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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Nuclear Data Sheets for A = 33

The experimental data are evaluated for known nuclides of mass number A=33 (Ne, Na, Mg, Al, Si, P, S, Cl, Ar, K). Detailed evaluated level properties and related information are presented, including adopted values of level and {gamma}-ray energies, decay data (energies, intensities and placement of radiations), and other spectroscopic data. This work supersedes earlier full evaluations of A=33 published by 1990En08 (also 1998En04 update) and 1978En02. Experimental studies for the identification of the {sup 33}Ne and {sup 33}K nuclides have not met with success as yet. Values of S(n) and S(p) from systematics for {sup 33}Ne and {sup 33}K, respectively suggest that these nuclei are likely to be unstable to particle emission. The radioactive decay schemes of {sup 33}Na and {sup 33}Al seem incomplete in view of large Q values and known excitations much below than allowed by the Q values. The {sup 33}S and {sup 33}Cl nuclides remain as the most extensively studied from many different reactions and decays.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

In-beam {gamma}-ray spectroscopy of {sup 35}Mg and {sup 33}Na

Excited states in the very neutron-rich nuclei {sup 35}Mg and {sup 33}Na were populated in the fragmentation of a {sup 38}Si projectile beam on a Be target at 83 MeV/u beam energy. We report on the first observation of {gamma}-ray transitions in {sup 35}Mg, the odd-N neighbor of {sup 34}Mg and {sup 36}Mg, which are known to be part of the 'island of inversion' around N=20. The results are discussed in the framework of large-scale shell-model calculations. For the A=3Z nucleus {sup 33}Na, a new {gamma}-ray transition was observed that is suggested to complete the {gamma}-ray cascade 7/2{sup +}{yields}5/2{sup +}{yields}3/2{sub gs}{sup +} connecting three neutron two-particle-two-hole intruder states that are predicted to form a close-to-ideal K=3/2 rotational band in the strong-coupling limit.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Engineering Database of Liquid Salt Thermophysical

The purpose of this report is to provide a review of thermophysical properties and thermochemical characteristics of candidate molten salt coolants, which may be used as a primary coolant within a nuclear reactor or heat transport medium from the Very High Temperature Reactor (VHTR) to a processing plant; for example, a hydrogen-production plant. Thermodynamic properties of four types of molten salts, including LiF-BeF 2 (67 and 33 mol%, respectively; also known as FLiBe), LiF-NaF-KF (46.5, 11.5, and 52 mol%, also known as FLiNaK), and KCl-MgCl 2 (67 and 33 mol%), and sodium nitrate-sodium nitrite-potassium nitrate (NaNO 3 –NaNO 2 –KNO 3 , 7-49-44 mol%, also known as Hitec® salt) have been investigated. Limitations of existing correlations to predict density, viscosity, specific heat capacity, surface tension, and thermal conductivity were identified. The impact of thermodynamic properties on the heat transfer, especially the Nusselt number, was also discussed. Stability of the molten salts with structural alloys and their compatibility with the structural alloys was studied. Nickel and high temperature alloys with dense Ni coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Of the chromium containing alloys, Hastelloy N appears to have the best corrosion resistance in fluorides, while Haynes 230 was the most resistant in chloride. In general, alloys with increasing carbon and chromium content are increasingly subject to corrosion by the fluoride salts FLiBe and FLiNaK due to attack and dissolution of the intergranular chromium carbide. Future research to obtain needed information was identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Comparison of Groundwater Sampling Technologies, Including Passive Diffusion Sampling, for Radionuclide Contamination - 20205

