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Justification to Disregard U-234 Mass in Certain Uranium Systems

The ANSI/ANS-8.1-1998 Standard regarding nuclear criticality safety in operations with fissionable material outside reactors states certain guidelines for dealing with uranium isotopes when setting mass limits. In ANSI/ANS-8.1-1998 235 U limits apply to mixtures “with 234 U, 236 U, or 238 U, provided that 234 U is considered to be 235U” in computing mass. In this evaluation calculations coupled with benchmark data show a superior way of handling the other common uranium isotopes when setting limits for 235 U. Currently 234 U mass must be considered as 235 U mass in order to account for the fact 234 U has a k ∞ = 1.70 meaning criticality can be reached with a finite amount of 234 U mass. 236 U and 238 U have k ∞ ’s equal to 0.71 and 0.31 meaning they remain subcritical in infinite amounts and thus no mass limits are required. In fresh uranium (un-irradiated or reprocessed) enriched to 94% 235 U by weight, bounding ratios of 234 U, 236U, and 238U exist due to the enrichment process. Utilizing these ratios it is shown that 234 U accounts for less than 0.20 of the total mass fraction of 234 U, 236 U, and 238 U and the remaining 0.80 mass fraction is in a ratio of at least 1:1 for 238 U: 236 U. Creating models conservatively based on these two material limits, the resulting k ∞ = 0.91367 ± 0.00208 suggests a mass limit for 234 U is no longer required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Three-Dimensional Geologic Characterization of Geothermal Systems: Astor Pass, Nevada, USA

Geothermal systems in the Great Basin, USA, are controlled by a variety of fault intersection and fault interaction areas. Understanding the specific geometry of the structures most conducive to geothermal circulation is crucial in order to both mitigate the costs of geothermal exploration (especially drilling) and to identify blind geothermal systems (no surface expression). Astor Pass, Nevada, one such blind geothermal system, lies near the boundary between two distinct structural domains, the Walker Lane and the Basin and Range, and exhibits characteristics of each setting. Both northwest-striking, left-stepping dextral faults of the Walker Lane and kinematically linked northerly striking normal faults associated with the Basin and Range are present at Astor Pass. Previous studies identified a blind geothermal system controlled by the intersection of northwest-striking dextral and north-northwest-striking normal faults. Wells drilled into the southwestern quadrant of the fault intersection yielded 94°C fluids, with geothermometers suggesting significantly higher maximum temperatures. Additional data, including reprocessed 2D seismic data and petrologic analysis of well cuttings, were integrated with existing and reinterpreted geologic maps and cross-sections to aid construction of a 3D geologic model. This comprehensive 3D integration of multiple data sets allows characterization of the structural setting of the Astor Pass blind geothermal system at a level of detail beyond what independent data interpretation can provide. Our analysis indicates that the blind geothermal system is controlled by two north- to northwest-plunging fault intersections.

15 GEOTHERMAL ENERGY↗

Off-gas Adsorption Model and Simulation - OSPREY

The absence of industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other countries. Thus, it is essential to model complex series of unit operations to simulate, understand, and predict inherent transient behavior. A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and REcoverY (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DISSOLUTION OF METAL OXIDES AND SEPARATION OF URANIUM FROM LANTHANIDES AND ACTINIDES IN SUPERCRITICAL CARBON DIOXIDE

This paper investigates the feasibility of extracting and separating uranium from lanthanides and other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of a counter current stripping technique, which would be a more efficient and environmentally benign technology for spent nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U, Pu, and Np) and europium were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, uranium/europium and uranium/plutonium extraction and separation in sc-CO2 modified with TBP is successful at nitric acid concentrations of less than 6 M and at nitric acid concentrations of less than 3 M with acetohydroxamic acid or oxalic acid, respectively. A scheme for recycling uranium from spent nuclear fuel by using sc-CO2 and counter current stripping columns is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Electrochemically modulated separations for material accountability measurements

