APPENDIX A: NUCLEAR FACILITIES The Nuclear Weapons Complex (Complex) is comprised of facilities located at 11 major U.S. Department of Energy (DOE) sites, distributed over 10 states. Summary descriptions of the Complex sites considered for tritium supply and recycling are presented in chapters 1, 3, and 4. This appendix examines in more detail four Complex sites and one other, the Idaho National Engineering Laboratory (INEL), which is being considered as a potential addition to the Complex. The five DOE sites include INEL, the Nevada Test Site (NTS), the Oak Ridge Reservation (ORR), the Pantex Plant (Pantex), and the Savannah River Site (SRS). The first section of this appendix provides detailed reference operation assumptions for each site examined in this Programmatic Environmental Impact Statement (PEIS). Information provided includes specific site location descriptions, current missions, facility operations, and environmental regulatory compliance activities associated with ongoing DOE Office of the Assistant Secretary for Defense Programs (DP), other DOE, and non-DOE programs. The next section of this appendix is divided into two parts. First is the detailed description of the tritium supply technologies. This is followed by a detailed description of the tritium recycling facilities which could be either new collocated facilities or, in the case of SRS, upgrade modifications to existing facilities. Each description includes specific information describing missions, assumptions, functional parameters, expected capabilities, process descriptions, special process requirements, utilities, chemicals used, operational resources, transportation, and environmental regulatory setting. The final section discusses tritium supply technology options which are outside the scope of this (PEIS). -TSAR_DOE_SECTION- A.1 Reference Operating Assumptions TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions The reference case for the PEIS for Tritium Supply and Recycling is No Action, which was previously described in section 3.2.1. No Action allows a comparison of tritium supply alternatives for the candidate sites, not against the current nuclear weapons facility configuration, but against the configuration as it would be expected to operate in 2010 and beyond. No Action assumes that all nuclear sites of the current Complex would continue their current missions only with existing facilities that can comply with environment, safety, and health (ES&H) requirements, and at a level that is consistent with current DOE guidance. The basic missions assigned to the sites include the following: research, development, and testing; maintaining nuclear weapons production and testing capability; processing and storage of nuclear materials; operation of an extensive trans- portation safeguards system to assure the safe, secure movement of weapons and strategic quantities of nuclear materials within the continental United States; non-weapon projects; energy programs; and, cooperation with the Department of Defense (DOD) in responding to nuclear accidents or incidents throughout the world. Under No Action, the siting and construction of a new tritium supply facility would not occur, there would be no upgrades/modifications of existing facilities, and future support of the nuclear weapons stockpile would be provided within the confines of the existing Complex capabilities. Some mission requirements for maintenance of the weapons stockpile in the future would not be met under No Action. However, No Action includes those mission requirements to represent a reference against which tritium supply and recycling alternatives that would meet the Department's Atomic Energy Act responsi- bilities could be compared. Sites would continue waste management programs to meet the legal requirements and commitments in formal agreements and would proceed with cleanup activities. Production facilities and support roles at specific sites, however, would be downsized or elim- inated in accordance with the reduced workload projected for the year 2010 and beyond. Detailed reference descriptions of the affected nuclear sites follows. These descriptions include discussion of the site location, missions, facility operations, and environmental regulatory compliance. -TSAR_DOE_SECTION- A.1.1 Idaho National Engineering Laboratory TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions A.1.1 Idaho National Engineering Laboratory Site Description. INEL is located in 4 counties and covers 890 square miles in southeastern Idaho near Idaho Falls. The main site is 34 miles west of Idaho Falls, 38 miles northwest of Blackfoot, and 22 miles east of Arco. There is also an annex in Idaho Falls. The facility covers more than 569,000 acres with approximately 277 miles of roads, both paved and unpaved, and 30 miles of railroad tracks (figure 4.2.1-1 and table 4.2.2.2-1). There are 450 buildings and 2,000 support structures at INEL with more than 3 million square feet (ft2) of floor space in varying conditions of utility. INEL has approximately 270,000 ft2 of covered warehouse space and an additional 200,000 ft2 of fenced yard space. The total area of the various machine shops is 32,665 ft2. The majority of experiment sites are no longer in use and are scheduled for decontamination and decommissioning (D&D) and restoration. There are three reactors operating continuously and eight sites that formerly housed reactors, now scheduled for D&D. Warehouses are located in all onsite areas and, on average, are filled to more than 85 percent of capacity. Also, there are 11 machine shops on the site to support all activities. Besides the main site, there is also an INEL annex in Idaho Falls which has office buildings, a computer center, and a research and development laboratory where site related technical, analytical, and administrative activities are conducted. Four separate management and operations contractors operate the Idaho Chemical Processing Plant, the Naval Reactor Facility, Argonne National Laboratory-West, and the remainder of the site for DOE. Transportation to and from INEL is provided to all persons employed onsite. There are no long-term housing facilities at INEL. INEL procures more than $25 million worth of material, supplies, and construction services in the region per year. Missions. The missions of INEL are: . Provide waste management functions. . Perform waste processing, technology research, and development. . Perform research on reactor safety operations, materials testing, and environmental monitoring. . Perform breeder reactor irradiation tests. . Perform irradiation service, geological and defense research, and develop nuclear instruments. . Maintain a standby facility for conduct- ing ship propulsion reactor research and training. Facility Operations. Activities at INEL have been divided among nine distinct and geographically separate functional areas (table 3.3.2-1). Fuel reprocessing activities at the Idaho Chemical Processing Plant have been terminated and operations are focused on spent fuel storage and high-level waste processing. The Radioactive Waste Management Complex is a storage and disposal facility intended primarily for radioactive materials from defense and nuclear energy research programs. The Complex consists of three main areas: an 88-acre subsurface disposal area, a 56-acre tran- suranic storage area, and an administrative area. The Power Burst Facility supports research and development for waste reduction programs. A reactor used for thermal fuels behavior studies is now in a standby mode. Major waste reduction facilities include the Waste Engineering Development Facility, the Waste Experi- mental Reduction Facility, and the Mixed Waste Storage Facility. The Test Area North complex is the northernmost facility within INEL and consists of several experimental reactor and support facilities conducting research and development activities on reactor performance. These include the Technical Support Facility, the Containment Test Facility, the Water Reactor Research Test Facility, and the Inertial Engine Test Facility. The Inertial Engine Test Facility has been abandoned with no future programs planned. The remaining facilities support ongoing programs that are expected to continue for the foreseeable future. The Auxiliary Reactor Area is the site where materials testing and environmental monitoring activities are conducted. The facilities in this area are scheduled for D&D. The Argonne National Laboratory-West is dedicated to breeder reactor development. The Test Reactor Area supports the Advanced Test Reactor which is used for irradiation testing of fuel, core materials, and instrumentation for naval reactors. Wastes from this facility are handled by the Radioactive Waste Management Complex. The Naval Reactors Facility is operated for DOE and U.S. Navy by Westinghouse Electric Corporation under jurisdiction of DOE's Pittsburgh Naval Reactors Office. Activities include the testing of advanced design equipment and new systems for current naval nuclear power propulsion plants and obtaining data for future design. The Central Facilities Area provides effective site-wide support services including transportation, shop services, health services, radiation monitoring, and administrative offices. Defense Program Activities. There are no defense program activities currently being performed at INEL. Other Department of Energy Activities. The Test Area North is used for light water reactor safety tests and specific manufacturing. The Naval Reactor Facility contains four reactors (being shutdown in 1995) where new designs were tested: the Submarine Prototype (S1W), the Large Ship Reactor (A1W), the Natural Circulation Submarine Prototype (S5G), and the Expended Core Facility. At the Test Reactor Area, the Advanced Test Reactor is currently operating to test reactor fuel, targets, core materials, and electronics. The Engineering Test Reactor and the Materials Test Reactor are in standby and the buildings are used for offices and test laboratories. The Auxiliary Reactor Area is being utilized for materials testing, but is scheduled for D&D in the near future. The Power Burst Facility reactor is on standby, but the area is used for low-level waste (LLW) engineering and development and for mixed waste storage. Argonne National Laboratory-West operates an Experimental Breeder Reactor for research and development. The Radioactive Waste Management Complex is used for examination and certification of INEL wastes. It also provides for storage of retrievable transuranic (TRU) waste and disposal of low-level radioactive (beta-gamma) waste. The Central Facilities Area provides support services for the entire site. There are a variety of support and service organizations onsite. These include the facilities to handle security, fire protection, facilities service and maintenance, food preparation, mail, transportation, medical, communication, warehousing, and machine shops. Non-Department of Energy Activities. There are non-DOE activities at INEL which include research being conducted by the National Oceanic and Atmospheric Administration, the U.S. Geological Survey (USGS), and various institutions of higher learning. These activities support the designation of INEL as a National Environmental Research Park. Environmental Regulatory Setting. The Environmental Oversight and Monitoring Agreement between DOE and the State of Idaho, signed May 21, 1990, was developed to assure the citizens of Idaho that their health and safety and the environment are being protected. This voluntary agreement addresses understandings and commitments between the parties regarding DOE's provision to the State of Idaho of technical and financial support for state activities to assess compliance with applicable laws and regulations at INEL. These activities consist of environmental oversight, monitoring, and evaluations of emergency response plans. The independent monitoring includes onsite discharges, groundwater and air quality, and offsite radioactivity, as well as evaluation of waste minimization planning and source reduction methods. This agreement has been extended until negotiations for a new agreement, begun in 1994, are concluded. The Department is working with Federal and state regulatory authorities to address compliance and cleanup obligations arising from its past operations at INEL. The Department is engaged in several activities to bring its operations into full regulatory compliance. These activities are set forth in negotiated agreements that contain schedules for achieving compliance with applicable requirements, and financial penalties for nonachievement of agreed upon milestones. On December 21, 1989, the Environmental Protection Agency (EPA) placed INEL on the National Priorities List (NPL) as a "Superfund Site" pursuant to the provisions of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). This determination was based on the contamination present due to past practices. Air. The INEL air emission inventory, completed in March 1991, catalogs all vents and stacks at INEL. The air toxics inventory for radioactive and other hazardous air pollutants is being compiled and, when added to the air emissions inventory, will serve as the basis for the operating permits required under Title V of the Clean Air Act (CAA) Amendments of 1990. INEL is in full compliance with National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities (40 CFR 61, Subpart H), with an effective dose equivalent to the public for 1991 at 0.004 millirem (mrem) per year (IN DOE 1992d:34). The Idaho Operations Office signed a Consent Order on February 11, 1992, with the State of Idaho whereby INEL agreed to pay a monetary penalty and to apply for the appropriate air quality permit for the construction and operation of potential air pollution sources. A Notice of Violation issued in June 1991 alleged these facilities had been constructed without the required permits. Water. The Snake River Plain aquifer, a sole source aquifer which lies beneath INEL, serves as the source for drinking water and crop irrigation in the Snake River Basin. Natural radioactivity is found in the Snake River Plain aquifer in areas upgradient, parallel to, and distant from INEL. Onsite and offsite water samples are collected routinely to monitor the movement of waste substances, both radioactive and nonradioactive, through the aquifer onsite. Onsite drinking water samples are collected monthly from production drinking water wells in use at active site facilities. Approximately 25 percent of all drinking water samples collected in 1991, contained detectable concentrations of gross alpha activity. The Idaho Operations Office submitted applications to the State of Idaho for two National Pollutant Discharge Elimination System (NPDES) permits in January 1992. These permit requests are for discharges of noncontact cooling water and for discharges of wastewater to the Big Lost River from the Idaho Chemical Processing Plant. No other INEL facility discharges liquid effluents to surface waters other than stormwater, and no streams or rivers flow from within INEL to locations outside the boundaries. In addition to the two NPDES permits, INEL has filed nine deep injection well permit applications with the State of Idaho. The injection wells are used to dispose of stormwater runoff. DOE continues to update the inventory of shallow injection wells. On October 7, 1992, the Idaho Operations Office signed a Consent Order with the State of Idaho to settle the Notice of Violation issued on June 7, 1991, for alleged violation of Idaho water quality regulations. The Notice of Violation alleges that the Idaho Operations Office is subject to a state permit program for systems treating wastewater by application to the land via percolation ponds and unlined sewage systems at the Idaho Chemical Processing Plant. DOE will obtain stormwater discharge permits under NPDES. Also, draft wastewater land application permits have been issued and are currently being negotiated. INEL is currently covered under two general permits: stormwater discharges associated with industrial activity and construction activity. Land. The Idaho Hazardous Waste Management Act and EPA delegation of authority to the state provides the Idaho Department of Health and Welfare authority to enforce hazardous waste management regulations and to provide oversight of Resource Conservation and Recovery Act (RCRA) related environmental restoration activities. INEL has submitted the RCRA Part B permit application for treatment, storage, and disposal. In 1989, the Governor of the State of Idaho placed a moratorium on the receipt of TRU waste from out-of-state. INEL is honoring the governor's moratorium even though solid radioactive wastes from Rocky Flats Environmental Technology Site (formerly known as the Rocky Flats Plant) have been received and buried at INEL since 1954. Offsite generators, including the Rocky Flats Environmental Technology Site and the Mound Plant, which had been approved to make routine waste shipments to the Radioactive Waste Management Complex at INEL, are adversely affected. INEL has also agreed to terminate TRU waste disposal activities on INEL. TRU waste not buried prior to this agreement is being stored at the Radioactive Waste Management Complex in compliance with applicable regulatory requirements, awaiting permanent disposal at a Federal repository. Three areas of confirmed releases of contamination to the environment at INEL have been identified: The Radioactive Waste Management Complex, where small amounts of volatile organic compounds have been measured in the groundwater aquifer and trace concentrations of TRU radionuclides were detected at a 110-foot depth above the aquifer. Both contaminants are suspected to be traceable to buried TRU waste. Additional contaminants that potentially may be released include petroleum products, acids, bases, volatile organic compounds, heavy metals, radionuclides, polychlorinated biphenyls (PCBs), and asbestos. The Test Reactor Area, where chromium compounds have been detected beneath a wastewater pond in groundwater that is perched above the Snake River Plain aquifer. Remedial action is underway. The Test Area North groundwater, where volatile organic compounds were measured in wells that supply drinking water. Remedial action is underway. DOE entered into a Federal Facility Agreement and Consent Order with EPA and the State of Idaho. The agreement, signed December 9, 1991, meets the requirement of Section 120(e) of CERCLA (42U.S.C. 9601 et seq.) for an interagency agreement with EPA for Federal agencies that have facilities included on the NPL. The Federal Facility Agreement and Consent Order is implemented by an Action Plan outlining the remedial action process, which will encompass investigation of hazardous substances and cleanup activities at INEL. The purposes of the Federal Facility Agreement and Consent Order Action Plan are to: Establish a procedural framework and schedule for developing and monitoring appropriate response actions at INEL in accordance with CERCLA, RCRA, and the Idaho Hazardous Waste Management Act. Facilitate cooperation, exchange of information, and participation of the parties in such actions. Minimize duplication of analyses and documentation. Expedite the clean-up process to the maximum extent possible consistent with protection of human health and the environment. Supersede the existing RCRA 3008(h) Consent Order and Compliance Agreement executed on July 10, 1987. In February 1990, INEL received a Notice of Noncompliance issued by EPA on January 29, 1990, for 28 alleged violations of RCRA regulations based on a June 5 to 9, 1989, inspection. The majority of violations have been resolved and long-term technical solutions for the Radioactive Waste Management Complex and Idaho Chemical Processing Plant have been agreed upon. On April 3, 1992, the Idaho Operations Office, the State of Idaho, and EPA Region 10 signed a Consent Order to settle the unresolved issues from the 1990 Notice of Noncompliance. The sodium in Building 703 (1,400 55-gallon drums) at the Argonne National Laboratory-West site has been declared a waste. DOE has agreed to manage the sodium in Building 703, "...in compliance with all applicable interim status provisions of the Hazardous Waste Management Act and the Rules, Regulations and Standards for Hazardous Waste..." (IN DOE 1993a:124). The Consent Order sets up a schedule for corrective actions to be taken in the management of radioactive sodium bearing liquid wastes currently stored at the Idaho Chemical Processing Plant in the tanks with concrete vault secondary containment. Because the concrete vaults would react with the acidic solutions stored in the tanks, they do not meet RCRA requirements for secondary containment of hazardous wastes. The corrective actions at the Radioactive Waste Management Complex, Argonne National Laboratory-West, and the Idaho Chemical Processing Plant could take more than 25 years. On October 7, 1992, the Idaho Operations Office signed another Consent Order with the State of Idaho to resolve the Hazardous Waste Notice of Violation issued by the state on June 5, 1991, for 23 alleged hazardous waste violations identified during the September 10 to 14, 1990, inspection. This Consent Order provides the schedule for corrective actions such as establishing satellite accumulation areas, closure of the Idaho Chemical Processing Plant percolation ponds, and discontinuing the discharge of non-RCRA wastewater into the percolation ponds prior to formal "clean" closure. Completion of these actions will resolve this Notice of Violation. INEL generates mixed wastes that are radioactive and contain RCRA hazardous wastes banned from land disposal. The primary INEL compliance concerns for mixed wastes are the radioactive-contaminated wastes containing solvents and California list wastes as defined in 40 CFR 268.5. These mixed wastes are subject to the RCRA land disposal restric- tions which ban the land disposal of certain listed hazardous wastes unless they meet specific treatment standards. Due to the nationwide shortage of treatment and disposal facilities for these types of waste, INEL is storing these mixed wastes until better treatment and disposal options are developed. Such storage violates the Land Disposal Restrictions provisions of RCRA which permits storage only for accumulation of sufficient quantities to facilitate proper treatment, recycle, or disposal. Therefore, the Idaho Operations Office has proposed to enter into a Compliance Agreement with DOE and the State of Idaho concerning the storage and continued generation of land disposal restriction waste until a treatment method is developed and WIPP or another suitable repository is open for disposal of mixed TRU waste. The land disposal restrictions also prohibit the storage of new restricted waste, including storage of new calcine, generated after May 1992, unless it is being accumulated to facilitate the proper treatment, disposal, or recovery, or there is an approved case-by-case extension. Because there are more than 1million pounds of radiologically contaminated lead at INEL, serious efforts are being made to develop a lead treatment process that can satisfy the land disposal restriction. Schedules to develop such technologies are proposed in the draft Site Treatment Plan. When finalized, the Site Treatment Plan will satisfy DOE's obligation under the Federal Facility Compliance Act of 1992 to develop and submit a mixed waste treatment plan for INEL. INEL is not presently in full compliance with the Toxic Substances Control Act (TSCA) due to the storage of some polychlorinated biphenyl (PCB)- contaminated equipment and materials. The equipment and materials are also contaminated by various radioactive elements and treatment and disposal technology does not currently exist for these materials. DOE has an aggressive program to develop this technology but until it does, these materials must be safely stored in violation of the TSCA ban on the storage of such materials. DOE is presently negotiating a compliance agreement with the State of Idaho for the continued storage of radioactive PCB-contaminated wastes at INEL until a method for their treatment or disposal can be developed and permitted. -TSAR_DOE_SECTION- A.1.2 Nevada Test Site TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions A.1.2 Nevada Test Site Site Description. NTS is located in Nye County, NV, and encompasses approximately 864,000 acres. It varies in width from 28 to 35 miles (east to west) and in length from 40 to 55 miles (north to south). To the north, east, and west, the rugged, mountainous, undeveloped, Federal-owned land masses of the Nellis Air Force Range provide a buffer zone, varying from 15 to 65 miles wide, between the test areas and public lands. The Bureau of Land Management manages the land which borders the southern and southwestern boundaries. United States Highway 95 and the town of Amargosa Valley are also to the south. The southeast corner of NTS is about 65 miles northwest of Las Vegas. NTS is unique in that it is a large open area for which access is tightly controlled, with adequate infrastructure, to handle and run tests with hazardous or radioactive materials. Approximately 25 percent of NTS is currently undeveloped or provides buffer zones for on-going programs and projects. Facility expansions are possible within all areas and encroachment from land development is not a concern. NTS is divided into numbered test areas to simplify the distribution, use, and control of resources (figure 4.3.1-1). The main entrance and the Desert Rock airstrip are at the southeast corner of the site (Area22). Mercury base camp is adjacent in Area 23 and provides administrative operations and general support. Offices for DOE, DOD, Defense Nuclear Agency, Lawrence Livermore National Laboratory, Los Alamos National Laboratory, Sandia National Laboratories, New Mexico, and all of the supporting contractors of these organizations are located in this area. Dormitory, cafeteria, recreation, and transpor- tation facilities are located here. North of Mercury is Frenchman Flat (Area 5), an area historic for its atmospheric nuclear tests. Just north of Frenchman Flat is Area 6. The Control Point One facility which provides control over and execution of nuclear detonations at NTS is located here as is a new work-camp for construction and craft support. A shallow, usually dry, lake bed, Yucca Lake, is also in this area. Further north is the broad valley of Yucca Flat, site of much of the more recent nuclear testing (Areas 1, 2, 3, 4, 7, 9, and 10). At the northern edge of this flat at the base of Rainier Mesa, is the center of DOD/Defense Nuclear Agency activities (Area12). The Area 12 Camp has logistic, service, and administration facilities, which, in busier times, supported the northern part of NTS but which has now been closed and consolidated with the Mercury and Area 6 camps. The Area 12 Camp provided ready access to the Defense Nuclear Agency tunnels mined into the face of Rainier Mesa. In the northwest section of NTS is Pahute Mesa. Its geology, combined with the greater distance from Las Vegas, allows its use for testing nuclear devices with larger yields (Areas 19 and 20). Due to its large size, the perimeter of NTS is not fenced. However, roving security guards patrol the test site. Security and hazardous areas are fenced and some are protected with armed guards and electronic security measures. Capital assets at NTS include about 1,200 buildings with 8,000 units of installed equipment, approximately 300 miles of primary and secondary surfaced roads, and 400 miles of unsurfaced roads. The NTS water system consists of 15 wells, pumps, booster pumps, and many sumps, reservoirs, chlorinator water softeners, and 100 miles of supply and distribution lines. This water system has an average weekly production of 8.5 million gallons. Total well capacity is 5,752 gallons per minute (gpm). Twelve wells supply water for domestic use on NTS. Mercury basecamp is supplied by 3 wells: 2 in Area 5 and 1 near Desert Rock Airstrip. Electrical power to NTS is supplied by Nevada Power Company and Valley Electric Association transmission lines. Both transmission lines are rated at 138 kilovolt (kV). The Nevada Power Company line is approximately 60 miles long and ties into the NTS transmission system near Mercury. The Valley Electric Association line is more than 100 miles long. It runs from the Amargosa Valley substation and ties into the NTS transmission system at Jackass Flats substation. This system (the Nevada Power Company/Valley Electric Association transmission lines) is capable of providing 45 megawatt electric (MWe) based on a single contingency failure. NTS has over 700 miles of overhead and underground transmission and distribution power lines. NTS also uses a small amount of fuel oil. Table 4.3.2.2-1 shows the annual usage of resources. Missions. The missions of NTS include: Provide the capability to conduct underground nuclear weapons tests. Provide technology, facilities and expertise for non-DOE customers. Support site characterization of Yucca Mountain. Conduct environmental assessment and remediation. Provide the DOE response for radiological or malevolent nuclear emergencies. Dispose of low-level radioactive waste for DOE. Conduct onsite and offsite technical monitoring for nuclear treaty compliance. Facility Operations. In December 1950, President Truman established the Nevada Proving Grounds (forerunner to NTS) as the Nation's on-continent nuclear weapons testing area. The first nuclear test at NTS occurred on January 27, 1951. At that time, the nuclear weapons program was administered by the Atomic Energy Commission Albuquerque Operations Office. Atomic Energy Commission employees were sent to the Nevada Proving Grounds for the duration of a test series and then returned to Albuquerque. As tests became more frequent during the 1960's, Atomic Energy Commission created the Las Vegas-based Nevada Operations Office, which officially opened on March 6, 1962, and has administered NTS operations since then. Approximately 40percent of the total Nevada Operations Office budget for fiscal year 1992 was for defense programs activities. NTS is operated by four major management and operations contractors. Contractor employment levels at the test site are dependent upon programmatic requirements and have varied greatly during the history of the site. During 1988, the peak year, roughly 11,000 contractor employees were assigned to Nevada Operations Office administered activities at NTS and miscellaneous other locations, including metropolitan Las Vegas and North Las Vegas, Nevada. From that peak employment level, and particularly in view of the test moratorium that was first implemented in the fall of 1992, contractor employment levels decreased to approximately 7,700 in 1992 and further to approximately 6,850 in 1994. The majority of the facilities at NTS were constructed 25 to 30 years ago as temporary structures; less than ten percent have been constructed in the past 15 years. Many of the facilities are also currently inadequate in one or more of the structural, mechanical, or electrical categories. Although Mercury basecamp has close to 100 percent occupancy and some forward areas are at 80 percent occupancy levels, fiscal pressures are forcing the closure of the forward area facilities and consolidation at the Mercury basecamp. Consequently, most of the Mercury facilities are occupied with little reserve capacity. Desert Rock Air Strip is located southwest of Mercury. The airstrip has in busier times provided scheduled air service by DOE aircraft between NTS and Lawrence Livermore National Laboratory, Los Alamos National Laboratory, and Sandia National Laboratories, New Mexico, for access by researchers and testing personnel. Currently it is used only for high priority shipments. Only one major new facility, the Device Assembly Facility, is currently under construction. However, modification of existing facilities on an as needed basis, is an ongoing activity. Drilled holes for groundwater monitoring are always in the process of being selected, designed, and developed. A waste management facility is being considered for handling TRU waste from DOE facilities; this is the only major nondefense program facility anticipated for NTS. Defense Program Activities. Historically, most of the work carried out onsite has been related to defense program activities. Since it was established in December 1950, NTS has been the primary testing location for the Nation's nuclear explosives program. As of September 30, 1992, the United States had conducted 1,054 nuclear tests, 928 which were on the NTS and 828 of which were underground. A breakdown of the categories of the 928 nuclear tests conducted within the boundaries of the NTS from 1951, through 1992, is as follows: 100 atmospheric tests, 16 of which were safety tests that by design produced little or no nuclear yields. 10 cratering tests (i.e., shallow burial of a nuclear device). 