Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Fuel Reprocessing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 541 records · Page 30

Delayed Gamma-ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

This project addresses the need for improved non-destructive assay techniques for quantifying the actinide composition of spent nuclear fuel and for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle. High-energy delayed gamma-ray spectroscopy following neutron irradiation is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Chloride-Based Volatility for Waste Reduction and/or Reuse of Metallic-, Oxide- and Salt-Based Reactor Fuels

The objective of the chloride based volatility project (CBV) was to demonstrate the ability to separate uranium from used fuel to enable process improvements resulting in 10x reduction of waste volume, while maintaining safeguards standards and global backend costs at $\$$1/MW-hr. Current industrial practices perform separations of used fuel using solvent media in the form of aqueous and molten salt processes, resulting in contaminated process waste. The CBV approach utilized solid state chemistry with no solvent media and was successful in chlorinating uranium and fission product oxides that sublimed into the gas phase and were collected in targeted condensation zones based on temperature gradients of chlorinated products. Recovery of better than 95% of initial uranium in the form of UCl4 was demonstrated when simulated used nuclear fuel was used. Laser induced breakdown spectroscopy, LIBS, and ultraviolet-visible spectroscopy, UV-Vis, were combined into a high temperature flow cell design and utilized as process monitoring techniques to observed chlorinated products leave the reaction vessel in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Safeguards Challenges for Molten Salt Reactors

There are now a wide variety of molten salt reactor (MSR) technologies being proposed or developed that are gaining international interest and momentum toward deployment. There are two sub-categories of this class of reactors - those that are liquid fueled (i.e., the molten salt is both the fuel and the coolant) and those that have solid fuel with the molten salt used as coolant only. The liquid-fueled MSRs present unique challenges to current safeguards approaches. These challenges exist because there are currently no conceptual approaches to applying safeguards to unique, tightly coupled nuclear energy systems with the reactor core, balance of plant, and fuel cycle (everything else including salt processing and material separations) combined in a single facility. The application of safeguards to liquid-fueled MSRs will have to take into consideration several technical factors including: The homogeneous mixture of fuel, coolant, fission products, and actinides (with the attendant extremely high radiation field); Continuous variation in isotopic concentrations in the fuel salt, including removal (passive or active) of fission products, rare earth elements, and noble metals; High operating temperatures of the fuel salt; Online processing where some fraction of the inventory can be removed while the reactor is operational; Unique refueling schemes, including the ability to continuously feed the core with fresh fissile or fertile material; The presence of frozen fuel, potentially requiring a different safeguards process from the liquid fuel; and The presence of fuel outside the vessel. Additionally, If the thorium fuel cycle is employed, there will be additional complications since the resulting radiation signatures will be different from those of the uranium-based fuel cycle. The existing IAEA inspection regimes are based on the uranium-plutonium fuel cycle, item counting for nuclear reactors, and bulk material accountancy for the front and back end of the nuclear fuel cycle. These techniques and associated instrumentation for bulk accountancy have been developed predominantly for enrichment, fuel fabrication, and aqueous reprocessing. However, none of these bulk accountancy measures can be directly applied to liquid-fueled MSRs. This paper will explore some of the possible challenges in applying safeguards to molten salt reactors.

Kovacic, Donald N↗

Gas-phase stability of large lanthanide:diglycolamide clusters evaluated using collision-induced dissociation

Efficient utilization of long-term deep geologic storage repositories is critical for the large-scale deployment of nuclear energy. As the minor actinides (americium and curium) are the largest contributor to the decay heat of used nuclear fuel after a few hundred years, their removal from used fuel prior to disposal can significantly increase the amount of waste that can be stored in a given volume. The development of ligands that can effectively separate the minor actinides from other components of used nuclear fuel are crucial to efficient utilization of geological storage repositories. N,N,N’,N’-tetraoctyl diglycolamide (TODGA) is a promising extractant for the separation of lanthanides and the minor actinides from other components of used nuclear fuel. While the use of TODGA in f-element separations has been investigated in process-based formulations, gas-phase metal ion cluster experiments enable the study of covalent interactions in reprocessing systems absent from solvent effects. Exploring fundamental differences in lanthanide-ligand covalent interactions can impact the development and implementation of actinide-lanthanide separation systems in nuclear fuels reprocessing. In this study, lanthanide-TODGA clusters were synthesized in the gas-phase and identified using mass spectrometry fragmentation experiments. Large europium and samarium clusters were identified that contained up to 8 and 10 bound TODGA ligands, respectively; this was surprising due to the size and multidentate binding that is normal for TODGA. In addition, while smaller clusters showed evidence of sequential ligand fragmentation, larger clusters displayed the loss of neutral TODGA with applied collision voltage. Interestingly, the voltage required for this removal decreased as more TODGA ligands were bound to the metal, suggesting that the metal coordination sphere was becoming more saturated and TODGA ligands were more weakly bound as the clusters got larger.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Glass structure and crystallization in boro-alumino-silicate glasses containing rare earth and transition metal cations: a US-UK collaborative program

