ML18038B677
| ML18038B677 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/24/1996 |
| From: | Williams J NRC (Affiliation Not Assigned) |
| To: | Kingsley O TENNESSEE VALLEY AUTHORITY |
| References | |
| NUDOCS 9604290038 | |
| Download: ML18038B677 (50) | |
Text
gP,R RECT 0O IVl0
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++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 24, 1996 Mr. Oliver D. Kingsley, Jr.
President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
Dear Mr. Kingsley:
SUBJECT:
DRAFT 1982-83 ACCIDENT SEQUENCE PRECURSOR REPORT Enclosed for your information are excerpts from the draft Accident Sequence Precursor (ASP) Report for 1982-83.
This report documents the ASP Program analyses of operational events which occurred during the period 1982-83.
We are providing the appropriate sections 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.
Two of these precursors occurred at the Browns Ferry Nuclear Plant (BFN).
One event took place at BFN Unit 1 on February 5,
1983.
This event involved a reactor scram with a main steam relief valve and its associated vacuum breaker stuck open, as reported in LER 50-259/83-006 and 50-259/83-007.
The conditional core damage probability for this event is estimated as 4.4E-S.
The second event involves a
BFN Unit 2 scram on November 10, 1983 with high pressure coolant injection inoperable.
This event was reported in LER 50-260/83-074.
The conditional core damage probability for this event is estimated as 3.2E-5.
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.
1 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.
We emphasize that you are under no licensing obligation to review and comment on the enclosures.
9604290038 960424 PDR ADOCK 05000259 S
PDR MC FILE CEIIIYHMPV ~F.i
If
Nr..c. Kingsley, Jry 2
The draft report contains detailed documentation for all precursors with conditional core damage probabilities
~ 1.0E-5.
However, the relatively large number of precursors identified for the period 1982-83 necessitated that only summaries be provided for precursors with conditional core damage probabilities between 1.0E-6 and 1.0E-5.
We will begin revising the report about May 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 section of the final report.
Any response to this letter on your part is entirely voluntary and does not constitute a licensing requirement.
The Tennessee Valley Authority is on distribution for the final report.
Sincerely, Original sigtled by Joseph F. Williams, Project Manager Project Directorate II-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Docket Nos.
50-259, 50-260, and 50-296
Enclosures:
l.
ASP Evaluation of BFN Unit 1 Scram - February 5,
1983 2.
ASP Evaluation of BFN Unit 2 Scram - November 10, 1983 3.
Selection Criteria and guantification 4.
ASP Models cc w/enclosures:
See next page Distribution w enclosures:
Docket File PUBLIC BFN Reading SVarga Jzwolinski OGC ACRS EMerschoff, RII w o enclosures:
POReilly MBoyle DOCUMENT NAME:
G: iBFNNBPB2 03.BFN To receive a copy of this document, Indicate ln the box:
C
~ Copy without attachment/enclosure "E
ee Copy with attachment/enclosure "ff ~ No copy OFFICE PD I I -3/LA C'PD I I-3/PM PDI 1-4/D
~
HAifE BCIayton JMI II Iams DATE 04/
/96 04/7 /96 FHebdon 04, Q/96
, OFFICIAL RECORD COPY
B.8 LER No. 259/83-006 and -007 Event
Description:
Scram, MSRV and its vacuum breaker fail open Date of Event:
February 5, 1983 Plant:
Browns Ferry 1
B.8.1 Summary Unit 1 was operating at approximately 100% power when a reactor scram occurred.
A main steam relief valve (MSRV) was opened manually to control reactor pressure but when operators attempted to close it, they were unable to do so.
The MSRV tailpipe vacuum relief valve failed open during the event, venting steam into the drywell instead of directing it to the suppression pool.
The conditional core damage probability estimated for the event is 4.4 x 10'.
B.8.2 Event Description On February 5, 1983, unit 1 was operating at 100% power when the operator began a surveillance test of the main turbine overspeed trip system.
At that time something, presumably a turbine trip, occurred which caused a reactor scram.
MSRV 1-1-22 was manually opened to control reactor pressure but operators were unable to close it.
At the same time, the vacuum relief valve for the MSRV 1-1-22 tailpipe stuck open.
The main steam relief valves at Browns Ferry sit on stub headers attached to the main steam lines.
Each MSRV's exhaust is routed via a tailpipe to a quencher submerged in the suppression pool. After MSRV operation, steam in the tailpipe willcondense, drawing a vacuum.
