ML20027B318
| ML20027B318 | |
| Person / Time | |
|---|---|
| Site: | 07001007 |
| Issue date: | 08/23/1982 |
| From: | Book H, Brock B, Pang J, Thomas R, Zurakowski P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20027B310 | List: |
| References | |
| 70-1007-82-02, 70-1007-82-2, NUDOCS 8209200269 | |
| Download: ML20027B318 (6) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION V Report No.
70-1007/82-02 License No. - SNM-54 Priority II Category A
Licensee:
General Electric Company 175 Curtner Avenue San Jose, California 95125 Facility Name:
same Inspection at:
Building J Inspection conducted:
August 5,1982 Inspectors:
I A
Buddy Brick, Fuel Facilities Inspector Date Signed NW Of&
Paul Zurakowski, Radiation Specialist
/Dat4 Signed T. %... ~Pw 2 u 5y J. Frank Pang, RaMation Specialist Date Signed
~ #4/
Approved by:
R. D. Thomas, Chief, Materials Radiation Protection
/ Date Signed Section W
A*
W a
H. E. Book, Chief, Radiological Safety Branch Date Signed Summary:
Inspection of August 5,1982 (Report No. 70-1007/82-02)
The licensee requested NRC to conduct a confirmatory survey of his facility to establish that the buildings can be released for unrestricted use prior to the transfer of the operations to the state license and the termination of NRC License SNM-54.
On August 5, 1982, three Region V inspectors began the confirmatory survey in Building J of the facility.
Based on the results of this survey, it was det. ermined that the facility care not be released for unrestricted use and that additional decontamination will be required in one area of Building J.
A total of fifteen inspector hours were expended in conducting this survey.
8209200269 820901
{DRADOCK 0700100J RV Form 219(1)
l Summary (cont.) Results This survey, conducted with Micro-R portable gamma scintillation survey instruments, air proportional alpha survey instruments, and contamination wipes, found the contamination level in one area of Building J in excess of the contamination limitations specified in the NRC guidelines. Due to the contamination still present inside Building J, the NRC survey was discontinued until the licensee resurveys the facility, performs the necessary decontamination, conducts the final survey, and sutratts the survey results for evaluation by the NRC.
DETAILS A.
Persons Contacted Alan H. Lane, Nuclear Safety Engineer B.
Background
1.
In correspondence to the NRC dated April 14, 1982, the licensee stated the intention to terminate NRC license SNM-54 because of the elimination of reactnr fuels activities involving significant quantities of SNM (enriched uranium) and to transfer the remaining activities performed under that license to the State of California license 0418-59.
The licensee stated in correspondence that both Buildings H and J, where licensed activities had been conducted, have been decontam-inated to an unconditional release status except for:
(1) the liquid radwaste treatment annex, (2) the underground line between Building J and the radwaste treatment facility, and (3) one subfloor pipe in Building H.
2.
On August 5,1982, three Region V inspectors, expending a total of 15 inspector hours, conducted a confirmatory survey of the licensed facilities. The survey criteria was based on the NRC requirements established by the " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Ternination of Licenses for Byproduct, Source, or Special Nuclear Material" dated July 1982 (Table I).
3.
The field radiation detection instruments selected to perform this survey were:
a.
LLL Alpha Air Proportional Count Rate Meter Serial No. A-90 due for calibration September 10, 1982 with an efficiency of 49% for plutonium alphas.*
b.
LLL Alpha Air Proportional Count Rate Meter Serial No. A-234 due for calibration December 10, 1982 with an efficiency of 41% for plutonium alphas.*
c.
Eberline PRM-7 Micro-R Meter, Serial No. 247 due for calibration March 31, 1983.
d.
Eberline PRM-7 Micro-R Meter, Serial No. 453 due for calibration March 22, 1983.
- The efficiency was obtained using a Pu-239 Source, Serial No. 7685.
4.
The USNRC Region V NMC Model PC-55 gas flow proportional counter was used to count wipes for removable contamination.
An efficiency of 27% was used to determine the activity on the
. wipes. This counter efficiency was determined from the Pu-239 Source, Serial No. 7685.
Buildings H and J had been used for reactor fuel development and fabrication activities under License SNM-54 involving the use of enriched uranium. Some neutron sources containing plutonium had also been used at the facility but leak tests conducted by the licensee have not indicated any leakage.
The acceptable contamination levels established by the NRC guidelines for U-238 and U-235 in the absence of the more hazardous radionuclides are summarized in Table I.
Table I Acceptable Levels Average Maximam Removable 2
2 5,000 dpra/100 cm 15,000 dpm/100 cm 1,000 dpm/100 cm C.
Survey Results The survey was initiated in and limited to the former UF conversion 6
area in Building J using Micro-R scintillation and alpha air proportional count rate meters. The Micro-R meters were used to locate areas of potential contamination and the alpha air propor-tional count rate meters were used to evaluate the levels of contamination for compliance with the NRC guidelines. Areas surveyed with the alpha survey meters included all visible exposed floor surfaces where the asphalt tiles had been removed as well as several areas of potential contamination underneath the tiles which were located with the Micro-R meters. The ceiling girders and ceiling light fixture above the elevated platform in the room were also surveyed. Swipes for removable contamination were collected and analyzed on the USNRC Region V NMC Model PC-55 gas flow propor-tional counter.
Survey results are given in Table II.
With the exception of one measurement, all of the survey results were within the NRC guideline limits.
The exception was a small area underneath an asphalt tile in the 2
north end of the scrap recovery area which measured 48,260 dpm/100 cm with the alpha survey meter. A yellowish powder was observed to be embedded in the tile cement, which Mr. Lane identified as am[ponium 1
- diuranate (ADU). A wipe of this area measured 784 dpm/100cmd, which was within the NRC guideline limits.
D.
Conclusion 2
One small area of about 100 cm underneath an asphalt tile in the UF6 conversion area of Building J was found to have fixed contamination exceeding the NRC guideline limits for the maximum permissible l evel. The contamination was visibly evident as a yellow powder and was identified by Mr. Lane as ammonium diuranate.
Because of this finding, the survey which was limited to the UFS conversion area was discontinued until the licensee performs the necessary decontamination, surveys the area, and submits the results to the NRC regional office for evaluation. The confirmatory survey will be continued by NRC after the actions by the licensee have been completed.
Mr. Lane agreed with the inspectors findings.
It was suggested to Mr. Lane that a portable gamma scintillation survey meter be used to conduct a general survey of the areas to locate possible con-tamination.
Mr. Lane stated that the items of exceptions mentioned in paragraph B.I.
of the Background Section of this report will also be removed and di.sposed of as radioactive waste as part of the decontamination efforts to be carried out.
r
I Table II Survey Results 2
2 Radiation Fixed Contamination, dpm/100 cm Removable Contamination, dpm/100 cm Location pr/hr Average Maximum UF conversion area Note 1 Note 2 3
Bare concrete floor 37" x 26" 1810 14480 92 9" x 27" 230 420 Door sill to 1640 7240 loading room No. end of scrap recovery area (tile removed) 50 48260 784 Acid dissolution tank area (tiles removed) 20-50 240 240 g
Ceiling light fixture above platform in UF conversion area BG BG 1
6 Beam No. I above platform in UF conversionarek BG BG 0
Beam No. 2 above platform in UF conversionarek BG BG 0
Beam No. 3 above platform in UF6 conversion area BG BG 1
Under desk in Q.C.
lab 15 Note 1 - Based on an efficiency of 41% obtained from counting a Pu-239 source.
Note 2 - Based on an efficiency of 27% obtained from counting a Pu-239 source.