ML20204B672
| ML20204B672 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 03/12/1999 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Kingsley O COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 9903220205 | |
| Download: ML20204B672 (31) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. spee64em j
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March 12, 1999 i
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l Mr. Oliver D. Kingsley, President Nuclear Generation Group Commonwealth Edison Comoany Executive Towers West 111 1400 Opus Place, Suite 500 Downers Grove,IL 60515 SUDJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF OPERATIONAL CONDITION AT BYRON STATION, UNIT 1
Dear Mr. Kingsley:
Enclosed for your review and comment is a copy of the preliminary Accident Sequence l
Precursor (ASP) analysis of an operational condition which was discovered at Byron Station, Unit 1 (Byron 1) on September 12,1998 (Enclosure 1), and was reported in Licensee Event Report (LER) No. 454/98-018. This analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL). The results of this preliminary analysis indicate that this condition may be a precursor for 1998. 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.
In order for us to incorporate your comments, perform any required re-analysis, 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 resciution of your comments.
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9903220205 990312 PDR ADOCK 05000454 P
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4 2-l We have also enclosed several items to facilitate your review. Enclosure 2 contains specific I
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-identifed additional equipment or specific actions in recovering from the event, and describes the specific information that you should provide to support such a claim. Enclosure 3 is a copy of LER No.
454/98-018, which documented the event.
Please contact me at (301) 415-3017 if you have any questions regarding this request. This request is covered by the existing OMB clearance number (3150-0104) for NRC staff follow up 4
review of events documented in LERs. Your response to this request is voluntary and does not i
constitute a licensing requirement.
Sincerely, 30 ys John B. Hickman, Project Manager Project Directorate ill-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. STN 50454 Enciosures: As stated.
ec w/ encl: See next page e.
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<e
i March 12,1999 i i We have also enclosed severalitems to facilitate your review. Enclosure 2 contains specific guidance for performing the requested review, identifies the criteria which we will apply to j
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. Enclosure 3 is a copy of LER No.
454/98-018, which documented the event.
Please contact me at (301) 415-3017 if you have any questions regarding this request. This 1
request is covered by the existing OMB clearance number (3150-0104) for NRC staff follow up review of events documented in LERs. Your response to this request is voluntary and does not constitute a licensing requirement.
Sincerely, ORIG. SIGNED BY John B. Hickman, Project Manager Project Directorate 111-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. STN 50-454
Enclosure:
As Stated cc w/ encl: See next page j
Qis1ribution.
IDochet FilsQi PUBLIC PDlll-2 r/f JZwolinski/SBlack SRichards CMoore MJordan, Rill JHickman SMays, RES PO'Reilly, RES OGC,015B18 ACRS, T2E26 i
DOCUMENT NAME: G:\\PD3-2\\CM\\ BYRON \\BY-ASP.WPD To RECElvE A COPY OF THis DoCuME WT lNDICATE IN THE box: "C" = COPY WITHouT ENCLOSURES *E" = COPY WITH ENCLOSURES "N" 1
i OFFICE PM:PDA,,( PM-2 DD:PD3-2 NAME J.H M kE SRICHARDSN DATE 3/ / R /99 3/ h/99 3/ lb /99 3/
/99 OFFICIAL RECORD COPY
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l O. Kingsley Byron Station Commonwealth Edison Company Units 1 and 2 cc:
Regional Administrator, Region lli Mrs. Phillip B. Johnson U.S. Nuclear Regulatory Commission 1907 Stratford Lane 801 Warrenville Road Rockford, Illinois 61107 Lisle, Illinois 60532-4351 Attomey General lilinois Department of Nuclear Safety 500 S. Second Street Office of Nuclear Facility Safety Springfield, Illinois 62701 1035 Outer Park Drive Springfield, Illinois 62764 Commonwealth Edison Company Byron Station Manager Document Control Desk-Licensing 4450 N. German Church Road Commonwealth Edison Company Byron, Illinois 61010-9794 1490 Opus Place, Suite 400 Downers Grove, Illinois 60515 Commonwealth Edison Company Site Vice President - Byron Ms. C. Sue Hauser, Project Manager 4450 N. German Church Road Westinghouse Electric Corporation Byron, Illinois 61010-9794 Energy Systems Business Unit Post Office Box 355 Mr. David Helwig Pittsburgh, Pennsylvania 15230 Senior Vice President Commonwealth Edison Company j
Joseph Gallo Executive Towers West lil Gallo & Ross 1400 Opus Place, Suite 900 1025 Connecticut Ave., N.W., Suite 1014 Downers Grove, Illinois 60515 Washington, DC 20036 Mr. Gene H. Stanley Howard A. Learner PWR Vice President
)
Environmental Law and Policy Com.monwealth Edison Company Center of the Midwest Executive Towers West lli 35 East Wacker Drive 1400 Opus Place, Suite 900 Suite 1300 Downers Grove, Illinois 60515 Chicago, Illinois 60601 Mr. Christopher Crane U.S. Nuclear Regulatory Commission BWR Vice President Byron Resident inspectors Office Commonwealth Edison Company 4448 North German Church Road Executive Towers West lli Byron, Illinois 61010-9750 1400 Opus Place, Suite 900 Downers Grove, Illinois 50515 Ms. Lorraine Creek RR 1, Box 182 Manteno, Illinois 60950 Chairman, Ogle County Board Post Office Box 357 l
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O. Kingsley Byron Station Commonwealth Edison Company Units 1 and 2 Mr. R. M. Krich Vice President - Regulatory Services Commonwealth Edison Company Executive Towers West lli 1400 Opus Place, Suite 500 Downers Grove, Illinois 60515 Commonwealth Edison Company Reg. Assurance Supervisor-Byron 4450 N. German Church Road Byron, Illinois 61010-9794 Ms. Pamela B. Stroebel Senior Vice President and General Counsel Commonwealth Edison Company P.O. Box 767 Chicago, Illinois 60690-0767
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.'d LER No. 454/98-018 LER No. 454/98-018 Event
Description:
Long-term unavailability of an emergency diesel generator Date of Event: September 12,1998 Plant: Byron Station, Unit 1 Event Summary Byron Station, Unit 1 (Byron 1), had been in Mode I for six months following a refueling outage. During a monthly surveillance test on the l A emergency diesel generator (EDG), the EDG tripped on a low lube oil pressure signal durir.g the first minute of the test run. Personnel determined that the 1A EDG had been susceptible to tripping on a low lube oil pressure signal for ~18 days until the EDG was repaired and tested ratisfktority. He core damage probability (CDP) at Byron 1 increased because ofthe increased susceptibility that would result from a loss of offsite power (LOOP) which could progress to a station blackout. The 4
estimated increase in the CDP (i.e., the importance) for this event is 6.9 x 10.
Event Description On September 12,1998, operators were starting the 1 A EDG for the planned monthly surveillance test. He 1 A EDG was started locally in the slow start mode. The 1 A EDG experienced a test-mode trip on an " engine lube oil pressure low" signal during the first minute of the test ren. Concurrent with this alarm were an " engine lobe oil pessure low" signal and a " turbo lube oil pressure low" signal. An immediate inspection of the 1 A EDG failed to reveal any leaks or obvious compr
- failures; all piping components were in the correct configuration.'
