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At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Rescuing a Treasure Uranium-233

Uranium-233 (233U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium (232Th). At high purities, this synthetic isotope serves as a crucial reference for accurately quantifying and characterizing natural uranium isotopes for domestic and international safeguards. Separated 233U is stored in vaults at Oak Ridge National Laboratory. These materials represent a broad spectrum of 233U from the standpoint isotopic purity the purest being crucial for precise analyses in safeguarding uranium. All 233U at ORNL currently is scheduled to be down blended with depleted uranium beginning in 2015. Such down blending will permanently destroy the potential value of pure 233U samples as certified reference material for use in uranium analyses. Furthermore, no replacement 233U stocks are expected to be produced in the future due to a lack of operating production capability and the high cost of returning to operation this currently shut down capability. This paper will describe the efforts to rescue the purest of the 233U materials arguably national treasures from their destruction by down blending.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Utilization of accelerators for transmutation and energy production

Given the increased concern over reliable, emission-free power, nuclear power has experienced a resurgence of interest. A sub-critical accelerator driven system (ADS) can drive systems that have either safety constraints (waste transmutation) or reduced fissile content (thorium reactor). The goals of ADS are some or all of the following: (1) to significantly reduce the generation or impacts due to the minor actinides on the packing density and long-term radiotoxicity in the repository design, (2) preserve/use the energy-rich component of used nuclear fuel, and (3) reduce proliferation risk. ADS systems have been actively studied in Europe and Asia over the past two decades and renewed interest is occurring in the U.S. This talk will cover some of the history, possible applicable fuel cycle scenarios, and general issues to be considered in implementing ADS systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of Waste Arising from Future Nuclear Fuel Cycle

A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Actinide Solubility and Speciation in the WIPP [PowerPoint]

The presentation begins with the role and need for nuclear repositories (overall concept, international updates (Sweden, Finland, France, China), US approach and current status), then moves on to the WIPP TRU repository concept (design, current status--safety incidents of February 5 and 14, 2014, path forward), and finally considers the WIPP safety case: dissolved actinide concentrations (overall approach, oxidation state distribution and redox control, solubility of actinides, colloidal contribution and microbial effects). The following conclusions are set forth: (1) International programs are moving forward, but at a very slow and somewhat sporadic pace. (2) In the United States, the Salt repository concept, from the perspective of the long-term safety case, remains a viable option for nuclear waste management despite the current operational issues/concerns. (3) Current model/PA prediction (WIPP example) are built on redundant conservatisms. These conservatisms are being addressed in the ongoing and future research to fill existing data gaps--redox control of plutonium by Fe(0, II), thorium (analog) solubility studies in simulated brine, contribution of intrinsic and biocolloids to the mobile concentration, and clarification of microbial ecology and effects.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Neutronic analysis of a fusion hybrid reactor

In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Gaseous fission product management for molten salt reactors and vented fuel systems

Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Hydrogeologic Modeling at the Sylvania Corning FUSRAP Site - 13419

