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Production of high specific activity (HSA) 60Co has recently resumed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The technical steps of performing in-canal assay of HSA 60C...

Feedstock cost is a major variable cost component in conversion to biofuels and chemicals. Consistent feedstock quality is critically important to achieve high product yield and maximum onstream time. Traditionally, raw biomass materials are delivered directly to the biorefineries where they are preprocessed to feedstock prior to being converted to products. Since many types of biomass materials-including agricultural residues, energy crops, and logging residues-are harvested according to growth cycles and optimal harvesting time, just-in-time steady supply of raw biomass to the biorefineries is not possible. Instead, biomass materials are stored, then delivered to the biorefineries as needed. Experience to date indicates that this approach has caused many issues related to logistics, biomass losses due to microbial degradation and fire, and inconsistent feedstock quality due to variability in the properties of as-delivered biomass. These factors have led to high feedstock cost, low throughput, and low product yield for the biorefineries. Idaho National Laboratory has developed a new strategy to address the problems encountered in the traditional approach in biomass feedstock supply, storage, and preprocessing mentioned above. The key components of this strategy are (1) preservation and preconditioning of biomass during storage, (2) utilization of all the biomass, including minor components that are normally considered wastes or contaminants, and (3) maximization of the value of each component. This new approach can be accomplished using feedstock preprocessing depots located near the biomass-production sources.

The EPA's Water Security Test Bed (WSTB) facility is a full-scale representation of a drinking water distribution system. In collaboration with the Idaho National Laboratory (INL), EPA designed the WSTB facility to support full-scale evaluations of water infrastructure decontamination, real-time sensors, mobile water treatment systems, and decontamination of premise plumbing and appliances. The EPA research focused on decontamination of 1) Bacillus globigii (BG) spores, a non-pathogenic surrogate for Bacillus anthracis and 2) Bakken crude oil. Flushing and chlorination effectively removed most BG spores from the bulk water but BG spores still remained on the pipe wall coupons. Soluble oil components of Bakken crude oil were removed by flushing although oil components persisted in the dishwasher and refrigerator water dispenser. Using this full-scale distribution system allows EPA to 1) test contaminants without any human health or ecological risk and 2) inform water systems on effective methodologies responding to possible contamination incidents.

In 2010, the Idaho National Laboratory was in the process of removing legacy materials from one of their hot cells. As part of this clean-out effort, five metal capsules and some loose zeolite material were identified as test specimens produced in the late 1970s as part of research and development (R&D) conducted under the Airborne Waste Management Program. This specific R&D effort examined the encapsulation of 85Kr within a collapsed zeolite structure for use as a potential waste form for long-term storage. These reclaimed capsules and loose material presented a unique opportunity to study a potential 85Kr waste form after three half-lives have elapsed. Of the five capsules, the walls of two had been cut or breached during previous experiments. The aim of this study was to produce mounted samples from the two breached samples that could be handled with minimal shielding, assess the physical condition and chemical composition of the capsule walls for each breached sample, and determine if any loss...

A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO2–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications made to a version of the MELCOR code by the Idaho National Laboratory (INL) that allows its application to alternate fuel-cladding systems. Preliminary analysis results are presented in this paper for candidate silicon carbide and FeCrAl cladding concepts.

Integrated Energy System (IES) is a core initiative at Idaho National Laboratory where energy to molecules and materials (E2M2) pillar plays a critical role, due to the widely available distributed feedstock, energy input (renewable and/or nuclear energy associated electricity and excess heat) as well as the resulting industrial electrification. These factors offer new opportunities to those technologies with smaller scale, modularized/distributed, and driven by an integrated system of renewable and other energy sources that can show promise for increased productivity with decreased energy, capital, and operating costs. Intermediate temperature electrochemical processing through the solid oxide cells is a promising strategy because it can efficiently utilize electrons and heat while overcoming and/or mitigating the challenges of their lower and higher temperature counterparts, including 1) the use of precious metal; 2) sluggish kinetics; 3) fast degradation; 4) costly interconnect materials. Despite of these promising features, this technology still faces many challenges that include but not limited to immature process, relatively low catalyst activity, and unproven operation durability. At Idaho National Laboratory, we have made significant achievements on electrosynthesis of intermediates, fuels and chemicals using natural gas, steam, carbon dioxide and nitrogen through protonic ceramic membrane reactors. The research outcomes have the potential to establish a globally competitive U.S. industry for converting the Nation’s abundant supply of feedstock into conversion-ready intermediates and/or high demand fuels and chemicals.