Subsequent troubleshooting revealed that a fibrous material, consistent with that of the engine's main tube oil filter element medium, had clogged both lube oil strainers. De fibrous material was found to have covered the entire intemal surface of the strainer element. An intemal inspection of the lobe oil filter housing unit showed that none of the filter elements had undergone a catastrophic failure. Regardless, the licensee decided to replace all 146 filter elements in order to perform a closer inspection of the removed elements. While replacing the filters, personnel noted that one filter element was missing its cartridge guide and many other filter elements were slightly crushed. A root <:ause analysis determined that an inadequate maintenance practice was a factor in allowing a significant amouat of unfiltered oil to bypass the filter elements and dislodge and transport the filter material to the lube oil strainers.'
Additional Event-Related Information ne lube oil circulating pymp for each EDG runs continuously during standby conditions so that % mternal engine parts remain lubricated. His facilitates a rapid start of the diesel engine, ne Erkis at Byron are designed to trip on a low lube oil pressure condition when manually started or when th9 manual test mode 1
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LER No. 454/98-018 switch is seiceted at the main control board Although the 1 A EDG should have successfully started in an emergency, the ability of the EDG to continue to perform its required function with a low lube oil pressure condition was questionable.'
No fibrous material was discovered in any other part of the 1 A EDG lube oil system Additionally, although no fibrous material was found in the lube oil filters on the turbocharger, the filters were replaced.'
The l A EDG was returned to service on September 14,1998. A review of the 1 A EDG operating history revealed that the lube oil relief valve had liAed on September 3,1998. The licensee subsequently determined
)
that the reliefvalve had lined because of the strainer blockage. Therefore, the 1 A EDG was considered to be unavailable for at least 11 days - from September 3,1998, until September 14,1998, when the 1 A EDG was retumed to service. The licensee could not determine the actual point of failure before September 3,1998. The last successful surveillance test on the 1 A EDG was completed on August 19,1998. Additionally, the 1 A EDG operated without incident for approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during a LOOP event on August 4,1998.
The licensee verified that the IB EDG was continually available between August 19,1998, and September 14, 1998.'
Modeling Assumptions AAer reviewing the 1 A EDG records, the licensee considered the l A EDG to be unavailable for the ll-day period between September 3,1998, when the tube oil relief valve liAed, and September 14,1998, when the EDG was returned to service. Becausc plant personnel could not precisely determine the actual failure point of the l A EDQ for this analysis, the EDG was assumed to be unavailable for one-half of the 15 day interval between the last successful surveillance test (August 19,1998) and the point when the clogged strainers had the potential to be positively identified (September 3,1998). Therefore, this event was modeled as an 18 day
.(432 h) condition assessment with the 1 A EDG failed. Additionally, because a root-cause analysis determmed that an inadequate maintenance practice along with an inadequate maintenance procedure contributed to the
- clogging of the lube oil strainers, a potential for a common-cause failure of any of the other station EDGs existed.
Because the ability to cross-tie the A and B emergency buses between Byron I and 2 exists, the EDGs from Unit 2 were added to the Integrated Reliability and Risk Analysis System (IRRAS) model for Unit 1. The probabilities that either of the opposite unit EDGs fails to start and run (basic events EPS-DGN-FC-2A and EPS-DGN-FC-2B) were set to the base probability of the Unit 1 EDGs (3.8 x 102). Based on data from NUREGICR-5496, Ewuluation ofLoss of0fsite Power Events atNuclear Power Plants: 1980-1996 (Table ES-3), given a LOOP has occurred at one unit, the probability that both units are affected is 0.11.8 Therefore, a new basic event, MULTI-UNIT LOOP, was added to the Byron model. This new basic event accounts for the probability that the emergency buses and EDGs at Unit 2 might not be available to supply ac power during a LOOP at Unit 1 (i.e., the conditional probability of a LOOP occurring at both units simultaneously given one unit is experiencing a LOOP = 0.11). In addition, because operators must manually cross-tie the emergency buses between units, a baic event was added to reflect the probability that the operator fails to start and load 2
J
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LER No. 454/98-018 the alternate EDG (basic event EPS-XHE-XM-OU). Basic event EPS-XHE-XM-OU was set to 8.0 x 10-2 based on a human error analysis provided in the Byron indisidual plant examination (IPE).8 De common-cause failure probability of the emergency power system for the base case was based on the two EDGs at Byron 1. His was adjusted based on the availability of four EDGs and was developed based on data distributions contained in NUREG/CR-5497, Common-Cause Failun Parameter Estimations (Ref. 3, Table 5-9: alpha factor distribution summary - fail to start, CCCG = 4, a,s = 0.0116; and Table 5-12: alpha factor distribution summary - fail to run, CCCG = 4, a. = 0.0146). Because a. is equivalent to the py5 factor of the multiple Grock letter (MGL) method used in the IRRAS models, the base case common cause failure probability of the EDGs (basic event EPS DGN-CF-ALL) was adjusted from 1.4 x 108 based on two EDGs to 4.6 x 10' based on four EDGs.
Based on the failure of the 1 A EDG with common-cause failure potential, the basic event, EPS-DGN-CF-LL, was adjusted for this event from 4.6 x 10 to 1.5 x 102 based on the MGL method (Ref. 3, Table 5-5 Summary of MGL Parameter Estimations - Fail to Run). De portion of the base case EDG common cause failure probability for an engine start was not altered because the low lube oil pressure trip is not in effect following an emergency EDG start. However, the portion of the base case EDG common-cause failure probability for failure to run for the mission time was adjusted based on the failure mechanism described.
Analysis Results i
ne increase in the CDP (i.e., the importance) as the result of an 18 day failure ofthe 1 A EDG with common-cause failure-to run implications for this event is estimated to be 6.9 x 104. The base probability over the same 4
18-day period (the CDP) for all sequences is 9.0 x 10, resulting in a conditional core damage probability (CCDP) of 7.8 x 104. As expected, station blackout (SBO) sequences dominate. He dommant core damage sequence for this event (Sequence 18 on Fig. I and Sequence 18 9 on Fig. 2) involves the following events-a LOOP, a successful reactor trip, a
a failure of the emergency power system, e
a successful initiation of auxiliary feedwater, e
successful control of reactor coolant system pressure such that the PORVs remain closed, a
a failure of the reactor coolant pump (RCP) seals, and a failure of the operators to restore ac power before core damage.
His sequence accounts for almost 24% of the total contribution to the increase in the CDP. A second SBO sequence where the RCP seals do not fail, but the operators fail to restore ac power before the battenes are depleted (sequence 18-2) accounts for an additional 21% of the increase in the CDP. A third SBO sequence where the PORVs fail open accounts for 20% of the increase in the CDP (sequence 18-20).
Definitions and probabilities for selected basic events are shown in Table 1. De conditional probabilities associated with the highBt probability sequences are shown in Table 2. Table 3 lists the sequence logic 3
W i
I LER No. 454/98-018 l
associated with the sequences listed in Table 2. Table 4 describes the system names associated with the dominant sequences. Minimal cut sets associated with the donunant sequences are shown in Table 5.