A comparative groundwater hydrogeologic modeling analysis is presented herein to simulate potential contaminant migration pathways in a sole source aquifer in Nassau County, Long Island, New York. The source of contamination is related to historical operations at the Sylvania Corning Plant ('Site'), a 9.49- acre facility located at 70, 100 and 140 Cantiague Rock Road, Town of Oyster Bay in the westernmost portion of Hicksville, Long Island. The Site had historically been utilized as a nuclear materials manufacturing facility (e.g., cores, slug, and fuel elements) for reactors used in both research and electric power generation in early 1950's until late 1960's. The Site is contaminated with various volatile organic and inorganic compounds, as well as radionuclides. The major contaminants of concern at the Site are tetrachloroethene (PCE), trichloroethene (TCE), nickel, uranium, and thorium. These compounds are present in soil and groundwater underlying the Site and have migrated off-site. The Site is currently being investigated as part of the Formerly Utilized Sites Remedial Action Program (FUSRAP). The main objective of the current study is to simulate the complex hydrogeologic features in the region, such as numerous current and historic production well fields; large, localized recharge basins; and, multiple aquifers, and to assess potential contaminant migration pathways originating from the Site. For this purpose, the focus of attention was given to the underlying Magothy formation, which has been impacted by the contaminants of concern. This aquifer provides more than 90% of potable water supply in the region. Nassau and Suffolk Counties jointly developed a three-dimensional regional groundwater flow model to help understand the factors affecting groundwater flow regime in the region, to determine adequate water supply for public consumption, to investigate salt water intrusion in localized areas, to evaluate the impacts of regional pumping activity, and to better understand the contaminant transport and fate mechanisms through the underlying aquifers. This regional model, developed for the N.Y. State Department of Environmental Conservation (NYSDEC) by Camp Dresser and McKee (CDM), uses the finite element model DYNFLOW developed by CDM, Cambridge, Massachusetts. The coarseness of the regional model, however, could not adequately capture the hydrogeologic heterogeneity of the aquifer. Specifically, the regional model did not adequately capture the interbedded nature of the Magothy aquifer and, as such, simulated particles tended to track down-gradient from the Site in relatively straight lines while the movement of groundwater in such a heterogeneous aquifer is expected to proceed along a more tortuous path. This paper presents a qualitative comparison of site-specific groundwater flow modeling results with results obtained from the regional model. In order to assess the potential contaminant migration pathways, a particle tracking method was employed. Available site-specific and regional hydraulic conductivity data measured in-situ with respect to depth and location were incorporated into the T-PROG module in GMS model to define statistical variation to better represent the actual stratigraphy and layer heterogeneity. The groundwater flow characteristics in the Magothy aquifer were simulated with the stochastic hydraulic conductivity variation as opposed to constant values as employed in the regional model. Contaminant sources and their exact locations have been fully delineated at the Site during the Remedial Investigation (RI) phase of the project. Contaminant migration pathways originating from these source locations at the Site are qualitatively traced within the sole source aquifer utilizing particles introduced at source locations. Contaminant transport mechanism modeled in the current study is based on pure advection (i.e., plug flow) and mechanical dispersion while molecular diffusion effects are neglected due to relatively high groundwater velocities encountered in the aquifer. In addition, fate of contaminants is ignored hereby to simulate the worst-case scenario, which considers the contaminants of concern as tracer-like compounds for modeling purposes. The results of the modeling analysis are qualitatively compared with the County's regional model, and patterns of contaminant migration in the region are presented. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Reducing Contingency through Sampling at the Luckey FUSRAP Site - 13186

Typically, the greatest risk in developing accurate cost estimates for the remediation of hazardous, toxic, and radioactive waste sites is the uncertainty in the estimated volume of contaminated media requiring remediation. Efforts to address this risk in the remediation cost estimate can result in large cost contingencies that are often considered unacceptable when budgeting for site cleanups. Such was the case for the Luckey Formerly Utilized Sites Remedial Action Program (FUSRAP) site near Luckey, Ohio, which had significant uncertainty surrounding the estimated volume of site soils contaminated with radium, uranium, thorium, beryllium, and lead. Funding provided by the American Recovery and Reinvestment Act (ARRA) allowed the U.S. Army Corps of Engineers (USACE) to conduct additional environmental sampling and analysis at the Luckey Site between November 2009 and April 2010, with the objective to further delineate the horizontal and vertical extent of contaminated soils in order to reduce the uncertainty in the soil volume estimate. Investigative work included radiological, geophysical, and topographic field surveys, subsurface borings, and soil sampling. Results from the investigative sampling were used in conjunction with Argonne National Laboratory's Bayesian Approaches for Adaptive Spatial Sampling (BAASS) software to update the contaminated soil volume estimate for the site. This updated volume estimate was then used to update the project cost-to-complete estimate using the USACE Cost and Schedule Risk Analysis process, which develops cost contingencies based on project risks. An investment of $1.1 M of ARRA funds for additional investigative work resulted in a reduction of 135,000 in-situ cubic meters (177,000 in-situ cubic yards) in the estimated base volume estimate. This refinement of the estimated soil volume resulted in a $64.3 M reduction in the estimated project cost-to-complete, through a reduction in the uncertainty in the contaminated soil volume estimate and the associated contingency costs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Lessons Learned from a Complex FUSRAP Site - Sylvania Corning FUSRAP Site - 12269

