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FFTF Plant transition function analysis report

Description: The document contains the functions, function definitions, function interfaces, function interface definitions, Input Computer Automated Manufacturing Definition (IDEFO) diagrams, and function hierarchy charts that describe what needs to be performed to deactivate FFTF.
Date: September 1, 1995
Creator: Lund, D.P. & Group, FFTF Working
Partner: UNT Libraries Government Documents Department

FFTF Plant transition mission analysis report

Description: FFTF (Fast Flux Test Facility) is a 400-MW(t) sodium-cooled, fast flux test reactor at Hanford, designed to test fuels and materials for advanced nuclear power plants; it has no capability for generating electric power. Since a long-term mission could not be found for FFTF, it was placed in standby, and a recommendation was made that it be shut down. Purpose of the FFTF Transition Project is to prepare it for Decontamination and Decommissioning; this will be accomplished by establishing a passively safe and environmentally secure configuration, that can be preserved for several decades. This report presents the results of the mission analysis, which is required by Hanford systems engineering procedures.
Date: May 25, 1995
Creator: Lund, D.P.
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning of the JANUS reactor at the Argonne National Laboratory-East site

Description: Argonne National Laboratory has begun the decontamination and decommissioning (D&D) of the JANUS Reactor Facility. The project is managed by the Technology Development Division`s D&D Program personnel. D&D procedures are performed by sub-contractor personnel. Specific activities involving the removal, size reduction, and packaging of radioactive components and facilities are discussed.
Date: May 1, 1997
Creator: Fellhauer, C.R. & Garlock, G.A.
Partner: UNT Libraries Government Documents Department

Innovative technology summary report: High-speed clamshell pipe cutter

Description: The Hanford Site C Reactor Technology Demonstration Group demonstrated the High-Speed Clamshell Pipe Cutter technology, developed and marketed by Tri Tool Inc. (Rancho Cordova, California). The models demonstrated are portable, split-frame pipe lathes that require minimal radial and axial clearances for severing and/or beveling in-line pipe with ranges of 25 cm to 41 cm and 46 cm to 61 cm nominal diameter. The radial clearance requirement from the walls, floors, or adjacent pipes is 18 cm. The lathes were supplied with carbide insert conversion kits for the cutting bits for the high-speed technique that was demonstrated. Given site-specific factors, this demonstration showed the cost of the improved technology to be approximately 30% higher than the traditional (baseline) technology (oxyacetylene torch) cost of $14,400 for 10 cuts of contaminated 41-cm and 61-cm-diameter pipe at C Reactor. Actual cutting times were faster than the baseline technology; however, moving/staging the equipment took longer. Unlike the baseline torch, clamshell lathes do not involve applied heat, flames, or smoke and can be operated remotely, thereby helping personal exposures to be as low as reasonably achievable. The baseline technology was demonstrated at the C Reactor north and south water pipe tunnels August 19--22, 1997. The improved technology was demonstrated in the gas pipe tunnel December 15--19.
Date: September 1, 1998
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning activities photobriefing book FY 2000.

Description: A major milestone was reached in Fiscal Year (FY) 2000--the completion of the Chicago Pile-5 (CP-5) D and D Project. CP-5, the first reactor built on the Argonne National Laboratory-East (ANL-E) site, was a 5-megawatt, heavy water-moderated, enriched uranium-fueled reactor. It was the principal reactor on the Argonne site used to produce neutrons for scientific research from 1954 to 1979. The reactor was shut down and defueled in 1979, and placed in a lay-up condition. In 1990, funding was provided to begin the decontamination and decommissioning (D and D) of this facility, and work began in June 1991. D and D tasks were performed by both ANL-E personnel and subcontractor personnel from Duke Engineering and Services, Marlborough MA, under the management of ANL-E D and D Program personnel. In July 2000, the final project report was presented to the Department of Energy, and the facility was formally decommissioned and transferred to the landlord. Total project duration was 97 months, and total project cost was $29.5M. Also, in FY 2000, work began on the 60 Cyclotron D and D Project. An accelerator used for basic research, this facility produced beams of deuterons, helium ions, singly charged hydrogen molecules, and neutrons of a broad energy spectrum. This machine was built in 1952; operations ended in 1992. Late in FY 1999, MOTA Corporation, Columbia SC, was selected as subcontractor. MOTA used a custom-built 76-inch reciprocating saw manufactured by Reverse Engineering, Columbia SC, to cut the cyclotron's 220-ton steel yoke into blocks that were approximately 17,000 pounds each. This project is scheduled to be completed in March 2001. In addition to a photographic chronology of FY 2000 activities at the CP-5 D and D Project and the 60 Cyclotron D and D Project, brief descriptions of other FY 2000 activities and of projects planned ...
Date: February 5, 2001
Creator: Fellhauer, C. F.
Partner: UNT Libraries Government Documents Department