9 tests in unstemmed holes to minimize, but not eliminate, the release of radioactivity to the atmosphere. 742 underground tests in drilled or mined shafts. 67 tunnel tests. In typical defense program weapons development tests during the past several decades, a nuclear device was emplaced in a vertical drilled hole (86 to 190inches in diameter) at depths of from 1,200 feet to 2,500 feet in the Yucca Flat basin or at Pahute Mesa. The nuclear device was assembled into the desired testing configuration and mated with a diag- nostic equipment canister at NTS. Operational tests were performed repeatedly, both before and after the nuclear device was placed "down-hole," to ensure proper functioning of the telemetry. After the test package, up to 200 feet in length, was lowered to its proper depth, the vertical emplacement hole was backfilled with different materials in order to contain the resulting radioactive debris when the nuclear device was exploded. After detonation, sample recovery holes were bored into the test cavity to obtain samples of the nuclear debris for subsequent radiochemical analysis by the weapons design laboratory that furnished the nuclear device. Underground testing was controlled at the Area 6 Control Point One. This facility contains the technical, managerial, and safety infrastructure to control the site. As has previously been noted, since the U.S. Nuclear Testing Moratorium Act went into effect in early October 1992, no nuclear tests have been conducted by the United States. On the day immediately following China's October 4, 1993, nuclear test, President Clinton issued a directive to DOE to continue to maintain indefinitely a state of readiness for possible resumption of U.S. testing. The Device Assembly Facility is the only new major facility for defense program activities at NTS. This 100,000 ft2 facility was authorized in 1984, and is under construction now. Physically, it is located just south of Control Point One. It will combine and cen- tralize all the functions and facilities of the existing device assembly area. Once the Device Assembly Facility is operational, Lawrence Livermore National Laboratory and Los Alamos National Laboratory will be able to conduct multiple operations with high explosive(s) (HE) and nuclear devices simultaneously. All aspects of the operations will be handled in this one facility due to its multiple processing areas which include; assembly cells, assembly bays, high bays, radiographic facilities, special nuclear materials laboratories, HE staging, special nuclear materials staging, shipping and receiving areas, and associated administrative and support areas. In addition, the facility will provide for increased overall security and permit easier entrance/exit for the workers during hazardous operations. There will be no manufacturing or machining of special nuclear materials at this facility; only the assembly/disassembly and material storage/staging functions would be handled at this facility. The Nevada Operations Office has been delegated the lead Federal role in maintaining the capability to respond to certain kinds of national emergencies. It will provide the leadership when a Federal Radiological Monitoring and Assessment Center is estab- lished. Additionally, a team of highly trained DOE and contractor radiological specialists known as the Nuclear Emergency Search Team trains, tests equipment for search and detection, and stores equipment for rapid deployment under the auspices of the Nevada Operations Office. It can be mobilized in case of accidents involving radioactive materials or a terrorist threat involving nuclear weapons. Other Department of Energy Activities. Although the principal activity at NTS is the underground testing of nuclear devices, DOE is also involved in a number of other activities. These activities include liquified gaseous fuels spill testing, radioactive and mixed waste disposal, and the Yucca Mountain characterization programs. NTS has also been designated a DOE National Environmental Research Park. The Liquified Gaseous Fuels Spill Test Facility in Area 5 was completed in 1986. It was operated on a fee basis for commercial users as a basic research tool for studying the dynamics of accidental releases of hazardous materials and to evaluate the effectiveness of various foams and fire retardants in accidents involving chemicals and hazardous materials. Test facility personnel discharge a measured volume of hazardous test fluid at a controlled rate onto a surface specially prepared to meet the test requirements and record close-in and downwind meteorological data and gaseous concentration levels. NTS is participating in a program sponsored by DOE to establish a Nevada Solar Enterprise Zone. As part of this program, a 100 megawatts (MW) solar power plant may be built at NTS. The power from this plant would support Government needs in the area with the remainder being sold to the commercial grid. This size plant can be supported with the existing transmission lines at NTS. There is also the potential to expand the solar power capability at NTS to approximately 500 MW in the future. However, this expansion would require substantial infrastructure upgrades including new transmission lines. It is assumed in this PEIS that the first 100 MW will be in place by the 2005 No Action time frame. NTS also operates radioactive waste disposal facilities. The Radioactive Waste Management Site, located in Area 5, accepts LLW materials which were generated in the Nation's defense programs activities. This 92-acre facility consists of trenches and pits for burial of short-lived LLW, and aboveground storage of TRU waste awaiting transfer to the Waste Isolation Pilot Plant (WIPP). Also located at the Area 5 Radioactive Waste Management Site is the Greater Confinement Disposal Facility which consists of a 10foot diameter lined shaft 120 feet deep. This facility is used for disposal of waste not suited for shallow land burial because of potential for migration into biopathways. Nonradioactive hazardous wastes are also accumulated at the Area 5 Radioactive Waste Management Site awaiting shipment to offsite disposal facilities. In Area 3, the Bulk Waste Management Site uses surface subsidence craters (that were formed by underground nuclear tests) for the emplacement and burial of LLW in bulk form (such as debris collected from atmospheric nuclear test locations). The Yucca Mountain Site is located at the edge of NTS. It is being considered by DOE for the disposal of spent power-reactor fuel and vitrified high-level waste (HLW), the latter resulting principally from defense program activities. The Yucca Mountain Site Characterization Project staff reports directly to DOE's Office of Civilian Radioactive Waste Management. However, because it has elements based on NTS, the Nevada Operations Office provides some administrative and operational support services to the project. Current designs for the Yucca Mountain Project reflect an electrical demand of 10 MWe; this design has been downsized from earlier estimates of 30 MWe. Recently, NTS has been designated as a DOE National Environmental Research Park with a purpose of consolidating previous ecological reports, filling in a significant gap in the existing DOE research park network, and providing a unique opportunity for research in the arid desert environment. This not only enables NTS scientists to link into the existing ParkNet computerized data system, but also makes the extensive accumulation of envi- ronmental research collected over the history of NTS available to students and scientists throughout the world. NTS's location in the transition zone between the Southern and Northern Basin and Range Ecological Regions, and its inclusion of vast undisturbed areas of mountain ridges, closed basins and diverse ecological communities makes it particularly valuable. From 1959 through 1973, NTS supported a series of open-air nuclear reactor, nuclear engine, and nuclear furnace tests in Area 25 at the Nuclear Research and Development Area. Another series of tests with a nuclear ramjet engine was conducted in Area 26. Non-Department of Energy Activities. The most significant NTS activity involving non-DOE organizations has been the Defense Nuclear Agency's nuclear testing facility. Congressional legislation (the Hatfield Amendment), however, limited nuclear testing to those tests that support the safety and reliability of the U.S. nuclear stockpile. This may preclude further Defense Nuclear Agency nuclear tests which are done to support research into nuclear weapons effects. Defense Nuclear Agency nuclear tests occurred in horizontal tunnels mined beneath Rainier Mesa. The nuclear devices for these tests were designed, built, funded, controlled, and executed by DP. Defense Nuclear Agency's nuclear testing provided the data base and design information for both nuclear effects and survivability. Nuclear weapons effects were studied for all U.S. tactical and strategic weapons systems that were required to operate in a nuclear warfare environment. These tests played a major role in maintaining high confidence in the nuclear stockpile and nuclear capable weapon systems. The weapons effects tests were conducted to study a number of nuclear effects including x-ray, gamma- ray, neutron, stress (thermal, electrical, and mechanical), electromagnetic pulse, airblast, ground and water shock propagation, and temperature. These tests assessed both weapons effects and the survivability of military systems in a nuclear environment. Area 25 has been used for a variety of purposes, including U.S. Army ballistic research using depleted uranium and transporter testing for the proposed mobile MX missile. Various military exercises and training activities are also conducted in and around Area 25. The Desert Research Institute, EPA, the University of Utah, and the Nevada Operations Office operate the Community Radiation Monitoring Program. This program provides the community surrounding NTS with an increased understanding of it's activities and the natural radiation environment. Other activities have been, and will likely continue to be, carried out for other Federal departments and agencies. Representatives from the EPA, USGS, and National Oceanic and Atmospheric Administration are onsite to assist and monitor conditions. Environmental Regulatory Setting. Underground and aboveground testing at NTS has resulted in contamination of surface and subsurface soils and water, and the release of some radioactive isotopes and byproducts, such as metals, into the environment. A Memorandum of Understanding between DOE and the State of Nevada covers radiological releases on NTS and the required notifications. DOE has signed a Programmatic Agreement with the State of Nevada to cover archaeological and historical preservation activities. All future activities at NTS must comply with both the Memorandum of Understanding on radiological releases and the Programmatic Agreement. The Department is working with Federal and state regulatory authorities to address compliance and cleanup obligations arising from its past operations at NTS. The Department is engaged in several activities to bring its operations into full regulatory compli- ance. These activities are set forth in negotiated agreements that contain schedules for achieving compliance with applicable requirements, and financial penalties for nonachievement of agreed upon milestones. This section discusses the more important agreements and other regulatory issues that must be considered before making a decision that would affect NTS. An Agreement in Principle with the State of Nevada was signed in October 1990, and provides DOE funding to Nevada for oversight of ES&H activities, including the environmental restoration activities at NTS. The Agreement in Principle also provides for understanding between, and commitment of, the parties regarding DOE's provision to the state for technical and financial support in return for environmental oversight and monitoring. NTS environmental permits include 43 different State of Nevada air quality operating permits involving emissions from construction, operation of facilities, boilers, storage tanks, and open burning. Five permits for onsite drinking water systems, eight permits for hauling septage, and one consolidated permit for four sewage discharges to onsite lagoons or septic tank fields also have been issued by the State of Nevada. The RCRA Part B permit application for hazardous waste was approved by the state in 1990 as number NV3890090001. Air. Air pollution sources at NTS formerly included aggregate production, stemming activities, surface disturbances, fugitive dust from unpaved roads, fuel burning equipment, open burning, and fuel storage facilities. The air pollutant of major concern at NTS from these former activities was particulate matter which included primarily fine sands emitted during stemming and tunneling operations, and carbon particles emitted during fuel combustion. While most of these activities stopped with the cessation of underground testing, the NTS air quality operating permits still contain clauses limiting the emission of particulates. Radionuclide emissions are not a problem as no significant emissions of radioactive material from venting, ventilation, or seeps have reached the uncontrolled areas surrounding NTS since 1980. Many of these air quality operating permits require annual reports on operating hours, production summaries, occurrences of open burning, and other similar information. For example, the Nevada Air Quality Officer must be notified of each burn no later than five days following the burn. During 1990, three open burns of explosives-contaminated debris in Area 27 were reported for this permit. Also, the Nevada Air Quality Officer must be notified by telephone at least two working days in advance of each training exercise for Class A flammables, and a written summary of each exercise must be submitted within 15 days following the exercise. During 1990, seven burns were conducted for radiological emergency response training and ten training burns were conducted by onsite fire protection services. Water. Effluents at NTS are primarily the result of equipment cleaning and sanitary wastes. Discharges also result from groundwater seeping into the tunnels in Rainier Mesa. Tunnels have been sealed with the exception of P-Tunnel which drains to a holding tank. There are no NPDES permits for Nevada Operations Office facilities because there are no wastewater discharges to onsite or offsite surface waters. Water monitoring at NTS was limited to sampling wastewater effluent to lagoons and ponds under a series of State of Nevada permits. Land. NTS manages two radioactive waste management sites and one hazardous waste accumulation site. As a result of its vast area, prior aboveground testing activities, and remote location, NTS serves as a disposal site for LLW generated by onsite and offsite DOE nuclear weapons program activities and facilities. NTS also serves as an interim storage site for TRU and mixed TRU wastes from its own activities and from Lawrence Livermore National Laboratory activities, prior to eventual shipment to WIPP for final disposal. Extensive environmental surveillance is conducted on NTS to monitor the radioactive waste management program sites. NTS previously served as a disposal site for DOE-nuclear-weapons-program-generated mixed wastes received under an interim status provision granted by the State of Nevada. In May 1990, mixed waste disposal operations were discontinued due to EPA issuance of the RCRA land disposal restrictions for the Third Thirds wastes. Active mixed waste disposal operations at the NTS will not commence until issuance of a State of Nevada RCRA Part B permit. The State of Nevada has been delegated authorization to enforce RCRA by EPA. The Nevada Division of Environmental Protection provides RCRA overview of NTS. The RCRA Part B permit application, submitted in 1990, for disposal of hazardous waste and storage of mixed waste, was approved in April 1995 for storage. Activities continue to permit additional mixed waste management facilities. The Hazardous Waste Accumulation Site consists of an impervious concrete pad with 6-inch curbs to contain spillage and runoff and a roof to protect the wastes from rain. This site is used to collect nonradioactive hazardous wastes from satellite accumulation areas prior to shipping offsite to a RCRA- permitted commercial hazardous waste disposal facility. On August 10, 1992, the state notified the Nevada Operations Office of its opinion that the uranium oxide material called Cotter Concentrate stored at NTS is probably a waste that has not yet been declared a waste. The state has requested chemical and physical data regarding the material and, if known to contain RCRA constituents or is uncharacterized, justification should be provided for continuing to classify the material as strategic material, rather than as a waste. On January 8, 1993, DOE declared that this material has no further programmatic use and that it be transferred to waste management for final disposition. The Radioactive Waste Management Site at Area 5 is a 732-acre remote radioactive waste storage and disposal facility. Area 5 contains LLW disposal units consisting of pits and shallow trenches. It also provides the Greater Confinement Disposal Unit which consists of a 10-foot diameter shaft 120 feet deep used for experimental disposal of wastes not suited for shallow land burial because of high specific activity or a potential for migration into biopathways. Area 5 also serves as temporary storage for TRU and mixed TRU wastes on a curbed asphalt pad. Approximately 800 cubic yards (yd3) of TRU waste are stored in 55-gallon drums on pallets and in various assorted steel boxes pending shipment to WIPP. DOE and the State of Nevada signed a Settlement Agreement on June 23, 1992, regarding alleged violations of storage of mixed TRU waste at NTS. Currently, the Nevada Operations Office plans to retain the existing inventory of mixed TRU waste pending acceptance of the waste at WIPP. The storage will likely remain under the Settlement Agreement until WIPP opens. This agreement resolves an alleged violation issued on November 1, 1990, by Nevada Division of Environmental Protection as a result of an August 1990, inspection alleging storage of mixed waste without a permit and lack of proper waste characterization. The agreement also resolves a finding of alleged violation issued on June 24, 1991, alleging an increase in the design capacity of the TRU storage pad without prior state approval. The NTS Waste Analysis Plan, submitted in April 1991, as amended by Nevada Division of Environmental Protection, outlines the characterization procedures to be used as a result of the agreement. As a result of this Settlement Agreement DOE will: Limit mixed TRU waste storage at Area 5 to the current inventory of approximately 800 yd3 (150,000 gallons). Obtain a Nevada hazardous waste permit before additional mixed TRU wastes are to be stored at NTS. Document why the current inventory of mixed TRU waste cannot be removed until after WIPP becomes operational. Report annually on its progress in certifying that the TRU waste stored at NTS meets the WIPP Waste Acceptance Criteria. Operate the Area 5 TRU waste storage pad until the waste is removed in accordance with 40 CFR 265, Subpart I. Obtain approval for, and construct, a cover for the waste. A Consent Decree between the State of Nevada and Reynolds Electrical & Engineering Company, Inc., was filed on August 24, 1992, regarding alleged inadequate sampling of pondcrete received from the Rocky Flats Plant (now known as the Rocky Flats Environmental Technology Site) to be disposed at NTS in the Radioactive Waste Management Site. The resolution was a monetary payment to Nevada and the submission of a schedule for development of a sampling plan for the adequate characterization of the remaining pondcrete to be disposed at NTS. The other Radioactive Waste Management Site is the Area 3 Bulk Waste Management Facility which accepts bulk LLW that cannot be packaged for disposal at the Area 5 Radioactive Waste Management Site. Much of the waste material buried there is contaminated soil and metal remaining onsite from the atmospheric testing of nuclear weapons at NTS. The materials are deposited in two waste management units, each consisting of two subsidence craters with the area between the craters excavated to form a large oval disposal unit. In 1987, a DOE task force determined that underground nuclear device testing areas are subject to the provisions of CERCLA. Under CERCLA all releases of hazardous or extremely hazardous substances that exceed reportable quantities must be reported to the National Response Center. Preliminary Assessment/Site Investigation reports required by CERCLA were prepared and provided to the EPA in 1988. The contaminants of concern at NTS are the results of historic aboveground and underground testing, the disposal of incidental wastes generated in support of the testing operations, and nuclear rocket experiments. Prior to 1963, soils contaminated during the atmospheric nuclear testing were consolidated and disposed of in fill areas. Soils and equipment contaminated with radionuclides as a result of drillback or tunnel operations were also disposed of in the drill areas at NTS. The types of possible contaminants found on the surface include: radionuclides; organic compounds; metals such as beryllium, lead, and hydrocarbons; and residues from plastics, epoxy, and drilling muds used during test boring drilling and instrumentation. Soils contaminated by plutonium are also a concern. EPA will use the revised Hazard Ranking System to determine if any NTS sites are to be included on the NPL. If NTS is placed on the NPL, then DOE anticipates entering into an interagency agreement with the State of Nevada and EPA. -TSAR_DOE_SECTION- A.1.3 Oak Ridge Reservation TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions A.1.3 Oak Ridge Reservation Site Description. ORR consists of approximately 34,700 acres of Federal-owned lands located directly to the west and south, but within the incorporated city limits of Oak Ridge, TN. The city of Oak Ridge and ORR are within the region known as the Great Valley of the Tennessee River, which lies between the Comburant and Southern Appalachian mountain ranges. About 10 miles to the northwest, the Comburant mountains rise to an elevation of 3,000feet or more while the Great Smoky Mountains National Park reaches to heights over 6,600 feet some 70 miles to the southeast. The largest city in the area, Knoxville, is located approximately 30 miles to the southeast. Land use in the five-county area surrounding ORR varies from the heavily populated and highly developed urban areas around Knoxville, to the sparsely populated areas immediately surrounding ORR. The largest single land use for each of the five counties is forestry; the second most common use is agriculture. DOE has three primary complexes within ORR. These are the Y-12 Plant (Y-12), the K-25 Site (formerly the Oak Ridge Gaseous Diffusion Plant), and the Oak Ridge National Laboratory. Originally built in the early 1940's for large-scale production of fissionable material for the world's first nuclear weapon, they continue to be used today as research, development, and manufacturing facilities (figure 4.4.1-1 and table 4.4.2.2-1). Y-12 is situated on 811 acres in the eastern end of ORR in an area known as Bear Creek Valley. The primary missions of Y-12 include dismantling nuclear weapon components returned from the national arsenal, maintaining nuclear production capability and stockpile support, and providing storage for special nuclear materials. Y-12 also supports other Federal agencies through a Work for Others program. In addition, a technology transfer program has been established to support the U.S. industrial base by applying Y-12 expertise to a wide range of manufacturing problems. All of the uranium parts used in building U.S. nuclear weapons were fabricated at Y-12. The plant itself consists of 494buildings containing more than 7 million ft2 of floor space. Although the primary mission of K-25 has been to provide enriched uranium for U.S. nuclear weapons and, later, an industrial toll enrichment service by which uranium is enriched for use in power reactors around the world, the gaseous diffusion process at K-25 designed for that function was placed in permanent shutdown in 1987, because of a lack of weapons or commercial requirements. Today, K-25 serves as an operations center for environmental res- toration and waste management programs. The site is also the home of DOE's Center for Environmental Technology and Center for Waste Management. Oak Ridge National Laboratory programs focus on basic and applied research, technology development, and technology that has been designated important to DOE and the Nation. It also performs work for non- DOE sponsors when such activities complement DOE missions and address significant national or international issues. In addition, Oak Ridge National Laboratory supplies radioactive and stable isotopes that are not available from the private sector. The 604 buildings on ORR outside Y-12 contain more than 17 million ft2 of floor space. Most of these buildings and structures are located within K-25 or Oak Ridge National Laboratory; however, several buildings and structures owned or leased by DOE are outside ORR. The onsite buildings and structures outside the major plant sites consist of the Scarboro Facility, the Central Training Facility, the Transportation Safeguards Division Maintenance Facility, and some ancillary structures. Most physical facilities used by the various plant protection and security groups are within the primary plant's fenced area; however, the target ranges are outside the fence but within the buffer zones of the main plant areas. Small-arms ranges are located on the east end of Y-12 and north of the west end of Oak Ridge National Laboratory. The Scarboro Facility, located within ORR and south of Y-12, houses the Oak Ridge Institute for Science and Education's Medical and Health Sciences Division's Large Animal Research Program. The Central Training Facility is shared by the site security force; DOE's Transportation Safeguards Division; and other contractor and agency security personnel. This facility also consists of a small office building, an indoor firing range, classroom and storage trailers, onsite parking, fitness facilities, and numerous outdoor firing ranges. The site, located less than 1mile southeast of K-25, currently consists of approximately 140 acres including a buffer area. The Transportation Safeguards Maintenance Facility is the former Stone & Webster warehouse, located about 1 mile east of K-25. The offsite buildings and structures consist of the Oak Ridge Operations Office, the DOE Office of Scientific and Technical Information, the Oak Ridge Institute for Science and Education facilities, the American Museum of Science and Energy, the prime contractor's "Townsite" facilities, the National Oceanic and Atmospheric Administration's Atmospheric Turbulence and Diffusion Laboratory, and others. With the exception of the Federal Office Building and space leased from the private sector, all buildings and structures used for DOE functions are situated on DOE-owned land. Missions. The missions of ORR include: Maintain capability to fabricate uranium and lithium components and parts for nuclear weapons. Store uranium and lithium materials and parts. Dismantle nuclear weapons components returned from the stockpile. Process uranium, some of which is subject to International Atomic Energy Agency storage requirement provisions. Provide support to design agencies as requested. Perform waste management and D&D activities at Oak Ridge National Laboratory, Y-12, and K-25. Operate the Oak Ridge National Laboratory to perform basic research and development in energy, health, and environment and to produce radioactive and stable isotopes not available elsewhere. Sponsor Oak Ridge Institute for Science and Education programs in the areas of health, environment, and energy. Perform projects to support other Federal programs. Maintain programs to transfer unique technologies developed at Oak Ridge to private industry. Perform meteorological and atmospheric diffusion research. Facility Operations. The complexes at ORR are managed by a management and operations contractor, under a contract administered by the Oak Ridge Operations Office. Current missions and functions can be grouped into the following categories: defense program activities; DOE Office of the Assistant Secretary for Environmental Management (EM) activities; other DOE activities; and non-DOE activities. Defense Program Activities. All defense program activities at ORR are conducted within Y-12. The site supports Nuclear Weapons Production and Surveillance and Nuclear Materials Production mission assignments housed in approximately 425 buildings and utilizing some 5.4 million ft2 of floor space. Y-12 also stores depleted uranium. Another important mission of Y-12 is the processing of uranium. Uranium materials are also recovered from the fabrication process and the disassembly of retired weapons. In addition to its functions related to uranium materials, Y-12 performs precision machining and assembles components, provides fabrication support to DOE's weapon design laborato- ries, and produces components for design evaluation for these customers and most of the test devices used at NTS. Other Department of Energy Activities. Other DOE activities conducted at ORR include missions and programs of K-25, Oak Ridge National Laboratory, Y-12 nondefense programs, the Oak Ridge National Environmental Research Park, the Oak Ridge Institute for Science and Education, and the American Museum of Science and Energy. K-25 contains approximately 1,700 acres and is located 6miles northwest of Y-12. The site consists of 250 buildings with approximately 12.2 million ft2 of floor space. The original mission of K-25 was to separate uranium-235 for use in atomic weapons. In December 1987, DOE permanently shut down the gaseous diffusion processes and K-25 was placed on the list of facilities slated for decontamination and decommissioning. Today, K-25 serves as the operations center for Environmental Restoration and Waste Management Programs. It is also the home of DOE's Center for Environmental Technology and Center for Waste Management. Missions and activities include technology development, technology transfer, engineering technology, uranium enrichment support, and the central functions of business management, engineering, computing and telecommunications. The Oak Ridge National Laboratory complex consists of approximately 2,900 acres located 4 miles southwest of Y-12. The site has approximately 240buildings containing 2.7 million ft2. Activities at Oak Ridge National Laboratory include basic and applied research, technology development, and other technology important to DOE and the Nation. Oak Ridge National Laboratory also performs research and development for non-DOE sponsors when such activities complement DOE missions and address significant national or international issues. Missions and activities include energy production and conservation technologies, physical and life sciences, scientific and technological user facilities, environmental protection and waste management, science and technology transfer, and education. Oak Ridge National Laboratory also designs and provides research facilities for the scientific and technical community and supplies radioactive and stable isotopes that are not available from the private sector. Major portions of ORR are used by Oak Ridge National Laboratory in their aquatic habitat, flora, fauna, and other environmental sciences research programs. In addition to defense program activities described above, Y-12 provides processing of radioactive source materials and support for other government agencies. Some 47 buildings containing 1.5 million ft2 located on Y-12 grounds are utilized by Oak Ridge National Laboratory in support of nondefense program missions. Oak Ridge National Laboratory employs some 450 people at Y-12. Also located on the Y-12 site are approximately 20 buildings containing 300,000 ft2 which house the DOE construction manager, the water plant maintenance contractor for ORR, and several organizations of the Oak Ridge Operations Office. Employment in these activities include 175 in DOE and 550 in construction manager organizations. The Oak Ridge National Environmental Research Park, established in 1980, consists of 13,590 acres on the ORR. As one of seven DOE research parks, its purpose is to provide protected land areas for research and education in the environmental sciences and to demonstrate that energy technology development is compatible with a quality environment. There are 53 active Environmental Sciences Research Sites consisting of 3,562 acres on ORR. In addition, there are 15 inactive sites on a total of some 323 acres. Oak Ridge Institute for Science and