ML17229A711
| ML17229A711 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/23/1998 |
| From: | Gleaves W NRC (Affiliation Not Assigned) |
| To: | Plunkett T FLORIDA POWER & LIGHT CO. |
| References | |
| NUDOCS 9805060031 | |
| Download: ML17229A711 (17) | |
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0 UNITED STATES NUCLEAR REGULATORY COMM1SS1ON WASH)NGTON, D.C. 20555-0001 April 23, 1998 Mr. T. F. Piunkett President - Nuclear Division Florida Power and I ight Company P.O. Box 14000 Juno Beach, Florida 33408-0420 SUBJECT; REVIEW OF PRELIMINARYACCIDENT SEQUENCE PRECURSOR ANALYSIS OF OPERATIONAL CONDITION AT ST. LUCIE UNIT 1 Enclosed for your review and comment is a copy of the preliminary Accident Sequence Precursor (ASP) analysis of an operational event which occurred at St. Lucie Unit 1, on November 2, 1997 (Enclosure 1), and was reported in Licensee Event Report (LER)
No. 335/97-011.
This analysis was prepared by our,contractor at the Oak Ridge National Laboratory.
The results of this preliminary analysis indicate that this condition may be a precursor for 1997.
In assessing operational events, an effort was made to make the ASP models as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators. We realize that licensees may. have additional systems and'emergency procedures, or other features at their plants that might affect the analysis.
Therefore, we are providing you an opportunity to review and comment on the technical adequacy of the preliminary ASP,analysis, including the. depiction of plant equipment and equipment capabilities.
Upon receipt and evaluation of your comments, we will revise the conditional core damage probability calculations where necessary to consider the specific information you have provided.
The object of the review process is to provide as realistic an. analysis of the significance of the event as possible.
ln order for us to incorporate your comments, perform any required reanalysis, and prepare the final report of our analysis of this event in a timely manner, you are requested to complete your review and to provide any comments within 30 days of receipt of this letter.
We have streamlined the ASP Program with the objective of significantly improving the time after an event in which the final precursor analysis of the event is made publicly available.
As soon as our final analysis of the event has been completed, we will provide for your information the final precursor analysis of the event and the resolution of your comments.
We have also enclosed several items to facilitate your review. contains specific guidance for performing the requested review, identifies the criteria which we will apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or specific actions in recovering from the event, and describes the specific information that you should provide to support such a claim. is a copy of LER No. 335/97-011, which documented the event.
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T. F. Plunkett April 23, 1998 Please contact me at (301) 415-1479 if you have any questions.
This request is covered by the existing OMB clearance number (3150-0104) for NRC staff followup review of events doc0mented,in LERs. Your response to this request is voluntary and does not constitute a licensing requirement.
Sincerely,
/s/
William C. Gleaves, Project Manager Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-335
Enclosures:
As stated cc w/encls:
See next page
- Docket-File------
PUBLIC St. Lucie Rdg.
JZwolinski FHebdon BClayton WGleaves PO'Reilly, AEOD MBoyle LPlisco, Rll OGC ACRS 1
To receive a copy of this document, indicate in the box: "C"'
Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE PM:PDII-3
+ LA:PDII-3 D:PDII-3 D(
):DRPE K
NAME DATE ave 4/~98 4/ 0/98 4/~98 BCIayton
'Hebdon ki i 4
/98
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/98
/
/97 DOCUMENT NAMEiSTLUCIEiSLS197ST.WPD OFFICIAL RECORD COPY
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T. F. Plunkett April 23, 1998 Please contact me at (301) 415-1479 if you have any questions.
This request is covered by the existing OMB clearance number (3150-0104) for NRC staff followup review of events documented in LERs. Your response to this request is voluntary and does not constitute a licensing requirement.
