Event Notification Report for November 03, 2011

ISFSI BUILDING AIR VENT BLOCKED FOR 12 MINUTES

"At 1009 (all times MDT) today, 11/3/11, a Security Officer performing routine rounds noted 95-100% blockage on the inlet screens at the Fort St. Vrain Independent Spent Fuel Storage Installation (FSV ISFSI). The FSV ISFSI is located near Platteville, Colorado. The FSV ISFSI safely stores used fuel from the Fort St. Vrain Nuclear Generating Station (FSV NGS). The NGS has been decommissioned and released for unrestricted use. The fuel is stored at the ISFSI under NRC license SNM-2504. The inlet screens are in place to provide a cooling path for the used fuel.

"The Security Officer immediately notified the Emergency Coordinator, who directed the screens to be cleared of the blockage at 1010. The blockage was removed at 1021, at which time the event was terminated. The blockage was caused by frost, which built up due to dense fog, high humidity, and low temperatures. Removal was accomplished by lightly hitting the screens by hand. No further action is necessary.

"Per FSV ISFSI Limiting Condition for Operation (LCO) 3.1, inlet screen blockage which equals or exceeds 95 percent must be cleared within 24 hours. The blockage was cleared in 12 minutes. Thus the REQUIRED ACTION was satisfactorily completed, and the CONDITION was exited.

"Per the FSV ISFSI Emergency Plan Implementing Procedure (EPI)-102, Emergency Action Level 1NE.6, 95% or greater blockage of the inlet screens constitutes a 1 hour NON-EMERGENCY event. Required notifications were made by the Warning Communications Center (WCC) in Idaho."

There was no increase in building temperature during this event.

UNANALYZED CONDITION FOR A POSTULATED CONTROL ROOM FIRE

"During the 2011 Triennial fire protection inspection, it was determined that the off normal procedure for control room evacuation due to fire has two defects. It does not adequately protect the steam generators from overfilling and possibly damaging the turbine driven auxiliary feedwater pump. In addition it does not protect the reactor coolant system pressurizer from filling to above 100% indicated water level, possibly causing the primary system to go solid. Both of these issues are results of inadequate assumptions used in the Post Fire Safe Shutdown Analysis of a fire in the Control Room.

"Compensatory measure of hourly fire watch for the control room is in place. The procedure for control room evacuation due to fire is being revised to include compensatory actions that will address the above events."

The NRC Resident Inspector has been notified.

DIESEL GENERATOR STARTING AIR NOT COMPLIANT WITH APPENDIX R REQUIREMENTS

"During the conduct of a system vulnerability assessment by Engineering on the emergency diesel system. a discussion was held regarding the impact of a fire on the DG starting air system.

"Subsequently it was recognized that components of the system were vulnerable to damage during a fire. A review of the Appendix R analysis was conducted and it was determined that this vulnerability was not analyzed in the evaluation.

"This is reportable to Unit 2, as Unit 1 has a separate diesel building for the 1-Alpha emergency diesel."

The NRC Resident Inspector has been notified.

* * * RETRACTION ON 12/29/11 AT 1556 EST FROM GIOFFRE TO HUFFMAN * * *

"The Appendix R analysis was updated and no changes to the plant or procedures resulted. Therefore, no unanalyzed condition that significantly degraded plant safety existed. This event is being retracted. The NRC Resident Inspector has been notified."

R1DO (Joustra) notified.

UNANALYZED CONDITION - CONTROL ROOM VENTILATION SINGLE POINT VULNERABILITY

"On November 3, 2011, at 1550 PDT, operators determined that control room ventilation system (CRVS) contained a single failure vulnerability whereby unfiltered air supplied to the control room could exceed the flowrates used in the licensing basis analyses of design basis accident (DBA) consequences. This vulnerability was discovered during performance of control room inleakage testing required by TS SR 3.7.10.5.

"It was determined that the control room pressurization system airflow could bypass the supply filter if the CRVS booster fan in the associated train was not operating. This would allow as much as 800 cubic feet per minute of unfiltered air to be delivered to the control room following an accident that results in initiation of the CRVS pressurization mode. Operators would correct the condition approximately 10 minutes after a safety injection by manually selecting the train's redundant booster fan in accordance with existing proceduralized actions specified in the DCPP emergency procedure E-0 Appendix E. This period of unfiltered air supply to the control room due to a single failure of a CRVS booster fan had not been previously analyzed and could have potentially resulted in operator dose greater than contained in plant analyses.

"Plant staff verified that all components and redundant components in each ventilation train are currently OPERABLE. Plant staff has implemented additional compensatory measures by issuing a shift order to require that TS Action 3.7.10.A be entered for unavailability of either of the two CRVS booster fans in each CRVS train. Additionally, evaluation of the new unfiltered inleakage may result in more restrictive administrative controls to ensure operator doses are maintained less than the FSAR accident analyses."

