Event Notification Report for April 30, 2002

VA REPORTS LOST RADIOACTIVE MATERIAL LATER LOCATED IN A LANDFILL

"1. Part 20.2201(a) [telephone report]:

a. A telephone report was not made since the circumstances related to the loss of licensed materials were identified to the licensee by the Nuclear Regulatory Commission (NRC) and are not based on specific information verified by the license or on official records.

b. Neither the NRC nor the state regulatory agency (Illinois Department of Nuclear Safety) with purview over the landfill has provided the licensee any information other than telephone summaries and digital photographs.

C. The estimate of the quantity of lost licensed materials was completed post hoc and is not based on measurements performed by the licensee.

d. The licensee became aware of loss of licensed materials on March 29, 2002.

2. Part 20.2201(b)(i) [description of the licensed material involved, including kind, quantity, and chemical and physical form]:

a. The licensed material was I-131 as identified by spectral analysis at the landfill by the state regulatory agency. Since the material could not be recovered, the exact chemical form could not be determined. However, in light of subsequent investigation, the form was probably inorganic NaI bound to Sephadex beads.

b. The gamma factor for I-131 is 2.2 R./hr per mCi @1cm.

c. The state regulatory agency reported the following exposure rates measured using a Bicron internal chamber meter.

(1) 125 mR/hr @ contact

(2) 50 mR/hr through side of truck

(3) 0.65 mR/hr @ 1 meter

d. The three different readings appear to be somewhat contradictory. However, since the actual distance for the first two measurements is uncertain and less error is likely in exposure rate measurements at a known distance. the third measurement is considered the most accurate to use to estimate activity.

e. The waste materials from the laboratory were not specifically recovered or identified. The following three possible basis were used to estimate activity: the exposure rate measurements by the state regulatory agency, the usual laboratory protocols, and recreation of exposure rate measurements in a laboratory setting.

(1) Using corrected distances for the exposure rate measurements and the gamma factor for I-131, the measurement at one meter calculates to an estimated activity of approximately 3 mCi.

(2) The usual laboratory protocol is to use either 1 or 2 mCi I-131 run through a column. The highest activity possible on the column is 1.9 mCi, corresponding to a maximum doping of 2 mCi, with the poorest binding success of 5%.

(3) Laboratory measurements were taken using a 16 uCi I-131 capsule and a GM detector calibrated to CS-137 and with a thick-walled tube. The exposure rate measurement at contact with the thin plastic container with the capsule nearest the probe was 45 mR/hr. This represents the closest likely proximity implied by contact. Under this geometry, a reading of 125 mR/hr could be produced by as little as 44 uCi I-131. However, the rate drops to less than 0.05 mR/hr at 6 inches, due in part to the influence of the beta component. This indicates an activity in the millicurie range was involved in the incident, and the contact exposure rate measurement was likely not taken at contact, but at about 8 cm.

(4) The estimate for activity is approximately 2 mCi.

3. Part 20.2201(b)(ii) [a description of the circumstances under which the loss or theft occurred]:

a. The loss of licensed materials involved the Roxanna, Illinois, landfill. The loss was identified by the landfill operator during routine radiation monitoring for incoming waste shipments on March 28, 2002. The landfill operator reported higher than expected radiation monitoring results and notified the state regulatory agency. The state contacted the NRC. The NRC contacted the VA National Health Physics Program who then contacted the licensee.

b. The loss involved I-131 contaminated glassware which was identified in a truck hauling waste from the licensee to the landfill. The landfill is approximately 20 miles from the licensee, John Cochran Division, in downtown St. Louis: The waste was probably generated in a research laboratory in Building 1 of the John Cochran Division.

c. Workers assigned to the licensee Environmental Management Service (EMS) collect waste at the end of each work day from research laboratories. The waste is initially placed in carts and then relocated to a dumpster at the loading dock. The contract waste carrier empties the waste collected in the dumpster on a daily basis, normally at night. The waste is then transported to a transfer station and consolidated with other waste. The consolidated waste is relocated to the landfill using a semi-trailer.