Sincerely, 4
William C. Gleav s, Project Manager Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-335
Enclosures:
As stated cc w/encls:
See next page
Mr. T. F. Plunkett
'lorida Power and Light Company ST. LUCIE PLANT r
CC:
Senior Resident Inspector St. Lucie Plant U.S. Nuclear Regulatory Commission 7585 S. Hwy A1A Jensen Beach, Florida 34957 Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 M. S. Ross, Attorney Florida Power & Light Company P.O. Box 14000 Juno Beach, FL 33408-0420 John T. Butler, Esquire Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr. Douglas Anderson County Administrator St. Lucie County 2300 Virginia Avenue Fort Pierce, Florida 34982 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 61 Forsyth Street, SW., Suite 23T85 Atlanta, GA 30303-341 5 H. N. Paduano, Manager Licensing '5 Special Programs
'lorida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 J. A. Stall, Site Vice President St. Lucis Nuclear Plant
~ 6351 South Ocean Drive Jensen Beach, Florida 34957 Mr. J. Scarola Plant General Manager St. Lucie Nuclear Plant 6351 South Ocean Drive.
Jensen Beach, Florida 34957 Mr. Kerry Landis U.S. Nuclear Regulatory Commission 61 Forsyth Street, SW., Suite 23T85 Atlanta, GA 30303-3415 E. J. Weinkam Licensing Manager St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957
LER No. 335/97-011 LER No. 335/97-011 Event
Description:
Non~nservative recirculation actuation signal set point Date ofEvent:
November 2, 1997 Plant:
St. Lucie, Unit 1 Event Summary St. Lucie, Unit 1 was defueled in support ofa steam generator replacement refueling outage. Utilitypersonnel determined that the engineered safety feature actuation system (ESFAS) recirculation actuation signal (RAS) bistable set point for indicating the water level in the refueling water tank (RWT) had been set less conservatively than the Technical Specification set point. Plant personnel determined that the unit was more susceptible to core damage in the event ofa large-break losswf~lant accident (LBLOCA)since changing the span ofthe RWT level indication during a 1993 refueling outage.'he nominal core damage probability (CDP) over a 1-year period estimated using the IRRAS model for St. Lucie is 1.7
>< 10'. The increase in the CDP (i.e., the importance) for this event over a 1-year period because ofan improper RAS set point is 1.7 x 10'. Hence, the less conservative RAS set point results in an estimated conditional core damage probability (CCDP) for a 1-year period ofoperation of3.4 x 10'. Uncertainty in the &equency ofa LBLOCA(none have occurred) and uncertainty in the amount of emergency core cooling system (ECCS) ficw required when recirculation is initiated contribute to a substantial uncertainty in this estimate.
Event Description On October 27, 1997, St. Lucie, Unit 1 was defueled in support ofa steam generator replacement refueling outage. As part ofthe outage, obsolete ESFAS bistables were being replaced to improve system reliabilityand calibration methods. The equipment to be replaced included all four channels ofthe RWT low level bistables.
Asignal fromthese RWTlow-level bistables causes the operating mode ofthe safety injection system to change
&om the injection mode to the recirculation mode followinga losswf~lant accident (LOCA).
Because ofthe RWT bistable changes, a system engineer performed additional verification to ensure that the RWT level set point agreed with the instrument loop scaling requirement.
This review showed that the Technical Specification set point of48 in&omthe bottom ofthe RWTcorrelated to a bistable set point of5.28
'A.'he functional test procedure required an assigned set point of4.96 mA, which corresponds to a water level in the RWT of36 in above the tank bottom.'he less conservative set point dictated by the functional test procedure was applied to all 4 channels ofthe Unit 1 RWT level instrument bistables.
In January 1993, an engineering calculation was issued by the St. Lucie engineering staKto change the span ofthe RWT level measurement loop such that zero feet would reference the bottom ofthe tank. The level instruments are actually 1 ftabove the bottom ofthe RWT and the 0 ftmark previously referenced the height ofthe level instruments.
Before the change in the span ofthe RWT level, the set point procedure correctly Enc1osure 1
LER No. 335/97-011 initiated the RAS 48 in above the true bottom ofthe RWT (36 in set point+ 12 in instrument height).'fter the change in the span ofthe RWT level, the set point procedure was not changed.
Subsequently, the set point procedure incorrectly initiated RAS at 36 in above the bottom ofthe RWT.
Additional Event-Related Infonuation The RAS signal causes the suction source for the ECCS system to transfer &om the RWT at the end ofthe injection phase to the containment sump.
This begins the recirculation phase.