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1 LER No. 454/98-018 1
l Acronyms AFW auxiliary feedwster system CCDP conditional core damage probability CDP core damage probability EDG emergency diesel s..;or HPI high pressureinjection PE individual plant examination IRRAS Integrated Reliability and Risk Analysis System LOCA loss-of-coolant accident LOOP loss ofoffsite power MGL multiple Greekletter PORV power-operated relief valve RCP reactor coolant pump SBO station blackout References
- 1. LER 454/98-018, Rev. O," Inoperable Unit 1 Diesel Generator Due to1.ow Lube Oil Pressure Condition,"
October 9,1998.
- 2. C. L. Atwood, et. al., Evaluation ofLoss ofOfsite Power Events at Nuclear Power Plants: 1980-1996, NUREG/CR 5496, November 1998.
- 3. ComEdByron andBraidwoodStations individualPlant Examinations, March 1997.
- 4. Marshall, Rasmuson, and Mosleh, Commorf-Couse Failune Parameter Estimations, NUREG/CR-5497, October 1998.
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y_ :1, LER No. 454/98-018 Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 454/98-018 Modified Event Base Current for this name Description probability probability Type event IEIDOP Initiating Event-IDOP (excludes the 1.6 E 005 1.6 E 005 No Probability ofRecovering Offsite Powerin the Short Term)
IE-SOTR Initiating Event-Steam Generator 1.6 E 006 1.6 E 006 No Tube Rupture IE-SLOCA Initiating Event-Small-Break LOCA 23 E 006 23 E 006 No IE TRANS Initiating Event-Transient 2.5 E-004 2.5 E 004 No AFW-EDP-FC 1B AuxiliaryFeedwater(AFW)Desel.
2.0 E 002 2.0 E 002 No Driven Pun p Fails AFW FMP CF-ALL Common-Cause Failure ofAFW 2.1 E404 2.1 E404 No Pumps EPS-DON CF ALL Common Cause Failure ofEDOs 4.6 E 004 1.5 E-002 Yes EPS-DON-FC 1A EDO IA Fails 3.8 EA)02 1.0 E+000 TRUE Yes EPS-DON-FC 1B EDO 1B Fails 3.8 E 002 3.8 E 002 No EPS XHE-XM OU Operator Fails to Cross Connect ESF 8.0 E 002 8.0 E 002 NEW No Bus Without se Power to Opposite
)
Unit HPI-MDP CF-ALL Common Cause Failure ofHigh 7.8 E 004 7.8 E 004 No Pressureinjection (HPI) Pumps HPI-MDP-FC 1B HPIMotor-Driven Pump Fails 3.8 E 003 3.8 E-003 No HPI-XHE XM-FBL OperatorFails toInitiate Feed and 1.0 E 002 1.0 E 002 No Bleed Cooling LOOP 17-NREC IDOP Sequence 17 Nonvocovery 2.2 E 001 2.2 E 001 No Probability-Failure to Recover AFW-L (0.26) and Feed & Bleed (0.8)
LOOP 18 02 NREC LOOP Sequence 18 02 N.-
e i 8.0 E 001 8.0 E 001 No Probability-Failure toRecovw Electric Power (EP)
LOOP 18-09-NREC LOOP Sequence 18 09 Nonrecovery 8.0 E 001 8.0 E 001 No Probability-Failure to Recover EP j
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i LER No. 454/98-018 Table 1. Definitions and Probabilities for Selected Basic Events for l
LER No. 454/98-018 (Continued)
Modified Event Base Current for this name Description probability probability Type event LOOP 18-11-NREC LOOP Sequenos 1811 Nonrecovery 8.0 E 001 8.0 E 001 No Probability-Failure to Recover EP j
LOOP 18-18-NREC LOOP Sequence 18-18 Nonrecovery 8.0 E-001 8.0 E 001 No Probability-Failure to Recover EP LOOP-18-20-NREC LOOP Sequence 18-20 Nonrecovery 8.0 E 001 8.0 E 001 No Probability-Failure to RecoverEP LOOP 18-22-NREC LOOP Sequence 18-22 Nonrecovery 2.7 E 001 2.7 E 001 No Probability-Failure to Recover EP (0.8) and AFW-L(0.34)
MULTI-UNIT LOOP LOOP Affects Second Unit Given 1.1 E 001 1.1 E 001 NEW No First Unit Affected OEP-XHE-NOREC-Operator Fails to Recover ac Power 2.0 E 002 2.0 E 002 No BD Before Battery Depictum OEP-XHE.NOREC-OperatorFails to Recover oc Power 63 E4X)!
63 E 001 No SL Before Core Damage Results From a SealLOCA OEP-XHE-NOREC-Operator Fails to Recover oc Power 53 E-001 53 E 001 No ST in the Short Term PPR-SRV CC-PRV1 PORV I Fails to Open on Demand 63 E-003 63 E 003 No PPR-SRV CC-PRV2 PORV 2 Fails to Open on Demand 63 E 003 63 E 003 No PPR SRV CO SBO Safety / Relief Valves Open During an 3.7 E 001 3.7 E-001 No SBO PPR-SRV OO-PRV1 Powetoperated ReliefValve 3.0 E 002 3.0 E 002 No (PORV)1 Fails to Recent Aner Opening PPR SRVMPRV2 PORV 2 Fails to Rosent AAer 3.0 E 002 3.0 E 002 No Opening RCS-MDP-LK-RCP Seals Fail without Cooling and 3.5 E 002 3.5 E 002 No SEALS Injection 9
,o LER No. 454/98-018 Table 2. Sequence Conditional Probabilities for LER No. 454/98-018 Conditional Event tree Sequence core damage Core damage importance Percent name number probability probability (CCDP-CDP) contribution *
(CCDP)
(CDP)
LOOP 18-09.
1.7 E-006 6.7 E-008 1.6 E-006 23.8 LOOP 18-02 1.5 E-006 5.8 E-008 1.4 E-006 20.8 LOOP 18-20 1.5 E-006 5.6 E-008 1.4 E-006 20.1 LOOP 18-18 1.0 E-006 3.9 E-008 9.6 E-007 14.0 LOOP 18-11 8.8 E-007 3.4 E-008 8.4 E-007 12.2 LOOP 18-22 4.5 E-007 1.8 E-008 4.3 E-007 6.3 LOOP 17 1.7 E-007 2.0 E-008 1.5 E-007 2.2 Total (all sequences) 7.8 E-006 9.0 E-007 6.9 E-006 1
i
' Percent contribution to the total importance 1
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t LER No. 454/98-018 Table 3. Sequence Logic for Dominant Sequences for LER No. 454/98-018 Event tree name Sequence Logic number LOOP 18 09
/RT-L, EP, /AFW-L, /PORV-SBO, SEALLOCA, OP-SL LOOP 18-02
/RT-L, EP, /AFW-L, /PORV-SBO,
/ SEAL'LOCA, OP-BD i
l LOOP 18-20
/RT-L, EP, /AFW-L, PORV-SBO, PRVL-RES,
)
ACP-ST LOOP 18-18
/RT-L, EP, /AFW-L, PORV-SBO, /PRVL-RES, SEALLOCA, OP-SL LOOP 18-11
/RT-L, EP, /AFW-L, PORV-SBO, /PRVL-RES, /SEALLOCA, OP-BD l
LOOP 18-22
/RT-L, EP, AFW-L, ACP-ST l
LOOP 17
/RT-L, /EP, AFW-L, FAB-L i
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LER No. 454/98-018
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Table 4. System Names for LER No. 454/98-018 System name Logic ACP-ST Offsite Power Recovered in Short Term AFW-I, No or Insufficient AFW System Flow During LOOP EP Emergency Power System Fails FAB-L Failure to Provide Feed and Bleed Cooling OP-BD Operator Fails to Recover ac Power Before Battery Depletion OP SL Operator Fails to Recover ac Power Before Core Damage Results Following an RCP Seal LOCA PORV SBO PORVs Open During a SBO PRVL-RES PORVs and Block Valves Fail to Reclose RT-L Reactor Fails to Trip During a LOOP SEALLOCA RCP Seals Fail During a LOOP
+
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LER No. 454/98-018 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 454/98-018 Cut set Percent nuinber contribution CCDP' Cut sets
- LOOP Sequence 18-09
' l.7 E-006
$$@M3M?M3l@MEj$$$M$$MMij 1
67.5 1.1 E-006 EPS-DGN CF-ALL, /PPR-SRV CO SBO, RCS-MDP-LK-SEALS, OEP.
