ML20133G979

Forwards Event Rept 352/95-008 on 950911 Re Safety Relief Valve Fails Open,Reactor Scram,Suppression Pool Strainer Fails

ML20133G979
Person / Time
Site:	Limerick
Issue date:	12/23/1996
From:	Shea S NRC (Affiliation Not Assigned)
To:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
References
NUDOCS 9701170007
Download: ML20133G979 (26)

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l December 23, 1996 Mr. George A. Hunger, Jr. 1 Director-Licensing, MC 62A-1 PECO Energy Company Nuclear Group Headquarters Correspondence Control Desk P.O. Box No. 195 Wayne, PA 19087-0195

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Dear Mr. Hunger

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SUBJECT:

FINAL ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT LIMERICK  !

GENERATING STATION, UNIT 1 l l

Enclosed for your information is a copy of the final Accident Sequence j Precursor analysis of the operational event at Limerick Generating Station, Unit 1, reported.in Licensee Event Report No. 352/95-008. This final analysis (Enclosure 1) was ' prepared by our contractor at the Oak Ridge National Laboratory based on review and evaluation of your comments on the preliminary 1 analysis and. comments received..from the NRC staff and from our independent contractor,* Sandia' National Laboi atories. Enclosure 2 contains our responses j to'your specific comments. Our, review of.your comments employed the criteria contained'in'the: material which accompanied the preliminary analysis. The results' of the final analysis indicate that this event is a precursor for

1995. ,

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• Please,contactmeat(301)'415-144) if you have any questions regarding the enclosures. We re' cognize and ' appreciate the effort expended by you and your staff,in reviewing [and: providing comments on the preliminary analysis.

Sincerely, 1

/s/ J. Shea for i Frank Rinaldi, Project Manager l Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-352

Enclosures:

As stated cc w/encls: See next page DISTRIBUTION Docket File WPasciak, RGN-I PUBLIC ACRS, TWF i PDI-2 Reading I g SVarga/JZwolinski -

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0% Ji* UNITED STATES NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20666 4 001

\h,,,g +/ December 23, 1996 Mr. George A. Hunger, Jr.

Director-Licensing, MC 62A-1 PECO Energy Company Nuclear Group Headquarters Correspondence Control Desk P.O. Box No. 195

) Wayne, PA 19087-0195

SUBJECT:

FINAL ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT LIMERICK GENERATING STATION, UNIT 1 i

Enclosed for your information is a copy of the final Accident Sequence Precursor analysis of the operational event at Limerick Generating Station, Unit 1, reported in Licensee Event Report No. 352/95-008. This final analysis (Enclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory based on review and evaluation of your comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories. Enclosure 2 contains our responses

, to your specific comments. Our review of your comments employed the criteria contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1995.

Please contact me at (301) 415-1447 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.

Sinc ly, I

j% r Rinaldi, Project Manager r ect Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-352

Enclosures:

As stated cc w/encls: See next page

l

, Mr. George A. Hunger, Jr. Limerick Generating Station, l'ECO Energy Company l Units 1 & 2 i j cc:

J. W. Durham, Sr., Esquire Mr. Rich R. Janati, Chief Sr. V.P. & General Counsel Division of Nuclear Safety PECO Energy Company PA Dept. of Environmental Resources

2301 Market Street P. O. Box 8469 f

Philadelphia, Pennsylvania 19101 Harrisburg, Pennsylvania 17105-8469

} Mr. David P. Helker, MC 62A-1 Mr. Michael P. Gallagher j Manager-Limerick Licensing Director - Site Engineering PECO Energy Company Limerick Generating Station 965 Chesterbrook Boulevard P. O. Box A

] Wayne, Pennsylvania 19087-5691 Sanatoga, Pennsylvania 19464 l Mr. Walter G. MacFarland, Vice President

! Limerick Generating Station Mr. James L. Kantner

! Post Office Box A Manager-Experience Assessment

Sanatoga, Pennsylvania 19464 Limerick Generating Station P. O. Box A l Mr. Robert Boyce Sanatoga, Pennsylvania 19464

Plant Manager

, Limerick Generating Station Library

P.O. Box A US Nuclear Regulatory Commission

Sanatoga, Pennsylvania 19464 Region I 475 Allendale Road Regional Administrator King of Prussia, PA 19406

U.S. Nuclear Regulatory Commission 2

Region I Mr. Ludwig E. Thibault

475 Allendale Road Senior Manager - Operations

King of Prussia, PA 19406 Limerick Generating Station

P. O. Box A i

Mr. Neil S. Perry Sanatoga, Pennsylvania 19464 Senior Resident Inspector US Nuclear Regulatory Commission Dr. Judith Johnsrud P. O. Box 596 National Energy Committee Pottstown, Pennsylvania 19464 Sierra Club '

433 Orlando Avenue Mr. Darryl P. Lequia State College, PA 16803

. Director - Site Support Services Limerick Generating Station P.O. Boy A Sanatoga, Pennsylvania 19464 ,

Chairman Board of Supervisors '

i of Limerick Towaship 646 West Ridge Pike '

q Linfield, PA 19468 1,

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LER No. 352/95-008 LER No. 352/95-008 l

Event

Description:

Safety / relief valve fails open, reactor scram, suppression pool strainer fails Date of Event: September 11,1995 Plant: Limerick 1 Event Summary Limerick Unit I was manually scrammed frcm 100% power after a safety / relief valve (SRV) failed open. Residual heat removal (RHR) pump A was in the suppression pool cooling (SPC) mode of operation and was being used to remove heat from the suppression pool to compensate for various SRV steam leaks when an SRV failed open, forcing the manual scram. RHR pump A was secured and declared inoperable after oscillations in the pump motor current and decreasing pump flow were observed. Subsequent examination revealed that the pump suction strainer had become obstructed with debris from the suppression pool. The conditional core damage probability (CCDP) estimate for the one-year potential 4

unavailability of the Emergency Core Cooling Systems (ECCS) dependent upon the suppression pool is 1.3 = 10 This is an increase of 9.0 x 10* over the nominal core damage probability (CDP) of 4.0 x 104 for the same period. The CCDP for the actual transient event is 2.5 = 10*

Event Description-i Lunerick I was operating at 100% power at 1245 hours0.0144 days <br />0.346 hours <br />0.00206 weeks <br />4.737225e-4 months <br /> on September 11,1995, when SRV "M" failed open. When plant operators were unable to reclose the valve within 2 min., they manually scrammed the reactor in accordance with technical specification requirements. At the time of the SRV failure, RHR pump A was in service to remove heat from the suppression pool to compensate for various SRV steam leaks.

After the scram, operators aligned RHR B pump for SPC as well. At 1307 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.973135e-4 months <br />, the pressure in the reactor had l decreased from 1005 psig to 410 psig. Even though a closed indication was received for the "M" SRV, reactor pressure continued to decrease. Typically, technical specifications for boiling water reactors (BWRs) require a controlled o

depressurization if the temperature in the suppression pool exceeds 120'F. In such a case, the cooldown rate is typically limited to less than 100*F/h. During this event, however, the uncontrolled depressurization resulted in a cooldown rate of approximately 130*F/h and the temperature in the suppression pool peaked at 124 *F.

