ML053190327

Regulatory Conference Kewaunee Turbine Building Flooding Issues

ML053190327
Person / Time
Site:	Kewaunee
Issue date:	11/08/2005
From:	Satorius M Division Reactor Projects III
To:
References
EA-05-176
Download: ML053190327 (9)

Text

Regulatory Conference Kewaunee Turbine Building Flooding Issues Nuclear Regulatory Commission - Region III Lisle, Illinois November 8, 2005

Introduction and Opening Remarks Mark Satorius Director Division of Reactor Projects

Agenda

1. Introduction and Opening Remarks

2. Overview of Event and Inspection Finding

a. Description of Flooding Issues

b. Description of Performance Deficiency

c. Preliminary Safety Significance Determination and Apparent Violation

3. Licensee Presentation

4. NRC Caucus

5. Closing Remarks

Description Scenario: Large rupture of pipe or tank in turbine building Flood water enters safeguards alley due to insufficient barriers (doors, drains) and affects equipment vital to safe shutdown Susceptible Class I Equipment: AFW, ESF busses, EDGs, safe shutdown panel Past precedent: May 2003 tank overflow resulted in turbine building sump overflow; water flowed into safeguards alley trench; extent of condition and corrective actions inadequate Extensive modifications made including: robust flood barriers, drain line check valves, circulating water pump trips

Performance Deficiency Failure to meet requirement 10 CFR 50, Appendix B, Criterion III, Design Control Requires, in part, adequate measures be established to assure that design basis for safety-related SSCs are correctly translated Requires, in part, measures to verify and check adequacy of design Cause within licensees ability to foresee and correct 1972 letter to AEC stating, in part, that non-Class I pipe breaks could affect Class I equipment, however, plant design (redundancy, design arrangement) would not be jeopardized in the event of a pipe break Safeguards alley trench flood due to turbine building flooding in May 2003

Safety Significance Determination (Preliminary)

Phase III Analysis (IMC 0609, Appendix A)

Multiple trains of equipment degraded Licensees preliminary evaluation resulted in change in CDF representative of Yellow safety significance Several assumptions under NRC scrutiny Preliminary Greater-than-Green Finding and Apparent Violation of 10 CFR 50, App. B, Criterion III, Design Control, issued October 6, 2005

Licensee Presentation NRC Caucus Closing Remarks Mark Satorius Director Division of Reactor Projects