4. Part 20.2201(b)(iii) [a statement of disposition. or probable disposition, of the licensed material involved]

a. Disposition of the radioactive materials was under the purview of the state regulatory agency.

b. The state regulatory agency directed the landfill operator to bury the waste materials.

5. Part 20.2201(b)(iv) [exposures of individuals to radiation, circumstances under which the exposures occurred, and the possible total effective dose equivalent to persons in unrestricted areas]:

a. The following individuals had the most potential for exposure

(1) EMS (Environmental Management Service, i.e. Housekeeper) personnel picking up waste and transporting it to dumpster

(2) Driver of truck hauling waste from VA to transfer point ( first truck )

(3) Driver of truck hauling consolidated waste from transfer point to landfill ( 2nd truck )

(4) State of Illinois (IDNS) responding personnel

(5) Landfill personnel were not involved in the effort to recover the contaminated object. The potential exposure of personnel other than the above must be considered small in comparison.

b. Estimated exposure

(1) EMS workers

Estimated duration of exposure = 10 min (0.16 hr)
Average proximity 3 feet (bag of waste is carried in large cart with other bags)
Inverse-square correction, using 0.65 mR/hr at 1.13 meters 0.83 mR/hr
Add 20% for lack of shielding by dump truck =1.0 mR/hr
Estimated exposure 1.0 mR/hr x 0.16 hr = 0.16 mR

(2) Driver of first truck

Estimated duration of exposure 1 hr
Average proximity = 10 feet (3 m)
Inverse square correction, using at 0.65 mR/hr @ 1.13 m. through truck bed

0.65 mR/hr x (1.13/3)squared = 0.09 mR/hr

Total estimated exposure: 1 hr x 0.09 mR/hr = 0.09 mR

(3) Driver of 2nd truck (semi-trailer) .

Estimated duration of exposure = 1 hr
Average proximity 17 feet (5 m)
Inverse square calculation, staring at 0.65 mr/hr @ 1.13 m, through truck bed

0.65 mR./hr x (1.13/5) squared = 0.03 mR/hr

Total estimated exposure: 1 hr x 0.03 mR/hr =0.03 mR

(4) IDNS responding personnel = N/A; occupationally exposed and monitored

6. Part 20.2201(b)(v) [actions that have been taken, or will be taken, to recover the material]:

a. Disposition of the radioactive materials was under the purview of the state regulatory agency.

b. The state regulatory agency directed the landfill operator to bury the waste materials.

c. The licensee attempted to recover the material by driving to the landfill, prior to being informed that the material had been buried. The licensee does not have any authority to take further actions to recover the material nor is recovery currently possible.

7. Part 20.2201(b)(vi) [procedures or measures that have been, or will be, adopted to ensure against a recurrence of the loss or theft of licensed material]:

a. Investigation by the licensee verified that current policies and procedures regarding waste control are adequate and should have prevented such an incident. At some point, therefore, procedures were apparently not followed. Lab workers can not recall any breach of procedures. Similar waste has been produced many times in the past without incident, and the ergonomics of the laboratory do not point to any probable mis-step. Therefore, remediation efforts will be broader than a focus on handling of the specific type of waste in question.

b. The corrective actions will involve screening of all waste generated from areas where external radiation detectable by NaI instruments is produced. All laboratories possessing photon-producing isotopes of energies greater than approximately 20 KeV, in quantity greater than 1 micro-curie, will be required to screen all sanitary and DIS waste with a NaI detector prior to release into the environment. The new standards will require that equipment, procedures, and competency of screeners be reviewed and approved by the RSO for each area. The level of detail in the existing procedure for screening of DIS waste has been increased to include removing interposed shielding, measuring each container at multiple positions, using the most sensitive scale, and a standard for the background reading. This more detailed procedure will be applied to all labs utilizing DIS, and will also be applied to screening of sanitary waste by those labs required to do so. The procedure will be reviewed by the Radiation Safety Committee on May 2, 2002. Users will be required to implement the new procedure by June 1, but will probably do so earlier. NaI detectors have been ordered; first deliveries are expected by the May 1. Until the equipment arrives and the program is fully implemented, routine trash pickups of affected areas has been halted. All waste is screened by the RSO, using the limited NaI technology currently available, before release.