Evaluation of the Molten Salt Reactor Experiment drain tanks for reuse in salt disposal, Oak Ridge National Laboratory, Oak Ridge, Tennessee

Description: This report was prepared to identify the source documentation used to evaluate the drain tanks in the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). The evaluation considered the original quality of the tanks, their service history, and their intended use during the removal of fluoride salts. It also includes recommendations for a quality verification plan. The estimates of corrosion damage to the salt containing system at the MSRE are low enough to lend optimism that the system will be fit for its intended use, which is disposal of the salt by transferring it to transport containers. The expected corrosion to date is estimated between 10 and 50 mil, or 2 to 10% of the shell wall. The expected corrosion rate when the tanks are used to remove the salt at 110 F is estimated to be .025 to 0.1 mil per hour of exposure to HF and molten salt. To provide additional assurance that the estimates of corrosion damage are accurate, cost effective nondestructive examination (NDE) has been recommended. The NDE procedures are compared with industry standards and give a perspective for the extent of additional measures taken in the recommendation. A methodology for establishing the remaining life has been recommended, and work is progressing towards providing an engineering evaluation based upon thickness and design conditions for the future use of the tanks. These extra measures and the code based analysis will serve to define the risk of salt or radioactive gases leaking during processing and transfer of the salt as acceptable.
Date: May 1, 1998
Partner: UNT Libraries Government Documents Department

Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

Description: Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.
Date: April 1, 1997
Creator: Owen, M.B. Smith, R. & McNeil, J.
Partner: UNT Libraries Government Documents Department

Deactivation of the EBR-II complex

Description: In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D&D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D&D plan has necessitated this current action. The EBR-II is a pool-type reactor. The primary system contains approximately 87,000 gallons of sodium, while the secondary system has 13,000 gallons. In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility has been built to react the sodium to a dry carbonate powder in a two stage process. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in the primary and secondary systems must be either reacted or inerted to preclude future concerns with sodium-air reactions that generate explosive mixtures of hydrogen and leave corrosive compounds. Residual amounts of sodium on components will effectively {open_quotes}solder{close_quotes} components in place, making future operation or removal unfeasible.
Date: December 31, 1997
Creator: Michelbacher, J.A.; Earle, O.K. & Henslee, S.P.
Partner: UNT Libraries Government Documents Department

Quality assurance plan for the Molten Salt Reactor Experiment Remediation Project at the Oak Ridge National Laboratory. Phase 1 -- Interim corrective measures and Phase 2 -- Purge and trap reactive gases

Description: This Quality Assurance Plan (QAP) identifies and describes the systems utilized by the Molten Salt Reactor Experiment Remediation Project (MSRERP) personnel to implement the requirements and associated applicable guidance contained in the Quality Program Description Y/QD-15 Rev. 2 (Energy Systems 1995f). This QAP defines the quality assurance (QA) requirements applicable to all activities and operations in and directly pertinent to the MSRERP Phase 1--Interim Corrective Measures and Phase 2--Purge and Trap objectives. This QAP will be reviewed, revised, and approved as necessary for Phase 3 and Phase 4 activities. This QAP identifies and describes the QA activities and procedures implemented by the various Oak Ridge National Laboratory support organizations and personnel to provide confidence that these activities meet the requirements of this project. Specific support organization (Division) quality requirements, including the degree of implementation of each, are contained in the appendixes of this plan.
Date: November 1, 1995
Partner: UNT Libraries Government Documents Department

Reactor D and D at Argonne National Laboratory - lessons learned.