Domestic production of high specific activity 60Co was halted after a target rupture in 2012 at the Advanced Test Reactor (ATR). The Isotope Program (IP) within the US Department of Energy (DOE) Office of Science tasked a multilaboratory team of researchers and managers from Oak Ridge and Idaho National Laboratories with the redesign the radioisotope capsule. The objective of this effort was to create a more robust and reliable design, compared to the pre-2012 target. The team successfully completed this task to produce the DOE-IP cobalt (Co) production capsule design. Furthermore, 66 capsules were successfully fabricated by Oak Ridge National Laboratory (ORNL) and delivered to Idaho National Laboratory (INL) for irradiation in the ATR between January 2014 and October 2016. This paper describes the efforts of the team to prepare and disposition the two initial DOE-IP Co production capsules that were processed in March 2020. These efforts include performing accurate production predictions, experimentally validating predictions with assay measurements, shipping with the Orano-furnished Battelle Energy Alliance Research Reactor shipping package, and disassembling capsules at the isotope vendor site.

A full-scale process has been developed to immobilize fission products that accumulate within the Mark IV electrorefiner (ER) electrolyte at Idaho National Laboratory. ER salt was blended with treatment additives, followed by pressureless consolidation (PC) in a furnace to produce a durable ceramic waste form (CWF). The goal is the development of a process to consolidate actual radioactive ER salt into a form suitable for transportation and disposal. Four batches (300 to 400 kg per batch) of full-scale pre-qualification material preparation runs have been prepared. From these four batches of nonradioactive salt-loaded surrogate material, three full-scale PC trials have been conducted. The first PC test run, established equipment parameters with a basic CWF container design. The second trial included a modified CWF container design, real-time measurement of CWF consolidation, and an audio recording to identify cracking during the CWF cool-down. During the third trial, salt was doped (from the fourth material preparation batch) to create a nonradioactive salt material and to more closely represent actual ER salt. The second and third trials were also used to validate a model developed for the CWF. The CWF model is beneficial for understanding and predicting the physical processes that occur during the heat cycle. This would be particularly useful when the CWF is located in a hot cell, which makes accessing and examining a CWF difficult.

High temperature steam electrolysis (HTSE) is a promising technology for large-scale hydrogen production. However, research on HTSE performance above the kW level is limited. This paper presents the results of 4 kW HTSE long-term test completed in a multi-kW test facility recently developed at the Idaho National Laboratory (INL). The 4 kW HTSE unit consisted of two solid oxide electrolysis stacks electrically connected in parallel, each of which included 40 electrode-supported planar cells. A current density of 0.41 A cm −2 was used for the long-term operating at a constant current mode, resulting in a theoretical hydrogen production rate about 23 slpm. A demonstration of 830 h stable operation was achieved with a degradation rate of 3.1% per 1000 h. The paper also includes detailed descriptions of the piping layout, steam generation and delivery system, test fixture, heat recuperation system, hot zone, instrumentation, and operating conditions. This successful demonstration of multi-kW scale HTSE unit will help to advance the technology toward near-term commercialization.

