ENS 48845
ENS Event | |
|---|---|
00:00 Nov 20, 2012 | |
| Title | Control Rod Drive Check Valve in Service Leak Test Failure |
| Event Description | Control Rod Drive (CRD) Check valves 3-0399-593 and 3-0399-594 are installed on the CRD supply header inside the reactor building to prevent unfiltered radiological release from primary containment past secondary containment through the CRD supply header. The [Design Basis Accident] DBA [Loss of Coolant Accident] LOCA dose analysis does not postulate the CRD supply piping as a release path. These valves were added in response to NRC IN 90-78. Their function is described in the Dresden UFSAR section 4.6.4.6.
CRD check valves 3-0399-593 and 3-0399-594 failed their as found In Service Testing (IST) seat leakage test during refuel outage D3R22. Based on the condition of as-found failed seat leakage tests, Engineering determined that CRD check valves 3-0399-593 and 3-0399-594 would not have been able to perform their design functions as described in the UFSAR to maintain the offsite dose and control room dose within regulatory limits. The valves were repaired in D3R22 and successfully retested. The licensee notified the NRC Resident Inspector.
Based on an evaluation of the current plant design and the overly conservative assumptions made in the 1996 dose estimate, the dose that would be experienced as a result of a DBA LOCA with LOOP [Loss of Offsite Power] due to back leakage through the control rod drive system to the condensate storage tanks [CST] has been reassessed. The initial estimations of dose did not consider the as-built piping configuration of the CRD system. The backflow would be greatly reduced due to the lower differential pressure between the Drywell and CSTs that would exist following a DBA-LOOP. This differential pressure experienced following a DBA-LOOP would not result in a significant backflow due to excess flow check valves and restricting orifices in the current CRD piping configuration. Additionally, in the event that any backflow did occur, the volume of water in the CST would provide a high decontamination factor and dilution. Therefore backflows that could be experienced following a DBA-LOOP would not produce doses that would exceed current limits. Based on this assessment, the unanalyzed condition that was reported on March 22 (EN 48845) is being retracted. The Licensee has notified the NRC Resident Inspector. Notified R3DO (Kozak). |
| Where | |
|---|---|
Dresden  Illinois (NRC Region 3) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | |
| Time - Person (Reporting Time:+2948.97 h122.874 days <br />17.553 weeks <br />4.039 months <br />) | |
| Opened: | Patrick Haarhoff 19:58 Mar 22, 2013 |
| NRC Officer: | Pete Snyder |
| Last Updated: | May 20, 2013 |
| 48845 - NRC Website
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Unit 3 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
WEEKMONTHYEARENS 532042018-02-12T18:00:00012 February 2018 18:00:00
[Table view]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 En Revision Imported Date 10/14/2019 ENS 488452012-11-20T00:00:00020 November 2012 00:00:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Control Rod Drive Check Valve in Service Leak Test Failure ENS 441682008-04-25T16:30:00025 April 2008 16:30:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Non-Conservative Core Spray Flow in Dresden Nuclear Power Station Loca Analysis ENS 413702005-02-04T01:15:0004 February 2005 01:15:00 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 4160 Volt Relaying and Metering Single Failure Vulnerability ENS 403112003-11-10T13:20:00010 November 2003 13:20:00 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 Bypassing the Pressure Suppression Function of the Torus 2018-02-12T18:00:00 | |