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Event Notification Report for August 12, 2011

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
08/11/2011 - 08/12/2011

** EVENT NUMBERS **


46230 47111 47141 47142 47143 47147

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General Information Event Number: 46230
Rep Org: GE HITACHI NUCLEAR ENERGY
Licensee: GE HITACHI NUCLEAR ENERGY
Region: 1
City: WILMINGTON State: NC
County:
License #:
Agreement: Y
Docket:
NRC Notified By: DALE E. PORTER
HQ OPS Officer: ERIC SIMPSON
Notification Date: 09/03/2010
Notification Time: 15:23 [ET]
Event Date: 09/03/2010
Event Time: [EDT]
Last Update Date: 08/12/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
RICHARD CONTE (R1DO)
EUGENE GUTHRIE (R2DO)
TAMARA BLOOMER (R3DO)
RICK DEESE (R4DO)
MIKE CHEOK (NRR)
PART 21 GP via email ()

Event Text

PART 21 - FAILURE TO INCLUDE SEISMIC INPUT IN REACTOR CONTROL BLADE CUSTOMER GUIDANCE

The following is text of a facsimile submitted by the vendor:

"GE Hitachi Nuclear Energy (GEH) has identified that engineering evaluations that support the guidance provided in SC 08-05, Revision 1, do not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. Note that the seismic loads are not a consideration in the scram timing, but rather the ability to insert the control blades. In other words, the control blades must be capable of inserting during the seismic event, but not to the timing requirements of the Technical Specifications. GEH is evaluating the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures in the BWR/2-5 plants. The ability to scram for the BWR/6 plants is not adversely affected by the seismic events. Additional evaluation is required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants in SC 08-05 Revision 1 is affected by the addition of seismic loads.

"GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10 CFR 21.21 (a)(2) to allow additional time to for this evaluation to be completed."

Affected US plants previously notified by vendor and recommended for surveillance program include: Nine Mile Point, Units 1 and 2; Fermi 2; Columbia; FitzPatrick; Pilgrim; Vermont Yankee; Grand Gulf; River Bend; Clinton; Oyster Creek; Dresden, Units 2 and 3; LaSalle, Units 1 and 2; Limerick, Units 1 and 2; Peach Bottom, Units 2 and 3; Quad Cities, Units 1 and 2; Perry, Unit 1; Duane Arnold; Cooper; Monticello; Brunswick, Units 1 and 2; Hope Creek; Hatch, Units 1 and 2; and Browns Ferry, Units 1and 2.

Affected US plants previously notified by vendor and provided information include: Susquehanna, Units 1 and 2 and Browns Ferry, Unit 3.

* * * UPDATE FROM DALE PORTER TO ERIC SIMPSON AT 1556 ON 09/27/2010 * * *

The following update was received via fax:

"This letter provides a revision to the information transmitted on September 2, 2010 in MFN 10-245 concerning an evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic input in channel-control blade interference customer guidance. Two changes have been made in Revision 1:

"1) A statement was added regarding the applicability of this issue to the ABWR and ESBWR design certification documentation.

"2) The original MFN 10-245 referenced the Safety Communication SC 08-05 R1 that was transmitted to the US NRC via MFN 08-420. The references to SC 08-05 were changed to MFN 08-420 to prevent possible confusion.

"As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR 21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420."

Notified the R1DO (Gray), R2DO (Hopper), R3DO (Orth), R4DO (Farnholtz), NRR EO (Lee) and Part 21 Group (via email).

* * * UPDATE FROM DALE PORTER TO MARK ABRAMOVITZ AT 1723 ON 12/15/2010 * * *

The following update was received via fax:

"This letter provides information concerning an on-going evaluation being performed by GE Hitachi Nuclear Energy (GEH) regarding the failure to include seismic loads in the guidance provided in MFN 08-420. As stated herein, GEH has not concluded that this is a reportable condition in accordance with the requirements of 10CFR21.21(d) and continued evaluation is required to determine the impact of a seismic event on the guidance contained in MFN 08-420.

"GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with an Operating Basis Earthquake (OBE), and a Safe Shutdown Earthquake (SSE) on BWR/2-5 plants."

GEH expects the task to be completed by August 15, 2011.

Notified the R1DO (Holody), R2DO (Henson), R3DO (Kozak), R4DO (Werner), NRR EO (Evans) and Part 21 Group (via email).

* * * UPDATE AT 1808 EDT ON 08/11/11 FROM DALE PORTER TO JOE O'HARA * * *

The following was received via fax:

"GE Hitachi Nuclear Energy (GEH) identified, in July 2010, that engineering evaluations did not address the potential impact of a seismic event on the ability to scram as it relates to the channel-control blade interference issue. GEH provided status of the on-going evaluation in [December 2010]. GEH has not completed the evaluation of the impact of the seismic loads between the fuel channel and the control blade associated with a bounding Safe Shutdown Earthquake (SSE) on BWR/2-5 plants. The scram capability is expected to be affected due to the added seismic loads at low reactor pressures [less than 1000 psig] in the BWR/2-5 plants. Additional evaluations are required to determine to what extent the maximum allowable friction limits specified for the BWR/2-5 plants are affected by the addition of SSE seismic loads at low reactor pressures.

