IR 05000187/1985002
| ML20141E903 | |
| Person / Time | |
|---|---|
| Site: | 05000187 |
| Issue date: | 04/04/1986 |
| From: | Brown G, Cillis M, Fish R, Garcia E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20141E872 | List: |
| References | |
| 50-187-85-02, 50-187-85-2, NUDOCS 8604220418 | |
| Download: ML20141E903 (19) | |
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I U. S. NUCLEAR REGULATORY COMMISSION
REGION V
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Report No.
50-187/85-02 Docket No.
50-187 License No.
R-90 Licensee:
Northrop Research and Technology Center Northrop Corporation One Research Park Palos Verdes, California 90274 Facility Name:
Northrop Triga Reactor Inspection at:
Hawthorne, California Inspection conducted:
December 30, 1985 through January 24, 1986 Inspector:
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/5 E. M. Garcia, Emergency Response Coordinator Da'te Signed k.
Y YSC h M. 2 h 6 46 111s, Radiation Specialist Date Si ned G. A. Brown. Emergency Preparedness Analyst Date Signed Approved By:
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R. F. Fish, Chief Date Signed Emergency Preparedness Section Summary:
Inspection on December 30, 1985 through January 24, 1986 including onsite inspection on January 7 through 10, 1986 (Report No. 50-187/85-02)
Areas Inspected: Special announced team inspection by regionally based inspectors of the Northrop Triga Reactor Facility, Building 3-48 (See Figures 2 and 3).
This inspection was to verify that the release criteria, as established in the June 24, 1985 Order Authorizing Dismantling of Facility and Disposition of Component Parts from the Office of Nuclear Reactor Regulation and presented in the licensee's Decommissioning Plan of January 10, 1985, had been satisfied. The inspection included review of records, interviews with personnel and independent measurements. The inspectors also followed up on allegation RV-85-A-0055. The guidance provided in Inspection Procedure 83890 was utilized.
8604220418 860404 PDR ADOCK 05000187 G
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Results: 'The inspection confirmed the findings described in the' licensee's
"Triga Reactor Facility Decommissioning Program Final Report, December 1985,"
with the exception that eight sealed sources remained in the possession of the licensee. Nuclear fuel has been transferred to other licensees, and the
" Unrestricted Use" release criteria as defined in section 4.4 of the " Safety Evaluation Supporting the Order Authorizing Dismantling of Facility and Disposition of Component Parts" appears to have been satisfied. The inspection also identified that there remains residual activation products in concrete structures. Resolution on the proper disposition of this material must be obtain from the Office of Nuclear Reactor Regulation. No violations or deviations were identified.
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DETAIL.j 1.
Persons Contacted
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Northrop Corporation
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J. Benveniste, Chief Scientist -- Chairman Corporate Radiation Committee H. Woo, Project Manager G. Cozens, Radiation Safety Officer J. Woods, Health Physicist Consultants to Northrop W. Crandall, Private Consultant J. Chalmers, Private Consultant F. Gardner, Project Manager, Chem-Nuclear Corporation (Prime Contractor)
State of California, Department of Occupational Health and Safety K. Wong, Sr. Health Physicist
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L. Carter, Health Physicist Los Angeles County Sanitation District (LACSD)
E. Esfandi, Project Engineer (contacted by telephone)
All individuals, with the exception of E. Exfandi, were present at the exit interview.
2.
Release Criteria for Unrestricted Use The radiation levels for release of the reactor facility to unrestricted use were established in Section 4.4 of the " Safety Evaluation Supporting Order Authorizing Dismantling of Facility and Disposition of Component Parts," dated June 24, 1985. The safety evaluation references Table 1 of Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," for residual radioactive contamination levels and the staff position of 5 microroentgen per hour (pr/hr) above background at one meter from any surface. The safety evaluation also stated that "if it can be shown that the maximum radiation exposure to an individual would be less than..
10 mR per year, considering potential occupancy in the vicinity of the radiation, then levels greater than 5 pr/br would be acceptable."
3.
Disposition of Materials The reactor has been dismantled and the fuel shipped offsite. Records maintained by the licensee indicate that radioactively contaminated components have been transferred to an agreement state licensed radioactive material disposal sit n
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Disposition of the fuel elements was as follows:
Number of Transfer Activity Transferred to License No.
