifone assumes, following an operational event in which core cooling was successful, that components observed failed were "failed"with probability 1.0, and components that functioned successfully were "successful" with probability 1.0, then one can conclude that the risk ofcore damage was zero, and that the only potential sequence was the combination ofevents that occuned. In order to avoid such trivialresults, the status ofcertain components must be considered latent.
In the ASP program, this latency is associated with components that operated successfully these components are considered to have been capable offailing during the operational event.
Quantification ofprecursor significance involves the determination ofa conditional probability ofsubsequent core damage given the failures and other undesirable conditions (such as an initiating event or an unexpected reliefvalve challenge) observed during an operational event. The effect ofa precursor on systems addiessed in the core damage models is assessed by reviewing the operational event specifics against plant design and operating infomiation, and translating the results ofthe review into a revised model for the plant that reflects the observed failures. The precursors's significance is estimated by calculating a conditional probability ofcore damage given the observed failures.
The conditional probability calculated in this way is useful in ranking because itprovides an estimate ofthe measure ofprotection against core damage remaiiung once the observed failures have occurred.
A.l.l Types of Events Analyzed Two differaittypes ofevents aie addressed in precursor quantitative analysis. In the first, an initiating event such as a loss of offsite power (LOOP) or small-break loss of coolant accident (LOCA) occurs as a part of the precursor.
The probability of core damage for this g~ of event is calculated based on the required plant response to the particular initiating event and other failures that may have occurred at the same time. This type ofevent includes the "windowed" events subsetted for the 1982-83 ASP program and discussed in Section 2.2 ofthe main report.
The second type ofevent involves a failure condition that existed over a period oftime during which an initiating event could have, but did not occur. The probability ofcore damage is calculated based on the required plant response to a set ofpostulated initiating events, considering the failures that were observed.
Unlike an initiating event assessment, where a particular initiatingevent is assumed to occur with probability 1.0, each initiating event is assumed to occur with a probability based on the initiating event &equency and the failure duration.
ASP MODELS
A-3 A.1.2 Modification ofSystem Failure Probabilities to Reflect Observed Failures The ASP models used to evaluate 1982-83 operational events describe sequences to core damage in terms of combinations of mitigating systems success and failure following an initiating event.
Each system model represents those combinations oftrain or component failures that willresult in system failure. Failures observed during an operational event must be represented in terms of changes to one or more of the potential failures included in the system models.
Ifa failed component is included in one ofthe trains in the system model, the failure is reflected by setting the probability for the impacted train to I.Q. Redundant train failure probabilities are conditional, which allows potential common cause failures to be addressed. Ifthe observed failure could have occurred in other similar components at the same time, then the system failure probability is increased to represent this. Ifthe failure could not simultaneously occur in other components (for example, ifa component was removed &om service for preventive mainteiiance), then the system failure probability is also revised, but only to reflect the "removal" of the unavailable component &om the model.
Ifa failed component is not specifically included as an event in a model, then the failure is addressed by setting elements impacted by the failure to failed. For example, support systems are not completely developed in the 1982-83 ASP models. A breaker failure that results in the loss ofpower to a group ofcomponents would be represented by setting the elements associated with each component in the group to failed.
Occasionally, a precursor occurs that cannot be modelled by modifying probabilities in existing system models.
In such a case, the model is revised as necessary to address the event, typically by adding events to the system model or by addressing an unusual initiating event through the use ofan additional event tree.
A.1.3 Recovery from Observed Failures The models used to evaluated 1982-83 events address the potential for recovery ofan entire system ifthe system fails.
This is the same approach that was used in the analysis of most precursors through 1991.'n this approach, the potential for recovery is addressed by assigning a recovery action to each system failure and initiating event.
Four classes were used to describe the different types of short-term recovery that could be involved:
'ater precursor analyses utilize Time-Reliability Correlations to estimate the probability of failing to recover a failed system when recovery is dominated by operator action.
ASP MODELS
~
~
E 0
Recovery Class RI R4 Likelihood ofNon-Recovery 0.55 0.10 0.01 Recovery Characteristic Thc failure did not appear to bc recovcrablc in thc rcquircd period, either from thc control room or at the failed equipmcnt.
The failure appeared rccovcrablc in thc rcquircd period at thc failed cquipmcnt, and thc cquipmcnt was acccssiblc; recovery from the control room did not appear possible.
Thc failure appcarcd rccovcrablc in ihc rcquircd period from the control room, but rccovcry was not routine or involved substantial operator burden.
The failure appcarcd rccovcrable in thc rcquircd period from thc control room and was considered routine and procedurally based.
The assignment ofan event to a recovery class is based on engineering judgment, which considers the specifics ofeach operational event and the likelihood ofnot recovering &om the observed failure in a'moderate to high-stress situation followingan initiating event.
Substantial time is usually available to recover a failed residual heat removal (RHR) or BWRpower conversion system (PCS).