At 1320 hoars, operators observed a decrease and fluctuations in flow from the A RHR pump as well as osAm.a u.

its motor current. Operators, attributing these signs to suction strainer fouling, secured the A RHR pump and declared it inoperable. After it was checked, the A pump was restarted but at a reduced flow rate of 2,000 gpm. No problems were observed so the flow rate was gradually increased to 8,500 gpm and no problems were observed. A pressure gauge located on the pump suction was observed to 1.4ve a gradually lower reading, which was believed to be indicative of an increased pressure drop across the pump suction strainers located in the suppression pool. At 0227 hours0.00263 days <br />0.0631 hours <br />3.753307e-4 weeks <br />8.63735e-5 months <br /> on September 12,1995, reactor pressure was reduced below 75 psig, with one loop of shutdown cooling in service. By 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />, the unit was in cold shutdown with a reactor coolant temperature of 194 *F.

Additional Event-Related Information SRV "M" was removed and sent to a laboratory for testing, where it was found to have been damaged by steam erosion of the pilot valve seat. Failure of the pilot valve caused a pressure differential across the SRV main disk, which resulted i

ENCLOSUl:E 1

LER No. 352/95-008 in spurious operation of the main SRV valve. The SRV was reported to have been leaking for more than a year before its failure. Four other SRVs were found to have seat damage and were also replaced During an inspection of the A RHR pump suction strainer assembly, a mat of brown, fibrous material and a sludge of oxide corrosion products were found covering most of the assembly. The sludge material was determined to have come from the suppression pool. Upon inspection, personnel discovered that most of the suction strainer assembly for the B RHR pump was covered with a thinner layer of the same material. However, the B RHR pump ran normally during and aner the event. The other strainers in the suppression pool for the pumps which ~re not employed during this event also had minor sludge accumulations. It is not known to what extent the blowdown caused by the SRV opening increased the rate of debris accumulation on the strainers. Approximately 1,400 lb of debris (wet weight, dry weight would be less) was removed from the suppression pool. A similar amount of material had been removed previously from the Unit 2 suppression pool.

Modeling Assumptions Two assessments were required to analyze this event. First, a transient event assessment was performed to analyze the actual event. Second, a condition assessment was performed because of the prolonged potential unavailability of those ECCS system which are dependent on the suppression pool.

Transient event assessment his event was modeled as a scram with one SRV failed open and one train of RHR unavailable in all modes except SDC because RHR train A was declared inoperable and secured when debris from the suppression pool clogged its suction strainer assembly. Similar debris was found on other strainers, and 1,400 lb of debris (wet weight) was later removed fmm the suppression pool. Reference 4 indicates that, under some circumstances, debris could have migrated and caused obstruction of additional pump strainers. This effect could depend on a number of factors, including the amount of suppression pool agitation caused by shock waves from SRV discharge, the amount of debris in the suppression pool, which specific pumps were placed in service; what flow rates were demanded, how long the pumps were operated, etc.

The potentia! for common-cause failure of all strainers was modeled by adding an additional basic event to the model for each appropriate system. He event "RHRSTRAINERS" was added to the suppression pool cooling models (SPC, SPC/L), the low pressure coolant injection models (LCI/L), and the containment spray system models (CSS /L). In addition, this event was added to the low pressure core spray system models (LCS/L), as core spray is also dependent upon the suppression pool for water. No change was made to the high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC) systern models, which may take suction frorn the suppression pool, because these systems also are provided with an alternate water supply from the condensate storage system.

The CCDP calculated for this event is dependent upon assumptions made regardmg the likelihood that the foreign matter in the suppression pool could cause failure of additional ECCS pumps Research cited in Reference 4 indicates that the debris concentrations present in the Limerick suppression pool (1400 lb sludge /135,000 n' suppression pool water volume x 62.4 lb ft'= 0.02% wt % sludge) were easily sufficient to obstmet multiple ECCS system strainers. Based on Reference 5, a common-cause strainer failure probability of 0.135 was used in the analysis. A sensitivity analysis also was performed, assuming a common-cause strainer failure probability of 1.0.

Condition assessment In addition to the analysis of N reponed transient event, an analysis was made of the prolonged potential unavailability of the ECCS systems that are dependent upon the suppression pool for water. The debris in the suppression pool was 2

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i LER No. 352/95-008 assumed to have been present throughout the operating year (6132 h, assuming a 70% availability), and it was assumed to have the potential to cause failure of LCI, LCl/L, LCS/L, SPC/L, and CSS /L. This event was modeled with one train of RHR unavailable because, during the actual demand, train A of RHR was declared inoperable and secured when debris from the suppression pool clogged its suction strainer assembly. A common-cause strainer failure probability of 0.135 was used in this analysis (RHRSTRAINERS), and a sensitivity case was evaluated for a common-cause failure probability of 1.0. Potential recovery of the power conversion system (PCS) was credited with event PCS-LONG, as it was in the transient assessment.

Analysis Results The CCDP estimate for the one-year potential unavailability of ECCS systems dependent upon the suppression pool is 1.3 x 104 This is an increase of 9.0 x 10

• over the nominal CDP for the same period of 4.0 x 10 4 The CCDP for the actualtransient event is 2.5 x 104 In both cases, the dominant sequence, highlighted as sequence number 4 on the event tree in Fig.1, involves

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the reactor successfully serams,

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the PCS initially fails, RHR system fails, personnel fail to recover PCS in the long term, and containment venting fails Sequence number 4 is still the dominant sequence if a common-cause strainer failure probability of 1.0 is assumed (versus the 0.135 probability used for the actual event analysis). A CCDP of 7.1 x 104 with an importance of 6.7 x 104 is estimated for the long-tenn unavailability of the ECCS. The importance increased 7 times for this sensitivity analysis 4 4 (from 9.0 x 10 to 6.7 x 10 ). The CCDP for the sensitivity analysis for the transient event is 1.4 x 10 ,4or an increase 4

of about 6 times over the CCDP for the actual transient event of 2.5 x 10 It should be noted that main feedwater success coincident with PCS failure is possible in the Limerick model because some failures which render the PCS incapable of functioning as a sink for reactor decay heat do not render it incapable of supportmg main feedwater (e g., turbine trips or load rejections with failures of the turbine bypass valves).

Defmitions and probabilities for selected basic events are shown in Table 1. Table 2 describes the system names associated with the dominant sequences for both the condition assessment and the initiating event assessment The conditional probabilities associated with the highest probability sequences for the condition assessment are shown in Table 3. Table 4 lists the sequence looic awociated with the requences listed in Table 3. Minimal cut sets associated with the dominant sequences for the condnion assessment are shown in Table 5. The conditional probabilities associated with the highest probability sequences for the initiating event assessment are shown in Table 6. Table 7 lists the sequence logic associated with the sequences in Table 6. Minimal cut sets associated with the dominant sequences for the initiating event assessment are shown in Table 8.