XHE NOREC-SL, LOOP.I8 09-NREC 2
18.8 3.2 E-007 EPS-DON-FC-1 A,EPS-DGN-FC 1B, MULTI UNIT LOOP,
/PPR-SRV CO SBO.RCS-MDP-LK SEALS,OEP-XHE-NOREC SL, LOOP 18-09-NREC 3
13.7 2.3 E-007 EPS-DON FC 1 A, EPS-DGN-FC 1B, EPS-XHE XM OU,
/PPR-SRV CO-SBO,RCS-MDP-LK-SEALS,OEP-XHE-NOREC-SL, LOOP 18 09 NREC kb!!(ifMMMihM[MMM3M$1%M LOOP Sequence 18-02 1.5 E-006 1
67.5 1.0 E-006 EPS-DGN-CF-ALL,/PPR SRV CO SBO,/RCS-MDP-LK-SEALS, OEP-XHE-NOREC-BD, LOOP 18 02-NREC 2
18.8 2.8 E-007 EPS-DGN-FC 1 A,EPS-DGN-FC 1B,MUL11-UNIT LOOP,
/PPR SRV CO SBO,/RCS-MDP-LK-SEALS,OEP-XHE-NOREC-BD, LOOP 1842-NREC 3
13.7 2.0 E-007 EPS-DGN-FC I A,EPS-DGN FC-1B,EPS-XHE XM OU,
/PPR SRV CO-SBO,/RCS-MDP-LK-SEALS,OEP-XHE-NOREC-BD, LOOP-18 02-NREC fM t
&kfiff*
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- 4 W@m 6lQP!"c%g-egn'" ]@?*K'M*"gy&me g"g"g"fg LOOP Sequence 18-20 1.5 E-006 4%
1 33.7 4.9 E-007 EPS-DGNCF-ALL,PPR-SRV CO-SBO,PPR SRV OO-PRV1, OEP XHE NOREC ST, LOOP 18-20-NREC 2
33,7 4.9 E-007 EPS-DGN4F-ALL, PPR SRV CO-SBO,PPR-SRV OO-PRV2, OEP-XHE-NOREC ST, LOOP 18 20-NREC 3
9.4 1.4 E-007 EPS-DON-FC I A, EPS-DON-FC-1B, MULTI-UNIT LOOP, PPR-SRV CO SBO, PPR-SRV OO PRVI. OEP-XHE NOREC-ST, LOOP-18-20-NREC 4
9.4 1.4 E-007 EPS-DGN-FC 1 A,EPS-DGN-FC 1B, MULTI-UNIT-LOOP, PPR-SRV CO-SBO,PPR SRV OO-PRV2,OEP-XHE-NOREC-ST, LOOP-18 20-NREC 5
6.8 -
9.9 E-008 EPS-DON-FC 1A,EPS-DGN-FC 1B EPS XHE-XMOU, PPR SRV CO-SBO,PPR-SRV OO-PRVI,OEP-XHE-NOREC-ST, d
LOOP 18-20-NREC l
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LER No. 454/98-018 Cut set Percent number contribution CCDP' Cut sets" 6
6.8 9.9 E-008 EPS-DGN-FC-1 A, EPS-DON-FC-1B, EPS-XHE XM OU, PPR SRV CO SBO,PPR SRV OO PRV2,OEP-XHE-NOREC-ST, LOOP 18-20-NREC
$$hpM $ d[NM7pM$[M}$$@$M M $
LOOP Sequence 1818 1.0 E-006 1
67.5 6.7 E-007 EPS-DGN CF-ALL,PPR-SRV CO-SBO, RCS-MDP ILSEALS, OEP-XHE-NOREC-SL, LOOP-1818-NREC 2
18.8 1.9 E-007 EPS DGN-FC 1 A, EPS-DGN-FC 1B,MULT1-UNIT IDOP, PPR SRV CO-SBO, RCS-MDP-LK-SEAIJ, OEP XHE-NOREC-SL, LOOP 1818-NREC 3
13.7 1.4 E-007 EPS-DON-FC l A,EPS-DGN FC 1B,EPS XHE-XMOU, PT,t SRV CO SBO, RCS-MDP-LK-SEALS,OEP-XHE NOREC-SL, LOOP 1818-NREC
%p M W %y, m d: h [ %p a [,; [ lift ? k m !;y p o
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g gg; g
y Bj tj LOOP Sequence 18-11 8.8 E-007 1
67.5 5.9 E-007 EPS-DGN CF-ALL, PPR SRV CO SBO, /RCS-MDP-LK-SEAIE, OEP-XHE-NOREC BD, LOOP-18-11 NREC 2
18.8 1.6 E-007 EPS-DGN-FC l A, EPS-DGN-FC 1B, MULTI-UNIT IDOP, PPR-SRV CO SBO,.RCS DP-LK SEALS,OEP XHE NOREC-BD, LOOP 18-11 NREC 3
13.7 1.2 E-007 EPS-DGN-FC IA,EPS-DGN-FC 1B,EPS-XHE-XMOU,
+
PPR SRV CO-SBO,/RCS-MDP-LK-SEALS,OEP-XHE NOREC-BD, LOOP-1811-NREC
!fyeghgg/WwwrnesQ,wmejenwspggpq a p p m &- wqppagpam LOOP Sequence 18 22 4.5 E-007 l
66.6 3.0 E-007 EPS DGN-CF-ALL, AFW-EDP-FC 1B,OEP-XhE-NOREC-ST, LOOP 18 22-NREC 2
18.6 8.2 E-008 EPS-DON-FC 1A,EPS-IviN-FC 1B, MULTI-UNIT-LOOP, AFW-EDP-FC 1B OEP XHE-NOREC-ST, LOOP-18-22-NREC 3
13.5 6.0 E-008 EPS-DGN-FC-1A,EPS-DGN-FC-1B EPS-XHE XMOU, AFW-EDP-FC 1B,OEP-XHE-NOREC ST, LOOP-18-22-NREC
, M M 3 $ $ [ 4 $ i k ( $ $ $ M y $ $ $)
LOOP equence17 1.7 E-007 I
19.6 3.3 E-003 EPS-DGN-FC-1 A MULTI-UNIT-LOOP, AFW-EDI-FC-1B, HPI-XHE-XM-EL, LOOP 17-NREC 2
14.3 2.4 IL008 EPS-DGN-FC 1A,EPS-XHE-XMOU, AFW EDP-FC 1B, HPI-XHE XM-FBL, LOOP 17-NREC 3
12A" 2.1 E-008 EPS DON-FC-1A,MUL11-UNIT-LOOP, AFW-EDP FC 1B, PPR SRV CC-PRVI,1DOP 17-NREC a
14
+
LER No. 454/98-018 Cut set Percent i
number contribution CCDP Cut sets
9.0 1.5 E-003 EPS-DON-FC 1 A,EPS XHE XMOU, AFW-EDP-FC 1B, PPR SRV CC-PRVI, LOOP 17 NREC 6
9.0 1.5 E-003 EPS DON-FC 1A,EPS XHE-XM4U AFW-EDP-PC 1B, PPR-SRV CC-PRV2, LOOP 17-NREC 7
7.5 1.3 E-003 EPS-DON-FC 1 A. MULTI-UNIT LOOP, AFW-EDP-PC 1B, HPI-MDP-FC 1B,1DOP-17 NREC 8
5.4 9.2 E-009 EPS-DON-FC IA,EPS XHE-XMOU, AFW-EDP-IC 1B, HPI-MDP-FC hB, LOOP 17-NREC 9*
1.9 3.2 E-009 AFW-FMP CF-ALL,HPI XHE-XM-FBL, LOOP-17-NREC 10 15 2.6 E-009 EPS-DON-FC IA, MULTI-UNIT LOOP, AFW-EDP-FC 1B, HPI-MDP CF ALL,IDOP 17-NREC 11 1.2 2.0 E-009 AFW-PMP CF-ALL,PPR-SRV CC-PRVI, LOOP 17-NREC 12 1.2 2.0 E-009 AFW-PMP CF ALL,PPR-SRV CC-PRV2,1DOP 17-NREC 13 1.1 1.9 E-009 EDS-DON-FC 1 A, EPS-XHE-XM OU, AFW-EDP-FC-1B, HPI-MDP CF-ALL, LOOP-17-NRLC Total (ell sequences) 7.8 E-006
!@NMS55f$$MS@$f$$$@$M[d
- The change in conditional probability (imponance)is determined by calculating the conditional probability for the period it Miich the condition existed, and subuscting the rWitional probability for the same period but with plant equipment assumed to be operstmg nonunaDy. The conditional probabiliO' fomh cut set withir, a sequ nee is determined by multiplying the probability that the portion of the sequence that makes the pacursor visible (e.g., the system with a failure is demanded) will occur during the duration of the event by the probebilities of the remeining basic events in the munimal eut set. This can be approximated by 1 - e*, wh:.re p is deterrriwd by multiplyinphe expected number ofin; tutors that occur during the duration of the event by the probabilities of the basic events in that minimal eut not. The expected number ofinitiators is given by At, where lis the frequency of the initiating event (given on a per-hour
'essis), and Iis the duration time of the event. This approximation is conservative for precursors made visible by the initiating event. The fregrencies ofinterest for this event are: Arna,. = 2.5 x 10%,1uia, = 1.6 x 10 */ho laoes = 2.3 x 10 */h,and i, ara = 1.6 x 10 % The duration time for this event is C2 h.