Using traditional high-flow purge methods for long-term water quality monitoring of deep groundwater wells can be expensive, affect contaminant migration, and produce excessive volumes of discharge water that can be difficult to manage. The use of low-flow pumping methods and depth discrete bailers (DDBs) can reduce the cost of sampling deep groundwater wells. In general, using different pumping methods to obtain reproducible and representative groundwater can be challenging. Passive diffusion samplers (PDSs) have successfully been used in long-term monitoring for volatile organic compounds, major ions, and trace-elements, but application has been limited for stable and radioactive isotopes. Beginning in 2018, the United States Geological Survey (USGS) conducted three sampling events to test the ability of PDSs and DDBs to obtain reproducible and representative groundwater samples. The sampled well is completed in a regional, permeable carbonate aquifer and is one in a set of wells that have historically been used for tritium tracer testing. All samples were obtained at 180 m (590 ft) below land surface (bls) in a 13.97 cm (5.5 inch) uncased well that has a total depth of 202 m (662 ft) bls. The first sampling event deployed a regenerated cellulose dialysis membrane (RCDM) PDS for 14 days with deionized water as the blank. The second sampling event deployed a RCDM for 27 days also with deionized water as the blank. The third sampling event deployed a Dual Membrane (DM) PDS for 65 days using a blank of tritium-dead carbonate water. The DM PDS was used to assess the effect of longer term deployment on tritium concentrations and address whether or not the PDSs reached equilibrium with ambient groundwater. The day after each of the three passive samplers were retrieved a DDB was used to obtain discrete non-integrated groundwater samples. For each DDB sampling day, the bailer was lowered into the well 10 consecutive times to determine if the water chemistry changed from the first to the last bailed sample. Quality assurance samples including blanks and duplicates were obtained during all three sampling events. All blank waters had tritium concentrations less than 21±33 pCi/L. Major ion (e.g. calcium, chloride, sodium, and sulfate) results were compared between all samples obtained with RCDM and DDB. Major ion concentrations showed a coefficient of variation of less than 6% between all RCDM and DDB samples; however, the coefficient of variation between the different deployment times and the two different methods for trace-element concentrations was much larger, particularly for manganese (82%), lead (41%), and zinc (33%). Stable isotope values were compared between the RCDM and DDB samples. The DDB sample results all fell within analytical uncertainty and were considered representative of the formation groundwater. The stable isotope values from the RCDM samples indicated that a longer deployment time was necessary to gain equilibrium and to obtain representative groundwater samples. Tritium results from groundwater samples obtained from the RCDM, DM, and DDB indicate that groundwaters obtained with PDSs produced tritium concentrations 2 to 3 times higher (between 1,426 and 3,060±87 pCi/L) than groundwaters obtained with a DDB (479 to 1,219±51 pCi/L). The longer the passive diffusion samplers were deployed, the higher the tritium concentration, suggesting that equilibrium with tritium was not reached within a 27-day deployment. DDB samples showed tritium results declining from the first to the last bailed sample for all three sampling events. This research suggests that tritium results from groundwater samples obtained from PDSs are more reproducible than samples obtained from DDBs. Also, PDSs likely do not accumulate isotopes of water but rather equilibrate with the ambient groundwater. On the other hand, both PDSs and DDBs were able to provide representative groundwater samples for major ions and have the potential to produce representative groundwater samples for stable isotopes of water (δD and δ{sup 18}O). Utilizing the most effective method for sampling groundwater is important for long-term groundwater monitoring. Quantifying the ability of passive diffusion groundwater samplers to obtain representative radioactive chemistry samples will help determine if these methods can be a useful alternative to pumping and purging wells for long-term monitoring. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Strongly coupled rotational band in Mg 33

The “island of inversion” at N≈20 for the neon, sodium, and magnesium isotopes has long been an area of interest both experimentally and theoretically due to the subtle competition between 0p-0h and np-nh configurations leading to deformed shapes. However, the presence of rotational band structures, which are fingerprints of deformed shapes, have only recently been observed in this region. In this work, we report on a measurement of the low-lying level structure of Mg33 populated by a two-stage projectile fragmentation reaction and studied with the Gamma Ray Energy Tracking In-Beam Nuclear Array (GRETINA). The experimental level energies, ground-state magnetic moment, intrinsic quadrupole moment, and γ-ray intensities show good agreement with the strong-coupling limit of a rotational model.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