Elements Date (Curies)
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Apr 11, 1985 54.0 University of Texas R-92
Apr 16, 1985 52.2 University of Texas R-92
Apr 18, 1985 32.4 Kansas State University R-88
Apr 22, 1985 19.8 University of Illinois R-117 A review of copies of the recipients licenses, possessed by Northrop, confirmed that each recipient was authorized to receive the transferred fuel. Records of the transfer of the material indicate that the material was properly packaged, radiologically surveyed and transported, thus fulfilling the conditions and limitations specified in 49 CFR 177.421.
Receipt of fuel was acknowledged by each recipient and verified by letter to Northrop. At the time of the inspection, the following eight sealed radioactive sources remained at the licensee's facility:
Nuclide Quantity Units Description Serial No.
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Ci Start up Neutros None U-235 1.5431 g
Fission Chamber G207 U-235 0.034 g
Fission Chamber C215 Pu-239 0.000416 pCi Eberline Alpha Ca.libration 8798 Pu-239 0.0034 pCi Eberline Alpha Calibration 8799 Pu-239 0.0352 pCi Eberline Alpha Calibration 8800 Pu-239 0.265 pCi Eberline Alpha Calibration 8801 Pu-239 0.002 pCi Alpha Calibration P6044
The licensee was informed that prior to the license being terminated the J
sealed sources discussed in this paragraph must be transferred to another licensee or another license not being terminated. Northrop informed the inspectors of their intention to transfer these sources to other licensees. This matter was discussed at the exit interview.
No violations were identified in this area.
4.
Records of Radiation Surveys and Reports The inspectors reviewed the licensee's "Triga Reactor Facility Decommissioning Program Final Report" submitted to the Commission on December 15, 1985, and the supporting records.available at the site.
This review indicated that the licensee made approximately 20,000 measurements representative of 2,880 square meters or almost 44% of the total surfaces in the facility. Based on these measurements the licensee concluded that the facility had been decontaminated and that the released criteria as described in Paragraph 2 above had been meet.
For fixed contamination measurements the licensee used Eberline E-120 and E-520 survey meters with HP-210 " pancake" Geiger-Mueller detectors and Eberline PAC-4G proportional Alpha counters. Removable contamination measurements (surface smears) were analyzed using a Tennelec LB 5100
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Series II automatic proportional counter. These instruments if properly calibrated and used can detect radiation levels below those established in the release criteria.
Licensee records indicated that beta-gamma and alpha surface contamination was well below the release crit'eria limits as established in Table 1 of Regulatory Guide 1.86.
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The inspector reviewed instrument calibration, testing and survey records. These records clearly identified the particular instrument used for each survey. All instruments used appear to have been properly calibrated with sources traceable to the National Bureau of Standards.
Recalibrations were performed in a timely manner. Operational checks were performed on each instrument prior to each use.
Four instruments, which f ailed to pass the operational checks, were removed f rom service.
For determining the residual gamma dose rate at a meter from any surface, the licensee used Ludlum Model 12S Micro-R-Meters. The licensee's final survey, with one exception, determined that the gamma dose rate criteria was met.
The one exception was the measurements made in the exposure room. The licensee found that due to the small dimensions of the room the detector was affected by the ceiling and other walls. The measured
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dose rate was 21.3 pr/hr.
In order to demonstrate achievement of the 5
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pr/hr criterion without resorting to a geometrical correction, a swath was cut through the concrete structure. This swath went from the hot cell at the north end through the exposure room and the south wall of the biological shield. With the concrete removed the gamma radiation level is comparable to the background levels. The removed concrete was collected in three piles (See Photographs 1 through 3).
At the time of the inspection the licensee had not made measurements of the gamma radiation levels from the resulting piles. At the time of the inspection the licensee intended to dispose of the concrete under State of California regulatory provisions similar to those in 10 CFR 20.302.
Agreement en the proper disposition of this material must be obtain from the Of fice of Nuclear Reactor Regulation. This matter was discussed at the exit interview.
Micro-R-Meters use a sodium iodide thallium drifted detector (NaI(TI)).
A copy of the manufacturer's energy response curve is included as Figure 1.
This instrument is basically an event counter and can not be used to precisely determine the true dose rate in environments of polyenergetic and variable gamma fields. Therefore, the measurements made by the licensee can be considered to provide only relative dose rates from one location to another. However, as noted in Paragraph 6, the independent measurements made by the inspectors using a true dose instrument did not identify any instance where the release criteria was exceeded. This matter was discussed at the exit interview.
A review of the licensee's personnel monitoring and exposure records showed that all individuals working on the decommissioning project were monitored by both pocket ion chambers and film badges. The collective dose to all workers for the entire dismantling operation was 0.630 man-REM. All exposures to individuals were less than the limits established by 10 CFR 20. The licensee's records appeared accurate and well maintained. They plan to preserve the records indefinitely in the Northrop archives.
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No violations or deviations were identified in this area.
5.
Followup on Allegation RV-85-A-0055
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ATS No.: RV-85-A-0055
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a.
Characterization Northrop Triga Reactor Facility discharged a large amount of licensed radioactive material into the sanitary sewers at the plant.
b.
Implied Significance to Safety If the licensee released licensed radioactive materials in excess of those permitted under the provisions of 10 CFR 20.303, the licensee would have violated the regulations and an environmental impact exceeding that analyzed by the regulations may have taken place.
c.
Assessment of Safety Significance
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A review of records maintained by the licensee, interviews with the.
licensee's staff and a telephone conversation with an official of Los Angeles County Sanitation District (LACSD) indicate that the licensee did release licensed radioactive materials to the sanitary sewer, but the releases did not exceed the limits established in 10 CRF 20.303.
The licensee released the content of the holdup tank (Facility Sump)
and the reactor pool during 1985. Prior approval for the disposal of these releases was obtained from LACSD and an inspector from that agency was present when releases were initiated.
LACSD collected samples on both occasions. The licensee had an independent laboratory analyze samples from the released water prior to each release. Review of the analyses results indicate that the concentrations and total quantities were below the levels permitted by 10 CFR 20.303.
d.
Staff Position The inspector concluded that the actions of the licensee were permitted by the provisions of 10 CFR 20.303, no violations were identified, and no unevaluated environmental impact resulted from these releases.
e.
Action Required None
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6.
Independent Confirmatory Measurements The inspectors conducted selected radiation surveys to independently confirm the results of the licensee measurements. These sdrveys
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consisted of wipes for detecting removable contamination, direct measurement of surfaces for nonremovable contamination, and dose rate measurements at one meter from the surfaces.
In conjunction with the dose rate measurements, gamma spectrums were collected to determined if the gamma emissions were due to naturally occurring nuclides or licensed radioactive material.
a.
Removable Contamination Survey A total of 264 surface wipe samples were collected from areas and objects in the facility. These samples were analyzed on the NRC's Tennelec LB-5100 thin window gas proportional counter system. The samples were analyzed for gross alpha and beta contamination. The instrument's efficiencies for Pu-239 and Tc-99 were used in determining the activity. The instrument was calibrated and sample background determined for each batch of samples analyzed. The
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batches ranged in size from 15 to 62 samples each. At the 95%
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r confidence level, contamination was less than 8 dpm/100 ca alpha, 2 beta. These values are well below the and less than 52 dpm/100 cm limits for removable contamination in Table 1 of Regulatory Guide 1.86.
b.
Alpha and Beta Gamma Nonremovable (Fixed) Contamination Survey The inspectors surveyed the facility for fixed alpha, beta and gamma contamination.
Inspector selected areas of floors, walls, and major components were surveyed with the instruments described below.
Surveys with the Micro-R-Meter (PRM-7) were made throughout the facility and in the immediate environs adjacent to the facility.
The PRM-7 was used only to identify areas of relatively higher dose rates. A more detailed survey, as described in Paragraph 6c, was conducted in those areas. The NRC instruments used were:
Calibration Radiation Model Detector NRC No.
Date Type LRL Alpha Proportional 000374 Nov 18 1985 Alpha LRL Alpha Proportional 000375 Aug 20 1985 Alpha Ludlum 12 Proportional 003564 Sep 04 1985 Alpha Eberline E-520 HP-260 006385 Dec 06 1985 Beta & Gamma Eberline E-520 HP-260 007907 Dec 06 1985 Beta & Gamma Eberline E-520 HP-260 010965 Dec 06 1985 Beta & Gamma Eberline PRM-7 NaI(TI)
006383 Dec 06 1985 Gamma Eberline PRM-7 HaI(TI)
008596 Dec 06 1985 Gamma Eberline PRM-7 NaI(Tl)
010839 Sep 27 1985 Gamma The alpha and beta measurements did not identify any residual activity above backgroun '.
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c.
Residual Gamma Emitters Survey To independently confirm that the unrestricted use criterion of 5 pr/hr at a meter from any surface was met, the inspectors conducted measurements in selected locations. These locations were selected based on the surveys with the PRM-7 and where higher activation products would be expected from the operation of the facility. The measurements consisted of precise gamma dose rate determinations with a Reuter Stokes Environmental Pressurized Ion Chamber, Model RSS-Ill, serial number Z-3999 (see photograph 4).
This instrument was calibrated on July 19, 1985. To determine " background",
measurements were also performed at two locations of similar architecture and construction history but cutside the restricted area. A total of 17 precise dose rates measurements were made.
Figures 2 and 3 identify the locations of the measurements, and the results are summarized below:
NO.
LOCATION GROSS RATE NET RATE pr/br pr/hr 1.
Near Beam Port 13.8 4.5
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Near Beam Port 13.3 4.0 3.
Exposure Room Remains 13.8 4.5 4.
Exposure Room Remains 13.3 4.0 5.
Plug Door Remains 12.7 3.4 6.
North Reactor Outer Wall 11.4 2.1 7.
North of Exposure Room 11.4 2.1 8.
Near Truck Door 10.0 0.7
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East Reactor Outer Wall 9.8 0.5 10.
South Reactor Outer Wall 12.2 2.9 11.
Bottom of Sump 12.7 3.4 12.
Reactor Pool 14.0 4.7 13.
Reactor Pool 13.6 4.3 14.
Reactor Pool 13.8 4.5 15.
Upper Pool Deck 8.5 BKG.
16.
Building 3-55 (Background)
9.6 N. A.
17.
Building 3-10 (Background)
9.3 N. A.
To determined if the measured dose rates were due in part to the presence of residual licensed gamma emitters, spectrums were
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collected using a portable intrinsic germanium detector and a Nuclear Data ND Six portable multichannel analyzer (see photographs 5 and 6).
The spectrums were used for the qualitative determination of the nuclides present. An energy calibration was performed using National Bureau of Standards gamma standard number 4275-B. The collected data was later analyzed using a Nuclear Data ND-6620 computerized multichannel analyze,
These measurements identified the presence of naturally occurring and licensed radionuclides. The results are summarized below:
NO.
LOCATION NUCLIDES IDENTIFIED-LICENSED NATURAL 1.
Near Beam Port Mn-54, Co-60, K-40, Pb-212, Eu-152 Pb-214 2.
Exposure Room Remains Co-60, Eu-152 K-40, T1-208, Pb-212, Pb-214 Bi-214 3.
Plug Door Remains Co-60, Eu-152 K-40, T1-208, Pb-212, Pb-214 Bi-214 4.
Chemistry Storage Well Na-24 K-40, T1-208, Pb-212, Pb-214 Bi-214, Ac-228 5.
Debris Pile None K-40, T1-208, Pb-212, Pb-214 Bi-214, Ac-228 6.
Pool Bottom Eu-152 K-40, T1-208,
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Pb-212, Pb-214
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Bi-214 7.
Building 3-10 (Background) None K-40, T1-208, Pb-212, Pb-214 Bi-214, Ac-228 8.
Building 3-55 (Background) None K-40, T1-208, Pb-212, Pb-214 Bi-214, Ac-228 The licensee stated that after the license is terminated the remaining standing structures will be demolished, and all concrete debris will be disposed.
7.
Conclusions c
With the exception of the eight sealed sources, discussed in Paragraph 3, and the residual activation products in the concrete all licensed materials have been properly disposed of.
The evaluations and measurements made by the licensee and the NRC indicate that the criteria in Table 1 of Regulatory Guide 1.86 have been satisfied. The criterion for gamma radiation of 5 pr/hr above background at a meter from any surface also appears to have been satisfied. Agreement on the proper disposition of the residual activation products in concrete structures must be obtained from the Office of Nuclear Reactor Regulation.
8.
Exit Interview At the conclusion of the inspection the inspectors met with the individuals denoted in Paragraph 1.
The preliminary findings of the inspection were presented. The licensee was informed that, prior to the license being terminated, the sealed sources discussed in Paragraph 3 must be transferred to another licensee or another license not being terminated, and agreement on the proper disposition of the residual
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activation products in the concrete pile generated from the demolition of the exposure room must be obtained from the Office of Nuclear Reactor Regulation. The licensee was informed that no violations or deviations had been identified.
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Subsequent to the inspection the licensee submitted NRC Form 314,
" Certificate of Disposition of Materials," documenting the transfer of the sealed sources to other licensees.
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