Nuclear wastes generated from reprocessing of used nuclear fuel tend to contain a large fraction of rare earth (RE, e.g., Nd3+), transition (TM, e.g., Mo6+, Zr4+), alkali (A, e.g., Cs+), and alkaline earth cations (AE, e.g., Ba2+, Sr2+). Various strategies have been considered for immobilizing such waste streams, varying from nominally crystal-free glass to glass-ceramic to multi-phase ceramic waste forms. For glass and glass-ceramic waste forms, the added glass-forming system is generally alkali-alkaline earth-aluminoborosilicate (i.e., Na-Ca-Al-B-Si oxide). In a US-UK collaborative project, summarized here, we investigated the glass structure and crystallization dependence on compositional changes in simulated nuclear waste glasses and glass-ceramics. Compositions ranged in complexity from five – to – eight oxides. Specifically, the roles of Mo and rare earths are investigated, since a proposed glass-ceramic waste form contains crystalline phases such as powellite [(AE,A,RE)MoO4] and oxyapatite [(RE,AE,A)10Si6O26], and the precipitation of molybdenum phases is known to be affected by the rare earth concentration in the glass. Additionally, the effects of other chemical additions have been systematically investigated, including Zr, Ru, P, and Ti. A series of studies were also undertaken to ascertain the effect of the RE size on glass structure and on partitioning to crystal phases, investigating similarities and differences in glasses containing single RE oxides of Sc, Y, La, Ce, Nd, Sm, Er, Yb, or Lu. Finally, the effect of charge compensation was investigated by considering not only the commonly assessed peralkaline glass but also metaluminous and peraluminous compositions. Glass structure and crystallization studies were conducted by spectroscopic methods (i.e., Raman, X-ray absorption, nuclear magnetic resonance (NMR), optical absorption, photoluminescence, photoluminescence excitation, X-ray photoelectron spectroscopy), microscopy (i.e., scanning electron microscopy, transmission electron microscopy, electron probe microanalysis), scattering (i.e., X-ray and neutron diffraction, small angle measurements), and physical characterization (i.e., differential thermal analysis, liquidus, viscosity, density). This paper will give an overview of the research program and some example unpublished results on glass-ceramic crystallization kinetics, microstructure, and Raman spectra, as well as some examples of the effects of rare earths on the absorption, luminescence, and NMR spectra of starting glasses. The formal collaboration described here has resulted in the generation of a large number of results, some of which are still in the process of being published as separate studies.

McCloy, John S.↗

Closed Fuel Cycle Waste Treatment Strategy

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Tritium Content in and Release from Pressurized-Water-Reactor Fuel Cladding

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels; >30 yrs). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritiated water (THO or T{sub 2}O). In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy'R cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste attributed to Zircaloy'R clad fuel by recovering the zirconium from the cladding for reuse. Treatment options for Zircaloy cladding include recycling to recover the significant value of the zirconium and eliminate costs associated with the disposal of transuranic-contaminated Zircaloy. If LWR fuel is reprocessed and the fuel is dissolved without tritium pretreatment, the tritium in the cladding will remain bound to the cladding. If tritium pretreatment is included in the flow sheet, some portion of the tritium in the cladding may be released along with the tritium from the fuel matrix. To design and operate reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the UNF must be understood and quantified. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0-96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. The present study was undertaken to understand how tritium pretreatment at standard tritium pretreatment conditions (oxidation in air at 480 to 600 deg. C) affects the tritium content in the Zircaloy cladding and the extent to which the tritium content could be reduced with modest increases in the tritium pretreatment temperature. The data indicate that <20% of the tritium in PWR cladding will be removed by pretreatment at 500-700 deg. C for 24 h. Essentially 100% of the tritium will likely be released if the heating time at 700 deg. C is extended to 96 h. Small amounts of tritium are expected to be released if the cladding is heated at ≤ 600 deg. C for up to 96 h. The results of this study indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the pretreatment system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in any subsequent processing of the cladding to recover/recycle the zirconium. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development and Optimization of Voltammetric Methods for Real Time Analysis of Electrorefiner Salt with High Concentrations of Actinides and Fission Products