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LER No. 351/95-008 l

l Acronyms ADS automatic depressurization system BWR boiling water reactor CCDP conditional core damage probability l

CDP core damage probability CSS containment spray system ,

ECCS emergency core cooling system l

EPC electro-hydraulic control HPCI high pressure coolant injection LOCA loss-of-coolant accident LOOP loss of offsite power LPCI low pressure coolant injection  ;

LPCS low pressure core spray  !

MFW main feedwater PCS power conversion sys'.em RCIC reactor core isolation cooling RHR residual heat removal RHRSW residual heat removal service water RPS reactor protection systera SDC shutdown cooling SLC standby liquid coneol SPC suppression pool cooling SFV safety /reliefvalve TXANS transicat US NRC united states nuclear regulatory comnussion References

1. Licensee Event Report 352/95-00R %m PECO Enngy to USNRC: " Unusual Event and RPS actuation when the reactor was manually shutdown due to the inadvertent opening of a main steam safety relief valve caused by pilot valve seat leakage," 10/10/95.

2. NRC Bulletin 95-02: " Unexpected clogging of a residual heat removal (RHR) pump strainer while operating in a suppression pool cooling mode" USNRC,10/17/95.

3. NRC Information Notice 95-47: " Unexpected opening of a safety / relief valve and complications involving suppression pool cooling strainer blockage" USNRC,10/4/95.

4. Parametnc Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris, NUREG/CR-6224, Zigler, et al., Science and Engineering Associates for USNRC,1995.

5. Common-Cause Failure Data Collection and Analysis System, Vol Common-Cause Failure Parameter Estimations, INEL.74/0064, Marshall and Rasmuson, Idaho National agineering Laboratory for USNRC, 1995.

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l Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 352/95 408 Modified Event Base Current for this name Description probability probability Type even:

IE.TRAN Transient Initiator 4.5 E-004 1.0 E+000 Yes*

ADS-SRV CC VALYS Automatic Depressurization System 3.7 E403 3.7 E-003 No (ADS) Valves Fail to Open ADS-XHLXLERROR Operator Error Prevents 1.0 E 002 1.0 E-002 No Depressurization ADS-XIIE-XLNOREC Oprator Fails to Recover ADS 7.1 E-001 7.1 E-001 No ADI XiiE XLERROR Operator Fails to inhibit ADS and 1.0 E402 1.0 E 002 No Controllevel CDS-SYS-VF COND Condensate Hardware Components 3 4 E 001 3.4 E-001 No Fail CDS-XHE-XLNOREC Operator Fails to Recover 1.0 E+000 1.0 E+000 No Condensate CVS-XHLXLVEhrF Operator Fai!:to Vent Containment 1.0 E 002 1.0 E-002 No EPS-DON-fC-DOC Diesel Generater Failure 1.9 E-002 1.9 E402 No EPS-XIIE-J* % 3 REC Operators Failt a Recover Electric 5.0 E-001 5.0 E 001 No Power Symem hcl-TDP-FC-TRAIN liigh Pressure (cof ant injs.: tion 3 6 E-002 8.6 E-002 No (HPCI) Train level Failures hcl XHLXE-NOREC Operator Fails to Recover HPCI 7.0 E 001 7.0 E 001 No Irl-MOV CC-IDOPB Law Pressure Coolant Injection 3.1 E-003 3.1 E 003 No (LPCI) Train B Injection Valves Fail to Open III-XMLXE-NOREC Operator Fails to Recover LPCI 1.0 E+000 1.0 E+000 No I.CS XHLXE-NOREC Operstor Fails to Recover low 1.0 E+000 1.0 E+000 No Pressure Core Spray System MFW SYS-VF FEEDW Main Ferdwater System (MFW) 4 6 E 001 4 6 E-001 No Hardware Components Fail l 6

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LER No. 352/95-008 i

Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 352/95-008 l

Modified Event Base Current for this name Description probability probability Type event MFW XHE XE-NOREC Operators Fail to Recover MFW 3.4 E 001 3 4 E401 No PCS-LONO Operators Fail to Recover the Power 3 9 E-001 3.9 E-001 No Conversion System (PCS)in the Img Term PCS-SYS-VF-MISC PCS Hardware Components Fail 1.7 E-001 1.7 E-001 No PCS XHE-XE NOREC Operator Fails to Recover PCS 1.0E+000 1.0 E+000 No PPR-SRV-OO-l VLV One or less Safety / relief Valves 1.0 E+000 1.0 E+000 TRUE Yes' (SRVs) Fail to Close PPR SRV-00-2VLYS Two SRVs fant to Close 2.0 E 003 0.0 E+000 FALSE Yes' PPR.SRV OO-3VLYS More Than Two SRVs Fail to Close 2 0 E 004 0.0 E+000 FALSE Yes' RCI TDP FC-TRAIN Reactor Core Isolation Cooling 8.3 E-002 8.3 E-002 No Symem (RCIC) Train Component Failures RCI XIIE XE-NOREC Operator Fails to Recover RCIC 7.0 E 001 7.0 E 001 No RHR-MDP-CF MDPS Common Cause Failure of Residual 3.0 E-004 3.0 E-004 No Heat Removal (RHR) Pumps RilR MDP-FC-TRNA RHR Train A Components Fail 3 8 E 003 1.0 E+000 TRUE Yes' RHR-MOV OO-BYPSB RHR 1mp B Valve to Bypass Heat 3.0 E 003 3.0 E-003 No Exchangers Fails RHRSTRAINERS Common cause Failure of All 0.0 E+000 1.4 E 001 Yes' Strainers RPS-NONREC Nonrecoverable Reactor Protection 2.0 E 005 2.0 E 005 No System (RPS) Trip System Failures RPS-SYS.FC-MECH Mechanical Failures of the RPS 1.0 E 005 1.0 E 005 No RRS XHE-XE-ERROR Operator Fails to Trip the 10 E 002 1.0 E-002 No Recirculation Pumps SDC-MOV CC SUCT Shutdown Coolmg System (SDC) 6.0 E-003 6.0 E 003 No Suction Valves Fail to Open SDC-XHE-XE-ERROR Operator Fails to Abgn/ Actuate the 1.0 E-002 1.0 E-002 Na SDC SDO XHE XE-NOREC Operator Fails to Recover the SDC 1.0 E400 1.0 E+000 No 7

LER No. 352/95-008 Table 1. Definitioes and Probabilities for Selected Basic Events for LER No. 352/95-008 Modified Event Base Currect for this manie Description probability probability Type event SLE CKV CC-INJEC 'Ihe injection Check Valves in the 2.0 E-004 2.0 E-004 No Stan&y Liquid Cornrol Spiem (SLC) Fail SLC-EPV CF VALVS The Explosive Valves in the SLC 2.6 E-004 2.6 E-004 No Fail From Common cause SLC-MDP-CF-MDPS 'the Motor-Dnven Pumps in the 6.3 E-004 6.3 E-004 No SDC Fail From Common Cause SLE XilE-XE ERROR Operator Fails to Start' Control the 1.0 E-002 1.0 E-002 No SDC SLC XilE XE-NOREC Operator Fails to Recover SDC 1.0 E 000 1.0 E+000 h.)

