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Event Notification Report for March 31, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
03/30/2006 - 03/31/2006

** EVENT NUMBERS **


42453 42454 42455

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Hospital Event Number: 42453
Rep Org: INDIANA UNIVERSITY MEDICAL CENTER
Licensee: INDIANA UNIVERSITY MEDICAL CENTER
Region: 3
City: INDIANAPOLIS State: IN
County:
License #: 13-02752-03
Agreement: N
Docket:
NRC Notified By: MACK RICHARD
HQ OPS Officer: PETE SNYDER
Notification Date: 03/30/2006
Notification Time: 11:41 [ET]
Event Date: 03/29/2006
Event Time: 15:30 [CST]
Last Update Date: 03/30/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
Person (Organization):
SONIA BURGESS (R3)
GREG MORELL (NMSS)

Event Text

DIFFERENT LENGTH BUCKET AND APPLICATOR USED DURING BRACHYTHERAPY DOSE

Following a vaginal/cervical/uterine brachytherapy dose it was noted during a side by side comparison that a different length bucket and applicator were used. Because the applicator was shorter than the bucket, the applicator did not reach the end of the bucket during administration of the dose. Review of the x-ray taken to confirm placement during the exam confirmed that a different dose distribution was given to the patient than originally intended. It is estimated that the dose varied by greater than 20 percent. The dose was lower than the intended dose.

The prescribing physician did not note any apparent ill effects to the patient during a follow-up physical examination. This occurred because the licensee did not do a direct physical comparison of the bucket and applicator prior the exam. The licensee sorted all applicators and buckets following the event to create matched sets. The licensee is considering modifying procedures to include physical comparison of the applicator and bucket in the future. The oncology physician was to inform the patient of the differing dose.

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Power Reactor Event Number: 42454
Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: JOHN DAMPF
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/30/2006
Notification Time: 21:40 [ET]
Event Date: 03/30/2006
Event Time: 16:50 [CST]
Last Update Date: 03/30/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
DALE POWERS (R4)

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

Event Text

INADEQUATE OPERATOR RESPONSE TIME FOR COMPONENT COOLING WATER SYSTEM REALIGNMENT DURING A LARGE BREAK LOCA

"At 1650 on March 30, 2006, a concern was identified where the operators in the training simulator could not complete realignment of the component cooling water (CCW) flow to the residual heat removal (RHR) heat exchanger in a timely manner under certain accident scenarios. This could result in exceeding the maximum design temperature of the CCW system. In addition, assumptions made in the containment pressure and temperature analysis following a large break loss of coolant accident (LOCA) are non-conservative with respect to when CCW flow to the RHR heat exchangers is manually established in accordance with emergency operating procedures.

"Callaway plant FSAR indicates CCW system flow is manually aligned to the RHR heat exchangers prior to the recirculation phase of emergency core cooling system (ECCS). If the automatic transfer of the RHR pumps to cold leg recirculation, which happens at the Lo-Lo-1 level of the refueling water storage tank (RWST), occurs before CCW flow has been manually aligned to the RHR heat exchanger, containment sump water at temperatures up to 270F can be circulated through the RHR heat exchanger without CCW flow on the other side of the heat exchanger. The CCW side of the heat exchanger would contain stagnant water. This water can heat up quickly with no established flow and exceed the design rated temperature of the system.

"Recent simulator scenarios of large break LOCAs have shown that the CCW alignment is not reached before the Lo-Lo-1 RWST alarm level is reached. The CCW alignment is completed as part of procedure ES-1.3, Transfer to Cold Leg Recirculation. A review of two large break LOCA scenarios completed on 3-20-06 show that it takes between 1:00 and 1:30 minutes to initiate the step to align CCW to the RHR heat exchangers and takes between 3:00 and 4:30 minutes to complete the alignment.


"In addition to CCW system temperature concerns, an assumption that CCW flow is established to the RHR heat exchanger prior to reaching the Lo-Lo-1 level in the RWST is made in the containment temperature and pressure response analyses. As a result, a failure to establish CCW flow to the RHR switchover would result in an adverse impact on the inputs used in the Licensing Bases Containment Analysis. However, preliminary sensitivity runs using containment analyses codes indicate that post-peak temperature and pressure are not significantly affected by this issue.

"Actions taken:

"1650 Declared both trains of CCW inoperable. Declared both trains of ECCS inoperable and entered Technical Specification 3.0.3

"1710 Both trains of CCW aligned with flow through the RHR heat exchangers

"1711 Exited Technical Specification 3.0.3 "

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42455
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: WALT COOMBS
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 03/31/2006
Notification Time: 02:26 [ET]
Event Date: 03/30/2006
Event Time: 21:00 [CST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
SONIA BURGESS (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N N 0 Refueling 0 Refueling

Event Text

FEEDWATER CHECK VALVE LEAKAGE ABOVE ALLOWABLE LIMITS

"At 21:00 hours on 3/30/06, Unit Two feed water header check valves 2-0220-58A and 2-0220-62A LLRT [local leak rate test] results were both indeterminate, as a result both were greater than the allowable La containment leakage rate allowed by Tech Spec 5.5.12." La is defined as the maximum allowable leak rate at a specified pressure.

"This is reportable under 10CFR50.72(b)(3)(ii)."

The "indeterminate" leak rate was higher than the test equipment could read.

The licensee notified the NRC Resident Inspector.

Page Last Reviewed/Updated Friday, March 30, 2012
Friday, March 30, 2012