For these systems, the nonrecoveiy probabilities listed above are overly conservative.
Data in Refs. 2 and 3 was used to estimate the followirignonrecoveiy probabilities for these systems:
BWRRHR system BWR PCS PWR RHR system
~Setem nonrecove 0.016 (0.054 iffailures involve service water) 0.52 (0.017 for MSIVclosure) 0.057 Itmust be noted that the actual likelihood offailing to recover &om an event at a particular plant is difficultto assess and may vary substantially &om the values listed.
This diKculty is demonstrated in the genuine differaxes in opinion among analysts, operations and maintenance personnel, etc., concerning the likelihood of recovering specific failures (typically observed during testing) within a time period that would prevent core damage followingan actual initiating event.
A.1-4 Conditional Probability Associated with Each Precursor As described earlier in this appendix, the calculation process for each precursor involves a determination of initiators that must be modeled, plus any modifications to system probabilities necessitated by failures observed
'These nonrecovery probabilities are consistent with values specified in M.B. Sattison et al., "Methods Improvements Incorporated into the SAPHIRE ASP Models," Proceedings ofthe U.S. Nuclear Regulatory Commission Twenty-Second Water Reactor Safety Information Meeting, NUREG/CP4140, Vol. 1, April 1995.
eASP MODELS
A-5 in an operational event.
Once the probabilities that reflect the conditions ofthe precursor are established, the sequences leading to core damage are calculated to estimate the conditional probability for the precursor.
This calculational process is summarized in Table A.l.
Several simplified examples that illustrate the basics ofprecursor calculational process follow. It is not the intent of the examples to describe a detailed precursor analysis, but instead to provide a basic understanding ofthe process.
The hypothetical core damage model for these examples, shown in Fig. A.1, consists of initiator I and four systems that provide protection against core damage:
system A, B, C, and D. In Fig. A.l, the up branch represents success and the down branch failure foreach ofthe systems.
Three sequences result in core damage ifcompleted: sequence 3 P /A("/"represents system success) B C], sequence 6 (IA/B C D) and sequence 7 (I AB). In a conventional PRA approach, the &equency ofcore damage would be calculated using the &equency of the initiating event I, A(l), and the failure probabilities for A, B, C, and D [p(A), p(B), p(C), and p(D)].
Assuming L(l)= 0.1 yr'nd p(A(I)= 0.003, p(B)IA)= 0.01, p(C]1) = 0.05, and p(D)IC) = 0.1,'he &equency of core damage is determined by calculating the &equency ofeach ofthe three core damage sequences and adding the &equencies:
0.1 yr' (1- 0.003) x 0.05 x 0.1 (sequence 3)+
0.1 yr' 0.003 x (1 - 0.01) x 0.05 x 0.1 (sequence 6) +
0.1 yr" x 0.003 x 0.01 (sequence 7)
= 4.99 x 10+yr'~ (sequence 3) + 1.49 x 10+ yr
~ (sequence 6) + 3.00 x 10< yr
~ (sequence 7)
=5.03 x 10" yr'n a nominal PRA, sequence 3 would be the dominant core damage sequence.
The ASP program calculates a conditional probability ofcore damage, given an initiatingevent or component failures. This probability is different than the &equency calculated above and cannot be directly compared with it.
Exam le 1
Initi 'vent A sessment.
Assume that a precursor involving initiating event I occurs.
In response to I, systems A, B, and C start and operate correctly and system D is not demanded.
In a precursor initiating event assessment, the ~~ailiti.ofI is set to 1.0. Although systerrm A, B, and C were successful, nominal failure probabilities are assumed.
Since system D was not demanded, a nominal failure probability is assumed forit as well. The conditional probability ofcore damage associated withprecursor I is calculated by summing the conditional probabilities for the three sequences:
1.0 x (1 - 0.003) x 0.05 x 0.1 (sequence 3) +
1.0 x 0.003 x (1 - 0.010) x 0.05 x 0.1 (sequence 6) +
1.0 x 0.003 x 0.01 (sequence 7)
'he notation p(B ) IA)means the probability that B hils, given I occurred and A Med.
ASP MODELS
A-6
= 5.03 x 10 If,instead, B had failed when demanded, its probability would have been set to 1.0. The conditional core damage probability for precursor IB would be calculated as l.p x (1 - 0.003) x 0.05 x 0. 1 (sequence 3) + 1.0 x 0.003 x 1.0 (sequence 7) = 7.99 x 10 3, Since B is failed sequence 6 cannot occur.
le 2. Condition Assessment.
