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At least 451 records · Page 25

Confinement of Iodine Molecules into Triple-Helical Chains within Robust Metal–Organic Frameworks

During nuclear waste disposal process, radioactive iodine as a fission product can be released. The widespread implementation of sustainable nuclear energy thus requires the development of efficient iodine stores that have simultaneously high capacity, stability and more importantly, storage density (and hence minimized system volume). Here, we report high I 2 adsorption in a series of robust porous metal–organic materials, MFM-300(M) (M = Al, Sc, Fe, In). MFM-300(Sc) exhibits fully reversible I 2 uptake of 1.54 g g –1 , and its structure remains completely unperturbed upon inclusion/removal of I 2 . Direct observation and quantification of the adsorption, binding domains and dynamics of guest I 2 molecules within these hosts have been achieved using XPS, TGA-MS, high resolution synchrotron X-ray diffraction, pair distribution function analysis, Raman, terahertz and neutron spectroscopy, coupled with density functional theory modeling. These complementary techniques reveal a comprehensive understanding of the host–I 2 and I 2 –I 2 binding interactions at a molecular level. The initial binding site of I 2 in MFM-300(Sc), I 2 I , is located near the bridging hydroxyl group of the [ScO 4 (OH) 2 ] moiety [I 2 I ···H–O = 2.263(9) Å] with an occupancy of 0.268. I 2 II is located interstitially between two phenyl rings of neighboring ligand molecules [I 2 II ···phenyl ring = 3.378(9) and 4.228(5) Å]. I 2 II is 4.565(2) Å from the hydroxyl group with an occupancy of 0.208. Significantly, at high I 2 loading an unprecedented self-aggregation of I 2 molecules into triple-helical chains within the confined nanovoids has been observed at crystallographic resolution, leading to a highly efficient packing of I 2 molecules with an exceptional I 2 storage density of 3.08 g cm –3 in MFM-300(Sc).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Fuel cycles and envisioned roles of fast neutron reactors and hybrids

Future innovative nuclear fuel cycles will require insuring sustainability in terms of safe operation, optimal use of resources, radioactive waste minimization and reduced risk of proliferation. The present paper introduces some basic notions and fundamental fuel cycle strategies. The simulation approach needed to evaluate the impact of the different fuel cycle alternatives will also be shortly discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutronics for critical fission reactors and subcritical fission in hybrids

The requirements of future innovative nuclear fuel cycles will focus on safety, sustainability and radioactive waste minimization. Critical fast neutron reactors and sub-critical, external source driven systems (accelerator driven and fusion-fission hybrids) have a potential role to meet these requirements in view of their physics characteristics. This paper provides a short introduction to these features.

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Italian hybrid and fission reactors scenario analysis

Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

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The neutronics studies of fusion fission hybrid power reactor

In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

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Diagnostics for hybrid reactors

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

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Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Heavy metal inventory and fuel sustainability of recycling TRU in FBR design

Nuclear fuel materials from spent fuel of light water reactors have a potential to be used for destructive devices with very huge energy release or in the same time, it can be utilized as a peaceful energy or civil applications, for generating electricity, desalination of water, medical application and others applications. Several research activities showed some recycled spent fuel can be used as additional fuel loading for increasing fuel breeding capability as well as improving intrinsic aspect of nuclear non-proliferation. The present investigation intends to evaluate the composition of heavy metals inventories and fuel breeding capability in the FBR design based on the loaded fuel of light water reactor (LWR) spent fuel (SF) of 33 GWd/t with 5 years cooling time by adopting depletion code of ORIGEN. Whole core analysis of FBR design is performed by adopting and coupling codes such as SLAROM code, JOINT and CITATION codes. Nuclear data library, JFS-3-J-3.2R which is based on the JENDL 3.2 has been used for nuclear data analysis. JSFR design is the basis design reference which basically adopted 800 days cycle length for 4 batches system. Higher inventories of plutonium of MOX fuel and TRU fuel types at equilibrium composition than initial composition have been shown. Minor actinide (MA) inventory compositions obtain a different inventory trends at equilibrium composition for both fuel types. Higher Inventory of MA is obtained by MOX fuel and less MA inventory for TRU fuel at equilibrium composition than initial composition. Some different MA inventories can be estimated from the different inventory trend of americium (Am). Higher americium inventory for MOX fuel and less americium inventory for TRU fuel at equilibrium condition. Breeding ratio of TRU fuel is relatively higher compared with MOX fuel type. It can be estimated from relatively higher production of Pu-238 (through converted MA) in TRU fuel, and Pu-238 converts through neutron capture to produce Pu-239. Higher breeding ratio of MOX fuel and TRU fuel types at equilibrium condition are estimated from converted fertile material during reactor operation into fissile material of plutonium such as converted uranium fuel (converted U-238 into Pu-239) or additional converted fuel from MA into Pu-238 and changes into Pu-239 by capturing neutron. Loading LWR SF gives better fuel breeding capability and increase inventory of MA for doping material of MOX fuel; however, it requires more supply MA inventory for TRU fuel type.