Since its addition to the Formerly Utilized Sites Remedial Action Program (FUSRAP) in 2005, the Sylvania Corning FUSRAP Site (the Site) in Hicksville, New York, has provided challenges and opportunities from which to gain lessons learned for conducting investigation work at a complex multi-contaminant FUSRAP Site. The United States Army Corps of Engineers (USACE) and its contractors conducted a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Remedial Investigation (RI) and are currently in the Feasibility Study (FS) phase at the Site. This paper presents the planning, execution, and reporting lessons learned by USACE during the RI/FS. The Site, operated from 1952 to 1967 for the research, development, and fabrication of nuclear elements under the Atomic Energy Commission, and other government and commercial contracts. Previous investigations performed by the New York State Department of Environmental Conservation (NYSDEC) and the current property owner have identified uranium, thorium, nickel, and chlorinated solvents, as Site contaminants [1]. The property owner is currently under two separate voluntary agreements with NYSDEC to investigate and remediate the Site. USACE's work at the site has been independent of this voluntary agreement and has moved on a parallel path with any work the property owner has completed. The project at the Site is complex because of the radiological and chemical concerns in both soils and groundwater, high hydraulically conductive soils, lack of a shallow aquiclude/aquitard, and a principal water table aquifer underlying the site. Contaminants are migrating from the Site and may potentially impact local drinking water supplies (municipal wells). During the RI/FS process the project team has encountered many issues and has thus developed many resolutions. The issues are organized into three categories: Planning and Contracting, Execution, and Reporting. Planning and Contracting lessons learned include: how to incorporate an overwhelming volume of historical data; how to manage a complex team of three prime contractors innovatively, and how to implement a project under an Award Fee task order. Execution lessons learned include: characterization of investigation derived wastes, and proper approach to radiological scanning of direct-push borings and soil cores. Reporting lessons learned include: coordinating multiple phase (iterative) reporting, large dataset presentation, and the National Priorities List (NPL) designation. The goal of this paper is to provide a resource for other project delivery teams that encounter similar situations on their projects to optimize cost savings, realization of efficiency, shorten schedules, or simply ensure higher quality deliverables. Each FUSRAP project is unique but there are many lessons we can apply to each site to gain efficiency and work more effectively. The Sylvania Corning FUSRAP site is a complex site with both soils and groundwater contamination, contamination to depths of 182 meters, and a highly politically charged environment of PRP involvement. Many of the lessons the project team has learned during the life of the project to date are being shared with others as well as being applied back to this project for future work. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Solving Complex Radioactive Decay Chains for Future Assessment and Cleanup Decisions - 17046

In support of U.S. Environmental Protection Agency's on-line calculators for risk and dose assessment, a decay chain tool was developed to address the need to understand how radionuclide activity changes with time as the activity measured in the past will be different from current and future levels. When a radionuclide decays, its activity decreases exponentially as a function of time transforming into a different atom - a decay product. The atoms keep transforming to new decay products until they reach a stable state and are no longer radioactive. The series of decay products created to reach this balance is called the decay chain. For radionuclide chains, the daughter products can have significant implications in dosimetry and remediation. Thus, risk assessors evaluating sites with radioactive contamination need to plan for future progeny ingrowth, in addition to sampled radionuclides. These are important considerations for risk quantification during the characterization and cleanup plans, particularly when sampling may have occurred years before the remediation cleanup work begins. If a radionuclide half-life and current activity are known, then hand-calculating the future activity is straightforward. However, calculating the ingrowth of progeny quickly becomes cumbersome for longer chains such as the Thorium-232 decay series. For the more complex chains where many daughters are formed, possibly with multiple branches, this calculation involves solving a complex set of simultaneous differential equations known as the Bateman Equation. The Decay Chain Activity Projection Tool calculates the activity of radionuclides and their progeny as a function of time. This tool currently uses a combination of Perl, JavaScript, Plot.ly, and jQuery to automatically construct the radionuclide decay chains, solve the resulting Bateman Equation, and provide the user with tabular solution output and plots. Additional software may be used as development advances. The risk assessor can then use the data for exposure assessment and cleanup decisions without further costly sampling. The benefits of providing this calculator are to prevent overestimation of risk due to decay of the parent isotope and to avoid underestimation of risk due to ingrowth of progeny more toxic than the parent. Graphs are provided to show the simultaneous decay rates and ingrowth, as well as tabular output at time-point T and all the decay half-life intervals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