Corrective Action Management Units (CAMUs) were established by the Environmental Protection Agency (EPA) to streamline the remediation of hazardous waste sites. Streamlining involved providing cost saving measures for the treatment, storage, and safe containment of the wastes. To expedite cleanup and remove disincentives, EPA designed 40 CFR 264 Subpart S to be flexible. At the heart of this flexibility are the provisions for CAMUs and Temporary Units (TUs). CAMUs and TUs were created to remove cleanup disincentives resulting from other Resource Conservation Recovery Act (RCRA) hazardous waste provisions--specifically, RCRA land disposal restrictions (LDRs) and minimum technology requirements (MTRs). Although LDR and MTR provisions were not intended for remediation activities, LDRs and MTRs apply to corrective actions because hazardous wastes are generated. However, management of RCRA hazardous remediation wastes in a CAMU or TU is not subject to these stringent requirements. The CAMU at Sandia National Laboratories in Albuquerque, New Mexico (SNL/NM) was proposed through an interactive process involving the regulators (EPA and the New Mexico Environment Department), DOE, SNL/NM, and stakeholders. The CAMU at SNL/NM has been accepting waste from the nearby Chemical Waste Landfill remediation since January of 1999. During this time, a number of unique techniques have been …

For assessments of hard-to-detect radionuclides (HD-RN) contents in various type of radwastes at the NPP-A1, available empirical data referenced to 137Cs (actinides, 90Sr, 99Tc, 63Ni, 14C) and the theoretical assessment for the remaining HD-RN using calculated RN inventory and a simple model with effective relative (137Cs) spent fuel release fractions was applied. The analytical data of extended radiochemical analysis for the existing available operational radwaste forms have been reviewed for this purpose. 137Cs, 90Sr and 241Am were set up as release markers for partial spent fuel release groups of HD-RNs within which the total fractions of HD-RN released to the operational radwastes were assumed to be constant. It was shown by the assessment carried out that 137Cs and HD-RNs 129I, 99Tc, and partly 79Se and 14C are the main contributors to the disposal dose limit for the radioactive concentrate at NPP A-1. In the case of the radioactive sludge from the operational radwaste system the role of predominant dose contributors belongs to actinides 239,240Pu and 241Am. In the case of clearance of radioactive material from the NPP-A1 site, only the reference radionuclide, 137Cs was predicted to be the most dominant dose contributor. In all of these cases the estimated contributions …

A series of stepwise procedures to prepare a new organic-inorganic composite magnetic resin with phenolsulphonicformaldehyde and freshly formed iron ferrite was established, based upon wet-and-neutralization method for synthesizing iron ferrite and pearl-polymerization method for synthesizing rigid bead-type composite resin. The composite resin prepared by the above method shows stably high removal efficiency (maximally over 3.1 meq./gresin) to Co(II) species from wastewater in a wide range of solution pH. The wide range of applicable solution pH (i.e. pH 4.09 to 10.32) implies that the composite resin overcomes the limitations of the conventional ferrite process that is practically applicable only to alkaline conditions. It has been found that both ion exchange (by the organic resin constituent) and surface adsorption (by the inorganic adsorbent constituent) are major reaction mechanisms for removing Co(II) from wastewater, but surface precipitation results in the high sorption capacity to Co(II) beyond normal ion exchange capacity of the phenolsulphonic-formaldehyde resin. Standard enthalpy change derived from van't Hoff equation is 32.0 kJ{center_dot}mol-1 conforming to the typical range for chemisorption or ion exchange. In a wide range of equilibrium Co(II) concentration, the overall isotherm is qualitatively explained by the generalized adsorption isotherm concept proposed by McKinley. At the experimental conditions where …