The injection phase is not influenced by the set point error and, under worst case conditions, the injection phase willlast for at least the initial20 min followinga LOCA.'hen the RAS set point is reached, the containment sump isolation valve is opened in-30 s and the RWT isolation valve is closed in-90 s. The operator is directed by the emergency operating procedure(EOP) to initiate recirculationmanually ifthe automaticsignal failsto initiaterecirculation when RWT level reaches 48 in from the bottom ofthe tank.'he ECCS suction pipe outer diameter is -24 in. The top ofthe inner diameter ofthe ECCS suction pipe is 42.25 in above the bottom ofthe RWT. The bistable set point error would delay the automatic initiation of recirculation until the water level in the RWT was -6 in below the top ofthe ECCS suction piping. Without operator action, a condition known as open channel flowwould occur. From calculations made by the licensee, open channel flow conditions could not support full ECCS flow (-13,000 gpm).
The mismatch between available open channel flowand fullECCS flowwoulddrain down the level in the ECCS suction piping leading to air ingestion and reduction ofthe available net positive suction head for the ECCS pumps.
Under these conditions, chugging flowwould result until the ECCS pumps became air bound. This interruption ofECCS flowwould prevent a further drain down ofthe RWT and the lower RAS set point might never be reached.'he licensee further calculated that ECCS flows below 7,000 gpm could be supported by open channel conditions untilthe lower RAS set point was reached.'his limits concern forECCS failure, as a result ofthe less conservative RWTbistable set point, to a LBLOCA. Additionally,the peak containment temperature and pressure willbe mitigated withinthe first 20'min followinga LOCA; when containment pressure falls below 5 psig, operators are directed by the EOPs to secure the containment spray pumps.
Because the spray pumps provide more than 6,000 gpm flow(6,750 gpm, Ref. 2, Table 6.34), this would reduce the total ECCS flow required below 7,000 gpm before reaching open channel flowconditions.
Modeling Assumptions This event was modeled as a Mure ofthe automatic recirculation actuation signal for a LBLOCAduring a 1-year period. Afailure to initiate recirculation could air bind all ECCS pumps followinga LBLOCA,leading to core damage, Therefore, the significance ofthis event can be estimated directly &omthe change in ECCS pump failure probabilities because ofthe improper RAS set point and the probability ofa LBLOCAduring a 1-year period. The St. Lucie individual plant examination (1PE) estimates the &equency ofa LBLOCAto be 2.7 > 10"/year.
The increased probability ofcore damage because ofECCS pump failures is estimated by considering the prooability that the operators willfailto initiate recirculation manually when the water level inthe RWT drops
LER No. 335/97-011 to 48 in and the probability that the operators willfail to secure the spray pumps when conditions allow.
Hence, the increased probability ofcore damage is P(LBLOCA) > P(operator fails to initiate recirculation manually) x P(operator fails to secure spray pumps) erator fails to initiate recirculation manuall Manual initiation ofrecirculation following a LOCA is directed by the EOPs ifthe automatic RAS signal should fail. The operators had correct indication ofthe RWT level in the control room and the EOP directs operator attention to the RWT level when the water level drops to 6 to 8 ft..Manual initiationofrecirculation when directed by the EOP would prevent any damage to the ECCS pumps because ofthe RAS set point error.
Itwas assumed that the high-pressure injection (HPI) pumps would not be required to prevent core damage for the LBLOCAofconcern in this event. Therefore, the low-pressure injection (LPI) pumps were expected to be the key to preventing core damage.
Ref.
1 indicated 240 s were available after reaching the nominal RAS set point before the LPI pumps would fail.
@he HPI pumps would fail within 90 s.) A simulator test suggested that, operating crews maintained positive control of the RWT level and recognized the need to manually initiate recirculation -40 s afler the nominal automatic set point was reached. The Human Reliability and Safety Analysis Data Handbook'uggests that for post-accident task actions, a 1-min delay for traveVmanipulation should be allowed (Ref. 3, p. 66). So, 100 s was considered the median time response for an operating crew. Allowing-10 s for the containment sump valves to reposition enough to allow significant flow,the critical time was assumed to be 230 s before the LPI pumps would fail because ofgas binding (i.e.,
240 s 10 s). The probability ofthe operating crew failingto initiate recirculation manually can be estimated by assuming that the failure probability can be represented as a time-reliabilitycorrelation (TRC) as described inHuman ReliabilityAnalysis.'perator response was assumed to be rule-based and without hesitancy. For the 230-s period ofinterest, a failure probability of 1.2 x 10'as estimated.
