Enriched Uranium Research Articles

Molybdenum and Tungsten isotope compositions of UO2 fuel pellets: Implications for isotopically enriched taggants

The addition of isotopically enriched taggants to material at the front end of the nuclear fuel cycle could be a powerful tool used to assist law enforcement authorities should material outside of regulatory control be found. Two potential candidates for this purpose are molybdenum (Mo) and tungsten (W) as both elements have five or more stable isotopes and are trace elements contained within nuclear fuel. However, there is a concern that Mo and W could undergo isotope fractionation during processes like uranium enrichment and to date, it is unknown if nuclear fuels have natural Mo and W isotope compositions. If Mo and W isotopic variability is present in nuclear fuels, this would hinder the use of these elements as isotopic taggants because it would be difficult to discern the original taggant isotope composition with high confidence. Therefore, a set of 16 low enriched uranium (LEU) fuel pellets from US commercial producers was analyzed using multi collector-inductively coupled plasma mass spectrometry (MC-ICPMS) to determine Mo and W isotope compositions (i.e., 94Mo/92Mo, 95Mo/92Mo, 96Mo/92Mo, 97Mo/92Mo, 98Mo/92Mo, 183W/182W, 184W/182W, and 186W/182W). Relative to terrestrial standards, LEU fuel pellets have variable Mo and W isotope compositions, thereby complicating the use of these elements as isotopically enriched taggants. As such, this work demonstrates that the isotope composition of any potential taggant must be well characterized in the base nuclear fuel prior to any taggant addition. Furthermore, these results suggest that Mo and W are not ideal candidates for isotopically enriched taggants.

Polyphosphonate-segmented macroporous organosilicon frameworks for efficient dynamic enrichment of uranium with in-situ regeneration

Efficient separation and enrichment of uranium from radioactive effluents is of strategic significance for sustainable development of nuclear energy and environmental protection. Macropore structure of adsorbent is conducive to accessibility of the pore and transport of the adsorbate during dynamic adsorption. However, the low specific surface area results in fewer ligand sites and subsequently reduces the adsorption capacity. Herein, we present a novel strategy for efficient dynamic uranium enrichment using polyphosphonate-segmented macroporous organosilicon frameworks (PMOFs). PMOFs are constructed through the copolymerization of diethyl vinylphosphonate and triethoxyvinylsilane, followed by hydrolysis and condensation of the oligomers. The introduction of polyphosphonate segments into the frameworks endows PMOFs with a macroporous structure (31 µm) and a high ligand content (up to 72 wt%). Consequently, the optimized PMOF-3 demonstrated an ultrahigh dynamic adsorption capacity of 114.8 mg/g among covalently conjugated silicon-based materials. Additionally, PMOF-3 achieves a high enrichment factor (120) in the dynamic enrichment of uranium on a fixed bed column, which can be in-situ regenerated with 1 M NaHCO3 as the eluent. This work presents a new strategy for efficient dynamic enrichment of nuclides, which can be extended to the separation of other specific pollutants, shedding new light on adsorbent design and technical innovation.

Micrococcal nuclease sequencing of porcine sperm suggests enriched co-location between retained histones and genomic regions related to semen quality and early embryo development.