SRV One or less SRVs Fail to Close 2.2 E 003 2.2 E-003 No SSW MOV CC FLOOD Valve Fails to Open 6.I E 003 6.1 E 003 No SSW XIIE XE ERROR Operator Fails to Align RJIR I.0 E-002 1.0 E 002 No Service Water (RiiRSW)

SSW XHE XE-NOREC Operator Fails to Recover RilRSW 1.0 E+000 1.0 E+000 No I

• Applicable to the initiating event assessment only.

• 'Ihe probabihty was set to 0.0 E+000 (FALSE) for the initialmg event assessment. For the conditional event assessment, the base probabilny was not changed in the model.

• The base probability was changed for both the initiating event assessment and the conditional event assessment.

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Table 2. System Names for LER No. 35245-008 l

Systeem name I g ic l 1

ADI Failure to Inhibit ADS and Control Reactor Level  ;

ADS Automatic Depressurization Fails CDS Failure of the Condensate System CVS Containment (Suppression Pool) Venting

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EPS Emergency Power System Fails l l

hcl IIPCI Fails to Provide Suflicient Flow to the Reactor i Vessel LCl Low Pressure Coolant injection Fails  !

LCIL Low Pressure Coolant injection Fails Durmg a LOOP LCS Low Pressure Core Spray Fails LCSL Low Pressure Core Spray Fails During a LOOP MFW Failure of the Main Feedwater System P2 Two SRVs Fail to Close PCS Power Conversion System RCI RCIC Fails to Provide Suflicient Flow to RCS RHRL Residual Heat Removal System Fails During a LOOP RHRPCS Residual Heat Removal System Fails RPl Reactor Shutdown Fails RPS Reactor Shutdown Fails RRS Recirculation Pump Trip SLC Stanu"oy Liquid Control Fails SRV One or Less SRVs Fail to Ckm SSW RHR Senice Water Makeup Fails SSWL RHR Senice Water Makeup Fails During a LOOP 9

LER No. 352/95-008 Table 3. Sequence Conditional Probabilities for the Condition Assessment for LER No. 352/95-008 Conditional Event tree core Core damage Importance Percent name Sequence damage probability (CCDP. CDP) contribution' name probabluty (CDP)

(CCDP) i TRANS 04 4.6 E-006 S.7 E-007 4.0 E-006 44.6 LOOP 03 1.9 E-006 2.4 E-007 1.7 E-006 19.1 TRANS 44 8.3 E-007 0.0 E+000 8.3 E-007 9.2 d

LOOP 20 7.2 E-007 0.0 E+000 7.2 E-007 7.9 TRANS 07 7.1 E-007 8.7 E-008 6.2 E-007 6.9 LOOP 34 4.1 E-007 0.0 E+000 4.1 E-007 4.5 TRANS 27 2.2 E-007 0.0 E+000 2.2 E-007 2.5 LOOP 06 1.1 E-007 1.4 E-008 1.0 E-007 1.1 Condition Assessment 1.3 E-005 4.0 E-006 9.0 E-006 Total (all sequences)

• Percent contribution to the total importance l

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1 Table 4. Sequence legic for Dominant Sequences for the Condition Assessment for LER 352/95408

, Event tree name Sequence name logic TRANS 04 /RPS, PCS, /SRV, /MFW, RHRPCS, CVS LOOP 03 /RP1, EPS, /SRV,41CI, RHRL, CVS i

TRANS 44 /RPS, PCS, P2, /HCI, CDS, LCS, LCI, SSW LOOP 20 /RPI, EPS, /SRV, HCI, RCI,

/ ADS, LCSL, LCI/L, SSW1.

TRANS 07 /RPS, PCS, /SRV, MFW, /HCI, RHRPCS, CVS LOOP 34 /RP1, EPS, P2,41Cl, LCSL, LCIL, SSWL TRANS 27 /RPS, PCS, /SRV, MFW, HCI, RCl, / ADS, CDS, LCS, LCI, SSW LOOP 06 /RPl, EPS, /SRV, HCI, /RCI, RHRL, CVS i

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l Table 5, Conditional Cut Sets for Higher Probability Sequences for the Condition Assessnent I for LER No. 352/95-008 Cut set Percent Conditional no. contribution probability' Cut sets ,

1 TRANS Sequence 04 4.6 E-006 ' ' l 1 54.5 2.5 E-006 PCS-WNG, PCS-SYSAT MISC, PCS-XIIE XE NOREC, /SRV, 1

SDC-X11LXE ERROR, RHRSTRAINERS, CVS-XilE-XE-VENT l

2 32.4 1.5 E-006 PCS-WNO, PCS-SYSAT MISC, PCS XHE-XE-NOREC, /SRV, RRRSTRAINERS, SDC-MOV-CC-SUCT, SDC-XHLXLNOREC,

CVS XHE XEA'ENT 3 12.0 5.6 E-007 PCS-WNO, PCS-SYSAT-MISC, PCS-XHE XLNOREC, /SRV,

, RJIR MDP CF-MDPS, SDC X11E XE-NOREC, CVS-XIIE-XE-VENT LOOP Sequence 03 1.9 E-006 1 53.6 1.0 E-006 /SRV, RHRSTRAINERS, SDC-XHE-XLERROR. CVS-XHE-XEAINT 2 32.2 6.1 E-007 /SRV, RiiRSTRAINERS, SDC-MOV CC-SUCT, SDC-XHE XLNOREC, CVS XilE XEAINT 3 11.9 2.4 E-007 /SRV, RiiR-MDP-CF-MDPS, SDC-XHLXE-NOREC, CVS-XHE-XLVENT 4

4 1.1 2.3 E-008 ISRV, RHR-MOV.OO BYPSB, EPS-DON FC-DOC, EPS-XHE XE-NOREC, i SDC X11E XE-NOREC, CVS-XHLXE-VENT I TRANS Sequence 44 8.3 E-007 s 1 52.1 4.3 E-007 PCS-SYSAT-MISC, PCS XHE-XE-NOREC, PPR SRV OO 2VLVS, i

CDS-SYS-VF4OND, CDS-XHLENOREC, RHRSTRAINERS, 1 IIS-XHE XLNOREC, LCI XHE-XE NOREC, SSW-XHE-XE-ERROR 2 31.8 2.6 E-007 PCS-SYSAT MISC, PCS-XHE-XE NORIC, PPR SRV OO 2VLVS, i CDS-SYSAT-COND, CDS XIIE-XLNOREC, RIIRSTRAINERS, l LIS-XHE-XE-NOREC,III XHLXE-NOREC, SSW-MOV-CC-FWOD,  ;

