ML20112H553
| ML20112H553 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/11/1996 |
| From: | Mccollum W DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M95254, NUDOCS 9606180283 | |
| Download: ML20112H553 (13) | |
Text
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I DukeIbus Company WanutR. McColuSt. Jr Catauha Nacient Gemrathn ikpartmerd ficehnident
' MOD C<wicordR<xid (803)3314100 Offite York, SC29715 -
(8W)3314C6 Fat DUKEPOWER
' June-11, 1996
-U.
S. Nuclear Regulatory Commission A'I*rN : Document Control Desk Washington, D. C.
20555 t
Subjects Catawba Nuclear Station Docket No. 50-414 Response To The Preliminary Accident Sequence Precursor Analysis.Of' Loss Of Offsite Power Event At Catawba Unit 2 (TAC M95254)
Thank you for the opportunity to provide comments on Oak Ridge National Laboratory's (ORNL) preliminary Accident Sequence Precursor analysis of the February 6, 1996 loss-of-of fsite power (LOOP) event at Catawba Unit l
2.
We also appreciate the timely preliminary analysis, which facilitates our review and comments.
Attachment A provides the response to the preliminary Accident Sequence Precursor analysis of the February 6, 1996 loss-of-offsite power (LOOP) event at Catawba Unit 2.
Also enclosed are Attachment B (Dominant Core Damage Sequence) and Attachment C (Station Emergency AC Power Reliability Quarterly Report - Fourth Quarter, 1995) which are referenced by Attachment A.
If there are any questions concerning this response, please contact Jeff Lowery, Catawba Regut+cory Compliance, at (803) 831-3414.
M i
A W. R. McCollum,'Jr.
Catawba Nuclear Station Attachments
- l'70029 9606180283 960611
-gDR ADOCK 05000414 i
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U. G. Nuclear Regulatory Commission
,' Juna 11, 1995 Page 2 xc:
S. D.
Ebneter, Regional Administrator, Region II P. S. Tam, ONRR R. J. Freudenberger, Catawba Senior Resident Inspector i
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U. S. Nuclear Regulatory Commission Juns 11, 1995 Page 3 bxc ELL EC050 M. S. Kitlan CN01RC J.
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Snyder MG01RC G. A. Copp EC050 J. E. Burchfield ON03RC P. R. Newton PBOSE J. E. Stoner EC09N K. R. Caraway EC09N P. M. Abraham EC08I R. M. Glover CNO3CE J. W. Caldwell CN03CE R. A. Dickard CNO3CE J. D. Glasser CNO3CE SREC NCMPA-1 NCEMC PMPA Group File: CN-801.01 Document Control File:
801.01 d
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s ATTACHMENT A i
DUKE POWER COMPANY RESPONSE TO THE PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF LOSS OF OFFSITE POWER EVENT CATAWBA NUCLEAR STATION, UNIT 2
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e ATTACIDENT A
Response
To The Preliminary Accident Sequenco Precursor Analysis I,oss Of Offsite Power Event Catawba Nuclear Station, Unit 2 Ccessent 1.
Modeling Approach In the preliminary ORNL analysis only Emergency Diesel Generator l
(EDG) 2A was considered available for the mission time (7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) of interest.
EDG 2B, which was in maintenance at the time of the event but returned to rervice at about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, is treated as in maintenance but potentially recoverable, with a recovery probability of 0.48.
An alternate approach, which perhaps more closely resembles the actual post-event condition, would be to consider only one EDG to i
be available during the first three hours, with potential recovery of the other EDG, and both EDGs to be available subsequently.
Attachment B presents the development of the dominant related sequence.
1 At the time of the event, EDG 2B was out of service to perform maintenance on the EDG 2B battery charger.
In the event EDG 2B was needed after the LOOP event because no other source of ac power was readily available, plant personnel would have attempted to place EDG 2B into service by clearing the out-of-service tags and closing the breakers.
EDG 2B battery is considered to have adequate capacity to start the EDG without the charger.
The
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estimated time to place EDG 2B into a functional status for this scenario is estimated to be in the range of 1 to 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Since EDG 2A was supplying the load, EDG 2B was not needed.
It was placed into operation at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 51 minutes after the initiating event.
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i comument 2.
EDG Reliability l
The-preliminary ORNL analysis uses the EDG start and run failure probabilities of 0.03 and 0.002/hr, respectively.
The estimated Catawba EDG start and run failure probabilities, as reported in 1
the Catawba IPE are 0.007 and 0.0046/hr, respectively.
In fact, the current 3-year average values, as shown in Attachment C, are 0.003 and 0.0015/hr.
Use of the current plant-specific values should change the preliminary conditional core damage probability from 0.0033 to approximately 0.001 without any other changes.