This tends to draw a slug of water up from the suppression pool into the tailpipe.
Subsequent operation ofthe associated MSRV could propel this water slug into the quencher, causing damage.
To prevent this, each tailpipe is equipped with two vacuum breakers which open after MSRV operation to limitnegative pressure in the tailpipe.
During the event, the disk in a vacuum breaker for MSRV 1-1-22 became partially separated from its hinge arm and jammed in the open position, allowing steam to vent continuously to the drywell. The peak drywell pressure attained and the specific leak rate were not noted in the Licensee Event Report for this event, but it was noted that leakage into the drywell exceeded a technical specification limitof 5 gpm. Therefore, it may be assumed that steam leakage into the drywell was in excess of 200 cubic feet per minute, possibly substantially in excess.
It is unclear why the drywell pressure did not exceed the 2.45 psig LOCA setpoint during this event.
IER No. 259/83-006 Zh'CLOSURE
<>
q4 ti
B.S.3 Additional Event-Related Information The failure of the MSRV was initially attributed to its pilot valve so, after the unit was shut down, the pilot valve on 1-1-22 was replaced.
The vacuum reliefvalve for the MSRV was found stuck open and was replaced also.
Four other vacuum relief valves were also found to be damaged and were replaced.
Three days after the event described, the unit was restarted.
The intention was to test the repaired MSRV when reactor pressure reached 250 psig.
However, when reactor pressure reached 178 psig, MSRV 1-1-22 opened spontaneously and could not be closed until the reactor was shut down again.
In addition, the movement indicating arm attached to the tailpipe vacuum breaker for MSRV 1-1-22 was apparently bent out of its normal alignment, causing the vacuum breaker to again stick open.
Once again, an unspecified quantity of steam leakage greater than 5 gpm entered the drywell until the unit was shut down.
This time, MSRV 1-1-22 was replaced.
More rigorous inspection of the MSRV found that the pilot valve inlet tube mounting bracket had broken, and a piece of debris had lodged under the seat of the main valve.
The position indicating arms were removed from all vacuum relief valves to prevent further valve failures.
A safety evaluation report provided with licensee event report 259/83-007 indicates that this event, involving a stuck-open SRV relieving directly to the drywell instead of through quenching headers to the suppression pool, may be represented as a LOCA with a break area of approximately 0.15 square feet.
B.8.4 Modeling Assumptions This event was modeled as a LOCA and itwas assumed that the initiating event was non-recoverable.
A break area of 0.15 square feet is defined as an intermediate break in Browns Ferry's analyses.
To represent the medium-break LOCA, the low fiow-rate makeup systems, CRD and RCIC, were set to failed, since they would not be adequate to make up losses.
The second event occurred three days after shutdown, with a reduced decay heat load and at temperature and pressure conditions only slightly above those which would permit alignment of the residual heat removal (RHR) system.
Presumably, given a stuck-open relief valve, reactor pressure was sufficiently reduced to allow resumption ofshutdown cooling within a short time. The second event was therefore considered to be little different from a routine trip during startup, and the event was not analyzed.
B.8.5 Analysis Results The conditional core damage probability estimated for this event is 4.4 x 10'.
The dominant core damage sequences, highlighted on the event tree in Figure B.8.1, involve a failure to trip following the LOCA (this ATWS sequence is not developed in the model), and the observed LOCA, failure of PCS, MFW, HPCI and ADS. Excluding the potentially conservative ATWS sequence, a conditional core-damage probability of 8.6 x 10 is estimated.
LER No. 259/83-006
Smal LOCA Rx SHUT-DOWN HPCI HPCS RCIC SRVsl ADS CRD PUMPS (INJ)
STATE NO.
