NUKEM monitor FAMOS is designed to measure the burn up of fuel assemblies (FA’s) and allows the calculation of the content of fissile material and other actinides. The knowledge of the isotopic composition and activities of FA’s is required for a safe handling and storage of the elements. The monitor analyses the measurement results on the basis of burn up calculations which improve the quality of the results, with respect to previous systems, which did not use those.

The United States Department of Energy (US DOE) is conducting research and development (R&D) activities within the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R&D activities are ongoing at nine national laboratories, and are divided into two major topical areas: (1) storage and transportation research, and (2) disposal research. Storage R&D focuses on closing technical gaps related to extended storage of UNF. For example, uncertainties remain regarding high-burnup nuclear fuel cladding performance following possible hydride reorientation and creep deformation, and also regarding long-term canister integrity. Transportation R&D focuses on ensuring transportability of UNF following extended storage, addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R&D focuses on identifying multiple viable geologic disposal options and addressing technical challenges for generic disposal concepts in various host media (e.g., mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal in crystalline rock). R&D will transition to site-specific challenges as national policy advances. R&D goals at this stage are to increase confidence in the robustness of generic disposal concepts, to reduce generic sources of uncertainty that may impact the viability of disposal concepts, and to develop science and engineering tools that will support the selection, characterization, and ultimately licensing of a repository. The US DOE has also initiated activities that can be conducted within the constraints of the Nuclear Waste Policy Act to facilitate the development of an interim storage facility and supporting transportation infrastructure.

The purpose of this paper is to examine the energy and economic value of used nuclear fuel and depleted uranium. In the USA these materials are considered to be wastes. As such they are candidates for permanent disposal. The disposal of these materials in a manner that isolates them from human contact is a scientific, engineering, economic and political problem. Isolation can be defined requiring no potential human contact during a stated period of time due to failed containment. The period of isolation is expected to be in excess of one million years. The uranium in the used fuel and the depleted uranium left over from the enrichment process represent about 99.5 percent of the uranium that was removed from the ground by mining uranium ores. If these materials can be utilized they would not be considered to be wastes. In addition they would carry a positive economic value. The value of these materials, based on the energy that can be extracted from the uranium, exceeds $100 trillion. Based on this the conclusion is that the material is a very valuable resource and definitely is not a waste.

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications.

Treatment of used nuclear fuel through solvent extraction separation processes is hindered by radiolytic damage from radioactive isotopes present in used fuel. The nature of the damage caused by the radiation may depend on the radiation type, whether it be low linear energy transfer (LET) such as gamma radiation or high LET such as alpha radiation. Used nuclear fuel contains beta/gamma emitting isotopes but also a significant amount of transuranics which are generally alpha emitters. Studying the respective effects on matter of both of these types of radiation will allow for accurate prediction and modeling of process performance losses with respect to dose. Current studies show that alpha radiation has milder effects than that of gamma. This is important to know because it will mean that solvent extraction solutions exposed to alpha radiation may last longer than expected and need less repair and replacement. These models are important for creating robust, predictable, and economical processes that have strong potential for mainstream adoption on the commercial level. The effects of gamma radiation on solvent extraction ligands have been more extensively studied than the effects of alpha radiation. This is due to the inherent difficulty in producing a sufficient and confluent dose of alpha particles within a sample without leaving the sample contaminated with long lived radioactive isotopes. Helium ion beam and radioactive isotope sources have been studied in the literature. We have developed a method for studying the effects of high LET radiation in situ via 10 B activation and the high LET particles that result from the 10 B(n,α) 7 Li reaction which follows. Our model for dose involves solving a partial differential equation representing absorption by 10 B of an isentropic field of neutrons penetrating a sample. This method has been applied to organic solutions of TBP and CMPO, two ligands common in TRU solvent extraction treatment processes. Rates of degradation of TBP and CMPO and their respective degradation products in the presence of high LET radiation are presented and discussed. These results are also compared to gamma studies performed in our lab and other gamma and alpha studies found in the literature. The possible application of this method to a variety of other solvent extraction ligands to study the effects of high LET radiation is also considered.