In 1998, the U.S. Department of Energy (DOE) awarded the Management and Integration (M&I) contract for all five of the Oak Ridge Operations (ORO) facilities to Bechtel Jacobs Company LLC (BJC). At Oak Ridge National Laboratory (ORNL), a world renowned national laboratory and research and development facility, the BJC mission involves executing the DOE Environmental Management (EM) program. In addition to BJC's M&I contract, UT-Battelle, LLC, a not-for-profit company, is the Management and Operating (M&O) contractor for DOE on the ORNL site. As part of ORNL's EM program, legacy inactive facilities (i.e., reactors, nuclear material research facilities, burial grounds, and underground storage tanks) are transferred to BJC and are designated as remediation, decontamination and decommissioning (D&D), or long-term surveillance and maintenance (S&M) facilities. Facilities operated by both UT-Battelle and BJC are interspersed throughout the site and are usually in close proximity. Both UT-Battelle and BJC have DOE-approved Radiation Protection Programs established in accordance with 10 CFR 835. The BJC Radiological Control (RADCON) Program adapts to the M&I framework and is comprised of a combination of subcontracted program responsibilities with BJC oversight. This paper focuses on the successes and challenges of executing the BJC RADCON Program for BJC's ORNL Project through …

Hull wastes and some scrapped equipment are typical radioactive wastes generated from reprocessing process in Tokai Reprocessing Plant (TRP). Because hulls are the wastes remained in the fuel shearing and dissolution, they contain high radioactivity. Japan Nuclear Cycle Development Institute (JNC) has started the project of Hull Waste Treatment Facility (HWTF) to treat these solid wastes using compaction and incineration methods since 1993. It is said that Zircaloy fines generated from compaction process might burn and explode intensely. Therefore explosive conditions of the fines generated in compaction process were measured. As these results, it was concluded that the fines generated from the compaction process were not hazardous material. This paper describes the outline of the treatment process of hulls and results of safety evaluation.

Geostatistical interpolation of groundwater characterization data to visualize contaminant distributions in three dimensions is often hindered by the sparse distribution of samples relative to the size of the plume and scale of heterogeneities. Typically, placement of expensive monitoring wells is guided by the conceptualized plume rather than geostatistical considerations, focusing on contaminated areas rather than thoroughly gridding the plume boundary. The resulting data sets require careful analysis in order to produce plausible plume shells. A purely geostatistical approach is usually impractical; kriging parameters based on the observed data structure can extrapolate contamination far beyond the demonstrated extent of the plume. When more appropriate kriging parameters are selected, holes often occur in the interpolated distribution because realistic kriging ranges may not bridge large gaps between data points. Such artifacts obscure the probable location of the plume boundary and distort the contaminant distribution, obstructing quantitative modeling of remedial strategies. Two methods of constraining kriging can successfully eliminate these geostatistical artifacts. Laterally, the plume boundary may be controlled using a manually constructed mask that delineates the plan-view extent of the plume. After kriging, the mask is used to set all grid cells outside of the plume to a concentration of zero. Use of …

The United States Department of Energy is using or evaluating several Alternatives-to- Incineration (ATI) technologies for treating hazardous wastes and low-level mixed wastes. ATI treatment technologies may have the potential for generating gaseous or other emissions of polychlorinated dioxins/furans, a class of highly toxic compounds which are regulated to very low levels. At present, the emission limit for dioxins/furans from hazardous waste incinerators is 0.2 ng TEQ/dscm (0.4 ng TEQ/dscm w/TC). Emissions from ATI technologies are expected to be subject to similar restrictions.

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Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: GDH impact on an adjacent GDH and its attached piping; and GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.

Experiments on direct-contact heat exchange between molten metal and water for steam production were conducted. These experiments involved the injection of water into molten lead-bismuth eutectic for heat transfer measurements in a 1-D geometry. Based on the initial results of the experiments, the effects of the water flow rate and the molten metal superheat (temperature difference between molten metal and saturated water) on the volumetric heat transfer coefficient were discussed.