The mammalian spermatozoon has a unique chromatin structure in which the majority of histones are replaced by protamines during spermatogenesis and a small fraction of nucleosomes are retained at specific locations of the genome. The sperm's chromatin structure remains unresolved in most animal species, including the pig. However, mapping the genomic locations of retained nucleosomes in sperm could help understanding the molecular basis of both sperm development and function as well as embryo development. This information could then be useful to identify molecular markers for sperm quality and fertility traits. Here, micrococcal nuclease digestion coupled with high throughput sequencing was performed on pig sperm to map the genomic location of mono- and sub-nucleosomal chromatin fractions in relation to a set of diverse functional elements of the genome, some of which were related to semen quality and early embryogenesis. In particular, the investigated elements were promoters, the different sections of the gene body, coding and non-coding RNAs present in the pig sperm, potential transcription factor binding sites, genomic regions associated to semen quality traits and repeat elements. The analysis yielded 25,293 and 4,239 peaks in the mono- and sub-nucleosomal fractions, covering 0.3% and 0.02% of the porcine genome, respectively. A cross-species comparison revealed positional conservation of the nucleosome retention in sperm between the pig data and a human dataset that found nucleosome enrichment in genomic regions of importance in development. Both gene ontology analysis of the genes mapping nearby the mono-nucleosomal peaks and the identification of putative transcription factor binding motifs within the mono- and the sub- nucleosomal peaks showed enrichment for processes related to sperm function and embryo development. There was significant motif enrichment for Znf263, which in humans was suggested to be a key regulator of genes with paternal preferential expression during early embryogenesis. Moreover, enriched positional intersection was found in the genome between the mono-nucleosomal peaks and both the RNAs present in pig sperm and the RNAs related to sperm quality. There was no co-location between GWAS hits for semen quality in swine and the nucleosomal sites. Finally, the data evidenced depletion of mono-nucleosomes in long interspersed nuclear elements and enrichment of sub-nucleosomes in short interspersed repeat elements.These results suggest that retained nucleosomes in sperm could both mark regulatory elements or genes expressed during spermatogenesis linked to semen quality and fertility and act as transcriptional guides during early embryogenesis. The results of this study support the undertaking of ambitious research using a larger number of samples to robustly assess the positional relationship between histone retention in sperm and the reproductive ability of boars.

Multiple usage of thorium-based fuel in a VVER-1000 reactor

This paper considers the use of unconventional fuel in nuclear power reactors, using the example of a VVER-type unit, in order to find out the possibility of saving natural fissile uranium nuclei. Saving fissile uranium is one of the important tasks, the solution of which will give time for the development of a two-component nuclear power industry that will have no problems with fuel resources. However, at present, the reserves of cheap uranium can provide the existing level of global nuclear energy for only 80–100 years. The main components of this proposed fuel are 232Th and fissile isotopes of uranium: 235U (loaded) and 233U (produced from thorium). All the uranium isotopes and added 235U nuclei at the beginning of the campaign account for about 6% of the number of thorium nuclei and uranium isotopes. The abbreviated name of this fuel is TORUR-5. To keep fissionable nuclei in the fuel cycle after the spent fuel is unloaded, it is envisaged that all the heavy nuclei will be returned back to the reactor after they have been cleaned from fission fragments, i.e., the fuel cycle will be closed. At the same time, the principle of annual movement of fuel assemblies (as they burn up) is the same as in the existing VVER-1000 reactors. Using the Serpent software, a reactor model was built, the composition and dimensions of which were close to the parameters of the VVER-1000 serial unit. The main results of calculations were the quantitative compositions of isotopes annually loaded into the reactor as well as the amounts of 235U and thorium added also annually. The analysis of the obtained results allowed us to make the following conclusions. The annual reloading of 235U during the computation period is required almost at a constant level and, in comparison with uranium fuel, is about half as much. This is feasible for the following reasons. Part of the fissions of 235U is replaced by the fission of 233U produced from 232Th. In addition, fissionable nuclei are kept in the closed Th-U fuel cycle. This is the first “advantage” of the proposed fuel. TORUR-5 requires uranium enriched to at least 90%, the cost of which is several times higher than that of 3–5% enriched uranium. But since much less highly enriched uranium is required, the cost of fuel for a TORUR-5-fueled VVER-1000 reactor is significantly lower. This is the second “advantage” of the proposed fuel. The negative characteristic of TORUR-5, which requires further investigation, is that, after the initial loading, several uranium isotopes appear in the returned fuel, the total radioactivity of which, according to estimates, exceeds the radioactivity of traditional 3–5% enriched uranium fuel by several thousand times. At the same time, the radioactivity of discharged spent conventional fuel exceeds the radioactivity of fresh fuel by millions of times, and this problem has been solved at NPPs both organizationally and technically. Therefore, it will be necessary to develop a technology for loading TORUR-5, taking into account the estimated radioactivity.