Radioactive waste systems and structures (RWSS) are safety-critical facilities in need of monitoring over prolonged periods of time. Structural health monitoring (SHM) is an emerging technology that aims at monitoring the state of a structure through the use of networks of permanently mounted sensors. SHM technologies have been developed primarily within the aerospace and civil engineering communities. This paper addresses the issue of transitioning the SHM concept to the monitoring of RWSS and evaluates the opportunities and challenges associated with this process. Guided wave SHM technologies utilizing structurally-mounted piezoelectric wafer active sensors (PWAS) have a wide range of applications based on both propagating-wave and standing-wave methodologies. Hence, opportunities exist for transitioning these SHM technologies into RWSS monitoring. However, there exist certain special operational conditions specific to RWSS such as: radiation field, caustic environments, marine environments, and chemical, mechanical and thermal stressors. In order to address the high discharge of used nuclear fuel (UNF) and the limited space in the storage pools the U.S. the Department of Energy (DOE) has adopted a “Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste” (January 2013). This strategy endorses the key principles that underpin the Blue Ribbon Commission’s on America’s Nuclear Future recommendations to develop a sustainable program for deploying an integrated system capable of transporting, storing, and disposing of UNF and high-level radioactive waste from civilian nuclear power generation, defense, national security, and other activities. This will require research to develop monitoring, diagnosis, and prognosis tools that can aid to establish a strong technical basis for extended storage and transportation of UNF. Monitoring of such structures is critical for assuring the safety and security of the nation’s spent nuclear fuel until a national policy for closure of the nuclear fuel cycle is defined and implemented. In addition, such tools can provide invaluable and timely information for verification of the predicted mechanical performance of RWSS (e.g. concrete or steel barriers) during off-normal occurrence and accident events such as the tsunami and earthquake event that affected Fukushima Daiichi nuclear power plant. The ability to verify the conditions, health, and degradation behavior of RWSS over time by applying nondestructive testing (NDT) as well as development of nondestructive evaluation (NDE) tools for new degradation processes will become challenging. The paper discusses some of the challenges associated to verification and diagnosis for RWSS and identifies SHM technologies which are more readily available for transitioning into RWSS applications. Fundamental research objectives that should be considered for the transition of SHM technologies (e.g., radiation hardened piezoelectric materials) for RWSS applications are discussed. The paper ends with summary, conclusions, and suggestions for further work.

On 26 April 1986, the worst nuclear power plant accident in history occurred at the Chernobyl plant in Ukraine (then part of the Soviet Union). The destruction of Unit 4 sent highly radioactive fallout over Belarus, Russia, Ukraine, and Europe. The object shelter — a containment sarcophagus — was built in November 1986 to limit exposure to radiation. However, it has only a planned 25-year lifespan and would probably not survive even a moderate seismic event in a region that has more than its share of such events. It was time to take action. One of the largest tasks that are in progress is the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant Unit. One of the major mechanical handling systems to be installed in the new safe confinement is the Main Cranes System. The planned decontamination and decommissioning or dismantling activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the new safe confinement, will require large and sophisticated cranes. The article will focus on the current progress of the new safe confinement and of the main cranes system for the decommissioning or dismantling activities.

