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| {{Adams
| | #REDIRECT [[05000456/LER-2010-001, Regarding Reactor Trip Due to Water Intrusion in Breakers Causing Circulating Water Pump Trips and Resulting in Loss of Condenser Vacuum]] |
| | number = ML110620091
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| | issue date = 03/02/2011
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| | title = Regarding Reactor Trip Due to Water Intrusion in Breakers Causing Circulating Water Pump Trips and Resulting in Loss of Condenser Vacuum
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| | author name = Enright D
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| | author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
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| | addressee name =
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| | addressee affiliation = NRC/NRR, NRC/Document Control Desk
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| | docket = 05000456
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| | license number = NPF-072
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| | contact person =
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| | case reference number = BW110016
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| | document report number = LER 10-001-01
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| | document type = Letter, Licensee Event Report (LER)
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| | page count = 5
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| }}
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| {{LER
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| | Title = Regarding Reactor Trip Due to Water Intrusion in Breakers Causing Circulating Water Pump Trips and Resulting in Loss of Condenser Vacuum
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| | Plant =
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| | Reporting criterion = 10 CFR 50.73(a)(2)(i), 10 CFR 50.73(a)(2)(vii), 10 CFR 50.73(a)(2)(ii)(A), 10 CFR 50.73(a)(2)(viii)(A), 10 CFR 50.73(a)(2)(ii)(B), 10 CFR 50.73(a)(2)(viii)(B), 10 CFR 50.73(a)(2)(iii), 10 CFR 50.73(a)(2)(ix)(A), 10 CFR 50.73(a)(2)(iv)(A), 10 CFR 50.73(a)(2)(x), 10 CFR 50.73(a)(2)(v)(A), 10 CFR 50.73(a)(2)(v)(B), 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(v), 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(iv)(B)
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| | Power level =
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| | Mode =
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| | Docket = 05000456
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| | LER year = 2010
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| | LER number = 1
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| | LER revision = 1
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| | Event date =
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| | Report date =
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| | ENS =
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| | abstract =
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| }}
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| =text=
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| {{#Wiki_filter:E 10 CFR 50.73 March 2, 2011 BW110016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456
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| ==Subject:==
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| Supplemental Licensee Event Report 2010-001 Unit 1 Reactor Trip Due to Water Intrusion in Breakers Causing Circulating Water Pump Trips and ReSUlting in Loss of Condenser Vacuum This is a supplement to Licensee Event Report 2010-001-00 which was submitted on October 15, 2010. This supplement contains updated information regarding the root cause and corrective actions of the issue.
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| There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Ronald Gaston, Regulatory Assurance Manager, at (815) 417-2800.
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| Daniel J. Enrigh Site Vice President Braidwood Station
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| ==Enclosure:==
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| LER 2010-001-01 cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)
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| NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010) the NRC may digits/characters for each block) not conduct or sponsor. and a person is not required to respond to. the information collection.
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| : 13. PAGE Braidwood Station, Unit 1 05000456 1 of 4
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| : 4. TITLE Reactor Trip Due to Water Intrusion in Breakers Causing Circulating Water Pump Trips and Resulting in Loss of Condenser Vacuum
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| : 5. EVENT DATE
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| : 6. LER NUMBER
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| : 7. REPORT DATE
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| : 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIALIREV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
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| MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 08 16 2010 2010 - 001 - 01 03 02 2011 N/A N/A
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| : 9. OPERATING MODE
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| : 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C)
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| [gI 50.73(a)(2)(vii) 1 o 20.2201(d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)
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| : 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1)(ii)(A)
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| [gI 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) 100 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71(a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) o 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
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| Specify in Abstract below or in Following the reactor trip, the AF pumps auto started on low-low steam generator [SJ] water levels. Due to the loss of bus 133V, steam dumps [SB] were unavailable and core cooling was maintained from the main steam [SB] power operated relief valves (PORVs).
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| Operator response to the trip was proper and safety systems and controls performed as expected with the exception of the following:
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| The main steam relief valve 1MS016D lifted early due to age related spring relaxation, and did not reseat until main steamline [SB] pressure was reduced to 918 psig. Unit 1 subsequently transitioned to Mode 5 (cold shutdown) and the valve replaced.
