IR 05000324/1990003
| ML20058L779 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/11/1990 |
| From: | Munro J, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058L777 | List: |
| References | |
| RTR-NUREG-1021 50-324-90-03, 50-324-90-3, 50-325-90-03, 50-325-90-3, NUDOCS 9008080095 | |
| Download: ML20058L779 (7) | |
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UNITED ST ATES
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NUCLE AR REGULATORY COMMisslON
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Report Nos.: 50-325/90-03 and 50-324/90-03 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.:
50-325 and 50-324 License Nos.: DPR-71 and DPR-62 Facility Name:
Brunswick 1 and 2 Operational Evalu ons conducted: June 9-10, 1990 7!4!7" Inspector:
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J. A nro
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Date Signed Team members:
C. Payne, RII B. Wetzel, NRR/0LB T. Bettendorf, PNL J. Pellet, RIV Approved by:
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T. A. Pedbles T,hief pate'51gned Operations Branch Division of Reactor Safety SUMMARY
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' Scope:
This was a special announced inspection of crew evaluations on the Brunswick-simulator in accordance with NUREG-1021 ES-601.
Both the facility and NRC evaluators assessed the ability of three crews of licensed operators to safely
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function and operate as a team prior to allowing their return to licensed
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duties.
Results:-
The NRC determined that two of three crews evaluated, and sixteen of 21 o)erators performed satisfactorily during these evaluations.
As a result, tie licensee had sufficient licensed operators available to man a four shift rotation and therefore could startup and resume power operation of both units.
j Overall, the crews evaluated demonstrated significant improvement in their operational ability with respect to the weaknesses previously identified in the_ Requalification Program Rep (ort (Examination Report OL-50-325/90-01)
and the Operational Evaluation Report Report OL-50-324/90-02).
However, significant j
weaknesses in cognitive skills with safety systems, weaknesses in senior
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reactor operator command / control and cognitive skills with Emergency Operating
Procedures. were again. demonstrated primarily by one crew during one
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- examination scenario.
Further Operational Evaluations will be conducted on July 25 and 26 for the purpose of qualifying a fifth and sixth operating crew.
9008080093 900711 i
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REP 0PT DETAILS 1.
Persor,s Contacted Licensee Employees A. Watson, Senior Vice President A. Cutter, Vice President Nuclear Services R. Starkey, Brunswick Site Vice President
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J. Harness, Brunswick Site Manager J. Moyer, Assistant to Site Manager L. Martin, Corporate Training Director J. O'Sullivan, Director Training Brunswick R. Poulk, Project Specialist Operator Training Other instructors and technicians.
NRC Representatives E. Merschoff, Acting Director, DRS, RII J. Munro, Chief, OLS 1, RII C. Pryne, Examiner RII B. Wetzel, Examiner NRR T. Bettendorf, Examiner PNL J. Pellet, Chief, OLS, RIV l
NOTE: Acronyms and initialisms used throughout this report are listed in the last section.
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2.
June 9-10, 1990, Site Visit L
The evaluation team observed simulator performance of three shift crews.
The crews consisted of an SOS, SF, SCO, and two R0s.
Two of the three
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crews were administered three simulator scenarios.
One crew was administered two simulator scenarios.
Scenarios were designed to evaluate previously demonstrated generic weaknesses and were administered and evaluated in accordance with the guidance of revision 5 to the i
l Examiner Standards, NUREG-1021.
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The following list outlines the crews and the specific scenarios on which
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they were evaluated.
Crew C:
Scenario #22 - Recirculation Pump Trip Thermal Hydraulic Instability Requiring Scram, ATWS, Loss of MCC-2XG (SLC Pump Failure), Steam Line Break Causing MSIV Isolation, L
Loss of HP Injection Requiring Emergency Depressurization
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Scenario # 18 - Leak in DW, Spray Logic Failure, HPCI Failure
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Scenario # 19 - HPCI Logic Failure, RBCCW to Drywell
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Failure, Loss of 4KV Bus 20 Pipe Break in Drywell
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Note:
Scenario #19 was an additional scenario run at licensee's request.
Crew D:
Scenario #20 - NI Failure, Loss of Off-Site Power, Diesel i
Generator Failure, HPCI Malfunction, large Leak in Primary Containment Scenario #19 - See above.