This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) advanced gas reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the advanced test reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended [1]. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight) [1]. The results of the FSAR for TREAT can be found in [2]. The TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for themore » fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO2 vs U3O8 [1]. In this report, the results are documented pertaining to the choice mentioned above (UO2 vs U3O8). The conclusion in favor of using UO2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and current literature were reviewed. This analysis was supplemented by the construction and examination of the U-O and U-O-C phase diagrams at pressure close to negligent, thereby mimicking the conditions in which nuclear fuel is supposed to function inside the zirconium-based cladding in the reactor. Finally, our conclusion in favor of the UO2 down selection was summarized and explained in the last Section of this document.« less

This report describes conditions, as required by the state of Idaho Wastewater Reuse Permit (#LA-000141-03), for the wastewater land application site at the Idaho National Laboratory Site’s Central Facilities Area Sewage Treatment Plant from November 1, 2014, through October 31, 2015.

This report summarizes radiological monitoring results from groundwater wells associated with the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds Reuse Permit (I-161-02). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

One of the primary INL missions is to support development of advanced fuels with the goal of creating reactor fuels that produce less waste and are easier to store. The Advanced Fuel Cycle Initiative (AFCI) Glovebox in the Fuel Manufacturing Facility (FMF) is used for several fuel fabrication steps that involve transuranic elements, including americium. The AFCI glove box contains equipment used for fuel fabrication, including an arc melter – a small, laboratory-scale version of an electric arc furnace used to make new metal alloys for research – and an americium distillation apparatus. This overview summarizes key findings related to the investigation into the releases of airborne radioactivity that occurred in the AFCI glovebox room in late August and early September 2014. The full report (AFCI Glovebox Radiological Release – Evaluation, Corrective Actions and Testing, INL/INL-15-36996) provides details of the identified issues, corrective actions taken as well as lessons learned

The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

Source term information is required for to reconstruct a device used in a dispersed radiological dispersal device. Simulating a radioactive environment to train and exercise sampling and sample characterization methods with suitable sample materials is a continued challenge. The Idaho National Laboratory has developed and permitted a radioactive response training range (RRTR), an 800 acre test range that is approved for open air dispersal of activated KBr, for training first responders in the entry and exit from radioactively contaminated areas, and testing protocols for environmental sampling and field characterization. Members from the Department of Defense, Law Enforcement, and the Department of Energy participated in the first contamination exercise that was conducted at the RRTR in the July 2011. The range was contaminated using a short lived radioactive 82Br isotope (activated KBr). Soil samples contaminated with KBr (dispersed as a solution) and glass particles containing activated potassium bromide that emulated dispersed radioactive materials (such as ceramic-based sealed source materials) were collected to assess environmental sampling and characterization techniques. This presentation summarizes the performance of a radioactive materials surrogate for use as a training aide for nuclear forensics.

The Idaho National Laboratory prepared a preliminary technical and functional requirements (TF average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 °C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

As part of the Department of Energy’s Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

The amount of plutonium discarded as wastes from the U.S. nuclear weapons complex appears to be significantly greater than the U.S. Department of Energy's 1996 declaration of its plutonium holdings. This is due to in part to improved radioactive waste characterization and the disposal of plutonium residues originally intended for use in weapons. The Hanford site in Washington State has the largest quantity of plutonium wastes, which pose potentially serious human risks to ground water and the near shore the Columbia River. The department should revise its accounting for plutonium, and take steps to remove plutonium discarded to the environment at Hanford, as it is required to do at Idaho National Laboratory.

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: high temperature gas reactor fuels behaviour, high temperature materials qualification, design methods development and validation, hydrogen production technologies energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

The U.S. Department of Energy, in concert with industry, is developing a high-temperature gas-cooled reactor at the Idaho National Laboratory (INL) to demonstrate high temperature heat applications to produce hydrogen and electricity or to support other industrial applications. A key part of this program is the production of hydrogen from water that would significantly reduce carbon emissions compared to current production using natural gas. In 2009 the INL led the methodical evaluation of promising advanced hydrogen production technologies in order to focus future resources on the most viable processes. This paper describes how the evaluation process was systematically planned and executed. As a result, High-Temperature Steam Electrolysis was selected as the most viable near-term technology to deploy as a part of the Next Generation Nuclear Plant Project.