A method for the accurate and timely analysis of accountable materials is critical for safeguards measurements in nuclear fuel reprocessing plants. Non-destructive analysis (NDA) methods, such as gamma spectroscopy, are desirable for their ability to produce near real-time data. However, the high gamma background of the actinides and fission products in spent nuclear fuel limits the use of NDA for real-time online measurements. A simple approach for at-line separation of materials would facilitate the use of at-line detection methods. A promising at-line separation method for plutonium and uranium is electrochemically modulated separations (EMS). Using an electrochemical cell with an anodized glassy carbon electrode, Pu and U oxidation states can be altered by applying an appropriate voltage. Because the affinity of the actinides for the electrode depends on their oxidation states, selective deposition can be turned “on” and “off” with changes in the applied target electrode voltage. A high surface-area cell was designed in house for the separation of Pu from spent nuclear fuel. The cell is shown to capture over 1 µg of material, increasing the likelihood for gamma spectroscopic detection of Pu extracted from dissolver solutions. The large surface area of the electrode also reduces the impact of competitive interferences from some fission products. Flow rates of up to 1 mL min-1 with >50% analyte deposition efficiency are possible, allowing for rapid separations to be effected. Results from the increased surface-area EMS cell are presented, including dilute dissolver solution simulant data.

36 MATERIALS SCIENCE↗

Evaluation of Static Mixer Flow Enhancements for Cryogenic Viscous Compressor Prototype for ITER Vacuum System

As part of the U.S. ITER contribution to the vacuum systems for the ITER fusion project, a cryogenic viscous compressor (CVC) is being designed and fabricated to cryopump hydrogenic gases in the torus and neutral beam exhaust streams and to regenerate the collected gases to controlled pressures such that they can be mechanically pumped with controlled flows to the tritium reprocessing facility. One critical element of the CVC design that required additional investigation was the determination of flow rates of the low pressure (50 to 1000 Pa) exhaust stream that would allow for complete pumping of hydrogenic gases while permitting trace levels of helium to pass through the CVC to be pumped by conventional vacuum pumps. A sub-scale prototype test facility was utilized to determine the effectiveness of a static mixer pump tube concept, which consisted of a series of rotated twisted elements brazed into a 2-mm thick, 5-cm diameter stainless steel tube. Cold helium gas flow provided by a dewar and helium transfer line was used to cool the exterior of the static mixer pump tube. Deuterium gas was mixed with helium gas through flow controllers at different concentrations while the composition of the exhaust gas was monitored with a Penning gauge and optical spectrometer to determine the effectiveness of the static mixer. It was found that with tube wall temperatures between 6 K and 9 K, the deuterium gas was completely cryopumped and only helium passed through the tube. These results have been used to design the cooling geometry and the static mixer pump tubes in the full-scale CVC prototype

Duckworth, Robert C↗

Overview of reductants utilized in nuclear fuel reprocessing/recycling

Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold promises as a replacement for AHA. FHA undergoes hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. Unfortunately, FHA powder was not stable in the experiments we ran in our laboratory. In addition, AHA and FHA also decompose to hydroxylamine which may undergo an autocatalytic reaction. Other reductants are available and could be extremely useful for actinides separation. The review presents the current plutonium reductants used in used nuclear fuel reprocessing and will introduce innovative and novel reductants that could become reducers for future research on UNF separation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Electrochemically Modulated Separation for Plutonium Safeguards

Accurate and timely analysis of plutonium in spent nuclear fuel is critical in nuclear safeguards for detection of both protracted and rapid plutonium diversions. Gamma spectroscopy is a viable method for accurate and timely measurements of plutonium provided that the plutonium is well separated from the interfering fission and activation products present in spent nuclear fuel. Electrochemically modulated separation (EMS) is a method that has been used successfully to isolate picogram amounts of Pu from nitric acid matrices. With EMS, Pu adsorption may be turned "on" and "off" depending on the applied voltage, allowing for collection and stripping of Pu without the addition of chemical reagents. In this work, we have scaled up the EMS process to isolate microgram quantities of Pu from matrices encountered in spent nuclear fuel during reprocessing. Several challenges have been addressed including surface area limitations, radiolysis effects, electrochemical cell performance stability, and chemical interferences. After these challenges were resolved, 6 µg Pu was deposited in the electrochemical cell with approximately an 800-fold reduction of fission and activation product levels from a spent nuclear fuel sample. Modeling showed that these levels of Pu collection and interference reduction may not be sufficient for Pu detection by gamma spectroscopy. The main remaining challenges are to achieve a more complete Pu isolation and to deposit larger quantities of Pu for successful gamma analysis of Pu. If gamma analyses of Pu are successful, EMS will allow for accurate and timely on-site analysis for enhanced Pu safeguards.

