05000456/LER-2010-006
| Docket Number | |
| Event date: | 11-12-2010 |
|---|---|
| Report date: | 01-11-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 4562010006R00 - NRC Website | |
A. Plant Conditions at Time of Condition Discovery and Background Information:
Event Date: November 12, 2010� Event Time: 1444 CDT Unit: 1 MODE: 1 Reactor Power: 100 percent Unit: 2 MODE: 1 Reactor Power: 100 percent Unit 1 Reactor Coolant System [AB]:� Normal operating temperature and pressure Unit 2 Reactor Coolant System:� Normal operating temperature and pressure Pertinent original design basis of the Component Cooling Water (CC) ICC1 system:
The CC system is a shared systembetweenthebwounitsond consists of five pumps (two per unit and one common), three heat exchangers (one per unit and one common), and two surge tanks (one per unit). Make-up water to the surge tanks is not safety-related.
The bounding design basis scenario for the CC system is the simultaneous Loss of Coolant Accident (LOCA) and Loss of Offsite Power on one unit and the normal shutdown on the other unit. To account for the design requirement to mitigate a passive single failure in the long term, the CC system trains on the accident unit must be separated preemptively at the onset of a design basis LOCA scenario.
B. Description of Event:
On September 27, 1987, a License Amendment Request (LAR) was submitted for Byron and Braidwood Stations to request an increase in the Allowed Outage Times (AOTs) from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to seven days for several systems, including the CC System and an Emergency Core Cooling System (ECCS) sub-system (i.e., the Residual Heat Removal (RH) [BP] System). This LAR was approved by the NRC via License Amendments 14 and 4 for Byron and Braidwood Stations, respectively, issued January 21, 1988. This LAR was based on the 1984 Westinghouse Commercial Atomic Power (WCAP) -10526, "Byron Generating Station Limiting Condition for Operation Relaxation Program," which provided a Probabilistic Risk Assessment (PRA) technical justification for the AOT extensions.
On July 7, 2010, an issue was identified concerning an apparent discrepancy in WCAP-10526. Based on CC design discrepancies involving the common CC system's pump, that have been known to exist approximately since the 1987 timeframe, the CC system description contained in the WCAP was incorrect and therefore, was likely modeled incorrectly in the PRA analysis.
In the WCAP-10526 CC system description, the common CC system's pump was described as a maintenance spare that could be substituted for any of the CC system's unit-specific pumps. Subsequent to the issuance of the WCAP, it became known that if the common CC system pump is aligned to substitute for either unit's B CC train pump, then it would be isolated from its unit surge tank upon splitting of the CC system's trains. Also, with the common CC system pump in this configuration, it would be powered by electrical division II while providing cooling to the A train of the RH system. With either of these discrepancies, the common CC system pump could not be considered an operable pump while aligned to either unit's B CC train. Though these design discrepancies were known, the Technical Specifications (TS) implications were not recognized and the B CC trains were considered operable when the common CC pump was aligned to them.
An assessment was conducted of the significance of this design discrepancy on the PRA modeling assumptions and conclusions of WCAP-10526. On November 12, 2010, it was concluded the CC system design discrepancies would result in a negative impact on the PRA analysis results which justified the AOT extension for the CC system.
In addition, another potentially significant discrepancy was discovered in both the CC and RH system analyses, in that it did not correctly account for the operational requirement to preemptively split CC trains in a post accident scenario.
Re-creation of the 1984 vintage PRA modeling and analysis was not feasible in order to determine a quantitative value to this discrepancy. Therefore, it is unknown whether this negative impact would have been significant enough to have impacted NRC approval of the LAR for CC and RH. However, Exelon Generation Company, LLC (EGC) concluded that it would have been significant enough to impact the NRC's approval of the LAR and that the AOT for TS 3.7.7, "Component Cooling," and TS 3.5.3, "ECCS- Operating" (RH Sub-system) should be considered non-conservative and the provisions of NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That are Insufficient to Assure Plant Safety," be invoked.
The following administrative controls have been implemented at Byron and Braidwood Stations pending modifications to address the CC design discrepancies:
- The AOT for TS 3.7.7, "Component Cooling" Condition B for one required CC pump inoperable has been restricted to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- The AOT for TS 3.5.2, "ECCS-Operating" Condition A has been restricted to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for an inoperable RH train
This condition is reportable to the NRC pursuant to 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degraded plant safety, and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of a system's safety function.
C. Cause of Event
The most probable causes of the event are limited procedural guidance for preparing correspondence, and ambiguity in intended system operation. The root cause of the inaccurate LAR in the 1987 timeframe was indeterminate due to the timeframe when the event occurred. The processes in place for preparing and reviewing LARs in the 1987 timeframe were not as robust as the current processes.
D. Safety Consequences:
The design basis safety function of the CC system is to remove the post LOCA heat load from the containment sump during the ECCS recirculation phase. The containment sump is the suction source for the ECCS pumps during the recirculation phase.
There were no actual safety consequences resulting from this condition. No actual loss of a safety function occurred. However, the potential existed for more severe conditions to have developed, when the common CC system pump was aligned to replace either unit's B CC train pump and CC trains split. With the postulation of design basis assumptions, a loss of the CC safety function could have occurred and, if not mitigated, would in turn lead to a loss of the ECCS.
A three year historical review of times the common CC system pump replaced either unit's B train pump while in the mode of applicability resulted in finding two instances for Unit 1 and two for Unit 2. The duration time frames were 80 and 138 hours0.0016 days <br />0.0383 hours <br />2.281746e-4 weeks <br />5.2509e-5 months <br /> for Unit 1, and 59 and 281 hours0.00325 days <br />0.0781 hours <br />4.646164e-4 weeks <br />1.069205e-4 months <br /> for Unit 2.
E. Corrective Actions:
The corrective actions include:
- Investigate modification of the CC system to eliminate the design discrepancies with the common CC pump and the need to pre-emptively split CC trains.
- A review of the current LAR preparation and review process concluded it is sufficiently robust to minimize potential inaccurate information from not being identified.
- Training will be conducted to appropriate Site personnel to raise awareness of the circumstances and missed opportunities for recognizing the significance and implications of the design discrepancies.
- An extent of condition review will be conducted.
F. Previous Occurrences:
There have been no previous, similar Licensee Event Reports identified at the Braidwood Station.
G. Component Failure Data:
Manufacturer� Nomenclature� Model�Mfg. Part Number N/A� N/A� N/A N/A