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Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

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Neutronics of accelerator-driven subcritical fission for burning transuranics in used nuclear fuel

We report the development of a conceptual design for accelerator-driven subcritical fission in a molten salt core (ADSMS). ADSMS is capable of destroying all of the transuranics at the same rate and proportion as they are produced in a conventional nuclear power plant. The ADSMS core is fueled solely by transuranics extracted from used nuclear fuel and reduces its radiotoxicity by a factor 10,000. ADSMS offers a way to close the nuclear fuel cycle so that the full energy potential in the fertile fuels uranium and thorium can be recovered.

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Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy

A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.

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Phase evolution studies in CaZrTi{sub 2}O{sub 7}−RE{sub 2}Ti{sub 2}O{sub 7} (RE = Nd{sup 3+}, Sm{sup 3+}) system: Futuristic ceramic host matrices for nuclear waste immobilization

Series of compositions with general stoichiometry as Ca{sub 1−x}Zr{sub 1−x}RE{sub 2x}Ti{sub 2}O{sub 7} (RE = Nd{sup 3+}, Sm{sup 3+}) were prepared by solid state reaction and characterized by powder x-ray diffraction technique to unravel the phase fields in the title systems. The phase fields in CaZrTi{sub 2}O{sub 7−}Nd{sub 2}Ti{sub 2}O{sub 7} and CaZrTi{sub 2}O{sub 7−}Sm{sub 2}Ti{sub 2}O{sub 7} systems differed significantly at the rareearth rich regions. The common phase fields like zirconolite-2M, zirconolite-4M, cubic perovskite are observed at the zirconolite rich regions of both systems. Depending on the structure of RE{sub 2}Ti{sub 2}O{sub 7} phase, the cubic pyrochlore or monoclinic RE{sub 2}Ti{sub 2}O{sub 7} phases are observed in the studied system. The observed phase fields in these two systems indicate ionic radius of the rare-earth ion has a dominating role in the phase relations. Further details of the phases and their homogeneity are explained in the text of the manuscript.

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'Strategy is a commodity, implementation is an art' - 2 years of implementation of the UK national LLW strategy

The Low Level Waste Repository (LLWR) is the primary facility for disposal of Low Level Waste (LLW) in the United Kingdom (UK), serving the UK nuclear industry and a diverse range of other sectors. Management of LLW in the UK historically was dominated by disposal to the LLWR. The value of the LLWR as a national asset was recognised by the 2007 UK Governmental Policy on management of solid LLW. At this time, analysis of the projected future demand for disposal at LLWR against facility capacity was undertaken identifying a credible risk that the capacity of LLWR would be insufficient to meet future demand if existing waste management practices were perpetuated. To mitigate this risk a National Strategy for the management of LLW in the UK was developed by the Nuclear Decommissioning Authority (NDA), partnered with LLW Repository Ltd. (the organisation established in 2008 to manage the LLWR on behalf of NDA). This strategy was published in 2010 and identified three mechanisms for protection of the capacity of LLWR - application of the Waste Hierarchy by waste producers; optimised use of existing assets for LLW management; and opening of new waste treatment and disposal routes to enable diversion of waste away from the LLWR. (authors)

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Preliminary waste management plan of European spallation