Education's primary missions are to provide educational and research programs in the areas of health, environment, and energy for DOE, other federal agencies, and private industry. The American Museum of Science and Energy is located at a site contiguous to the Oak Ridge Institute for Science and Education campus. The museum contains historical displays and exhibits about energy in its various forms as well as topical matter on the growth of the nuclear power industry. Non-Department of Energy Activities. Non-DOE activities pursued at ORR include missions and programs of the National Oceanic and Atmospheric Administration which conducts meteorological and atmospheric diffusion research that is supported by both itself and DOE. This work is done at the Atmospheric Turbulence and Diffusion Laboratory and field sites on ORR. The laboratory also provides services to DOE contractors and operates the Weather Instrument Telemetering Monitoring System for DOE. Environmental Regulatory Setting. ORR consists of three separate sites in and around the city of Oak Ridge; however, all Federal and state environmental agreements deal with ORR as a single entity. The State of Tennessee, which has regulatory authority for air, water, solid waste, hazardous waste, and mixed waste, entered into a 5-year Monitoring and Oversight Agreement with DOE on May 13, 1991, to assure Tennessee citizens that their health, safety, and environment are being protected during ORR facility operations. The Tennessee Department of Environment and Conservation is the lead state agency for implementation of this agreement in which DOE provides financial support to allow Tennessee to carry out its commitment under the Monitoring and Oversight Agreement and the Federal Facility Agreement regarding cleanup activities. In addition, ORR performs its own environmental monitoring of effluents and surveillance of the environmental media to characterize and quantify contaminants, assess radiation exposures of members of the public, demonstrate compliance with applicable standards and permit requirements, and assess the effects, if any, on the local environment. The Monitoring and Oversight Agreement also ensures that DOE complies with all applicable laws, regulations, and orders. The Department is working with Federal and state regulatory authorities to address compliance and cleanup obligations arising from its past operations at ORR. The Department is engaged in several activities to bring its operations into full regulatory compli- ance. These activities are set forth in negotiated agreements that contain schedules for achieving compliance with applicable requirements, and financial penalties for nonachievement of agreed upon milestones. This section discusses the more important agreements and other regulatory issues that must be considered before making a tritium supply decision that would affect ORR. On December 21, 1989, EPA placed ORR on the NPL as a "Superfund Site" pursuant to the provisions of CERCLA. This determination was based on the contamination present due to past practices. Air. Regulation of radionuclide air emissions at ORR is governed by the Federal Facility Compliance Agreement for CAA (Rad-NESHAP) signed May 26, 1992, and the Radionuclide National Emissions Standards for Hazardous Air Pollutants (NESHAP) Compliance Plan for ORR, dated April 17, 1991. The National Emissions Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities (40 CFR 61, Subpart H) requires sampling and reporting to demonstrate compliance with the 10mrem per year effective dose equivalent standard. Continuous emission sampling is required for any emission point with the potential to cause a dose exceeding 0.1 mrem per year. During 1992, 65 of 81 continuously monitored stacks were judged to have the potential to emit radioactive effluents that contribute greater than 0.1 mrem per year effective dose equivalent to an offsite individual. On March 26, 1993, EPA Region IV certified that DOE had completed all of the actions required by the ORR Rad-NESHAP Federal Facility Compliance Agreement and is considered to be in compliance with the Rad-NESHAP regulations. The annual offsite dose to the maximally exposed member of the public for ORR was 1.4 mrem in 1992, well below the 40 CFR 61 standard of 10 mrem per year (ORDOE 1993a:xxxviii). Water. A DOE water treatment facility supplies potable water to the city of Oak Ridge and other ORR facilities. The water is treated, chlorinated, and fluoridated before distribution and meets health standards. Activities are underway to reduce discharges of priority pollutants, high temperature water, and toxic agents such as chlorine to the East Fork Poplar Creek. Two dechlorination systems were installed in late 1992, at key outfalls on East Fork Poplar Creek to help control discharges of chlorine from noncontact cooling water systems and to help to eliminate chronic fish kills in the upper reaches of the creek. Additional efforts relating to reducing non-point-source pollutants to surface streams and cleaning up mercury pollution in the East Fork Poplar Creek are being planned. NPDES permits are required for each ORR facility. Y-12 is operating at the standards of the permit which expired in May 1990, and submitted a renewal in November 1989, with an addendum submitted in January 1993. Oak Ridge National Laboratory is operating at the standards of the permit which expired in April 1991, and submitted a renewal on October 28, 1990. Oak Ridge National Laboratory submitted a request to the state for modification of its NPDES permit based on evidence that past exceedance of total suspended solids, oil, and grease limits in the past have not impacted watershed water quality. A renewed NPDES permit was issued to K-25 on October 1, 1992, that will eliminate many of the permit exceedances which occurred repeatedly under the expired permit, but were not related to discharges from K-25. The new permit requires monitoring of storm drain discharges into settling ponds and streams, not at the outlets of these streams and ponds as did the expired permit. Personnel at Y-12 operate a sanitary collection sewer system. Sanitary wastewater is discharged to the city of Oak Ridge under an industrial pretreatment permit. A new monitoring station was completed January 16, 1993, which allows for more accurate monitoring of the sanitary sewage discharges by Y-12. In addition to the sanitary sewer system, ORR submitted individual stormwater permit applications in October 1992, for all three sites, as required by the Clean Water Act (CWA) and Tennessee regulations. ORR had been performing extensive sampling since June 1991, in preparation for these applications. On January 17, 1992, Friends of the Earth, a nonprofit corporation, filed a lawsuit against DOE in Federal District Court in Knoxville, TN. The lawsuit alleges that DOE is violating the NPDES permits because discharges of certain quantities of various pollutants into tributaries of the Clinch River have exceeded the allowable discharge limits of the NPDES permits. Friends of the Earth filed a motion for summary judgement in October 1992, and DOE filed a cross-motion for denial of summary judgement in January 1993. Land. To satisfy the requirement under Section 104 of CERCLA for an interagency agreement, DOE, EPA Region IV, and Tennessee completed a Federal Facility Agreement effective January 1, 1992. The purposes of the Federal Facility Agreement are to: establish a procedural framework and schedule for developing, implementing, and monitoring appropriate response actions at ORR in accordance with CERCLA, RCRA, NEPA, appropriate guidance and policy, and Tennessee state law; coordinate future assessments and most of the remedial action activi- ties planned at ORR pertaining to environmental restoration activities under CERCLA with state laws and existing corrective actions required under the RCRA permit issued to DOE for ORR effective on October 25, 1986; minimize the duplication of investigations, analytical work, and documentation; ensure the quality of data management; expedite response actions with a minimum of delay; and achieve a comprehensive remediation of the site. The Federal Facility Agreement also addresses technical standards for new and existing liquid LLW storage tank systems; conduct of remedial investigation/feasibility studies and remedial design/remedial actions in accordance with timetables for sites listed in Appendix E of the Federal Facility Agreement; setting of annual priorities; and provision of responsive guidance by regulators when requested. ORR facilities are being operated with a combination of RCRA Part B permits and interim status regulations. The RCRA Part B permit applications have been submitted for all of the active storage and treatment units listed on the Part A permit. Some are approved and other Part B applications are still awaiting action by the State of Tennessee. ORR facilities generate hazardous wastes, and personnel operate hazardous waste treatment and storage facilities. However, there are no units actively used for the disposal of hazardous wastes. Closure actions are complete for several previously used hazardous waste disposal units. DOE declared a moratorium in May 1991, on the shipment of hazardous waste to non-DOE sites to prevent waste, potentially contaminated with radioactive material, from being shipped to a facility which is not licensed to handle radioactive material. Current DOE policy on No Rad Added, which resulted from the moratorium, effectively requires all RCRA hazardous waste generated at ORR (and other DOE facilities) to be managed as mixed waste until ORR Waste Management provides sufficient evidence to convince an independent review board for each offsite shipment that selected waste streams are free of radioactivity. The No Rad Added policy, therefore, restricts the ability of ORR to ship offsite RCRA hazardous waste generated at ORR facilities. (OR DOE 1993a:xxxix). DOE and EPA signed a Federal Facility Compliance Agreement effective June 12, 1992, covering concerns related to the RCRA land disposal restrictions. This Agreement recognizes that DOE is currently storing and will continue to generate and store hazardous wastes, including mixed wastes which contain a hazardous waste component, subject to land disposal restrictions: solvent waste, California-list waste, and Third Thirds waste. The agreement requires DOE to submit for EPA review a plan for a waste minimization program at ORR that provides for segregation of hazardous wastes from mixed wastes, substitution of nonhazardous solvents for hazardous solvents where technically practicable, and the minimization of the generation of hazardous waste throughout ORR. In addition to agreeing to implement the waste minimization program, DOE agreed to submit the following additional plans to EPA for review and approval: storage of the land disposal prohibited wastes identified in the Federal Facility Compliance Agreement; the treatment method, facility, and schedule for treating the land disposal prohibited wastes with identified existing treatment; the strategy for conducting treatability studies, technology development, and prioritization of treatment method options for land disposal prohibited wastes without identified existing treatment; and the treatment method, facility, and a schedule for the completion of such treatment for wastes without identified existing treatment. This last plan is required not later than March 1995. The RCRA- Land Disposal Restrictions Federal Facility Compliance Agreement and the information provided in the required plans would form the basis for the site-specific mixed waste treatment plan required by the Federal Facility Compliance Act of 1992, for DOE facilities storing mixed wastes. The ORR underground storage tank program regulates approximately 80 tanks and includes some that are deferred or exempt from external regulation. The tanks store petroleum and hazardous substances. ORR is ahead of its schedule for upgrading and/or replacing the underground storage tanks to implement leak detection, spill and overflow protection, and corrosion protection on all regulated tanks by 1998. ORR facilities have PCBs from past practices and continue to generate PCB-contaminated wastes. These wastes are either shipped to a commercial disposal site, or if contaminated with uranium, stored until the TSCA Incinerator at K-25, which began operations in 1990, can process the waste. Some PCB-contaminated wastes have been stored in excess of 1 year due to: specific constituents in the waste that render it unacceptable at the TSCA incinerator; the burn priority at the incinerator; or the use of other treatment methods. TSCA requires PCB-contaminated wastes to be disposed of within 1 year of its initial placement in storage. Due to its radioactive nature, treatment and disposal technology does not exist for most of these wastes. DOE has an aggressive program to develop this technology, but until it is available, these PCB- contaminated wastes must be safely stored, although not in compliance with TSCA. On June 11, 1992, DOE formally requested negotiation of a Federal Facility Compliance Agreement with EPA that would allow development of an ORR treatment and disposal schedule for radioactive PCB-contaminated waste and storage or disposal per the Agreement. -TSAR_DOE_SECTION- A.1.4 Pantex Plant TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions A.1.4 Pantex Plant Site Description. Pantex is located in the panhandle of Texas, in Carson County. It is about 17 miles northeast of downtown Amarillo and 40 miles southwest of Pampa. The plant is located on a 10,000-acre portion of the former Pantex Army Ordnance Plant. Pantex was constructed in the first half of the 1940's by the U.S. Army for the production of conventional ordnance. At the end of World War II, the plant was deactivated and the property eventually reverted to the War Assets Administration. In 1949, the entire installation was sold to Texas Technological College (now Texas Technological University) for one dollar. The land was to be used for experimental farming, but was subject to recall under the National Security Clause. Following an extensive survey of World War II ordnance plants, Pantex was chosen in 1951, by the Atomic Energy Commission for expansion of its nuclear weapons assembly facilities. The Army Ordnance Corps reclaimed the site for the Atomic Energy Commission and contracted a civilian contractor to rehabilitate it. Pantex consists of 425 buildings containing approximately 2,500,000 ft2 of floor space. Within the 10,000 acres, approximately 2,000 acres are dedicated to active facility operations. Approximately 8,000 acres are devoted to storage, disposal, and miscellaneous activities in support of plant operations. In addition, about 5,800 acres are leased from Texas Technological University (Texas Tech) to provide a Government-controlled safety and security zone on the south side of the facility. The buffer zones are used by Texas Tech for agricultural research with some acreage leased for private farming. An additional 1,080 acres northeast of the facility provide supplemental water rights and include a playa formerly used for the disposal of treated sewage effluent. Plant acreage also contains some 47 miles of paved roads, 17 miles of railroad tracks, 67 miles of fence, and 4 playas. All the land within a 3-mile radius of the plant site is used for agricultural purposes, either farming or grazing. Approximately 2,000 people live within 5 miles of the outside boundary of Pantex. A significant population concentration occurs southwest of the Pantex facility near the Amarillo International Airport and includes the Texas State Technical Institute and the Highland Park Village. Highland Park Village consists of 500 single- and multiple-family housing units (duplexes) with an occupancy rate averaging about 90 percent. Approximately 100 students are housed in a Texas State Technical Institute student dormitory. The main mission of Pantex is the manufacture of HE components and the assembly and disassembly of nuclear weapons. Activities in support of the mission are conducted in portions of the facility referred to as zones. The major zones are shown in figure 4.5.1-1 and Pantex baseline characteristics are summarized in table 4.6.2.2-1. Almost 3,000 personnel support Pantex's mission. This includes area office personnel from DOE, Sandia National Laboratories, employees actively working in quality evaluation, the courier section personnel tasked with weapon transportation, and contractor employees that operate the plant. Plant operation includes direct and support manufacturing operations, management and administrative services, protective services, and maintenance and utilities. During normal operations, approximately 600 people are assigned primarily to the Zone12 South Material Access Area. Half of these individuals routinely handle the radioactive components that are used in weapon production. The Material Access Area is a highly restricted area where special nuclear materials are staged and assembled. Missions. Pantex missions include: Manufacture HE components for use in nuclear weapons. Assemble nuclear weapons. Retrofit, maintain, and repair nuclear weapons in the stockpile. Provide quality assurance evaluations of nuclear weapon systems. Assemble nuclear weapon-like devices for testing and training programs. Disassemble nuclear weapons for disposal or maintenance purposes. Dismantle nuclear weapons no longer required. Provide development support to weapons design agencies as requested. Store strategic reserve quantities of plutonium. Facility Operations. DOE directs all operations conducted at Pantex. Current missions and functions can be grouped into broad categories of defense program and environmental management activities. No other missions or functions are expected to be added or removed except as described in section 4.5.1 under tritium supply and recycling alternatives for Pantex. Defense Program Activities. The main missions of Pantex are the manufacture of HE components and the assembly/disassembly and maintenance of nuclear weapons. As previously stated, activities in support of the mission are conducted in portions of the facility referred to as zones. The principal operations performed at Pantex are the assembly of nuclear weapons from components received from other DOE facilities; fabrication of chemical HE components for nuclear weapons; operation of the chemical HE synthesis and characterization group; surveillance testing and disposal of chemical HE; disassembly of obsolete nuclear weapons for retirement; and maintenance, modification, repair, and nonexplosive testing of nuclear weapons components. Weapons assembly and stockpile surveillance activities involve handling significant quantities of uranium components, plutonium components, and tritium components, as well as a variety of nonradioactive toxic chemicals. Brief descriptions of all the above mission activities follow. HE component production includes manufacturing main charge subassemblies and mock components for use in weapon test assemblies, manufacturing small HE components, producing a variety of explosive materials from chemical reactants and commercially-produced explosives, and evaluating explosive materials and components through a variety of analytical, mechanical, and explosive tests. New production is defined as the final assembly of a new nuclear weapon to be added to the stockpile. Pantex receives weapons components and other materials from throughout the Complex. The first step in the new production process is the mating of the HE main charge subassemblies and the special nuclear materials, which takes place within an assembly cell. Assembly bays house the remainder of the assembly process. This is where the nuclear subassembly produced in the assembly cell is built into a complete weapon. After final assembly, the items produced at Pantex are shipped either to other facilities within the DOE Complex or to military facilities by truck. The tasks of modification, maintenance, and repair involve the disassembly of a stockpiled nuclear weapon so that one or more of the components can be repaired, replaced, or modified. After replacing the components, the weapon is reassembled and returned to the stockpile. Activities of this type can be placed into one of the following categories: Major Modification-Replacement or modification of such components as the HE charges, the firing sets, or the electrical subassemblies. Limited-Life Component Exchange- Parts that must be periodically replaced. Some systems require extensive disassembly to accomplish this task. Pantex is responsible for disassembling these weapons, replacing the limited-life component and returning the weapons to the stockpile. Repairs-Weapons that have sustained damage. An example of such repair is the replacement of mechanical or electrical parts damaged by military users in field locations. Pantex performs many quality assurance evaluation activities on both new and stockpiled nuclear weapons. These tests involve the disassembly of a weapon, the laboratory testing of various components, and the rebuilding of the weapon for shipment back to the stockpile. Five evaluations are performed at Pantex: New Material Laboratory Testing, New Material Flight Testing, Stockpile Laboratory Testing, Stockpile Flight Testing, and Accelerated Environmental Aging and Materials Compatibility Testing. These evaluations are outlined below: New Material Laboratory Testing-Disassembly of a randomly selected newly-produced weapon before it is shipped to the stockpile. Various components are subjected to either destructive or nondestructive tests. After testing, the weapon is rebuilt and shipped to the stockpile. New Material Flight Testing-Similar to New Material Laboratory Testing. Units are selected at random before delivery to the stockpile and may be assembled into a nonnuclear, usually nonexplosive, joint test assembly for flight testing. These assemblies are tested aboard aircraft to verify the functioning of the components under in-flight conditions. After the test flight, the unit is returned to Pantex for further examination when possible. Stockpile Laboratory Testing-Similar to the New Material Laboratory Testing, but with two major differences. First, Stockpile Laboratory Testing is performed on units randomly selected from the stockpile. Second, some weapons selected for Stockpile Laboratory Testing are not rebuilt after the test, but are disassembled and the components destructively tested to ensure that DOE has an accurate estimate of system reliability. Stockpile Flight Testing-Has characteristics of both Stockpile Laboratory Testing and New Material Flight Testing. Components from the selected stockpiled unit are assembled into a joint test assembly for in-flight testing, and undergo similar post-test examination. Accelerated Environmental Aging and Materials Compatibility-Determines the effects of aging on the integrity of weapon systems over time. These tests involve subjecting newly-produced units to an artificial aging process or to environmental stresses to determine whether or not they retain their chemical and physical properties, and to ensure that they will react in a predictable manner after an extended period of time. Also, some testing is performed at the Gas Analysis Laboratory, which evaluates samples taken from accelerated aging units, material compatibility tests, development activities, material certification tests, and production operations. Nuclear weapons no longer needed in the stockpile must be disassembled and disposed of; Pantex accomplishes this in one of the following ways: Quality Assurance Disposal-Weapons selected for Stockpile Laboratory Testing that are not to be rebuilt are disposed of. After the parts selected for testing have been removed, the remaining portion of the weapon is disassembled and the parts disposed of after component testing. Dismantlement Surveillance-Units selected for disposal in this manner are completely disassembled and selected parts subjected to testing. Dismantlement-Dismantlement activities consist of the disassembly of a weapon and the disposition (i.e., staging or destruction) of the components. Limited quality testing and evaluations may be performed. In addition to the primary efforts associated with weapons assembly/disassembly, disposal, and quality assurance, Pantex provides development support and services to the nuclear weapon design agencies and to other government entities, as requested. Pantex contains a number of facilities that stage (i.e., temporarily store) weapon components that are destined either for the assembly cells or for shipment back to other DOE facilities. Staging procedures may involve the leak testing of staging containers, inventory procedures to verify the number and contents of containers, and unpacking and repacking to physically verify and test contents. Assembly and disassembly activities are conducted in Zone 12. There are special nuclear materials staging areas, HE staging and production areas, and component warehouses in Zone 12. Other Zone 12 activities include electrical component testing at the Sandia Systems Test Laboratory located there, as well as administrative office functions, craft shops, the command center, fire station, and other general support facilities such as the cafeteria and sewage treatment. Stockpile maintenance, testing, disassembly, and disposal missions are conducted in Zones 5 and 12. Zone 4 West is used for staging of weapons and special nuclear materials and interim storage of plutonium pits. Zone 4 East is used for staging of HE and other material. Additional sources of support to the Pantex mission include warehousing and landfill operations in Zone 10, water wells and a water treatment plant in Zone 15, and a vehicle maintenance facility in Zone 16. The HE development area in Zone 11 consists of facilities for synthesizing and characterizing new HE. HE and HE scrap are disposed of onsite at the Burning Ground. The test firing site has several reinforced concrete bunkers containing control rooms and camera areas. Experimental HE configurations are detonated on firing pads surrounded by earthen bunkers. Selected samples of HE components are detonated for quality assurance testing. Other Department of Energy Activities. At present, no other DOE activities are pursued at Pantex. Non-Department of Energy Activities. At present, there are no non-DOE activities pursued at Pantex. Environmental Regulatory Setting. In 1989, the Secretary of Energy invited the host state of each DOE facility to independently determine and verify any plant operational impacts to the environment. In response to this initiative, DOE entered into an Agreement in Principle with the State of Texas to focus on waste management, emergency response, and environmental monitoring. DOE provides required information to the State of Texas, and the State conducts sampling and research activities. DOE also issued a Grant-in-Aid for hydrogeologic characterization studies at Pantex. On September 1, 1993, the Texas legislature created the Texas Natural Resources Conservation Commission and transferred to it the responsibilities of the Texas Water Commission and merged the Texas Air Control Board into the new agency. This agency is now organized to include the Office of Water Resource Management, Office of Air Quality, and Office of Waste Management, among others. Pantex provides office space for Commission officials who are assigned to the plant. EPA has delegated to the State of Texas regulatory authority for air and solid and hazardous waste. The Department is working with Federal and state regulatory authorities to address compliance and cleanup obligations arising from its past operations at Pantex. The Department is engaged in several activities to bring its operations into full regulatory compliance. These activities are set forth in negotiated agreements that contain schedules for achieving compliance with applicable requirements, and financial penalties for nonachievement of agreed upon milestones. On May 31, 1994, EPA placed Pantex on the NPL as a "Superfund Site" pursuant to the provisions of CERCLA. This determination was based on the contamination present due to past practices. Air. The emission of radionuclides from Pantex is regulated under the National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities (40 CFR 61, Subpart H) by EPA. The standard is that level of emissions of radionuclides that would cause any member of the public to receive in any year an effective dose equivalent of 10 mrem or less. The effective dose equivalent to any member of the public from emissions of radionuclides from Pantex in 1993 was less than 1 percent of 10 mrem. To demonstrate compliance with the 40 CFR 61 standard, Pantex performed periodic confirmatory monitoring as prescribed in 40 CFR 61.93 for DOE facilities that emit less than 1 percent of the 10 mrem per year allowablelimit. The Burning Ground, where explosives, explosive components, and explosive-contaminated materials are demilitarized as required by the Atomic Energy Act, operates as a RCRA Interim Status Unit and under a written Grant of Authority from the Texas Natural Resources Conservation Commission. The Burning Ground is also used to thermally treat explosive waste and explosive-contaminated waste. The Texas Natural Resources Conservation Commission has indicated its desire to modify the provisions of this authority. Discussions on the terms of the proposed permit modification between the State of Texas, DOE, and parties to the hearing process are ongoing. Water. The city of Amarillo operates a major water supply well field immediately north and down gradient of Pantex. Pantex also receives its drinking water from the Ogallala aquifer via five groundwater wells located on the northeast corner of the plant. The water is treated onsite and tested in accordance with requirements for public drinking water systems. The domestic water supply at Pantex meets all of the national primary and secondary drinking water standards for noncommunity, nontransient public water supply systems. On December 20, 1992, Texas Natural Resources Conservation Commission repre- sentatives inspected the domestic water supply system at Pantex. The inspection revealed that the system is being operated and maintained in compliance with Texas statutes and regulations. Onsite monitoring wells, installed in 1990 between the former chemical burn pit and the Amarillo/Pantex water supply, have not detected any contamination. Other onsite monitoring wells have detected contamination in the upper level or perched aquifer. Pantex no longer releases aqueous waste streams containing hazardous waste or hazardous waste constituents to the land surface. EPA issued DOE a RCRA Administrative Order on Consent effective December 10, 1990, requiring Pantex to conduct a RCRA Facility Investigation to identify, assess, and correct actual and potential threats to human health or the environ- ment resulting from the release or potential release of hazardous wastes or constituents at the facility. DOE completed the RCRA facility investigation and has completed Phase I on eight work plans and Phase II on one work plan. In response to a request by the DOE, the EPA determined that an NPDES permit was applicable to operations conducted at Pantex. HE-contaminated wastewater is filtered and treated at individual building filter systems. Most of the filtered water flows into the wastewater lagoon. Since 1980, pursuant to the Texas Water Code, the Texas Natural Resources Conservation Commission has permitted Pantex to discharge its wastewater under a no discharge wastewater permit (Permit Number 02296). No discharge permits allow wastewater discharges not going to "Waters of the State." In 1980, the state did not consider playas to be "Waters of the State." On December 26, 1990, the DOE filed a permit application to modify its permit. The permit application was resubmitted in May 1992, at the request of the Texas Natural Resources Conservation Commission to change the permit from a no discharge to a discharge permit. Land. Pantex is registered with the State of Texas (Texas Natural Resources Conservation Commission Solid Waste Registration Number 30459) and operates under a hazardous waste permit HW-50284 and EPA Identification Number TX4890110527. Pantex currently operates units both under its hazardous waste permit and under interim status. On April 25, 1991, the Texas Natural Resources Conservation Commission and EPA issued a permit to Pantex to store containers and tanks and to treat hazardous waste in tanks. The permit specifically excluded 17 RCRA units at the Burning Ground, but continued the interim status of those units. Pantex thermally treats explosive waste and explosive-contaminated waste at the Burning Ground. The Burning Ground operates as a RCRA Interim Status Unit and under a written Grant of Authority from the Texas Natural Resources Conservation Commission. In November 1991, the DOE formally