Knowledge of the radiological state of processes and equipment of a nuclear facility is essential to supervise a wide variety of sensitive tasks: building of intervention scenarios in order to optimize maintenance or dismantling operations, optimization of waste categorization, monitoring the effectiveness of decontamination processes, monitoring of nuclear facility decommissioning, etc. In order to meet the diversity of the issues involved, the CEA has developed in situ radiological characterization methods and techniques to acquire reliable radiological data. The data gathered is necessary to build robust radiological models which can be used as input data for dismantling studies. Over the last 30 years, the main nuclear measurement techniques, such as gamma imaging and gamma spectrometry, have been widely deployed by the CEA on many facilities under dismantling and more recently, on the Phénix nuclear power plant. Phénix was a small-scale prototype of a sodium-cooled fast breeder reactor, located at the Marcoule nuclear site. These techniques have been implemented on this reactor in order to meet the increased need for radiological knowledge to prepare for future dismantling operations following its final shutdown in 2009. This paper will focus on the description of three radiological characterization methods which take advantage of advanced nuclear measurement techniques. For each method, an example of a specific application on the Phénix reactor will be presented. Firstly, the so-called “gamma scanning” method will be explained. The objective of this method is to determine the activity profile of equipment based on collimated gamma spectrometry measurements with compact probes like CdZnTe. This method was applied to a neutron shielding of the reactor core to estimate the 60 Co activity profile. Then the measured activities helped to validate the theoretical activities resulting from neutron activation calculations. Secondly, this paper will focus on the interest of combining different measurement techniques such as gamma imaging, gamma spectrometry and collimated/uncollimated dose rate mapping to characterize complex equipment or processes. In this case, a specific methodology was developed to define the radiological state of a shielded cell used for the processing of irradiated nuclear fuels. Finally, an isotopic characterization technique using a high purity germanium detector will be discussed. This technique was applied to a non-irradiated fertile fuel sub-assembly in order to determine the level of uranium enrichment. The processing was carried out by three types of analysis: two automated, with the MGA-U and IGA software, and one manual.

A major prerequisite in order for civilian commercial nuclear energy production to qualify as sustainable energy production is that systems for the management of the nuclear waste legacy are in operation. These waste types are present in a range from very low short lived waste (VLLW) to long lived high level waste (HLW) (including the used nuclear fuel). The second prerequisite is that financial responsibilities or other constraints must not be passed on to coming generations. The first condition for qualification corresponds to the Polluters Pays Principle (PPP) which demands that the responsibility for the waste management rests solely with the polluter. The second qualification corresponds to the principle of fairness between generations and thus concerns the appropriate distribution of responsibilities between the generations. It is important to note that these two conditions must be met simultaneously, and that compliance with both is a necessary prerequisite in order for commercial use of nuclear power to qualify as a semi-sustainable energy source. Financial and technical planning for dismantling and decommissioning of nuclear installations cannot be regarded as successful unless it rests upon a distinctive way to describe and explain the well-founded values of different groups of stakeholders. This cumbersome task can be underpinned by transparent and easy to grasp models for calculation and estimation of future environmental liabilities. It essential that a systematic classification is done of all types of costs and that an effort is done to evaluate the precision level in the cost estimates. In this paper, a systematic and transparent way to develop a parametric approach that rest upon basic accounting standards is combined with data about younger stakeholder’s values towards decommissioning and dismantling of nuclear installation. The former entity rests upon theoretical and practical methods from business administration, whilst the latter is based on current survey data retrieved from 667 personal interviews in one town in Poland and one town in Slovakia with a near 100 % response rate. The main conclusions from this field study may be summarised as follows: • Sustainable energy sources are prioritised. • Around one quarter of the respondents regards nuclear power as a future semi-sustainable commercial energy production mode subject to that the waste is managed in a sustainable, environmental friendly and safe way. • The values are to a significant degree positioned on health, safety and environmental (HSE) attributes. • The polluter pays principle is honoured. • There are doubts regarding the compliance with these principles due to risks for delays in the implementation phase of repositories for disposal of the nuclear residues. • 1/5 th of the respondents expressed an openness to reprocessing (which is linked to the concept of “new nuclear power”).

The underwater spectrometric system for survey the bottom of material science multi-loop reactor MR ponds was elaborated. This system uses CdZnTe (CZT) detectors that allow for spectrometric measurements in high radiation fields. The underwater system was used in the spectrometric survey of the bottom of the MR reactor pool, as well as in the survey located in the MR storage pool of highly radioactive containers and parts of the reactor construction. As a result of these works irradiated nuclear fuel was detected on the bottom of pools, and obtained estimates of the effective surface activity detected radionuclides and created by them the dose rate.