0 eratorfailstosecures ra um s
Iftotal ECCS flowcould be reduced below 7,000 gpm before reaching the automatic RAS set point, then the potential for ECCS pump failure, because ofthe non~nservative bistable set point, is eliminated (Ref. 1, p.
9). Analysis of containment pressure curves in the FSAR'ndicates that contairiment pressure should be reduced to 5 psig in -20 min following a LBLOCA.,Ifthe RWT level were at the technical specification minimum level [401,800 gal (Ref. 2)] and ECCS flows were at the maximum indicated by reference 1 (13,000 gpm), then a minimum of25.8 min would be available until the intended automatic RAS set point is reached
[66,200 gal (Ref. 2)]. The EOPs direct the operating crew to secure the containment spray pumps when containment pressure returns below 5 psig.
The Human Reliabilityand Safety Analysis Data Handbook'uggests that for post-accident task actions, a 5-min delay fordiagnosis/analysis be allowed and a 1-min delay for traveVmanipulation be allowed (Ref. 4, p. 66). Withthis median response prediction, the probability of the operating crew failingto secure the spray pumps can be estimated by assuming that the Mure probability can be represented as a TRC as described in Human ReliabilityAnalysis.s Operator response was assumed to be rule-based, but with hesitancy, because conditions to secure the spray pumps are not met immediately.
This operator action was assumed to be independent of the operator action to manually establish the recirculation line-up, because ofthe way the EOP directs these activities be performed and the independence ofthe instrumentation required. Allowing0.2 min for equipment response, a failure probability of5.2 ~ 10'
LER No. 335/97-011 was estimated forthe 5.6-min period ofinterest f(25.8 0.2) min].
Asensitivity study on the operator response time is presented at the end ofthe Analysis Results section, Analysis Results The nominal core damage probability (CDP) over a 1-year period estimated using the IRRAS model for St.
Lucie is 1.7 > 10'. The increase in the CDP (i.e., the importance) forthis event over a 1-year period because ofan improper RAS set point is 1.7 ~ 10'. Hence, the less conservative RAS set point results in an estimated conditional core damage probability (CCDP) for a 1-year period of operation of 3.4
< 10 s.
A large uncertainty is associated with this estimate because it relies on an estimated LBLOCAfrequency (none have occurred) and estimations ofoperator response during accident conditions.
The dominant core damage sequence (sequence 4 in Fig. 1) for this event involves
~
a LBLOCA,
~
successful injection from the safety injection tanks,
~
a failure ofautomatic RAS signal,
~
operator fails to manually backup the RAS, and operators fail to secure the containment spray pumps before the RAS set point is reached.
The conditional core damage probabilities are shown in Table 1, while Table 2 lists the sequence logic associated with the sequences listed in Table 1. Table 3 provides the definitions and failure probabilities for event tree branch points in Pig. 1.
The HPI pumps were expected to fail 90 s following a failure ofthe RAS.'fit is assumed that HPI pump failure impacts adequate decay heat removal followingsome LOCAevents ofconcern, the probability ofthe operating crew failingto initiate recirculation manually before HPI pump failure can be estimated by assuming that the failure probability can be represented as a TRC as described inHuman ReliabilityAnalysis.'perator response was assumed to be rule-based and without hesitancy.
Again allowing 10 s for valves to reposition and allow significant flow, a failure probability of6.2 x 10'as estimated forthe 80-s period ofinterest. In this case, the estimated increase in the core damage probability is 8.7 ~ 10'for a 1-year period with the RAS set point too low. The CCDP then is increased to 1.0 < 10".