c. The requirement for sanitary waste screening will become a condition during both permit approval and renewal, and whenever an affected isotope is added to the authorization. This condition will apply equally in Nuclear Medicine and Research laboratories at the St. Louis site of use. RSO review of equipment specifications and operator competency will be included in the approval process. This program has already been implemented.

d. Certain logistics of the waste-handling process, such as: how long waste is accumulated before screening, whether waste storage is centralized, arid how unscreened waste is stored, etc., can only be developed through practice. However, once best practices are developed, affected labs will be required to adopt diem, and will not be allowed to change them without review and approval of the RSO. Logistics will be under review through May, 2002; final practices will be in place by June 1, 2002.

e. DIS procedures and compliance will continue to be monitored during routine inspections and the results reported to the Radiation Safety Committee.

PRESSURIZER SAFETY VALVE SETPOINT DRIFT AT POINT BEACH UNIT 2

"On April 24, 2001, a vendor conducting offsite testing of the Point Beach Nuclear Plant (PBNP) Unit 2 RCS pressurizer safely valves' reported that safety valve 2RC-435 failed to lift at a test pressure of 2660 psig. The lift pressure specification for this valve is from 2440 to 2551 psig. In accordance with code requirement, the second in-service Unit 2 pressurizer safety valve was sent to the vendor for set point testing. The vendor also initiated an investigation to determine why the first valve failed to lift.

"On April 29, 2002, our engineering department received a telecopy report of that investigation. After evaluation, we determined this event to be reportable at 1129, 4-30-02. The valve at position 2RC-435 was last tested in November 2000 by the same vendor. At that time the valve set point was determined to be within specification. In accordance with the vendor's normal practice, a jack and lap procedure was done following the set point test to lap the valve disc and nozzle to insure that the valve is leaktight with undamaged seats. It was during this process that the valve became incapable of lifting at the specified pressure. The valve assembly was corrected and the set point retested with satisfactory results. The second safety valve that was shipped for testing subsequent to the failure of the first valve has also been tested and was found to lift within the specified range. Our investigation of the circumstances of the event is continuing.

"The required design capacity for the RCS pressurizer safety valves assumes the use of 2 valves based on RCS pressure not exceeding the maximum code allowable 110% of design pressure for the maximum calculated in surge of reactor coolant into the pressurizer. We have concluded that we operated PBNP Unit 2 with one inoperable pressurizer safety valve for the past cycle. The second Unit 2 safety valve has been tested and proven to be operable. This event is reportable both as a condition prohibited by the Technical Specification (LCO 3.4.10) and as an event or condition that resulted in an unanalyzed condition that had the potential to significantly degrade plant safety. An evaluation of the significance of this condition will be performed."

The NRC Resident Inspector has been notified.

UNUSUAL EVENT AT ANO2 DUE TO BORON RESIDUE AT PRESSURIZER HEATER SLEEVE N-2

"On 4/30/02, while performing a planned Reactor Coolant System Integrity Inspection following the 2R15 refueling outage, with the Reactor Coolant System at normal operating temperature and pressure (Mode 3), an inspector discovered small traces of boric acid on pressurizer heater sleeve N-2. Based on the subsequent evaluation, pressurizer heater sleeve N-2 has been determined to have had pressure boundary leakage. No current leakage is detectable. The Unit 2 Tech Spec 3.4.6.2 does not allow operation in Modes 1-4 with any pressure boundary leakage. Therefore a plant cooldown is being initiated to place Unit 2 in Mode 5. Based on this criteria, the condition is reportable upon declaring a Notification of Unusual Event (EAL 2.1) and Degraded Condition which is reportable per 1OCFR50.72(b)(3)(ii)(A)".

The NRC Resident Inspector was notified.


* * * UPDATE ON 5/1/02 @ 0331 BY AHO TO GOULD * * *

NOUE terminated at 0221CDT. Plant in mode 5 (cold shutdown)

The NRC Resident Inspector will be notified.

Notified FEMA(Zapata), REG 1 RDO(Spitzberg), EO(Black) and IRO(McGinty)