Uranium-Bearing Layers of Sandstone Type Uranium Deposits Identification and Three-Dimensional Reconstruction in the Northern Ordos Basin, North-Central China

Sandstone type uranium is the most valuable and has the most potential for mining among the known uranium deposits. In the process of forming, the hydrolytic migration and enrichment of uranium require special basin sedimentary environment and tectonic background. Therefore, the mineralization process of sandstone type uranium deposits has certain layering characteristics and distribution rules in the underground vertical depth space. It is important to mine the spatial distribution characteristics of vertical uranium-bearing layers, and thus, reconstruct the three-dimensions of uranium orebodies. In this paper, according to the metallogenic law and distribution characteristics of sandstone type uranium in the underground vertical space, a nonlinear uranium-bearing layers identification (NULI) method of sandstone type uranium is proposed by using different types, resolutions and scales of borehole data. Then, the depth of uranium mineralization for the Daying uranium deposit within northern Ordos Basin is identified accurately and the spatial distribution characteristics of the uranium-bearing layer on the exploration line are obtained. Finally, the occurrence mode of the underground uranium orebodies are presented by using three-dimensional reconstruction analysis. It provides a basis for the prediction, exploration and mining of sandstone type uranium deposits within the Ordos Basin.

A Review of the Metallogenic Mechanisms of Sandstone-Type Uranium Deposits in Hydrocarbon-Bearing Basins in China

As a valuable mineral resource, uranium is extensively utilized in nuclear power generation, radiation therapy, isotope labeling, and tracing. In order to achieve energy structure diversification, reduce dependence on traditional fossil fuels, and promote the sustainable development of energy production and consumption, research on the metallogenic mechanisms and related development technologies of uranium resources has been one of the focuses of China’s energy development. Sandstone-type uranium deposits make up approximately 43% of all deposits in China, making them the most prevalent form of uranium deposit there. Sandstone-type uranium deposits and hydrocarbon resources frequently coexist in the same basin in China. Therefore, this study summarizes the spatial and chronological distribution, as well as the geological characteristics, of typical sandstone-type uranium deposits in China’s hydrocarbon-bearing basins. From the perspectives of fluid action, geological structure, and sedimentary environment, the metallogenic mechanisms of sandstone-type uranium deposits in hydrocarbon-bearing basins are explored. According to the research, the rapid reduction effect of oil and gas in the same basin is a major factor in the generation of relatively large uranium deposits. Additionally, ions such as CO32− and HCO3− in hydrothermal fluids of hydrocarbon-bearing basins, which typically originate from dispersed oil and gas, are more conducive to uranium enrichment and sedimentation. This study provides guidance for efficient sandstone-type uranium deposit exploration and production in hydrocarbon-bearing basins and helps to achieve significant improvements in uranium resource exploitation efficiency.

Structural insights into regulation of the PEAK3 pseudokinase scaffold by 14-3-3

PEAK pseudokinases are molecular scaffolds which dimerize to regulate cell migration, morphology, and proliferation, as well as cancer progression. The mechanistic role dimerization plays in PEAK scaffolding remains unclear, as there are no structures of PEAKs in complex with their interactors. Here, we report the cryo-EM structure of dimeric PEAK3 in complex with an endogenous 14-3-3 heterodimer. Our structure reveals an asymmetric binding mode between PEAK3 and 14-3-3 stabilized by one pseudokinase domain and the SHED domain of the PEAK3 dimer. The binding interface contains a canonical phosphosite-dependent primary interaction and a unique secondary interaction not observed in previous structures of 14-3-3/client complexes. Additionally, we show that PKD regulates PEAK3/14-3-3 binding, which when prevented leads to PEAK3 nuclear enrichment and distinct protein-protein interactions. Altogether, our data demonstrate that PEAK3 dimerization forms an unusual secondary interface for 14-3-3 binding, facilitating 14-3-3 regulation of PEAK3 localization and interactome diversity.

Using optimal squared-off cascades instead of optimal square cascades to produce enriched uranium required for fresh fuel in the equilibrium cycle in power reactors

Using square cascades instead of symmetrical tapered cascades to produce enriched uranium in the equilibrium cycle of a power reactor, on the one hand, eliminates the dilution unit in the enrichment facilities, and on the other hand, reduces the separation efficiency of the cascades. In this work, in order to approach the separation efficiency of square cascades to symmetrical conical, the arrangement of square cascades is changed to squared-off cascades in the form of 2-section, 3-section and 4-section. The developed software codes “SQC-PSOA”, “2SSQ-PSOA”, “3SSQ-PSOA” and “4SSQ-PSOA” are used for the design and optimization. The results show: a) The average separation capacity of a centrifuge in 2-section, 3-section and 4-section squared-off cascades shows an increase of 1.2–6.3%, 24.4–28.8% and 28.1–32.3%, respectively, compared to the square cascade; b) The total number of square cascades compared to 2-section, 3-section and 4-section squared-off cascades show an increase of 3.4–5.9%, 16.6–20.8% and 16.7–23.5%, respectively.