The BNFL Group of Companies owns and operates a number of nuclear licensed sites in the UK, Europe, and US. These cover fuel manufacture and reactor services; power reactors; spent fuel management; and nuclear decommissioning and clean up. To implement its environmental policy, BNFL needs to have tools and techniques that allow it to: (a) Respond appropriately to Environmental Trigger Events (ETEs), and to (b) Provide assurance that BNFL is able to manage contaminated land in the short to medium term (prior to site closure). As a consequence, over the past five years, BNFL has developed the Environmental Response Handbook (ERH). ETEs on a nuclear licensed site cover a number of scenarios: proactive action to remediate known contamination; change in behavior or location of known contamination (e.g. mobility increase); revision of permitted environmental limits on contaminants; other changes in regulatory regime; precedent set by a third party; and discovery of previously unknown contamination or new contaminating event. The main themes of the ERH are: global considerations for remediation on an operational site; detailed consideration of the application of remediation to the current ETE(s); a maintained ''toolkit'' of favored remediation techniques; and case studies and action plans In this paper the …

This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been …

In this paper, the performance of adsorbing and retarding 125I (substituted for 129I) for mixed minerals as buffer, backfill material was investigated. The distribution coefficient Kds by batch sorption experiments were determined for four kinds of minerals and one kind of bentonite under atmosphere. Sorption and desorption of radioiodine on several minerals were studied under low oxygen ambience at first time in the domestic, and apparent diffusion coefficient Da of 125I- was determined for mixed minerals under atmosphere. The results as follows: Distribution coefficient Kds of 125I- under atmosphere: bentonite is 3.23 ml{center_dot}g-1 , chalcopyrite is 72.42 ml{center_dot}g-1 , galena is 118.9 ml{center_dot}g-1 , pyrite is 1.93 ml{center_dot}g-1 , cinnabar is 55.48 ml{center_dot}g-1 , and the corresponding Kds under low oxygen ambience: galena is 88.48 ml{center_dot}g-1 , chalcopyrite is 6.47 ml{center_dot}g-1 . when pH of solution was in the range of 2.25-12.26, Kds of 125I- on chalcopyrite , galena, pyrite and cinnabar decreased with increase of pH under atmosphere. Kds of 125I- on several minerals increased with increase of mineral ratio in mixed materials under atmosphere. Under the same condition, Kds of 125I- on chalcopyrite and galena were larger than Kds of 125IO3 -. Sorption of 125I- on galena seems …

The INEEL has developed a software called TOWER (Tools for Optimal Waste and Exposure Reduction) to provide a new way to visualize and safely interact with hazardous facilities and equipment. This gives workers a comprehensive tool for planning and executing work in hazardous areas and the dismantling of complex and often dangerous sites. TOWER incorporates innovative technologies that can significantly reduce worker hazardous material doses and decrease the volume of waste going to disposal sites compared with currently used approaches. TOWER creates a 3-dimensional simulation of a facility that reveals in detail its solid components (pipes, valves, and pumps) as well as its invisible hazards (radiation, chemical, gas, and electric fields). Suggested segmentation can be shown on the facility model. TOWER also lets operators move simulated workers through the work site while monitoring their instantaneous and cumulative hazardous materials doses to help plan the best operational strategies. Because of its ability to visualize complex facilities, chart optimal dismantling and packaging steps, and track worker hazardous material dose, TOWER is a significant new way to plan and document work in hazardous areas including D&D projects, train workers, reduce worker exposure to hazardous materials, and lower project and waste disposal costs.

The design and performance capabilities of the Advanced Vitrification System (AVS) are described, together with the results of experimental tests. The AVS is an in-can melting system in which high-level waste (HLW) is vitrified directly inside the final disposal container. The AVS container, or module, consists of an outer stainless steel canister and an alumina-lined, inner graphite crucible, which is thermally insulated from the outer stainless canister. The graphite crucible is inductively heated to very high temperatures (up to 1500 C) by an external low frequency (30 Hertz) alternating current (AC) transformer coil. The actively cooled outer stainless canister remains at near ambient temperature. The HLW/frit mixture is fed into the hot graphite crucible, where it is vitrified. After cooldown, the HLW/frit feed and off-gas pipes are disconnected from the top of the module, which is then sealed and readied for shipment or storage. All radioactively contaminated melter components inside the module are disposed of along with the vitrified waste. The graphite crucible also provides a geologically stable barrier for the vitrified product. The AVS potentially can double HLW loading over that obtained from Joule melters; lower vitrification costs by about half; reduce the number of disposal canisters required by …