SSW-XHE-XLNOREC l

3 16.I 1.3 E-007 PCS-SYSAT-MISC, PCS-XIILXE-NOREC, PPR-SRV OO-2VLYS, CDS-SYS-VF COND, CDS-XHLXE-NOREC, RHRSTRAINERS, LCS-XHE XE NOREC,Irl-MOV-CC-WOPB,III XHLXLNORIC, SSW XHLXE-NOREC LOOP Sequence 20 7.2 E-007 1 51.6 3.7 E-007 /SRV, HCI-TDP-FC-TRAIN, HCI XIIE-XLNOREC, RCI-TDP-FC TRAIN, RCI-X11LXLNOREC, RHRSTRAINERS, ifs-XilE-XE-NOREC, Lff XHLXE NOREC, SSW XHE XE ERROR 2 31.5 2.3 E-007 /SRV, hcl-TDP-FC-TRAIN, HCI-XHLXE-NOREC, RCI TDP-FC-TRAIN, RCI-XHLXE-NOREC, RHRSTRAINERS,1fS-XHE XE-NOREC, IfI X11E XE-NOREC, SSW-MOV CC-FIDOD, SSW XHE-XLNOREC 12

I 1

LER No. 352/95-008 1

Table 5. Conditional Cut Sets for Higber Probability Sequences for the Condition Assessment for LER No. 352/95-008 Cut set Percent Conditional so. contribution probability

• Cut sets 3 16.0 1.2 E-007 /SRV, HCI TDP FC TRAIN, hcl-XHLXE NOREC, RCI-TDP-FC-TRAIN, RCI XHE-XE-NOREC, RHRSTRAINERS, LCS-XHE XE-NOREC, III MOV-CC IDOPB, III-XHLXE-NOREC, SSW-XHE-XE-NOREC TRANS Sequence 07 7.1 E-007 I 54.9 3.9 E-007 PCS-IDNO, PCS-SYSAT MISC, PCS XHE-XE NOREC, /SRV, j MFW SYS-\T-FEEDW, MFW XHE-XE-NOREC, RHRS TRAINFPF, 1 SDC-XHE-XE-ERROR, CVS-XHE-XE-VENT 2 32.9 2.3 E-007 PCS-IDNO, PCS-SYS-VF MISC, PCS-XHLXE-NOREC, /SRV, l MFW SYS VF-FEEDW, MFW XHE-XE-NOREC, RHRSTRAINERS, SDC-MOV-CC-SUCT, SDC-XHE-XE NOREC, CVS-XHE XE-VENT 3 12.2 8.7 E-008 PCS-IDNO, PCS-SYS-VF-MISC, PCS-XHE-XE-NOREC, /SRV, MFW SYS47-FEEDW, MFW XHE-XE-NOREC, RHR-MDP-CF.MDPS, l SDC-XHE-XLNOREC, CVS-XHE XE VENT LOOP Sequence 34 4.1 E-007 1 52.I 2.1 E-007 PPR-SRV CO-2VLYS, RHRSTRAINERS,1fS-XHE XLNOREC, LCI-XHLXE-NOREC, SSW XHE XE-ERROR 2 31.8 1.3 E-007 PPR SRV40 2VLYS, RHRSTRAINERS,14S-XHE-XE-NOREC, Ifi-XHE XLNOREC, SSW-MOV CC-FLOOD, SSW XHLXLNOREC 3 16.1 6.6 E-008 PPR SRV.OO 2VLYS, RHRSTRAINERS,1fS XHE-XE-NOREC, ifI-MOV CC-IDOPB,If!-XHLXE NOREC, SSW XHE-XF-NOREC TRAN Sequence 27 2.3 E-007 -

9 w

^* '

1 52.1 1.2 E-007 PCS SYS-VF-MISC, PCS-XHE XE-NOREC, /SRV, MFW SYS-\T-FEEDW, MFW XHLXE-NOREC, HCI TDP FC-TRAIN, HCI XHE-XLNOREC, RCI-TDP-FC TRAIN, RCI XHE-XE-NOREC, CDS-SYSAT-COND, CDS-XHE-XE-NOREC, RHRSTRAINERS, ifs-XHE XE-NOREC, Irl-XHE-XLNOREC, SSW XHLXE-ERROR 2 31.8 7.3 E-008 PCS.SYSAT MISC, PCS-XHLXE-NOREC, /SRV, MFW-SYSAT-FEEDW, MFW XHE XE-NOREC, HCI TDP-FC-TRAIN, HCI XHE XE-NOREC, RCI TDP-FC TRAIN, RCI-XHE-XE-NOREC, CDS-SYSAT COND, CDS-XHLXLNOREC, RHRSTRAINERS, IIS-XHE-XLNOREC, Ifl-XHLXE-NOREC, SSW-MOV CC-FIDOD, SSW-XHE-XE NOREC 3 16.1 3.7 E-008 PCS-SYSAT-MISC, PCS-XHE-XE-NOREC, /SRV, MFW SYS-\T-FEEDW, MFW XHLXE-NOREC, hcl-TDP FC-TRAIN, hcl-XHE-XE-NOREC, RCI-TDP-FC-TRAIN, RCI-XHE-XE-NOREC, CDS-SYSAT COND, CDS-XHLXE-NOREC, RHRSTRAINERS, LCS-XHLXE-NOREC, Irl-MOV CC 1DOPB,Irl-XHLXLNOREC, SSW XHE XE-NOREC LOOP Sequence 06 1.1 E-007 13

1 i

LER No. 352/95-008 Table 5. Conditional Cut Sets for Higher Probability Sequences for the Coodhion Assessment i for LER No. 352/95-008 1

Cut set Pereest Conditional '

me. contribution probability

• Cut sets 1 54.9 6.0 E-008 /SRV, HCI TDP-FC TRAIN, hcl-XHE-XE NOREC, RHRSTRAINERS, SDC-XHE-XE-ERROR, CVS-XHE XE-VENT 2 32.9 3.6 E-008 /SRV. HCI-TDP-FC TRAIN, HCI-XHE XE-NOREC, RHRSTRAINERS, SDC MOV CC-SUCT, SDC-XHE-XE NOREC,CVS-XHE-XE-VENT 3 !2.2 1.3 E-008 /SRV, HCI-TDP FC TRAIN, HCI-XHE-XE NOREC, RHR-MDP-CF-MDPS, SDC-XHE-XE-NOREC, CVS-XHE-XE4'ENT Condition Assessment 1.3 E-005 _ . . _ . ~.

