ENS 50244
ENS Event | |
|---|---|
02:46 Jul 1, 2014 | |
| Title | Unanalyzed Condition That Could Have Resulted in an Increased Maximum Flood Level |
| Event Description | On June 27, 2014, TVA identified in a reanalyzed hydrologic analysis for Sequoyah Nuclear Plant (SQN) a deviation from the current hydrologic analysis. The flooding analysis in Section 2.4.3 of the SQN UFSAR assumes that the Watts Bar West Saddle Dike fails completely and instantaneously at approximately 1.5 feet of overtopping during a Peak Maximum Flood [PMF]. This assumption exists in the original design basis analysis and the revised analysis which supports SQN-TS-12-02, "Application to Revise Sequoyah Nuclear Plant Units 1 and 2 Updated Final Safety Analysis Report Regarding Changes to Hydrologic Analysis".
The results of recent studies of the West Saddle Dike, conducted as part of the Fukushima Order 2.1 flooding review, indicate that the complete and instantaneous failure of the Watts Bar West Saddle Dike may not be a valid assumption. If the dike does not fail, analyses performed using the codes and methods consistent with those used in original plant design show that the east floodwall of the Watts Bar Dam would overtop. As a result of this overtopping, the east floodwall is assumed to fail. Based on this assumption and analysis, failure of the east floodwall of the Watts Bar Dam would result in an increase in the flood level at the SQN Plant Site. The current licensing basis PMF level for SQN is 719.6 feet as stated in Section 2.4.2.2 of the SQN UFSAR. In addition, it should be noted that by letter dated August 10, 2012, as supplemented by letters dated April 5, 2013 and January 16, 2014, TVA proposed a revised PMF level of 722.0 feet. Introducing non failure of the Watts Bar West Saddle Dike indicated a potential increase of approximately 1.5 feet over the revised PMF level. TVA performed additional analysis using current industry standard for flooding analysis. Specifically, TVA modeled the condition using the United States Army Corps of Engineers Hydrologic Engineering Center River Analysis System (HEC-RAS) tool. TVA's analysis of the condition using HEC-RAS determined that all required safety equipment for SQN would not be impacted and are considered operable based on a Prompt Determination of Operability completed on June 30, 2014. This report addresses a condition as described in 10 CFR 50.72 (b)(3)(ii)(B). TVA is making this report consistent with the guidance of NUREG-1022 regarding the application of engineering judgment to the evaluation of reportability of an unanalyzed condition. The NRC Resident Inspector has been notified of this condition.
On June 30, 2014, SQN reported (Event 50244) that during a re-analysis conducted as part of the Fukushima Order 2.1 flooding review, a probable maximum flood (PMF) design assumption that the Watts Bar Dam west saddle dike fails completely and instantaneously at approximately 1.5 feet of overtopping, was determined to be a non-conservative flood model assumption (i.e., invalid). As a result, TVA postulated that Watts Bar Dam's east floodwall would fail, increasing the site flood level at Sequoyah Nuclear Plant (SQN) by 1.5 feet; a condition that was beyond the current licensing basis. Through subsequent analysis, TVA has demonstrated that although the west saddle dike may not completely and instantaneously fail during a PMF (as previously assumed), the consequential increase in reservoir levels does not result in a failure of the Watts Bar Dam east floodwall and would not result in an increase in the flood level at SQN. Therefore, the previously reported 10 CFR 50.72(b)(3)(ii)(B) event is being retracted. The NRC Resident Inspector has been informed of this event retraction. Notified the R2DO (Hickey). |
| Where | |
|---|---|
Sequoyah  Tennessee (NRC Region 2) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | |
| Time - Person (Reporting Time:+-3.58 h-0.149 days <br />-0.0213 weeks <br />-0.0049 months <br />) | |
| Opened: | Bill Harris 23:11 Jun 30, 2014 |
| NRC Officer: | Howie Crouch |
| Last Updated: | Aug 21, 2014 |
| 50244 - NRC Website
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Unit 1 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
Unit 2 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
WEEKMONTHYEARENS 544872020-01-22T03:18:00022 January 2020 03:18:00
[Table view]10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition ENS 521872016-08-17T21:22:00017 August 2016 21:22:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition Due to Not Meeting Single Failure Criteria ENS 519942016-06-08T19:26:0008 June 2016 19:26:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition That Could Potentially Degrade Plant Safety ENS 519352016-05-17T01:05:00017 May 2016 01:05:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition That Significantly Degrades Plant Safety ENS 518542016-04-07T19:07:0007 April 2016 19:07:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition - High Pressure Fire Suppression Isolated from Containment ENS 515272015-11-10T20:02:00010 November 2015 20:02:00 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 Unanalyzed Condition Due to Debris Dropped Into Reactor Cavity Equipment Pit ENS 502442014-07-01T02:46:0001 July 2014 02:46:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition That Could Have Resulted in an Increased Maximum Flood Level ENS 496902014-01-03T20:00:0003 January 2014 20:00:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Abgts Potentially Inoperable for Both Units Under Worst-Case Conditions ENS 485842012-12-13T00:14:00013 December 2012 00:14:00 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Risk of Possible Flooding to Ercw Building During Design Basis Flood ENS 487252009-07-28T05:00:00028 July 2009 05:00:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition That Could Have Resulted in an Increased Maximum Flood Level ENS 448142009-01-28T19:25:00028 January 2009 19:25:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Unanalyzed Condition Related to a Postulated Appendix R Fire Scenario 2020-01-22T03:18:00 | |