The Springfields facility manufactures nuclear fuel products for the UK’s nuclear power stations and for international customers. Fuel manufacture is scheduled to continue into the future. In addition to fuel manufacture, Springfields is also undertaking decommissioning activities. Today it is run and operated by Springfields Fuels Limited, under the management of Westinghouse Electric UK Limited. The site has been operating since 1946 manufacturing nuclear fuel. As part of the decommissioning activities, there was a need was to quantify contamination in a large redundant building. This building had been used to process uranium derived from uranium ore concentrate but had also processed a limited quantity of recycled uranium. The major non-uranic contaminant was Tc-99. The aim was to be able to identify any areas where the bulk activity exceeded 0.4 Bq/g Tc-99 as this would preclude the demolition rubble being sent to the local disposal facility. The problems associated with this project were the presence of significant uranium contamination, the realisation that both the Tc-99 and the uranium had diffused into the brickwork to a significant depth and the relatively low beta energy of Tc-99. The uranium was accompanied by Pa-234m, an energetic beta emitter. The concentration/depth profile was determined for several areas on the plant for Tc-99 and for uranium. The radiochemical analysis was performed locally but the performance of the local laboratory was checked during the initial investigation by splitting samples three ways and having confirmation analyses performed by 2 other laboratories. The results showed surprisingly consistent concentration gradients for Tc-99 and for uranium across the samples. Using that information, the instrument response was calculated for Tc-99 using the observed diffusion gradient and averaged through the full 225 mm of brick wall, as agreed by the regulator. The Tc-99 and uranium contributions to the detector signal were separated using a simple absorber, which essentially eliminated the Tc-99 count rate and reduced the uranium contribution only marginally. The outcome of the project was that it was possible to demonstrate that the complete building met the criterion for acceptance at the local waste facility.

It is essential the environmentally safe industrial production of nuclear fuel especially in the case of uranium extraction by In Situ Leaching, when the environment and the deep extraction of uranium are important problems. In the presented paper it is studied the feasibility of the application of an electrodialysis method for the deep extraction of uranium and radium from liquid (acid) streams. It is proposed to apply a new electro-hydro-metallurgical [1] extraction and refining method to ensure the necessary extraction level of elements. In the same time the new method ensures the recycling of acids used in the process.

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR’s), Boiling Water Reactors (BWR’s), Gas Cooled Reactors (GCR’s) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals segmentation and packaging project. The purpose of this paper is to provide an overview of the Westinghouse reactor internals segmentation experience by illustrating projects related to various types of reactors and providing feedback from project execution.

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technologies (OFCT) has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel (UNF) and high level radioactive waste (HLW). The U.S. has, in accordance with the U.S. Nuclear Waste Policy Act (as amended), focused efforts for the past twenty-plus years on disposing of UNF and HLW in a geologic repository at Yucca Mountain, Nevada. The recent decision by the U.S. DOE to no longer pursue the development of that repository has necessitated investigating alternative concepts for the disposal of UNF and HLW that exists today and that could be generated under future fuel cycles. The disposal of UNF and HLW in a range of geologic media has been investigated internationally. Considerable progress has been made by in the U.S and other nations, but gaps in knowledge still exist. The U.S. national laboratories have participated in these programs and have conducted R&D related to these issues to a limited extent. However, a comprehensive R&D program investigating a variety of storage, geologic media, and disposal concepts has not been a part of the U.S. waste management program since the mid 1980s because of its focus on the Yucca Mountain site. Such a comprehensive R&D program is being developed and executed in the UFDC using a systematic approach to identify potential R&D opportunities. This paper describes the process used by the UFDC to identify and prioritize R&D opportunities. The U.S. DOE has cooperated and collaborated with other countries in many different “arenas” including the Nuclear Energy Agency (NEA) within the Organisation for Economic Co-operation and Development (OECD), the International Atomic Energy Agency (IAEA), and through bilateral agreements with other countries. These international activities benefited the DOE through the acquisition and exchange of information, database development, and peer reviews by experts from other countries. Recognizing that programs in other countries have made significant advances in understanding a wide range of geologic environments, the UFDC has developed a strategy for continued, and expanded, international collaboration. This paper also describes this strategy.