OK 301 OK 302 Ce 303 OK 304 CD 305 OK 306 CD 307 OK 308 CD 309 OK 310 CD 311 OK 312 CD 313 OK 314 CD 315 CD 316 OK 317 CD 318 CD 319 CD 320
B.8-4 CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIOHS Event Identifier:
259/83-006 Event
Description:
February 5, 1983 Plant:
Browns Ferry 1
INITIATING EVENT NON-RECOVERABLE INITIATING EVENT PROBABILITIES LOCA 1.DE+00 SEQUENCE CONDITIONAL PROBABILITY SUMS End State/Initiator CD LOCA Total Probability 4.4E-05 4.4E-05 SEQUENCE COHDITIOHAL PROBABILITIES (PROBABILITY ORDER)
Sequence 320 LOCA rx.shutdown 319 LOCA -rx.shutdown pcs mfw hpci RCIC srv.ads CRD(INJ) 303 LOCA -rx.shutdown pcs
-mfw rhr
- non-recovery credit for edited case SEQUENCE CONDITIONAL PROBABILITIES (SEQUENCE ORDER)
End State CO CD CD Prob N Rec**
3.5E-05 1.0E-01 6.8E-06 1.7E-01 1.8E-06 1.4E-02 Sequence 303 LOCA -rx.shutdown pcs
-mfw rhr 319 LOCA -rx.shutdown pcs mfw hpci RCIC srv.ads CRD(IHJ) 320 LOCA rx.shutdown End State CD CD CD Prob 1.8E-06 6.8E-06 3 'E-05 H Rec**
1.4E-02 1 'E-01 1.0E-01
- non-recovery credit for edited case SEQUEHCE MODEL:
d:iaspimodelslbwrc8283.cmp BRANCH MODEL:
d:iaspimodelslbrown1.82 PROBABILITY FILE:
d:iaspimodeislbwr8283.pro Ho Recovery Limit BRANCH FREQUENCIES/PROBABILITIES Branch trans loop LOCA Branch Model:
INITOR Initiator Freq:
System 1.7E-03 1.6E-05 3.3E-06 ) 3.3E-06 3.3E-06 Hon-Recov 1 'E+00 2.4E-O'I 6.7E-01 1 AL OE+00 Opr Fail LER No. 259/83-006
B.S-5 rx.shutdown pcs srv.ftc.<2 USE-04 1.7E-01 1.DE+00 1.0E-01 1.DE+00 1.DE+00 Event Identifier: 259/83-006 srv.ftc.2 srv.ftc.>2 mfw hpci RCIC Branch Model:
1.0F.1 Train 1
Cond Prob:
srv.ads CRD( IHJ)
Branch Model:
1.0F.1+opr Train 1
Cond Prob:
cond lpcs lpci rhrsw(inj) rhr rhr.and.pcs.nrec rhr/-lpci rhr/lpci rhr(spcool )
rhr(spcool)/- lpci ep ep.rec rpt slcs ads.inhibit man.depress branch model file
- forced 1.3E-03 2.2E-04 4.6E-01 2.9E-02 6.0E-02
> 1.DE+00 6.0E-02
> 1.DE+00 3.7E-03 1.0E-02
> 1.DE+00 1.0E-02
> 1.DE+00 1.DE+00 1.7E-03 1.1E-03 2.0E-02 1.5E-04 1.5E-04 O.DE+00 1 'E+00 2.1E-03 2.0E-03 7.5E-03 1.4E-01 1.9E-02 2.0E-03 O.DE+00 3.7E.03 1.DE+00 1.DE+00 3.4E-01 7.0E-01 7.0E-01
> 1.DE+00 7.0E-01 1.DE+00 3.4E-01 1.DE+00 1 'E+00 1.DE+00 1.6E-02 8.3E-03 1.DE+00 1.DE+00 1.DE+00 1.0E+00 8.7E-01 1.DE+00 1.DE+00 1.DE+00 1.DE+00 1.DE+00 1.0E-02 1.0E-02 1 'E.03 1.0E-02 1.0E-05 1.0E-05 1.0E-05 1.0E-05 1 'E-03 1.0E-03 1.0E-02 1.0E-02 1.0E-02 Dolan 01-04-1996 17:20:13 Event Identifier: 259/83-006 LER No. 259/S3-006
h 4
B.9 LER No. 260/83-074 Event
Description:
Trip with HPCI inoperable Date of Event:
November'10, 1983 Plant:
Browns Ferry 2 B.9.1 Summary Unit 2 was operating at approximately 98% power when a reactor scram occurred.
Reactor vessel level dropped sufficiently to provide an auto-initiation signal to the High Pressure Coolant Injection (HPCI) system.
HPCI started and immediately isolated when a turbine exhaust rupture diaphram ruptured, rendering HPCI inoperable.
The conditional core damage probability estimated for the event is 3.2 x 10'.
B.9.2 Event Description On November 10, 1983, while operating at essentially full power, unit 2 experienced a scram.
Reactor vessel level dropped sufficiently to result in HPCI auto-initiation, however HPCI immediately isolated when its turbine exhaust rupture diaphragm ruptured.