"GEH issues this 60-Day Interim Report in accordance with the requirements set forth in 10CFR 21.21 (a)(2) to allow additional time for this evaluation to be completed."

The following sites are noted as having channel-control blade concerns:
Region 1: Nine Mile Point, Fitzpatrick, Pilgrim, Vermont Yankee, Oyster Creek, Limerick, Peach Bottom, Susquehanna, and Hope Creek
Region 2: Browns Ferry, Brunswick, Hatch,
Region 3: Fermi, Clinton, Dresden, LaSalle, Quad Cities, Perry, Duane Arnold, Monticello
Region 4: Columbia, Grand Gulf, River Bend, Cooper.

Notified R1DO (Powell), R2DO (Hopper), R3DO (Dickson), R4DO (Farnholtz) and NRR Part 21 Grp via email.

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Fuel Cycle Facility Event Number: 47111
Facility: PADUCAH GASEOUS DIFFUSION PLANT
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
                   6903 ROCKLEDGE DRIVE
                   BETHESDA, MD 20817 (301)564-3200
Region: 2
City: PADUCAH State: KY
County: McCRACKEN
License #: GDP-1
Agreement: Y
Docket: 0707001
NRC Notified By: BILLY WALLACE
HQ OPS Officer: JOE O'HARA
Notification Date: 08/01/2011
Notification Time: 11:27 [ET]
Event Date: 07/31/2011
Event Time: 10:30 [CDT]
Last Update Date: 08/11/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
76.120(c)(1) - UNPLANNED CONTAMINATION
Person (Organization):
ALAN BLAMEY (R2DO)
PETER HABIGHORST (NMSS)

Event Text

UNPLANNED CONTAMINATION AREA

"On 07/31/11 a water leak was found coming through the C-310 North NaF (Sodium Fluoride) trap room ceiling. The leak was coming from a steam condensate header located over the trap room. The leak had created a puddle, approximately 8' by 6', in the trap room. The area was checked by Health Physics and the spill area in the trap room was upgraded to a Contamination Areas at 1030 [CDT]. The condensate water is not contaminated but is spreading contamination from a contamination area above the ceiling. Decontamination of the trap room has been performed; however, due to a small amount of water that continues to leak into the area the area is being recontaminated. Access to the area has been restricted with the contamination area requirements for more than 24 hours.

"This event is reportable as a 24 hour event in accordance with 10 CFR 76.120(c)(1)(i) 'An unplanned contamination event that: Requires access to the contaminated area, by workers or the public, to be restricted for more than 24 hours by imposing additional radiological controls or by prohibiting entry into the area.'

"The NRC Senior Resident Inspector has been notified of this event.

"PGDP Problem Report No. ATRC11-1944: PGDP Event Report No. PAD-2011-12."

* * * RETRACTION ON 08/11/2011 AT 1356 EDT FROM CALVIN PITTMAN TO DAN LIVERMORE * * *

"Once the floor was decontaminated and down posted, which was within the 24-hour limit, a Nuclear Criticality Safety (NCS) approved 5.5 gallon container was placed on the floor to contain the continuing leak. Due to the fact that the water dripping into the container was contaminated, the container was posted as a contamination area. The above statement that the area was being recontaminated by a small amount of water that continued to leak is referring to the water dripping into the bucket. A review of the sequence of actions to contain the water leak and decontaminate the area resulted in the conclusion that the contaminated area was down posted within the 24-hour limit. The fact that the NCS approved 5.5 gallon container placed on the floor to contain the leak was posted as a contamination area is not considered a continuation of the spread of contamination event. The spread of contamination onto the floor had been stopped and the floor decontaminated. With the leak contained, the floor decontaminated, and returned to the radiological control level that existed prior to the leak within the 24-hour limit, the event notification was not required. Thus, the notification is being retracted."

The licensee has notified the NRC Resident Inspector.

Notified R2DO (Hopper) and NMSS EO (Hiltz).

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Power Reactor Event Number: 47141
Facility: NINE MILE POINT
Region: 1 State: NY
Unit: [ ] [2] [ ]
RX Type: [1] GE-2,[2] GE-5
NRC Notified By: JOHN APRIL
HQ OPS Officer: BILL HUFFMAN
Notification Date: 08/11/2011
Notification Time: 01:05 [ET]
Event Date: 08/11/2011
Event Time: 00:16 [EDT]
Last Update Date: 08/11/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
RAY POWELL (R1DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 M/R Y 15 Power Operation 0 Hot Shutdown

Event Text

MANUAL REACTOR SCRAM INSERTED DUE TO A FEEDWATER LEAK ON A MIN FLOW LINE

"NMP2 inserted a manual reactor scram due to a through wall leak on feedwater pump 2FWS-P1A minimum flow line. 2FWS-P1A has been removed from service to minimize the leak. All control rods inserted and all systems functioned as designed. The unit will remain in hot shutdown until plant startup."