High Sodium Solubility in Magnesiowüstite in Iron‐Rich Deep Lower Mantle

Abstract (Mg,Fe)O ferropericlase‐magnesiowüstite has been proposed to host the majority of Earth's sodium, but the mechanism and capacity for incorporating the alkali cation remain unclear. In this work, experiments in the laser‐heated diamond anvil cell and first‐principles calculations determine the solubility of sodium and favorability of sodium incorporation in iron‐rich magnesiowüstite relative to (Mg,Fe)SiO 3 bridgmanite. Reaction of Mg/(Mg + Fe) (Mg#) 55 and 28 olivine with NaCl at 33–128 GPa and 1600–3000 K produces iron‐rich magnesiowüstite containing several percent sodium, while iron‐rich bridgmanite contains little to no detectable sodium. In sodium‐saturated magnesiowüstite, sodium number [Na/(Na + Mg + Fe)] is 2–5 atomic percent at pressures below 60 GPa and drastically increases to 10–20 atomic percent at deep lower mantle pressures. For these two compositions, there is no significant dependence of the results on Mg#. Our calculations not only show consistent results with experiments but further indicate that such an increase in solubility and partitioning of Na into magnesiowüstite is driven by the spin transition in iron. These results provide fundamental constraints on the crystal chemistry of sodium at lower‐mantle conditions. If the sodium capacity of (Mg,Fe)O is not strongly dependent on Mg#, (Mg,Fe)O in the lower mantle may have the capacity to store the entire sodium budget of the Earth.

Geochemistry & Geophysics↗
DOE OSTI2013

Laboratory characterization of PM emissions from combustion of wildland biomass fuels

Particle emissions from open burning of southwestern (SW) and southeastern (SE) U.S. 17 fuel types during 77 controlled laboratory burns are presented. The fuels include SW 18 vegetation types: ceanothus, chamise/scrub oak, coastal sage scrub, California sagebrush, 19 manzanita, maritime chaparral, masticated mesquite, oak savanna, and oak woodland as 20 well as SE vegetation types: 1-year, 2-year rough, pocosin, chipped understory, 21 understory hardwood, and pine litter. The SW fuels burned at a higher Modified 22 Combustion Efficiency (MCE) than the SE fuels resulting in lower particulate matter 23 (PM) mass emission factor (EF). Particle size distributions for six fuels and particle 24 number emission or all fuels are reported. Excellent mass closure (slope = 1.00, r2=0.94) 25 between ions, metals, and carbon with total weight was obtained. Organic carbon 26 emission factors inversely correlated (= 0.72) with MCE, while elemental carbon (EC) 27 had little correlation with MCE (=0.10). The EC/total carbon (TC) ratio sharply 28 increased with MCE for MCEs exceeding 0.94. The average levoglucosan and total Poly 29 Aromatic Hydrocarbons (PAH) emissions factors ranged from 25-1272 mg/kg fuel and 30 1790-11300 μg/kg fuel, respectively. No correlation between MCE and emissions of 31 PAHs/levoglucosan was found. Additionally, PAH diagnostic ratios were observed to be 32 poor indicators of biomass burning. Large fuel-type and regional dependency was 33 observed in the emission rates of ammonium, nitrate, fluoride, chloride, sodium, and

Hosseini, SeyedEhsan↗
DOE OSTI2015

MC2-3 / DIF3D Analysis for the ZPPR-15 Doppler and Sodium Void Worth Measurements

This manuscript covers validation efforts for our deterministic codes at Argonne National Laboratory. The experimental results come from the ZPPR-15 work in 1985-1986 which was focused on the accuracy of physics data for the integral fast reactor concept. Results for six loadings are studied in this document and focus on Doppler sample worths and sodium void worths. The ZPPR-15 loadings are modeled using the MC2-3/DIF3D codes developed and maintained at ANL and the MCNP code from LANL. The deterministic models are generated by processing the as-built geometry information, i.e. MCNP input, and generating MC2-3 cross section generation instructions and a drawer homogenized equivalence problem. The Doppler reactivity worth measurements are small heated samples which insert very small amounts of reactivity into the system (< 2 pcm). The results generated by the MC2-3/DIF3D codes were excellent for ZPPR-15A and ZPPR-15B and good for ZPPR-15D, compared to the MCNP solutions. In all cases, notable improvements were made over the analysis techniques applied to the same problems in 1987. The sodium void worths from MC2-3/DIF3D were quite good at 37.5 pcm while MCNP result was 33 pcm and the measured result was 31.5 pcm. Copyright © (2015) by the American Nuclear Society All rights reserved.