The cause of the failure was not determined with certainty, but an exhaust diaphragm rupture which occurred during testing five days earlier had been attributed to inadequate draining of condensate from HPCI steam lines. Apparently, the November 5 rupture disk failure may have been caused by the impact of a slug of water which accelerated in the steam exhaust line after the turbine started.
Testing conducted later, in February of 1984, suggested that improper HPCI control system behavior could lead to exhaust line pressure fluctuations, perhaps great enough to cause failure of the rupture disk.
Adjustments were made to the control system to minimize these fluctuations.
B.9.3 Additional Event-Related Information High-pressure makeup sources at Browns Ferry include the turbine4riven main feedwater pumps, HPCI, the reactor core isolation cooling system (RCIC) and the control rod drive (CRD) pumps.
For events involving isolation of the reactor vessel, only HPCI can provide high flow-rate (5,000 gpm) makeup to the reactor.
B.9.4 Modeling Assumptions This event was modeled as a scram with HPCI assumed unavailable and not recoverable.
Because the HPCI auto-initiation reported indicates that reactor vessel level had dropped to -51.5 inches below instrument zero, it can be assumed that the main steam isolation valves (MSIVs) isolated, causing an initial loss of main feedwater and power conversion systems.
The nonrecovery probability for PCS was revised to 0. 11 to reflect initial assumed closure of MSIVs.
LER No. 260/83-074 RlCLOSURE 2
B.9-2 B.9.5 Analysis Results The conditional core damage probability estimated for this event is 3.2 x 10~.
The dominant core damage
- sequence, highlighted on the event tree in Figure B.9.1, involves the observed trip, failure of the power conversion system, failure of two SRVs to close, unvailability of HPCI, and failure of ADS.
LER No. 260/83-074
TRANS Rx SHUT.
DOWII SAYs CLOSE HPCI HPCS SRVal ADS CRD PUMPS (INJ)
END STATE SEO.
2 vs 0
>2 vtvs Opt0 OK OK CD OK CD OK CD OK CD OK CO OK CD OK CD CO OK CD CD OK CD OK CD OK CD OK CD CD GK CD OK CD OK OK CD CO CO OK CD OK CD OK CD OK CD CD ATWS 101 102 103 104
$08 108 107 108 109
$ 10ill 112 113 114 118 118 117 118
$ 19 120 121 122 123 124
$ 28 128 127 128 129 130 131
$32
$ 33 134
$35 138 137 138 139 140 141 142 143 144
$45 148 142
~ '
B.9-4 COHDITIOHAL CORE DAHAGE PROBABILITY CALCULATIONS Event Identifier:
260/83-074 Event
Description:
Scram with HPCI inop Event Date:
November 10, 1983 Plant:
Browns Ferry 2 INITIATING EVENT NOH-RECOVERABLE INITIATING EVENT PROBABILITIES TRANS 1.DE+00 CD End State/Initiator TRANS Total SEQUENCE CONDITIONAL PROBABILITY SUHS Probability 3 'E-05 3.2E-05 SEQUENCE CONDITIONAL PROBABILITIES (PROBABILITY ORDER)
Sequence End State Prob N Rec**
138 trans -rx.shutdown 103 trans -rx.shutdown 119 trans -rx.shutdown rd(inj) 107 trans -rx.shutdown S.HREC 414
~
trans rx.shutdown PCS srv.ftc.2 HPCI srv.ads PCS srv.ftc.<2
-HFW RHR.AND.PCS.NREC PCS srv.ftc.<2 HFW HPCI rcic srv.ads c
CO CD
'pt PCS srv.ftc.<2 HFW HPCI -rcic RHR.AND ~ PC CD 1.6E-05 6.8E-06 3.6E-06 3.3E-06 6.7E-07 7.0E-01 1 'E-03 1.7E-01 6.0E-04 1.0E-01
- non-recovery credit for edited case SEQUENCE CONDITIOHAL PROBABILITIES (SEQUEHCE ORDER)
Sequence 103 trans -rx.shutdown PCS srv.ftc.<2
-HFW RHR.AND.PCS.NREC 107 trans -rx.shutdown PCS srv.ftc.<2 HFW HPCI -rcic RHR.AHO.PC S.HREC 119 trans -rx.shutdown PCS srv.ftc.<2 HFW HPCI rcic srv.ads c
