ENS 47096

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v t e ENS Event
20:00 Jul 26, 2011
Title	Unanalyzed Conditions Involving the Safety Related Direct Current (Dc) System
Event Description	Information was received in regards to an old design issue identified in a Component Design Basis Inspection Unresolved Item. Two issues were identified with the Safety-Related Direct Current (DC) System: 1. The plant's licensing basis states that non-safety-related electrical equipment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety-related electrical equipment required safety functions, is qualified as required. However, the Reactor Coolant Pump (RCP) backup lift oil pump motors and the Containment Emergency Lighting Panel L49E1 are located inside containment and are not environmentally qualified. This could challenge the adequacy of electrical separation between the potentially grounded non-safety related equipment and the safety related batteries. 2. Automatic transfer switches are installed to automatically transfer non-safety related loads such as non-nuclear instrumentation, station annunciators, plant computer, and integrated control system between two non-safety related inverters, which receive power from the safety-related DC power system. If a ground fault existed on one of these switches, the fault could be transferred from one power source to the redundant source, potentially impacting the ability of both safety-related DC power sources to perform their required functions. This type of transfer is not permitted by the plant's licensing basis. The breakers for the 4 RCP backup lift oil pump motors and for the Containment Emergency Lighting were opened. One train of instrumentation power was placed on its alternate power source from the Alternating Current (AC) system, eliminating the potential to impact both trains of the DC power system. This condition is being reported per 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(A-D) as an event or condition that could have prevented fulfillment of a safety function. The licensee has notified state and local authorities and the NRC Resident Inspector.

ENS 47096 +/-
Where
Davis Besse Ohio (NRC Region 3)
Reporting

10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
LER:	05000346/LER-2011-004
Time - Person (Reporting Time:+-3.22 h-0.134 days <br />-0.0192 weeks <br />-0.00441 months <br />)
Opened:	Tom Cobbledick 16:47 Jul 26, 2011
NRC Officer:	Bill Huffman
Last Updated:	Jul 26, 2011
47096 - NRC Website
v • d • e

v t e Emergency Notification System
Davis Besse with 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor, 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat, 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
WEEKMONTHYEAR2024-01-05T15:40:00 [Table view]
[ENS list for Davis Besse]

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