" Basic event EPS-DON-FC 1 A,is a TRUE type event which is not normally included in the output of fault tree reduction programs but has baon added to aid in understanding the sequences to potential core damage associated with the event.
IS
n o-S GUIDANCE FOR LICENSEE REVIEW OF PRELIMINARY ASP ANALYSIS
Background
The preliminary precursor analysis of an operational event that occurrad 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-site 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 j
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.
Modeling Techniques i
The models used for the analysis of 1998 event were developed by the Idaho National Engineer ng Laboratory (INEL). ~
.iodels were developed using the Systems Analysis,
Programs for Hands-on IntegS oeliability Evaluations (SAPHIRE) software. The models are based on linked fault t,ses. Four types of initiating events are considered: (1) transients, (2) loss-of-coolant accidents (LOCAs), (3) losses of offsite power (LOOPS), and (4) steam generator tube mptures (PWR only). Fault trees were developed for each top event on the event trees to a supercomponent level of detail. The only support system 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 j
circumstances of the event being analyzed.
Guidance for Peer Review Comments regarding the analysis should address:
e Does the " Event Description" section accurately describe the event as it occurred?
e Does the " Additional Event-Related Information" section provide accurate additional information concoming the configuration of the plant and the operation of and procedures associated with relevant systems?
"e 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 potentthi to occur under the event conditions? This also includes assumptions regarding the likelihood of equipment recovery.
.. =
=..._:.:.:.=====---'
4 j
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j Appendix G of Reference 1 provides examples of comments and responses for previous ASP analyses.
5 j
Criteria for Evaluating Comments 1
Modifications to the event analysis mcy be made based on the comments that you provide, i
Specific documentation will be required to consider modifications to the event analysis.
References should be made to portions of the LER, AIT, or other event documentation conceming the sequence of events. System and component capabilities should be supported by references to the FSAR, IPE, plant procedures, or analyses. Comments related to operator response times and capabilities should reference plant procedures, the FSAR, the IPE, or applicable operator response models. Assurrptions used in determining failure probabilities
]
should be cleariy stated.
Criteria for Evaluating Ad$tional Recovery uessures Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis. However, to assess the viability and effectiveness of the equipment and methods, the appropriate documentation must be included in your response.
This includes:
normal or emergency operating procedures.*
piping and instrumentation diagrams (P&lDs),*
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 equipment, 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 auxinary feedwater (AFW) system is unavailable. Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW unavailable. The AFW modeling would be pattemed 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 unavailabl!Ry would be mitigated by the use of the standby feedwater sy; tem The Revision or practices at the time the event occurred.
-~
4
- e O
mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:
standby feedwater system characteristics cre docurr.ented 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 syster.1 pr;or to the event, a clear diagrarn of the system is available (either in the FSAR, IPE, or supplied by the licensee),
previous analyses have indicated that there would be sufficient tiric available to implement the procedure successfully under the circumstances of me event under analysis, l
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 raduce the likelihood of recovering failed AFW equipment or initiating feed-and-bleed due to time and personnel constraints.
Materials Provided for Review The following materials have been provided in the package to facilitate your review of the preliminary analysis of the operational event.
e The specific LER, augmented inspection team (AIT) report, or other pertinent reports.
A summary of the calculation results. An event tree with the dominant sequence (s) e highlighted. Four tat.!es in the analysis indicate: (1) a summary of the relevant basic evants, including modifications to the probabilities to reflect the circumstances of the event, (2) the dominant enre damage sequences, (3) the system names for the systems cited in the dominant core damap9 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.
R. J. Ealles et al., " Precursors to Potential Severe Core Damage Accidentc 1997, A Status Report," USNRC Report NUREG/CR-4674 (ORNidNOAC-232) Volume 26 Lockheed Martin Energy Research Corp., Oak Ridge National 1.aboratory, and Science Applications intemational Corp., Oak Ridge, Tennessee, November 1998.
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4 4.