rd(inj) 138 trans -rx.shutdown PCS srv.ftc.2 HPCI srv.ads 414 trans rx.shutdown rpt
- non-recovery credit for edited case End State CD CD CD CO Prob 6.8E-06 3.3E-06 3.6E-06 1.6E-05 6.7E-07 N Roc**
1.2E-03 6.0E-04 1.7E-01 7.0E-01 1.0E-01 SEQUENCE HOOEL:
BRANCH HOOEL:
PROBABILITY FILE:
No Recovery Limit d:Kaspimodelslbwrc8283.cmp d:iaspimodelslbrown2.82 d:iaspimodelslbwr8283.pro LER No. 260/S3-074
B.9-5 BRANCH FREQUENCIES/PROBABILITIES Branch Event Identifier: 260/83-074 System Non-Recov Opr Fail trans loop loca rx.shutdown PCS Branch Model:
Train 1
Cond srv.ftc.<2 srv.ftc.2 srv.ftc.>2 HFII Branch Hodel:
Train 1
Cond HPCI Branch Hodel:
Train 1
Cond rci c sl'v.ads crd(inj) cond lpcs lpci rhrsw(inj) rhr RHR.ANO ~ PCS.NREC Branch Hodel:
Train 1
Cond Train 2 Cond Train 3 Cond Train 4 Cond rhr/- lpci rhr/lpci rhr(spcool) rhr(spcool)/- lpci ep ep.rec rpt slcs ads.inhibit man.depress 1.0F ~ 1 Prob:
1.OF.1 Prob:
1.0F. 1 Prob:
1.0F.4+opr Prob:
Prob:
Prob:
Prob:
2'E-03 1.6E-05 3.3E-06 3.5E-04 1.7E-01 1.7E-01 1.0E+00 1.3E-03 2.2E-04 4.6E-01 1 AL OE+00
> 1.0E+00
> 1.0E+00 2.9E-02 6.0E-02 3.7E-03 1.0E-02 1 AL OE+00 1.7E-03
- 1. 1E-03 2.0E-02 1.5E-04 USE-04 1.0E-02 1.0E-01 3.0E-01 5.0E-01 O.OE+00 1.0E+00 2.'1E-03 2.0E-03 7.5E-03 1.4E-01 1.9E-02 2.0E-03 O.OE+00 3.7E-03
> 1.0E+00
> 1.5E-04 4.6E-01
> 1.0E+00 2.9E-02
> 1.0E+00 1.0E+00 2.4E-01 6.7E-01 1.0E-01 1.0E+00 1.0E+00 1 AL OE+00 1.0E+00 3.4E-01 7.0E-01
> 1.0E+00 7.0E-01 7.0E-01 1.0E+00 3.4E-01 1.0E+00 1.0E+00 1.0E+00 1.6E-02 8.3E-03
> 1.8E-03 1.0E+00 1.0E+00 1.0E+00 1 AL OE+00 8.7E-01 1.0E+00 1.0E+00 1 AL OE+00 1.0E+00 1.0E+00 1.0E-02 1.0E-02 1.0E-03 1.0E-02 1.0E-OS 1.0E-05 1.0E-05 1.0E-OS 1.0E-03 1.0E-03 1.0E-02 1 'E-02 1.0E-02 branch model file
~* forced Oolan 01-10-1996 18:54:55 Event Identifier: 260/83-074 LER No. 260/83-074
2-1 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 challenged 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-of-offsite power (LOOP), small-break loss-of-coolant accident (LOCA),and steam generator tube ruptures (SGTR) (PWRs only). These four initiators are primarily associated with loss of core 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 of the SCSS data base at the Nuclear Operations Analysis Center (NOAC) of the Oak Ridge National Laboratory was conducted to identify LERs that met minimum selection criteria forprecursors.
This computerized search identified LERs potentially involving failures in plant systems that provide protective functions for the 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 subset 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 of the 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 study are documented in Appendix E. Events were also eliminated from further review ifthey had little impact on core damage sequences or provided littlenew information on the risk impacts ofplant operationfor example.
short-term single failures in redundant systems, uncomplicated reactor trips, and LOFW events.
Selection Criteria and Quantification P"CLOSURE 3
2-2 LERs requiring rcvicw Does the event only involve:
~ component failure (no loss of redundancy)
~ loss of redundancy (single system)
~ seismic qualiflcation/design error
~ environmental qualiflication/design error
~ pre.critical event
~ structural degradation
~ design error discovered by re.analysis
~ bounded by trip or LOFW
~ no appreciable safety system impact
~ shutdown related evem
~ post core damage impacts only No Can event be reasonably analyzed by PRA based models?