Pratt, Sandra H.↗

Supercritical Fluid Extraction and Separation of Uranium from Other Actinides

This paper investigates the feasibility of separating uranium from other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of an extraction and counter current stripping technique, which would be a more efficient and environmentally benign technology for used nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U(VI), Np(VI), Pu(IV), and Am(III)) were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, the separation of uranium from plutonium in sc-CO2 modified with TBP was successful at nitric acid concentrations of less than 3 M in the presence of acetohydroxamic acid or oxalic acid, and the separation of uranium from neptunium was successful at nitric acid concentrations of less than 1 M in the presence of acetohydroxamic acid, oxalic acid, or sodium nitrite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluating Safeguards Benefits of Process Monitoring as compared with Nuclear Material Accountancy

This paper illustrates potential safeguards benefits that process monitoring (PM) may have as a diversion deterrent and as a complementary safeguards measure to nuclear material accountancy (NMA). This benefit is illustrated by quantifying the standard deviation associated with detecting a considered material diversion scenario using either an NMA-based method or a PM-based approach. To illustrate the benefits of PM for effective safeguards, we consider a reprocessing facility. We assume that the diversion of interest for detection manifests itself as a loss of Pu caused by abnormally operating a dissolver for an extended period to accomplish protracted diversion (or misdirection) of Pu to a retained (unconditioned) waste stream. For detecting the occurrence of this diversion (which involves anomalous operation of the dissolver), we consider two different data evaluation and integration (DEI) approaches, one based on NMA and the other based on PM. The approach based on PM does not directly do mass balance calculations, but rather monitors for the possible occurrence of anomaly patterns related to potential loss of nuclear material. It is thus assumed that the loss of a given mass amount of nuclear material can be directly associated with the execution of proliferation-driven activities that trigger the occurrence of an anomaly pattern consisting of series of events or signatures occurring at different unit operations and time instances. By effectively assessing these events over time and space, the PM-based DEI approach tries to infer whether this specific pattern of events has occurred and how many times within a given time period. To evaluate the goodness of PM, the 3 Sigma of the estimated mass loss is computed under both DEI approaches as function of the number of input batches processed. Simulation results are discussed.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Toward a Fieldable Atomic Mass Spectrometer for Safeguards Applications: Sample Preparation and Ionization

The International Atomic Energy Agency’s (IAEA’s) long-term research and development plan calls for the development of new methods to detect misuse at nuclear fuel cycle facilities such as reprocessing and enrichment plants. At enrichment plants, for example, the IAEA’s contemporary safeguards approaches are based on a combination of routine and random inspections that include collection of UF6 samples from in-process material and selected cylinders for subsequent analyses. These analyses include destructive analysis (DA) in a laboratory (typically by mass spectrometry [MS]) for isotopic characterization, and environmental sampling (ES) for subsequent laboratory elemental and isotopic analysis (also both typically by MS). One area of new method development includes moving this kind of isotope ratio analytical capability for DA and ES activities into the field. Some of the reasons for these developments include timeliness of results, avoidance of hazardous material shipments, and guidance for additional sample collecting. However, this capability does not already exist for several reasons, such as that most lab-based chemical and instrumental methods rely on laboratory infrastructure (highly trained staff, power, space, hazardous material handling, etc.) and require significant amounts of consumables (power, compressed gases, etc.). In addition, there are no currently available, fieldable instruments for atomic or isotope ratio analysis. To address these issues, Pacific Northwest National Laboratory (PNNL) and collaborator, Clemson University, are studying key areas that limit the fieldability of isotope ratio mass spectrometry for atomic ions: sample preparation and ionization, and reducing the physical size of a fieldable mass spectrometer. PNNL is seeking simple and robust techniques that could be effectively used by inspectors who may have no expertise in analytical MS. In this report, we present and describe the preliminary findings for three candidate techniques: matrix-assisted laser desorption/ionization (MALDI) MS, liquid sampling-atmospheric pressure glow discharge (LS-APGD), and laser ablation/ionization (LAI) MS at atmospheric pressure. Potential performance metrics for these techniques will be presented, including detectability, response, isotope ratio accuracy and precision, and ease of use.