NRC FORM 366 U.S, NUCLEA,R REGULATORY COMMIS$10N APPROVED BY OMB No. 3150 0104 8N EXPIRES C4/30/98 i
ESTIMATED BURDEN PER RESPONSE TO COMPLY W'TN TMS MANDATORV WFORMArl0N COLLECTION REQUEST: 50.0 Mb$. REPORTED LESSONS LEARNED ARE LICENSEE EVENT REPORT (LER) ar*PDRATED WTO THE UCEGWG PROCES$ M0 FE0 BACK TO WDUSTRt r0RWARD COMMENTS REGARDING BURDEN [$Til4 ATE TO THE WFORMATIDN Als0 (See reverse for required number of
[l ANT 0IN'Ei $0 Y #E r PERY$a"Rt0Nis PYOx"CYo sa degits/ characters for each block) 0104. OmCE Of MANAGEMENT AND BUDGET, WASNWGTON. DC 20Mn Picauty mane sta pocarv myMasa as case on BYRON NUCLEAR POWER STATION, UNIT 1 05000454 1OF7 T!TLE to)
INOPERABLE UNIT 1 DIESEL GENERATOR DUE TO LOW LUBE OIL PRESSURE CONDITION EVENT DATl! (El LER NUMBER 86)
' REPORT DATE (7)
I oTHER FACILITIES INVOLVED IS)
R I f Aciuiv NAME DOCKET NUMSER gg
,M
(;0 NTH DAY VEAR VEAR MONTH DAY YEAR l gg ll f ACluf y NAME DOCKET NUMBER 09 12 98 88 018 00 10 09 98 05000 OPERATIhrG THit REPORT 18 SudMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 8: (Chec4 one or motel (til MODE (fa) 20.2201(b) 20.2203(aH2Hvi X
E0.73teH2Ho 50.73(aH2Hvm)
POWER 20.2203(aH1) 20.2203(aH3Hd 50.73(aH2Hul 50.73(aH2Hx)
LEVEL (10) g 20.2203(aH2H0 20.2203(aH3Hu) 50.73(aH2Huo 73.71 20.2203(aH2Hu) 20.2203(aH4) 50.73(aH2Hiv)
OTHER 20.2203(aH2Hud 50.36(CHI) 50.73(aH2Hv) specifv m Abstreet below
^
20.2203(aH2Hiv) 50.36(cH2) 50.73(aH2Hvn)
LICENSEE CONTACT FOR THis LER 612)
NAME TELEPMONE NUMBER (Include Area Godel Dinnis Nicol 815 234 5441, X3065 J se Dubon 815 234-5441, X4504 COMPLETE ONE La#E FoR EACH COMPoNEP FAILURE DEsCRisED IN THis REPORT < 131 Af A0 CAUSE SYSTEM COMPONENT MANUFACTURER Af0 N P
f CAUSE SYSTEM COMPONENT MANUFACTURER 0
SUPPLEMENTAL REPORT EXPECTED (146 l
MONTH DAY HAR EXPECTED.
1ES r.UBMisSiON (if yes, complete EXPECTED SUBMISSION DATE).
x E0 l
DATE(16)
ABSTRACT tLimit to 1400 spaces,i.e., approximately 15 sino6e-spaced typewritten hnes) (18)
At 0901, on September 12,1998, while performing the Diesel Generator monthly surveillance, the 1 A Diesel Gtnerator, a Cooper Bessimer KSV 20, tripped on low lube oil pressure. When Operations began to perform the monthly Unit 1 Emergency Diesel Generator (EDG) 1 A surdilance, a low lube oil alarm initiated which subsequently Irad to an engine trip. The surveillance was being conducted in accordance with Comed procedure 1BOS 8.1.1.2.a 1, Revision 14. The 1 A EDG experienced a Test Mode Trip on " Engine Lube Oil Pressure Low" ccncurrent with alarms for " Engine Lube oil Pressure Low" and " Turbo Lube Oil Pressure Low." The event was experienced during the local slow start and it occurred during the first minute of the monthly surveillance test run.
The muitiple indications of low lube oil pressure (trip and alarms) indicated an actual low lube oil pressure condition cxisted. This,instead of a spurious signal or out of tolerance pressure sender condition, was verified via multiple prgssure alarms for the engine and turbo bearings. The engine is designed to trip upon a low lube oil pressure condition when manually started or when the manual test mode switch is selected at the Main Control Board (MCB).
An immediate investigation was conducted by Maintenance and Systems Engineering personnel to determine the cruse of the trip and the basis for the alarms. The investigation revealed that the cause of the trip was due to low lobe oil pressure caused by clog $ed strainers. As part of the trouble shooting investigation, the two parallel lube oil strainers were inspected. Both parallel lube oil strainers were found clogged with a fibrous material consistent with th t of the engine main lube oil filter element media. The material was subsequently confirmed through a laboratory (nalysis to be that of the filter element media, it vias subsequently t'3termined that the 1 A EDG was historically inoperable from 9/3/98 until 9/14/98 which is a condition prohibited Dy the Byron Station Technical Specifications.
This event is reportable per 10CFR50.73(a)(2)(i)(b).
90102100.-9sM POR R0 4
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l NRc FORM 366A I4 951 U.s. NUCLEAR REGULATORY commission LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (31 l
YEAR SEQUENTIAL REVISION I
BYRON NUCLEAR POWER STATION, UNIT 1 NUMBER NUMBER 2 OF 7
l 05000454 98 -- 018 --
00 l
i
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l TEXT lif more space is required, use additional copies of NRC Form 366A) (175 PLANT CONDITIONS PRIOR TO EVENT:
m.
Unit: 1 Event Date: 09/12/98 Event Time: 0901 Unit 1 Mode 1 Rx Power 100%
RCS [AB) Temperature / Pressure: NOT/NOP Unit 2 Mode 2 Rx Power 100%
RCS (AB] Temperature / Pressure: NOT/NOP B.
DESCRIPTION OF EVENT:
On September 12,1098, while Unit 1 was at 100% power, Surveillance Procedure 1BOS 8.1.1.2.a-1,1 A Diesel Generator Operability (staggered) and Semi-Annual (staggered) Surveillance, was being performed. The purpose of the procedure is to verify the Unit 1,1 A Emergency Diesel Generator (EDG) OPERABILITY. The EDG experienced a test mode trip on " Engine Lube Oil Pressure Low" concurrent with alarms for " Engine Lube Oil Pressure Low" and " Turbo Lube Oil Pressure Low." The event occurred at the initiation of a local slow start while performing the monthly surveillance run. The trip occurred during the first minute of the surveillance test run local start. The multiple indications of low lube oil pressure (trip concurrent with alarms) j indicated that an actual low lube oil pressure condition existed and, therefore, was not considered ? ' be a spurious signal or an out of tolerance pressure sensor / sender condition. Immediately following not.iication of the event to the Main Control Room (MCR), the Limiting Condition for Operation Action Requirement (LCOAR) for A.C. Sources (TS 3.8.1.1, Action a.) was entered as required by the Byron Station Technical Specifications (TS).
i l
In an attempt to determine the cause of the event, an immediate investigation by Maintenance and System Engineering personnel was conducted. This investigation included a visualinspection for leaks and obvious incorrect alignments and component failures. The inspection revealed that there were no leaks es mvious component failures and that all piping components were in the correct configuration. The visua; mvestigation was immediately followed by a trouble shooting investigation. The trouble shooting included an internal inspection of the parallel strainers and the filter housing. The trouble shooting revealed that both lube oil strainers were clogged with a fibrous material consistent with that of the engme main lube oil filter element i
media. The fibrous material was found to have thinly covered the entire internal surface of the strainer element.
l The discovery of the fibrous material in the strainers lead to an internal inspection of the lube oil filter housing l
l unit. This inspection revealed that none of the filter elements had undergone a catastrophic failure. However, l
a decision was made to replace the entire one hundred forty six (146) filter elements in order to inspect them for signs of metal wear particles and anything unusual that would indicate a breakdewn of the filter function.