Smith, Micheal A.↗
DOE OSTI2026

Evaluating CO 2 mitigation strategies in SAF biorefineries: Techno-economic and life cycle analysis

The aviation sector requires scalable decarbonization strategies, and lignocellulosic sustainable aviation fuel (SAF) represents a promising pathway. This study comparatively evaluates the techno-economic analysis and life cycle assessment (LCA) of three CO 2 management strategies integrated within a U.S.-based gasification–Fischer–Tropsch SAF biorefinery: (i) catalytic hydrogenation of captured CO 2 to methanol, (ii) geological CO 2 sequestration, and (iii) mineralization to sodium bicarbonate (NaHCO 3 ). Techno-economic analysis indicates that methanol synthesis requires approximately 26% higher capital investment and 33% higher operating costs than mineralization. Although methanol co-production generates the highest gross revenue, NaHCO 3 production reduces the SAF minimum selling price by approximately 38% relative to both methanol synthesis and geological sequestration pathways, reflecting a more balanced cost allocation through mineral co-product valorization. Geological sequestration lowers operating costs by nearly 50% compared with methanol synthesis but remains highly dependent on carbon credit mechanisms. LCA reveals substantial divergence in climate performance. Relative to methanol synthesis, sequestration improves net greenhouse gas performance by approximately 163%, transitioning the system from net-positive to net-negative emissions. Mineralization further enhances carbon mitigation, achieving roughly 85% greater carbon reduction than sequestration and over sixfold improvement relative to methanol synthesis within the defined system boundary. Sensitivity analysis identified hydrogen price, co-product market value, and process emissions as dominant drivers. Under baseline assumptions, CO 2 mineralization is found to offer the most balanced pathway.

Carbon capture and storage↗
DOE OSTI2016

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Giant piezoelectricity in oxide thin films with nanopillar structure

High-performance piezoelectric materials are critical components for electromechanical sensors and actuators. For more than 60 years, the main strategy for obtaining large piezoelectric response has been to construct multiphase boundaries, where nanoscale domains with local structural and polar heterogeneity are formed, by tuning complex chemical compositions. We used a different strategy to emulate such local heterogeneity by forming nanopillar regions in perovskite oxide thin films. We obtained a giant effective piezoelectric coefficient d 33 , f * of ~1098 picometers per volt with a high Curie temperature of ~450°C. Our lead-free composition of sodium-deficient sodium niobate contains only three elements (Na, Nb, and O). The formation of local heterogeneity with nanopillars in the perovskite structure could be the basis for a general approach to designing and optimizing various functional materials.

Science & Technology - Other Topics↗
DOE OSTI2017

Modeling Chemical Evolution of Cement Seals

The prevention of the migration of fluids is a field of study with high demand in both the private industry as well as in government research and development. More particularly, the prevention of carbon dioxide (CO 2 ) migration or return to the atmosphere. This, in turn, causes measurably significant atmospheric climate change that results in immense damage to the environment. In this study, the interaction between 0.5 molar sodium chloride (NaCl) brine solution and a cement interface over extended periods of time is computationally investigated, as well as the interaction of a 33 millimolar CO 2 dissolved in the afore-mentioned brine solution for the same period of time. A 2-meter-long saturated brine solution reacted with an 800-meter-long cement block in PFLOTRAN, an open-source reactive flow transport software whose output was later visualized in ParaView, a separate, open-source, visualization software.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Evaluation of fuel-borne sodium effects on a DOC-DPF-SCR heavy-duty engine emission control system: Simulation of full-useful life