All systems functioned as required and the scram was uncomplicated.

The licensee plans to issue a press release and has informed the NRC Resident Inspector and the New York State Public Service Commission.

* * * UPDATE AT 1035 EDT ON 8/11/11 FROM MANLEY TO HUFFMAN * * *

The licensee will not be issuing a press release regarding this manual scram. R1DO (Powell) notified.

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Power Reactor Event Number: 47142
Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: JEFFREY A. MIELL
HQ OPS Officer: STEVE SANDIN
Notification Date: 08/11/2011
Notification Time: 07:02 [ET]
Event Date: 08/11/2011
Event Time: 03:27 [CDT]
Last Update Date: 08/11/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
50.72(b)(3)(v)(B) - POT RHR INOP
Person (Organization):
BILLY DICKSON (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 100 Power Operation 68 Power Operation

Event Text

INITIATION OF TECHNICAL SPECIFICATION REQUIRED SHUTDOWN

"At 0327 [CDT] on August 11, 2011, NextEra Energy Duane Arnold commenced a reactor shutdown as required by TS 3.7.2.B for both River Water Supply subsystems inoperable. At 0008 [CDT], both River Water Supply subsystems were declared inoperable as a result of high differential pressure across the intake travelling screens. TS 3.7.2.B requires that one River Water Supply subsystem be restored or commence reactor shutdown. While the River Water Supply system is currently supplying rated flow, high differential pressure across the travelling screens indicates the potential for eventual degraded flow. This event is being reported pursuant to the requirements of 10 CFR 50.72(b)(2)(i), as a TS required shutdown.

"Additionally, the current condition of both trains of River Water Supply inoperable also potentially affects the plant capability to remove residual heat. Therefore, this event is also being reported under 10 CFR 50.72(b )(3)(v)(B), as an event that could prevent fulfillment of a safety function.

"The NRC Resident Inspector has been notified."

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Power Reactor Event Number: 47143
Facility: DUANE ARNOLD
Region: 3 State: IA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: BOB MURRELL
HQ OPS Officer: DAN LIVERMORE
Notification Date: 08/11/2011
Notification Time: 15:29 [ET]
Event Date: 08/11/2011
Event Time: 11:38 [CDT]
Last Update Date: 08/11/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
BILLY DICKSON (R3DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 15 Power Operation 0 Hot Shutdown

Event Text

UNPLANNED PRIMARY CONTAINMENT ISOLATIONS

"At 1138 CDT, while in the process of shutting down as required by Technical Specifications (Reference EN# 47142), with the reactor at approximately 15 percent power, a manual scram was inserted in order to complete the TS Required Action of being in Mode 3 within 12 hours. Upon inserting the manual scram, reactor water level dropped below 170 inches resulting in Primary Containment Isolation System (PCIS) Groups 2, 3 and 4 being received. This reactor water level response is considered normal following a reactor scram from power due to void collapse in the reactor vessel. Reactor water level is currently being controlled in the normal band. All PCIS group isolations went to completion and were subsequently reset. The PCIS isolations all functioned properly. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A)."

The NRC Resident Inspector has been notified.

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Power Reactor Event Number: 47147
Facility: TURKEY POINT
Region: 2 State: FL
Unit: [3] [ ] [ ]
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: NICHOLAS DESANTIS
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 08/11/2011
Notification Time: 21:07 [ET]
Event Date: 08/11/2011
Event Time: 16:31 [EDT]
Last Update Date: 08/11/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
GEORGE HOPPER (R2DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
3 N Y 100 Power Operation 100 Power Operation

Event Text

LOSS OF INTAKE COOLING WATER FOR 20 MINUTES

"At 1631 hours on Thursday, August 11, 2011, Turkey Point Unit 3 lost the function of the Intake Cooling Water (ICW) System for approximately 20 minutes. This is reportable in accordance with 10 CFR 50.72(b)(3)(v)(B) and (D). Loss of function occurred when manual valve 3-50-406, ICW/Component Cooling Water (CCW) Heat Exchanger Outlet Valve, failed closed. Valve 3-50-406 is located in the common discharge line preventing flow through both ICW headers to the CCW heat exchangers. Operations Department personnel opened the bypass valves associated with the failed valve and restored ICW flow to the CCW heat exchangers at approximately 1651 hours.

"Presently, Unit 3 is in Mode 1 at 100% power with all systems stable."

The failed valve is a manual 24" butterfly valve. The bypass valve can maintain full flow. During this event, Operations personnel monitored CCW temperature. Unit 3 briefly entered T.S. 3.0.3 before restoring flow using the bypass. A high RCP temperature alarm was received following the momentary loss of ICW flow. The valve failure was accompanied with a loud bang.

The licensee notified the NRC Resident Inspector.

Page Last Reviewed/Updated Thursday, March 29, 2012
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