This paper describes the design and operation of the Combined Technology Automated Waste Characterisation System (CTA-WCS) at JRC Ispra. The WCS was designed for the measurement of fission products and uranium and plutonium containing waste arising from nuclear fuel and nuclear materials processing and reactor operations. The WCS covers a range of activity including Low and Intermediate Level Waste (LLW and ILW). The system is designed to measure the waste in 200 and 400 (440) litre drums with a maximum drum weight of 1500 kg. Gamma-ray measurements of radio-nuclide content are performed by a gamma ray measurement station which functions as either a Segmented Gamma Scanner (SGS) or Tomographic Segmented Gamma-ray Scanner (TSGS). Either of these two techniques may be employed to perform the functions of drum screening, non-destructive examination (NDE) and, where appropriate, final drum assay. Coupled to the gamma ray station is a surface dose-rate measurement station, which employs 6-shielded Geiger-Muller detectors. Active and passive neutron measurements are performed by an advanced, graphite lined Differential Die-away (DDA) system, which comprises the neutron measurement station. The DDA performs conventional passive neutron totals, coincidence, and multiplicity counting and active DDA total neutron counting. Data analysis is based on the use of a range of matrix calibrations, some determined by Monte Carlo analysis. Linking the gamma ray and neutron measurement stations is an automated roller conveyor with a 20 drum buffer capability and a weight measurement station. Drums are identified by bar code reading technology. Once loaded, the system performs automatic assay of up to 20 drums and then returns the drums to the buffer position on the conveyor. The first WCS of this type was supplied to the European Commission at the Joint Research Centre (JRC) Ispra in northern Italy and it was commissioned at the end of 2007. It is now in a phase of pilot operation. Results will be presented from the first drum measurement campaign.

CIEMAT, formerly the Nuclear Energy Board (JEN), is the Spanish Centre for Energy-Related, Environmental and Technological Research. Located in Madrid, it used to have more than 60 facilities in operation that allowed a wide range of activities in the nuclear field and in the application of ionising radiations. Particularly significant among these facilities were the research reactors, particle accelerators, hot cells and nuclear fuel manufacturing and processing plants. At present the centre, which is authorised as a single nuclear facility, includes various installations, some of them are now obsolete, shut down and in dismantling phases. In 2000 CIEMAT started the “Integrated plan for the improvement of CIEMAT installations (PIMIC)”, which includes activities for the decontamination, dismantling and rehabilitation of obsolete facilities. This paper will describe the decommissioning process carried out at CIEMAT Research Centre, and will review the key aspects of the project, including site remediation and release.

Located in the scenic hills of Western New York, 35 miles south of Buffalo, the 68 hectare West Valley Demonstration Project (WVDP) is a unique and challenging environmental cleanup project that currently manages High Level, Transuranic, and Greater than Class C wastes. Before the U.S. Department of Energy (DOE) assumed the responsibility of cleaning up the site, the site was the location of the only commercial spent nuclear fuel (SNF) reprocessing facility to operate in the United States. Operated by Nuclear Fuels Services from 1966–72, the site was owned by the State of New York and licensed by the Atomic Energy Commission. During operations, the plant reprocessed approximately 640 metric tons of commercial and defense nuclear fuel. When commercial operations were discontinued and the facilities were returned to New York State, there were nearly 2,271,247 liters (nearly 600,000 gallons) of liquid high-level radioactive waste (HLW) in an underground storage tank, there were approximately 750 unprocessed fuel assemblies, and there was a highly contaminated Main Plant Process Building (MPPB). The West Valley Demonstration Project Act of 1980 (the Act) authorized DOE to conduct a cleanup of the site, in cooperation with the New York State Energy Research and Development Authority (NYSERDA), the State agency responsible for managing the property. The Nuclear Regulatory Commission (NRC) license was placed in abeyance and DOE arrived on site in 1981. Since that time, several significant cleanup milestones have been completed, including vitrification (solidification in a glass matrix) of the high level liquid waste, containing approximately 15 million curies, into 275 high level waste canisters and the disposition of more than 19,000 subsequent drums of low-level radioactive waste (LLW); the shipment of the remaining 125 SNF assemblies that were in storage on site; and the shipment of more than 304,800 cubic meters (1,000,000 cubic feet) of LLW for offsite disposal. More recent accomplishments at West Valley have included deactivation and decontamination of the still-very-contaminated MPPB, management of the underground high level waste tanks (4) and their associated vaults (3), and processing and packaging of both legacy waste and waste generated as a result of decommissioning activities. This paper focuses on the high level, greater than Class C, and Transuranic waste challenges at West Valley and the options to address those challenges.