Comunent 3. SSF Model The preliminary ORNL analysis attempts to include the Standby Shutdown Facility (SSF) feature to mitigate a loss of all ac power condition.
However, inclusion of both the SSFB and SSFC logic is not correct.
The SSFC cut sets are a subset of the SSFB cut sets. (Please see Table A.
18-7 of the Catawba IPE.)
The j
main differences between SSFB and SSFC are that (i) SSFB requires operator action within 15 minutes, while for the SSFC case, action can be delayed for up to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and (ii) SSFB contains J
additional failure modes involving the Reactor Coolant pump seal injection components.
In the Catawba Probabilistic Risk Assessment (PRA), success of SSFC does not prevent a core melt; it simply changes the plant damage state.
In addition, cut sets representing unavailability of.the SSF due to maintenance (2.57E-2) and maintenance on the SSF diesel generator (2.96E-3) could be deleted when determining the probability associated with SSFB, since this equipment was available during the event.
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Thus, SSFC cut sets and SSF maintenance events should be deleted from the ORNL preliminary analysis.
Incorporating these suggestions results in an SSF failure probability of 0.19.
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l Comment 4.
Power Availability From Unit 1 Catawba has two 4 kV transformers (SATA and SATB) which can power the two essential 4 kV switchgears in one unit from the ac power system from the other unit.
The operator action to make use of this feature is contained in the plant emergency procedure, and operators are trained on this action. Catawba's estimate to perform this action is 30 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Considering that this action is required in the emergency procedure, that the operators are given training on it, and that the available time is about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the operator failure probability is not considered to be significant.
The Duke calculations use 0.17 as the failure probability of this event, derived primarily from the data on LOOP events where more than one unit suffered a LOOP in multi-unit sites.
Even with the assumption of a 30 minute cognitive time and a 90 minute action time, the Human Cognitive Reliability model (Hannaman et.al., " Human Cognitive Reliability Model for PRA Analysis," NUS-4531, December 1994) yields the operator failure probability as 0.03, with the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> available time.
Therefore, the value of 0.35 used in the preliminary ORNL analysis for failure probability of using the Unit 1 ac power when it was indeed available seems a bit conservative.
It should be noted that the procedural difficulty encountered during the event should not be viewed as a significant factor since one EDG was operating and supplying necessary loads and the use of Unit 1 ac power was not critically needed at that time.
The Duke analysis of this event, based on Catawba-specific EDG reliability data, the two distinct EDG availability representations, and the applicable SSF failure modes, yielded a conditional core damage probability of approximately 4 E-4.
This calculation is based on a base case failure probability of Unit 1 power of 0.17.
Since offsite power was available through Unit 1 throughout the mission time, the 0.17 value is conservative.
As a sensitivity analysis, changing this value to 0.03 and then to 0.5 produces results of 7 E-5 and 1 E-3, respectively.
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Commsent 5.
Description of EDG Run Tisse The "Modeling Assumptions" section states that a mission time of 7.5 h was used for EDG 2B.
It appears that this mission time was actually used to compute the failure probability of the EDG 2A.
Comument 6.
Auxiliary Feedwater Pump Floor Drain Leakage The following description of the auxiliary feedwater (CA) pump room and associated floor drain system is provided to understand the significance of leakage found in the floor drain sump.
The three CA pumps are located in the CA pump room, which has a floor drain sump (4x3x7 ft) with two load-shed floor drain sump pumps.
Each CA pump is located in a pit (17x17x14.5 ft) with two sump pumps for the turbine driven pump and one sump pump for each of the motor driven CA pumps. The CA pump pit sump pumps (each with 50 gpm capacity) are powered by the essential ac power system, and one of the turbine driven CA pump sump pumps is backed up by the SSF power.
i The CA pump pit sump pumps and the CA pump room sump pumps discharge into a common header.
During the LOOP event some water was found to be accumulating in the CA pump room floor and is attributed to leakage of check valves between the room sump pumps and the common header for this floor drain system.
The leakage was estimated to be no greater than 20 gpm, the expected maximum floor drain requirement for the turbine driven CA pump pit.
Considering that the floor area outside the CA pump pits is about 2231.6 square foot and that the curb is about 1.5 feet high, the estimated time for the leakage to spill into the CA pump pits is about 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.
The leakage into the turbine driven pump pit is within the capability of the operating sump pump.
If this sump failed, an additional 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> would be available before the leakage or the water accumulation 1
in the turbine driven pump pit could fail the pump.
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l For the motor driven CA pump B pit, the estimated time to fill the pump pit is about 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> once the leakage starts into the pit and if the sump rump is not operating.
Thus the leakage into the CA pump room from part of the CA pump pit floor discharge should not influence the mission times and success criteria for CA punps or battery depletion time considerations.