Assume that during a monthly test system B is found to be failed, and that the failure could have occum8 at any time during the month. The best estimate for the duration ofthe failure is one halfofthe test period, or 360 h. To estimate the probability ofinitiating event I during the 360 h period, the yearly &xluencyoflmust be converted to an hourly rate. IfI can only occur at power, and the plant is at power for 70% ofa year, then the frequency for I is estimated to be 0.1 yr'/(8760 h/yr x 0.7) = 1.63 x 10f, as in example 1, B is always demanded followingI, the probability ofI in the 360 h period is the probability that at least one I occurs (since the failure ofB willthen be discovered), or e F3 x fjdggehgggjgg 1
e 1.638.5 x 360 5 85 x lp.3 Using this value for the probability ofI, and setting p(B) = 1.0, the conditional probability ofcore damage for precursor B is calculated by again summing the conditional probabilities for the core damage sequences in Fig.
A.1:.
5.85 x 10 x (1 - 0.003) x 0.05 x 0.1 (sequence 3) + 5.85 x lp 3 x 0.003 x 1.0 (sequence 7)
=4.67 x 10 As before, since B is failed, sequence 6 cannot occur.
The conditional probability is the probability of core damage in the 360 h period, given the failure ofB. Note that the dominant core damage sequence is sequence 3, with a conditional probability of2.92 x 10'. This sequence is unrelated to the failure ofB. The potential failure ofsystems C and D over the 360 h period stilldrive the core damage risk To iuiderstand the significance ofthe failiueofsystem B, another calculation, an importance measure, is required.
The importance measure that is used is equivalent to risk achievement worth on an interval scale (see Ref. 4).
In this calculation, the increase in core damage probability over the 360 h period due to the failure of B is estimated:
p(cd ) B) - p(cd). For this example the value is 4 67 x Ip-s 2 94 x Ip's 1 73 x lps where the second term on the leR side ofthe equation is calculated using the previously developed probability ofI in the 360 h period and nominal failure probabilities forA, B, C, and D.
For most conditions identified as precursors in the ASP program, the importance and the conditional core damage probability are numerically close, and either can be used as a significance measure for the precursor.
- However, for some eventstypically those in which the components that are failed are not the primary mitigating plant features the conditional core dainage probability can be significantly higher than the importance. In such cases, it is important to note that the potential failure of other components, unrelated to the precursor, are still dominating the plant risk ASP MODELS
A-7 The importance measure for unavailabilities (condition assessments) like this example event were previously referred to as a "conditional core damage probability" in annual precursor reports before 1994, instead ofas the increase in core damage probability over the duration ofthe unavailability. Because the computer code used to analyze 1982-83 events is the same as was used for 1984-93 evaluations, the results for 1982-83 conditions are also presented in the computer output in terms of "conditional probability," when in actuality the result is an importance.
A.2 Overview of 1982-83 ASP Models Models used to rank 1982-83 precursors as to significance consist ofsystem-based plantelass event trees and simplified plant-specific system models. These models describe mitigation sequences for the followinginitiating events: a nonspecific reactor trip [which includes loss offeedwater (LOFW) withinthe model], LOOP, small-break LOCA, and steam generator tube rupture [SGTR, pressurized water reactors (PWRs) only].
Plant classes were defined based on the use ofsimilar systems in providing protective functions in response to transients, LOOPs, and small-beak LOCAs. System designs and specific nomenclature may differ among plants included in a particular class; but functionally, they are similar in response.
Plants where certain mitigating systems do not exist, but which are largely analogous in their initiator response, are grouped into the appropriate plant class. ASP plant categorization is described in the followingsection.
The event trees consider two end states:
success (OK), in which core cooling exists, and core damage (CD), in which adequate core cooling is believed not to exist. In the ASP models, core damage is assumed to occur followingcore uncovcry. Itis acknowledged that clad and fuel damage willoccur at later times, depending on the criteria used to definc "damage," and that time may be available to recover core cooling once core uncovcty occurs but before the onset ofcore damage.
However, this potential recovery is not addressed in the models.
Each event tree describes combinations ofsystem failures that willprevent core cooling, and makeup ifrequired, in both the short and long term. Piiinaiy systems designed to provide these functions and alternate systems capable ofalso performing these functions are addressed.
The models used to evaluate 1982-83 events consider both additional systems that can provide core protection and initiating events not included in the plant-class models used in the assessment of 1984-91 events, and only partially included in thc asscssinent of 1992-93 events.
Response to a failure to trip the reactor is now addressed,
~ as is an SGTR in PWRs. In PWRs, the potential use ofthc residual heat removal system followinga small-break LOCA (to avoid sump recirculation) is addressed, as is the potential recovery ofsecondary-side cooling in the long tcimfollowingthc initiationoffeed and bleed. In boiling water reactors (BWRs), the potential use ofreactor core isolation cooling (RCIC) and the control rod drive (CRD) system for makeup ifa single reliefvalve sticks open is addressed, as is the potential long-term recovery ofthe power conversion system (PCS) for decay heat removal in BWRs. These models better reQect the capabilities ofplant systems in preventing core damage.
ASP MODELS
)
t