05000333/LER-2015-003
| James A. Fitzpatrick Nuclear Power Plant | |
| Event date: | 07-20-2015 |
|---|---|
| Report date: | 09-18-2015 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material |
| Initial Reporting | |
| ENS 51242 | 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material |
| 3332015003R00 - NRC Website | |
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James A. FitzPatrick Nuclear Power Plant 05000333
Background
The Secondary Containment [EIIS identifier: NG] boundary surrounds the primary containment and refueling equipment. The boundary forms a control volume to contain, dilute, and hold up fission products. The Secondary Containment consists of four systems which include the Reactor Building, the Reactor Building Isolation and Control System, the Standby Gas Treatment System, and the Main Stack. Secondary Containment is designed to provide containment for postulated design basis accident scenarios: loss-of-coolant accident and refueling (fuel handling) accident. Since pressure may increase in Secondary Containment relative to the environmental pressure, support systems are required to maintain a differential pressure vacuum such that external atmosphere would leak into containment rather than fission products leak out. The Reactor Building roof structure consists of a metal deck with a built-up roof (an overlay of insulation and an asphalt felt paper coating with rock ballast).
The systems which maintain a differential pressure vacuum inside Secondary Containment include the normal Reactor Building Ventilation and Cooling (RBV) System [VA] (during normal plant operations) and the safety- related Standby Gas Treatment (SBGT) System [BH] for post-accident conditions.
Event Description
On July 20, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when the planned work activity of replacing the Reactor Building built-up roof (materials above the metal deck) started at approximately 0630. The work consisted of removal of the materials above the roof metal deck (insulation, felt papers, and asphalt / gravel mixture) and replacement with a membrane roof system (insulation, roof board, and single ply EPDM membrane). Shortly after work started on removing roofing materials, Reactor Building D/P started to degrade. Degradation was halted when the SBGT was started for other plant activities. The Reactor Building Ventilation system was then isolated, resulting in a momentary improvement in D/P immediately followed by a degradation in D/P reaching less negative than 0.25 inches of vacuum water gauge.
Work activities continued with subsequent improvement in D/P conditions. The period of time that the differential pressure was below 0.25 inches of vacuum water gauge was approximately ninety-two minutes.
During the period in which Secondary Containment did not meet the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 for differential pressure, the TS Limiting Condition of Operation (LCO) 3.6.4.1 required action was to restore secondary containment to Operable status in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While secondary containment D/P did improve and become more negative than 0.25 in vacuum water gauge, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time for operability was not met (based upon ongoing roof restoration work and demonstration of operability by testing was not complete) and Condition B required action to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was entered. After secondary containment was declared Operable / Degraded, Condition B was exited. An NRC notification was made via ENS 51242. This Licensing Event Report (LER) is being submitted per 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of safety function to control the release of radioactive material.
Event Analysis Timeline of Events on 7/20/2015 0630 — 0645 Entergy Construction Supervisor obtains Operations permission to start work on replacing the RB roof insulation and built-up roof (asphalt and gravel) per Work Order 00297697 and EC 56686.
0654 RB D/P was noted to be degrading upon commencement of roof maintenance, starting at about 1.0 inches of vacuum water gauge and slowly degrading to about 0.4 inches of vacuum water gauge.
0716 Operators started Alpha train of the SBGT system to support a nitrogen feed to the Drywell and a vent of the Torus air space to maintain the required D/P between the Drywell and the Torus (not related to RB roof work). SBGT operations results in a halt of the degrading RB D/P trend at approximately 0.55 inches of vacuum water gauge.
0739 Operators isolated RBV. After a short term initial improvement of RB D/P, it immediately degraded from about 0.55 inches of vacuum water gauge to less negative than 0.25 inches of vacuum water gauge. Operators then started Bravo SBGT train resulting in RB D/P stabilizing at about 0.15 inches of vacuum water gauge.
0740 Operations declared Secondary Containment INOPERABLE due to inability to maintain >/= 0.25 inches of vacuum water gauge with both trains of SBGT in service. TS LCO 3.6.4.1 Condition A was entered to restore secondary containment to OPERABLE status in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The roofing vendor supervisor was directed to stop work and proceed with resealing the roof. This action mitigated the RB D/P degradation. Continuation of roof replacement activities eventually resulted in improving RB D/P conditions.
0911 RB D/P becomes more negative than 0.25 inches of vacuum water gauge 1140 TS LCO 3.6.4.1 Condition B was entered to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
1145 The inner and outer Reactor Track bay door seals were deflated in support of testing to demonstrate the FSAR design function of secondary containment.