The cancellation of the Yucca Mountain repository program in the Unites States raises the prospect of very long-term storage (i.e., >120 years) and deferred transportation of used fuel at the nuclear power plant sites. While long-term storage of used nuclear fuel in dry cask storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSIs) is already a standard practice among U.S. utilities, recent rule-making activities of the U.S. Nuclear Regulatory Commission (NRC) indicated additional flexibility for the NRC licensees of ISFSIs and certificate holders of the DCSSs to request initial and renewal terms for up to 40 years. The proposed rule also adds a requirement that renewal applicants must provide descriptions of aging management programs (AMPs) and time-limited aging analyses (TLAAs) to ensure that the structures, systems, and components (SSCs) that are important to safety in the DCSSs will perform as designed under the extended license terms. This paper examines issues related to managing aging effects on DCSSs for very long-term storage (VLTS) of used fuels, capitalizing on the extensive knowledge and experience accumulated from the work on aging research and life cycle management at Argonne National Laboratory (ANL) over the last 30 years. The technical basis for acceptable AMPs and TLAAs is described, as are generic AMPs and TLAAs that are being developed by Argonne under the support of the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign for R&D on extended long-term storage and transportation.

A way to recover pure uranium as UO 4 from UO 2 and (U,Gd)O 2 nuclear fuel scraps by using carbonate solution with hydrogen peroxide was studied and the nuclear fuel fabrication characteristics of the UO 2 powder prepared from the UO 4 was evaluated.

A permeable treatment wall (PTW) was designed and installed at the West Valley Development Project (WVDP), a former commercial nuclear fuel reprocessing facility in New York, to passively remove and contain the expansion of strontium-90 (Sr-90) in the site groundwater. AMEC engineers and geologists have collaborated with researchers at the State University of New York at Buffalo (UB) and West Valley Environmental Services LLC to design and install an approximately 260 meter (860-foot) long by 0.9 meter (3-foot) thick zone of granular zeolite (a natural aggregate composed of approximately 85 percent of the mineral clinoptilolite) that will remove Sr-90 in situ from groundwater through ion-exchange reactions. The PTW was designed to meet the functional objectives for up to 20 years; performance monitoring will be conducted regularly and will be used to assess the lifetime efficacy of the PTW. The zeolite-filled PTW was the selected remedial alternative due to its hydraulically passive operations and lower life-cycle cost over other more traditional active treatment alternatives such as pump and treat. The design relied heavily on detailed site characterization of site soils and groundwater conditions and bench and pilot scale evaluations of various zeolite materials. The design specified the use of a one-pass trencher to simultaneously remove unconsolidated aquifer material composed of glaciofluvial-derived silt, sand, and gravel from ground surface to depths up to approximately 9 meters (30 feet) and replace the excavated zone with zeolite along the entire alignment while keying the PTW at least 0.9 meter (3 feet) into the underlying low-permeability glacial till. Several technology demonstrations were conducted to assess implementability using the one-pass trencher prior to completing the final design. During full-scale implementation, excavated sediment was conveyed directly into a prefabricated containment structure pending final disposition. The passive design provides a cost effective and sustainable alternative for treatment of Sr-90 and potentially other exchangeable radioactive ions in groundwater where these constituents migrate in unconsolidated materials.