Crew E:
Scenario #22 - See above.
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Scenario #18 - See above.
Scenario #20 - See above, a.
Crew C (1) Senario #22.
The BOP failed to properly prevent injection from the Condensate /Feedwater System as directed per E0P-01-LPC.
The operator did not recognize that isolation valve V-6 was failed open.
With reactor pressure lowering and the Condensate System injection not properly terminated, an uncontrolled injection of unborated water to the reactor occurred.
This caused a significant power (approximately 38 percent) and pressure excursion before injection could be terminated.by direction of the SOS and SF.
The actions of the SOS and SF to mitigate the transient caused by the 80P's error were evaluated as
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satisfactory.
The NRC and licensee evaluators graded the B0P's performance as unsatisfactory due to his failure to properly prevent injection from the Condensate /Feedwater System - IS C #18.
The NRC evaluators commented that although overall
communications were satisfactory, no acknowledgements were given to the SF as he directed different controlling level bands.
(2) Scenario #18.
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The crew acknowledged the initibtion of the ADS timers.
The crew did not understand, address, or report, the increased rate of reactor inventory loss indicated by the actuation of the ADS timers.
The SF exhibited difficulty in his utilization and
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direction of E0P-01-EPP.
The SOS advised the SF at least twice to inhibit the ADS system.
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The SF improperly ordered ADS to be inhibited with reactor level less than TAF.
With the operator in the process of inhibiting
ADS, a proper automatic actuation occurred; the operator's actions to inhibit then caused an improper override of the actuation.
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The increased size of the leak caused reactor pressure to decrease rapidly and resulted in injection of the LP systems.
The SF improperly ordered systems secured with level below TAF.
Before this direction could be corrected by the SOS, the R0 secured a RHR pump.
The NRC and licensee evaluators graded the crew and individual operator's performance as unsatisfactory due to the failure to properly perform the following ISCTs:
R0-1)
Improperly shutdown RHR pump with reactor level less
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than TAF.
(PostExam),
2)
Failed to report reactor level to SF with level decreasing and less than TAF.
(PostExam).
B0P. 1)
Did not properly operate RCIC with a high drywell pressure.
(PostExam),
2)
See ISCT #2 above listed for R0.
SF 1)
Improperly ordered ADS to be-inhibited with reactor level less than TAF.
(PostExam).
2)
Failed to perform an Emerge
' Depressurization in accordance with E0P-01-EPP.
PostExam).
Did not properly direct or assist the SF to mitigate SOS
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significant safety related adverse consequences, ISCT
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This scenario was not run as planned due to an error made by the licensee's simulator operator.
The scenario was designed to increase the size of the leak by changing a 1 percent steam leak to. a 1 percent - 2 percent water leak.
The simulator operator increased the size of the -leak to a 10 percent water leak.
This caused the rate of inventory loss to increase to a higher than planned rate.
Following the scenario, licensee and
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NRC evaluators reviewed plant data, re-ran the transient in the simulator play-back mode and reviewed crew performance on the video tape record.
The NRC considered the available information and acknowledged that the difficulty of the simulator problem had increased but determined that the scenario was still valid, realistic and within the scope of the E0Ps.
The licensee evaluators completed the co-evaluation process as described above.
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The licensee however indicated in a June 14, 1990, letter to the NRC that they may still choose to rebut the results of this scenario. No formal rebuttal has yet been submitted.
(3) Scenario #19.
This scenario was administered at the licensee's request based on their concerns with scenario #18.
i The performance of the crew and individual operators on the scenario and related ISCTs was satisfactory.
The crew was rated unsatisfactory on scenario #18, and also due to a total of seven ISCTs being performed unsatisfactorily.
Four operators were rated unsatisfactory on scenario #18.
An additional different operator, the B0P, was rated unsatisfactory on scenario #22.
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Crew D (1) Scenario #20.
The SF exhibited difficulty in his utilization and direction of procedures E0P-01-FP-04 and E0P-01-EPP.
The SF required continuing assistance from the SOS to adequately accomplish the procedures.
The performance of the crew and individual operators on the scenario and related ISCTs was satisfactory.
This scenario was not run as planned due to an error made by the licensee's simulator operator.