This project addresses the problem of achieving accurate material control and accountability (MC&A) around pyroprocessing electrorefiner systems. Spent nuclear fuel pyroprocessing poses a unique challenge with respect to reprocessing technology in that the fuel is never fully dissolved in the process fluid. In this case, the process fluid is molten, anhydrous LiCl-KCl salt. Therefore, there is no traditional input accountability tank. However, electrorefiners (ER) accumulate very large quantities of fissile nuclear material (including plutonium) and should be well safeguarded in a commercial facility. Idaho National Laboratory (INL) currently operates a pyroprocessing facility for treatment of spent fuel from Experimental Breeder Reactor-II with two such ER systems. INL implements MC&A via a mass tracking model in combination with periodic sampling of the salt and other materials followed by destructive analysis. This approach is projected to be insufficient to meet international safeguards timeliness requirements. A real time or near real time monitoring method is, thus, direly needed to support commercialization of pyroprocessing. A variety of approaches to achieving real time monitoring for ER salt have been proposed and studied to date—including a potentiometric actinide sensor for concentration measurements, a double bubbler for salt depth and density measurements, and laser induced breakdown spectroscopy (LIBS) for concentration measurements. While each of these methods shows some promise, each also involves substantial technical complexity that may ultimately limit their implementation. Yet another alternative is voltammetry —a very simple method in theory that has previously been tested for this application to a limited extent. The equipment for a voltammetry system consists of off-the-shelf components (three electrodes and a potentiostat), which results in substantial benefits relative to cost and robustness. Based on prior knowledge of electrochemical reduction potentials for each of the species of interest, voltammetry can be used to quantify concentrations of a variety of elemental species—including uranium, plutonium, minor actinides, and rare earths. Various methods have been tested by other researchers to date—including cyclic voltammetry, square wave voltammetry, normal pulse voltammetry, etc. In most cases, it has been observed that there is a very limited concentration range for which the output can be readily correlated with concentration in the salt. Furthermore, testing to date has been limited to simple ternary salts with only a single element being quantified. While incomplete for application to MC&A for pyroprocessing, these results lead us to believe that voltammetry can be optimized based on salt properties and fundamental electrochemical rate processes to yield a highly accurate and robust method. This project is divided into four tasks jointly executed by three university research groups. This includes experimental measurement of key physical data on the systems of interest, development of a predictive voltammetry model, experimental validation of the voltammetry model, and design/verification of an optimized measurement method. This project supports the goals of the US-ROK Joint Fuel Cycle Study in addition to the NA-24 Office of the National Nuclear Security Agency and the International Atomic Energy Agency (IAEA).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios – (1) a sealed reactor with releases only during operational shutdown and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (PWR, BWR, RBMK). In the cases where continuous reprocessing of the fuel salt occurred both the radioxenon and radioiodine signatures are nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 1015 – 8×1016 Bq/d for 133Xe if no abatement is used. An assessment was also made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios: (1) a sealed reactor with releases only during operational shutdown, and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (Pressurized Water Reactor (PWR), BWR, RBMK). In cases where continuous reprocessing of the fuel salt occurred, both the radioxenon and radioiodine signatures were nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 10^15-8×10^16 Bq/d for 133Xe if no abatement is used. An assessment was made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1).

Johnson, Christine M.↗

A Model from the Mind to the Operating Facility - 16448

The reprocessing of Spent Magnox fuel has been undertaken at the Sellafield site in the United Kingdom (UK) for more than 50 years and is now entering the final few years of its operating programme. The age of this facility has resulted in additional operational challenges which have led to the need for a sustained programme of continuous improvement. It has therefore been necessary to assess what potential plant improvements are key to delivering a successful end to the operating programme. In 2003 the UK National Nuclear Laboratory (NNL) (then Nexia Solutions) worked with Sellafield Ltd to produce a visual discrete event simulation (DES) model. The DES approach models plant equipment (such as a storage pond crane or fuel element processing equipment), and the key processes that these items undertake. Process times, availability and process logic are all combined to provide a prediction of the plant performance. Through the use of 'what-if' scenarios, changes to any model parameter can be undertaken. The benefits of such an approach have been in the ability for Sellafield Ltd to understand which changes to make and which do not add value. This then allows the operating facility to focus the improvement programme on quantitatively underpinned benefits, thus increasing the potential to complete the reprocessing programme ahead of schedule, which is a key aim for Sellafield Ltd. Maximising the benefit of this model has required a close working relationship between NNL, Sellafield Ltd Technical and the Sellafield Ltd Operations teams to ensure that the model is using the correct information and is addressing the real issues at hand. Over the past 12 years many ideas have been assessed and three examples have been selected to demonstrate the range of use of the model, and the different approaches used to identify issues and opportunities, agree an approach and provide solutions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Current status of fluoride volatility method development

