ML20135E566
| ML20135E566 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/04/1997 |
| From: | Chandu Patel NRC (Affiliation Not Assigned) |
| To: | Dugger C ENTERGY OPERATIONS, INC. |
| Shared Package | |
| ML20135E569 | List: |
| References | |
| TAC-M77487, NUDOCS 9703070148 | |
| Download: ML20135E566 (11) | |
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n UNITED STATES a
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NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 30666-0001 March 4, 1997 i
Mr. Charles M. Dugger Vice President Operations Entergy Operations, Inc.
P. O. Box B Killona, LA 70066
SUBJECT:
REVIEW OF WATERFORD STEAM ELECTRIC STATION, UNIT 3, INDIVIDUAL PLANT EXAMINATION SUBMITTAL - INTERNAL EVENTS (TAC. NO. M77487)
Dear Mr. Dugger:
Enclosed is our Staff Evaluation Report (SER) of the Waterford Steam Electric Station, Unit 3 (Waterford 3), Individual Plant Examination (IPE) submittal for internal events and internal flood.
A " Step 1" review was performed which examined the IPE results for their
" reasonableness" considering the design and operation of Waterford 3.
The staff employed Brookhaven National Laboratory to review the front-end i
analysis, human reliability analysis, and back-end analysis of the IPE submittal. Their Technical Evaluation Report (TER) is attached as an appendix to the SER. This contractor TER was reviewed by the IPE " Senior Review Board" (SRB) as part of Office of Nuclear Regulatory Research (RES), quality assurance process. The SR'3 is comprised of RES staff and consultants at Sandia National Laboratory with Probabilistic Risk Assessment (PRA) expertise.
The IPE has estimated a core damage frequency (CDF) of 1.7E-05 per reactor-year from internally initiated events, plus an additional contribution from internal floods of 1.1E-06. Transients (including loss of offsite power) contribute 52 percent to the CDF; loss of coolant accidents (LOCAs),
40 percent; internal floods, 7 percent; steam generator tube rupture (SGTR),
5 percent; interfacing systems LOCA (ISLOCA), 3 percent; anticipated transients without scram (ATWS), 1 percent.
(The total exceeds 100 percent because internal flood is not part of the reported CDF.)
The Entergy Operations, Inc. (E01) defined a vulnerability as either an extremely high sequence CDF (substantially greater than IE-04), a greater than 50 percent contribution to CDF from a single sequence, or an event that contributes in an unusual or substantial way to the risk profile.
Based on this definition, E01 did not identify any vulnerabilities.
In addition, no plant improvements were identified, hs tu CENIER GOPY 9703070148 970304 PDR ADOCK 05000382 P
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Mr. Charles M. Dugger Based on the " Step 1" review, we conclude that the E01 has met the intent of Generic Letter 88-20, although we have noted a few weaknesses in the submittal that may limit its future usefulness without modifications. A " Step 2" review will not be conducted by the Nuclear Regulatory Commission staff.
Sincerely, dMM O [
Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-382 i
Enclosure:
Staff Evaluation Report w/ attachment cc w/ encl:
See next page 2
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. a Mr. Charles M. Dugger -,
Based on the " Step 1" review, we conclude that the E01 has met the intent of Generic Letter 88-20, although we have noted a few weaknesses in the submittal that may limit its future usefulness without modifications. A " Step 2" review will not be conducted by the Nuclear Regulatory Commission staff.
Sincerely, ORIGINAL SIGNED BY:
Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-382
Enclosure:
Staff Evaluation Report w/ attachment cc w/ encl:
See next page DISTRIBUTION i
t0ccket File 71 PUBLIC PD4-1 r/f C. Patel W. Beckner C. Hawes J. Roe J. Dyer, RIV E. Adensam (EGA1)
'T. Harris (TLH3)
Document Name: WAT77487.LTR '
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Mr. Charles A. Dugger Entergy Operations, Inc.
Waterford 3 cc:
Administrator Regional Administrator, Region IV Louisiana Radiation Protection Division U.S. Nuclear Regulatory Commission Post Office Box 82135 611 Ryan Plaza Drive, Suite 1000 Baton Rouge, LA 70884-2135 Arlington, TX 76011 Vice President, Operations Resident Inspector /Waterford NPS Support Post Office Box 822 Entergy Operations, Inc.