Numerical Simulations of Flow-Induced Deflections in MITR LEU Fuel Plate Due to Channel Size Disparity

The hydromechanical stability of the fuel plates in parallel coolant channels of a Materials Testing Reactor (MTR) fuel element design is of great importance to the safety of research and test reactors. Previous analytical, experimental, and numerical efforts focused on parallel channels with the same or similar size; also, in the prior numerical simulations, the fuel plate was often assumed to be perfectly flat. This work presents the results of a fluid-structure interaction simulation performed to evaluate the flow-induced deflections of the fuel plates in the low-enriched uranium (LEU, <20 wt% 235U) fuel element design for the conversion (from highly enriched uranium) of the Massachusetts Institute of Technology Reactor (MITR-II, also referred to as MITR). Various manufacturing and assembly tolerances of the MITR LEU elements are considered in the analysis, and the effects of channel size disparity, nonideal plate shape, and flow rate uncertainty are investigated. Results show that, for all cases analyzed, the deflection occurs toward the larger channel, and the change in any channel stripe remains small (less than 0.021 mm) compared to fabrication tolerances. In addition to simulation work, a hydraulic performance test of the MITR LEU fuel element is currently planned to support conversion to the use of LEU fuel.

Lineage segregation in human pre-implantation embryos is specified by YAP1 and TEAD1.

Lineage segregation in human pre-implantation embryos is specified by YAP1 and TEAD1.

Development of schemes from a single cascade to multi-cascades for separation of regenerated uranium

Returning precious fissile materials containing 235U, obtained from regenerated uranium, to the nuclear fuel cycle is important for the production of sustainable nuclear energy. Herein, starting from the basic single cascade separation scheme, schemes for enriching regenerated uranium are developed to satisfy the requirements for low-enriched uranium with a commercial grade and several supplementary separation conditions to the greatest extent for different considerations involving the consumption of natural uranium, scale of separation cascade, and material recovery. Double- and triple-cascade schemes are derived from the consideration of improving the basic scheme by eliminating its drawbacks such as the contamination of the entire cascade by artificial nuclides. All schemes are optimized based on a Q-model cascade and are compared. The results demonstrate that the triple-cascade scheme exhibits a good separation performance and may be a good choice among the existing schemes. However, the concentration of 235U in this scheme exceeds the limitation specified by the international standard for handling fissile materials; therefore, this scheme requires further improvement.

Application of SCALE code for reactor physics analysis of the as-built Nigeria miniature neutron source reactor operating with low enriched uranium fuel

This study presents a broad reactor physics analysis including, neutronic, burnup, and decay heat of the ‘as-built’ NIRR-1 LEU core performed using SCALE 6.2.3 code package. First, verification of neutronic parameters of the NIRR-1 was performed by modeling the reactor ‘as-built’ LEU core. Thereafter validation of the model was done by comparing calculated result with measured data during the commissioning of the reactor. The results showed that the estimated neutronic parameters using KENO-VI code of SCALE 6.2.3 code package are in close agreement with the measured data. Hence, KENO-VI can be used for detailed neutronic assessment of MNSRs. The total decay heat at the end of core life after operation for 216 EFPD was estimated to be 1620.15 W and was majorly due to the contribution from fission products which was estimated to be 1608.10 W. This value explains why MNSRs are cooled through natural convection of the coolant during operation and at shutdown conditions. Additionally, the burnup of the core was determined to be 0.53 wt% which is within the expected burnup for a typical MNSRs (less than1 wt%) with a core lifetime of about 200 EFPD (∼10 years). Furthermore, the mass of 239Pu buildup after 216 EFPD was far less than the IAEA category III nuclear material and indicates a huge advantage of research reactor conversion in reducing proliferation risk.

