05000446/LER-2011-002
| Comanche Peak Nuclear Power Plant Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(B), System Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4462011002R00 - NRC Website | |
L DESCRIPTION OF REPORTABLE EVENT
A. REPORTABLE EVENT CLASSIFICATION
10CFR50.73(aX2XivXA), Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section including:
- 10CFR50.73(a)(2)(iv)(B)(1), Reactor protection system (RPS) including: reactor scram or reactor trip
- 10CFR50.73(aX2Xiv)(B)(6), PWR auxiliary or emergency feedwater system
B. PLANT OPERATING CONDITIONS PRIOR TO THE
EVENT
On May 19, 2011, CPNPP Unit 2 was in Mode 1 operating at 100 percent power.
C. STATUS OF STRUCTURES, SYSTEMS, OR COMPONENTS THAT WERE INOPERABLE AT THE START OF
THE EVENT AND THAT CONTRIBUTED TO THE EVENT
There were no inoperable structures, systems, or components that contributed to the event.
D. NARRATIVE SUMMARY OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES
At 1350 on May 19, 2011 a Condensate Polishing System alarm occurred on the Comanche Peak Unit 2 Main Control Board followed immediately by alarms from other secondary sample points including the Steam Generator Slowdown Sample Panel (SGBD). At 1357, Unit 2 operators (Utility, Licensed) entered Abnormal Conditions Procedures Manual, ABN-304, "Main Condenser and Circulating Water System Malfunction" due to indications of a condenser [EIIS: CONDI tube leak.
Sodium concentration in the Unit 2 steam generators [EUS: SG] increased rapidly to approximately 3000 parts per billion.
In accordance with ABN-304, the Unit 2 reactor was manually tripped at 1407 on 5/19/2011, due to steam generator sodium concentration greater than 1000 parts per billion.
MI control rods fully inserted and both Motor Driven Auxiliary Feedwater Pumps [EIIS: (BA)(P)] started as expected as a result of the reactor trip. The Turbine Driven Auxiliary Feedwater Pump also started on receipt of low level signals from more than one steam generator immediately after the trip. All systems responded normally during and after the event.
There were no safety consequences impacting plant or public safety as a result of this event.
E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE, OR PROCEDURAL OR
PERSONNEL ERROR
Control board indications and alarms, as well as local chemistry instrumentation readings alerted Unit 2 reactor operators (Utility, Licensed) to the condenser tube leak.
H. COMPONENT OR SYSTEM FAILURES
A. FAILURE MODE, MECHANISM, AND EFFECTS OF EACH FAILED COMPONENT
Not applicable — there were no component failures associated with this event.
B. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE
Not applicable — there were no component or system failures associated with this event.
C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE All FECTED BY FAILURE OF COMPONENTS WITH
MULTIPLE FUNCTIONS
Not applicable — there were no component or system failures associated with this event.
D. FAILED COMPONENT INFORMATION
Not applicable — there were no component failures associated with this event.
111. ANALYSIS OF THE EVENT
A. SAFETY SYSTEM RESPONSES THAT OCCURRED
The Reactor Protection System (RPS) was manually actuated to initiate this event. Both Motor Driven Auxiliary Feedwater Pumps and the Turbine Driven Auxiliary Feedwater Pump started as expected as a result of the reactor trip.
B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY
Not applicable — No safety system was rendered inoperable.
C. SAFETY CONSEQUENCES AND IMPLICATIONS
There were no inoperable structures, systems, or components at the time that contributed to this event. The plant responded as designed for the manual reactor trip followed by the automatic initiation of the auxiliary feedwater system. This event is bounded in severity by the analysis of the loss of normal feedwater event presented in Section 15.2.7 of the CPNPP Final Safety Analysis Report (FSAR), which assumes conservative initial conditions that bound the plant operating range and other assumptions which could reduce the capability of safety systems to mitigate the consequences of the transient.
High concentrations of sodium are a long term concern for steam generator tube integrity; prolonged exposure to out-of-specification sodium levels could shorten steam generator tube life and increase the possibility of steam generator tube failure. Based on the short duration of exposure, this event is not expected to impact nuclear safety.
Based on the above, it is concluded that the health and safety of the public were unaffected by this condition and this event has been evaluated to not meet the definition of a safety system functional failure per 10CFR50.73(aX2)(v).
IV. CAUSE OF THE EVENT
The Unit 2 reactor was manually tripped due to high sodium concentrations in the steam generators. The sodium excursion was the result of a falling object causing the sudden failure of two tubes in the Unit 2 main condenser allowing lake water to enter the condensate system and accumulate in the steam generators. The sodium concentration in the steam generators rapidly exceeded the ABN-304 manual trip criteria of 1000 parts per billion.
Evaluation of the condenser tubes using Sodium hexafluoride (SF6) gas and dimple plug testing identified the two breached tubes. Subsequent boroscope and eddy current testing (ECT) revealed indications the two tubes were struck by a falling object. Six additional tubes were identified to have external denting that did not exist at the conclusion of the previous refueling outage. All of the tube damage occurred in the same general location approximately 4.5 feet from the inlet side tube sheet.
V. CORRECTIVE ACTIONS
The two ruptured tubes and six dented tubes were plugged, along with an additional seven adjacent tubes as a preventative measure.
Following the tube plugging activities, an Engineering evaluation was performed to determine the acceptability of returning the Unit 2 main condenser to service. The Unit 2 main condenser was then returned to service, secondary and steam generator chemistry restored to limits, and Unit 2 was returned to 100 percent power.
VI. PREVIOUS SIMILAR EVENTS
There have been no similar events reported from CPNPP in at least the last three years.