Killona, LA 70066 P. O. Box 31995 Jackson, MS 39286 Parish President Council St. Charles Parish Director P. O. Box 302 Nuclear Safety & Regulatory Affairs Hahnville, LA 70057 Entergy Operations, Inc.
P. O. Box B Executive Vice-President Killona, LA 70066 and Chief Operating Officer Entergy Operations, Inc.
Wise, Carter, Child & Caraway P. O. Box 31995 P. O. Box 651 Jackson, MS 39286-1995 Jackson, MS 39205 Chairman General Manager Plant Operations Louisiana Public Service Commission Entergy Operations, Inc.
One American Place, Suite 1630 P. O. Box B Baton Rouge, LA 70825-1697 Killona, LA 70066 Licensing Manager Entergy Operations, Inc.
P. O. Box B Killona, LA 70066 a
Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 1
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006Nm01
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WATERFORD 3 NUCLEAR POWER PLANT INDIVIDUAL PLANT EXAMINATION STAFF EVALUATION REPORT I.
INTRODUCTION In August 1992, Entergy Operations, Inc. submitted the Wate'rford Steam Electric Station, Unit 3 Naterford 3), Individual Plant Examination (IPE) submittal in response to Generic Letter (GL) 88-20 and associated supplements.
On January 22, 1996, the staff sent questions to the licensee requesting additional information. The licensee responded by letters dated April 30, 1996, and August 29, 1996.
A " Step 1" review of the Waterford 3 IPE submittal wa', performed including the efforts of contractor Brookhaven National Laboratr,ry in helping the staff evaluate the front-end analysis, back-end analysis, and human reliability 4
analysis (HRA) portions of the submittal. The Step 1 rev;m fccused on whether the licensee's method was capable of identifying vulnerabilities.
Therefore, the review considered:
(1) the completeness of the information, and (2) the reasonableness of the results given the Waterford 3 design, i
operation, and history. A more detailed review, a " Step 2" review, was not performed for this IPE submittal. A summary of staff's findings is provided below. Details of the contractor's findings are in the attached technical evaluation report (Appendix) of this staff evaluation report (SER).
In accordance with GL 88-20, Waterford 3 proposed to resolve Unresolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements." No other specific USIs or generic safety issues (GSIs) were proposed for resolution as part of the Waterford 3 '/E.
I The submittal states that the licensee intends to maintain a "living" probabilistic risk assessment.
II.
EVALUATION Waterford 3 is a Combustion Engineering pressurized-water reactor with a large dry containment.
The Waterford 3 IPE has estimated a core damage frequency (CDF) of 1.7E-05 per reactor-year from internally initiated events, plus an additional contribution from internal floods of 1.lE-06.
The Waterford 3 CDF compares reasonably with that of other Combustion Engineering plants.
Transients (including loss of offsite power) contribute 52 percent to the CDF; loss of coolant accidents (LOCAs), 40 percent; internal flood, 7 percent; l
steam generator tube rupture (SGTR), 5 percent; interfacing systems LOCA (ISLOCA), 3 percent; anticipated transients without scram (ATWS), I percent.
i (The total exceeds 100 pvcent because internal flood is not part of the reported CDF.)
ENCLOSURE
, The most important system / equipment basic event contributors to the estimated CDF are:
(1) loss of offsite power nonrecovery (within 50 minutes and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), (2) failure of turbine driven emergency feedwater pump to start, (3) emergency diesel generator failure to start and/or run, (4) common cause failure of the emergency diesel generators to run, (5) common cause failure of the containment sump recirculation valves to operate, (6) dry cooling tower maintenance unavailability, and (7) failure of auxiliary component cooling water pump.
The licensee's Level 1 analysis examined significant initiating events and dominant accident sequences, and appears complete.
In addition the initiating event frequencies are generally reasonable and in agreement with other Probabilistic Risk Assessments (PRAs). However, some weaknesses, discussed below, have been identified.
I 1.
The IPE took credit for recovery of offsite power. However, compared to Electric Power Research Institute (EPRI) document NSAC-147 which contains industry average data, Waterford 3 appears to be quite optimistic regarding the likelihood of power recovery. As measured by non-recovery probability, they appear to be between 2.5 and 10 times more optimistic than average data indicates. Sensitivity studies indicate that raising this non-recovery an order of magnitude, to a range more in keeping with the rest of the nuclear industry, raises the CDF by about a factor of three.
