ML20217Q987
| ML20217Q987 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/01/1998 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1920-98-20236, NUDOCS 9805130065 | |
| Download: ML20217Q987 (7) | |
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GPU Nuclear, Inc.
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floute 441 South NUCLEAR Post Office Box 480 Middletown, PA 17057-0480 Tel 717-944 7621 May 01,1998 3
1920-98-20236 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 GPU Nuclear Review of Preliminary Accident Sequence Precursor Analysis : f Opert.tional Event at Three Mile Island Nuclear Station, Unit No.1 As requested in your March 30,1998 letter, GPU Nuclear has reviewed the preliminary Accident Precursor Sequence (ASP) analy sis of the operational event that occurred at TMl-1 on June 21, 1997. Consideration of the technical adequacy of the preliminary ASP analysis, including its depiction of plant equipment and the equipment capabilities, resulted in comments in the attachment which are provided and organized in accordance with the guidance questions of the letter's Enclosure 2.
Ifyou have any questions regarding this review, please contact William Heysek of the TMI Nuclear Safety and Licensing Department at (717) 948-8191.
Sincerely, SUk LO.
m V"~t er.
James W. Langenbach
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Vice President and Director, TMI
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WGH Attachment cc: Administrator, Region 1 - Hubert J. Miller TMI Senior Resident Inspictor - Wayne L. Schmidt File 98070 9905130065 990501 PDR ADOCK 05000289 S
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1920-98-20236 Attachment Page lof 6 Review of the Preliminary Accident Sequence Precursor Analysis for LER Nos. 289/97-007,008, and 010 The Preliminary Accident Sequence Precursor Analysis for LER Nos. 289/97-007,008, and 010 was reviewed and comments have been organized in accordance with the guidance questions in :
1.
Does the " Event Description" section accurately describe the event as it occurred?
Page 1, under Event Summary, the fifth sentence states,"The unit was cooled by a.
natural circulation cooling until offsite power was restored." This sentence should i
state, "The unit was cooled by natural circulation cooling until offsite power and forced cooling were restored." Per LER 97-009 pages 4 and 5 it took 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for offsite power to be restored. Another 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> passed before forced cooling (RC-P-1 A/B/C/D) was restored.
b.
Page 1, under Event Description, first paragraph, second sentence states, "On June 21,1997 the B phase of the 230-kv power transformer developed a fault causing severe overheating and the subsequent ejection of the bushing and conductor from the breaker housing of output breaker GBI-02 (Fig.1)." This sentence should state, "On June 21,1997 the B phase of the 230-kv generator output breaker GBI-02 (Fig.1) developed a fault causing severe overheating and the subsequent ejection of the bushing and conductor from the breaker housing." Per LER 97-007 there is no evidence that the fault location was in the power transformer.
c.
Page 1, under Event Description, second paragraph, second to the last sentence states, "All control rods inserted to the % insertion position within 2.2 s." This sentence should state, "All control rods inserted to the % insenion position well I
within 3.0 s." Per LER 97-008 there is no claim made that at the time of the reactor trip that all control rods inserted to the % position within 2.2 s. The only claim is that re-tested rods came in at less than 2.2 seconds. The LER does indicate that the rods did insert to the % position well within 3.0 s at the time of the reactor trip.
d.
Page 1, under Event Description, third paragraph, second sentence states,"Non-vital loads, including circulating water and main condenser vacuum pumps, were not energized." This sentence should state,"Non-vitalloads, including main feedwater, condensate, circulating water, and main condenser vacuum pumps, were not energized." The more :mportant non-vital loads per LER 97-007 page 2 were omitted from the event description.
i e.
Page 2, under Event Description, last paragraph, last sentence states, "After the operators.., the reactor coolant pumps were restaned, returning the reactor
1920-98-20236 Attachment Page 20f 6 coolant system to forced circulation cooling." This should state, "After the operators., the reactor coolant pumps were restarted. Since the natural circulation cooling of the reactor coolant system was stable, the reactor coolant system was not returned to forced circulation cooling until nine hours afler the reactor trip."
f.
Page 6, Fig.1; change breaker identification from 18-14 to IB-12 per Gilbert / Commonwealth drawing 229-002.
2.
Does the " Additional Event-Related Information" section provide accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?
a.
Page 2, under Additional Event-Related Information, the first and second paragraphs are considered to place too much emphasis on the PORV being inoperable. The Reactor Coolant System pressure did not approach the PORV setpoint at any time during the LOOP event. Therefore, the PORV being inoperable was not a factor during the LOOP event.
b.
Page 2, under Additional Event-Related Information, last paragraph states, the second and third sentences state " Additionally one EDG from TMl 2 is available as an alternate ac power source during a station blackout (SBO). The alternate EDG which is manually started from the control room, can be aligned to either i
engineered safeguards bus 1D or lE within 10 min following an SBO." These sentences more correctly should state," Additionally one EDG previously from TM12... The alternate EDG which is manually started from the control room, can be aligned to either engineered safeguards bus 1D or 1E or balance of plant (BOP) bus 1C within 10 min following an SBO."