Barinaga, Charles J.↗

Eddy Current for Sizing Cracks in Canisters for Dry Storage of Used Nuclear Fuel

The storage of used nuclear fuel (UNF) in dry canister storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSI) sites is a temporary measure to accommodate UNF inventory until it can be reprocessed or transferred to a repository for permanent disposal. Policy uncertainty surrounding the long-term management of UNF indicates that DCSSs will need to store UNF for much longer periods than originally envisioned. Meanwhile, the structural and leak-tight integrity of DCSSs must not be compromised. The eddy current technique is presented as a potential tool for inspecting the outer surfaces of DCSS canisters for degradation, particularly atmospheric stress corrosion cracking (SCC). Results are presented that demonstrate that eddy current can detect flaws that cannot be detected reliably using standard visual techniques. In addition, simulations are performed to explore the best parameters of a pancake coil probe for sizing of SCC flaws in DCSS canisters and to identify features in frequency sweep curves that may potentially be useful for facilitating accurate depth sizing of atmospheric SCC flaws from eddy current measurements.

dry storage↗

Climate-change effects on soils: Accelerated weathering, soil carbon and elemental cycling

Climate change [i.e., high atmospheric carbon dioxide (CO2) concentrations (≥400 ppm); increasing air temperatures (2-4°C or greater); significant and/or abrupt changes in daily, seasonal, and inter-annual temperature; changes in the wet/dry cycles; intensive rainfall and/or heavy storms; extended periods of drought; extreme frost; heat waves and increased fire frequency] is and will significantly affect soil properties and fertility, water resources, food quantity and quality, and environmental quality. Biotic processes that consume atmospheric CO2, and create organic carbon (C) that is either reprocessed to CO2 or stored in soils are the subject of active current investigations, with great concern over the influence of climate change. In addition, abiotic C cycling and its influence on the inorganic C pool in soils is a fundamental global process in which acidic atmospheric CO2 participates in the weathering of carbonate and silicate minerals, ultimately delivering bicarbonate and Ca2+ or other cations that precipitate in the form of carbonates in soils or are transported to the rivers, lakes, and oceans. Soil responses to climate change will be complex, and there are many uncertainties and unresolved issues. The objective of the review is to initiate and further stimulate a discussion about some important and challenging aspects of climate-change effects on soils, such as accelerated weathering of soil minerals and resulting C and elemental fluxes in and out of soils, soil/geo-engineering methods used to increase C sequestration in soils, soil organic matter (SOM) protection, transformation and mineralization, and SOM temperature sensitivity. This review reports recent discoveries, identifies key research needs, and highlights opportunities offered by the climate-change effects on soils.

accelerated weathering↗

Reactive Transport of the Uranyl Ion in Soils, Sediments, and Groundwater Systems