Radioactive Fission Waste of The Conversion of High-Enriched Uranium to Low-Enriched Uranium Target on 99Mo Production

In the past, 99mTc produced by PT. Industri Nuklir Indonesia (PT. INUKI) was carried out through the processing of a 99Mo parent radioisotope that was produced using high enriched uranium target. Due to restrictions on the use of high-enriched uranium (HEU) by the International Atomic Energy Agency, PT. INUKI has converted the process using a low enriched uranium (LEU) target. This study aims to determine the change of radioactive fission waste characteristics in 99Mo production due to the conversion process. The characteristics of radioactive fission waste generated from low-enriched uranium targets were calculated using ORIGEN 2.1 program. The results show that the radioactive fission waste contains radionuclides of activation products, actinides, and fission products. The conversion produced smaller activity of 99Mo compared to that of using the HEU target, which was 397 compared to 1010 Ci. The conversion generated radioactive fission waste with a smaller content of remaining 235U but greater content of actinides, particularly 239Pu. The activity of 239Pu in the radioactive fission waste from LEU was 29.1 μCi, approximately 19 times higher compared to that of using HEU, which was 1.52 μCi. After 50 years of decay, this radioactive fission waste was calculated to have a specific activity of 6.54x108 Bq/g, lower than that of the highly enriched uranium target, which was 3.01x109 Bq/g. This radioactive fission waste requires management with a high safety level.

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)-Iso-Butyramide (DEHiBA) at High Metal Loadings

ABSTRACT Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. By increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. The extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

In suit establishment of Ti3C2 MXenes exchanger via Na+, K+ and NH4+ intercalation: Exceptionally UO22+ loading capability and hijacking mechanism

Uranium naturally occurs over four billion tonnes in the seawater, which can be extracted for use as nuclear fuel. Herein, an efficient Ti3C2 MXenes catcher was constructed by a facile and environmental friendly chemistry intercalation strategy and directed at uranium capture. We demonstrate the promising loading capability of MXenes with expansion interlayer spacing due to the intercalation of Na+, K+ and NH4+. We report high sequestration capacity (qm-Na = 1099.40 mg/g and qm-K 876.05 mg/g), large KdU values (104 –105 mL/g, despite the presence of a tremendous excess of Na+, K+, Ca2+ and Mg2+ system). Remarkable, even in the 104-folds of V5+ system, the adsorption efficiency of Na+-MXenes still reached 51.98 %. Real uranium spiked seawater extraction experiments indicated that the sample exhibits promising harvesting uranium performance with ∼ 95.92 ∼ 3643.84 μg/g adsorption capacity after 24 h contact in ∼ 30.34 ∼ 1702.37 μg/L initial concentration. Importantly, the uranyl ion in the exchanged MXenes products could be easily eluted by hydrochloric acid and hydroxide during continuous adsorption–desorption cycles. HRTEM, XRD, FTIR and XPS tests confirmed that ion exchange and surface coordination were the main uptake mechanisms. These advantages of intercalation functionalization method coupled with the very high loading capacity, easy fabrication and environmentally friendly nature providing a new insight for further development of highly efficient uranium enrichment materials.

Integrative transcriptomics and cell systems analyses reveal protective pathways controlled by Igfbp-3 in anthracycline-induced cardiotoxicity.

Anthracyclines such as doxorubicin (Dox) are effective chemotherapeutic agents; however, their use is hampered by subsequent cardiotoxicity risk. Our understanding of cardiomyocyte protective pathways activated following anthracycline-induced cardiotoxicity (AIC) remains incomplete. Insulin-like growth factor binding protein (IGFBP) 3 (Igfbp-3), the most abundant IGFBP family member in the circulation, is associated with effects on the metabolism, proliferation, and survival of various cells. Whereas Igfbp-3 is induced by Dox in the heart, its role in AIC is ill-defined. We investigated molecular mechanisms as well as systems-level transcriptomic consequences of manipulating Igfbp-3 in AIC using neonatal rat ventricular myocytes and human-induced pluripotent stem cell-derived cardiomyocytes. Our findings reveal that Dox induces the nuclear enrichment of Igfbp-3 in cardiomyocytes. Furthermore, Igfbp-3 reduces DNA damage, impedes topoisomerase IIβ expression (Top2β) which forms Top2β-Dox-DNA cleavage complex leading to DNA double-strand breaks (DSB), alleviates detyrosinated microtubule accumulation-a hallmark of increased cardiomyocyte stiffness and heart failure-and favorably affects contractility following Dox treatment. These results indicate that Igfbp-3 is induced by cardiomyocytes in an effort to mitigate AIC.

