05000387/LER-2003-006
| Susquehanna Steam Electric Station - Unit 1 | |
| Event date: | 09-24-2003 |
|---|---|
| Report date: | 11-19-2003 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3872003006R00 - NRC Website | |
,Susquehanna Steam Electric Station - Unit 1 4 05000387 , � 2003 � - � 006 � - 00 2 � OF 5
EVENT DESCRIPTION
At 00:53 on September 24, 2003 with Unit 1 in Mode 1 at 100% power, an automatic reactor scram occurred in response to low reactor water level conditions. A human performance error that occurred during testing of the Reactor Feed Pump Turbine (RFPT; EIIS Code: SJ) initiated the event sequence that led to this Reactor Protection System (RPS; EIIS Code: JC) actuation. The control room operator (Utility, Licensed) had successfully performed Emergency Governor and Trip Lockout testing on the 'A' and 'B' RFPT's as directed by station procedures. During testing of the 'C' Feed Pump Turbine, the operator incorrectly manipulated the 'C' RFPT lockout key switch Instead of the Reset pushbutton thus causing the RFPT to trip. The Reactor Recirculation pumps (EIIS Code: AD) ran back to approximately 45% speed as water level began to drop in response to the partial loss of Feedwater flow. Although the 'A' and 'B' Reactor Feed Pumps increased speed in an attempt to maintain reactor inventory levels, the reactor automatically scrammed when water level reached the Low-Level RPS initiation setpoint. All rods successfully inserted. Reactor power was approximately 81% at the time of the scram. Level dropped to approximately — 48". The Reactor Core Isolation Cooling system (RCIC; EIIS Code: BN) and the High Pressure Coolant Injection system (HPCI; EIIS Code: BJ) automatically initiated to assist the 'A' and 'B' Reactor Feed Pumps In level restoration. Both Reactor Recirculation pumps tripped upon receipt of an Automatic Transient Without Scram-Recirc Pump Trip (ATWS-RPT) low water level signal. Both pumps were, however, manually restarted to avoid thermal stratification.
Numerous Primary Containment Isolations (EIIS Code: JM) also occurred as designed during the transient. Following the scram, reactor pressure was controlled using Main Steam Bypass Valves (EllS Code: JI). Post-scram conditions were maintained within procedural operating requirements.
Susquehanna was designed to withstand a single Reactor Feed Pump Turbine trip, at high power operation, without experiencing a Rx Low-Level Scram. Early plant testing demonstrated that the Unit 1 operating margin for a loss of Feed Pump transient was at least 15 inches above the Rx Low-Level Scram setpoint. As such, the scram of September 24, 2003 was not anticipated. This event suggests that previous changes made at Susquehanna, from the time of original design until September 24, 2003, have affected the plant's integrated response to the trip of a single Reactor Feed Pump Turbine.
These changes involved Reactor Recirculation Runback performance, Reactor Feed Pump response and capabilities, Power to Flow Map operating domains, steam generation rates, and other parameters. The cumulative effect of these changes has resulted in the unintentional reduction of the original operating margin. This reduction of margin has led to an automatic reactor scram and the subsequent initiation of emergency core cooling systems following a RFPT trip, a transient not expected to attain such results.
CAUSE OF EVENT
The RFPT trip has been attributed to human performance error and a less than adequate use of essential human performance tools. Most significantly, self-checking, peer checking, and 3-part communication were not effectively utilized. Also, inadequate identification of critical activities and critical steps within procedures were identified as causal factors.
CAUSE OF EVENT (continued) The cause of the automatic reactor scram has been attributed to a reduction in operating margin resulting from inadequate identification of operating margin requirements in plant change processes.
EXTENT OF CONDITION
Investigation of this event has revealed that changes to the Susquehanna Unit 1 and Unit 2 operating domain (Power/Flow Map) have been implemented since the start of commercial operation which have compromised the reliability of a design function originally intended to sustain power operations (Le., Recirculation flow runback associated with the #2 limiter). Thus, the plant's operational margin has been reduced.