The European Spallation Source (ESS) is the European common effort in designing and building a next generation large-scale user facility for studies of the structure and dynamics of materials. The proposed schematic layout of the ESS facility is based on a linear driver (linac) directing the proton beam (5 MW of 2.5 GeV) of 2.8 ms long pulses with a 20 Hz on a tungsten target where neutrons are produced via spallation reactions. Further the neutrons will be moderated to thermal and sub-thermal energies in a couple of moderators placed around the target. The moderators feed 22 beamlines guiding the neutrons to the scattering instruments, mainly for neutron scattering research, as has been previously mentioned. The objective of this work is to develop a waste management plan for ESS facility. In this respect two important aspects are analyzed. First the present status of the problem is outlined as follow. Estimate types and quantities of waste that the ESS project will generate at different stages: commission, operation, decommissioning were derived using: i) precise Monte Carlo calculations ii) scaling the activity from the operation experience of the existing spallation source installations for waste such it is difficult to predict level of activation or for components of the facility in stage of the preconceptual model. Associated waste treatment/conditioning options and final disposal route were further analyzed in order to define the waste type and packet descriptions in agreement with Swedish regulations and policy. It was found that the compilation of completely new waste type descriptions for qualification of the ESS waste for disposal will be necessary. Particular attention was devoted to 'problematic waste' as Beryllium reflector, C-14 from graphite used as core zone of the beam-dump and collimators or waste arising from the purification systems of both Helium and water cooling circuits. Management of waste on ESS site: collection/segregation systems, characterization system, storage options, is also described. In the second step, the acquired information is used for planning and implementing actions involving all participants (ESS, treatment facility operator, disposal operator, regulatory body and other authorized authorities). (authors)

07 ISOTOPE AND RADIATION SOURCES↗

The long-term safety and performance analyses of the surface disposal facility for the Belgian category a waste at Dessel

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium. This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology. An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal. This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified. (authors)

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Limitation of tritium outgassing from tritiated solid waste drums

In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO{sub 2}, Ag{sub 2}O, Pt and molecular sieve, the three first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient.

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Enhanced capacity and stability for the separation of cesium in electrically switched ion exchange

Electrically switched ion exchange (ESIX) can be used to separate ionic contaminants from industrial wastewater, including that generated by the nuclear industry. The ESIX method involves sequential application of reduction and oxidation potentials to an ion exchange film to induce the respective loading and unloading of cesium. This technology is superior to conventional methods (e.g electrodialysis reversal or reverse osmosis) as it requires very little energy for ionic separation. In previous studies, ESIX films have demonstrated relatively low ion exchange capacities and limited film stabilities over repeated potential applications. In this study, the methodology for the deposition of electro-active films (nickel hexacyanoferrate) on nickel electrodes was modified to improve the ion exchange capacity for cesium removal using ESIX. Cyclic voltammetry was used to investigate the ion exchange capacity and stability. Scanning electron microscopy (SEM) was used to characterize the modified film surfaces. Additionally, the films were examined for the separation of cesium ions. This modified film preparation technique enhanced the ion exchange capacity and improves the film stability compared to previous methods for the deposition of ESIX films. (authors)

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Ion Exchange Kinetics Testing with SRF Resin

The U.S. Department of Energy (DOE) Hanford Site contains more than 53 million gallons of legacy waste generated as a byproduct of plutonium production and reprocessing operations. The wastes are a complex mixture composed mostly of NaNO 3 , NaNO 2 , NaOH, NaAlO 2 , Na 3 PO 4 , and Na 2 SO 4 , with a number of minor and trace metals, organics, and radionuclides stored in underground waste tanks. The DOE Office of River Protection (ORP) has contracted Bechtel National Incorporated (BNI) to build a pretreatment facility, the River Protection Project-Waste Treatment Plant (RPP-WTP), that will separate long-lived transuranics (TRU) and highly radioactive components (specifically 137 Cs and, in selected cases, 90 Sr) from the bulk (non-radioactive) constituents and immobilize the wastes by vitrification. The plant is designed to produce two waste streams: a high-volume low-activity waste (LAW) and a low-volume high-activity waste (HLW).

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