It could be assumed that the operator focus on current EOP steps and accident conditions overall would preclude any consideration ofsecuring the containment spray pumps so early in the event. In this case, the probabilityofMingto secure the spray pumps before reaching the RAS set point would be 1.0. The estimated CCDP for this event increases to 4.9 > 10 ~ for a 1-year period with the RAS set point lower than the design basis. On the other hand, withincreased attention on RWTwater management per the EOP, itis possible that the operating crew would be quick to secure containment spray pumps when it became permissible. Ifthe median response were considered to be 3 min instead ofthe 6 min assumed previously, the probability that the
'perator fails to secure'the spray pumps is 1.9 < 10'. The estimated CCDP forthis event decreases to 2.3 >
10'or a 1-year period with the RAS set point too low. Similar results are obtained ifthe water level in the
LER No. 335/97-011 RWT is assumed to start well above the minimum water level allowed by Technical Specifications when demanded.
Acronyms CCDP CDP ECCS EFW EOP ESFAS HPI IPE IRRAS LBLOCA LOCA LPI RAS RWT TRC conditional core damage probability, core damage probability emergency core cooling system emergency feedwater system
'mergency operating procedure engineered safety feature actuation system high-pressure injection integrated plant examination Integrated Reliability and Risk Analysis System large-break losswf~lant accident losswf~lant accident low-pressure injection recirculation actuation signal refueling water tank time-reliability correlation References 1.
LER 335/97-011, Rev. 0, "Non-Conservative Recirculation Actuation Signal Set Point Resulted in Operation Prohibited by the Technical Specifications," December 2, 1997.
2.
St. Lucie, Final Safety Analysis Report (Updated Version).
3.
D. I. Gertman and H. S. Blackman, Human Reliabilityand Safety Analysis Data Handbook, John Wiley and Sons, 1994.
4.
E. M. Dougherty and J. R. Fragola, Human Reliability Analysis, John Wiley and Sons, 1988.
IHBJXA SPRAY ST. llCFASP ERG LARGE8%%UX'ABIENlTREE
~ p LER No. 335/97-011 Table 1. Sequence Conditional Probabilities for LER No. 335/97-011 Event tree name C
LBLOCA LBLOCA Sequence number Conditional core damage probability (CCDP) 3.4 E-005 2.7 E-008 Core damage probability (CDP) 1.7 E-005 2.7 E-008 Importance (CCDP-CDP) 1.7 E-005 0.0 E+000 Percent contribution 99.9 0.1 Total (all sequences) 3.4 E-005 1.7 E-005 1.7 E-005 Table 2. Sequence Logic for Dominant Sequences for LER No. 335/97-011 Event tree name LBLOCA LBLOCA
'Sequence number Logic
/SIT, RAS, RAM, SPRAY Table 3. System Names for LER No. 335/97-011 System name IE-LBLOCA SPRAY Description Initiating Event-LBLOCA The RAS bistable fails to change the safety injection mode &om injection to recirculation Operator Guls to manually initiate recirculation when the water level in the RWT drops to 48 in The safety injection tanks fail to inject water properly Operator fails to secure the spray pumps when conditions allow Failure probability 2.7 E-004 1.0 E+000 1.2 E-001 1.0 E-004 5.2 E-001
GUIDANCE FOR LICENSEE REVIEW OF PRELIHINARY ASP ANALYSIS
. Background
The preliminary precursor analysis of an operational event that occurred at your plant has been provided for your review.
This analysis was performed as a part of the NRC's Accident Sequence Precursor (ASP) Program.
The ASP Program uses probabilistic risk assessment techniques to provide estimates of operating event significance in terms of the potential for core damage.
The types of events evaluated include actual initiating events, such as a loss of off-si'.e power (LOOP) or loss-of-coolant accident (LOCA), degradation of plant
-.conditions; and safety equipment failures-.or-unavailabilities that could increase the probability of core damage from postulated accident sequences.
This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), individual plant examination (IPE),
and the licensee event report (LER) for this event.
Yodeling Techniques The models used for the analysis of 1995 and 1996 events were developed by the Idaho National Engineering Laboratory (INEL).
The models were developed using the Systems Analysis Programs for Hands-on Integrated Reliability'Evaluations (SAPHIRE), software.
The models are based on linked fault trees.
Four types of initiating events are considered:
(I) transients, (2) loss-of-coolant accidents (LOCAs), (3) losses of offsite power (LOOPs),
and (4) steam generator tube ruptures (PWR only).