Since plant operation In an expanded domain has prevented the fulfillment of an original design function (#2 limiter), efforts were Initiated to determine if other plant design functions may have been impacted by operation beyond the original design envelope.
This effort, which included review of selected FSAR chapters, licensing submittals, and NRC Safety Evaluation Reports (SERs) associated with key projects that expanded Susquehanna's operating envelope has concluded that:
- The #2 recirculation pump speed limiter is solely intended to maintain power operations in the event of abnormal Balance of Plant circumstances. Its failure to successfully perform its function represents a reduction in power generation reliability, or operational margin.
- The #2 recirculation pump speed limiter does not provide a function related to the safe operation of the Susquehanna Units. The inability of the limiter to prevent a reactor scram does not impact the plant's margin to safety. Erosion of safety related margins in other plant systems or areas, because of plant operation in the extended operating domain, was not evident following performance of this review.
- With the exception of the #2 recirculation pump speed limiter, no additional examples were identified where plant operation in the expanded operating domain had adversely impacted a design function intended to sustain power operations.
NRC FORM 3613R (7.2001) Susquehanna Steam Electric Station - Unit 1 05000387 2003 � - � 006 � - � 00 4 � OF 5
REPORTABILITY/SAFETY CONSEQUENCES ANALYSIS
Actual: This event was determined to be reportable under 10 CFR 50.73(a)(2)(iv)(A) in that unplanned actuations of Engineered Safety Features (ESF) occurred as a result of operator error during testing of the Unit i "C" Reactor Feed Pump Turbine. The error resulted in actuation of RPS, initiation and injection of the HPCI and RCIC systems, and numerous primary containment isolations.
All safety systems functioned as designed. All control rods inserted and post scram reactor conditions were maintained within procedural requirements. There were no challenges to Primary or Secondary containment (EIIS Code: NH) and integrity was maintained. Water level was restored to the nominal operating band.
Main Steam Line (EIIS Code: SB) and Offgas (EIIS Code: WF) radiation levels indicated no abnormal variations. Reactor coolant activity, in-plant radiation levels and vent release rates (Ens Code: IL) were all normal based on plant conditions and were well within acceptable limits. There was no evidence of any fuel failure.
Since all ESF systems and components functioned properly and per design, there were no safety consequences or compromises to the health or safety of the public.
Potential: Human performance errors can challenge safety functions of plant structures, systems, and components (SSC). Reduced operating margins challenge Operations personnel and increase the frequency of safety system initiations.
CORRECTIVE ACTIONS
The following corrective actions for this event have been completed:
- The Unit 1 RFPT High Speed Stops were adjusted to increase available speed.
- Ul and U2 Reactor Recirculation pump controller reset rates were increased to provide a faster runback speed to recover lost operating margin.
- Involved individuals (Utility. Licensed) have been coached/counseled on the effective use of error prevention tools and techniques.
The following corrective actions are planned to address Human Performance:
- � Error prevention techniques will be reinforced to support desired human performance attributes.
The following corrective actions are planned to address loss of operating margin:
- � Station operating margin controls will be strengthened within plant change processes.
- � U2 RFPT High Speed Stops will be adjusted to increase available speed.
Susquehanna Steam Electric Station - Unit 1 05004387 � _ 2003 � - � 006 � - � 00 5 � OF 5
ADDITIONAL INFORMATION
Failed Component information:
Component — Unit 1 'C' Reactor Feed Pump Turbine Model — DRV 631122913 Manufacturer - General Electric
Previous Similar Events:
At Susquehanna, there is no history of scrams resulting from a loss of a single Reactor Feed Pump. In September 1999, loss of a single Reactor Feed Pump resulted in a Recife #2 limiter runback. At that time, the remaining operational Feed Pumps restored water level to normal levels to avoid a low-water level scram.
NRC FORM OW (74001)