The Fluoride Volatility Method is based on a separation process, which comes out from the specific property of uranium, neptunium and plutonium to form volatile hexafluorides whereas most of fission products (mainly lanthanides) and higher transplutonium elements (americium, curium) present in irradiated fuel form nonvolatile tri-fluorides. Fluoride Volatility Method itself is based on direct fluorination of the spent fuel, but before the fluorination step, the removal of cladding material and subsequent transformation of the fuel into a powdered form with a suitable grain size have to be done. The fluorination is made with fluorine gas in a flame fluorination reactor, where the volatile fluorides (mostly UF{sub 6}) are separated from the non-volatile ones (trivalent minor actinides and majority of fission products). The subsequent operations necessary for partitioning of volatile fluorides are the condensation and evaporation of volatile fluorides, the thermal decomposition of PuF{sub 6} and the finally distillation and sorption used for the purification of uranium product. The Fluoride Volatility Method is considered to be a promising advanced pyrochemical reprocessing technology, which can mainly be used for the reprocessing of oxide spent fuels coming from future GEN IV fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Impact of Storage Time on the Needed Capture Efficiency for Volatile Radionuclides - 13369

During the processing of used nuclear fuel (UNF), volatile radionuclides will be discharged from the facility stack if no recovery processes are in place to limit their release. The volatile radionuclides of concern are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. There are three key regulations that address the release of these radionuclides to the environment- 40 CFR 61, 40 CFR 190, and 10 CFR 20. These regulations apply to the total radionuclide release and establish dose limits for the maximum exposed individual (MEI) in the public both in terms of whole body dose and dose to specific organs such as the thyroid. Each radionuclide released to the environment contributes to the total dose to some degree. In this paper we attempt to evaluate the efficiency requirements for the capture processes to limit the doses to the MEI to regulatory levels. Since the total amount of each volatile radionuclide present in the UNF changes with the age of the fuel, the respective capture requirements also change with time. Specifically, we are interested in the impact of the decreasing contribution of {sup 3}H and {sup 85}Kr, which have relatively short half-lives, 12.32 y and 10.76 y, respectively, with the increasing age of the fuel (i.e., time between when the UNF is removed from the reactor and the time it is processed) on the capture requirements. In this paper we examine the capture requirements for these four radionuclides for three fuel types (pressurized water reactor [PWR] with uranium oxide fuel [UOX], PWR with mixed oxide fuel [MOX], and an advanced high temperature gas-cooled reactor [AHTGR]), several burnup values, and time out of reactor extending to 200 y. We calculate doses to the MEI with the EPA code CAP-88 and look at two dose contribution cases. In the first case, we assume that the total allowable dose is attributed to only the four volatile radionuclides. This establishes the lowest capture efficiency value possible. Since this is unrealistic, because it assumes zero dose contribution from all other radionuclides, we also examine a second case, where only 10% of the allowable dose is assigned to the four volatile radionuclides. We calculate the required decontamination factors (DFs) for both of these cases for the three fuel types, multiple fuel burnups, and fuel ages and determine whether or not the dose to the whole body or to the thyroid that drives the capture requirements would require additional effluent controls for the shorter half-life volatile radionuclides based on dose considerations. This analysis indicates that the principal isotopes of concern are generally {sup 3}H and {sup 129}I, the latter requiring the highest DFs. The maximum DF value for {sup 129}I is 8000 for the evaluated cases and assumptions used. ∼60 for fresh fuels. The DF for {sup 14}C could be as high as 30 for certain fuels. These values are based on just meeting the regulatory limits, and additional engineering margins (perhaps 3x to 10x or higher) should be applied to provide a safety factor for compliance. However, by assuming less conservative dose allocations, taller stacks, different radionuclide speciation, fuel aging, and other reprocessing facility design and location parameters, the DF requirements could be significantly reduced. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Noria: A Highly Xe‐Selective Nanoporous Organic Solid

Abstract Separation of xenon and krypton is of industrial and environmental concern; the existing technologies use cryogenic distillation. Thus, a cost‐effective, alternative technology for the separation of Xe and Kr and their capture from air is of significant importance. Herein, we report the selective Xe uptake in a crystalline porous organic oligomeric molecule, noria, and its structural analogue, PgC‐noria, under ambient conditions. The selectivity of noria towards Xe arises from its tailored pore size and small cavities, which allows a directed non‐bonding interaction of Xe atoms with a large number of carbon atoms of the noria molecular wheel in a confined space.

Patil, Rahul S.↗