Ycs Ycs No Reject identify as potentially significant but impractical to analyze Perform detailed review. analysis. and quantification Define impact ofevent in terms of initiator observed and trains of systems unavailable.
Modifybranch probabilities to reflect event.
ASP models lant drawings.
system descriptions.
FSARs. etc.
Calculate conditional probabiliry associated with event using modifled event uees.
Docs operational event involve:
~ a core damage inhiator
. a total loss of a system
~ a loss of redundancy in two or morc systems
~ a reactor trip with a degraded mitigating system No Reject Yes Is conditional probability g IO+
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 typicaily 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 of events 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 affected 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 of interest.
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 it and 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
'lg '
2-4 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 of subsequent 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 Significant Shutdown-Related Events No cold shutdown events were analyzed in this study because the lack of information 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 of RHR cooling during conditions of high 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 of information or inability to reasonably model within a probabilistic risk assessment (PRA) framework, considering the 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.
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 Program.
Descriptions of events considered impractical to analyze are provided in Appendix E.
Selection Criteria and Quantification
il,
% 1
- 4
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.
2.1.5 "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 reflect 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 it provides an estimate of the 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 frequencies 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 are plant-specific in nature.
Because of this, the conditional probabilities determined for each precursor cannot be rigorously associated with the probability of severe core damage resulting from the actual event at the specific reactor plant at which it occurred. Appendix A documents, the accident sequence models used in the 1982-83 precursor analyses, and provides examples of the probability values used in the calculations.
The evaluation of precursors in this report considered equipment and recovery procedures 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 of the reactor core isolation cooling system to supply makeup following a small-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 Quantification
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 of a 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 of the significance of precursors involving unavailable 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 1;igher 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 precursors.
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 of the 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 for review and to the NRC staff for review. Comments received from the licensees within 30 days will be considered during resolution of comments 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) >1.0 x 10're contained in Appendix B and those with CCDPs between 1.0 x 10'nd 1.0 x 10~ are summarized in Appendix C. For the events in Appendix B, a description of the event Selection Criteria and Quantification
2-7 is provided with additional information relevant to the assessment of the 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 of the
~ 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 of the LERs were evaluated after a computerized search of the 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 Abnornial 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 of precursors 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. Requirements 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 of interest 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 of certain kinds of 1982-83 events. Specifically, through use of the "Grey Books" (Licensed Operating Reactors Status Report, NUREG-0200)'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 iA 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 QuantiTication
Q
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/2 of the component's test/surveillance period, and which was believed not to have been demanded during the trip, was assumed to be unavailable concurrent with the trip. (Ifthe component had been demanded and failed, the failure would have been reported; ifit had 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 probabiliry data.
The event trees used in the analysis are plant-class 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 differences.
Modeling improvements that address these problems are being pursued in the ASP Program.
Because of the sparseness ofsystem 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-break LOCAs). Because of this, 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 are 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.
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 uniformly available; much ofit was based on FSAR and IPE documentation available at the time this report was prepared. 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.
Difficultyin determining the potential for recovery offailed equipment.
Assignment of recovery credit for an event can have a significant impact on the assessment of the event. The approach used to assign recovery credit is described in detail in Appendix A. The actual likelihood of failing 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 of recovering 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 precursors 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 l for a more comprehensive discussion of test interval assumptions.
Selection Criteria and Quantification
Appendix A:
ASP MODELS ASP MODELS ENCLOSURE 4
,i J
"~l yC
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 occumxi. In order to avoid such trivial.results, 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 addressed in the core damage models is assessed by reviewing the operational event specifics against plant design and operating information, 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 different types ofevents are 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 type 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 respoiise to a set ofpostulated initiating events, considering the failures that were observed.
Unlike an initiating event assessment, where a particular initiatingevent is assuined to occur withprobability 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 >s-
~
A-3 A.l.2 Modification ofSystem Failure Probabilities to Refiect 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 ofchanges 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 refiected by setting the probability for the impacted train to 1.0. Redundant train failure probabilities are conditional, which allows potential common cause failure to be addressed. Ifthe observed failure could have occurred in other similar compoiients 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 maintaiance), then the syste: m 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 system-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 recoveiy action to each system failure and initiating event.