It is recommended that the preliminary ORNL analysis be revised to take into account the comments provided herein.
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ATTACHMENT B 5
DUKE POWER COMPANY j
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RESPONSE TO THE REVIEW OF PRELIMINARY ACCIDENT SEQUENCE l
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PRECURSOR ANALYSIS OF LOSS OF OFFSITE POWER EVENT CATAWBA NUCLEAR STATION, UNIT 2
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DOMINANT CORE DAMAGE SEQUENCE i
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ATTACHMENT B Dominant Core Damage Sequence
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First 3 bra follooring LOOP event Nant 4.5 bra Failure Event Probability Failure Event Probability LOOP occurs 1.0 LOOP occurs 1.0 EDG 2B is in 0.48 EDG 2B fails to 0.00975 i
maintenance and is not start or run recovered (0.003 + 4.5 h*0.0015/h)
EDG 2A fails to start 0.0075 EDG 2A fails to run 0.00675 or run (4.5 h*0.0015/h)
(0.003 + 3 h*0.0015/h)
SSF fails to provide 0.19 SSF fails to provide 0.19 seal inj.
seal inj.
Operators fail to 1.0 Operators fail to 1.0 recover offsite power recover offsite j
power Operators fail to 0.17 operators fail to 0.17 recover using Unit 1 recover using Unit 1 power power Subtotal:
1.2E-4 Subtotal:
2.1E-6 Total Sequence Probability = 1.2E-4 i
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ATTACHMENT C i
DUKE POWER COMPANY RESPONSE TO THE REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF LOSS OF OFFSITE POWER EVENT CATAWBA NUCLEAR STATION, UNIT 2 l
STATION EMERGENCY AC POWER RELIABILITY REPORT (FOURTH QUARTER, 1995) 1 l
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Attachment C j
i Station Emergency AC Power Reliability j
Quarterly Report j
(Fourth Quarter,1995)
KELIABILI1 Y CALCULATIONS (1959 - 1993)
McGUIRE TOTAL D/G DEMANDS l 1014 l MNS TOTAL D/G DEMAND FAILURES l
9 l D/G Fails to Start on Demand =
0.0089 /d 3Yr Rolling Average (93 95) =
0.0025 /d TOTAL D/G LOAD RUN IIOURS l 1692.4 l TOTAL D/G RUN FAILURES (note 1) l 7
l D/G Fails to Run =
0.0041/hr 3Yr Rolling Average (93 95) =
0.0046 /hr AVG D/G UNAVAILABLE 110URS l 680.6 l AVG CRITICAL llOURS [(U1 + U2)/2]
l 47422.6 l D/G Unavailability (notes 2&4) =
1.44 %
3Yr Rolling Avg. [93 95] (notes 2&4) =
0.83 %
CATA WBA TOTAL D/G DEMANDS l 827 l CNS TOTAL D/G DEMAND FAILURES l
4 l D/G Fails to Start on Demand =
0.0048 /d j
3Yr Rolling Average (93 95) =
0.0030 /d TOTAL D/G LOAD RUN liOURS l 1590.6 l TOTAL D/G RUN FAILURES (note 3) l 2
l D/G Fails to Run =
0.0013 /hr 3Yr Rolling Average (93-95) =
0.0015 /hr AVG D/G UNAVAILABLE ilOURS l 1138.9 l AVG CRITICAL llOURS [(Ul + U2)/2]
l 49602.1 l D/G Unavailability (notes 2&4) =
2.30 %
3Yr Rolling A vg. [93 95] (notes 2&4) =
0.74 %
OCONEE TOTAL UNIT DEMANDS l 4289 l ONS TOTAL UNIT DEMAND FAILURES l
22 ] Keowee Unit Fails to Start =
0.0051/d 3Yr Rolling Average (93 95) =
0.0032 /d TOTAL UNIT RUN 110URS l 7608.8 l TOTAL UNIT RUN FAILURES l
5 l Keowee Unit Fails to Run =
0.00066 /hr i
3Yr Rolling Avg (93 95) =
0.00064 /hr AVG UNIT UNAVAILABLE 110URS l 891.1 l Keower Station Unavailability (KPRA value)=
0.52 %
TOTAL PERIOD llOURS l 61345.0 l Keowee Unit Unavailability (notes 2&4) =
1.45 %
3Yr Rolling Avg. [93 95] (notes 2&4) =
1.57 %
l NOTES:
I - Total number of McGuire D/G run failures reduced by one due to data QA process.
2 Estimated hours excluded from unavailability calculation.
3 - Catawba D/G run failures increased by one due to D/G 2B run failure in the 4th quarter.
4 - Unavailability reported on per D/G or per Keowee unit basis.
Page C-1 BY: CMM 1/31/96