Description: This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals.
Date: March 23, 1998
Creator: Fellhauer, C. R.
Partner: UNT Libraries Government Documents Department

Stationary low power reactor No. 1 (SL-1) accident site decontamination & dismantlement project

Description: The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft{sup 2} two story, cinder block administrative building; two 4,000 ft{sup 2} single story, steel frame office buildings; a 850 ft{sup 2} steel framed, metal sided PL condenser building, and a 550 ft{sup 2} steel framed decontamination and laydown building.
Date: November 1, 1995
Creator: Perry, E.F.
Partner: UNT Libraries Government Documents Department

The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

Description: If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of ...
Date: May 27, 1994
Partner: UNT Libraries Government Documents Department

ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

Description: This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted.
Date: February 27, 2003
Creator: Morton, M.R.; Nielson, R.R. & Trevino, R.A.
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning of the EBR-I Complex. Final report

Description: This final report covers the Decontamination and Decommissioning (D and D) of the Experimental Breeder Reactor No. 1 (EBR-I) Complex funded under Contract No. AT(10-1)-1375. The major effort consisted of removal and processing of 5500 gallons of sodium/potassium (NaK) coolant from the EBR-I reactor system. Tests were performed to assess the explosive hazards of NaK and KO$sub 2$ in various environments and in contact with various contaminants likely to be encountered in the removal and processing operations. A NaK process plant was designed and constructed and the operation was successfully completed. Lesser effort was required for D and D of the Zero Power Reactor (ZPR-III) Facility, the Argonne Fast Source Reactor (AFSR) Shielding, and removal of contaminated NaK from the storage pit. The D and D effort was completed by 13 June 1975, ahead of schedule. (auth)
Date: July 1, 1975
Creator: Kendall, E.W. & Wang, D.K.
Partner: UNT Libraries Government Documents Department

Comments received on proposed rule on radiological criteria for decommissioning and related documents

Description: The Nuclear Regulatory Commission (NRC) is conducting an enhanced participatory rulemaking to establish radiological criteria for the decommissioning of NRC-licensed facilities. As a part of this action, the Commission published in the Federal Register (59 FR 43200), on August 22, 1994, a proposed rule on radiological criteria for decommissioning, soliciting comments both on the rule as proposed and on certain specific items as identified in its supplementary statement of considerations. A draft Generic Environmental Impact Statement (GEIS) in support of the rule, also published in August 1994 as NUREG-1496, along with its Appendix A (NUREG-1501), were also made available for comment. A staff working draft on regulatory guidance (NUREG-1500)was also made available. This report summarizes the 1,309 comments on the proposed rule and supplementary items and the 311 comments on the GEIS as excerpted from 101 docketed letters received associated in the Federal/Register notice. Comments from two NRC/Agreement-States meetings are also summarized.
Date: March 1, 1996
Creator: Page, G.; Caplin, J. & Smith, D.
Partner: UNT Libraries Government Documents Department

Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - appendices. Final report

Description: The NRC staff is in need of decommissioning bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2) located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.
Date: July 1, 1996
Creator: Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. & McDuffie, P.N.
Partner: UNT Libraries Government Documents Department

A survey of commercially available manipulators, end-effectors, and delivery systems for reactor decommissioning activities

Description: Numerous nuclear facilities owned by the U.S. Department of Energy (DOE) are under consideration for decommissioning. Currently, there are no standardized, automated, remote systems designed to dismantle and thereby reduce the size of activated reactor components and vessels so that they can be packaged and shipped to disposal sites. Existing dismantling systems usually consist of customized, facility-specific tooling that has been developed to dismantle a specific reactor system. Such systems have a number of drawbacks. Generally, current systems cannot be disassembled, moved, and reused. Developing and deploying the tooling for current systems is expensive and time-consuming. In addition, the amount of manual work is significant because long-handled tools must be used; as a result, personnel are exposed to excessive radiation. A standardized, automated, remote system is therefore needed to deliver the tooling necessary to dismantle nuclear facilities at different locations. Because this system would be reusable, it would produce less waste. The system would also save money because of its universal design, and it would be more reliable than current systems.
Date: May 1, 1996
Creator: Henley, D.R. & Litka, T.J.
Partner: UNT Libraries Government Documents Department