Four classes were used to describe the diQerent types ofshort-tarn 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
l r
A-4 Recovery Class RI R4 Likelihood ofNon-Recovery'.55 0.10 0.01 Recovery Characteristic Thc failure did not appear to bc rccovcrablc in the required period, either from thc control room or at thc failed equipment.
Thc failure appeared rccovcrablc in thc rcquircd period at thc failed equipment, and the equipmcnt was accessible; recovery from thc control room did not appear possible.
The failure appcarcd rccoverablc in thc required period from thc control room, but rccovcry was not routine or involved substantial operator burden.
Thc failure appeared recoverablc in thc required period from thc control room and was considcrcd routine and procedurally based.
The assigtunent 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 nonrecovery probabilities listed above are overly conservative.
Data in Refs. 2 and 3 was used to estimate the followingnonrecovery probabilities for these systems:
BWR RHR system BWR PCS PWR RHR system
~Setem nonrecove 0.016 (0.054 iffailures involve service water) 0.52 (0.017 forMSIVclosure) 0.057 It must be noted that thc actual likelihood offailing to recover &om an event at a particular plant is diQicult to assess and may vary substaatially &om the values listecL This difIicultyis demonstrated in the genuine difference ia opinion amoag analysts, operations and maintenance personael, etc., concerning the likelihoodof recovering speci6c failures (typically observed during testing) within a time period that would prevent core damage foHowing an 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 aecessitated by failures observed
'Ihese 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 banty-Second Water Reactor Safety Information Meeting, NUREG/CP-0140, Vol. 1, April 1995.
ASP MODELS
UW
A-5 in an operational event.
Once the probabilities that reflect the conditions of the 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 simplifled 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.l, 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 tepresents success and the down branch failure for each ofthe systems.
Three sequence 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, X(1), and the failure probabilities for A, B, C, and D jp(A), p(B), p(C), and p(D)].
Assutning L(I)= 0.1 yt'nd p(Ajl)= 0.003, p(BjIA)= 0.01, p(Cjl) = 0.05, and p(DjIC)= 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) ld 0.05 x 0.1 (sequence 3) +
0.1 yr'f 0.003 ld (1- 0.01) tf 0.05 x 0.1 (sequence 6)+
0.1 yr'f 0.003
>< 0.01 (sequence 7)
= 4.99 hf 10 yr'sequence 3)+ 1.49 ~ 10 yr'sequence 6)+3.00 hf 10~ yr'sequence 7)
= 5.03 ld 10" yt'n a nominal PRA, sequence 3 would be the dominant core damage sequence.
The ASP program calculates a conditional probability ofcore damage, given an initiating event or component failures. This pmbability is different than the &equency calcuiated above and cannot be directly compared with it.
Exam le 1
Initiatin Event A sment.
Assume that a precursor involving initiating event I occurs.
In response to I, systetns A, B, and C start and operate correctly and system D is not demanded.
In a precursor initiating event assessmcnt, the gmmbabili g oft is set to 10. Although systctns 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 weH. 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
>c 0.1 (sequence 3)+
1.0 hf 0.003 ld (1- 0.010) hf 0.05 x 0.1 (sequence 6)+
1.0 hf 0.003 x 0.01 (sequence 7)
~ The notation p(B jIA)means the probability that B Ms, given I occurred and A Med.
ASP MODELS
~A
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 1.0 x (1 - 0.003) x 0.05 x 0.1 (sequence 3) + 1.0 x 0.003 ~ 1.0 (sequence 7) = 7.99 x 10'.
Since B is failed sequence 6 cannot occur.
Exam le 2.
ondition Assessment.
Assume that during a monthly test system B is found to be failed, and that the failure could have occur@xi 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 ofinitiatingevent Iduring the 360 h period, the yearly Gequency ofImust be converted to an hourly rate. IfI can only occur at power, and the plant is at power for 70% ofa year, then the &equency for I is estimated to be 0.1 yr'/(8760 h/yr < 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 e40) ~ failure dursuca 1
e.1.638-$
~ 360 5 85
>< 10.3 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 I0< x (1 0.003) x 0.05 ~ 0.1 (sequence 3) + 5 85 x IQ ~ x 0.003 x 1.Q (sequence 7) 467 x 10-s 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 donntumt 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 undersbmd the significance ofthe failure ofsystem B, another calculation, an importance measure, is required.