Uranium is a ubiquitous trace component in rocks ranging from 1.2 to 1.3 µg g-1 in sedimentary rocks, 2.2 to 15 µg g-1 in granites, and 20 to 120 µg g-1 in phosphates (Langmuir, 1997; Plant et al., 1999). Uranium (U) is released to natural waters in dilute concentrations (generally < 10-7 mole L-1) from the weathering of these sources, with water concentrations in uraniferous geologic terrains, such as the southwestern U.S. (USGS, 2011), being higher (~ 10-6.5 mol L-1). Elevated water-borne concentrations are associated with the weathering of natural ore bodies [~10-6 mol L-1; e.g, (Payne and Airey, 2006)], the extraction and mining of U for armaments (Jiang and Aschner, 2009; WHO, 2001) and nuclear fuels [10-6 to 10-3 mol L-1; (Abdelouas et al., 1999)], and the disposal of waste solids and liquids from nuclear fuels reprocessing and arms production [~ 10-6 to 10-2 mol L-1; e.g., (Wan et al., 2009; Zachara et al., 2007)]. The form of U present in natural waters at high concentration is generally the uranyl ion [e.g., UO22+] which is quite soluble. Groundwater in many parts of the world contains dissolved U originating from natural and anthropogenic sources (ATSDR, 2011; EFSA, 2009). Low levels of dissolved U in drinking water are considered a health concern, causing renal and other effects (Kurttio et al., 2002; Kurttio et al., 2005; Limson Zamora et al., 1998; Nriagu et al., 2012; Raymond-Whish et al., 2007; Selden et al., 2009). The U.S. Environmental Protection Agency has established a regulatory drinking water standard of 30 µg L-1 (1.26 x 10-7 mol L-1) or 30 pCi L-1, whichever is exceeded first. The World Health Organization has recommended an even lower drinking water standard of 15 µg L-1 [6.3 x 10-8 mol L-1; (WHO, 2005)]. Human exposure to U through drinking water is expected to rise as world-wide reliance on groundwater sources increase (ESS, 2010).

Environmental Molecular Sciences Laboratory↗

Development of the Actinide-Lanthanide Separation (ALSEP) Process

Separating the minor actinide elements (Am and Cm) from acidic high-level raffinates arising from the reprocessing of irradiated nuclear fuel is an important step in closing the nuclear fuel cycle. Most proposed approaches to this problem involve two solvent extraction steps: 1) co-extraction of the trivalent lanthanides and actinides, followed by 2) separation of the actinides from the lanthanides. The objective of our work is to develop a single solvent-extraction process for isolating the minor actinide elements. We report here a solvent containing N,N,N',N'-tetra(2 ethylhexyl)diglycolamide (T2EHDGA) combined with 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) that can be used to separate the minor actinides in a single solvent-extraction process. T2EHDGA serves to co-extract the trivalent actinide and lanthanide ions from nitric acid solution. Switching the aqueous phase chemistry to a citrate buffered solution of N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid at pH 2.5 to 4 results in selective transfer of the actinides to the aqueous phase, thus affecting separation of the actinides from the lanthanides. Separation factors between the lanthanides and actinides are approximately 20 in the pH range of 3 to 4, and the distribution ratios are not highly dependent on the pH in this system.

actinide separation↗

Persistent Monitoring of Acyclic Low-Level Chemical Effluents in Restricted Areas

Detection of acyclic, transitory low-level chemical releases in restricted areas is not possible for remote sensing. A prototype for near-field stealthy persistent monitoring was developed, including a self-organizing semantic sensor field self-management capability. The capability was focused on nuclear conversion and reprocessing, but the platform and software is general.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

openPMD-viewer

Many simulation software produce data in the form of a set of field values or of a set of particle positions. (one such example is that of particle-in-cell codes, which produce data on the electromagnetic fields that they simulate.) However, each particular software uses its own particular format and layout, for the output data. This makes it difficult to compare the results of different simulation software, or to have a common visualization tool for these results. However, a standardized layout for fields and particles has recently been developed: the openPMD format ( HYPERLINK "http://www.openpmd.org/"www.openpmd.org) This format is open- source, and specifies a standard way in which field data and particle data should be written. The openPMD format is already implemented in the particle-in-cell code Warp (developed at LBL) and in PIConGPU (developed at HZDR, Germany). In this context, the proposed software (openPMD-viewer) is a Python package, which allows to access and visualize any data which has been formatted according to the openPMD standard. This package contains two main components: - a Python API, which allows to read and extract the data from a openPMD file, so as to be able to work with it within the Python environment. (e.g. plot the data and reprocess it with particular Python functions) - a graphical interface, which works with the ipython notebook, and allows to quickly visualize the data and browse through a set of openPMD files. The proposed software will be typically used when analyzing the results of numerical simulations. It will be useful to quickly extract scientific meaning from a set of numerical data.

Lehe, Remi↗

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

Beatty, Randy L.↗