In-field alpha spectrometry system for uranium enrichment verification of uranium hexafluoride

A portable In-Field Alpha Spectrometry (IFAS) system was developed to support on-site uranium enrichment determination for samples taken at facilities working with gaseous UF6. It consists of a modified commercial alpha spectrometer, a vacuum pump, a laptop computer, and associated connectors and cables. The IFAS uses specially designed Single-Use Destructive Assay (SUDA) samplers to capture gaseous UF6 and convert and store it as UO2F2 in a thin-film zeolite. To examine the quantitative performance of the IFAS system a large number of replicate SUDA samples was acquired for natural uranium (NU, 0.72 wt%, 23 samples), low enriched uranium 1 (LEU1, 2.33 wt% 235U, 10 samples), and low enriched uranium 2 (LEU2, 4.62 wt% 235U, 10 samples). The standard deviation (SD) values for uranium enrichment determination with the IFAS for NU, LEU1, and LEU2, in 20-hour assays, were 0.037 wt% 235U, 0.063 wt% 235U, and 0.072 wt% 235U, respectively. A separate analysis suggests that comparable SD values can be achieved in less time, approximately 8–10 h. For context, these SD values are well below the International Atomic Energy Agency’s International Target Values for the determination of uranium enrichment in non-destructive assay measurements.

A cascade electro-dehydration process for simultaneous extraction and enrichment of uranium from simulated seawater

Uranium extraction from seawater has become a crucial issue that has raised tremendous attention. The transport of water molecules along with salt ions through an ion-exchange membrane is a common phenomenon for typical electro-membrane processes such as selective electrodialysis (SED). In this study, a cascade electro-dehydration process was proposed for the simultaneous extraction and enrichment of uranium from simulated seawater by taking advantage of water transport through ion-exchange membranes and the high permselectivity of membranes for monovalent ions against uranate ions. The results indicated that the electro-dehydration effect in SED allowed 1.8 times the concentration of uranium with a loose structure CJMC-5 cation-exchange membrane at a current density of 4 mA/cm2. Thereafter, a cascade electro-dehydration by a combination of SED with conventional electrodialysis (CED) enabled approximately 7.5 times uranium concentration with the extraction yield rate reaching over 80% and simultaneously desalting the majority of salts. Overall, a cascade electro-dehydration is a viable approach, creating a novel route for highly effective uranium extraction and enrichment from seawater.

Plutonium Signatures in Molten-Salt Reactor Off-Gas Tank and Safeguards Considerations

Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.

Thermal Assessment of Repository Footprint Requirement for Advanced Fuel Cycles

Decay heat is an important constraint for repository size and design because it can drive processes that affect performance and compromise critical materials. This paper investigates how compliance with repository thermal limits is affected by three decay heat management strategies: waste package loading, waste package spacing, and surface storage time. In particular, this paper focuses on how repository area, a result of package spacing, is impacted by waste loading and surface storage time. A two-part analytical heat transfer model is presented and executed iteratively to determine the minimum allowable repository area. The analysis considers three generic close-contact repository designs along with the wastes generated from the 40 fuel cycle analysis examples used to generate metric data for the Fuel Cycle Evaluation and Screening study. Detailed results are presented for two fuel cycles: the once-through use of low-enriched uranium fuel in light water reactors and the continuous recycling of U and Pu in sodium fast reactors. Two limits for surface storage time are identified: the time required for disposal to be possible at all and the time at which further surface storage time yields no gains. The impact of waste loading is also diminished with increasing surface storage time. In general, the generic salt repository is most flexible to accept high-heat-generating wastes with less surface storage time than other repository environments. Limited-recycle fuel cycles are shown to pose a disposal challenge because of elevated, sustained decay heat generation in the waste.