The Department of Energy Idaho Operations Office (DOE-ID) is making preparations to close two underground high-level waste (HLW) storage tanks at the Idaho National Engineering and Environmental Laboratory (INEEL) to meet Resource Conservation and Recovery Act (RCRA) regulations and Department of Energy (DOE) orders. Closure of these two tanks is scheduled for 2004 as the first phase in closure of the eleven 300,000 gallon tanks currently in service at the Idaho Nuclear Technology and Engineering Center (INTEC). Design, development, and deployment of a remotely operated tank cleaning system were completed in August 2001. The system incorporates many commercially available components, which have been adapted for application in cleaning high-level waste tanks. The system also uses existing waste transfer technology (steam-jets) to remove tank heel solids from the tank bottoms during the cleaning operations. By using this existing transfer system and commercially available equipment, the cost of developing custom designed cleaning equipment can be avoided. Remotely operated directional spray nozzles, automatic rotating wash balls, video monitoring equipment, decontamination spray-rings, and tank specific access interface devices have been integrated to provide a system that efficiently cleans tank walls and heel solids in an acidic, radioactive environment. This system is also compliant with …

The vitrification of high-level waste is the internationally recognized standard to minimize the impact to the environment resulting from waste disposal as well as to minimize the volume of conditioned waste to be disposed of. COGEMA has been vitrifying high-level waste industrially for over 20 years and is currently operating three commercial vitrification facilities based on a hot metal crucible technology, with outstanding records of safety, reliability and product quality. To further increase the performance of vitrification facilities, CEA and COGEMA have been developing the cold crucible melter technology since the beginning of the 1980s. This type of melter is characterized by a virtually unlimited equipment service life and a great flexibility in dealing with various types of waste and allowing development of high temperature matrices. In complement of and in parallel with the vitrification process, a glass formulation methodology has been developed by the CEA in order to tailor matrices for the wastes to be conditioned while providing the best adaptation to the processing technology. The development of a glass formulation is a trade-off between material properties and qualities, technical feasibility, and disposal safety criteria. It involves non-radioactive and radioactive laboratories in order to achieve a comprehensive matrix qualification. …

Three samples of {sup 239}Pu-{sup 241}Am-doped ceramics obtained from previous research were used for alteration experiments simulating corrosion of waste forms in ion-saturated solutions. These were ceramics based on: pyrochlore, (Ca,Hf,Pu,U,Gd){sub 2}Ti{sub 2}O{sub 7}, containing 10 wt.% Pu and 0.1 wt.% Am; zircon, (Zr,Pu)SiO{sub 4}, containing 5-6 wt.% Pu and 0.05 wt.% Am; cubic zirconia, (Zr,Gd,Pu)O{sub 2}, containing 10 wt.% Pu and 0.1 wt.% Am. All these samples were milled in an agate mortar to obtain powder with particle sizes less than 30 micron. Sample of granite taken from the depth 500-503 m was studied and then used for preparing ion-saturated water solutions. A rock sample was ground, washed and classified. A fraction with particle size 0.10-0.25 mm was selected for alteration experiments. Powdered ceramic samples were separately placed into deionized water together with ground granite (approximately 1gram granite per 12-ml water) in special Teflon{trademark} vessels and set at 90 C in the oven for 3 months. After alteration experiments, the ceramic powders were studied by precise XRD analysis. Aqueous solutions and granite grains were analyzed for Am and Pu contents. The results show that alteration did not cause significant phase transformation in all ceramic samples. For all altered samples, …

The U.S. Department of Energy (DOE) is considering the possible recommendation of a site at Yucca Mountain, Nevada, for the potential development of a geologic repository for the disposal of high-level radioactive waste and spent nuclear fuel. To facilitate public review and comment, in May 2001 the DOE released the Yucca Mountain Science and Engineering Report (S&ER) (1), which presents technical information supporting the consideration of the possible site recommendation. The report summarizes the results of more than 20 years of scientific and engineering studies. Based on internal reviews of the S&ER and its key supporting references, the Total System Performance Assessment for the Site Recommendation (TSPA-SR) (2) and the Analysis Model Reports and Process Model Reports cited therein, the DOE has recently identified and performed several types of analyses to supplement the treatment of uncertainty in support of the consideration of a possible site recommendation. The results of these new analyses are summarized in the two-volume report entitled FY01 Supplemental Science and Performance Analysis (SSPA) (3,4). The information in this report is intended to supplement, not supplant, the information contained in the S&ER. The DOE recognizes that important uncertainties will always remain in any assessment of the performance of …