submitted a request to the Texas Natural Resources Conservation Commission for a Class 3 Modification to add the units at the Burning Ground to the permit. Pursuant to the public notice published on October 31, 1991, interested parties requested a hearing before a hearing examiner on the permit. On May 31, 1992, the Texas Natural Resources Conservation Commission Office of Air Quality, recommended draft hazardous waste permit provisions for the Burning Ground. The Texas Natural Resources Conservation Commission, the DOE, and parties to the hearing process are continuing discussions on terms of the proposed permit modification. Pursuant to the requirements of the Federal Facility Compliance Act of 1992, DOE prepared and submitted to the EPA and Texas Natural Resources Conservation Commission an inventory of all mixed waste stored at Pantex on April 23, 1993. Pantex also submitted a Conceptual Site Treatment Plan to the Texas Natural Resources Conservation Commission in October 1993, as required by the Federal Facility Compliance Act of 1992. The Final Site Treatment Plan is to be submitted to the State of Texas by October 1995. One area of alleged noncompliance resulting from the Texas Natural Resources Conservation Commission RCRA inspection in July 1992, is the storage of mixed waste for a period greater than 1 year, which is prohibited by the land disposal restrictions. Pantex is addressing the issue of mixed waste storage in its Site Treatment Plan as required by the Federal Facility Compliance Act of 1992. The Pantex solution will be finalized in late 1995 upon acceptance by the state of the Site Treatment Plan and state agreement to a resulting Consent Order. As of December 31, 1993, all equipment and parts used at Pantex that contained PCBs had concentrations of less than 50 parts per million (ppm); thus, most requirements of the TSCA do not apply. TSCA regulated wastes, including asbestos and materials contaminated with less than 50 ppm PCBs, are transported to permitted facilities for treatment and disposal. -TSAR_DOE_SECTION- A.1.5 Savannah River Site TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.1 Reference Operating Assumptions A.1.5 Savannah River Site Site Description. SRS is located 12 miles south of Aiken, SC, and approximately 16 miles southeast of Augusta, GA. The site is located on approximately 198,000 acres in portions of Aiken, Allendale, and Barnwell counties. First established in 1950, SRS has been involved for more than 40 years in tritium operations and other nuclear material production. Today the site contains 15 major production, service, research, and development areas, not all of which are in operation at this time (figure 4.6.1-1). The developed areas of the site account for less than 5 percent of the land use and more than 99 percent of the total capital investment. There are more than 3,000 facilities at SRS, including 740 buildings, with 5.5 million ft2 of floor area. SRS baseline character- istics are summarized in table 4.6.2.2-1. The following discussion pertains to the active facilities. The K-Reactor, the last operational reactor at SRS, has been placed into cold standby status with no provision for restart, ending all reactor operations. The other four reactors (C, L, P, and R) already were in cold standby. The separations facilities of the 200-Area (the largest being the F- and H-Canyons) are located in the center of the site. This area contains approximately 17 major facilities including waste management, defense waste processing, naval fuel materials, and tritium separations. The new Replacement Tritium Facility, is also located in the separations area. The central shops are located in the 600-Area and perform major facility construction activities, while site administration, general support, and many of the research and development facilities are located in the 700-Area. Missions. The current missions of SRS are: Tritium recycling. Processing Pu-238 for space mission requirements. Processing irradiated targets and spent nuclear fuel. Interim plutonium storage. Waste management. Environmental monitoring and restoration. Research and development at the Savannah River Technology Center. Facility Operations. SRS is conducting tritium recycling operations in support of stockpile requirements using retired limited-life components as the tritium supply source. Some plutonium remaining from the production mission is in interim storage at SRS. Defense Program Activities. SRS conducts tritium recycling operations (tritium bottle recycling) in support of stockpile requirements. The Replacement Tritium Facility assumed this function in fiscal year 1994 operating at 30 to 35 percent capacity, using tritium from returned limited-life components. The facility will also assume some of the Nonnuclear Consolidation Plan functional transfers from the Mound Plant. These functions include as a minimum Reservoir Surveillance Operations and Gas Transfer Systems. The Replacement Tritium Facility would also be able to support new-build production require- ments, assuming the tritium is available. Subject to completion of an EIS, the separation facilities, the F- and H-Canyons, will assess acceptability of operations for material stabilization until onsite backlogs of fuel and target elements are processed. Upon completion of material stabilization activities, these facilities will be shut down permanently. Other Department of Energy Activities. The Savannah River Technology Center, in 700-Area and at TNX-Area provides technical support to all DOE operations at SRS. In this role it provides for continued updating and improvement of process efficiency to help reduce costs, waste generation, and radiation exposure. The new Defense Waste Processing Facility, now in the cold test phase, is designed for waste processing, and includes a vitrification plant and incinerator. Non-Department of Energy Activities. There are several facilities and operations at SRS that deal mainly with the ecological elements of the site. These are the Savannah River Forest Station, the Savannah River Ecology Laboratory, the South Carolina Wildlife and Marine Resources Department, the Institute of Archaeology and Anthropology, and the Soil Conservation Service. The Savannah River Forest Station is an administrative unit of the U.S. Forest Service that provides timber management, research support, soil and water protection, wildlife management, secondary road management, and fire management to DOE. The Savannah River Forest Station manages 154,000acres, or approximately 80 percent of the site area. It has been responsible for reforestation and manages an active timber business. The improve- ment of the environment has seen an increase in wildlife populations. The station assists with the development and updating of site-wide land use planning and provides continual support with site layout and vegetative management. It also assists in long-term wildlife management and soil rehabilitation projects. This unit occupies buildings on 7 acres. The Senior Community Services Employment Program operated by the Department of Labor is collocated with the station. In addition, there are also offices used by those employed to harvest the timber managed by the station. The Savannah River Ecology Laboratory is operated for DOE by the Institute of Ecology of the University of Georgia. It has established a center of ecological field research where faculty, staff, and students perform interdisciplinary field research to provide an understanding of the impact of energy technologies on the ecosystems of the southeastern United States. This information is communicated to the scientific community, government agencies, and the general public. The South Carolina Wildlife and Marine Resources Department manages a program to restore turkeys to their former range throughout the state. The Institute of Archaeology and Anthropology is operated by the University of South Carolina to survey the archaeo- logical resources of the SRS. This survey is used by DOE when planning new facility additions or modifications and is referred to in the operational management of the site. A soil survey report for land use management at the site has been completed. Environmental Regulatory Setting. The South Carolina Department of Health and Environmental Control is the state regulatory agency with authority for air, water, solid waste, mixed, and hazardous waste. DOE and South Carolina signed a Memorandum of Agreement on April 8, 1985, designed to meet both parties' interests in maintaining the environmental qualities on SRS. DOE recognizes South Carolina's delegated authority under CAA, CWA, Safe Drinking Water Act (SDWA), and RCRA, and agrees to abide by South Carolina environmental laws. South Carolina recognizes that SRS has unique Federal obligations regarding the Federal budget process, NEPA requirements, and classified information. Both parties agreed to a First Amendment to this Memorandum of Agreement on May 5, 1988, stating that DOE shall comply with the South Carolina Underground Storage Tank Regulation. The Department is working with Federal and state regulatory authorities to address compliance and cleanup obligations arising from its past operations at SRS. The Department is engaged in several activities to bring its operations into full regulatory compli- ance. These activities are set forth in negotiated agreements that contain schedules for achieving compliance with applicable requirements, and financial penalties for nonachievement of agreed upon milestones. This section discusses the more important agreements and other regulatory issues that must be considered before making a reconfiguration decision that would affect SRS. On December 21, 1989, EPA placed SRS on the NPL as a "Superfund Site" pursuant to the provisions of CERCLA. This determination was based on the contamination present due to past practices. Air. SRS has 14 air quality permits covering 141point sources for air emissions. According to EPA Region IV, the air monitoring systems at SRS do not adequately provide for monitoring of radionuclide emissions from point sources as specified by National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities (40 CFR 61, Subpart H), effective December 15, 1989. This requirement states that emissions of radionuclides from DOE facilities shall not exceed those amounts which would cause any member of the public to receive an effective dose equivalent of 10 mrem per year. In order to determine compliance, DOE must provide for monitoring of radionuclide emissions from point sources with the potential to emit 0.1 mrem per yr. Therefore, DOE and EPA entered into the Federal Facility Compliance Agreement for Radionuclide NESHAP on October 31, 1991, allowing SRS to continue operations and at the same time upgrade facilities to come into compliance with the NESHAP monitoring requirements of 40 CFR 61, Subpart H. Water. SRS has 76 outfalls on NPDES permit, SC0000175, which has been administratively extended since 1988. Since 1986, the compliance rate for NPDES regulated discharges has remained above 99 percent. SRS has two other NPDES permits: SC0044903 and one general stormwater discharge permit. A new combined NPDES permit is expected to be issued in 1995. SRS has documented groundwater contamination. Concentrations of at least one of the following constituents: tritium, gross alpha or trichloroethylene (TCE) in excess of primary drinking water standards were observed in at least one monitoring well at the landfill in the first half of 1989. TCE concentrations up to 400 parts per billion (ppb) have also been detected approximately 800 feet from the SRS boundary. Quarterly status reports of its groundwater quality assessment programs for F-, H-, and M-Areas is required by Administrative Consent Order 85-70-SW dated November 7, 1985, and amended on September 14, 1988. SRS has accomplished all one-time requirements of this order and continues the routine assessment and corrective action as required by the order. DOE entered into a Settlement Agreement on February 27, 1990, to satisfy violations of discharging wastewater into the environment of the state without a permit. DOE began construction of a Sanitary Wastewater Treatment Plant that will handle effluent discharged from Outfall A008 of the A-Area Powerhouse and other discharges as well. Construction will be completed in 1995. DOE entered into two settlement agreements under the CWA on June 5, 1990, agreeing to address high temperature discharges and related fish kills. The Thermal Mitigation Settlement Agreement, dated June 5, 1990, alleges failure to comply with both the thermal requirements of the NPDES permit and Consent Order 84-4-W, dated January 3, 1984, and requires DOE to address corrective actions for any outfall which exceeds 90 F. On June 15, 1990, DOE issued the SRS Plan of Study for Minor Thermals to the state and is awaiting approval of the plan that will satisfy this agreement. The Fish Kill Settlement Agreement requires compliance with the Fish Kill Remedial Action Plan submitted by SRS on July5,1990, and commented on by the Department of Health and Environmental Control on October25,1990, to mitigate high temperature discharges from the L- and P-Reactors which may have resulted in fish kills. On December 12, 1991, South Carolina approved the SRS request to delay remedial actions at Pond C of Par Pond and at the L-Lake due to changes in status of the P-Reactor. SRS was not able to comply fully with thermal discharge limitations contained in the NPDES permit and, therefore, conducted a Thermal Mitigation Study. The study showed a need for major construction. South Carolina issued Consent Order 84-4-W which was later amended to require construction of the K-Reactor mitigation alternative not later than December 31, 1992, and to submit a 316(a) plan of study for the D-Area coal-fired facility. The 316(a) study was submitted and is considered conditionally complete. On December22, 1992, SRS submitted documentation to the state that the K-Reactor mitigation alternative (the cooling tower) had been completed and would allow SRS to discharge K-Reactor cooling effluent in compliance with laws and permit requirements for SRS. Completion of the cooling tower accomplished one of the last thermal mitigation requirements in Consent Order 84-4-W. However, in March 1993, DOE placed the K-Reactor in cold standby with no provision for restart. Land. EPA identified approximately 150 potential operable units and 20 RCRA-regulated units on SRS. In accordance with Section 120 of CERCLA, DOE entered into a Federal Facility Agreement effective August 16, 1993, with the EPA and South Carolina to coordinate cleanup activities at SRS under one comprehensive strategy. The plan also directs interim status corrective action for releases of hazardous waste or hazardous constituents. The Federal Facility Agreement expands the ongoing RCRA Facility Investigation Program Plan under RCRA Sections 3004(u) and 3004(v) to include those releases as a CERCLA cleanup which are not already included in the RCRA Hazardous Waste Management Act permit. The purposes of the Federal Facility Agreement are to: Meet the requirements of CERCLA for an interagency agreement and assure compliance with Federal and state hazardous waste laws for matters covered by this agreement. Provide for state involvement in the initiation, development, selection, and enforcement of corrective/remedial actions. Identify operable units for investigation and possible corrective/remedial action. Establish requirements for the performance of investigations to determine fully the nature and extent of the threat to the public health or welfare. Identify the nature, objective and schedule of response actions to be taken at SRS; establish a basis for a determination that the DOE has completed the remedial actions required. Establish requirements for the SRS high-level radioactive waste tank system to ensure structural integrity, containment and detection of releases. SRS has further agreed in the Federal Facility Agreement to ensure structural integrity, containment and detection of releases, and source control for the SRS high-level radioactive waste tank systems identified in Appendix B of the agreement. SRS will also develop a process to review the risk associated with each of the regulated single-wall storage tanks on a tank-by-tank basis. This process will result in a tank removal or abandonment program which is protective of human health and the environment and which is in compliance with the provisions of RCRA (40CFR 264, Subpart J), the agreement, and all other applicable regulations. According to the Waste Removal Plan and Schedule submitted to the state on November 15, 1993, the schedule to remove waste from the tanks that do not meet secondary containment requirements is the year 2028. Radioactive waste is contained in 23 of the underground storage tanks. Further construction of the single-wall type tank for storage of radioactive waste is prohibited by the Federal Facility Agreement. SRS must meet storage tank/container standards in 40 CFR 264 and provide a schedule for submitting complete RCRA permit applications for new facilities to come on line. Schedules have been agreed upon within the Federal Facility Agreement to provide compliance through construction of treatment facilities and other measures. The Federal Facility Compliance Agreement signed by EPA and DOE on March 13, 1991, addresses SRS compliance with the Land Disposal Restrictions of the Hazardous and Solid Waste Amendments of 1984 pertaining to past, ongoing, and future generation of mixed wastes (mostly solvents, dioxin and California list wastes contaminated with radioactive tritium from rags and wipes used to remove radioactive-contaminated and nonradioactive solvents and wastes from equipment), that are prohibited from land disposal without treatment to standards. This Federal Facility Compliance Agreement supersedes the July30, 1987, Federal Facility Compliance Agreement for Past Disposal at the Mixed Waste Management Facility and incorporates any remaining provisions in addition to the new requirements. In this agreement, the parties agreed that the high-level mixed waste in the tank farms must be allowed to decay to an appropriate level of radioactivity before processing at the Defense Waste Processing Facility and that such aging shall be considered an accumulation of such quantities of hazardous waste as is necessary to facilitate proper treatment, recovery or disposal so long as the other requirements contained in 40 CFR 268.50 are met. The Federal Facility Compliance Agreement allows SRS to continue to operate, generate, and store mixed wastes subject to the Land Disposal Restrictions and outlines a plan for mixed waste compliance that, when approved by South Carolina, would allow SRS to come into compliance with the mixed waste storage provisions of the Federal Facility Compliance Act of 1992. In return, SRS will report to EPA the characterization of all solid waste steams disposed in land disposal units at SRS, will identify within 60 days any new hazardous waste streams introduced, and will submit a plan for waste minimization to EPA for review. DOE submitted the required waste minimization plan on time. SRS is also required to segregate restricted wastes, excluding those previously located on the TRU storage pads, from other mixed wastes and report on a recurring basis the identity, quantity, and compliance problems associated with restricted mixed wastes and the impact of Land Disposal Restrictions requirements. Under the provisions of the Federal Facility Compliance Agreement, SRS agreed to report construction and operational data and waste processing milestones for the Consolidated Incineration Facility, Defense Waste Processing Facility, and other hazardous waste storage, treatment, and disposal facilities planned to be built at SRS. This reporting is required because their use is key to long-term compliance with Land Disposal Restrictions for mixed waste. SRS is required to prepare TRU waste for shipment to WIPP, but the agreement allows continued storage of certified TRU waste at the TRU waste storage pads until its suitability has been determined pursuant to the provisions of 40 CFR 191 and 40 CFR 268 and it can be shipped to WIPP. In addition, DOE agreed to provide EPA with an annual progress report on WIPP, and this agreement must be modified if DOE determines that no TRU waste will be shipped from SRS by July 30, 1999. Finally, SRS agreed to provide hazardous waste training to EPA Region IV employees. On April 24, 1992, the agreement was amended to include Third Thirds restricted wastes and an alternative treatment strategy for M-Area waste. The periodic reporting requirements of the original agreement were also amended. On September 30, 1987, the South Carolina Department of Health and Environmental Control issued the RCRA Part B Permit (RCRA SC1890008989) that granted final status to four hazardous waste storage buildings and approved a postclosure plan for the M- Area Settling Basin and vicinity. The RCRA permit requires SRS to assess and, where necessary, take appropriate corrective action for any releases of hazardous wastes. Additional hazardous waste facilities, subject to RCRA interim status when the permit was approved, have been added to the RCRA permit. In January 1990, DOE notified the South Carolina Department of Health and Environmental Control that rags and wipes used with F-listed solvents for cleaning and for radioactive decontamination had been disposed in portions of the Low-Level Radioactive Waste Disposal Facility and in the SRS Sanitary Landfill. The Department of Health and Environ- mental Control advised DOE and SRS that certain solvent rags and wipes constituted hazardous waste subject to South Carolina Hazardous Waste Management Act regulation. As a result, DOE and South Carolina signed Settlement Agreement 91-51-SW effective August 26, 1991, to satisfy South Carolina allegations that SRS failed to comply with interim status standards for operation of the Low-Level Radioactive Waste Disposal Facility and the sanitary landfill in the disposal of spent solvent wastes without having applied for, or received, a permit for operation of these units. SRS agreed to refrain from further disposal of F-listed solvent contaminated rags and wipes in the Low-Level Radioactive Waste Disposal Facility and sanitary landfill or any other facility not specifically permitted for such disposal. An SRS Public Involvement Plan has been designed to facilitate public involvement in decision-making processes for permitting, closure, and selection of remedial alternatives. The plan addresses the requirements of CERCLA, RCRA, and NEPA, and forms the basis for continued involvement of stake-holders at SRS. As envisioned by the revised plan, SRS has formed a Citizen's Advisory Board to further public involvement. Currently, SRS is storing PCB-contaminated equipment and cleanup materials for which, due to their radioactive nature, treatment is not currently available. DOE is presently developing this treatment capability and SRS intends to store these PCB-contaminated radioactive equipment and materials onsite until such time as appropriate disposal methods, treatment capacity, and facilities become available. Currently, SRS is storing these PCB-contaminated radioactive materials in an onsite facility that meets storage requirements under 40CFR 761, but is out of compliance with the provision requiring removal of PCB-contaminated wastes from storage and their disposal within 1 year. As at other DOE sites, DOE is presently working with EPA to approve a treatability study to remove the PCB contamination and the radioactive materials from SRS as LLW. -TSAR_DOE_SECTION- A.2 Project Descriptions TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions This section contains detailed descriptions of the individual tritium supply and recycling technologies being evaluated in this PEIS. Each supply technology description represents a new facility and the recycling descriptions represent either new or upgraded facilities whose impacts are being considered in evaluating the program alternatives. Design bases for each of the tritium supply technology options have been previously developed and are as follows: Heavy Water Reactor (HWR)-All requirement documents developed in the New Production Reactor program for the HWR. Modular High Temperature Gas-Cooled Reactor (MHTGR)-All requirement documents developed in the New Production Reactor program for the MHTGR. Advanced Light Water Reactor (ALWR)-All requirements developed for the New Production Reactor program and the data, as gathered, from the Surplus Fissile Materials program. Accelerator Production of Tritium (APT)-All requirement documents developed in the Tritium Supply and Recycling program for the APT as amended. Each of the facility designs will, to the extent possible, incorporate the following objectives: maximum operational efficiency, ability to incorporate future technological advances, and ability to meet changing operating requirements. The upgrades/modifications option is designed to renovate existing tritium recycling facilities at SRS to meet mission requirements while complying with all ES&H criteria. The conceptual designs described in the following sections for the various tritium supply and recycling facilities were based on the following assumptions: The design would be based on a normal production rate to meet anticipated baseline tritium requirements (3/8 New Production Reactor goals). The enduring stockpile would consist of a mix of systems equivalent to the following existing weapons programs: B61, B83, W76, W78, W80, W87, and W88. The facilities would be designed to produce only materials or components required for nuclear weapons programs. The exception is the tritium recycling facility, which would provide tritium for commercial sales. The tritium facilities would be capable of producing materials and assembling new building components for two weapon systems in any given year. This capability would be achieved by either simultaneous or sequential campaigns, as long as the sum of the product shipments for the year meets the annual production goals. -TSAR_DOE_SECTION- A.2.1 Tritium Supply TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.1 Tritium Supply Tritium, a radioactive form of hydrogen gas, is an essential component of all the warheads in the existing and projected nuclear weapons stockpile. Since it decays at a rate of 5.5 percent per year (12.3-year half-life), the tritium in fielded weapons must be replenished periodically. In addition, because of its decay, the existing national inventory of tritium will become insufficient to meet projected nuclear weapon stockpile requirements around the end of the next decade. This means that a source of new tritium will be needed to prevent degradation of the national nuclear deterrent capability. The last operating tritium supply reactor, K-Reactor at SRS, has been placed in cold standby with no planned provision for restart. Therefore, DOE is evaluating the construction of a new tritium supply facility. There are a number of candidate tritium production technologies under consideration that could fulfill this mission. Each of these technologies has strengths and weaknesses that are dependent on: selected site, tritium production capacity, and/or supporting infrastructure. The technologies under consideration are listed below and will be discussed in the following sections: HWR; MHTGR; ALWR; and Linear Induction Accelerator, herein after referred to as APT. -TSAR_DOE_SECTION- A.2.1.1 Heavy Water Reactor TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.1 Tritium Supply A.2.1.1 Heavy Water Reactor Mission. The primary mission of the HWR facility would be to provide the full range of tritium supply functions performed by DOE, as well as waste treatment functions related to HWR operations. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. The HWR facility would produce tritium by bombarding an isotope of the element lithium with neutrons produced by nuclear fission reactions in uranium. The interaction of neutrons with the lithium creates an unstable lithium isotope that decays almost instantaneously to tritium and an alpha particle. The HWR facility would consist of the following major components: the reactor, target and fuel fabrication, tritium target processing, interim spent nuclear fuel storage, general service, and waste treatment. Figure A.2.1.1-1 depicts the HWR tritium production process and figure A.2.1.1-2 shows the HWR facility site plan. This figure shows a mechanical draft cooling tower system. Component Facility Functions Reactor. The reactor would be based on a low-pressure, low-temperature application of nuclear fission technology specifically designed to produce tritium. The reactor vessel and cooling system configuration (with primary and secondary cooling loops) would be similar to that used in commercial light water reactor nuclear power technology. The HWR would use heavy water as the reactor coolant and moderator. Heavy water, circulated through the core for cooling and moderation, also passes through heat exchangers that are external to the reactor tank. The heat is in turn carried away by the secondary cooling system. The heavy water in the tank surrounding the fuel would represent the bulk moderator. Less than 10 percent of the reactor heat would be generated in the moderator space because of gamma heating; the bulk of the heat would be generated in the core. The major engineered safety features would include the reactor containment building; the emergency core cooling system to provide makeup coolant flow heat removal for an extended period, following either a loss-of-coolant accident or a loss-of-pumping accident; emergency shutdown safety rods independent of the control rods; neutron poison injection of gadolinium nitrate into the moderator space of the reactor tank; and a backup residual heat removal system capable of removing decay heat from the reactor if the primary coolant fails to circulate. Engineered safety features of the HWR also include natural circulation cooling by the primary and secondary coolant systems and the boiling pot mode. The boiling pot mode would occur by circulation through the core of the water in which the system is submerged following complete discharge of the Borated Light-Water Reactor Storage Tank; the core will be cooled by a two-phase flow. Steam generated by this process condenses on the containment shell and interior structures. The condensate replenishes the water at the bottom of the containment. The containment shell is cooled by convective flow of outside air which flows between the steel containment and the concrete shield. All the engineered safety features combined would provide a "walk-away" design of the HWR to prevent and/or mitigate a beyond design basis accident event (severe accident). A circulating ventilation system with atmospheric driers would be used to collect and control tritiated water vapor during normal operation, thus reducing leakage of tritium. The containment system would include such features as heat removal sprays, hydrogen igniters, and long-term residual heat removal systems to mitigate potential radiological releases under severe accident conditions. The initial heavy water inventory for an HWR is available from existing inventories at SRS. Because of the relatively low exit coolant temperature in the HWR, a power conversion system would not be feasible. The heat dissipation system would ensure that the reactor is cooled by the flow of water circulated through the core in a closed primary loop under pressure through the reactor vessel and through coolant pipes connected in parallel to heat exchangers. The heat dissipation system selected, wet or dry, would be dependent on site characteristics. Both wet and dry cooling systems would use water as the heat exchange medium. Wet systems would use water towers and the evaporation process to carry off heat. Dry systems, designed for cold climates, would use water in closed nonevaporative cooling towers to carry off heat to the atmosphere by conduction through radiator-like vanes. In moderate climates, fans would be added to the dry cooling towers to move air over the vanes. There would be some water loss through evaporation in a dry system, but significantly less than with a wet tower. Dry cooling towers would be used for the reactors at all dry sites. Target and Fuel Fabrication Facility. The concentric tubes of lithium-aluminum and of uranium-aluminum would be fabricated and assembled into tritium target/fuel assemblies in the target and fuel fabrication facility. The facility would include areas for alloying, casting, machining, extrusion, inspection, and assembly. Auxiliary services would include a maintenance shop, an electronic and instrumentation shop, offices, and a metallurgical and chemical laboratory. There would also be several small storage buildings nearby for lithium and aluminum components. The facility would be designed to control the spread of contamination and prevent the uncontrolled release of radioactive material to the environment. Tritium Target Processing Facility. After irradiation and cooling in the spent fuel pool, the tritium target tubes would be withdrawn from the assemblies and sent to the tritium target processing facility for tritium extraction. The facility would house the pro- cesses, laboratory, and other activities associated with tritium handling. Most of the process operations and hot laboratory analyses would be performed in nitrogen-blanketed gloveboxes to ensure tritium containment. The hot cell would contain two or more vacuum furnaces for extracting tritium from the irradiated targets. Airlocks would separate the furnaces from the rest of the hot cell. Interim Spent Nuclear Fuel Storage Facility. Spent nuclear fuel would be sent to, and stored underwater at, the interim spent fuel storage facility. General Service Facility. The general service facility would contain the control equipment and various plant support equipment. Waste Treatment Facility. The waste treatment facility would receive all solid, liquid, and gaseous radioactive waste for storage, treatment, and packaging for either release or disposal at an appropriate permanent waste disposal facility. Description of Processes Tritium Target and Fuel Production. Lithium-aluminum tritium target alloy would be formed into billets and preextruded into logs which would then be machined and nested into coextrusion billets. These billets would then be coextruded to produce the target tubes. The uranium-aluminum fuel alloy would be similarly formed into fuel tubes and then nested with the target tubes into the tritium target/fuel assemblies described earlier. Tritium Production. The tritium target/fuel assemblies would be loaded into the reactor and irradiated. When the irradiation is complete, the assemblies would be removed and shipped to the disassembly basin. Tritium Extraction. At the disassembly basin, the target tubes would be separated from the fuel tubes and allowed to cool so that short-lived isotopes could decay. A 1-month cooling period for tritium targets is assumed. After cooling, the target tubes would be washed within the shielded transport cask, then placed in a crucible, inserted into a furnace, and heated to drive off the gases, including the tritium product. The gas stream from the furnace, mainly helium and hydrogen isotopes, including tritium, would pass through a diffusing process to separate the hydrogen isotopes from helium and other impurities. This process would cycle the gas between two beds of treated diatomaceous earth used as a filter aid. One bed would be heated to drive off the gases; the other bed would be cooled to promote gas absorption. The purified tritium would be sent to the tritium recycling facility, where tritium reservoirs are filled. Facility Utilities. Facility construction and operation utility requirements are shown in tables A.2.1.1-1 and A.2.1.1-2, respectively. Chemicals Required. Table A.2.1.1-3 depicts chemical resources required during operation. Personnel Requirements. Construction of the HWR would have a peak employment of 2,320 con- struction workers. Approximately 9,760 worker years would be needed during the 8-year construction period. Operation of the HWR would require 930workers with 230 of these badged for radiation detection. Transportation. Interfacility transfers would be made by rail, truck transport, or pipeline, as appropriate. Truck service would be needed for intrafacility transport. Waste Management. The solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Uncontaminated wastewa- ter would be used for soil compaction and dust control, and excavated soil would be used for grading and site preparation. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes generated during construction would consist of such materials as waste adhesives, oils, cleaning fluids, solvents, and coatings. Hazardous waste would be packaged in Department of Transportation (DOT) approved containers and shipped offsite to commercial RCRA- permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The facility design considers and incorporates waste minimization and pollution prevention. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycling. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials that generate hazardous or mixed waste. Production processes would be config- ured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future D&D considerations would also be incorporated into the design. Table A.2.1.1-4 presents the estimated annual spent nuclear fuel and waste volumes from the HWR facility during construction and operation. Solid and liquid waste streams would be routed to the waste management system. Figure A.2.1.1-3 depicts the waste management system at a dry site while figure A.2.1.1-4 illustrates the waste management system for a wet site. Solid wastes would be characterized and segregated into LLW, hazardous, and mixed wastes, then treated to forms suitable for disposal or storage within the facility. Liquid wastes would be treated onsite to reduce hazardous/toxic and radioactive elements before discharge or transport. All fire sprinkler water discharged in process areas would be contained and treated as process wastewater, when required. Spent Nuclear Fuel. Spent nuclear fuel would not be reprocessed. Spent nuclear fuel would be initially stored at the spent nuclear fuel storage facility in pools onsite. After the spent nuclear fuel decay heat has decreased sufficiently (approximately 230 days), the nested fuel tubes from four spent nuclear fuel assemblies would be loaded into stainless steel containers in the disassembly area. The loaded containers would be dewatered by displacement with air or an inert gas such as helium or argon and then sealed. The sealed containers would then be transferred to an interim onsite dry storage basin for horizontal emplacement on racks with poison plate separators. The facility design would have sufficient capacity to store the spent nuclear fuel for the life of the facility. The ROD (60 FR 28680) from the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement has the type of spent nuclear fuel generated by the HWR destined for consolidation at SRS to await availability of a geologic repository. Transuranic Waste. The HWR would not generate any TRU waste. Low-Level Waste. LLW would be generated by the operation of the reactor and support facilities. The aqueous releases from fuel and tritium-target fabrication would be liquid LLW from tube-cleaning operations. Process effluents would be temporarily held in storage tanks before treatment into solid LLW that is suitable for LLW disposal. The nonhazardous liquid effluent would then be discharged through a permitted NPDES outfall. The bulk of the solid LLW would be generated in the reactor, fuel fabrication, and tritium target processing facilities. Solid LLW would consist of contaminated equipment pieces, plastic sheeting, protective clothing, solidified slurry from the wastewater treatment facility, spent high efficiency particulate air (HEPA) filters, spent targets, spent catalysts and resins, and spent uranium beds. Solid LLW would be compacted as appropriate and then disposed of in a suitable disposal facility. Mixed Low-Level Waste. No liquid mixed LLW would be generated from the HWR facility. Solid mixed LLW may originate from wipes laden with contaminated oils and hydraulic fluids from the tritium extraction facility. Mixed LLW would be stored in an onsite RCRA-permitted storage facility until treatment in accordance with the site-specific treatment plan that was developed to comply with the Federal Facility Compliance Act of 1992. Hazardous Waste. Liquid hazardous wastes would be generated from degreasing agents, cleaning solvents, cutting oils, vacuum pump oils, film processing fluids, hydraulic fluids from mechanical equipment, antifreeze solutions, and paint. The cleaning solvent selected would be from a list of nonhalogenated solvents. Liquid hazardous wastes would be collected in DOT-approved containers and sent to an onsite hazardous waste accumulation area. The hazardous waste accumulation area would provide a 90-day staging capacity prior to shipment to an offsite commercial RCRA-permitted treatment, storage, and disposal facility using DOT-certified transporters. Solid hazardous wastes would be generated from nonradioactive materials such as wipes contaminated with oils, lubricants, and cleaning solvents that are used for equipment outside the main processing units. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT-approved containers and sent to a hazardous waste accumulation area for staging prior to shipment to an offsite commercial RCRA-permitted treatment, storage, and disposal facility using DOT-certified transporters. Nonhazardous Waste. Sewage wastewater would be treated in the sanitary wastewater treatment plant. Sewage wastewater is kept separate from all industrial and process wastewaters and normally contains no radioactive wastes from the facility. The sewage wastewater would be routinely monitored for radioactive contaminants. The sewage process sludge would be disposed of in a permitted landfill. All treated effluent would be discharged to a stream or river through an NPDES outfall (wet site) or a natural drainage channel (dry site). Cooling system blowdown would be treated and discharged to the river (wet site). Because the dry site design uses nonevaporative-type cooling towers, there would be no liquid discharges resulting from blowdown. The treated effluent from the utility wastewater treatment would be discharged to the river through an NPDES outfall (wet site) or a natural drainage channel. All sludges would be disposed of in a permitted landfill. Other nonrecyclable, solid, nonhazardous sanitary and industrial wastes would be compacted and disposed of in a permitted landfill. The facility design includes stormwater retention facilities with the necessary NPDES monitoring equipment. Rainfall within the Limited Area and Protected Area would be collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge to the natural drainage channels (dry site) or river (wet site). If the runoff is contaminated, it would be treated in the radioactive waste treatment system. Runoff from the Property Protection Area would be discharged directly into the natural drainage channels or river. Figure (Page A-35) Figure A.2.1.1-1.-Heavy Water Reactor Tritium Production Process. Figure (Page A-36) Figure A.2.1.1-2.-Heavy Water Reactor Facility (Typical). Figure (Page A-37) Figure A.2.1.1-3.-Heavy Water Reactor Waste Management System (Dry Site). Figure (Page A-38) Figure A.2.1.1-4.-Heavy Water Reactor Waste Management System (Wet Site). Table A.2.1.1-1.-Heavy Water Reactor Construction Material/Resource Requirements Material/Resource Consumption Electrical Energy (MWh) 87,000 Concrete (yd3) 220,000 Steel (tons) 45,000 Fuel (gal) 2,400,000 Water (gal) 170,000,000 Source: DOE 1995d. Table A.2.1.1-2.-Heavy Water Reactor Operation Utility Requirements Utility Consumption Electrical Energy (MWh per year) Wet site 370,000 Dry site 540,000 Electrical Load (MWe) Wet site 51 Dry site 69 Fuel Gas (ft3 per year) 240,000,000 Liquid (GPY) 82,000 Water (MGY) Wet site 5,900 Dry site 48 Source: DOE 1995d. Table A.2.1.1-3.-Heavy Water Reactor Annual Chemical Requirements Chemicals Quantity (tons) Aluminum 12 Uranium metal 0.75 Lithium metal 0.08 Stainless steel 0.125 Nitric acid 1.2 Sodium hydroxide 0 Sulfuric acid 0 Source: DOE 1995d. Table A.2.1.1-4.-Heavy Water Reactor Estimated Spent Nuclear Fuel and Waste Volumes - - Dry Site Wet Site - Annual Average Volume Annual Volume Annual Volume Annual Volume Annual Volume Generated From Generated From Effluent From Generated From Effluent From Construction Operations Operations Operations Operations Category (yd3) (yd3) (yd3) (yd3) (yd3) Spent Nuclear Fuel None 7 7a 7a 7a Low-Level Liquid None 10,400 None 10,400b None (2,100,000 gal) (2,100,000 gal) Solid None 5,200 1,870 5,200 1,870c Mixed Low-Level Liquid None None None None None Solid None 120 120 120 120 Hazardous Liquid Included in solid Included in solid Included in solid Included in solid Included in solid Solid 20 40 40 40 40 Nonhazardous (Sanitary) Liquid 79,220 238,000 238,000 11,600,000 11,600,000 (16,000,000 gal) (48,000,000 gal) (48,000,000 gal) (2,350,000,000 gal) (2,350,000,000 gal) Solid 7,800 7,600 2,530 7,600 2,530d Nonhazardous (Other) Liquid 2,570 Included in Included in Included in Included in (520,000 gal) sanitary sanitary sanitary sanitary Solid Included in 6,500 None 6,500f None sanitary -TSAR_DOE_SECTION- A.2.1.2 Modular High Temperature Gas-Cooled Reactor TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.1 Tritium Supply A.2.1.2 Modular High Temperature Gas-Cooled Reactor Mission. The primary mission of the MHTGR facility would be to provide the full range of tritium supply functions performed by DOE, as well as waste treatment functions related to MHTGR operations. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. The MHTGR would produce tritium by bombarding an isotope of the element lithium with neutrons produced by nuclear fission reactions in uranium. The interaction of neutrons with the lithium creates an unstable lithium isotope that decays almost instantaneously to tritium and an alpha particle. The MHTGR facility would consist of the following components: the specified number of reactors (three reactors are analyzed in this PEIS), target and fuel fabrication, tritium target processing, interim spent fuel storage, general service, waste treatment, and power conversion. Figure A.2.1.2-1 depicts the MHTGR tritium production process while figure A.2.1.2-2 shows the MHTGR facility site plan. Component Facility Functions. The MHTGR would be sized (i.e., the number of reactors) and designed for a specific tritium supply capability based on reasonably foreseeable long-term requirements. Reactor. Each individual reactor would be a moderate pressure, high temperature device designed to produce tritium. A reactor includes its vessel, a steam generator, a helium circulator, and interconnecting cross ducts housed in the reinforced concrete containment structure. The reactor vessel would contain the core, control rods and drives, the reflector, and the core supports. Top-mounted standpipes would house the control rod drives and the reserve shutdown system. A shutdown circulator mounted at the bottom of the core would provide cooling when the main circulator is out of service for maintenance. The core size and shape provide for natural convection heat transfer through to the pressure vessel walls in the event of the loss of circulation or if the pressure boundary is breached for any reason. Core power would be controlled by moveable control rods con- taining boron carbide as a nuclear poison. Reserve shutdown would consist of boronated graphite pellets normally contained in hoppers above the core. When needed, they would be dropped into holes in the fuel blocks, shutting down the nuclear reaction. Helium would be used as the coolant and graphite would provide the core structure and act as moderator. The coolant would be pumped by an electrically driven circulator located above the steam generator. The reactor core would be housed in a steel pressure vessel. A cross vessel would direct the helium to the steam generator. The helium heated by the core would flow over the steam generator tube bundle, giving up its heat to the feedwater, and producing superheated steam. This steam would be of sufficient quality that it could be used for the production of electricity. When shut down for maintenance or refueling, the decay heat of the core would be removed either by the normal cooling system, or by the shutdown cooling system, with its own circulator and independent heat exchanger. In the event of transients that would increase reactor fuel temperature, the increased temperature would decrease the power level. This would work two ways in that the mass of the core would absorb the additional heat generated, and as the temperature gradually rises, power would decrease. Even in the event of the loss of coolant at full power, the reactor would return to a lower temperature, core geometry would be maintained, and control rod insertion would not be hampered. Predictable reactor response and long time constants are a product of the high heat capacity of the core and the lack of phase changes in the primary coolant. All this would occur without the need for operator intervention. The heat dissipation system selected, wet or dry, would depend on site characteristics. Both wet and dry cooling systems would use water as the heat exchange medium. Wet systems would use water towers and the evaporation process to carry off heat. Dry systems, designed for cold climates, would use water in closed nonevaporative cooling towers to carry off heat by conduction to the atmosphere through radiator-like vanes. In moderate climates, fans would be added to the dry cooling towers to move air over these vanes. There would be some water loss through evaporation in a dry system, but significantly less than with a wet tower. Dry cooling towers would be used for the reactors at all dry sites. The use of wet cooling towers would be an option only for the power conversion facility and only when the facility would be located at a wet site. Target and Fuel Fabrication Facility. The target and fuel rods would be assembled for irradiation in the reactor. After irradiation, the assemblies would be returned for separation of the target and fuel rods. Tritium Target Processing Facility. Spent nuclear fuel rods would be sent to interim storage and the target rods would be sent to the tritium processing facility, which houses the processes, laboratory, and other activities associated with the production of tritium gas including a hot cell for receiving, storing, and processing irradiated target rods. Interim Spent Nuclear Fuel Storage Facility. This facility consists of three water-cooled fuel storage basins paired with individual reactors. Fuel elements containing spent fuel would be stored in dry canisters for up to 3 years in the storage basins. After the cooling period, the spent fuel elements would be encapsulated and then transferred to dry storage vaults for the life of the plant. Power Conversion Facility. This facility would consist of a fully independent turbine generator for each reactor. It would also include a maintenance building, makeup water treatment and auxiliary boiler building, station cooling towers, station transformers, circulating water pumphouse, and electric switchyard. General Service Facility. This facility would contain the control equipment and various plant support equipment. Waste Treatment Facility. This facility would receive all solid, liquid, and gaseous radioactive waste for storage, treatment, and packaging for either release or disposal at an appropriate permanent waste disposal facility. Descriptions of Processes Tritium Target Fabrication. Targets would be fabricated from pyrocarbon-coated lithium aluminate microspheres. The compacts would be loaded into graphite sleeves that are subsequently loaded into graphite blocks similar to those used to hold the fuel rods. Tritium Production. The fuel and target blocks would be loaded into the reactor and irradiated. When the irradiation is complete, the blocks would be removed and the target compacts would be separated from the fuel and sent to the tritium extraction facility for further processing. Tritium Extraction. The tritium in the irradiated target compacts would be extracted by heating and captured by chemical absorption. The spent targets would be disposed of as waste. The tritium would be transferred to the tritium recycling facility. Facility Utilities. Facility construction and operation utility requirements are shown in tables A.2.1.2-1 and A.2.1.2-2 respectively. Chemicals Required. Table A.2.1.2-3 lists chemical resources required during operation. Personnel Requirements. Construction of the MHTGR would have a peak employment of 2,210construction workers. Approximately 8,810worker-years would be needed during the 9-year construction period. Operation of the MHTGR would require 910 workers with 180 of these badged for radiation detection. Transportation. Interfacility transfers would bemade by rail, truck transport, or pipeline as appropriate. Truck service would be needed for intrafacility transport. Waste Management. The solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Uncontaminated wastewa- ter would be used for soil compaction and dust control, and excavated soil would be used for grading and site preparation. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes generated during construction would consist of such materials as waste adhesives, oils, cleaning fluids, solvents, and coatings. Hazardous waste would be packaged in DOT-approved containers and shipped offsite to commercial RCRA-permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The facility design considers and incorporates waste minimization and pollution prevention. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycling. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials that contribute to the generation of hazardous or mixed waste. Production processes would be configured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future D&D considerations would also be incorporated into the design. Table A.2.1.2-4 presents the estimated annual spent nuclear fuel and waste volumes from the MHTGR during construction and operation. Solid and liquid waste streams would be routed to the waste management system. Figure A.2.1.2-3 depicts the waste management system at a dry site while figure A.2.1.1-4 illustrates the waste management system for a wet site. Solid wastes would be characterized and segregated into low-level, hazardous, and mixed wastes, then treated to a form suitable for disposal or storage within the facility. Liquid wastes would be treated onsite to reduce hazardous/toxic and radioac- tive elements before discharge or transport. All fire sprinkler water discharged in process areas would be contained and treated as process wastewater, when required. Spent Nuclear Fuel. Spent nuclear fuel would not be reprocessed. Following removal from the reactor, spent nuclear fuel would be stored in the spent fuel storage wells in the reactor service building. After a 3-year cooling period, a fuel-handling machine would load spent nuclear fuel elements, one at a time, into a fuel cylinder. The cylinder would be plugged at each end with suitable shielding and filled with helium. After the end plates were welded on, the cylinder would then be leak-tested. The cylinder would be inserted into a dry storage canister. After the canister is filled with seven fuel cylinders, it would be filled with helium, welded shut, and leak-tested. To remove the dry storage canister from the reactor service building, specially designed equipment would be used to grapple the canister and insert it into a truck-mounted transfer cask. At the storage facility, the canister would be placed into a horizontal concrete storage module. The module would include a passive ventilation system for decay heat removal, metal heat shields to protect the strength of the concrete, and temperature monitors and alarms. These modules would be arrayed on slab-on-grade basemats in a rectangular configuration near the reactor service building. The facility design would have sufficient capacity to store the spent nuclear fuel for the life of the facility. The ROD (60 FR 28680) from the DOE Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement has the type of spent nuclear fuel generated by the MHTGR destined for consolidation at INEL to await availability of a geologic repository. Transuranic Waste. The MHTGR would not generate any TRU waste. Low-Level Waste. LLW would be generated by the operation of the reactor and support facilities. Process effluents would be temporarily stored in storage tanks before treatment into solid LLW that is suitable for LLW disposal. The nonhazardous liquid effluent would then be discharged through a permitted NPDES outfall. The bulk of the solid LLW would be generated in the tritium target processing facility. Solid LLW would consist of contaminated equipment pieces, spent targets, spent uranium beds, plastic sheeting, spent resins, spent HEPA filters, and protective clothing. Solid LLW would be compacted as appropriate and then disposed of in a suitable disposal facility. Mixed Low-Level Waste. No liquid mixed LLW would be generated from the MHTGR. Solid mixed LLW may originate from wipes laden with contaminated oils and hydraulic fluids from the tritium extraction facility. Mixed LLW would be stored in an onsite RCRA-permitted facility until treatment in accordance with the site-specific treatment plan that was developed to comply with the Federal Facility Compliance Act of 1992. Hazardous Waste. Liquid hazardous wastes would be generated from cleaning solvents, cutting oils, vacuum pump oils, film processing fluids, hydraulic fluids from mechanical equipment, antifreeze solutions, acids, and paint. The cleaning solvent selected for use would be from a list of nonhalogenated solvents. Liquid hazardous wastes would be collected in DOT-approved containers and sent to a hazardous waste accumulation area. The hazardous waste accumulation area would provide a 90-day staging capacity prior to shipment to an offsite commercial RCRA-permitted treatment, storage, and disposal facility using DOT-certified transporters. Solid hazardous wastes would be generated from nonradioactive materials such as wipes contaminated with oils, lubricants, and cleaning solvents that are used for equipment outside the main processing units. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT- approved containers and sent to a hazardous waste accumulation area for staging prior to shipment to an offsite commercial RCRA-permitted treatment, storage, and disposal facility using DOT-certified transporters. Nonhazardous Waste. Sewage wastewater would be treated in the sanitary wastewater treatment plant. Sewage wastewater would be kept separate from all industrial and process wastewaters and normally contains no radioactive wastes from the facility. The sewage wastewater would be routinely monitored for radioactive contaminants. The sludge would be disposed of in a permitted sanitary landfill. All treated effluent would be discharged to a stream or river through an NPDES outfall (wet site) or a natural drainage channel (dry site). Cooling tower blowdown would be treated and discharged to the river (wet site). Because the dry site design uses nonevaporative type cooling towers, there would be no liquid discharges resulting from blowdown. The treated effluent from the utility wastewater treatment would be discharged to the river through an NPDES outfall (wet site) or a natural drainage channel. All sludges would be disposed of in a permitted landfill. Other nonrecyclable, solid nonhazardous sanitary and industrial wastes would be compacted and disposed of in a permitted landfill. The facility design includes stormwater retention facilities with the necessary NPDES monitoring equipment. Rainfall within the Limited Area and Protected Area would be collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge to the natural drainage channels (dry site) or river (wet site). If contaminated, the runoff would be treated in the radioactive waste treatment system. Runoff from the Property Protection Area would be discharged directly into the natural drainage channel or river. Figure (Page A-45) Figure A.2.1.2-1.-Modular High Temperature Gas-Cooled Reactor Tritium Production Process. Figure (Page A-46) Figure A.2.1.2-2.-Modular High Temperature Gas-Cooled Reactor Facility (Typical). Figure (Page A-47) Figure A.2.1.2-3.-Modular High Temperature Gas-Cooled Reactor Waste Management System (Dry Site). Figure (Page A-48) Figure A.2.1.2-4.-Modular High Temperature Gas-Cooled Reactor Waste Management System (Wet Site). Table A.2.1.2-1.-Modular High Temperature Gas-Cooled Reactor Construction Material/Resource Requirements Material/Resource Consumption Electrical energy (MWh) 73,000 Concrete (yd3) 220,000 Steel (tons) 60,000 Fuel (gal) 3,200,000 Water (gal) 160,000,000 Source: DOE 1995e. Table A.2.1.2-2.-Modular High Temperature Gas-Cooled Reactor Operation Utility Requirements Utility Consumption Electrical Energy (MWh per year) Wet site 260,000 Dry site 360,000 Electrical Load (MWe) Wet site 36 Dry site 46 Fuel Gas (ft3 per year) 6,000,000 Liquid (GPY) 81,000 Water (MGY) Wet site 4,000 Dry site 30 Source: DOE 1995e. Table A.2.1.2-3.-Modular High Temperature Gas-Cooled Reactor Annual Chemical Requirements Chemical Quantity (tons) Aluminum oxide 2 Uranium oxide 1.1 Lithium nitrate 1.4 Stainless steel 0.08 Nitric acid 49 Sodium hydroxide 1.1 Sulfuric acid 0 Graphite 122 Source: DOE 1995e. Table A.2.1.2-4.-Modular High Temperature Gas-Cooled Reactor Estimated Spent Nuclear Fuel and Waste Volumes - - Dry Site Wet Site - Annual Average Volume Annual Volume Annual Volume Annual Volume Annual Volume Generated From Generated From Effluent From Generated From Effluent From Construction Operations Operations Operations Operations Category (yd3) (yd3) (yd3) (yd3) (yd3) Spent Nuclear Fuel None 80 80a 80a 80a Low-Level Liquid None 2,600 None 2,600b None (525,000 gal) (525,000) Solid None 1,300 468 1,300 468c Mixed Low-Level Liquid None None None None None Solid None <1 <1 <1 <1 Hazardous Liquid Included in solid Included in solid Included in solid Included in solid Included in solid Solid 20 100 100 100 100 Nonhazardous (Sanitary) Liquid 64,400 149,000 149,000 8,070,000 8,070,000 (13,000,000 gal) (30,000,000 gal) (30,000,000 gal) (1,630,000,000 gal) (1,630,000,000 gal) Solid 7,100 7,400 2,470 7,400 2,470d Nonhazardous (Other) Liquid 3,020 Included in sanitary Included in sanitary Included in sanitary Included in sanitary (610,000 gal) Solid Included in sanitary 6,400 None 6,400f None -TSAR_DOE_SECTION- A.2.1.3 Advanced Light Water Reactor TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.1 Tritium Supply A.2.1.3 Advanced Light Water Reactor Mission. The primary mission of the ALWR facility would be to provide the full range of tritium supply functions performed by DOE, as well as waste treatment functions related to ALWR operations. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. The ALWR would produce tritium by bombarding an isotope of the element lithium with neutrons produced by nuclear fission reactions in uranium. The interaction of neutrons with lithium creates an unstable lithium isotope that decays almost instantaneously to tritium and an alpha particle. Two ALWR design approaches, based on rated power (large and small reactor, designated Large ALWR and Small ALWR in the following discussion), are under consideration. For either design, the ALWR facility would consist of the following major components: the reactor, tritium target processing, interim spent fuel storage, power conversion, and waste treatment. FigureA.2.1.1-1depicts the ALWR tritium production process, while figure A.2.1.3-2 shows a typical ALWR facility site plan. There are two 1,300 MWe Large ALWR designs under consideration: the large pressurized water reactor and the large boiling water reactor. Two 600-MWe Small ALWR design concepts are also under consideration: a small pressurized water reactor and a small boiling water reactor. Component Facility Functions. Reactor. The reactor would be an improved version of existing commercial electric power generating reactors and would operate at or near rated power. An ALWR would use ordinary (light) water as both the moderator and coolant. Modifications to the design for tritium production would be minimal, with the principal modification being the replacement of a fraction of the fuel rods by target rods. The enrichment of the remaining fuel rods would be increased to compensate for the missing fuel rods, but derating of core power might be necessary. The core, contained within a steel pressure vessel, would be composed of bundles of fuel and target rods. The fuel rods would consist of tubes of zircaloy cladding filled with uranium dioxide fuel pellets, while the target rods would consist of steel clad lithium aluminate. Target rods would be removed from the core and replaced with new target rods once a year. Fuel rods would be obtained from offsite sources, while target rods would be fabricated onsite. The heat dissipation system selected, wet or dry, would depend on site characteristics. Both wet and dry cooling systems would use water as the heat exchange medium. Wet systems would use water towers and the evaporation process to carry off heat. Dry systems, designed for cold climates, would use water in closed nonevaporative cooling towers, to carry off heat by conduction to the atmosphere through heat exchangers. In moderate climates, fans would be added to the dry cooling towers to move air over the vanes of the heat exchangers. There would be some water loss through evaporation in a dry system, but significantly less than with a wet tower. Dry cooling towers would be used for the reactors at all dry sites. The use of wet cooling towers would be an option only for the power conversion facility and only when the facility would be located at a wet site. Tritium Target Fabrication Facility. This building would house the facilities and equipment needed to produce lithium-aluminate target rods. Tritium Target Processing Facility. After irradiation and a cooling period, the target rods are processed to remove the tritium. The facility houses the processes, laboratory, and other activities associated with handling irradiated elements and is also designed to control the spread of contamination within the facilities and prevent the uncontrolled release of radioactive material to the environment. Interim Spent Nuclear Fuel Storage Facility. Spent nuclear fuel would be sent to, and stored underwater at the interim spent fuel storage facility. Power Conversion Facility. This facility would contain a turbine generator, electrical equipment, control equipment, auxiliary systems, plant support systems, and other equipment. Waste Treatment Facility. This facility would receive all solid, liquid, and gaseous radioactive waste for storage, treatment, and packaging for either release or disposal at an appropriate permanent waste disposal facility. Description of Processes Tritium Target and Fuel Production. Fuel rod fabrication would be done offsite by commercial sources. Tritium-target rod fabrication begins with lithium carbonate and aluminum-oxide, which are blended, calcined, pressed into thin, annular, ceramic pellets, and sintered. The pellets are then ground to size, inspected, preassembled with the getter, and loaded into target rods clad in reactor grade stainless steel having an inner coat of aluminide. The rods are dried, backfilled with helium, and welded shut. The target rods are placed in containers and transported to the off-site fuel and target assembly facility where the target rods, fuel rods, and associated hardware are assembled into a fuel and target assembly. The fuel and target assemblies are shipped to the ALWR tritium supply plant. Tritium Production. Tritium would be produced within the steel clad lithium aluminate target rods, contained in fuel bundles in the reactor core. After about one year, when sufficient tritium would have collected within the target rods, the target rods would be removed from the reactor and placed in the interim fuel storage pool for approximately 1 month to allow for the decay of short-lived radioactive nuclides to acceptable levels. Target rods would then be sent to the onsite tritium extraction facility. Tritium Extraction. The tritium target rods would be removed from the storage pool and sent to the main hot cell. In the hot cell, the target tubes would be heated in a furnace to drive off the tritium, helium, and hydrogen gases. This gas stream would pass through a diffuser to separate the hydrogen isotopes from the other gases. Facility Utilities. Bounding values for the Large ALWR and the Small ALWR facility construction and operation utility requirements are shown in tables A.2.1.3-1 and A.2.1.3-2, respectively. Chemicals Required. Table A.2.1.3-1 depicts chemical resources required during operation. Personnel Requirements. Construction of the ALWR would have a peak employment of 3,500construction workers for a Large ALWR and 2,200workers for a Small ALWR. Approximately 12,600 worker-years would be needed for a Large ALWR and 7,100 worker years for a Small ALWR during the 6-year construction period. Operation of the Large ALWR would require 830 workers of which 210 would be badged. The Small ALWR would require 500 workers of which 125 would be badged. Transportation. Interfacility transfers would be made by rail, truck transport, or pipeline, as appropriate. Truck service would be needed for intrafacility transport. Waste Management. The solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Uncontaminated wastewa- ter would be used for soil compaction and dust control, and excavated soil would be used for grading and site preparation. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes generated during construction would consist of such materials as waste adhesives, oils, cleaning fluids, solvents, and coatings. Hazardous wastes would be packaged in DOT-approved containers and shipped offsite to commercial RCRA-permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The facility design considers and incorporates waste minimization and pollution prevention. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycle. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials that contribute to the generation of hazardous or mixed waste. Production processes would be configured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future D&D considerations would also be incorporated into the design. Tables A.2.1.3-4 and A.2.1.3-5 present the estimated annual spent nuclear fuel and waste volumes from the ALWR, Large and Small, respectively, during construction and operation. Solid and liquid waste streams would be routed to the waste management system. Figure A.2.1.3-3 depicts the waste management system at a dry site while figure A.2.1.3-4 illustrates the waste management system for a wet site. Solid wastes would be characterized and segregated into LLW, hazardous, and mixed wastes, then treated into a form suitable for disposal or storage within the facility. Liquid wastes would be treated onsite to reduce hazardous/toxic and radioactive elements before discharge or transport. All fire sprinkler water discharged in process areas would be contained and treated as process wastewater, when required. Spent Nuclear Fuel. Spent nuclear fuel would not be reprocessed. Fuel elements containing spent nuclear fuel would be stored for up to three years in water-cooled storage basins. The spent nuclear fuel pool would be equipped with an underwater canister loading system. Twelve spent nuclear fuel assemblies would be placed in fixed positions in a borated aluminum or stainless-steel basket for criticality safety. The basket would be contained in a canister whose lids would be seal-welded in place. After the 3-year cooling period, the canisters would be drained, vacuum dried, and backfilled with helium through lid penetrations in preparation for dry storage. The canisters would be transferred in a cask to the interim spent nuclear fuel storage facility. At the storage facility the canisters would be transferred into the final storage cask which would be made of precast concrete and would hold one canister each. Casks would be placed on a concrete basemat. Periodic visual inspections of the canisters and the cask vents would be required. Periodic testing for helium leaks might also be required. The facility design would have sufficient capacity to store the spent nuclear fuel for the life of the facility. The ROD (60 FR 28680) from the DOE Programmatic Spent Nuclear Fuel Management and the Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Envi- ronmental Impact Statement has the type of spent nuclear fuel generated by the ALWR destined for consolidation at SRS to await availability of a geologic repository. Transuranic Waste. The ALWR would not generate any TRU waste. Low-Level Waste. LLW would be generated by the operation of the reactor and support facilities. Process effluents would be temporarily stored in storage tanks before treatment into solid LLW that is suitable for disposal. The nonhazardous liquid effluent would then be discharged through a permitted NPDES outfall. The bulk of the solid LLW would then be generated in the reactor and tritium target processing facilities. Solid LLW would consist of contaminated equipment pieces, plastic sheeting, spent targets, spent uranium beds, and protective clothing. It would be compacted as appropriate and then disposed of in a suitable disposal facility. Mixed Low-Level Waste. No liquid mixed LLW would be generated from operating the ALWR. Solid mixed LLW may originate from wipes laden with contaminated oils and hydraulic fluids from the tritium extraction facility. Mixed LLW would be stored in an onsite RCRA-permitted storage facility until treatment in accordance with the site-specific treatment plan that was developed to comply with the Federal Facility Compliance Act of 1992. Hazardous Waste. Liquid hazardous wastes would be generated from cleaning solvents, cutting oils, vacuum pump oils, film processing fluids, hydraulic fluids from mechanical equipment, antifreeze solutions, acids, and paint. The cleaning solvent selected for use would be from a list of nonhalogenated solvents. Liquid hazardous wastes would be collected in DOT-approved containers and sent to an onsite hazardous waste accumulation area. The hazardous waste accumulation area would provide a 90-day staging capacity prior to shipment to an offsite commercial RCRA-permitted treatment, storage, and disposal facility using DOT-certified transporters. Solid hazardous wastes would be generated from nonradioactive materials such as wipes contaminated with oils, lubricants, and cleaning solvents that are used for equipment outside the main processing units. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT-approved containers and sent to a hazardous waste accumulation area for staging prior to shipment to an offsite commercial RCRA- permitted treatment, storage, and disposal facility using DOT-certified transporters. Nonhazardous Waste. Sewage wastewater would be treated in the sanitary wastewater treatment plant. Sewage wastewater would be kept separate from all industrial and process wastewaters and normally contains no radioactive wastes from the facility. The sewage wastewater would be routinely monitored for radioactive contaminants. The sewage process sludge would be disposed of in a permitted landfill. All treated effluent would be discharged to a stream or river through an NPDES outfall (wet site) or a natural drainage channel (dry site). Cooling tower blowdown would be treated and discharged to the river (wet site). Because the dry site design uses nonevaporative-type cooling towers, there are no liquid discharges resulting from blowdown. The treated effluent from the utility wastewater treatment would be discharged to the river through an NPDES outfall (wet site) or a natural drainage channel (dry site). All sludges would be disposed of in a permitted landfill. Other nonrecyclable, solid nonhazardous sanitary and industrial waste would be compacted and disposed of in a permitted landfill. The facility design includes stormwater retention ponds with the necessary NPDES monitoring equipment. Rainfall within the Limited Area and Protected Area would be collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge to the natural drainage channels (dry site) or river (wet site). If the runoff is contaminated, it would be treated in the radioactive waste treatment system. Runoff from the Property Protection Area would be discharged directly into the natural drainage channels or river. Figure (Page A-55) Figure A.2.1.3-1.-Advanced Light Water Reactor Tritium Production Process. Figure (Page A-56) Figure A.2.1.3-2.-Advanced Light Water Reactor Facility (Typical). Figure (Page A-57) Figure A.2.1.3-3.-Advanced Light Water Reactor Waste Management System (Dry Site). Figure (Page A-58) Figure A.2.1.3-4.-Advanced Light Water Reactor Waste Management System (Wet Site). Table A.2.1.3-1.-Advanced Light Water Reactor Construction Material/Resource Requirements - Consumption Material/ Resource Large Small Electrical energy 120,000 120,000 (MWh) Concrete (yd3) 380,000 200,000 Steel (tons) 68,000 50,000 Fuel (gal) 1,500,000 1,500,000 Water (gal) 200,000,000 120,000,000 Source: DOE 1995f. Table A.2.1.3-2.-Advanced Light Water Reactor Operation Utility Requirements - Consumption Utility Large Small Electrical Energy (MWh/yr) Wet site 700,000 380,000 Dry site 1,100,000 580,000 Electrical Load (MWe) Wet site 96 52 Dry site 140 75 Fuel Gas (ft3/yr) 0 0 Liquid (GPY) 200,000 110,000 Water (MGY) Wet site 16,000 7,200 Dry site 90 50 Source: DOE 1995f. Table A.2.1.3-3.-Advanced Light Water Reactor Annual Chemical Requirements - Quantity (tons) Chemical Large Small Aluminum oxide 0.9 0.9 Uranium oxide 0 0 Lithium carbonate 0.65 0.65 Stainless steel 8.5 4.65 Nitric acid 0.13 0.13 Sodium hydroxide 0 0 Sulfuric acid 0 0 Source: DOE 1995f. Table A.2.1.3-4.-Advanced Light Water Reactor (Large) Estimated Spent Nuclear Fuel and Waste Volumes - - Dry Site Wet Site - Annual Average Volume Annual Volume Annual Volume Annual Volume Annual Volume Generated From Generated From Effluent From Generated From Effluent From Construction Operations Operations Operations Operations Category (yd3) (yd3) (yd3) (yd3) (yd3) Spent Nuclear Fuel None 55 55a 55a 55a Low-Level Liquid None 24,800 None 24,800b None (5,000,000 gal) (5,000,000 gal) Solid None 710 567 710 567c Mixed Low-Level Liquid None None None None None Solid None 6 6 6 6 Hazardous Liquid Included in solid Included in solid Included in solid Included in solid Included in solid Solid 930 35 35 35 35 Nonhazardous (Sanitary) Liquid 134,000 446,000 446,000 31,100,000 31,100,000 (27,000,000 gal) (90,000,000 gal) (90,000,000 gal) (6,290,000,000 gal) (6,290,000,000 gal) Solid 15,000 6,900 2,300 6,900 2,300d Nonhazardous (Other) Liquid 2,480 Included in sanitary Included in sanitary Included in sanitary Included in sanitary (500,000 gal) Solid Included in sanitary 5,800 None 5,800f None Table A.2.1.3-5.-Advanced Light Water Reactor (Small) Estimated Spent Nuclear Fuel and Waste Volumes - - Dry Site Wet Site - Annual Average Volume Annual Volume Annual Volume Annual Volume Annual Volume Generated From Generated From Effluent From Generated From Effluent From Construction Operations Operations Operations Operations Category (yd3) (yd3) (yd3) (yd3) (yd3) Spent Nuclear Fuel None 36 36a 36a 36a Low-Level Liquid None 3,910 None 3,910b None (790,000 gal) (790,000 gal) Solid None 660 272 660 272b Mixed Low-Level Liquid None None None None None Solid None 6 6 6 6 Hazardous Liquid Included in solid Included in solid Included in solid Included in solid Included in solid Solid 850 35 35 35 35 Nonhazardous (Sanitary) Liquid 74,300 248,000 248,000 14,100,000 14,100,000 (15,000,000 gal) (50,000,000 gal) (50,000,000 gal) (2,850,000,000 gal) (2,850,000,000 gal) Solid 10,000 4,200 1,400 4,200 1,400d Nonhazardous (Other) Liquid 2,480 Included in sanitary Included in sanitary Included in sanitary Included in sanitary (500,000 gal) Solid Included in sanitary 3,500 None 3,500f None -TSAR_DOE_SECTION- A.2.1.4 Accelerator Production of Tritium TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.1 Tritium Supply A.2.1.4 Accelerator Production of Tritium Mission. The primary mission of the APT facility would be to provide the tritium supply functions performed by DOE, as well as waste treatment functions related to APT operations. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. APT would use a high-intensity beam of energetic protons produced by a linear accelerator to strike a heavy metal target. The interaction of the protons with the target material would generate a shower of neutrons which would be absorbed by surrounding helium-3 or lithium-6 atoms to produce tritium. The APT construction approach would be to build it in two phases with the first phase able to provide the steady-state tritium requirement and the second phase able to provide the baseline requirement. There are two different target designs (helium-3 and spallation-induced lithium conversion) under consideration for the baseline case which would each be based on a centrally located spallation neutron source (high Z material). The first phase of the APT would use only the helium-3 target design. The APT facility would consist of the following components: linear accelerator, beam transport and switchyard, beam stop, target area, secondary cooling, electric substations, emergency power, support, waste treatment, radioactive waste processing, and new target inspection and storage. Figure A.2.1.4-1 is a functional diagram of the APT facility, with the difference between phases noted, while figure A.2.1.4-2 depicts the APT facility site layout. Component Facility Functions Linear Accelerator. The linear accelerator would consist of a sequence of electromagnetically resonant cavities sustaining radio frequency electric fields that deliver energy to a continuous stream of proton bunches. The initial phase of the APT would use 300megahertz radio frequency energy in the low energy region and 700 megahertz radio frequency energy in the high energy region to deliver 100 milliamperes of 1 billion electron volt protons to the target chamber. The full-size APT would add another injector leg and a funnel to combine its proton beam with the other to deliver 200 milliamperes of 1 billion electron volt protons (or equivalent beam power at lower energies) to the target chamber. Beam Transport and Switchyard. At the end of the accelerator, the beam would be deflected to one of two target/blanket assemblies or a full power beam stop in the target area by a beam transport and switchyard. This system would consist of bending and focusing magnets and evacuated pipes through which the proton beam would travel. The system would also provide the correct expanded beam size and shape at the target, and switch the beam between target/blanket assemblies as required. Before the beam is directed onto the target, the beam would be sent to the full power beam stop to tune the accelerator. Beam Stop. The beam stop is a device that would be used to stop the proton beam during accelerator tuneup, or to dump the beam when necessary to divert the beam from the target area without shutting down the accelerator. The beam stop would accept the full power beam for an unlimited time while passively radiating the deposited power to an actively cooledstructure. Target Area. The target area concepts under consideration are detailed in the description of processes section. The target chamber would be located in a subterranean structure to place the target at the same elevation as the accelerator beam line. A confinement system would envelop the target area and those spaces housing the primary cooling loop components, as well as the compartments in which spent target handling, disassembly, and preparation for shipment are conducted. Tritium extraction would be dependent on the target system selected and is also discussed in the description of processes section. Secondary Cooling Facility. The secondary cooling facility would include cooling towers or cooling ponds for rejecting accelerator waste heat to the atmosphere. The major heat sources would include radio frequency power losses in the accelerator structure, heat produced in power supplies and klystrons, heat deposited in the target/blanket assembly by the proton beam, decay heat from the irradiated target, and heat produced in tritium extraction and target processing facilities. Electric Substations. The electric substations would receive electric power from the commercial grid, convert it to the required voltages for various functions, and distribute it to the accelerator, target/blanket chamber, processing systems, and other onsite facilities. Emergency Power. The emergency power facility would house diesel generators or gas turbines that would be used to provide short-term emergency power to support safety-related loads in the event of temporary failure of the offsite power supply. Support Facilities. The administration facility would house administrative, technical support, and clerical staff. It would include staff offices, a cafeteria, medical facilities, data processing, and a records center. The operations and utilities facility would provide for operational control and monitoring of the accelerator, target/blanket assembly, tritium extraction (as required by the target design), and associated support and safety systems. It would also provide for operator training, including simulators, serve as the center for site safeguards and security, provide personnel monitoring and access control, and house the safety related batteries providing an uninterruptible power supply for critical subsystems. The maintenance facility would provide shop and service areas for maintaining all accelerator, target/blanket, and processing system components. It would also provide warehousing and materials handling for consumables and spare parts for these systems. Waste Treatment Facility. The waste treatment facility would receive and treat all site-generated sanitary sewage for release to the environment in accordance with an NPDES permit. Radioactive Waste Processing Facility. This facility would receive all solid, liquid, and gaseous radioactive waste for storage, treatment, and packaging for either release or disposal at an appropriate permanent waste disposal facility. Water Treatment Facility. This facility would treat water to meet the required quality for the various plant cooling systems. The fire protection system would tap off the facility water supply upstream of the water treatment facility. New Target Inspection and Storage Facility. This facility would provide for the unloading, receipt, inspection and storage of new target/blanket assemblies fabricated offsite. The facility would also load packaged spent targets onto transport equipment for removal to the final disposal site. Description of Processes Accelerator Operation. The accelerator would operate in continuous wave mode, meaning that the cavities would be excited 100 percent of the time rather than part of the time as in a pulsed system. Several types of accelerating structure geometries would be used to couple radio frequency energy to the beam over different velocity ranges. At full energy, the protons would travel at approximately 0.9times the speed of light. The radio frequency field in the accelerating structures would be generated by a series of 1 megawatt (MW) klystrons (high power electron tubes) distributed along the length of the accelerator. For most of its 0.75 mile length, the accelerator would be located in a sealed, concrete-lined tunnel buried under about 40 feet of earth. A similar length building on the surface would house the klystrons, power supplies, and controls. The magnetic transport system would direct the proton beam from the end of the accelerator to the tritium production target/blanket assembly. When the beam is directed to one of the target assemblies, the other would be undergoing service or would be on standby. When the beam is not directed at a target, it would be directed to the full-power beam stop through the diagnostic beam channel used for tuning the accelerator. At the ends of the transport channels, the beam would be expanded to a large, rectangular, uniform current distribution form to match the power density requirements of the targets. Multiple redundant sensors distributed throughout the accelerator and beam lines would act within extremely short time frames to determine if the beam were out of acceptable limits in terms of position, energy, size, intensity distribution, etc. Malfunctions or out of tolerance values for any of these devices would terminate the beam. Heat transport loops would be needed for the target blanket system. The primary cooling loop would transport the heat generated in the target and blanket (separate cooling loops) to a secondary cooling loop. The secondary cooling loop would transfer heat through a heat exchanger system to the atmosphere. Tritium Production. The APT tritium production process would be dependent on the choice of APT design and target assembly employed. Two target concepts currently under consideration for the full-size APT are: helium-3 and spallation-induced lithium conversion. Helium-3-The target chamber would house a low-pressure, low-temperature target/blanket assembly consisting of a water-cooled inconel beam entrance window followed by a neutron-producing target of tungsten as depicted in figure A.2.1.4-3. That target would be cooled by circulating heavy water moderator with circulating, contained helium-3. The assembly would have lead in a surrounding annular region to provide neutron multiplication and moderation. About 60 percent of the tritium would be produced in the blanket region of the system; the remainder would be produced in a separate helium-3 volume in the spallation target. The purpose of the second helium-3 gas region is to act as a decoupler to reduce absorption of neutrons back into the tungsten source region. Tritium would be continuously removed from both regions by taking a side stream of the circulating helium-3 gas into a nearby processing area. Spallation-Induced Lithium Conversion-The target area, depicted in figure A.2.1.4-4 would contain a target system consisting of a heavy water-cooled lead spallation target, and light water-cooled lithium-aluminum target separated from the accelerator by a beam entrance window. The assembly would operate at relatively low temperature and low pressure. High energy protons would induce spallation of the lead target, thereby generating neutrons. Heavy water coolant would remove the heat of spallation and moderate the neutrons to thermal energy. The lithium-6 isotope within the light water-cooled lithium-aluminum target rods would capture the thermal neutrons and produce tritium. Tritium Extraction. The extraction process would be dependent on the APT design and target concept employed. Helium-3-In this process, a small slip-stream would be taken off the gas circulating through the heavy water source and blanket regions, and hydrogen and tritium would be extracted using a permeable membrane technique. The helium-3 would then be recirculated to the target area. The extracted hydrogen and tritium would be separated by cryogenic distillation. The tritium would be of very high purity. An area would be required for cooling and radiation shielding of spent tungsten removed from the target cavity. Remote handling apparatus would load packaged targets into shipping casks and onto rail or truck vehicles. Spent tungsten targets would go to existing waste management facilities for final disposal. The estimated volume of this material is small, typically on the order of a few cubic meters per year of operation. Figure A.2.1.4-5 provides the flow diagram for the helium-3 process. Spallation-Induced Lithium Conversion-Following irradiation, the targets would be removed and transported to the disassembly basin, where the target tubes would be allowed to cool for approximately 6 months so that short-lived isotopes could decay. After cooling, the target tubes would be washed within the shielded transport cask, placed in a crucible, inserted into a furnace, and heated to drive off the gases, including the tritium product. The gas stream from the furnace, mainly helium and hydrogen isotopes, including tritium, would pass through a diffusing process to separate the hydrogen isotopes from helium and other impurities. This process would cycle the gas between two beds of treated diatomaceous earth used as a filter aid. One bed would be heated to drive off the gases; the other bed would be cooled to promote gas absorption. Figure A.2.1.4-6 depicts the flow diagram for the spallation-induced lithium conversion process. Facility Utilities. Facility construction and operation utility requirements are shown in tables A.2.1.4-1 and A.2.1.4-2, respectively. Chemicals Required. Table A.2.1.4-3 depicts chemical resources required during operations. Personnel Requirements. Construction of the APT would have a peak employment of 2,760 construction workers. Approximately 6,380 worker years would be needed during the 5-year construction period. In the phased approach, the majority of civil engineering construction for both phases would be completed during initial construction. The upgrade from the first phase to the Full APT would involve installation of the equipment in the second injection line, the funnel, and additional radio frequency power systems. Operation of either APT facility would require 624 workers with 258 of these badged for radiation detection. Transportation. Interfacility transfers would be made by rail or truck transport. Truck service would be needed for intrafacility movements. Helium would be received in bulk by rail and unloaded into cylinders for storage. Waste Management. Solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Uncontaminated wastewater would be used for soil compaction and dust control, and excavated soil would be used for grading and site preparation. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes generated during construction would consist of such materials as waste adhesives, oils, cleaning fluids, solvents, and coatings. Hazardous waste would be packaged in DOT-approved containers and shipped offsite to commercial RCRA-permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The facility design incorporates waste minimization and pollution prevention. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycling. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials which contribute to the generation of hazardous or mixed waste. Production processes would be configured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future D&D considerations have also been incorporated into the design. Tables A.2.1.4-4 and A.2.1.4-5 present the estimated annual waste volumes from the Full APT facility for the two target designs during construction and operation. Table A.2.1.4-4 presents the estimated waste volumes for the Phased APT. All process and waste systems would be housed in enclosures designed to contain and allow recovery of tritium leaks. A gaseous waste system would recover residual tritium before releasing the waste to the atmosphere. Solid and liquid waste streams would be routed to the waste management system. Solid wastes would then be characterized and segregated into LLW, hazardous, and mixed wastes, then treated into a form suitable for disposal or storage within the facility. Liquid wastes would be treated onsite to reduce hazardous/toxic and radioactive elements before discharge or transport. All fire sprinkler water discharged in process areas would be contained and treated as process wastewater, when required. Spent Nuclear Fuel. The APT facility would not generate any spent nuclear fuel. Transuranic Waste. The APT facility would not generate any TRU waste. Low-Level Waste. LLW would be generated by operations in the target and blanket area and the target tritium-extraction facility. The process stripper would remove any tritium from process streams; thus eliminating liquid LLW generation. The bulk of the solid LLW would be generated in the tritium extraction facility. Solid LLW from the accelerator would result from klystron and ion pump refurbishment and cooling system maintenance. LLW generated in the spallation-induced lithium conversion (or lithium-aluminum) target area would include treatment residues, spent resin filters, piping and pressure tubes, the window, and job control wastes. From the helium-3 target area solid LLW would include treatment residues, spent resins and filters, inconel, zircaloy, and job control wastes. The lithium-aluminum target tritium-extraction system would generate job control wastes, crucibles, spent lithium-aluminum melts, piping, valves, and filters. Solid LLW from the helium-3 target tritium extraction system would include job control wastes, piping, valves, and filters. Solid LLW would be compacted as appropriate and then disposed of in a suitable disposal facility. Mixed Low-Level Waste. No liquid mixed LLW would be generated from the APT facility. Solid mixed LLW from the accelerator would include solvent rags and discarded batteries. The lithium-aluminum target area would generate solvent rags, lead aluminum targets, discarded batteries, and paint while the tritium target extraction facility sources of mixed LLW would include contaminated batteries, aerosol cans, and spent solvents. After removal from the target area, the spent lead targets would be stored in pools for one year. In this time the activity would decrease from 10 to 20 million curies (Ci) to 100,000Ci. The targets would then be processed for disposal as solid mixed LLW. Mixed LLW from the helium-3 target area would include discarded batteries and paint with the tritium target extraction facility generating contaminated batteries, aerosol cans, and spent solvents. Mixed waste would be processed in accordance with the site-specific treatment plan that was developed to comply with the Federal Facility Compliance Act of 1992. Hazardous Waste. Depending on the treatment technology selected, a small quantity (< 1 gal) of liquid hazardous wastes could be generated from the treatment of mixed LLW. Solid hazardous wastes would be generated from nonradioactive materials such as wipes contaminated with oils, lubricants, and cleaning solvents that are used for equipment outside the main processing units. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT-approved containers and sent to an onsite hazardous waste storage facility for staging prior to shipment to a commercial RCRA-permitted treatment, storage, and disposal facility using DOT- certified transporters. Nonhazardous Waste. Sewage wastewater would be treated in the waste treatment facility. Sewage waste-water would be kept separate from all industrial and process wastewaters and would normally contain no radioactive wastes from the facility. The sewage wastewater would be routinely monitored for radioactive contaminates. The sludge and other nonrecy- clable solid sanitary and industrial wastes would be disposed of in a permitted landfill. The treated effluent would be discharged to the river through an NPDES outfall (wet site) or a natural drainage channel (dry site). Cooling tower blowdown would be treated and discharged to the river (wet site) or recycled for reuse (dry site). The treated effluent from the utility wastewater treatment would be discharged to the river through an NPDES outfall (wet site) or a natural drainage channel (dry site). All sludges would be disposed of in a permitted landfill. The facility design includes stormwater retention facilities with the necessary NPDES monitoring equipment. Rainfall is collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge to the natural drainage channels (dry site) or river (wet site). If the runoff is contaminated, it would be treated in the radioactive waste treatment system. Runoff outside the facility area would be discharged directly into the natural drainage channels or river. Figure (Page A-67) Figure A.2.1.4-1.-Accelerator Production of Tritium Facility Functional Layout. Figure (Page A-68) Figure A.2.1.4-2.-Accelerator Production of Tritium Facility Site Layout (Typical). Figure (Page A-69) Figure A.2.1.4-3.-Accelerator Production of Tritium/Helium-3 Target System. Figure (Page A-70) Figure A.2.1.4-4.-Accelerator Production of Tritium/Spallation-Induced Lithium Conversion Target System (Exploded). Figure (Page A-71) Figure A.2.1.4-5.-Flow Diagram for Accelerator Production of Tritium/Helium-3 Target. Figure (Page A-72) Figure A.2.1.4-6.-Flow Diagram for Accelerator Production of Tritium/Spallation-Induced Lithium Conversion Target. Table A.2.1.4-1.-Accelerator Production of Tritium Construction Material/Resource Requirements - Consumption Material/ Full Phased Resource Electrical energy 40,000 40,000 (MWh) Concrete (yd3) 275,000 275,000 Steel (tons) 61,495 55,820 Fuel (gal) 2,110,000 2,110,000 Water (gal) 41,700,000 41,700,000 Source: SNL 1995a. Table A.2.1.4-2.-Accelerator Production of Tritium Operation Utility Requirements Consumption Utility Full Phased Electrical Energy 3,740,000 2,400,000 (MWh/yr) Electrical Load (MWe) 550 355 Fuel Gas (ft3 per year) 0 0 Liquid (GPY) 13,200 13,200 Water (MGY) 1,200 770 Source: SNL 1995a. Table A.2.1.4-3.-Accelerator Production of Tritium Annual Chemical Requirements - Quantity (tons) Chemical Full Phased Aluminum 50 13.3 Uranium metal 0.4 0 Lithium metal 1.2 0 Stainless steel 32 7.1 Sodium hydroxide 0.22 0.14 Sulfuric acid 1,375 880 Source: SNL 1995a. Table A.2.1.4-4.-Accelerator Production of Tritium (Helium-3 Target) Estimated Waste Volumes - - - Dry Site Wet Site - Annual Average Volume Annual Volume Generated Annual Volume Generated From Construction From Operations Effluent From Operations Category (yd3) (yd3) (yd3) (yd3) Low-Level Liquid None None None None Solid None 72 57 57a Mixed Low-Level Liquid None None None None Solid None 3.7 3.1 3.1 Hazardous Liquid Included in solid None 0.0003 0.0003 (0.06 gal) (0.06 gal) Solid 13 1.2 1.2 1.2 Nonhazardous (Sanitary) Liquid 1,570 1,210,000 <1,000 1,210,000 (317,000 gal) (245,000,000 gal) (<200,000 gal) (245,000,000 gal) Solid 5,500 1,240 413 413d Nonhazardous (Other) Liquid Included in sanitary Included in sanitary Included in sanitary Included in sanitary Solid Included in sanitary None Included in sanitary None Table A.2.1.4-5.-Accelerator Production of Tritium (Spallation-Induced Lithium Conversion Target) Estimated Waste Volumes - - - Dry Site Wet Site - Annual Average Volume Annual Volume Generated Annual Volume Generated From Construction From Operations Effluent From Operations Category (yd3) (yd3) (yd3) (yd3) Low-Level Liquid None None None None Solid None 544 221 221a Mixed Low-Level Liquid None None None None Solid None 6.8 3.9 3.9 Hazardous Liquid Included in solid None 0.003 0.003 (0.6 gal) (0.6 gal) Solid 13 2.5 2.5 2.5 Nonhazardous (Sanitary) Liquid 1,570 1,210,000 <1,000 1,210,000 (317,000 gal) (245,000,000 gal) (<200,000 gal) (245,000,000 gal) Solid 5,500 1,240 413 413d Nonhazardous (Other) Liquid Included in sanitary Included in sanitary Included in sanitary Included in sanitary Solid Included in sanitary None Included in sanitary None Table A.2.1.4-6.-Phased Accelerator Production of Tritium (Helium-3 Target Only) Estimated Waste Volumes - - - Dry Site Wet Site - Annual Average Volume Annual Volume Generated Annual Volume Generated From Construction From Operations Effluent From Operations Category (yd3) (yd3) (yd3) (yd3) Low-Level Liquid None None None None Solid None 68 54 54a Mixed Low-Level Liquid None None None None Solid None 3 2.5 2.5 Hazardous Liquid Included in solid None 0.0003 0.0003 (0.06 gal) (0.06 gal) Solid 13 1.2 1.2 1.2 Nonhazardous (Sanitary) Liquid 1,570 789,000 <1,000 789,000 (317,000 gal) (159,000,000 gal) (<200,000 gal) (159,000,000 gal) Solid 5,500 1,240 413 413d Nonhazardous (Other) Liquid Included in sanitary Included in sanitary Included in sanitary Included in sanitary Solid Included in sanitary None Included in sanitary None -TSAR_DOE_SECTION- A.2.2 Tritium Recycling TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.2 Tritium Recycling Tritium recycling would either be collocated with the tritium supply facility or be located at SRS. If the tritium supply facility is collocated with a new recycling facility, there may be a potential for some efficiencies to be realized through the sharing of common facilities such as sanitary wastewater treatment. However, these efficiencies are not considered large enough to affect either a technology or site selection and thus, are not analyzed in this PEIS. At SRS, an upgrade of the existing recycling facilities would be implemented instead of constructing a new facility. The descriptions of the new recycling facilities, and the upgrades to the existing facilities at SRS, are in the following sections. -TSAR_DOE_SECTION- A.2.2.1 New Recycling Facility TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.2 Tritium Recycling A.2.2.1 New Recycling Facility Missions. The primary mission of the new tritium recycling facility would be to provide the full range of tritium processing, recycling, and packaging functions performed by DOE, as well as associated testing and waste management functions. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. The tritium recycling facility would consist of two major buildings and associated support facilities as depicted in figure A.2.2.1-1. The tritium recycling building would house the systems that process inventories of tritium and tritium contaminated items along with the plant operations center. The auxiliary building, the other major building, would house the nontritium functions as well as the emergency power diesel generators for the tritium recycling building. The plant support buildings would provide general services such as administration, cafeteria, fire protection, medical, environmental, safety, and health. Component Facility Functions. Many modern nuclear weapons employ tritium gas or a mixture of tritium and deuterium gases, contained in reservoirs, to improve weapons performance. Radioactive decay reduces the reservoir tritium content, which means that stockpile reservoirs must be replaced periodically. The residual tritium from the returned res- ervoirs would be recovered for recycle, and the empty reservoir would be reclaimed and reused if possible. At the tritium recycling facility, reservoirs would be subjected to stringent environmental and performance tests to their integrity under all service con- ditions. The facility also would be the source for tritium used for commercial applications and for fusion research and development. Description of Processes. The tritium recycling facility processes are depicted in figure A.2.2.1-2. Tritium would be received in reservoirs returned from the field, or as virgin tritium from an extraction facility associated with a tritium supply facility. The reservoirs would be unpacked from their shipping containers and stored in a vault prior to being emptied. The reservoirs would be emptied and the contained gases processed to separate the hydrogen isotopes from other gases, primarily helium-3 (a product of the radioactive decay of tritium). The hydrogen isotopes would be separated into tritium, deuterium, and protium (normal hydrogen). Tritium and deuterium would be used to prepare a specified isotopic mixture for the reservoirs. Protium would be discharged to the stack. The empty reservoirs would be reclaimed, if possible, and refilled. Reservoirs that could not be reclaimed would be handled as solid LLW. The helium-3 would be purified to remove any residual tritium and other contaminants prior to packaging as a by-product. The reclaimed and/or new reservoirs would be filled with specific mixtures of gases. These gas mixtures could be obtained by using recovered gas of the proper specification, adding pure isotopes to the mix of recovered gases, or blending the pure isotopes. Once the reservoirs were filled, they would be sealed with a closure weld, trimmed, surface decontaminated, leak tested, inspected, marked, assayed for tritium content, and fitted with an explosively-actuated squib valve, if required. The reservoirs would then be placed in a storage vault until they were packaged and sent to the field for limited life component exchange in a weapon system. A sampling of the newly-filled reservoirs would be placed in the life storage area for surveillance operations. As these reservoirs age, they would be examined and tested to confirm predicted behavior and to ensure the integrity and function of the reservoirs in the field. Surveillance operations would include environmental testing, functional testing, calorimetry, flow testing, and burst testing. These tests would be performed to evaluate the behavior of the selected reservoirs (including reservoirs returned from the field or in life storage) under test conditions. The tritium recycling facility would also provide capabilities for metallurgical studies and gas sample composition analyses. The tritium recycling facility also would have the capability to fill containers with specified gas mixtures for commercial use and research applications. Facility Utilities. Facility construction and operation utility requirements are shown in tables A.2.2.1-1 and A.2.2.1-1, respectively. Chemicals Required. Table A.2.2.1-3 depicts chemical resources required during operation. Personnel Requirements. Construction of the tritium recycling facility would have a peak employment of 335 construction workers. Approximately 992 worker years would be needed during the 4-year construction period. Operation requirements of the tritium recycling facility would be 910 workers total with 400 of these badged for radiation detection. Transportation. Interfacility transfers would be made by DOE contract cargo carriers. Truck service would be needed for intrafacility transport. Waste Management. The solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Uncontaminated wastewa- ter would be used for soil compaction and dust control, and excavated soil would be used for grading and site preparation. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes such as waste adhesives, oils, and solvent rags would be packaged in DOT approved containers and shipped offsite to commercial RCRA-permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The facility design considers and incorporates waste minimization and pollution prevention. The facility recovery system would provide the necessary equipment to recover, purify, and package helium-3, and to recover tritium from tritiated wastes generated within the facility, possible offsite sources, and the glove box inert atmospheres. The recovery system consists of strippers, tritiated aqueous recovery, tritium scrap recovery, and helium-3 recovery. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycle. After evacuating and flushing with argon, empty reservoirs would be reclaimed, if necessary, and refilled. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials that contribute to the generation of hazardous or mixed waste. Tritium recycling operations would be configured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future D&D considerations would also be incorporated into the design. Table A.2.2.1-4 presents the estimated annual waste volumes from the tritium recycling facility during construction and operation. One can expect a significant reduction in waste generation once the facility is built and operational. Solid and liquid waste streams would be routed to the waste management system. Figure A.2.2.1-3 depicts the waste management system for the tritium recycling facility at a dry site, while figure A.2.2.1-4 illustrates the waste management system for a wet site. Solid wastes would be characterized and segregated into LLW, hazardous, and mixed wastes, then immobilized and packaged for disposal or storage within the facility. Liquid wastes would be treated onsite to reduce hazardous/toxic and radioactive elements before discharge or transport. All fire sprinkler water discharged in process areas during and after a fire would be con- tained, monitored, sampled, and if required, retained until disposed. Spent Nuclear Fuel. The tritium recycling facility would not generate any spent nuclear fuel. Transuranic Waste. The tritium recycling facility would not generate any TRU waste. Low-Level Waste. Tritium-contaminated wastewater would be generated by hydraulic burst testing and by the strippers. The tritium-contaminated wastewater is treated in the tritium recovery system to recover the tritium. Strippers are designed to remove hydrogen isotopes, water vapor, and methane from the process waste streams and the glove box recirculating inert atmosphere. Strippers consist of four types (process, primary, secondary, and purge) and would utilize a similar process to recover residual tritium from tritiated gases. The process stripper would be designed to remove hydrogen isotopes and water vapor from process waste streams by oxidizing elemental hydrogen isotopes and trapping the oxides on molecular sieves (Z-beds) prior to discharging to the stack. After the bed becomes saturated, it would be heated with the water vapor then being routed to tritium aqueous recovery to recover hydrogen isotopes. The primary stripper would be designed to continuously process and recirculate the glove box nitrogen or argon atmosphere in order to maintain acceptable tritium, water vapor, and oxygen levels inside the glove box. The secondary stripper would have the same design as a standby unit to process nitrogen gas from glove boxes when a tritium leak occurs, and also to serve as a spare for the primary stripper. The purge stripper employs the same type of process used in the other strippers to remove hydrogen isotopes and water vapor from the glove box airlock gas prior to discharging to the stack. The purge stripper also processes the blowdown from the primary and secondary strippers. Two primary tritiated aqueous recovery processes (reactive metal bed and electrolysis) would be utilized in the tritium recycling facility. It has been assumed that the reactive metal bed process would handle the stripper Z-bed tritiated water and the electrolysis process would handle the tritiated water from outside sources. Tritium scrap recovery would take tritiated solid wastes such as metals and pieces from reservoir reclamation, metallography, and burst testing; reduce them in size; and then transfer them to a container which would be baked in a vacuum oven. The tritiated off-gas would be evacuated to a storage tank for accountability of tritium prior to transfer to the process stripper. Any pyrophoric materials would be oxidized to allow for chemical stability. The solid components would then be cooled in the oven and managed in accordance with the solid LLW procedures outlined below. Solid LLW would consist of hydride beds, U-beds, Z-beds, stripper catalysts, retired process equipment, glove box tools wastes, process metal residues, unreclaimed reservoirs, and low-specific activity waste. The LLW solids are reduced in size, immobilized, stabilized, packaged, and staged in the Long-Term Waste Storage Building while awaiting shipment to a suitable LLW disposal facility. Mixed Low-Level Waste. Liquid mixed LLW would originate from potentially-contaminated used compressor and glove box bubblers (sealpot) oil. Solid mixed LLW would include solvent rags and oil-contaminated materials with trace quantities of tritium. Mixed LLW would be stored in the RCRA-permitted Long-Term Storage Building onsite until treatment and disposal in accordance with the site-specific treatment plan that is being developed to comply with the Federal Facility Compliance Act of 1992. Hazardous Waste. The cleaning solvents selected would be from a list of nonhalogenated solvents to the extent practicable. Solid hazardous wastes would be generated from nonradioactive materials such as solvent rags, containers with residual hazardous materials, and lead-acid batteries. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT-approved containers and sent to the Waste Management Building for staging prior to shipment to a commercial RCRA-permitted treatment, storage, and disposal facility using DOT- certified transporters. Nonhazardous Waste. Sewage wastewater would be treated in the facility sanitary wastewater treatment facility. Sewage wastewater would be kept separate from all industrial and process wastewaters and normally contains no radioactive wastes from the facility. The sewage wastewater would be routinely monitored for radioactive contaminants. The sludge would be disposed of in a permitted landfill. As illustrated in figure A.2.2.1-3, at the dry site the treated effluent would be recycled to the cooling tower and used as makeup, while figure A.2.2.1-4 shows the treated effluent at the wet site being discharged into the river through a permitted NPDES outfall. Process wastewater would be treated at the process wastewater treatment facility. The sludge would be disposed of in a permitted sanitary landfill and the water would be reclaimed. The facility design includes stormwater retention ponds with the necessary NPDES monitoring equipment. Rainfall within the main facility area would be collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge to the natural drainage channels (dry site) or river (wet site). If the runoff is contaminated, it would be treated in the process wastewater treating system. Runoff from the site outside of the main facility area would be discharged directly into the natural drainage channels or river. As depicted in figure A.2.2.1-3, at a dry site, the utility wastewater treatment system collects blowdown from the steam system and tower cooling water system. The blowdown water would be filtered and treated by conventional water treatment methods. The treated effluent serves as makeup to the tower cooling water system. The reject water would be evaporated and the condensate would be used for boiler makeup. The concentrated watery solution would be sent to a spray drier and the dry solids would be disposed of in a permitted landfill. As shown in figure A.2.2.1-4, at a wet site, the blowdown from the cooling tower would be discharged after treatment directly into the river through an NPDES outfall. Solid industrial, clean-shredded metal, after sanitization and demilitarization, and trash are collected and sent to a permitted landfill. Figure (Page A-81) Figure A.2.2.1-1 New Tritium Recycling Facility (Typical). Figure (Page A-82) Figure A.2.2.1-2.-New Tritium Recycling Facility Processes. Figure (Page A-83) Figure A.2.2.1-3.-New Tritium Recycling Waste Management System Process (Dry Site). Figure (Page A-84) Figure A.2.2.1-4.-New Tritium Recycling Waste Management System Process (Wet Site). Table A.2.2.1-1.-New Tritium Recycling Facility Construction Material/Resource Requirements Material/Resources Consumption Electrical energy (MWh) 10,000 Concrete (yd3) 32,000 Steel (tons) 5,600 Fuel (gal) 260,000 Water (gal) 6,100,000 Source: DOE 1995g. Table A.2.2.1-2.-New Tritium Recycling Facility Operation Utility Requirements Utility Consumption Electrical Energy (MWh/yr) 88,000 Electrical Load, (peak) (MWe) 16 Fuel Gas (ft3/yr) 7,000,000 Liquid (GPY) 50,000 Water (MGY) Wet site 37 Dry site 14 Source: DOE 1995g. Table A.2.2.1-3.-New Tritium Recycling Facility Annual Chemical Requirements Chemical Quantity (lb) Solid Adhesives, lubricants, and paints 360 Catalyst and molecular sieves 1,200 Depleted uranium (water 2,800 decomposer) Other 1,700 Liquid Organic solvents 3,700 Petroleum oils 600 Water treatment chemicals 9,800 Other 3,700 Gaseous Argon 190,000 Helium 500 Nitrogen 3,500,000 Other 8,900 Table A.2.2.1-4.-New Tritium Recycling Facility Estimated Waste Volumes - - Dry Site Wet Site - Annual Average Volume Annual Volume Annual Volume Annual Volume Annual Volume Generated From Generated From Effluent From Generated From Effluent From Construction Operations Operations Operations Operations Category (yd3) (yd3) (yd3) (yd3) (yd3) Low-Level Liquid None None None None None Solid None 350 117 350 117a Mixed Low-Level Liquid None 0.03 (6 gal) 0.03 (6 gal) 0.03 (6 gal) 0.03 (6 gal) Solid None 2 Liquid Included in solid None None None None Solid 0.5 1 1 1 1 Nonhazardous (Sanitary) Liquid 4,460 70,800 Nonee 119,000 119,000 (900,000 gal) (14,300,000 gal) (24,000,000 gal) (24,000,000 gal) Solid 163 7,400 2,470 7,400 2,470e Nonhazardous (Other) Liquid Included in sanitary Included in sanitary None Included in sanitary Included in sanitary Solid Included in sanitary 6,400 None 6,400f None -TSAR_DOE_SECTION- A.2.2.2 Tritium Recycling Facilities Upgrades at Savannah River Site TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.2 Project Descriptions A.2.2 Tritium Recycling A.2.2.2 Tritium Recycling Facilities Upgrades at Savannah River Site Missions. The primary mission of the upgraded tritium recycling facility would be to provide the full range of tritium processing, recycling, and packaging functions performed by DOE, as well as associated testing and waste management functions. Assumptions. The assumptions used in the design of this facility are listed in section A.2. General Functions and Layout. The tritium recycling facilities are centered on the new Replacement Tritium Facility (Building 233-H) and associated support facilities as depicted in figure A.2.2.2-1. The Replacement Tritium Facility was designed for gas handling operations (such as filling and emptying reservoirs), product separation, and enrichment activities. Support activities carried out in the following buildings are described by their assigned names: 232-H Tritium Extraction, Concentration, and Enrichment Facility; 232-1H Tritium Construction Pipe Shop; 232-H Maintenance Shop; 234-H Tritium Reservoir Finishing, Packaging, and Shipping Facility; 235-H Office Building; 236-H Burst Test Facility; 238-H Reservoir Reclamation Facility; 249-H Replacement Tritium Facility Support; and 720-H Central Alarm Station. The proposed unconsolidated upgrade is designed to meet DOE Natural Phenomenon Hazard Requirements affecting buildings 232-H, 232-1H, 238-H, and 249-H. These upgrades would involve the addition of wall bracing and cross bracing to beams, strengthening some exterior walls, and reinforcing building frames. Building 232-H would also require an anchor for the service area roof slab and an upgrade of the radiation control and monitoring system. Building 234-H upgrades include highly invulnerable encased safes for reservoir storage to protect the tritium filled reservoirs during a high wind and earthquake. A consolidated upgrade of SRS tritium recycling facilities is also possible. The consolidated upgrade entails the relocation of all tritium processing and handling functions from Building 232-H to buildings 233-H and 234-H in addition to the unconsolidated upgrade modifications. This upgrade would allow Building 232-H to be closed. Component Facility Functions. Many modern nuclear weapons employ tritium gas contained in reservoirs to improve weapon performance. Radioactive decay reduces the reservoir tritium content, which means that stockpile reservoirs must be replaced periodically. The residual tritium from the returned reservoirs is recovered for recycle, and the empty reservoir reclaimed and reused if possible. Reservoirs are subjected to stringent environmental and performance tests to ensure their integrity under all service conditions. The facility also would be the source for tritium used for commercial applications and for fusion research and development. Description of Processes. The tritium recycling facilities processes depicted in figure A.2.2.2-2 would be the same for both the unconsolidated and consolidated upgrades. Tritium would be received in reservoirs returned from the field, or as virgin tritium from an extraction facility associated with a tritium supply facility. The reservoirs would be unpacked from their shipping containers and stored in a vault prior to being emptied. The reservoirs would be emptied and the contained gases processed to separate the hydrogen isotopes from other gases, primarily helium-3 (a product of the radioactive decay of tritium). The hydrogen isotopes would be separated into tritium, deuterium, and protium (normal hydrogen). Tritium and deuterium would be used to prepare the reservoirs. Protium would be discharged to the stack. The empty reservoirs would be reclaimed, if possible and refilled. Reservoirs that could not be reclaimed would be handled as solid LLW. The helium-3 would be purified to remove any residual tritium and other contaminants prior to packaging as a by-product. The reclaimed and/or new reservoirs would be filled. The gas could be obtained by using recovered gas of the proper specification, adding pure isotopes to the mix of recovered gases, or blending the pure isotopes. Once the reservoirs have been filled, they would be sealed with a closure weld, trimmed, surface decontaminated, leak tested, inspected, marked, assayed for tritium content, and fitted with an explosively-actuated squib valve, if required. The reservoirs would then be placed in a storage vault until they were packaged and sent to the field for limited life component exchange in a weapon system. A sampling of the newly-filled reservoirs would be placed in the life storage area for surveillance operations. As these reservoirs age, they would be examined and tested to confirm predicted behavior and to ensure the integrity and function of the reservoirs in the field. Surveillance operations would include environmental testing, functional testing, calorimetry, flow testing, and burst testing. These tests would be performed to evaluate the behavior of the selected reservoirs (including reservoirs returned from the field or in life storage) under test conditions. The tritium recycling facilities would also provide capabilities for metallurgical studies and gas sample composition analyses. The tritium recycling facilities would have the capability to fill containers with specified gas mixtures for commercial use and research applications. Facility Utilities. Facility construction and operation utility requirements are shown in tables A.2.2.2-1 and A.2.2.2-2, respectively. Chemicals Required. Table A.2.2.2-3 depicts the chemical resources required during operation. Personnel Requirements. Upgrade of the tritium recycling facilities would have a peak employment of 26 construction workers for the unconsolidated upgrade and 36 workers for the consolidated upgrade. Approximately 62 worker years would be needed for the unconsolidated upgrade and 91worker years for the consolidated upgrade during the 3-year construction period. Operation under the unconsolidated upgrade would require 970 workers of which 400 workers would be badged. The consolidated upgrade would require 910 workers of which 400 would be badged. Transportation. Interfacility transfers would be made by DOE contract cargo carriers. Truck service would be needed for intrafacility transport. Waste Management. The solid and liquid nonhazardous wastes generated during construction would include concrete and steel construction waste materials and sanitary wastewater. The steel construction waste would be recycled as scrap material before completing construction. The remaining nonhazardous wastes generated during construction would be disposed of as part of the construction project by the contractor. Wood, paper, and metal wastes would be shipped offsite to a commercial contractor for recycling. Hazardous wastes generated during construction would consist of such materials as waste adhesives, oils, cleaning fluids, solvents, and coating. Hazardous waste would be packaged in DOT-approved containers and shipped offsite to commercial RCRA-permitted treatment, storage, and disposal facilities. No radioactive waste would be generated during construction. The upgrade design considers and incorporates waste minimization and pollution prevention. The recovery system would provide the necessary equipment to recover, purify, and package helium-3, and to recover tritium from tritiated wastes generated within the facility, possible offsite sources, and the glovebox inert atmospheres. The recovery system consists of strippers, Z-bed recovery, tritium scrap recovery, and helium-3 recovery. Activities that generate radioactive and hazardous wastes would be segregated, where possible, to avoid the generation of mixed wastes. Where applicable, treatment to separate radioactive and nonradioactive components would be performed to reduce the volume of mixed wastes and provide for cost-effective disposal or recycle. After evacuating and flushing with argon, empty reservoirs would be reclaimed, if necessary, and refilled. To facilitate waste minimization, where possible, nonhazardous materials would be substituted for those materials that contribute to the generation of hazardous or mixed waste. Tritium recycling operations would be configured with minimization of waste production given high priority. Material from the waste streams would be treated to facilitate disposal as nonhazardous wastes, where possible. Future decontamination and decommissioning con- siderations have also been incorporated into the design. Table A.2.2.2-4 presents the estimated annual waste volumes from the upgraded tritium recycling facilities at SRS during construction and operation. One can expect a reduction in waste generation and effluents once the facility is upgraded. As depicted in figure A.2.2.2-2, solid and liquid waste streams would be routed to the waste management system. Solid wastes would be characterized and segregated into LLW, hazardous, and mixed wastes, then immobilized and packaged for disposal. Liquid wastes would be neutralized, precipitated, and volume reduced via evaporation. The sludge would be immobilized and packaged for disposal. All fire sprinkler water discharged in process areas during and after a fire would be contained, monitored, sampled, and, if required, retained until it could be disposed of. Spent Nuclear Fuel. The upgraded tritium recycling facilities would not generate any spent nuclear fuel. Transuranic Waste. The upgraded tritium recycling facilities would not generate any TRU waste. Low-Level Waste. Liquid LLW would be generated by hydraulic burst testing and by the strippers; however, the tritium is recovered by Z-bed recovery. Thus, there is no solidification of any liquid LLW. Strippers would be designed to remove hydrogen isotopes, water vapor, and methane from the process waste streams and the glovebox recirculating inert atmosphere. Strippers consist of four types (process, primary, secondary, and purge) and would utilize a similar process to recover residual tritium from tritiated gases. The process stripper would be designed to remove hydrogen isotopes and water vapor from process waste streams by oxidizing elemental hydrogen isotopes and trapping the oxides on molecular sieves (Z-beds) prior to discharging to the stack. After the bed becomes saturated, it would be heated and the water vapor then routed to tritium aqueous recovery to recover hydrogen isotopes. The primary stripper would be designed to continuously process and recirculate the glovebox nitrogen or argon atmosphere in order to maintain acceptable tritium, water vapor, and oxygen levels inside the glovebox. The secondary stripper would have the same design as a standby unit to process nitrogen gas from gloveboxes when a tritium leak occurs, and also to serve as a spare for the primary stripper. The purge stripper employs the same type of process used in the other strippers to remove hydrogen isotopes and water vapor from the glovebox airlock gas prior to discharging to the stack. The purge stripper also processes the blowdown from the primary and secondary strippers. Tritium scrap recovery would take tritiated solid wastes such as metals and pieces from reservoir reclamation, metallography, and burst testing; reduce them in size; and then transfer them to a container that would be baked in a vacuum oven. The tritiated off-gas would be evacuated to a storage tank for accountability of tritium prior to transfer to the process stripper. Any pyrophoric materials would be oxidized to promote chemical stability. The solid components would then be cooled in the oven and managed in accordance with the solid LLW procedures outlined below. Solid LLW would consist of hydride beds, U-bed cartridges, Z-beds, stripper catalysts, retired process equipment, glovebox tools wastes, process metal residues, unreclaimed (empty) reservoirs, and low-specific activity waste. The LLW solids would be reduced in size, immobilized, stabilized, packaged, and then disposed of at one of the E-Area LLW disposal units. Hazardous Waste. Solid hazardous wastes would be generated from nonradioactive materials such as solvent rags, containers with residual hazardous materials, and lead-acid batteries. After compaction, if appropriate, the solid hazardous wastes would be packaged in DOT-approved containers and sent to an onsite/offsite RCRA-permitted treatment, storage, and disposal facility. DOT-certified transporters would be used for any offsite shipments. Mixed Low-Level Waste. Liquid mixed LLW would originate from potentially-contaminated used compressor and glovebox bubblers (sealpot) oil. Solid mixed LLW would include solvent rags and oil-contaminated materials with trace quantities of tritium. Mixed LLW would be stored in the RCRA-permitted facility onsite until the permanent disposal method could be determined. The overall management of mixed LLW would be in accordance with the Savannah River Site Treatment Plan, which is currently being developed to comply with the Federal Facility Compliance Act of 1992. Nonhazardous Waste. Sewage wastewater would be treated at the onsite Sewage Treatment Plant Number2 (607-H). Sewage wastewater would be kept separate from all industrial and process waste-waters and normally contains no radioactive wastes from the module. The sewage wastewater would be routinely monitored for radioactive contaminants. The sludge would be disposed of in a permitted landfill. As illustrated in figure A.2.2.2-3, the treated effluent would be discharged through a permitted NPDES outfall. Industrial wastewater would be treated and discharged through a permitted NPDES outfall. The sludge would be disposed of in a permitted sanitary landfill. Solid industrial, clean-shredded metal, after sanitization and demilitarization, and trash would be collected and sent to a permitted landfill. As shown in figure A.2.2.2-2 the blowdown from the cooling tower and process waste-water would be discharged after treatment directly into the river through an NPDES outfall. The upgrade design includes stormwater retention facilities with the necessary NPDES monitoring equipment. Rainfall within the facilities area would be collected separately and routed to the stormwater collection ponds and then sampled and analyzed before discharge into the river. If the runoff is contaminated, it would be treated in the process water treating system. Runoff from outside the facilities area would be discharged directly into the river. Figure (Page A-91) Figure A.2.2.2-1.