Page 2, under Additional Event-Related Information, last paragraph states, c.
" Operators must close two breakers and open any desired loads must be manually loaded onto the bus selected to be re-energized." This sentence should state,
" Operators must close two breakers and open/ lockout two breakers and any desired loads must be manually loaded onto the bus selected to be re-energized."
Per abnormal procedure 1202-2, the operator must open/ lockout two breakers in addition to closing two breakers to correctly align the SBO diesel with bus ID or I E.
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3a)
Does the "Modeling Assumptions" section accurately describe the modeling done for the i
event?
a.
Page 3, first paragraph should give a reference for the reliability value used for OEP-XHE-NOREC-SB in table 1.
1920-98-20236
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Attachment Page 30f 6 bl T'he two events described in Table 1 of Reference 1 EPS-HXE-NOREC (Operator Fails to Recover Emergency Power) and OEP-XHE-NOREC-SL (Operator Fails to Recover Offsite Power Before RCP Seals Fail) were substituted by the Nonrecovery values REA and REC given in Table B.1-12 of Reference 2. REA is 3
the nonrecovery factor for station blackout sequences in which the Emergency Feedwater operates successfully. REC is the nonrecovery factor for station blackout sequences in which the Emergency Feedwater does not operate 4
l successfully. Both REA and REC were evaluated under the assumption that at least two emergency diesel generators are recoverable.
i REA and REC were calculated by Monte Carlo simulation techniques as described in section B.l.3.4 of reference 2. Recovery of the TMl-1 diesel was not credited in the simulation. The result of the simulation shows that if EFW is unavailable, the mean time to core uncovery is 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 10.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the cc.se where EFW is available. The time interval of 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to core uncovery mentioned in Reference 2 was used as input to the STADIC code and was assumed to occur after loss of all AC power. The STADIC model did not assume that all onsite AC power would fail at time t=0 when offsite power is lost In other words, the time to core uncovery was assessed in (Reference 2, page B.1-16) as the time of onsite power failure, plus 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
j 3b) is the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions? This also includes assumptions regarding the likelihood of equipment recovery.
In general the probabilities for failure of restoration of offsite power, onsite a.
emergency power, and reactor coolant pump seals may be reduced from that assumed in the analysis because o' erators have the following procedures for p
guidance. Also, the 6/21/97 event was successfulin that offsite power was restored in 90 minutes, emergency onsite sources started and loaded as designed, and there was no seal damage.
Emergency Procedure 1202-2, Rev. 44, " Loss of Station Power", which was used during the TM1-1 event, includes steps (3.10 and Attachment 4) for restoring off site power. Because of these operator actions the probability of not restoring oft-site power (IE-LOOP) should be lower than 5.0 E-001.
Emergency Procedure 1202-2, Rev. 44, " Loss of Station Power", which was used during the TMI-l event, includes steps (3.11 and Attachment 1) l for starting and loading the emergency diesel generators (EDGs) assuming j
the EDGs did not automatically start and load. In addition, the procedure l
includes steps (3.12 and Attachment 1) for starting and loading the SBO l
diesel assuming the Class 1E diesel (s) did not automatically start. Because
l 1920-98-20236 Attachment Page 4cf 6 of these operator actions the probability of all diesels failing (EPS-DGN-CF-ALL) should be lower than 9.5 E-004.
Emergency Procedure 1202-2, Rev. 44, " Loss of Station Power", which was used during the TM1-1 event, includes Attachment 2 which may be used to restore reactor coolant pump seals. The restoration would be performed in a controlled manner to limit damage to the seals and pumps if injection and intermediate closed cooling water (ICCW) are lost for a long time period. Because of these operator actions the probability of RCP seal failure (RCS-MDP-LK-SEALS) should be lower than 4.0 E-002.