At the time of replacement, it was noted that one of the filter elements was missing its cartridge guide and many were slightly crushed.
l l
Other inspection actions taken were es follows:
The Ivbe oil strainer outlet 6" check valve was removed for inspection and found with no fibrous matenal present. The check valve was noted to be in good working condition.
The turbo lube. oil filters were replaced., No fibrous material was noted on tither of the two filter assemblies.
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i NHC FORM 36SA u.s. NUCLEAR REauLAToRY CoMMISsloN m
LICENSEE EVENT REPORT (LER) l TEXT CONTINUATION I
FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER BYRON NUCLEAR POWER STATION, UNIT 1 05000454 98 -- 018 00 TEXT ilt more space is required. use ad$tional copies of NRC Form 366A) (11)
B.
DESCRIPTION OF EVENT (cont.)
The diesel engine crankcase was completely drained an#d cleaned followed by a general inspection of the internals. No signs of fibrous material were found and no signs of internal damage were noted.
The drained engine oil was sent for laboratory analysis to inspect vor wear metal products and fibrous material; no concerns were identified.
I The engine lube oil was replaced with a fresh supply.
The circulating tube oil pump (pre-lube) relief valve was removed, inspected and replaced with a new valve. The removed valve was tested fo!!owing the inspection and found to be functioning properly.
I The tube oil temperature control valve,1DG5003A, was inspected and found to have no visible degradation, blockage or fibrous material.
Following the satisfactory completion of the above inspections, the 1 A EDG was tested, the pre-lube and post maintenance run oil pressures were verified normal. The EDG was then returned to ready for test condition status.
On September 14,1998, the 1 A EDG operability surveillance was successfully completed with no further events occurring, immediately following the surveillance, both lube oil strainers were disasseinbled, inspected -
and were found to be clean. Oil samples from before and after the test run, were sent for laboratory analysis l
and resul*.s were reported to be within specification requirements. On Seotember 14,1998, at 18:12 hours, the 1 A EDG was returned to service and the LCOAR for A.C. Sources (3.8.1.1) was exited.
C.
CAUSE OF EVENT:
The cause of the event was investigeted through the Comed Corrective Action Program's (CAP) Root Cause Analysis (RCA) investigation process. The investigation determined that the cause of the EDG engine trip event was a result of low lube oil pressure caused by clogged lube oil strainers. The lube oil strainers were determined to be obstructed with fibrous material that originated from the lubc oil filter media. This was confirmed from laboratory analysis of the material that caused the riogging effect.
Other factors that contributed to the event were determined to be an inadequate maintenance procedure and inadequate maintenance practice. The me;ntenance procedure did not provide specific instructions to insta:!
l the filter elements and the filter housing cover. Per the manuf acturer, the installation of t'he cover must be done carefully to ensure that the filter clements are not disturbed from their position during the lowering of the cover. Any sideways motion that occurs after the initial contact of the cover with the springs may result in one or several of the filter elements not being properly loaded by the springs and potential crushing or misalignment of the filter elements. The maintenance procedure did not contain this level of detail to warn and advise personnel on the proper method for re assembly of the filter housing.
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NRC FORM 3icA u.S. NUCLEAR REGULATORY Commission um LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FANLITY NAME til DOCKET LER NUMBER is)
PAGE (3) ygg SEQUENTIAL REVISION NUMBER NUMBER BYRON NUCLEAR POWER STATION, UNIT 1 05000454 98 -- 018 --
00 TEXT W more space os required, use additionet copies of NRC form 36GA) (11)
C.
CAUSE OF EVENT (cont.)
l The inadequate maintenance practice was determined to ts a lack of attention to detail. One of the filter elements was found to have been installed without its cartridge guide and many of the filter elements were
)
found to have been crushed slightly, it was determined that the crushing of the elements occurred when the springs that provide the force to keep the elements in place may have been tilted during cover installation.
The crushing of the elements was discussed with the manufacturer who confirmed the theory of the tilted springs causing uneven compression forces that may affect the proper function of the filtering action and could result in crushed elements. Lack of attention to detail resulted in one of the filter elements being installed without its cartridge guide which could have contributed to significant bypass of unfiltered oil.
Absence of the cartridge guide for an extended period of time would likely result in dislodging of the filter element during variable flow or pressure conditions resulting in this significant bypass of unfiltered oil.
D.
SAFETY ANALYSIS:
The safety of the plant and the public were not affected or challenged before, during, or after this event. The event is classified as a condition prohibited by the Plant's Technical Specifications. Subsequent to the event, it war determined that the plant w.is operated outside the Technical Specification Limiting Condition for Operation (LCO) for a period of :.me longer than that permitted by the Action Statement. The condition was discovered after the allowable time had elapsed and the condition was rectified immediately. The previous successful operability surveillance was performed on August 19,1998. The point in time from 8/19/98 when the operability status of the EDG became questionable could not be determined exactly. However, a review of the operating history of the 1 A EDG indicates the there was indication of a degraded condition on the engine on 9/3/98 when plant operators identified a lifting relief valve in the lube oil system upstream of the strainers.
As part of the historical review for this event, it was determined that the cause of the lifting relief valve was due to the lube oil system blockage at the strainers. This would have effectively created a low lube oil pressure condition in the engine at start-up. The ability of the 1 A EDG to perform its required function with a low lube oil pressure condition is uncertain and therefore, the operability of the EDG is questionable from the identification cf the lifting relief valve on 9/3/98. Byron Station has conservatively determined to report this event as required by 10CFR50.73 as a condition prohibited by the plant's Technical Specifications.
Note that the EDG would have successfully started in the event of an emergency and no other EDG was in an j
inoperable condition at the start and during this event.
l l
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_--m NRC FORM 366A U.s. NUCLEAR REGULATORY Commission was LICENSEE EVENT REPORT (LER)
+
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMsER (6)
PAGE (3)
YEAR SEQUENTIA1 REVISIDN NUMsER NUMsER 7
BYRON NUCLEAR POWER STATION, UNIT 1 06000454 98 -- 018 00 TWimore space is reeurred, use amonel comes of NRC Form 366M (11)
D.
SAFETY ANALYSIS (cont.)
The safety-related function of the EDG is to provide an independent emergency source of power in the event of a complete loss of offsite power. The diesel generator surplies all of the electrical loads which are required for reactor safe shutdown either with or without a Loss of Coolant Accident (LOCA). The nuclear safety concern is the question regarding whether or r.ot the EDG would have been able to start in an emergency as required by the Current Licensing Basis. The EDG is neigned to automatically trip on a low lube oil pressure condition when manually started or when the manual test mode switch is selected at the Main Control Board (MCB) for surveillance testing. This is a function designed to protect the engine from internal damage for those engine runs that are not associated with an emergency start. Additionally, the EDG is designed to auto start if needed during a low lube oil condition at the time of the auto start ESF signal. The lube oil acts in the same manner as cooling water; a continuous circulation of lube oil is provided during standby periods with a lube oil circulating pump. The circulating pump is in continuous operation during standby conditions to maintain the internal parts lubricated and thereby facilitate rapid starts. During engine runs, the main engine-driven lube oil ump takes oil from the oil cooler and to a header which supplies lube oil to variour parts of the engine. Therefore, the EDG would have been able to start if an actual emergency would have presented itself.