Total (ali sequences) '%' * "

• The conditional probabihty for each cut set is detemuned by muhiplying the probabihty that the portion of the sequence that makes the precursor visible (e.g., the system with a failure is demanded) will occur during the duration of the event by the probabilrties of the remauung basic events in the minimal cut set. This can be approdnated by 1 s', where p is desemuned by muhiplying the expected number ofinitiators that occur during the duration of the event by the probabilities of the basic events in that minimal cut set. The expected number ofinitiators is given by 11, where 1 is the frequency of the inrtisting event (given on a per-hour basis), and t is the duration tune of the event (in this case,6132 h). This approximation is conservative for precursors made visible by the initiating event. The frequencies ofinterest for this event are: 6 = 4.57 x 10%, and 6 = 1.29 j x10 %. j I

l 14

t h

. i i

LER No. 352/95-008 l

Table 6. Sequence Conditional Probabilities for the Initiating Es ent Assessment for LER No. 352/95-008 Conditional core Event tree damage Percent name Sequence name probability contribution (CCDP)

TRANS 04 1.6 E-006 65.3 TRANS 07 2.6 E-007 10.2 l TRANS1 80-15 2.2 E-007 8.7 TRANS- 80-16 2.0 E-007 8.0 TRANS- 27 8.2 E-008 3.2 TRANS 80-14 3.4 E-008 1.3 IE Assessment 2.5 E-006 Total (all sequences)

Table 7. Sequence logic for Dominant Sequences for the Initiating Event Assessment for LER 352/95-008 Event tree name Sequence name logic TRANS 04 /RPS, PCS, /SRV, /MFW, RHRPCS, CVS TRANS 07 /RPS, PCS, /SRV, MFW, / hcl, RHRPCS, CVS TRANS 80-15 RPS, /RRS, SLC TRANS 80-16 RPS, RRS TRANS 27 /RPS, PCS, /SRV, MFW, HCI, RCI, / ADS, CDS, LCS, LCI, SSW TRANS 80-14 RPS,/RRS,/SLC,PCS, AD1 15

LER No. 352/95-008 Table 8. Conditional Cut Sets for Higher Probability Sequences for the Initiating Event Assessment for LER No. 352/95-008 Cut set Percent Conditionel no. contribution probability

• Cut sets TRANS Sequence 04 1.6 E-006 ,,

1 53.7 8.9 E-007 PCS-WNO, PCS-SYSAT-MISC, PCS-XHE XE-NOREC, /SRV, SDC-XHE-XE ERROR, RHRSTRAINERS, CVS XHE-XEA'ENT 2 32.2 5.3 E-007 PCS-IDNO, PCS-SYS-\T-MISC, PCS-XHE XE-NOREC, /SRV, RHRSTRAINERS, SDC MOV CC-SUCT, SDC-XHE-XE-NOREC, CVS.XIIE-XEAINT 3 11.9 1.9 E-007 PCS-LONG, PCS-SYSAT MISC, PCS-XilE-XE-NOREC, /SRV, RHR-MDP CF-MDPS, SDC-XHE-XLNOREC, CVS-XHLXLVENT TRANS Sequence 07 2.6 E-007 1 53.7 1.4 E-007 PCS-!DNO, PCS-SYSAT MISC, PCS-XHE XLNOREC, /SRV, MFW-SYSAT FEEDW, MFW X11E-XE NOREC, RHRSTRAINERS, SDC-XHE XLERROR, CVS XIIE-XEA'ENT 2 32.2 8.3 E-008 PCS-IDNO, PCS-SYSAT-MISC, PCS-XHE-XE NOREC, /SRV, MFW-SYSAT-FEEDW, MFW X1tE-XE NOREC, RHRSTRAINERS, SDC-MOV CC-SUCT, SDC-XHE XE-NOREC, CVS-XHE-XLVENT 3 11.9 3.1 E-008 PCS-IDNO, PCS-SYS-VF-MISC, PCS-MIE-XE-NOREC, /SRV, MFW SYS-VF-FEEDW, MFW XHE-XE-NOREC, RHR-MDP-CF-MDPS, SDC-XHE XE NOREC, CVS XHE-XEA'ENT TRANS Sequence 80-15 2.2 E-007 I 89.5 2.0 E-007 RPS-NONREC, S!I-XHE-XLERROR 2 5.6 1.2 E-008 RPS-NONREC, SLC-MDP-CF MDPS, SLC-XHE-XE-NOREC 3 2.3 5.2 E-009 RPS-NONREC, Slf-EPV CF-VALVS, SLf-XIIE-XE-NOREC 4 1.7 4.3 E-009 RPS-NONREC, Sif-CKV CC-!NJEC, Sif-X11E-XLNOREC TRANS Sequence 80-16 1.0 E-007 1 97.6 2.0 E-007 RPS-NONREC, RRS-XHLXE ERROR TRANS Sequence 27 8.2 E-008 1 51.7 4.2 E-008 PCS-SYS VF MISC, PCS-XIIDXE-NOREC, /SRV, MFW-SYSAT-FEEDW, MFW XHE-XE-NOREC, HCI-TDP FC-TRAIN, HCI-XHLXE-NOREC, RCI TDP-FC-TRAIN, RCl-XHE XE-NOREC, CDS-SYSAT COND, CDS-XHE XE NOREC, RHRSTRAINERS, LIS X}!LXE NOREC, Irl XHE-XLNORff, SSW-XilE-XE-ERROR 16

LER No. 352/95-008

Table 8 Conditional Cut Sets for Higher Probability Sequences for the Initiating Event Assessment for LER No. 352/95-008 Cut set Percent ,

Conditional

no- contribution probability

• Cut sets

! 2 31.5 2.6 E-008 PCS-SYSAT-MISC, PCS. Nile-XE-NOREC, /SRV, MFW SYS-VF-FEEDW, MFW-XHE XE-NOREC, hcl TDP FC-TRAIN, HCI XHE XE NOREC, RCI TDP lC TRAIN, RCI XHE-XE-NOREC, CDS-SYS-\T COND, CDS-XilE XE-NOREC, RHRSTRAINERS, ifs-XHE-XE-NOREC, LCl XHE XE-NOREC, SSW-MOV CC-FLOOD, SSW XIIE XE-NOREC 3 16 0 1.3 E-008 PCS-SYS VF-MISC, PCS-XHE-XE NOREC, /SRV, MFW-SYS-VF-FEEDW, MFW XllE-XE-NOREC, HCI-TDP-FC-TRAIN, hcl-XHE XE NOREC,

, RCI TDP FC-TRAIN, RCI-XIIE-XE-NOREC, CDS SYSAT COND, CDS XIIE XE-NOREC, RHRSTRA'NERS, LCS-XHE XE-NOREC, Irl MOV CC-LOOPB, LCI-XIIE-XE-NOREC, SSW X11E-XE-NOREC TRANS Sequence 80-14 3.4 E-008 s 1 99.5 3.4 E-008 RPS-NONREC, PCS-SYS-VF-MIS' 9VT XllE-XE NOREC, ADI-X}{E XE-ERROR IE Assessment 2.5 E-006 Total (all sequences) 4 d

• The conditional probability for each cut set is deternuned by multiplying the probability of the initiating event by the probabilities of the basic events in that nummal cut set The probabihty of the imtiating events are given in Table 1 and begin with the designator 'IE." 1he probabihties for the basic

] events also are given in Table 1.