Based on the variation in the combination of events and modeling assumptions used by the NRC in the analysis of the operational event at TMI as described in the preliminary ASP, the estimate of Conditional Core Damage Probability of the Loss of Offsite Power (LOOP) was determined to be 9.6x10' Using the combination of events and modeling assumptions of the TMI-l PRA (Level 1) Update, December 1992, " Appendix B - Special Analysis" which gives consideration to both the REA and REC factors, GPU Nuclear estimates the Conditional Core Damage Probability of the Loss of Offsite Power (LOOP) to be 9.53x10
1920-98-20236 Attachment Page SCf 6 Tctal (all sequences) 9.53E-07 j
Cut Set DESCRIPTION Probability APUN-Cut Set #1 Sequence 26 EPS-DON-CF-ALL, RCS-MDP-LK-SEALS, IE-LOOP, REA 1.58E-07 EPS-DGN-CF-ALL Common Cause Failure of EDGs 9 50E-04 RCS-MDP-LK-SEALS RCP Seals Fail Without Cooling and injection 4.00E-02 lE-LOOP ~
Initiating Event LOOP 5.00E-01
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REA' Operator fails to recover Electric Power (2EDG EFW available) 8.31 E-03 Cut Set DESCRIPTION Probability GPUN4ut Set #2 Sequence 26 EPS-DONfC-1A*EPS-DGN-FC-1B*RCS-MDP-t.K-SEALS *lE-LOOP *EPS XHEfC-AAC*REA 1.23E48 i
EPS-DGN-FC-1 A 1 A EDG fails to Start and Run 4.20E-02 EPS-DGN-FC-1 B I B EDG fails to Start and Run 4.20E-02 EPS-DGibFC ACC Alternate ac EDG Fails to Start and Run 4 20E-02 RCS-MDP-LK-SEALS RCP Seals Fail Without Cooling and injection 4.00E-02 IE LOOP Initiating Event-LOOP 5.00E-01 REA Operator fails to recover Electric Power (2EDG, EFW availab!e) 8.31 E-03 Cut Set DESCRIPTION Probability GPUN-Cut Set #3 Sequence 26 EPS-DGN-FC-1A*EPS-DGN-FC 1B*RCS-MDP-LK-SEALS *lE-LOOP *EPS XHE XM-AAC"RdA 2 93E 09 EPS-DGN-FC-1 A 1 A EDG fails to Start and Run 4.20E-02 EPS-DGN-FC-1 B 1B EDG fails to Start and Run 4.20E-02 RCS-MDP-LK-SEALS RCP Seals Fail Without Cooling and injection 4.00E-02 IE LOOP Initiating Event. LOOP 5.00E-01 EPS-XHE-XM-AAC Operator Fails to Start or Load the AAC EDG 1.00E-02 REA Operator fails to recover Electric Power (2EDG, EFW available) 8.31 E-03 Cut Set DESCRIPTION Probability 2 PUN-Cut Set #1 Sequence 41 EPS-DON-CF-ALL, EFW TDP-FC-TDP, EFW XHE-NOREC-EP, IE-LOOP, REC 4.37213E47 EPS-DGN-CF ALL Common Cause Failure of EDGs 9.50E-04 EFW-TDP-FC-TDP EFW Turbine Driven Pump fails 3.20E-02 EFW-XHE-NOREC-EP Operator fails to Recover EFW Dunng SBO 3 40E-01 IE-LOOP Initiating Event-LOOP 5 00E-01 REC Operator fails to recover Electric Power (2EDG, EFW unavallable) 8 46E-02 h
1920-98-20236 Attachment Page 60f 6 Cut Set DESCRIPTION Probability GPUN-Cut Set #2 Sequence 41 EPS-DGN-FC 1A, EPS-DONfC-1B,EFW-TDP-FC TDP, EFW X'iE-NOREC-EP,IE-LOOP, REC,EPS. 3.40971E-08 DGNTC-ACC IE-LOOP Initiating Event. LOOP 5.00E-01 EFW-TDP-FC-TDP EFW Turbine Driven Pump fails 3.20E-02 EFW-XHE-NOREC-EP Operator fails to Recover EFW During SBO 3.40E-01 EPS-DGN-FC-1 A 1 A EDG fails to Start and Run 4.20E-02 i
EPS-DGN-FC-1 B 18 EDG fails to Start and Run 4.20E-02 EPS-DGN-FC-ACC Alternate ac EDG Fails to Start and Run 4.20E-02 REC Operator fails to recover Electric Power (2EDG, EFW unavailable) 8.46E-02 Cut Set DESCRIPTION Probability gpun-Cut Set #3 Sequence 41 EPS-DON-FC-1A, EPS-DON-FC-1B,EFW TDP-FC-TDP, EFW XHE-NOREC-EP,IE LOOP, REC,EPS.
1.62367E-C8 l
XHE-XM-ACC EPS-DGN-FC-1 A 1 A EDG fads to Start and Run 4.20E-02 EPS-DGN-FC-1 B 1B EDG falls to Start and Run 4.20E-02 EFW-TDP-FC-TDP EFW Turbine Driven Pump fails 3.20E-02 EFW-XHE-NOR EC-EP Operator fails to Recover EFW During SBO 3.40E-01 EPS-XHE-XM-AAC Operator Fails to Start or Load the AAC EDG 1.00E-02 REC Operator falls to recover Electric Power (2EDG, EFW unavailable) 8.46E-02 Cut Set DESCRIPTION Probability l
IPUN-Cut Set #1 Sequence 19 Same as ORNL 1.30E-07 j
GPUN Cut Set #2 Sequence is Same as ORNL 1.00E-08 j
aPUN-Cut Set #3 Sequence 19 Same as ORNL 2.60E-09 AH other sequences (ORNL)
Same as ORNL 1.60E-07 l
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