-Tritium Recycling Facilities Upgrades at Savannah River Site (Generalized). Figure (Page A-92) Figure A.2.2.2-2.-Upgraded Tritium Recycling Facilities Processes. Figure (Page A-93) Figure A.2.2.2-3.-Upgraded Tritium Recycling Facilities Waste Management System. Table A.2.2.2-1.-Upgraded Tritium Recycling Facilities Construction Material/Resource Requirements Material/Resources Consumption - Unconsolidated Consolidated Upgrade Upgrade Electrical energy 2,000 2,000 (MWh) Concrete (yd3) 1,900 2,100 Steel (tons) 210 240 Fuel (gal) 16,000 17,000 Water (gal) 130,000 140,000 Source: SR DOE 1995a. Table A.2.2.2-2.-Upgraded Tritium Recycling Facilities Operation Utility Requirements Utility Consumption - Unconsolidated Consolidated Upgrade Upgrade Electrical energy 24,000 24,000 (MWh/yr) Electrical load (MWe) 3 3 Coal (tons) 5,200 5,200 Fuel, liquid (GPY) 60,000 56,000 Water (GPY) 51,000,000 51,000,000 Source: SR DOE 1995a. Table A.2.2.2-3.-Upgraded Tritium Recycling Facilities Annual Chemical Requirements Chemical Quantity (lb) - Unconsolidated Consolidated Upgrade Upgrade Solid Adhesives, 330 330 lubricants and paints Catalyst and 800 800 molecular sieves Depleted uranium 1,500 1,900 (water decomposer) Other solid 1,500 1,500 chemicals Liquid Water treatment 43,000 39,000 chemicals Petroleum oils 550 550 Organic solvents 3,400 3,400 Other liquid 3,400 3,400 chemicals Gaseous Nitrogen 2,700,000 2,800,000 Argon 170,000 170,000 Helium 450 450 Other gaseous 8,100 8,100 chemicals Table A.2.2.2-4.-Upgraded Tritium Recycling Facilities Estimated Waste Volumes - Unconsolidated Consolidated Category Annual Average Annual Volume Annual Volume Annual Average Annual Volume Annual Volume Volume Generated Generated From Effluent From Volume Generated Generated From Effluent From From Construction Operations Operations From Construction Operations Operations (yd3) (yd3) (yd3) (yd3) (yd3) (yd3) Low-Level Waste Liquid None None None None None None Solid None 350 117 None 350 117a Mixed Low-Level Waste Liquid None 0.03 0.03 None 0.03 0.03 (6 gal) (6 gal) (6 gal) (6 gal) Solid None 2 2 None 2 2 Hazardous Waste Liquid Included in solid None None Included in solid None None Solid < 0.3 1 1 < 0.3 1 1 Nonhazardous Waste (Sanitary) Liquid 149 158,000 158,000 182b 153,000 153,000 (30,000 gal) (32,000,000 gal) (32,000,000 gal) (36,700 gal) (31,000,000 gal) (31,000,000 gal) Solid 14 7,800 2,600 15 7,400 2,470e Nonhazardous Waste (Other) Liquid Included in sanitary Included in sanitary Included in sanitary Included in sanitary Included in sanitary Included in sanitary Solid Included in sanitary 6,800 None Included in sanitary 6,400f None -TSAR_DOE_SECTION- A.3 Tritium Supply Technology Options TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options Light water reactors have undergone continual design improvements because they are used by the commercial power industry and the ALWR designs presented in this PEIS incorporate these improvements. Potential technological refinements to the MHTGR and HWR technologies that differ from those evaluated in this PEIS are provided below. This section also discusses the feasibility of using the tritium supply technologies to burn plutonium. -TSAR_DOE_SECTION- A.3.1 Technology Innovations TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.1 Technology Innovations -TSAR_DOE_SECTION- A.3.1.1 Gas Turbine Modular Helium Reactor TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.1 Technology Innovations A.3.1.1 Gas Turbine Modular Helium Reactor The initial MHTGR design for DOE's New Production Reactor Program was a 350 MWt steam cycle plant. The 600 MWt Gas Turbine Modular Helium Reactor represents a different technology, in which the primary helium coolant drives a turbine generator through a gas-compression/gas-expansion, heating/cooling cycle without a phase change. This rep- resents a departure from the conventional steam cycle used in the 350 MWt design, in which steam, produced in the steam generator of the secondary heat exchanger, is used to drive the main turbine generator. In this respect, the Gas Turbine Modular Helium Reactor can be compared to a boiling water reactor in that they are both direct-cycle plant designs. The exception is that the Gas Turbine Modular Helium Reactor uses helium in a single phase for cooling and for driving the turbine. This gas turbine design achieves higher efficiency and reduces the design complexity by eliminating the feedwater and steam systems in the smaller MHTGR plant. The operating conditions of the 600 MWt design plant are elevated above those of the 350 MWt design. The higher operating temperatures, espe- cially the hot helium (core outlet) temperature, provide a further increase in net plant efficiency over the 350 MWt design. Where three 350 MWt modules are considered necessary to meet the current tritium production baseline goal in this PEIS, only two 600MWt modules would be needed. Single Medium System. Hot helium exits the reactor core, flows through the center hot duct within the cross vessel, and expands through the turbine. The turbine directly drives the electrical generator, the low pressure compressor, and the high pressure compressor. Helium exits the turbine and flows through the high efficiency plate-fin recuperator to return as much energy as possible to the cycle, then finally flows through the precooler to reject heat to the ultimate heat sink. Cold helium enters the inter-cooler compressor and passes through the recuperator. The helium then flows from the recuperator exit, through the outer annulus within the cross vessel, past the reactor vessel walls for vessel cooling, and finally down through the core to complete the loop. Reactor Module. The reactor module arrangement of the 600 MWt design is similar to that of the 350MWt design. The steam generator vessel of the 350 MWt design is replaced with a slightly larger diameter power conversion vessel in the 600 MWt design. The reactor and power conversion vessels are vertically positioned and connected by a coaxial cross vessel. The fuel particle and fuel element designs are identical, as are the lithium target particles, compacts, and assemblies. Both reactor module designs refuel on the same schedule. Each design has a nominal 3-year residence time for the fuel, and one-third of the fuel blocks would be replaced annually. The target assemblies in all the fuel and reflector blocks would be replaced semiannually for the tritium supply mode. Power Conversion Module. The power conversion vessel would contain the turbomachine, generator, recuperator, precooler, and intercooler. The basic function of this power conversion module is to convert heat energy to electrical energy and to provide the motive force for the helium primary coolant. A significant feature of the 600 MWt design turbomachine assembly is that the turbine, the two compressors, and the submerged generator would all be mounted on a single, vertical shaft. This would enable using a vertical power conversion vessel, be the most space efficient, and allow the removal of the turbomachine by the reactor building crane. Using a generator that is submerged within the helium primary coolant would eliminate the need for rotating shaft penetrations of the primary pressure boundary. Also, the turbomachine assembly would incorporate an active magnetic thrust bearing and three to five magnetic journal bearings. With this noncontact type of bearing, no bearing lubricants or coolants would be required, and bearing power losses would be minimal. Recuperator. The recuperator is a helium-to-helium heat exchanger included in the 600 MWt design. This heat exchanger would recover energy from the helium exiting the turbine and would utilize that energy to preheat the compressed helium before it enters the reactor. This recovery process would assist in achieving a net plant efficiency of approximately 48 percent compared to approximately 38 percent for the 350 MWt steam cycle design. Precooler and Intercooler. The precooler and inter-cooler heat exchangers would be included in the power conversion loop of the 600 MWt design to reduce the temperature of the helium primary coolant prior to entering the inlets to the low pressure and high pressure compressors. This cooling process would reduce the work required to compress the helium in the compressors, thereby enhancing the net plant cycle efficiency. Environmental Impacts. There are no substantial overall differences between the two-module 600MWt Modular Helium Reactor and the three-module 350 MWt MHTGR. Compared to the MHTGR, the Modular Helium Reactor would have slightly more environmental impact in some resource areas, and slightly less environmental impact in other resource areas. A two-module 600 MWt Modular Helium Reactor would generate approximately 3percent more spent nuclear fuel, increase worker doses, increase normal operational releases of some radioactive constituents, and increase the accident source term. Potential decreases in environmental impacts could result from less construction, decreased tritium releases during normal operations, reduced operating personnel, and less cooling water requirements. -TSAR_DOE_SECTION- A.3.1.2 Small Advanced Heavy Water Reactor TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.1 Technology Innovations A.3.1.2 Small Advanced Heavy Water Reactor A reduced capacity HWR design does not now exist but could be developed to produce tritium at a level to meet the steady-state (3/16 goal) tritium requirement, but would have the capability, if necessary, to be modified to meet baseline (3/8 goal) tritium production requirements. The HWR evaluated in all other sections of this PEIS represents a downsized 8/8 goal quantity design from the original New Production Reactor Program operated at the steady-state level. The small advanced HWR would be designed for a nominal thermal output of approximately 470 MWt with a deep burn fuel cycle using uranium fuel. This would enable the small advanced HWR to produce steady-state (3/16 goal) tritium requirements. It would have the ability, with pre-planned plant upgrades and operational changes, to operate at a higher power level and produce the baseline 3/8 tritium goal quantity (DOE 1995a:1). Specific design and operational features would be incorporated into the basic design to facilitate the modifications necessary to increase the tritium production to the 3/8 goal in a timely manner. These include the addition of fuel/targets into vacancy posi- tions; operating at a higher capacity factor; reducing the fuel cycle duration; operating the reactor at a higher power density; and oversizing and making provisions for additional heat rejection system components (e.g., adding cooling tower cells). Key Design Improvements. The layout and design of the small advanced HWR would be similar to the 3/8 concept with the primary difference being a more compact, and modularly designed facility. The small advanced HWR facility would occupy about 40 to 50acres (within the limited access fence) as compared to 60 acres for the 3/8 design. The total area planned to be disturbed for the purpose of constructing and operating the small advanced HWR would be approximately 150 to 170 acres. Because of the smaller plant configuration, the construction schedule will be shortened from the current design. Construction of the small advanced HWR complex would require approximately 5 years and 1,800 workers during the peak construction period. Additional design improvements for the small advanced HWR would be a reduced number of systems consistent with passive reactor safety requirements; smaller heavy water inventory and lower primary coolant flow rates translating into smaller equipment and piping sizes; reduced shielding requirements based on reduced source terms and use of robotics; and smaller building sizes and concomitant reductions in bulk material quantities. The small advanced HWR fuel/targets, core configuration, and primary reactor coolant system would be similar to, but notably smaller than, the HWR described in sections 3.4.2.1 and A.2.1.1. Outlying buildings, cooling towers, switchyard, and overall plant footprints dimensions would also be reduced for the small advanced HWR. Some differences of key parameters between the small advanced HWR and the HWR design evaluated in this PEIS would be 470 MWt versus 990 MWt, 144 fuel/target assemblies versus 258, and a containment building which is less than 120 feet in diameter versus 140 feet. The design would still maintain the two reactor coolant loops with some reduction in the piping, heat exchanger, pressurizer and accumulator sizes. Also, the fuel cycle duration would be retained at 350 days with an annual capacity factor of 85percent. It should be emphasized that the description of the small advanced HWR represents a pre-preconceptual design based on knowledge of existing HWR designs, recent requirements for advanced light water reactors, and the varied production requirements defined in section 3.1.1. -TSAR_DOE_SECTION- A.3.2 Plutonium Disposition TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.2 Plutonium Disposition This PEIS analyzes technology options that have been determined to be reasonable for the mission of producing tritium required for continuing stockpile support only. The ALWR and the MHTGR technologies offer the added benefit of being capable of producing steam for electricity production that could prove to be desirable in offsetting operational and capital costs. The HWR cannot produce steam efficiently and the APT cannot produce steam at all. The ALWR and a Gas-Cooled Modular Helium Reactor are under consideration to fulfill roles in the ultimate disposition of fissile materials declared surplus to national defense needs, particularly plutonium. The current and planned dismantling of large quantities of the nuclear weapons in the Nation's stockpile has and will result in the generation of excess quantities of plutonium. Various alternatives have been considered that would place excess plutonium into relatively inaccessible forms conforming with the "spent fuel standard", in which the remaining plutonium is made as inaccessible for retrieval and weapons use as the residual plutonium in spent nuclear fuel from commercial power reactors. Among these is the use of plutonium-based reactor fuels and the immobilization of diluted plutonium through vitrification in ceramic or metal immobilization techniques. Other options use accelerator or reactor-generated neutrons to virtually destroy plutonium through deep burn fissioning. The reactor technologies considered for tritium production could be used to fulfill a part of or all of the requirements for plutonium disposition. However, other alternatives are also under consideration for plutonium disposition, such as direct disposal to deep boreholes and use of reactors in other countries, that do not use tritium production technologies. Clearly, plutonium disposition alternatives should not be limited to those technologies that support tritium production uniquely. However, if a reactor is to be selected as the tritium supply technology, it will be evaluated by the plutonium disposition program to determine if partial requirements can be met. The Office of Fissile Materials Disposition published an Implementation Plan in November 1994, which describes the alternatives that will continue to be evaluated for plutonium disposition. The National Academy of Science's report Management and Disposition of Excess Weapons Plutonium (NAS 1994a) did not consider a linkage between tritium production and plutonium disposition technologies to necessarily be desirable. The wider range of options available for plutonium disposition and the smaller capacity required for tritium production are cited as countervailing to linkage. The production of tritium in the same facilities that might also be used to dispose of weapons materials, which would fall under International Atomic Energy Agency safeguards, is also cited as a potential concern. The National Academy of Science committee also concluded that the cost savings from a dual purpose, tritium production/plutonium disposition technology strategy would not be great and might not be justified when balanced against the complications discussed above. Excess plutonium materials are considered appropriate for use in reactor fuel and disposal as reactor spent fuel for three basic reasons: It would meet the "spent fuel standard"; There is a path forward for material that meets the "spent fuel standard", i.e., ultimate disposition in a repository; and It may result in equivalent reciprocal actions from the Russians. Each of the tritium supply technology alternatives is discussed below for plutonium disposition. -TSAR_DOE_SECTION- A.3.2.1 Advanced Light Water Reactor Technology TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.2 Plutonium Disposition A.3.2.1 Advanced Light Water Reactor Technology This PEIS evaluates the ALWR technology alternative for both large and small reactor versions for tritium production. The reactor models under evaluation for plutonium disposition include light water reactors. The four primary design concepts are: ABB-Combustion Engineering Pressurized Light Water Reactor, System 80+. General Electric Advanced Boiling Water Reactor (GE ABWR). Westinghouse Pressurized Light Water Reactor, PDR-1400. Westinghouse Pressurized Light Water Reactor, AP-600. Without major modifications, typical light water reactors could burn a fuel consisting of mixed-oxides of plutonium-oxide and uranium-oxide in their reactor cores. Some light water reactors are designed to use mixed-oxides in 100 percent of their reactor cores. A number of operating or partially completed reactors could be modified to use full mixed-oxide cores, or a new full mixed-oxide reactor could be built with costs partly offset by later sales of electricity. Although the United States has no operating mixed-oxide fuel fabrication plant or mixed-oxide fuel-burning reactors today, this technology is techni- cally demonstrated by the burning of mixed-oxide fuels in several light water reactors in other countries, and has been demonstrated by use in several U.S. reactors in the 1970s. Burning mixed-oxide fuel in an ALWR would result in creating inaccessible forms of plutonium conforming with the "spent fuel standard." The ALWR designs could be adapted for dual tritium production and plutonium burning on essentially a noninterference basis. Some thermal power degradation could occur as a result of producing tritium, but not a significant amount. Also, the cyclic tritium production campaigns and the effect that this might have on the fuel burn cycle might not maximize the annihilation of plutonium fuel constituents. Only one reactor of the models shown would be required for tritium production. If only one Large ALWR is used as the principal technology to dispose of excess plutonium, approximately 50 metric tons of plutonium could be disposed of over the 40-year reactor life. For the AP-600, approximately one-half this amount could be disposed of over the 40-year reactor life. ABB-Combustion Engineering System 80+. The Combustion Engineering System 80+ is a standard pressurized water reactor design which incorporates evolutionary improvements to the System 80 design in operation at the Palo Verde Nuclear Generating Station. The Combustion Engineering System 80+ improvements include a ring-forged reactor vessel, greater design margins for major components, and improvements to safety systems. NRC issued final design approval on ABB-Combustion Engineering's 1,300 MWe System 80+ pressurized water reactor in July 1994. The Combustion Engineering System 80+ design would preserve the System 80 design feature of accommodating mixed-oxide fuel loadings up to and including a full mixed-oxide reactor core loading. The higher decay heat loads after shutdown generated by a full mixed-oxide reactor core would be accommodated in the design of the various heat removal systems, the shutdown cooling system, spent nuclear fuel pool cooling system, and the component cooling water system. The Combustion Engineering System 80+ design would maintain flexibility in its power output depending on the mode of power operation. As a uranium-oxide fueled power production plant, the core thermal rating would be 3,914 MWt. When loaded with mixed-oxides, the output would be 3,817MWt. And, when operating as a tritium producer while loaded with uranium-oxide, the output would be 3,410 MWt. Westinghouse Pressurized Water Reactors. The Westinghouse Electric Corporation's pressurized water reactor AP600 is a 600 MWe nuclear power plant that has a simpler overall design than other Westinghouse plants. The AP600 design includes enhanced safety margins in such areas as fuel rod thermal limits and corrosion protection features. The AP600 would use a natural circulation heat exchange loop connected to the reactor and located inside the containment for reactor residual heat removal. Application for approval is under review by the NRC for Westinghouse Electric Corporation's AP600 design. This advanced pressurized water reactor system is projected to have the capability of producing tritium with either plutonium-oxide or uranium-oxide as fuel. To operate in this mode, the core assembly structure would need to be altered such that some light water reactor tritium-producing rods would replace some fuel rods. The impact this would have on thermal-hydraulic compatibility and passive safety remains to be evaluated. The PDR-1400 is an evolutionary reactor similar to the four loop Westinghouse designs of existing pressurized water reactors. General Electric Boiling Water Reactor. General Electric has applied recent technology innovations to the development of an advanced design of its currently operating boiling water reactors. The 1,300MWe Advanced Boiling Water Reactor focuses on overall plant simplification and the use of safety systems that do not require operator actions. The major innovations that have been made in the systems would rely on gravity or stored energy to ensure core cooling and decay heat removal. The NRC issued final design approval for General Electric's 1,300 MWe Advanced Boiling Water Reactor in July 1994. The Advanced Boiling Water Reactor was originally designed to utilize full core loads of mixed-oxide fuel. The physical design of the core and the boiling water reactor core dynamics are such that no modifications to the reactor system are required for the use of mixed-oxide fuel over a wide range of plutonium loadings. If necessary, the Advanced Boiling Water Reactor has the flexibility to easily switch back to conventional uranium fuel, again without plant physical changes. -TSAR_DOE_SECTION- A.3.2.2 Modular High Temperature Gas-Cooled Reactor Technology TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.2 Plutonium Disposition A.3.2.2 Modular High Temperature Gas-Cooled Reactor Technology The MHTGR design evaluated in this PEIS for a tritium mission is based upon a 350 MWt steam cycle plant. In order to increase thermal efficiency, a direct cycle Gas Turbine Modular Helium Reactor pre-preconceptual design that could produce tritium as well as be used for plutonium disposition has been suggested. The module power rating for this reactor would be increased to 600 MWt. Where three 350MWt MHTGR modules would be required to meet baseline tritium production requirements, only two 600 MWt Modular Helium Reactor modules would be required. The new Gas Turbine Modular Helium Reactor design is under evaluation for plutonium disposition purposes and is represented by the General Atomics 600 MWt design described in Section A.3.1.1. This reactor is designed to actually destroy plutonium through deep burn fissioning which results in a net plutonium destruction rate of approximately 65 percent. The technical, research and development, and demonstration requirements necessary to be completed for the Modular Helium Reactor substantially increase the technical, schedule, and cost risks of bringing this concept to maturity. The Modular Helium Reactor could not be adapted for both tritium production and plutonium burning in a noninterference manner. The tritium production efficiency decreases by a factor of 4 when using plutonium fuel in a Modular Helium Reactor. Thus, utilizing the Modular Helium Reactor for tritium production of 3/8 goal quantities with plutonium fuel would require eight reactor modules as opposed to two with uranium fuel. This factor of 4 increase in the number of required 600 MWt Modular Helium Reactors is directly attributable to the fact that tritium targets can only be placed in the reflector material of the plutonium-fueled core and not within the fueled region of the core as they can with a uranium-fueled reactor. This limitation of the plutonium-consumption Modular Helium Reactor serves to reduce by a factor of 4 the number of targets in each reactor available to produce the required amount of tritium, and thus increases the number of required plutonium-fueled reactor modules by the same factor. Correspondingly, assuming that it is possible to extrapolate this data from the 600 MWt direct cycle gas turbine reactor technology to the 350 MWt MHTGR technology proposed for tritium production, the three 350 MWt reactors proposed to produce the 3/8 goal quantity of tritium utilizing uranium fuel would need to be increased by a factor of 4 if plutonium fuel is utilized. This would result in a requirement of 12 350 MWt MHTGR modules to produce the 3/8 goal quantity of tritium. However, only six 350 MWt MHTGRs would be required to produce the steady-state (3/16 goal) requirement. In the event that surge tritium production is required, some of the six 350 MWt reactors could be refueled with uranium fuel in order to allow for increased placement of target rods in the core, and thus meet the surge requirement. It is not expected that the six 350 MWt reactors could dispose of the same amount of plutonium as one large ALWR over the 40-year reactor life. The inefficiencies induced by using a dual purpose strategy are not entirely known; however, it could be expected that tritium production cycle interrupts would cause some minor delays in the plutonium disposition timetable. -TSAR_DOE_SECTION- A.3.2.3 Heavy Water Reactor Technology TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.2 Plutonium Disposition A.3.2.3 Heavy Water Reactor Technology An advanced HWR design for tritium production could possibly use plutonium fuel. The plutonium burn-up for such a reactor is estimated to be 35percent. If a single HWR at the power level required to produce the annual goal quantity of tritium is used for plutonium disposition then approximately 70 years would be required to dispose of 50tons of plutonium. Since this time-frame is greater than the expected life of such a reactor, two reactors would be required. If selected for tritium supply, the HWR would then be evaluated for a plutonium disposition mission. -TSAR_DOE_SECTION- A.3.2.4 Accelerator Production of Tritium Technology TABLE OF CONTENTS Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling Volume II APPENDIX A: NUCLEAR FACILITIES A.3 Tritium Supply Technology Options A.3.2 Plutonium Disposition A.3.2.4 Accelerator Production of Tritium Technology Accelerator-based conversion technologies have been investigated as possible plutonium disposition alternatives. Accelerator-based conversion would use a sub-critical reactor augmented by an accelerator to "deep burn" plutonium, which could eliminate in excess of 90 percent of the processed plutonium over the campaign. This technique would allow plutonium disposition to progress well beyond the "spent fuel standard." The technique involves the use of particle accelerator module, at the appropriate power level, that is used as the precursor for the APT alternative evaluated in this PEIS. The target systems are, however, quite different. For plutonium disposition, either molten salt or particle bed systems could be used. In both cases, materials containing surplus plutonium particles would be irradiated with neutrons until a large portion of the plutonium was eliminated through fissioning. Excess commercial power could not be generated with this scheme in order to exploit the energy value of the plutonium; however, enough power could be generated to operate the particle accelerator once started. In theory, an accelerator-based system could provide plutonium disposition services as well as produce tritium; however, one process would compete with the other for neutrons. For plutonium "deep burn" disposition, it is estimated that up to three accelera- tor-based facilities, each having four target blankets arrays, would be required for a 20year disposition campaign. Only one such facility with two lithium-6 target-blanket systems (or continuous helium-3 target systems) would be required to produce anticipated annual quantities of tritium. This option is only at the early paper study stage and would not be available on a large scale for decades. If the estimated performance could be attained, however, such systems could destroy plutonium at a rate (per unit of thermal energy) comparable to those of the other destruction-oriented options (e.g.,reactors) and could reach high reduction factors for plutonium inventory more rapidly than many of the other options.
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