The program to treat and dispose of the highly radioactive wastes stored in underground tanks at the U.S. Department of Energy's Hanford site has been studied. A strategy/management approach to achieve an acceptable (technically sound) end state for these wastes has been developed in this study. This approach is based on assessment of the actual risks and costs to the public, workers, and the environment associated with the wastes and storage tanks. Close attention should be given to the technical merits of available waste treatment and stabilization methodologies, and application of realistic risk reduction goals and methodologies to establish appropriate tank farm cleanup milestones. Increased research and development to reduce the mass of non-radioactive materials in the tanks requiring sophisticated treatment is highly desirable. The actual cleanup activities and milestones, while maintaining acceptable safety standards, could be more focused on a risk-to-benefit cost effectiveness, as agreed to by the involved stakeholders and in accordance with existing regulatory requirements. If existing safety standards can be maintained at significant cost savings under alternative plans but with a change in the Tri-Party Agreement (a regulatory requirement), those plans should be carried out. The proposed strategy would also take advantage of the lessons learned …

The Ashtabula River in northern Ohio contains a large amount of sediment containing quantities of NORM and TENORM from previous industrial activities at nearby mineral processing plants. Due to PCB contamination, these sediments were to be dredged and isolated in a landfill to be constructed by the responsible parties. Unfortunately, the State of Ohio has determined that these wastes may not be disposed of in this manner, and this determination has resulted in delaying the remediation project. Computer models performed using the RESRAD computer code indicate that isolating these wastes in this manner will reduce dose to the nearby population because the NORM/TENORM will be safely buried beneath a compacted clay cover and isolated from all sources of exposure. In fact, radiation doses (including radon emanation) from these wastes in a properly maintained landfill are significantly lower than in the present condition, and the reduction is even more marked for NORM/TENORM in tailings piles. This suggests that, in many cases, disposal of NORM/TENORM wastes in on-site landfills may be a cost-effective and dose-conscious method of disposal, if regulatory issues can be resolved.

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating the development of a comprehensive and quantitative risk model framework for environmental management activities at the site. Included are waste management programs (high-level waste, transuranic waste, low-level waste, mixed low-level waste, spent nuclear fuel, and special nuclear materials), major environmental restoration efforts, major decontamination and decommissioning projects, and planned long-term stewardship activities. Two basic types of risk estimates are included: risks from environmental management activities, and long-term legacy risks from wastes/materials. Both types of risks are estimated using the Environment, Safety, and Health Risk Assessment Program (ESHRAP) developed at the INEEL. Given these two types of risk calculations, the following evaluations can be performed: risk evaluation of an entire program (covering waste/material as it now exists through disposal or other e nd states); risk comparisons of alternative programs or activities; comparisons of risk benefit versus risk cost for activities or entire programs; ranking of programs or activities by risk; ranking of wastes/materials by risk; evaluation of site risk changes with time as activities progress; and integrated performance measurement using indicators such as injury/death and exposure rates. This paper discusses the definition and calculation of legacy risk measures and associated issues. …

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins.

While the Nuclear Regulatory Commission has had specific and dose-based standards for the release of liquids and gases for a long time, there are no regulatory mechanisms in place for the release of solid bulk materials from a nuclear power plant. Even though free releases of small quantities of solid materials continue under existing guidelines from the operating plants, the regulatory void creates major difficulties for the bulk materials that result from the decommissioning of a nuclear site. Decommissioning of a commercial nuclear power plant generates large quantities of solid bulk materials such as concrete, metal, and demolition debris. Disposition of such materials has a large impact on the overall decommissioning cost. Yet, there are no clear and cost-effective alternatives for the disposal of these materials from a regulatory perspective. This paper discusses the methodologies for clearance of solid materials1, their applicability to the disposition of bulk materials, and the impact of lack of consistent free release standards on the decommissioning projects and costs.