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| The motor control center 131X1 did not energize, preventing two valves from being energized - the safety injection pumps [BO] cold leg isolation valve 1S18835, and the residual heat removal [BP] to cold legs 1A and 1D isolation valve 1S18809A.
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| As a result of CW forebay material stirred up by the recycle flow from the 1B CW pump through the idle 1A and 1C CW pumps, essential service water (SX) [BI] pump discharge pressure was low and the differential pressure across the SX strainer was high. A second SX pump was started to restore SX system pressure.
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| This event is reportable under 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of any of the systems listed in 10 CFR 50.73(a)(2)(iv)(B) including any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical, and actuation of the PWR auxiliary feedwater system.
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| ~.
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| ==Cause of Event==
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| The root causes for this event were determined to be:
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| 1.
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| An inadequate design of the AF standpipes. Braidwood did not implement an effective design configuration which would have prevented the water spills from the AF suction standpipes.
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| 2.
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| A lack of sensitivity evolved that tolerated long-term uncontained water issues with inadequate evaluation for potential impacts. This organizational cause was considered a historical issue rather than a current issue and is being addressed via the corrective action program.
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| The installation of the standpipes in 1986 improved the low net positive suction head problem with the motor driven AF pumps. However, it resulted in the water level in the standpipe momentarily rising and overflowing, and introducec the potential for water spills from the standpipe during condenser hotwell rejection operations.
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| The CST header, in addition to containing the suction line to the AF pumps and the standpipe, is also used as the flo\\/\\;
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| path of condenser hotwell reject water from the hotwell to the CST. When the hotwell level becomes high, the excess water is rejected back to the CST through this line. With unit perturbations, ranging from unit trips to condensate and condensate booster pump swap operations, the level in the hotwell often changes. If the level becomes high enough, the controls initiate automatic hotwell rejection flow to the CST. When an automatic hotwell rejection occurs, the level in the standpipe will rise, and if sufficient, would result in an overflow. The overflow was onto the floor of the 451 elevation of the turbine building.
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| D.
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| ==Safety Consequences==
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| 2010 001 01 There were no safety consequences impacting plant or public safety as a result of this event.
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| For the loss of condenser vacuum, the systems and controls for managing this type of condition worked as expected.
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| The low vacuum trip setpoint removed the main turbine from service as expected. With the trip of the main turbine, an automatic reactor trip also occurred as expected. The main steam relief valve 1MS016D lifted early due to age related spring relaxation, and reseated when main steamline pressure was reduced to 918 psig during Unit 1 cooldown. All other safety systems and controls performed as expected.
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| The steam released from the opened 1MS016D valve and the PORVs contained tritium. After the 1MS016D valve reseated, steam release continued through the PORVs during the Unit 1 cooldown until shutdown cooling was established from the residual heat removal system [BP]. The offsite dose resulting from this release was not significant since tritium was the only isotope released and the dose impact of tritium is low. Calculated off-site dose was 4.59E-6 millirem.
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| This event did not result in a safety system functional failure.
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| ~.
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| ==Corrective Actions==
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| The corrective action to prevent recurrence is to install a design feature on the AF suction standpipe which prevents water spill events. The organizational cause was considered a historical issue rather than a current issue and is being addressed via the corrective action program.
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| Other corrective actions include:
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| CST level was limited to prevent another AF standpipe siphon event.
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| An operating configuration change was implemented to valves controlling a condenser hotwell rejection, to prevent AF suction standpipe overflows from hotwell rejections.
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| Develop a plan to identify long-term water leaks, spills, other uncontained fluid conditions or any degraded or abnormal conditions that are not captured by the station processes.
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| Develop a process for Byron and Braidwood to identify differences in their critical systems in design, alignment, operations or maintenance practices, to evaluate any differences for the best practices, define risks, and have actions assigned to gain commonality.
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| Both the inability to energize 1SI8835 and 1SI8809A valves and the issue with the high differential pressure across Unit 1 SX strainer have been addressed in the corrective action program.
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| ==Previous Occurrences==
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| There have been no previous, similar events identified at the Braidwood Station.
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| G.
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| ==Component Failure Data==
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| Manufacturer N/ANomenclature N/A Model N/A Mfg. Part Number N/A
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| }}
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