2.
Waterford 3 data are generally in agreement with NUREG/CR-4550 data.
However, the turbine driven pump failure to run rate is about two orders of magnitude lower than NUREG/CR-4550. This data was used for both the main feedwater and emergency feedwater pumps, the former runs continuously while the latter does not. The licensee stated that their failure rate data is based on four relavent generic data sources, all of which have a much lower failure rate than NUREG/CR-4550, which the licensee believes is too high, especially in light of the fact that, they report, their turbine driven main feedwater pumps, running continuously, demonstrate good reliability.
Using the higher failure rate for the emergency feedwater pump from NUREG/CR-4S50 increases the contribution of station blackout, currently estimated to be about 37 percent of CDF, to approximately 50 percent, and would also increase the overall CDF slightly. We believe the licensee should reexamine their turbine driven pump failure rate to ensure that it is based on the most appropriate data souce and is consistent with the equipment at Paterford 3.
3.
Common cause failure quan' tcaticn appears to be somewhat inconsistent.
On the one hand, comme f.e probabilities that were used were based on the procedure presentc6 NURE0/CR-4780 and the data presented in EPRI NP-3967. However, some equipment were not included.
For instance, chillers were stated tc have no common cause dependencies with each other if they were operated differently during operation, i.e., continuously every month (chiller B) rather than alternating monthly (chillers A and AB). This may be overly optimistic.
Similarly, other PRAs have treated
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i common cause failures which were neglected by Waterford 3, such as, circuit breakers (other than reactor trip breakers), electrical i
switchgear, air operated valves, air compressors, inverters, relays, and transmitters (except for miscalibration). Common cause failures between i
emergency feedwater pumps of different driver type, i.e., motor versus turbine, were not considered credible, either. However, this type of dependency has been modeled in other PRAs.
In conclusion, while the common cause failure rates used were generally reasonable, the omission of various equipment to common mode failures may cause the reported CDF to be somewhat understated.
f Based on the licensee's IPE process used to search for decay heat removal l
(DHR) vulnerabilities, and review of Waterford 3 plant-specific features, the staff finds the licensee's DHR evaluation consistent with the intent of the USI A-45 (DHR Reliability) resolution and is, therefore, acceptable.
The licensee performed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure j
events. The HRA considered both pre-initiator and post-initiator human actions. For pre-initiators, Waterford 3 considered both types of pre-initiators:
failure to restore systems after test, maintenance, or surveillance activities, and instrument calibrations.
Post-initiator human actions are those required in response to initiating events or related systems failures.
j The licensee identified the following operator actions (all but the last were post-initiator type) as important in the estimate of the CDF:
(1) Operator fails to recover from room cooling failure, (2) Operator fails to align HPSI pump train AB, (3) Operator fails to properly align emergency feedwater suction, and (4) Operator fails to restore air cooling unit after test / maintenance.
J Some weaknesses have been identified in the HRA analysis and are discussed i
below.
1.
Plant-specific performance shaping factors, reflected in the HRA by
" success likelihood indices," were assumed to have no effect on all but two events and, there, probabilities of human failure were lowered to reflect expected good perfomance.
By leaving success likelihood indices at their default values the analysts assumed Waterford 3 is an average plant. The result of this is that the analysis is somewhat generic rather than plant specific and may not completely represent the plant, since there was no evidence that detailed analysis was performed to assure the appropriateness of the generic values.
2.
Consideration of dependencies between separate tasks was essentially ignored by assuming they were independent. The licensee stated that task independence existed because of separation in time between the tasks and in the people performing the tasks. Howeve-in some cases, it may not be
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. i appropriate to assume independence in this manner since the context of the accident and the pattern of successes and failures can influence the probability of subsequent human error.
j 3.
There appears to have been no simulator exercises conducted to evaluate operator response nor walkdowns conducted to determine response times for operator actions outside the control room. These were based on interviews with operators and examination of procedures and training. Walkdowns were conducted, however, for the flooding analysis and to resolve specific front-end modeling issues.
In general, however, the application of the HRA method chosen by the licensee did not result in excessively low human error probabilities, nor do they suggest that identification of human action vulnerabilities was precluded, despite the limitations discussed.
The licensee evaluated and quantified the results of the severe accident progression through the use of a containment event tree and considered uncertainties in containment response through the use of sensitivity analyses.