4 4

17

i l *

LER No. 352/95-008 1

LER No. 352/95-008 Event

Description:

Safety / relief valve fails open, reactor scram, suppression pool strainer fails Date of Event: September 11,1995 Plant: Limerick 1 IJcensee Comments

Reference:

Letter from G. A. Hunger, Jr., Director - Licensing, PECO Nuclear, to U. S. Nuclear Regulatory Commission, " Limerick Generating Station, Unit 1 Comments Concerning Preliminary Accident Sequence Precursor Analysis of Suction Strainer Clogging Event," July 25,1996.

J Comment 1: First paragraph in the Event Summary, third sentence-Change "RHR pump A was declared inoperable when . " to "RHR pump A was secured and declared inoperable when .

Response 1: Word change was made as requested.

Comment 2: Third paragraph in the Event Summary, first sentence---Change " operators observed a decrease in flow from the A RHR pump . " to " operators observed a decrease and fluctuations in flow from the A RHR pump .

Response 2: The sentence was changed to " operators obsened a decrease and fluctuations in flow from the A RHR pump, l

Comment 3: First paragraph in the Event Description, third and forth sentences-delete " . 6t low flow rates.

As operators increased flow through the A RHR pump, they observed a pressure drop across the pump's suction strainer." Per the text of the LER, page 2 of 5, "At 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br />, following initial evaluation by the System Manager, Shift Supenision directed a restart of the "A" RHR pump (i c. in SPC mode), and no abnormal indications were observed." [ emphasis added). In addition, plant records indicate that the RHR ' A' pump was restarted and ramped up to 8500 gpm and retumed to SPC mode, which is not a " low flow rate". Finally, pressure drop across the pump's suction strainer would be expected with increasing flow rate, but did not hinder operation of the "A" RHR pump.

Response 3: This section was based on information in the LER and also on infonnation in NRC Bulletin (NRCB) 95-02. NRCB 95-02 indicates (p. 2, second paragraph):

18 ENCLOSURE 2

I LER No. 352/95-008 I

[. .] Approximately 30 minutes later, fluctuating motor current and flow was observed on the "A" l loop. Cavitation was believed to be the cause, and the loop was secured. AAer it was checked the "A" I pump was restarted, but at a reduced flowrate of 8ki/m [2,000 gpm]. No problems were observed, so the flow rate was gradually increased back to 32ki/m [8,500 gpm], the full flowTate for the RHR pumps when operating in suppression pool cooling mode. Again, no problems were observed, so the .

pump conth ed to be operated at a constant flow. A pressure gauge located on the pump suction was I

observed to have a gradually lower reading, which was beheved to be indicative of an increased pressure drop across the pump suction strainers located in the suppression pool. AAer about 30 l minutes of additional operation, the suction pressure remained constant.

^ l

\

, The sentence in the Event Description,"At about 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br />, operators restarted the A RHR pumps  :

and it appeared to operate normally at low flow rates." is believed to be consistent with this l information from NRCB 95-02. The quoted section of NRCB 95-02 also appears to suggest that

increasing suction strainer differential pressure was observed during pump operation, but the passage  !

may be subject to other interpretations. In order to avoid interpretation error, portions of the subject I J

paragraph in the Event Description have been replaced with verbatim excerpts from NRCB 95-02.

i Comanent 4: Second paragraph in Additional Event.Related Information-Aner the third sentence starting, "Upon inspection, personnel , " Add the sentence "However, the B RHR pump ran normally during 1

, and aAer the event." l Response 4: The sentence was added Note that this specific information was not given in the LER.

Comment 5: Second paragraph in Additional Event-Related Information-Reword the ftAh sentence from l

" Utility personnel reported that they were unable to determine if effects attributable to the SRV l 1 blowdown increased the rate of accumulation of debris on the strainers." to " Utility personnel reported that the SRV blowdown resulted in deposition of additional material on the strainer." per the first paragraph of page 5 of 5 of the LER 352/95-008.

1 Response 5: This statement was based upon the following information from NRC Information Notice 95-47 (p. 3, para. 4): [. .] Whether the blowdown caused by the SRV opening increased the rate of accumulation .

4 on the strainer is not known. I i ,

I l l Because the source of this information is unclear, the sentence in question has been revised to say,"It is not known to what extent the blowdown caused by the SRV openmg increased the rate of debris accumulation on the strainers."

Comment 6: Second paragraph in Additional Event-Related Information-Finally, second to last sentence, change "Approximately 1,400 pounds of debris was removed from the suppression pool." to 19

1

. i i

LER No. 352/95-008 1

"Approximately I A00 pounds (wet weight, dry weight is roughly 1/3 of wet weight) of debris was removed from the suppression pool." The 1,400 pounds reported was a net weight value. BWROG investigations have shown that the dry weight is roughly one-third of the wet weight.

Response 6: It would seem likely that the ratio of dry weight to wet weight would be dependent upon the type of debris encountered. Presumably this ratio would be smaller for fibrous material and larger for metal oxides. The dry weight of the debris was not reported in the LER and the comment implies that the net weight of the debris found in the Limerick suppression pool was not actually measured. Therefore it would seem difficult to determine what the ratio of dry weight to wet weight might have been for this event. However, a sentence has been added to indicate that the dry weight could be expected to be less than the wet weight.

Comment 7: Second paragraph in Modeling Assumptions-Change the first sentence "and one train of RHR unavailable in all modes because. " to "and one train of RHR unavailable in all modes except SDC because.

Response 7: This change has been incorporated.

I Cosiment 8: Second paragraph in Modeling Assumpdons-Also change second sentence "Similar debris was l found on other strainers and 1,400 pounds of debris . " to " Debris was also found on other strainers j and 1,400 pounds (wet weight, dry weight is roughly 1/3 of wet weight) of debris . " for the same explanation as given above.

Response 8: Consistent with the response to comment 6, the sentence in question has been revised to indicate that the 1,400 pounds was wet weight. l Comment 9: Second paragraph in Modeling Assumptions, last sentence-Finally, add "the amount of debris in the suppression pool" to the list of factors in the last sentence.

Response 9: This change has been incorporated.

Comment 10: Third paragraph in Modeling Assuusptions 'Ihe low pressure core spray (LPCS) system should not be grouped with the RHR system since LPCS can also take suction from the CST, similar to the RCIC and HPCI systems (see Figures 6.3 7 and 6.3-9 in the Limerick Generating Station's Updated Final Safety Analysis Report). The standard LPCS system operating procedure provides direction for 20

I j

t 4 i LER No. 352/95-008 l

alignment of LPCS to the CST. Thus the "RHRSTRAINERS" event should not be added to the LPCS model.