Fault trees were developed for each top event on the event trees to a supercomponent level of detail.
The only support sy-tern currently modeled is the electric power system.
The models may be modified to include additional detail for the systems/
components of interest for a particular event.
This may include additional equipment or mitigation strategies as outlined in the FSAR or IPE.
Probabilities are modified to reflect the particular circumstances of the event being analyzed.
Guidance for Peer Review Comments regarding the analysis should address:
Does the "Event Description" section accurately describe the event as it occurred?
Does, the "Additional Event-Related Information" section provide accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?
Does the "Modeling Assumptions" section accurately describe the modeling done for the event?
Is the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions?
This also includes assumptions regarding the likelihood of equipment recovery.
Encl osure '2
Appendix H of Reference I provides examples of comments and responses for
~ previous ASP analyses.
Criteria for Evaluating Comments Modifications to the event analysis may be made based on the comments that you provide.
Specific docume'ntation will be required to consider modifications to the event analysis.
References should be made to'o'rtions of the LER, AIT, or other event documentation concerning the sequence of events.
System and component capabilities should be supported by references to the
Comments related to operator response times and capabilities should reference plant procedures, the FSAR, the IPE, or appl,icable operator response models.
Assumptions used in determining failure
" probabilities should be clearly stated.
Criteria for Evaluating Additional Recovery Measures Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis.
However, to assess the viability and effectiveness of the eq'uipment and methods, the appropriate documentation must be included in your response.
This includes:
normal or emergency operating procedures.'iping and instrumentation diagrams (P&IDs),
electrical one-line diagrams, results of thermal-hydraulic
- analyses, and operator training (both procedures and simulator),
etc.
- Systems, equipment, or specific recovery actions that were not in place at the time of the event will not be considered.
Also, the documentation should address the impact (both positive and,negative) of the use of the specific recovery measure on:
the sequence of events, the timing of events, the probability of operator-error in using the system or equips ~nt, and other systems/processes already modeled in the analysis (including operator actions).
For example, Plant A (a PWR) experiences a reactor trip, and during the subsequent
'recovery, it is discovered that one train of the auxiliary feedwater (AFW) system is unavailable.
Absent any further information regrading'his
- event, the ASP Program would analyze it as a reactor trip with one train of AFW unavailable. 'he AFW modeling would be patterned after information gathered either from the plant FSAR or the IPE.
However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this unavailability would be Revision or practices at the time the event occurred.
mitigated by'he use of the standby feedwater system.
The mitigation
'ffect for the standby feedwater system would be credited in the analysis provided that the following material was available:
standby feedwater system characteristics are documented in the FSAR or accounted for in the IPE, procedures for using the system during recovery existed at the time of the event, the plant operators had been trained in the use of the system prior to the event, a clear diagram of the system is available (either in the
..previous. analyses
.have indicated.that-there. would be sufficient time ava'ilable to implem'ent 'the procedure successfully under the circumstance's of the event under analysis, the effects of using the standby feedwater system on the operation and recovery of systems or procedures that are already included in the event modeling.
In this case, use of the standby feedwater system may reduce the likelihood of recovering failed AFH equipment or initiating feed-and-bleed due to time and personnel constraints.
Materials Prov)ded for Review The following materials have been provided in the package to facilitate your review of the preliminary analysis of the operational event.
~
The specific
- LER, augmented inspection team (AIT) report, or other pertinent reports.
~
A su~mary of the calculation results.
An event tree with the dominant sequence(s) highlighted.
Four tables in the analysis indicate:
(I) a summary of the relevant basic events, including modifications to the probabilities to reflect the circumstances of the event, (2) the dominant core damage sequences, (3) the system names for the systems cited in the dominant core damage sequences, and (4) cut sets for the dominant core damage sequences.
Schedule Please refer to the transmittal letter for schedules and procedures for submitting your comments.
References 1.
L. N. Vanden Heuvel et al., Precursors to Potential Severe Core Damage Accidents:
- 1994, A Status
- Report, USNRC Report NUREG/CR-4674 (ORNL/NOAC-232) Volumes 21 and 22, Martin Marietta Energy Systems, Inc.,
Oak Ridge
'ational Laboratory and Science Applications International Corp.,
December 1995.