1255 The lowest RB D/P observed during the demonstration with Alpha SBGT train running, RBV isolated, and both inner and outer Track bay door seals deflated was 0.41 inches of vacuum water gauge.
1305 LCO 3.6.4.1 is exited after the FSAR design function for secondary containment was met. Secondary containment is declared OPERABLE-DEGRADED.
Cause
Secondary Containment differential pressure fell below 0.25 inches of vacuum water gauge when openings in the Reactor Building built-up roof resulted in air in-leakage exceeding the ability of support systems to maintain required vacuum. Inadequate documentation of the as-built configuration of the RB roof resulted in incorrect conclusions regarding what roof components were maintaining the RB roof ventilation boundary. It was not recognized to the extent that the outer layers of the RB roof (above the roof metal deck) were providing a ventilation boundary for secondary containment.
Similar Events Internal Events Equipment malfunction resulted in Secondary Containment D/P below TS limit External events:
Cooper Nuclear Station: LER 2014-001-00, Secondary Containment Declared Inoperable due to Rise in Differential Pressure when Operator inadvertently closed exhaust damper (ML14070A363).
Columbia Generating Station: LER 2013-007-01, Secondary Containment Pressure Exceeded During Severe Weather Conditions (ML14160B127).
Corrective Actions
Completed Actions
- Revised Work Order and revised Engineering Change to allow an alternate RB roof repair detail to preclude removal of existing RB built-up roof. Completed roof maintenance.
- Completed a Root Cause Evaluation on the degraded RB D/P.
Planned Actions
- Revise drawings to show appropriate level of detail for use during future evaluations.
- Update the FSAR to more accurately describe the RB roof construction Safety Consequence and Implications There were no actual consequences of this event relative to nuclear, industrial, or radiological safety.
Nuclear Safety There were no actual nuclear safety consequences related to this event.
Radiological safety There was no radiological consequence during this event.
The potential for a radiological consequence is only applicable during the time that Secondary Containment was below 0.25 inches water vacuum differential pressure. When this condition occurred, it lasted for approximately 92 minutes.
The potential safety implications of this event are primarily topical to radiological safety. The Secondary Containment serves as part of the primary success path in mitigating the consequences of certain postulated provides a control volume into which fission products that leak from primary containment, or are released directly to the secondary containment as a result of a refueling accident. With the Secondary Containment inoperable the potential exists for the Design Basis (DB) dose consequence analyses results and regulatory limits (10 CFR 100 & 10 CFR 50.67) to be exceeded. The two principal accidents for which credit is taken for Secondary Containment operability are the Loss of Coolant Accident (LOCA) and Refueling Accident (RA).
During the time the required Reactor Building D/P was not met, the Primary Containment system was being maintained Operable and the station was not moving fuel on the Refuel Floor (on Reactor Building El. 369'-6").
Therefore, no potential safety implications to the RA dose consequence exist.
The Design Basis Accident dose consequence analysis was reviewed to assess potential safety implications.
The analysis includes conservative assumptions that minimize the potential safety implications associated with this event, and demonstrates that post-LOCA radiation exposures will remain well below all regulatory guidelines. For exclusion area boundary and low population zone doses, the analysis conservatively assumes that all post-LOCA drywell leakage (leakage rate, La, of 1.5% of containment air weight per day) is released directly to the atmosphere without holdup or mixing in the Reactor Building. Additional conservatism is included in the assumption for post-LOCA drywell leakage rate. Based on recent testing associated with the Primary Containment Leakage Rate Testing Program, the actual drywell leakage rate would have been less than one- fourth of that assumed in the dose consequence analysis.
The event resulted in a loss of Secondary Containment (required RB D/P) for approximately 92 minutes (based upon plant information data). Secondary Containment maintains a differential pressure vacuum by the RBV system or one of two SBGT trains. During a postulated accident scenario RBV is placed in isolation and the SBGT is used to maintain differential pressure. During this event, the opening in secondary containment was such that the RBV and SBGT systems were not able to maintain the required D/P. The ability of RBV to isolate or SBGT to initiate was not affected. The opening was closed as part of the maintenance activity.
Industrial safety This event did not have any actual or potential impact on industrial safety as workers on the Refuel Floor were cognizant of activities on the RB roof and were aware of potential and actual area conditions associated with roof activities above them.
References
- Condition Report: CR-JAF-2015-03260, July 20, 2015, Reactor Building D/P Degraded
- Engineering Change EC 56686 Reactor Building Roof Replacement - 2015
- Work Order WO 00297697, Replace Roof on Reactor Building (RB-430-1)
- Technical Specifications