The U.S. Department of Energy (US DOE) remediation responsibilities include its Idaho National Laboratory. In 1989, the U.S. Environmental Protection Agency placed the Idaho site on its National Priority List for environmental cleanup. The site's contamination legacy from operations included inactive reactors and other structures, spent nuclear fuel, high-level liquid radioactive wastes, calcined radioactive wastes, and transuranic wastes. Documents governing cleanup include a 1995 Settlement Agreement between the US DOE and the US Navy as responsible parties, and the State of Idaho. The Subsurface Disposal Area contains buried transuranic wastes, lies above the East Snake River Plain Aquifer, and could be the “site's most nettlesome cleanup issue,” according to an outside observer. This article describes the technical and legal difficulties that have been encountered in remediating this area. © 2010 Wiley Periodicals, Inc.

A global collaboration is currently developing solvent extraction separations for the nuclear fuel cycle of the future. The goal is to recover fissionable material for recycle, mitigate proliferation concerns, and mitigate the environmental impact of nuclear waste disposal. Relying on selective metal complexing agents, the radiation stability of these solvent extraction ligands will determine the efficiency and recycle lifetime of any solvent intended for use in this high-radiation environment. This paper reviews work at the Idaho National Laboratory regarding the radiation chemistry of nuclear solvent extraction ligands, with particular emphasis on the reactions of nitrogen-centered radicals.

The Idaho National Laboratory (INL) has been at the forefront of developing methods to reduce the cost and schedule of deactivating spent fuel pools (SFP). Several pools have been deactivated at the INL using an underwater approach with divers. These projects provided a basis for the INL cooperation with the Dresden Nuclear Power Station Unit 1 SFP (Exelon Generation Company) deactivation. It represents the first time that a commercial nuclear power plant (NPP) SFP was decommissioned using this underwater coating process. This approach has advantages in many aspects, particularly in reducing airborne contamination and allowing safer, more cost effective deactivation. The INL pioneered underwater coating process was used to decommission three SFPs with a total combined pool volume of over 900,000 gallons. INL provided engineering support and shared project plans to successfully initiate the Dresden project. This report outlines the steps taken by INL and Exelon to decommission SFPs using the underwater coating process. The rationale used to select the underwater coating process and the advantages and disadvantages are described. Special circumstances are also discussed, such as the use of a remotely-operated underwater vehicle to visually and radiologically map the pool areas that were not readily accessible. A larger project, the INTEC-603 SFP in-situ (grouting) deactivation, is reviewed. Several specific areas where special equipment was employed are discussed and a Lessons Learned evaluation is included.

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

The long-term durability of WAXFIX, a paraffin-based grout, was evaluated for in situ grouting of activated beryllium wastes in the subsurface disposal area (SDA), a radioactive landfill at the Radioactive Waste Management Complex, part of the Idaho National Laboratory (INL). The evaluation considered radiological and biological mechanisms that could degrade the grout using data from an extensive literature search and previous tests of in situ grouting at the INL. Conservative radioactive doses for WAXFIX were calculated from the "hottest" (i.e., highest-activity) Advanced Test Reactor beryllium block in the SDA. These results indicate that WAXFIX would not experience extensive radiation damage for many hundreds of years. Calculation ofradiation-induced hydrogen generation in WAXFIX indicated that grout physical performance should not be reduced beyond the effects ofradiation dose on the molecular structure. Degradation of a paraffin-based grout by microorganisms in the SDA is possible and perhaps likely, but the rate of degradation will be at a slower rate than found in the literature reviewed. The calculations showed the outer 0.46-m (18-in.) layer of each monolith, which represents the minimum expected distance to the beryllium block, was calculated to require 1000 to 3600 yr to be consumed. The existing data and estimations of biodegradation and radiolysis rates for WAXFIX/paraffin do not indicate any immediate problems with the use of WAXFIX for grouting beryllium or other wastes in the SDA.