The licensee informed the NRC Resident Inspector.

AGREEMENT STATE REPORT - RADIOGRAPHY SOURCE FAILS TO RETRACT

The following report was received via e-mail:

"Valley Industrial X-ray and Inspection Services radiographers were unable to retract an Ir-192 source assembly into its fully shielded position and secure it in this position. The device is an Industrial Nuclear Co, Model IR-100 s/n 4100 camera which contained 49 Ci of Ir-192. Emergency procedures were enacted by evacuating the area 200 yards around the exposure device and calling the RSO for assistance. Radiation safety personnel were able to retract the source into the shield and the device was secured by 1600 [PDT]. The licensee removed the Ir-192 source from the camera and placed it into a spare shield. The IR-100 camera (without a source assembly) was sent to Industrial Nuclear Co. in San Leandro, CA for evaluation. No personnel overexposure occurred due to this event. RHB [Radiologic Health Branch] is following up with the licensee as to the cause of the event."

California Report: 5010-110311

AGREEMENT STATE REPORT - GADOLINIUM SOURCE LOST

The following report was received via e-mail:

"On November 3, 2011, at approximately 9:30AM, a 709.36 mCi Gadolinium-153 sealed source was discovered to be missing from the back of a LIXI, Inc. employee's RV in Phoenix, Arizona. The source was last seen on Monday, October 31st at approximately 8PM in Abilene, Texas.

"The investigation into this event is ongoing.

"The U.S. NRC, AZ governor's office, Texas, and Illinois have been notified."

Arizona Report: 11-012

* * * UPDATE ON 11/3/2011 AT 1541 FROM AUBREY GODWIN TO MARK ABRAMOVITZ * * *

The following information was received via e-mail:

"At 11:45AM, the licensee was notified that the source was left at the PepsiCo in Abilene, Texas. The licensee is flying back to Texas to pick up the source at approximately 7:30PM tonight.

"The U.S. NRC, AZ Governor's Office, Texas, and Illinois have been notified."

Notified the R4DO (Gaddy) and FSME (Camper).

THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL

Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf

This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint. Therefore is it being categorized as a less than Category 3 source.

INVALID ACTUATION OF A SAFETY INJECTION VALVE

"This telephone notification is being made in lieu of submitting a written LER under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of MOV 8803A, High Head Safety Injection (HHSI) to RCS Cold Leg, resulting in ECCS injection into the RCS.

"On November 3, 2011 at 03:48 CST, during a maintenance replacement activity for the K604 slave relay in Solid State Protection System, MOV 8803A unexpectedly opened. The relay was not recognized to be in a latched condition when the lead associated with MOV 8803A was landed causing it to stroke open. This actuation was invalid since the stroke signal was not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. MOV 8803A was subsequently closed from the control room in approximately 2 minutes and power was removed at 03:57 CST to prevent another stroke. After investigation, the relay was installed in the correct configuration and system tested to restore operability. This event did not adversely affect the safe operation of the plant or health and safety of the public.

"The following required information is being submitted per NUREG-1022, Rev. 2:
(a) MOV 8803A is an 'A' Train component.
(b) The stroke open of MOV 8803A and ECCS injection is a partial train actuation.
(c) Given the conditions, MOV 8803A functioned as expected and all other systems functioned per design."

The licensee notified the NRC Resident Inspector

AGREEMENT STATE REPORT- DELIVERED PATIENT DOSE DIFFERENT THAN PRESCRIBED

The following information was received by email:

"Based on protocol, a dose of 120 Gy (1.79 GBq) was prescribed. Upon completion of the treatment, survey of the Nalgene waste container measured higher than expected. Ensuing calculations resulted in dose delivered to be 85 Gy (1.24 GBq); greater than 20% variation from prescribed dose. All drapes, towels, etc were surveyed with no evidence of radioactivity present, therefore assuring no contamination present. Contents of the waste container were measured separately to locate the source of residual activity. The readings indicated minimal activity in the Y -90 vial; readings of the patient delivery microcatheter were indicative of residual microspheres. The treatment protocol was followed with no variations of procedure. As is typical, 3 saline flushes were made of the catheter including several vigorous flushes to dislodge any microspheres as recommended by Nordion, the product manufacturer. No high pressure was detected at any point during infusion which would trigger the pressure valve and deliver saline in the overflow vial. There was no build up of particles in the hub of the delivery catheters as inspected throughout the procedure. Measurements over the length of the catheter revealed greatest activity in the proximal portion of the catheter with little-to-no activity in the tip. Nordion has been contacted. In the future, survey of the catheter prior to disconnecting it for disposal may help detect the build-up of particles."

Oregon Incident: 11-0037

A Medical Event may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient.