ML20247C794
| ML20247C794 | |
| Person / Time | |
|---|---|
| Site: | 07000734 |
| Issue date: | 04/20/1998 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20247C764 | List: |
| References | |
| TAC-L30914, NUDOCS 9805140079 | |
| Download: ML20247C794 (19) | |
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DOCKET:
70-0734 LICENSE:
SUBJECT:
ENVIRONMENTAL ASSESSMENT OF IMPLEMENTATION OF THE GENERAL ATOMICS SITE DECOMMISSIONING PLAN (TAC NO.
l L30914)
1.0 INTRODUCTION
General Atomics (GA) has been authorized by the U.S. Nuclear Regulatory Commission (NRC) and its predecessor, the Atomic Energy Commission, to use special nuclear material in nuclear fuel fabrication and research and development for more than 30 years at its site in San Diego, Califomia. As operations changed at its site, GA initiated decommissioning activities affecting portions of the site beginning in the mid 1980s. By the early 1990s, fuel fabrication and operations involving special nuclear material at the facility had ceased, and in September of 1996, GA's Special Nuclear Material License, SNM-696, was amended to authorize only activities incident to decommissioning. GA also currently has State of California Radioactive Materials License No. 0145-37 to possess and use source and byproduct materials and NRC Reactor Licenses, R-38 and R-67, for two Training Reactor-Isotope-General Atomic (TRIGA) research reactors.
By application dated October 11,1996, and supplements dated December 5,1996; April 18, 1997; and January 15,1998; GA requested an amendment to SNM-696 to incorporate an overall Site Decommissioning Plan (DP). The NRC has prepared this environmental assessment pursuant to the Council of Environmental Quality regulations (40 CFR Parts 1500-1508) and NRC regulations (10 CFR Part 51), which implement the requirements of the National
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Environmental Policy Act of 1969, as amended. The purpose of this document is to assess the environmental consequences of approving and implementing the proposed site DP.
2.0 SITE HISTORY GA's Special Nuclear Material License was last renewed on January 23,1985. The licensee was authorized to possess uranium, enriched in uranium-233 and uranium-235, and plutonium to conduct research and development and to fabricate High Temperature Gas-Cooled Reactor (HTGR) and TRIGA fuel assemblies. The HTGR activities included the application of research on high-temperature materials, design and fabrication of reactor system components, and the development of new processes and process modifications related to the manufacture of the HTGR fuel assemblies. The TRIGA activities included the design, development, fabrication, and installation of research reactors and their fuel elements. In addition, research and development activities included laboratory-scale physical, metallurgical, chemical, and engineering investigations utilizing special nuclear material, pilot-scale process development, and Hot Cell operations involving irradiated fuel elements.
GA began partial decommissioning of the facility in the early to mid 1980s when the scope of activities involving special nuclear materials declined. At that time, HTGR fuel production ceased, TRIGA fuel production drastically decreased, and post irradiation examination of fuel elements in the Hot Cell ended. All fuel fabrication operations involving special nuclear material 9905140079 900420 PDR ADOCK 07000734 C
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ended by the early 1990s and, since that time, GA has been engaged only in decommissioning activities. These decommissioning activities are summarized in Section 5.1.
3.0 PROPOSED ACTION AND ALTERNATIVES 4
The proposed action is the amendment of GA s license to incorporate the site DP, which 4
describes the remaining decommissioning activities planned at the GA facility under License SNM-696 and subsequent release of the site for unrestricted use. GA intends to decommission to radiation levels required to release the site for unrestricted use and to terminate License SNM-696 for these areas. Soil will be remediated to levels specified in Option 1 of the Branch Technical Position (BTP), " Disposal or Onsite Storage of Thorium or Uranium Wastes from Past Operations," [46 FR 52061, October 23,1981]. Facilities and equipment will be decontaminated to levels specified in " Guidelines for Decontamination of Facilities and Equipment Prior to
- Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material," (USNRC, Policy and Guidance Directive FC 83-23, Division of industrial and Medical Nuclear Safety, November 4,1983).
The following subsections describe the areas and facilities to be decommissioned, the decontamination techniques, and the proposed effluent control and waste management practices that will be used during decommissioning Alternatives to the proposed action are then summarized.
3.1 Areas / Facilities to be Decommissioned The GA site DP has identified and described in detail the buildings and support areas that will require partial to complete decommissioning / decontamination / dismardlement (D&D). These areas are shown on a site map in Figure 1 and include:
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Building 2 - Science Laboratory Building, Laboratory 307, Underground Storage Tank.
The tank was used for temporary storage of radioactively contaminated liquids containing, primarily, Cs-137, Co-60, and Sr-90. The tank was removed and the soil remediated in 1984. However, the area was not decontaminated to release criteria described in Policy and Guidance Directive FC 83-23.
I Building 2 - Science Laboratory Building, Laboratory Sections A-C.
This building was used for research and development, testing, and experimental work with radioactive materials under both the State and NRC Special Nuclear Material Licenses. The building is currently inactive except for the Health Physics Laboratory, which is being moved to Building 10. To date, approximately 109 labs in this building have been released for unrestricted use and 33 labs are yet to be released.
Building 2 - Science Laboratory Building, Service Core and Drain Lines.
This refers to utility services to the laboratories plus drain lines from some of the I
laboratories. All known radioactively contaminated ducts in the core were rerroved in 1994. All other accessible ducts were surveyed and no radioactive contamination was l
found. However, two drain lines and one concrete area are known to be contaminated.
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3 2.
Building 9, Room 040.
This room was useo for TRIGA fuel fabrication prior to 197S. All other areas of Building 9 have been previously remediated. This room is currently being used for telecommunication equipment storage. The original floor, which may be potentially contaminated, is located about 8 inches below th current raised wooden floor.
3.
Building 22 - TRIGA Fuel Fabrication Facility.
Operations in this building have ceased and all equipment has been removed. Special nuclear material and depleted uranium are expected contaminants. Soil contamination is expected to be present.
4.
Building 25 - Liquid Waste Treatment Facility.
This facility is currently in use for treatment of liquid low-level radioactive waste for discharge to the sewer. All radionuclides used by GA under both the State and NRC licenses are expected to be present here. Soil contamination is possible.
5.
Building 41 - Nuclear Waste Processing Facility.
This facility is currently used for baling, compacting, solidification, drying, and repackaging. All radionuclides used by GA under both the State and NRC licenses are expected to be present here. Se contamination is possible.
u 6.
Building 27 - Experimental Area.
This area included radiochemistry and chemistry laboratories and was shut down in September 1996. All radionuclides used by GA under both the State and NRC licenses are expected to be present here. Soil contamination is possible.
7.
Building 27-1, EA-1 Bunker Facility.
This area was used for research and development associated with radiochemistry. The equipment has been removed and the area is ready for further remediation. All radionuclides used by GA under both the State and NRC licenses are expected to be present here. Soil contamination is possible, 8.
Building 30, Room 118 - Linac (Linear Accelerator) Facility.
Irradiated fuelis stored in casks in this room. GA intends to remove the fuel and transfer authority of the area to the State for continued use involving radioactive materials. No decontamination of the facility is expected to be necessary.
9.
Building 31, Room 103.
This room is currently used for storage, sampling, and combining of radioactive material.
All radionuclides used by GA under both the State and NRC licenses are expected to be present here.
10.
Building 37, "SVA-South."
This area was used for temporary storage of packaged low-level waste from decommissioning of the northern portion of Building 37 and sealed sources used in the calibration of TRIGA monitoring / control equipment. Currently no radioactive materials
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4 are used or stored in this facility. No radioactive contamination is exoected to be present here.
11.
Building 39, "SVBIPilot Plant."
This area was used for research and development of pilot plant fuel. Special nuclear material, depleted uranium, and thorium were all used in this area. Operations have ceased and equipment removal and characterization are in progress. Soil contamination is possible.
Although the site DP discusses decommissioning of the Hot Cell, the decommissioning plan for this area has been previously approved by NRC in license amendment dated May 1,1996, and l
is not considered in the proposed licensing action. In addition, the site DP listed the TRIGA l
reactor facility and various GA facilities, which are not regulated under License SNM-696.
Decommissioning of these areas will not be considered in the proposed licensing action.
Decommissioning of these areas must be conducted under the appropriate NRC Reactor or State Licenses. These activities were included in the site DP by GA to ensure development of a comprehensive plan for decommissioning the San Diego site.
3.2 Decommissioning Methods Because the level of contamination and the resulting levels of decontamination and decommissioning vary significantly for each area, GA has divided the D&D activities into three approaches-Approach A: Decontamination requires the removal of sealed sources. In these cases, no contamination is likely from the radioactive sources. GA will conduct a scoping survey and if no contamination is found, no additional decontamination will be
. required and the results will be documented in the final survey. This approach is expected to be appropriate for items 8 and 10, above.
Approach B: Removal of radioactive material and contaminated equipment is required.
Decontamination may include cleaning or removal of contaminated hoods, ducts and plenums. In addition, a limited number of smalllabs and rooms may require aggressive decontamination (scabbling, stripping, or otherwise abrading the floors, walls or ceiling surfaces). The areas requiring aggressive decontamination are localized and limited to a few small labs and/or rooms. The final radiation survey will confirm that the facilities meet the NRC release criteria. This approach is expected to be appropriate for items 2,6, and 9, above.
1 Approach C: Facility decontamination is widespread, and major decontamination /
dismantlement will be required. This effort consists of removal of contaminated material and equipment; decontamination of hoods, duct work, and plenums; and aggressive decontamination of surfaces. Some exhumation and removal of contaminated soil may also be required. Measures to control airborne radioactivity and air monitoring may be required to support the decontamination.
This is expected to be necessary for items 1, 3, 4, 5, 7, and 11, above.
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To support decommissioning of the facilities requiring Approach C decontamination, characterization radiological surveys will be conducted to determine the extent, volume and nature of the contamination. The characterization will also identify any pathways by which contamination may have migrated from the facility.
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GA will conduct decontamination of the areas and facilities described above by using simple and more passive decontamination methods first, such as vacuuming, damp cloth wiping, and to a limited degree washing / scrubbing operations. If these techniques are unsuccessful, GA will advance to more aggressive decontamination operations, such as abrasive scabbling.
Aggressive decontamination methods will be required to remove existing surface coatings such as paints, varnishes, and similar fixatives because of the potential for embedded contamination.
GA may use the following decontamination methods and equipment, which have been used in previous decommissioning operations at the facility:
Jack-hammering This may be used to remove contaminated concrete or asphalt from a surface.
Scabbling This technique may be used to clean surfaces. It uses tools having single or multiple bit piston heads, equipped with multi-point tungsten carbide bits. The pneumatically operated tool drives the bits against the surface, I
which causes the surface to abrade.
Plasma Arc Cutting _ Large metal pieces, including process equipment, may be cut with this equipment in preparation for shredding or as a volume reduction technique.
Flame Cutting Oxy-acetylene cutting systems may be used for the dismantlement and volume reduction of metal pieces.
Mechanical Cutting When modification or removal of structural material or equipment is required, it may be accomplished by using powered equipment, such as a power blade covering remover or a sectioning saw.
Core Drilling Floor sudaces throughout some facilities contain female and male anchor bolts where contamination may be present. The anchor bolts may be removed from the floors by core drilling.
Blastrac Machine This device may be used for cleaning concrete floors. The delivery L
system consists of an enclosed centrifugal blast wheel inside a cleaning head. As the wheel spins, metallic shot is hurled to the floor surfaces being cleaned. Abrasive medium and contaminants are drawn into a
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separation system where contaminants are removed by an attached dust collector and abrasive medium is recycled. Travel speed and shot' size may be adjusted to control the depth of surface removal. Residual metal j
shot left on floor surfaces is recovered by using a magnetic broom.
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Vacublast Machine This may be used for cleaning vertical surfaces and along interfaces between walls and floors. It uses compressed air to convey abrasive medium from a pressure generator through a hose and blast nozzle to the surface being cleaned. The medium and debris is vacuum recovered at the point of contact. Air then transports the collected medium, dust and debris to a reclaimer where it is air washed. The medium is recycled through the equipment. Dust and debris are drawn into a secondary cyclone separator and deposited in a collection chamber. The depth of abrasion is controlled by adjusting the size of the shot and travel speed of j
the unit.
l Compactor The compactor compacts waste into boxes, obtaining a volume reduction and/or Shredder factor as high as 10:1, depending on the type of material to be compacted. GA currently has a compactor onsite.
The shredder may be used to shred a significant quantity of contaminated materials packaged on the decommissioning project. Typical materials processed through the shredder include electrical conduit and small diameter piping, desks, chairs, benches, metal walls studs, metal sheet and beams and small pieces of equipment. GA has used a shredder in past decommissioning operations but does not currently have one onsite.
l GA was authorized by license amendment dated September 14,1990, to l
use a shredder / compactor system Liquid Abrasive This unit would be primarily used to decontaminate steel. It is a trailer Decontamination mounted system which uses recyclable metal medium delivered in a j
System water stream to the surface being cleaned. The operation would be performed in a stainless steel chamber having high efficiency particulate air (HEPA) filtered ventilation installed. The medium and debris are collected in a sump, pumped back into the supply and cleanup system and reused. GA does not currently have this system onsite.
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3.3 Effluent Control and Waste Management Solid, liquid, and airborne effluents will be generated during the decommissioning activities.
Solid radioactive waste willinclude contaminated soils and dismantled structures that can not be decontaminated. Class A radioactive wastes, as defined in 10 CFR Part 61, will be shipped to an authorized low-level radioactive waste disposal facility, it is anticipated that no Class B or C waste will be generated as a result of decommissioning any of the facilities regulated under
- License SNM-696. Radioactive waste from decommissioning activities is currently being shipped to two Department of Energy low-level radioactive waste disposal facilities. One is located at the Nevada Test Site and the other is at Hanford, Washington.
The primary mixed waste (radioactive and hazardous) is anticipated to be contaminated lead based paints and contaminated asbestos bearing tile. Environmental Protection Agency and waste disposal site requirements for such waste will be met. Hazardous wastes will be disposed
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l of through authorized hazardous waste disposal firms. Nonradioactive /non-hazardous waste will be removed to locallandfills.
Prior to disposal, solid radioactive waste may be subjected to various volume reduction techniques including decontamination, segmentation, shredding and/or compacting. GA has committed to use contamination and ventilation controls to prevent the spread of contamination to surrounding areas while using these techniques.
Liquid waste generated during decontamination operations may contain relatively small amounts of radioactive and/or chemical contaminants. These liquids will be solidified or filtered prior to l
disposal. Solidification will be performed at GA's Nuclear Waste Processing Facility (Buildings 41 and 25). The solidification process will consist of taking approximately 30 gallon batches of aqueous waste and placing each batch into a 55 gallon drum where it can be neutralized with an acid or base, if required. The contents of the drum will then be mixed with cement and mica (a filler), as needed, to form a monolithic mass. The material will be disposed of as solid radioactive waste. Alternatively, some low-level aqueous wastes will be filtered. After verification that local, state and federal regulations have been met, including the requirements of 10 CFR 20. 2003, the treated water will be discharged through the sanitary sewer system and records of each discharge will be maintained by GA.
Adsorption of radioactively contaminated organic liquids such as pump oils, solvents, etc. will be i
performed with solidification agents acceptable to radioactive waste disposal sites. The resultant solid material will be packaged in 55 gallon drums. This waste will only be sent to authorized waste disposal sites.
Airborne radioactive effluents may also be generated during the decommissioning activities. No significant non-radioactive effluents are expected to be generated. Airborne radioactive material will be minimized by utilizing engineering controls including (1) ventilation devices, such as in-place or portable HEPA filters and/or facility ventilation systems that include HEPA filtration; (2) containment devices, such as designed containers, plastic bags, tents, and glove bags; and (3) source term reduction, such as application of fixatives and misting of surfaces to minimize i
dust and resuspension. When enclosures are used to contain and control airbome contamination, the exhaust will be filtered through HEPA filters as appropriate to the operation being performed. HEPA systems will be rated at 99.95% efficient for particulate of 0.3 micron size. All discharge points that have the potential to exceed 10% of the radionuclides concentration limit derived from 10 CFR 20, Appendix B, Table 2, will be monitored as described i
in GA'S effluent and environmental monitoring program in License SNM-696.
3.4 Alternatives to the Proposed Action The attemative to the proposed action is to deny approval of the site DP and require that GA cease on-going decommissioning activities. Because GA is currently authorized under License SNM-696 to conduct decommissioning activities, several decommissioning operations discussed in the site DP are underway at the facility. Therefore, NRC would be required to issue an order based on health and safety to prevent GA from conducting these operations
8 After denial of the site DP, NRC would require GA to maintain the site in its contaminated state under the NRC license until GA revised the decommissioning plan to reduce environmental impacts and this plan was reviewed and approved by the NRC. Alternatively, GA could request license termination for the site as it is and impose area or site restrictions to protect the public from residual radioactivity.
4.0 AFFECTED ENVIRONMENT The GA facility consists of two areas on the Torrey Pines mesa in San Diego, California and is shown in Figure 1. The Main site includes the research and development laboratories, the waste processing facilities, the TRIGA reactors, and the former fuel fabrication processing areas.
It is located in the center of Torrey Pines Mesa Science Center, a 1,229,984 m (304 acre) 2 industrial park. The Sorrento Valley site includes the previous HTGR fuel production area and is located about 161 m (0.1 mi) north of the main site. The GA facility is approximately 91.2 m (300 ft) above sea level,1609 m (1 mi) from the Pacific Ocean, and 20,917 m (13 mi) northwest of downtown San Diego.
Detailed descriptions of the environment surrounding the GA facility can be found in a 1983 Environmental Impact Appraisal, performed by the NRC to analyze impacts from renewal of the GA license, and in a 1995 Environmental Assessment, performed by the U.S. Department of Energy to analyze impacts from decommissioning of the GA Hot Cell Facility. These descriptions include the climate, meteorology, air quality, land use and soils, water, geology and seismicity, ecology, demography (based on the 1990 census), and background radiological statistics at the facility. Because this information has not changed significantly, it is not repeated here.
5.0 EFFLUENT RELEASES AND MONITORING PROGRAMS GA has been conducting decommissioning activities since the mid-1980s. A summary of past decommissioning activities is provided in the following section. The effluent and environmental monitoring program and the monitoring data are then discussed.
5.1 Historic Decommissioning Activities Waste Processina Area in mid-1984, GA began remediation of the old nuclear waste processing area at the facility. This project included decommissioning of (1) the Solar Evaporation Ponds, which were four sets of three 6.1 m x 6.1 m x 0.3 m (20 ft x 20 ft x 1 ft) concrete ponds designed to contain contaminated liquid for evaporation; (2) the former Radioactive Waste incinerator site (the incinerator itself was decommissioned and removed in 1980); (3) the Nuclear Waste Processing Building, which was used to process and package solid waste (contaminated with enriched uranium and thorium and small quantities of mixed fiss and activation products) and to process and solidify acid waste (contaminated with uranium and thorium); and (4) a previous burial site for contaminated asphalt. The e uject also included removal of high-level, shielded storage facilities and underground storage wells 12 feet deep, soil remediation of the hillside and
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9 canyon area below the waste handling areas, and a survey of undeveloped land to demonstrate that no radioactive contamination was present.
Remediation of these areas involved the excavation, packaging, and shipment of approximately 8
2300 m (82,000 ft ) of contaminated soil containing an estimated activity of 480 mci. The primary contaminants found were U-235, Th-232, Sr-90, Co-60 and Cs-137, with maximum soil concentrations of 26.9,301.9,400.0,455.9, and 746.7 pCilg, respectively. In addition sludge containing maximum concentrations of up to 80,400 pCi/g Cs-137,2,400 pCi/g Co-60, 4,500 pCi/g Cs-134,580 pCl/g Ce-144,658 pCi/g Eu-154, and 200 pCi/g Eu-155 was removed from the ponds.-
Decommissioning was completed in accordance with a Decommissioning Plan dated October 1, 1985, which was approved by the NRC by letter dated November 26,1985. By license amendment dated June 22,1988, these areas were released from License SNM-696 for unrestricted use along with approximately 215 acres of primarily undeveloped land. The NRC conducted confirmatory surveys of all areas prior to release.
1 Experimental Buildina (Buildina 9)
In October 1986, GA initiated decommissioning of the Experimental Building. Prior to 1975, TRIGA fuel fabrication activities were conducted in this building. Since that time, there were no activities in the building involving quantities of enriched uranium in excess of the amount allowed i
under the State License (350 g U-235). HTGR fuel treatment methodology studies and demonstrations were performed in the building from 1971 until 1985. The only significant quantities of nuclear materials used in the studies were depleted uranium and natural thorium.
Very small quantities (<10 mCl) of short-lived radioactive tracers (e.g.,1-131 and Zr-95) were also used.
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2 About 10% of the building, approximately 500 m (5400 ft ), was required to be decontaminated using cleaning and scabbling techniques to meet unrestricted use criteria. Contaminated soil beneath the concrete was also removed. The average contaminant concentrations in soil were 5.44 pCi/g of Th-228,6.18 pCi/g of Ra-228, and 7.92 pCi/g of U-238. The D&D effort resulted in the generation of 271 m'(9,585 ft ) of contaminated equipment, debris and soil, which was disposed of at a licensed low-level waste disposal facility. The decommissioning project was completed by July 1987.
One room containing telecommunications equipment has not been officially released. The original floor, which may be contaminated, is located about 8 inches below the current raised wooden floor on which GA's telecommunications equipment is located.
Areas B1. B2 and B3 By license amendment dated October 3,1988, areas B1 and B3 were released for unrestricted 2
use. Area B1 included 6.1x104 m (15 acres) of steep, rough hillside located generally west and north of GA's TRIGA Reactor Facility and behind the Hot Cell and other nuclear facilities on 2
GA's Main site. Area B3 included about 1.2x104 m (3 acres) of an abandoned city sewege treatment facility (also known as the " Callan Ponds"). Remediation included removal of 8
digestors and approximately 280 m (10,000 ft ) of contaminated soil containing enriched uranium.
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10 Area B2 was removed from the license by amendment dated May 20,1992. This area was a small sewage pump station near GA's Main site. Sludge contaminated with Cs-137 was removed and a small area of the concrete holding tank was decontaminated.
Science Laboratorv Buildina (Buildina 2)
Also beginning in the 1980's, GA began decommissioning of many individual laboratories and rooms within Building 2. Building 2 is the Science Laboratory Building. It contains laboratories, offices, shops, and caves for work with low level radioactivity. Most of the research activities in metallurgy, chemistry, and experimental physics involving radioactive material were conducted in this building. A total of nine groups of labs in this building, each containing between 9 and 25 labs, have been remediated and released for unrestricted use.
Experimental Criticality Facility (Buildina 31-2)
The process area of Building 31-2, the Experimental Criticality Facility, was released for unrestricted use in 1992. Decommissioning involved cleaning of concrete surfaces and the removal of soil contaminated with enriched uranium from beneath the concrete surface.
HTGR Nuclear Fuel Fabrication Facility (SVA)
From October 1990 to August 1995, remediation was conducted of the HTGR Nuclear Fuel l
Fabrication Facility located in the northem half of Building 37 (referred to as GA's Sorrento Valley "A" (SVA) facility). This facility was used for the manufacture of HTGR fuel for the Ft. St.
Vrain HTGR reactor using high enriched uranium (93% U-235) and thorium. Operations included " dry mix" kernal formation, pyrocarbon and silicon carbide coating, compact fabrication, assembly, uranium recovery using solvent extraction, and quality control laboratory activities.
Phase I of decommissioning occurred from 1990-1993 and included removal of equipment and 2
2 decontamination of approximately 4,000 m (45,000 ft ) of the facility. Phase ll of the decommissioning occurred from 1993 to 1995 and included building dismantlement, removal of underground drain lines, and soil remediation. Following an NRC confirmatory survey, the facility was released from License SNM-696 by amendment dated August 1,1995.
The primary contaminants of this facility included Th-228, Ra-228, U-238, and U-235. The maximum radionuclides concentrations in concrete samples from the facility prior to remediation L
were 567 pCi/g of Th-228,587 pCi/g of Ra-228,63 pCi/g of U-238, and 367 pCi/g of U-235.
l Most of the contamination existed on the surface, with only three locations where contamination L
was found more than 0.64 cm (0.25 inches) into the concrete. Smearable contamination levels 2
2 ranged up to approximately 13,000 dpm/100 cm alpha and 7,000 dpm/100 cm beta. In addition, some floor tile was also found to contain asbestos.
8 The final packaged radioactive waste volume tutaled 4,000 m (143,000 ft ) and included l
3,700 m (131,200 ft ) of Class A low-level radioactive waste,320 m (11,200 ft ) of 8
8 contaminated asbestos, and 17 m (600 ft ) of potentially mixed waste. In addition,5.5 x108 kg 8
8 (5,390 long tons) of non-contaminated waste was disposed of in a nearby sanitary landfiH.
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Hot Cell Facility By license amendment dated May 1,1996 and letter dated January 29,1997, the NRC apprcved GA's Decommissioning Plan for the Hot Cell Facility (HCF), which includes GA's 2
Building 23 and an outdoor service yard. The interior of Building 23 has approximately 690 m 2
(7,400 ft ) of floor space consisting of offices, three hot cells, an operating gallery and auxiliary 2
2 areas. The service yard is 4,340 m (46,740 ft ) and includes several concrete pads used for staging heavy equipment and making material transfer into and out of the HCF building. There l
nre also two above ground waste storage tanks. The hot cells and their associated equipment j
i have been used for examining irradiated capsules and small fuel elements, mechanical testing, metallograhic preparation and examination, and photography.
From characterization of the facility, the highest survey rending was found to be 10 e dpm/100 cm and 2,800 mR/hr beta. The principal contaminants, in order of highest l
2 concentration to lowest concentration, are expected to be Cs-137, Co-60, Sr-90, Cs-134, Nb-94, Eu-154, Th-208, Sb 125 and Eu-155. The maximum soil concentrations found were 4.4 pCi/g Cs-137,2.1 pCi/g Cs-134, and 1.1 pCi/g of Co-60.
l Decommissioning activities include removal or decontamination of equipment, decontamination of building wrfaces and structural members, and packing and shinping of radioactive waste.
Approximately 840 m '(30,000 ft ) of contaminated equipment arid debris are expected to be 3
removed and, approximately 1,500 m (50,000 ft ) of contaminated soil may also have to be disposed of. Onsite soil treatment may be employed to reduce the volume of contaminated soil.
5.2 Effluent & Environmental Monitoring Program GA has committed to maintaining the effluent and environmental monitoring program during decommissioning activities as described in GA's License SNM-696. The monitoring program includes airborne effluent ;nonitoring, environmental air sampling, sewage sampling, area s)il l
sampling, and external radiation monitoring.
Airborne Effluent Monitorina All airborne effluents are continuously monitored where calculation indicates that radioactive material can be emitted to the site boundary at concentration levels, averaged for one calender quarter, which are equal to or greater than at least 10% of the appropriate effluent limit listed in 10 CFR Part 20. Samples are analyzed weekly for gross alpha / beta particle activities. Action levels are established for each vent / stack, which wht. exceeded, require implementation of appropriate corrective actions. These action levels are established to ensure the limits specified in 10 CFR Part 20 are not exceeded.
Sewaae Samolina Effluent sewage is sampled daily during normal work-weeks to determine gross alpha / beta particle activity concentrations. Effluents are held in a tank prior to release, until the results of these measurements are established in accordance with authorized release limits. Other than releases to the sanitary sewer, liquid effluents have not been released offsite from the GA faci'ity.
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12 Environmental Air Samolina Ambient air is sampled at a minimum of ten locations on, adjacent to, or near the site. Samples are collected continuously and analyzed weekly for gross alpha and beta particle activity. A composite sample is analyzed monthly for gross gamma activity.
Soil Samolina Soil is sampled annually at ten locations surrounding the GA facility for gross alpha, beta, and gamma activity.
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l External Gamma Radiation Measurements Thermoluminescent detectors (TLDs) are placed at selected air sampling locations and read quarterly.
5.3 Effluent and Environmental Monitoring Data Air Effluent Data Figure 2 shows the total activity of radionuclides releasert in airborne effluents from the GA facility, derived from GA's semi-annual effluent reports submitted in accordance with 10 CFR 70.59. This includes releases of uranium (various enrichments), thorium, mixed activation and l
fission products, and various radionuclides with atomic numbers between 3 and 83, but does not include releases from the TRIGA reactors. The figure demonstrates the generally decreasing trend in radioactive airborne effluents since the 1980s, as operations have declined and decommissioning activities have been conducted.
1 A conservative dose assessment was performed using this information from the period of 1990 I
to 1997. The maximally exposed individual was assumed to be an onsite, non-radiation worker; the closest offsite resident is approximately 1 mile from the facility. GA currently leases areas of l
the facility that have been previously released for unrestricted use. People working in these leased areas are considered members of the public for the purpose of this assessment, unless radiation exposure is part of their assigned duties. Onsite, non-radiation workers are assumed to be exposed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> / day,40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week,50 weeks per year. The average projected total.
effective dose equivalent (TEDE) of 0.15 mSv/yr (15 mrem /yr) is less than 20% of the dose limit for members of the public specified in 10 CFR 20.1301. Because dilution of the radionuclides concentrations from the release points to the individual was not accounted for, actual doses are expected to be far lower.
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Liould Effluent Data l
l As discussed previously, liquid effluents at the facility are sampled and then discharged to the sanitary sewer system. There are no liquid releases to surface or subsurface water at the i
facility, although there may have been historical leaching and shallow transport of radionuclides in the unsaturated zone beneath the site. Such contamination would be expected to be sorbed onto shallow soils beneath the site and will be remediated through exhumation and removal of contaminated soilin excess of release criteria. No radiological contamination has been detected l
in groundwater at the site. The groundwater beneath the main site is located at a depth of 91 m (300 ft). The groundwater below the Sorrento Valley site can be found at a depth of 2.5 to l
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13 3 m (8 to 10 ft) and is brackish. Trichlorethylene has been found in the groundwater at the l
Sorrento Valley site. The United States Environmental Protection Agency, the State of California Environmental Protection Agency, and the County of San Diego are aware of the contamination, and GA is currently working with the County on this issue.
The primary nuclides discharged through the sewage system from operations under License SNM-696 and the State License include U-235, U-238, Th-228, Ra-226, Ra-228, Cs-137, Cs-134,Co-60. Figure 3 shows the total activity in liquid effluents released to the sanitary sewer since 1984. The figure demonstrates that the amount of radionuclides released has decreased over time as operations at the facility have ceased and only decommissioning activities are taking place. From 1994 through the present, no releases have exceeded 10% of the limits specified in 10 CFR 20.2003 and 10 CFR Part 20, Appendix B, Table 3.
6.0 ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION AND ALTERNATIVES 6.1 Proposed Action The NRC staff performed a dose assessment to estimate the impact from airborne radioactive releases under the proposed action. Only radioactive effluents were considered because non-radioactive releases are expected to be insignificant. In addition, because liquid effluents were released only to the sanitary sewer system, the NRC assumed these discharges are insignificant compared to airborne releases due to dilution of the source-term prior to introduction into any potential drinking water source. The NRC again assumed that the maximally exposed individual was an onsite, non-radiation worker. Because the closect resident is approximately 1 mile from the facility, any discharges would be extensively diluted prior to exposure of the resident. The onsite, non-radiation worker was assumed to be exposed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> / day,40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week,50 weeks per year.
The primary nuclides expected to be released dunng decommissioning were estimated to be uranium (of various enrichments) and thorium from D&D of Building 39 (SVB/ Pilot Plant) and Building 22 (TRIGA Fuel Fabrication Facility), mixed activation and fission oroducts from D&D of Building 27 (Experimental Area) and Building 27-1 (EA-1 Bunker Facility), and radionuclides ranging from atomic number (Z) 3 to 105 from D&D of the Building 41 (Nuclear Waste Processing Facility), Building 2 (Science Laboratory Building), and Building 25 (Liquid Waste Treatment Facility). The annual average release concentration of uranium and thorium is based on the release concentration measured during D&D of the HTGR Nuclear Fuel Fabrication Facility (3.5 x 10 " pCi/ml) and on the average release concentrations from Buildings 39 (9.0 x 10-" pCi/ml) and 22 (4.0 x 10 " Ci/ml) from 1990 to 1997, while equipment removal and characterization were in progress. The release concentration of mixed activation and fission products is based on the release concentration (1.1 x 10 " Ci/ml) from D&D of the Hot Cell, where similar types of radionuclides are expected, but at higher levels. Finally, the release concentration of radionuclides ranging from Z3-105 was assumed to be three times the release concentration over the last seven years (5 x 10-" Ci/ml), during extensive decontamination of Building 2. The source terms for the dose assessment are given in Table 1.
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Table 1. Assumed release concentrations in airborne effluents.
j Radionuclides Type Average Annual Release l
Concentration in Ci/mi l
Uranium (various 5 x 10 "
enrichments) and Thorium (assumed to be all Th-232)
Mixed Activation and Fission 1 x 10 -"
l Products (assumed to be all Sr-90) l l
Radionuclides from Atomic 15 x 10 -"
l Number 3 to 105
)
Using these assumptions and the dose conversion factors of Federal Guidance Report No.11, j
l the TEDE to the maximally exoosed individual was estimated to be approximately 0.15 mSv/yr (15 mrem /yr), which is less than 20 percent of the dose limit for members of the public specified in 10 CFR 20.1301. The dose from external exposure was determined to be at least five orders l
of magnitude less than the inhalation dose. The dose assessment is considered to be l
conservative because it assumes that 6 commissioning of all areas of the facility will be l
conducted simultaneously, that Th-232 and Sr-90 are the predominant radionuclides released, I
and that there is no dilution from the release point to the individual exposed. Actual doses are exoected to be far lower. There are an estimated 2,000 employees at the GA facility, which results in a population dose of 0.3 person-Sv (30 person-rem), if all employees were exposed at the maximum projected dose. Doses to offsite members of the public are expected to be orders of magnitude lower due to dilution of the radionuclides concentrations.
1 An accident analysis was included in the 1985 Environmental Assessment, performed by the U.S. Department of Energy, to analyze impacts from decommissioning of the GA Hot Cell Facility. This analysis concluded that there was no significant risk from accidents during decommissioning of this facility. This analysis is considered bounding for the decommissioning activities of the proposed action and is not repeated here.
Cumulative impacts from the proposed action were also considered. As noted previously, l
substantial decommissioning activities have been conducted at the site since the mid 1980s.
Continuous environmental monitoring cf the site throughout this period until the present have not detected any significant environmental impacts. The only on-going activities authorized by the NRC are the decommissioning activities discussed in the proposed action and decommissioning of the Hot Cell Facility, which has been previously approved. The environmental assessment performed for the HCF decommissioning project by the U.S. Department of Energy estimated a l
dose of 4x10 " mSv/yr (0.04 mrem /yr) to the onsite, member of the public (who was considered l
to be the maximally exposed individual) and concluded that there were no environ antal, I
impacts. Cumulative impacts from decommissioning of the Hot Cell Facility and the operations specified in the site DP are, therefore, also expected to also be insignificant. The TRIGA L
15 research reactors are currently not operating. The environmental impact from decommissioning of these facilities will be considered under the NRC Reactor Licenses.
Alternative to the Proposed Action Under the first alternative to the proposed action, GA would cease decommissioning operations, maintain the site in its contaminated state, and then resume decommissioning activities at a later i
time with stricter effluent controls. These actions would reduce the quantity of radioactive effluents expected under the proposed action. However, if contamination is left in-place for a time, there is a potential for the spread of this material to unaffected areas. Decommissioning at a later time may then result in increased effluents or greater worker exposure.
Under the second alternative, GA would terminate the license for the facility under restricted conditions. This alternative would also reduce effluents but may result in a spread of contamination. In addition, the company may also be economically impacted by the inability to sell or lease the facilities and by the resources required to maintain the site. The public would '
also not have the opportunity to use these areas productively.
7.0 REGULATORY CONSULTATION The DP was approved by the State of California by License Amendment dated July 5,1996.
The NRC staff consulted GA and the State of California, Department of Hoalth Services, but did not consult any other State or Federal agencies in preparation of this Environmental Assessment.
8.0 CONCLUSION
Extensive decommissioning operations have been condycted at the GA facility since the mid-1980's. Effluent and environmental monitoring data indicate that all offsite radioactive releases have been below the effluent and dose limits established in 10 CFR Part 20 and have not resulted in any significant human health or environmental impact.
Future decommissioning operations are expected to be similar to decommissioning conducted previously by the facility and are, therefore, not expected to result in any significant environmentalimpact. This conclusion is also supported by a conservative dose assessment performed by the staff, which estimates a dose to the maximally exposed onsite individual of approximately 0.15 mSv/yr (15 mrem /yr). This is significantly below the dose limit for members of the public of 1 mSv/yr (100 mrem /yr) specified in 10 CFR Part 20.
j In addition, GA has committed to engineering controls, waste handling methoas, and an effluent and environmental sampling program to keep releases as low as reasonably achievable and to ensure continued compliance with applicable laws and regulations.
Given the engineering controls, waste handling procedures, projected doses to members of the public and workers, and demonstrated ability to conduct these activities without adverse impacts to the environment, the staff concludes that the proposed action can be implemented without significant environmental impacts.
1 s..
t 16 This environmental assessment was conducted based on preliminary characterization information. If further characterization data indicates that significantly greater concentrations of radionuclides or significantly different types of radionuclides may be released offsite, or if GA determines that significantly different decommissioning activities will be required that may result in significant impacts to workers or the environment, GA will be required to notify the NRC for i
review and approval of the proposed decommissioning activities, l
9.0 REFERENCES
1.
General Atomics, " Site Decommissioning Plan," September 19% and supplements dated December 5,1996; April 18,1997; and January 15,1998.
c 2.
Federal Guidance Report No.11, " Limiting Values of Radionuclides Intake and Air I
Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,"
1988.
l 3.
U.S. Nuclear Regulatory Commission Docket Files, Docket 70-734.
I l
I 4.
U.S. Nuclear Regulatory Commissicn, " Environmental Impact Appraisal Related to Special Nuclear Materials License No. SNM-696," June 1983.
5.
U.S. Department of Energy, " Final Environmental Assessment for Decontaminating and Decommissioning of the General Atomics Hot Cell Facility," August 1995.
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