The importance manure 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 10 - 2.94 x 10' 1.73 x 10', where the second term on the left 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 events typically those in which the components that are failed are not the primary mitigating plant features the conditional core datnage 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 temis of "conditional probability," when in actuality the result is an iillpoftailce.
A.2 Overview of 19S2-S3 ASP Models Models used to rank 1982-83 precursors as to significance consist ofsystem-based plant-class 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) within the 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-break LOCAs. System designs and specific nomenclature may differamong 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 uncovery. Itis acknowledged that clad and fuel damage willoccur at later times, depending on the criteria used to define "damage," and that time may be available to recover core cooling once core uncoveiy occurs but before the onset ofcore damage.
However, this potential recovay 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. Primary 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 plantwliss models used in the assessment of 1984-91 events, and only partially mciuded in the assessment of 1992-93 events.
Response to a failure to trip the reactor is now addressed,
. as is an SGTR in PWRs. h PWRs, the potential use ofthe 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 term followingthe mitiationoffeed and bleed. In boilingwater 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 systeins in preventing core damage.
ASP MODELS
.Dg
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 inciease in core damage probability over the duration ofthe unavailability. Because the computer code used to amilyze 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 inlpoffallce.
A.2 Overview of 1982-S3 ASP Models Models used to rank 1982-83 precursors as to significance consist ofsystem-based plantclass event trees and simplifiedplant-spccific system models. These models describe mitigation scquenccs for the followinginitiating events: a nonspecific reactor trip [which includes loss offecdwater (LOFW)withinthe model], LOOP, small-break LOCA, and stcam 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-break 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 systans 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.
'Hm 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 thc ASP models, core damage is assumed to occur followingcae uncovciy. Itis admowledged that clad and fuel damage willoccur at later times, depending on the criteria used to definc "dainage," and that time may be available to recover core cooling once core uncovery occurs but bcfae tbe oaset ofctxe damigc.
However, this potential recovery is not addressed in thc models.
Each event tice describes combinations ofsystem Mures that willprevent core cooling, and makeup ifrequired, in both the short and long tern. Mmaiy systcins dcsigncd to provide these functions and alternate systcins capable ofalso performing these fimctions 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 plantwlass models used in the assessment of 1984-91 events, and only partially mchded intbe asscssmatof 1992-93 events. Rcslxxise to a failure to trip the reactor is now addressed,
~ as is an SGTRin PWRs. In PWRs, the potential use ofthe residual heat removal system followinga sman-break LOCA(to avoid sump recirculation) is addressed, as is the potential rccovay ofsccondmy-side cooling in the kxigtean foGowing the initiate offeed and bleed. Inboilmgwater reactors (BWRs), the potential use ofreactor cae isolation cooling (RCIC) and the control rod drive (CRD) system formal+up ifa single reliefvalve sticks open is addressed, as is the potential long-tean recovery ofthc power conversion system (PCS) fordecay heat removal in BWRs. These models better rcQcct the capabilities ofplant systctns in preventing core damage.
ASP MODELS
V
Mr. Oliver D. Kingsley, Jr.
Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT CC:
Hr. 0. J. Eeringue, Sr. Vice President Nuclear Operations Tennessee Valley Authority 3B Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Hr. Mark 0. Medford, Vice President Engineering 8 Technical Services Tennessee Valley Authority 3B Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Hr.
R.
D. Hachon, Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O.
Box 2000
- Decatur, AL 35609 General Counsel Tennessee Valley Authority ET 10H 400 West Summit Hill Drive Knoxville, TN 37902 Hr. P.
P. Carier, Manager Corporate Licensing Tennessee Valley Authority 4G Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Hr. T.
D. Shriver Nuclear Assurance and Licensing Browns Ferry Nuclear Plant Tennessee Valley Authority P.O.
Box 2000
- Decatur, AL 35609 Hr. Pedro Salas Site Licensing Manager Browns Ferry Nuclear Plant Tennessee Valley Authority P.O.
Box 2000
- Decatur, AL 35609 TVA Representative Tennessee Valley Authority 11921 Rockville Pike, Suite 402 Rockville, HD 20852 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW., Suite 2900
- Atlanta, GA 30323 Hr. Leonard D. Wert Senior Resident Inspector Browns Ferry Nuclear Plant U.S. Nuclear Regulatory Commission 10833 Shaw Road
- Athens, AL 35611 Chairman Limestone County Commission 310 West Washington Street
- Athens, AL 35611 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, AL 36130-1701
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