The licensee's back-end analysis considered important severe accident phenomena, and it appears to be complete. According to the licensee's submittal, the Waterford 3 conditional containment failure probabilities are as follows:
Early containment failure is 26 percent with. loss of containment heat removal, coupled with the reactor coolant system at high pressure, leading to high pressure melt ejection challenges, the primary contributor; late containment failures is 20 percent with steam overpressure following loss of containment heat removal the primary contributor, and bypass is 8 percent with steam generator tube rupture and interfacing systems LOCAs the primary contributors. According to the licensee, the containment remains intact 46 percent of the time.
Subsequent to the IPE submittal, the licensee revised j
their containment failure curve which resulted in different, but generally similar, conditional contat1nment failure probabilities as those originally reported.
Early radiological releases are dominated by station blackout and small LOCA sequences with loss of safety injection and containment heat removal; late releases are dominated by small LOCA sequences. The release fractions i
predicted for SGTR sequences are less than those encountered in other IPEs because of the assumed availability of water scrubbing. However, since this scrubbing may not be available for all SGTR sequences, the release fractions reported in the submittal for these sequences may not be adequate for all SGTR sequences.
The licensee's response to containment performance improvement program recommendations is consistent with the intent of GL 88-20 and associated l
Supplement 3.
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. According to the licensee, some insights and unique plant safety features identified by the licensee at Waterford 3 are:
1.
There is no feed and bleed capability at Waterford 3: no pressurizer PORV
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exists and the HPSI/ charging pumps do not have the requisite head to lift the safety valves.
2.
The reactor coolant pump (RCP) seals are the Byron Jackson type which, according to the submittal, can sustain loss of component cooling water for 30 minutes without failure (no seal injection is provided). As a 1
result, spurious failures of the RCP seals was not considered a credible initiator.
3.
There are two emergency diesel generators. They require cooling by the component cooling water system, ventilation by dedicated fans, and de power provided by the station batteries.
4.
There are multiple pathways for secondary steam relief:
6 turbine bypass valves, 2 atmospheric dump valves, and 6 safety relief valves.
5.
There is no service water at this plant.
Instead, the ultimate heat sink is provided by the dry cooling towers.
6.
The component cooling water system is needed to cool the HPSI pumps, the LPSI pumps, the containment spray pumps, the shutdown heat exchangers (also used for containment spray recirculation cooling), the containment fans, the emergency diesel generators, and the central chillers used to provide heating, ventilation, and air conditioning cooling.
7.
The switchover to recirculation is automatic; however, the operator must manually close the suction to the refueling water storage pool.
8.
There are three plant atteries. The AB battery is used for turbine driven emergency feedwater pump control in station blackouts. The IPE pegged the de battery station blackout depletion time at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with load shedding (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without). Since the IPE, the batteries have been upgraded and a new non-safety battery has been added to take up non-safety loads previously serviced by the safety battery. These modifications have extended the depletion time to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The licensee defined a vulnerability as either an extremely high sequence CDF (substantially greater than IE-04), a greater than 50 percent contribution to CDF from a single sequence, or an event that contributes in an unusual or substantial way to the risk profile.
Based on this definition, the licensee did not identify any vulnerabilities.
j No plant improvements have been implemented as a result of the IPE, although some are currently under review.
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. III. CONCLUSION 4
Based on the above findings, the staff notes that:
(1) the licensee's IPE is complete with regard to the information requested by GL 88-20 (and associated guidance in NUREG-1335), and (2) the IPE results are reasonable given the Waterford 3 design, operation, and history. As a result, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the Waterford 3 IPE has met the intent of GL 88-20.
4 It should be noted that the staff's review primarily focused on the licensee's ability to examine Waterford 3 for severe accident vulnerabilities. Although certain aspects of the IPE were explored in more detail than others, the i
review is not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that stemmed from the examination.
Therefore, this SER does not constitute the Nuclear Regualtory Commission approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of GL 88-20. The staff has identified weaknesses in the front-end and HRA portion of the IPE.
Consequently, we believe that application of the IPE in support of risk-based regulatory applications, beyond those associated with GL 88-20, will require additional treatment in these areas.
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Attachment:
Technical Evaluation Report Principal Contributor: John C. Lane Daic: March 4, 1997
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APPENDIX CONTRACTOR TECHNICAL EVALUATION REPORT J
1 ATTACHMENT