Response 10

The LPCS system is normally aligned to the suppression pool and a number of operator actions would be required to align the system to the CST, including the opening of manual valves which are normally closed (and may even be locked closed). Therefore, it may be inappropriate to model the system as being unaffected by the potential strainer failure. Switching to take suction on the CST could be incorporated into the LPCS model as a recovery, but a typical non-recovery probability estimate for a non-routine ex-control room action such as this would be relatively large and the change would not materially affect the analysis results.

Comment 11: Fourth paragraph in Modeling Assumptions-The common cause strainer failure probability should be modeled as two populations of two strainers each, RHR A and B, and RHR C and D, rather than as a single group which contains all four strainers. This is due to the distinctly different operating histories of the two groups. RHR A and B are normally used for suppression pool cooling in routine operations, whereas RHR C and D are only run for required pump, valve, and flow tests. As the i failure mode is dependent upon the collection of material on the strainers over time as the pumps are used, these different profiles would clearly separate the two groups from common cause perspective.  ;

The analysis for both the event and condition a* essment should be reperformed with the increased j common cause value affecting only the A and B strainers, and with a much lower common cause l failure value affecting the C and D strainers (e g. 2a Qt = 0.2 x IE-4, or = 2E-5. Each group of RHR, l the RHR A and B group and the RHR C and D group, can be used in each mode of RHR operation, LPCI, SPC, SDC, and Containment Spray.

Response 11: The Licensee Event Report for this event indicates that the strainer fouling was due in part to the SRV discharge. ("The 'M' SRV discharge resulted in deposition of additional material on the strainer.")

The belief that turbulence from suppression pool blowdown would increase the suspension of debris and thereby the deposition of debris on pump suction strainers is consistent with experimental dats such as that described in NUREGICR-6224 (Parametric Study of the Potentialfor BHR ECCS Strainer Blockage due to LOCA Generated Debris, G. 2igler, et. al, SEA lnc.,0ctober 1995).

It is believed that anv nm taking suction from the suppression pool would have been subjected to a common increased chance of failure, had it been operated, due to the suspended debris. Neither the the Licensee Event Report nor the licensee's comments on the preliminary analysis provided suflicient information to permit estimation of the relative importance of the different mechanisms leading to the RHR pump suction strainer fouling.

Comment 12: Fourth paragraph in Modeling Assumptions, second sentence-The statement "Research cited in Reference 4 indicates that the sludge concentration . . were easily sufficient to obstruct multiple ECCS system scainers" is incorrect. Sludge by itself cannot cause the failure of an ECCS strainer due to the small particle size relative to the hole size of the strainer. A layer of fiber must be present to trap the sludge. From the results of diver inspections, it was found that no strainers other than the A and B RHR had any fiber rr.atting the strainer surfaces. Therefore, initially, the strainers could not have 21 l

LER No. 352/95-008  ;

plugged. A preliminary BWROG report indicates that appreciable settling of corrosion products could be expected in as little as 15 to 30 minutes following the end of a LOCA blowdown. Based on this analysis, it would be expected that by the time a fiber bed formed on any other strainers, the corrosion products required to foul the bed would have largely settled out. Therefore, no other ECCS suction strainers would be expected to plug. Therefore, a common cause strainer failure could not occur.

Response 12: The word " debris" has hen substituted for " sludge" in the subject sentence, since both fibrous material and oxide / sludge material were present.

The referenced report apparently pertains to large or medium break LOCA events and may not apply directly to events such as the one modeled in the analysis. Note that in the discussion above, blowdown to the suppression pool is assumed to suspend corrosion products. This would tend to support the assumption that a single common hazard could potentially impact some or a!! of the pumps taking suction from the pool.

In the actual event, the 'M' SRV remained stuck open over a prolonpd period, which is believed to have increased the amount of debris suspended in the suppession pool. In addition, experimental evidence indicates that "[i]nitiation of suppression pool cooling . . can induce high levels of turbulence in the suppression pool [which] may result in resuspension of debris." (NUREG/CR-6224, p. B-6)

The ASP analysis assumed that the same fibers and corrosion products which caused A RHR pump to be declared inoperable and which were found in lesser amounts on the B pump strainers could have led to failure of the B pump and then the other exposed pumps as well. Presumably, failure of B pump i would have cued operators to stan C or D pump, failure of that pump would have prompted them to start the remaining unaffected pump, and so on.

l Comment 13: For rth paragraph in Modeling Assumptions, third sentence-The alpha factor used from Reference 5 should be recalculated for 'he event assessment using the actual failure situation found a the plant (i.e. I failed (A), I could fail (B), and with the remaining two strainers (C and D) having an extremely low likelihood of failing in the same manner). I Response 13: The Licensee Event Report for this event indicates that the SRV discharge increased the deposition of debris on the operating pump strainers. As previously discussed, the same material which obstructed the A pump strainers, causing the pump to be declared inoperable, did deposit to a lesser extent on the B pump strainers and could have deposited on the C and D pump strainers had these pumps operated. Therefore, the modeling of the event assumed that the one pump which was reported to be inoperable was inoperable and it was assumed that the other pumps could have failed with a common-cause failure probability of 0.135. [The probability that 3 or more pumps might fail due to a common cause given that two failed due to that cause should be approximately unity.]

Comment 14: Fifth paragraph in Modeling Assumptions, last sentence-The common cause strainer failure probability should not be the same as in the event assessment, unless the A strainer is considered to be clogged in the condition assessment as well. The common cause strainer failure probability should be 0.135 = Qt, Qt being the random failure probability for the strainers. Unless the A RHR is assumed to be failed (as in the event asowent), Qt is less than 1.

22

LER No. 352/95-008 4

Response 14: During the actual event, the A strainer did obstruct sufficiently to cause the A pump to be declared inoperable. The Licensee Event Report attributed the strainer fouling to buildup of debris during

~

operation and to deposition of additional material due to the SRV discharge. There was not sufficient information to permit estimation of the relative importance of the different mec* w. isms leading to the strainer fouling. Since the A pump was declared inoperable when demanded ;t was assumed in the l condition assessment that that the A pump was inoperable. j l

Coniment 15: Anelysis Resuks--Considering the previous comments of, a) the appropriate use of the debris dry weight to estimate the strainer failuie probability.

b) the grouping of LPCS with HPCI and RCIC instead of with RHR since LPCS can also take suction j from the CST, and I

c) modeling the common cause strainer fai!ure probability as two populations of two strainers each l (RHR A and B, and RHR C and D) with a much lower common cause failure value affecting the RHR l C and D strainera of 2E-5.

a more realistic core damage probability for the transient event assessment below the current value would be obtained, and a more realistic core damage probability for the condition assessment would be less than 1.0E-5. j l

Response 15: These comments have been addressed as noted a) see Responses 6 and 8, l

b) see Response 10, c) see Responses 11 and 13.

The CCDP estimate for the one-year potential unavailability of ECCS systems dependent upon the 4

suppression poolis 1.3 x 10, an increase of 9.0 x 104over the nominal CDP of 4.0 x 10 The CCDP for the actual transient event is 2.5 x 104 l

l 23 i

I l