A loss of off-site power with a failure of both Diesel Generators was in progress.
- 3 was to be started by the simulator operator if its auto i
start pushbutton was depressed by the operator.
The operator
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was manipulating DG #4 controls and DG #3 was incorrectly started by the simulator operator.
Thus, no evaluation could be mada of the 80P's or crew's ability to properly regain emergency AC power (DG #3).
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(2) Scenario #19.
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L The crew did not trip the Recirculation Pumps within 10 minutes
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as required on a loss of RBCCW cooling.
The pumps tripped automatically on low level much later in the scenario.
The SF exhibited difficulty in his utilization and direction of
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E0P-01-FP-03 and E0P-01-EPP.
Again, the SOS was required te provide the SF' continuing assistance to adequately accomplish
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the procedures.
The SF incorrectly ordered reactor pressure be lowered with SRVs to assist the R0's attempts at restarting RCIC.
Without significant HP injection, reactor level decreased. rapidly from just above TAF to less than TAF.
The SOS properly mitigated this error by calling for an Emergency Depressurization.
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S Following depressurization, the crew incorrectly secured all LP injection at +60" with reactor level increasing rather than sequentially securing LP pumps.
This caused reactor level to decrease to TAF.
Proper mitigation of the error was accomplished by restert of one RHR pump.
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The errors noted above were identified and evaluated by both the licensee's and NRC evaluators.
It ms noted that proper mitigation was taken by the operators with the end result being
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that the safety status of the plant was not unduly challenged.
Therefore, the performance of the crew and individual operators
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on the scenario end related ISCTs was satisfactory.
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The crew and all individuals were rated satisfactory by both NRC and licensee evaluators.
It was also agreed that the weaknesses displayed by the SF on E0P usage and command and control were significant but did not merit failure.
Licensee management indicated that remedial training would be scheduled for the crew in three days with emphasis placed on the SF's weaknesses, c.
Crew E (1) Scenario #22 The performance of the crew and individual operators on the scenario and related ISCTs was satisfactory.
.(2) Scenario #18 The performance of the crew and individual operators on the l.
scenario and related ISCTs was satisfactory.
(3) Scenario #20
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The SF and R0 demonstrated a weakness in clearly identifying the correct level instrument to utilize with degraded Drywell conditions.
The R0 continued to announce level readings from
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N027 after it was no longer considered reliable due to a drywell temperature of 250 degrees F.
The SF did not give clear guidance to utilize NO36 vice N027.
The SF eventually took proper action to increase level before decreasing below t
TAF as indicated on NO36.
The performance of the crew and individual operators on the scenario and related ISCTs was satisfactory.
The crew and all individuals were rated satisfactory by both 11RC and licensee evaluators.
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Exit Interview The Team Leader ar,d the Acting Director, DRS, met with licensee managemera on June 10, 1990, after completion of the Operational Evaluations.
Licensee managen,ent was informed of the individual and crew failures.
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This meeting was followed by approval being given for restart of Brunswick Units 1 and 2 by the Region II Regional Administrator.
This approval was given via a phone call from S. D. Ebneter to R. A. Watson on June 10, 1990.
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Acronyms and Initialisms ADS Automatic Depressurizatien System ATWS Anticipated Transient Without Scram B0P'
Balance of Plant Operator DG Die:31 Generator DRS Division of Reactor Safety DW Drywell E0P Emergency Operating Procedure EPP End Path Procedure ERFIS Ems.*gency Response Facility Information System HP High Pressure HPCI High Pressure Coolant Injection ISCT Individual Simulator Critical Task LOCA Loss of Coolant Accident LP Low Pressure
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LPC Level Power Control
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MSIV Main Steam Isolation Valve Ni Nuclear Instrumentation RBCCW Reactor Building Closed Cooling Water RCIC Reactor Core Isolation Cooling System RBCCW Reactor Building Closed Cooling Water RHR Residual Heat Removal System
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R0 Reactor Operator SCO Senior Control Operator SCRAM Reactor Trip SF Shift Foreman SLC Standby Liquid Control i
l SOS Shift Operating Supervisor SR0 Senior Reactor Operator STA Shift Technical Advisor TAF Top of Active Fuel i
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