Low lube oil pressure does not keep the engine from starting from a standby condition. However, the ability of the 1 A EDG to continue to operate with low lube oil condition is questionable and provides the basis for the determination of historical operability.
The EDG performed well, without incident, during the Loss of Offsite Power (LOOP) event that occurred at Byron Station on August 4,1998, in addition, the 1 A EDG has performed well during 1998 for a total period of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> and 11 minutes which includes the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the LOOP event.
it is concluded that the EDG lube oil system strainers worked as designed. The strainers trapped the foreign material floating in the lube oil. The tube oil system performed its intended function correctly in that it filtered the oil, the instrumentation sensed the low pressure condition and the correct signals to trip the engine and protect it from unnecessary damage, due 11 the low lube oil pressure, were initiated.
A secondary evaluation was performed to evaluate the impact of the 1 A EDG inoperability during the p,eriod of September 3,1998 to September 14,1998. This evaluation determined that additional Technical Specification requirements would have been applicable during the period of the 1 A EDG inoperability besides the LCOAR for TS 3.8.1.1, Action a. An example is Action c.1 of TS LCOAR 3.8.1.1. This action requires that when one diesel generator is determined to be inoperable, it must be verified that "All required systems, subsystems, trains, components, and devices that depend on the remaining Operable diesel generator as a source of emergency power are also Operable,..." otherwise the LCOAR actions must be followed. This verification of other components was not performed from September 3,1998 until'ae failed surveillance on September 12,1998 since there was no indication at the time that the 1 A EDG was inoperable. Once the appropriate LCOAR was entered on September 12,1998 following the failed surveillance, all of the required action contained in TS 3.8.1.1 for AC Sources were followed. Note that the 1 A EDG was not called upon to perform an emergency function from August 19,1998 until the EDG was returned to service on September 14,1998. In addition, the 1B EDG was verified to have been always operable during this period of time. In addition, the emergencyfus cross-tie to t'1e Byron Unit 2 sources of AC power were always availa.ble during this period. Therefore, the health and safety of the public were never affected by this event and it was detennined that the overall safety impact for the plant due to this event was minimal.
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a N,.C FORM 36sA u S. NuCLF AR REGULATORY U.AMISSK)N Nm LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) ooCKET LER NUMBER (6)
PAGE (3)
YEAR SEQUENTML REVISION NUMBER NUMBER J
8YRON NUCLEAR POWER STATION, UNIT 1 05000454 9 8 -- 018 00 TEXT lit more space os required use additional copies of NftC Form 366A) (11l D.
SAFETY ANALYSIS (cont.)
A review of the LCOAR AC Sources - Operating Tech Spec LCO # 3.8.1.1 package was conducted to determine if all of the immediate actions taken successfully completed all of the surveillances required by the respective Tech Spec actions statements. This review concluded that all the respective surveillances as 4
required by the Tech Specs were completed on time and all components were determined to be OPERABLE.
1 E.
CORRECTIVE ACTIONS:
Immediate:
i The immediate corrective actions included:
The immediate entry into LCOAR for A.C. Sources TS 3.8.1.1, action a.
The verification of operability for the redundant EDG including both Unit 2 EDGs.
A complete inspection of each of the 1 A EDG lube oil system components.
An internal inspection of both parallel lube oil strainers.
A complete replacement of all the filter elements.
Lg.nq Term:
n Revise the Mechanical Maintenance Pro :edure BMP 3208 2 to provide specific instructions and inspection requirements for the lube oil filter assembly. (NTS # 454-180 98 SCAQ00018-01)
Systems Engineering will develop a pre-define work instruction to inspect the strainers after every filter element replacement until such time that it is determined this inspection is no longer needed. This will be performed after the lube oil has been re-circutating. (NTS # 454-200-98-SCA000021-04)
Design Engineering willinitiate a modification to provide appropriate indication of strainer pressure and/or Delta pressure to be permanently installed on all four Diesel Generator Engine lube oil systems.
(NTS # 454-200-98-SCAQOOO21-05)
Self Assessment #98-0037 on the 1 A Diesel Generator Lube Oil System failure operability determination will be completed and corrective actions as appropriate will be assigned to communicate lessons learned from the Lube Oil System safety relief valve chattering event that was inadequa'ely evaluated.
(NTS #454 201-98-CAQO2840)
An engineering evaluation to determine on line use of both Lube Oil Strainers versus the use of only one strainer at a time will be conducted and appropriate procedures, drawings, configuration control will be revised accordingly. (NTS #458 901-98-CAQ0?840-03) l l
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NRC F.RM 366A U.s. NUCLEAR REGULATORY COMMisslON ME LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMsER (6)
PAGE (3)
YEAR SEQUENTIAL flEVISION NUMBER NUMBER BYRON NUCLEAR POWER STATION, UNIT 1 05000454 98 -- 018 -- 00 TEXT ll1more spo ce is requored, use ed6toonel copoes of NRC Form 366A) (17)
E.
.C.QR,RfCTIVE ACTIONS (cont.)
R A follow up to the Root Cause Investigation willbe an Effectiveness Review conducted to re analyze this event following the next refueling outage to determine if the probable cause changes and or if the corrective actions were adequate. (NTS # 454-200-98-SCAQ00021-07)
Interim Actions:
/
8 A pressure gauge was installed down stream of the lube oil filter to monitor the lube oil strainer performance with the controls of the Temporary Alteration program. This gauge is inspected at least daily to monitor strainer condition until a permanent modification is installed.
F.
RECURRING EVENTS SEARCH AND ANALYSIS:
A review of previous reportable events was conducted and concluded that there have not been any previous similar events reported on LERs relating to EDG trip due to a low lobe oil system condition.
G.
COMPONENT FAILURE DATA:
None, this event did not involve a component failure.
4 i
l e
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11)n>n Gc,ncr.iting Station 4 450 Wrth Germ.in Church Road
- 11) ntn. IL 6101(t9*94 Tcl HI 54.44-54 4 I October 13, 1998 LTR:
BYRCN 98-0286 FILE:
3.03,0800 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 i
Dear Sir:
The enclosed Licensee Event Report from Byron Generating Station is being re-transmitted to you in accordance with the requirements of 10CFR50.73 (a) (2) (i). A typo was discovered after the LER was mailed.
The wrong number was used.
This report is number 98-018 and not 98-008; Docket No. 50-454.
Sincerely, e
f f11 N
b-tt William Levis i l Station Manage;
[
Byron Nuclear Power Station i
)
WL/JD/cb g,
Enclosure:
Licensee Event Report No.98-018 cc: Carl J.
Paperiello, NRC Region III Administrator NRC Senior Resident Inspector INPO Record Center Comed Distribution List i
e 9810210031 981'013 POR -ADOCK 05000454
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