Planning and supervision of reactor defueling using discrete event techniques

Description: New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on the progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.
Date: December 31, 1995
Creator: Garcia, H.E.; Imel, G.R. & Houshyar, A.
Partner: UNT Libraries Government Documents Department

Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

Description: The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.
Date: July 1, 1996
Creator: Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. & McDuffie, P.N.
Partner: UNT Libraries Government Documents Department

Reactor surface contamination stabilization. Innovative technology summary report

Description: Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m{sup 2} or 2116 ft{sup 2}) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage.
Date: November 1, 1998
Partner: UNT Libraries Government Documents Department

US graphite reactor D&D experience

Description: This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).
Date: February 1, 1997
Creator: Garrett, S.M.K. & Williams, N.C.
Partner: UNT Libraries Government Documents Department

Completion of decommissioning: Monitoring for site release and license termination

Description: To request termination of a license upon completion of dismantling or decommissioning activities, documenting any residual radioactivity to show that the levels are environmentally acceptable will be necessary. When the regulators approve the decommissioning plan, they establish what the release criteria for the decommissioned site will be at the time of the site release and license termination. The criteria are numeric guidelines for direct radiation in soils and on surfaces. If the regulatory body finds that the measured on-site values are below the guidelines, the site will be acceptable for unrestricted release (no radiological controls or future use). If areas are found above those values, more decontamination or cleanup of these areas may be required unless the regulatory body grants an exemption.
Date: August 1, 1997
Creator: Boing, L.E.
Partner: UNT Libraries Government Documents Department

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

Description: This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.
Date: April 1, 1997
Creator: Owen, M.B.
Partner: UNT Libraries Government Documents Department

Passivation of fluorinated activated charcoal

Description: The Molten Salt Reactor Experiment (MSRE), at the Oak Ridge National Laboratory has been shut down since 1969 when the fuel salt was drained from the core into two Hastelloy N tanks at the reactor site. In 1995, a multiyear project was launched to remediate the potentially hazardous conditions generated by the movement of fissile material and reactive gases from the storage tanks into the piping system and an auxiliary charcoal bed (ACB). The top 12 in. of the ACB is known by gamma scan and thermal analysis to contain about 2.6 kg U-233. According to the laboratory tests, a few feet of fluorinated charcoal are believed to extend beyond the uranium front. The remainder of the ACB should consist of unreacted charcoal. Fluorinated charcoal, when subjected to rapid heating, can decompose generating gaseous products. Under confined conditions, the sudden exothermic decomposition can produce high temperatures and pressures of near-explosive characteristics. Since it will be necessary to drill and tap the ACB to allow installation of piping and instrumentation for remediation and recovery activities, it is necessary to chemically convert the reactive fluorinated charcoal into a more stable material. Ammonia can be administered to the ACB as a volatile denaturing agent that results in the conversion of the C{sub x}F to carbon and ammonium fluoride, NH{sub 4}F. The charcoal laden with NH{sub 4}F can then be heated without risking any sudden decomposition. The only consequence of heating the treated material will be the volatilization of NH{sub 4}F as a mixture of NH{sub 3} and HF, which would primarily recombine as NH{sub 4}F on surfaces below 200 C. The planned scheme for the ACB denaturing is to flow diluted ammonia gas in steps of increasing NH{sub 3} concentration, 2% to 50%, followed by the injection of pure ammonia. This report summarizes ...
Date: October 1, 1997
Creator: Del Cul, G. D.; Trowbridge, L. D.; Simmons, D. W.; Williams, D. F. & Toth, L. M.
Partner: UNT Libraries Government Documents Department