Here we report that for renewable fuels to displace petroleum, they must be compatible with emissions control devices. Pure biodiesel contains up to 5 ppm Na + K and 5 ppm Ca + Mg metals, which have the potential to degrade diesel emissions control systems. This study aims to address these concerns, identify deactivation mechanisms, and determine if a lower limit is needed. Accelerated aging of a production exhaust system was conducted on an engine test stand over 1001 h using 20% biodiesel blended into ultra-low sulfur diesel (B20) doped with 14 ppm Na. This Na level is equivalent to exposure to Na at the uppermost expected B100 value in a B20 blend for the system full-useful life. During the study, NOx emissions exceeded the engine certification limit of 0.33 g/bhp-hr before the 435,000-mile requirement. Replacing aged diesel oxidation catalyst (DOC), diesel particulate filter (DPF), and selective catalytic reduction (SCR) devices with new degreened parts showed that each device contributed equally to the NOx increase. Following this systems-based evaluation, a detailed investigation of the individual components was completed. Na was determined to have minimal impact on DOC activity. For this system, it is estimated that B20-Na resulted in 50% more ash into the DPF. However, the Na did not diffuse into the cordierite DPF nor degrade its mechanical properties. The SCR degradation was found to be caused by a small amount of precious group metals (PGM) contamination that increased NH3 oxidation, and lowered NOx reduction. Therefore, we determined that the primary effect of Na in this study is through increased ash in the DPF rather than deactivation of the catalytic activity.

42 ENGINEERING↗
DOE OSTI2011

Preparation, crystal structures and rapid hydration of P2- and P3-type sodium chromium antimony oxides

Two new Na{sub x}[Cr{sub (1+x)/2}Sb{sub (1-x)/2}]O{sub 2} compounds have been prepared by solid-state reactions in argon. Their structures have been determined by the X-ray Rietveld method. Both new phases together with NaCrO{sub 2}-based solid solution comprise brucite-like layers of edge-shared (Cr,Sb)O{sub 6} octahedra but differ by packing mode of the layers and coordination of the interlayer Na{sup +} ions. A P3 phase exists at x{approx}0.5-0.58. It is rhombohedral (R3-bar m), a=2.966, c=16.937 A at x{approx}0.58, with 29% Na{sup +} occupancy of trigonal prisms. A P2 phase exists at x{approx}0.6-0.7. It is hexagonal (P6{sub 3}/mmc), a=2.960, c=11.190 A at x{approx}0.7, with 37% and 33% Na{sup +} occupancy of two non-equivalent trigonal prisms. Both P2 and P3 phases rapidly absorb moisture in air; packing mode is preserved, the a parameter changes slightly but c increases by 24-25%. Very high sodium ion conductivity is predicted for both P2 and P3 anhydrous phases. -- Graphical Abstract: Polyhedral presentation of layered structures of Na{sub x}[Cr{sub (1+x)/2}Sb{sub (1-x)/2}]O{sub 2}. (Cr,Sb)O{sub 6} octahedra are green. Part of the sodium prisms are open to show short distances between sodium sites which cannot be occupied simultaneously. Display Omitted highlights: > Two new layered Na{sub x}[Cr{sub (1+x)/2}Sb{sub (1-x)/2}]O{sub 2} phases, P2 and P3, have been prepared in argon. > Interlayer O-O distances are larger than in Na{sub x}(M,Ti)O{sub 2} analogs (M=Cr,Ni,Li). > As a result, both rapidly hydrate in air with c-axis expansion of 24-25%. > Bottleneck radii are also larger, and this predicts very high Na{sup +}-ion conductivity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Electrical Conductivity and Electrochemical Characteristics of Na{sub 3}V{sub 2}(PO{sub 4}){sub 3}-Based NASICON-Type Materials

NASICON-type materials with the compositions Na{sub 3}V{sub 2–x}Al{sub x}(PO{sub 4}){sub 3}, Na{sub 3}V{sub 2-} {sub x}Fe{sub x}(PO{sub 4}){sub 3}, Na{sub 3+} {sub x}V{sub 2–x}Ni{sub x}(PO{sub 4}){sub 3}, and Na{sub 3}V{sub 2-} {sub x}Cr{sub x}(PO{sub 4}){sub 3} (x = 0, 0.03, 0.05, and 0.1) have been prepared and characterized by X-ray diffraction analysis, electron microscopy, and impedance spectroscopy. The results demonstrate that the highest electrical conductivity among the samples studied is offered by the material doped with 5% Fe: Na{sub 3}V{sub 1.9}Fe{sub 0.1}(PO{sub 4}){sub 3}. The activation energy for low-temperature conduction in the doped materials decreases from 84 ± 2 to 54 ± 1 kJ/mol and that for high-temperature conduction is ~33 kJ/mol. The discharge capacity of Na{sub 3}V{sub 1.9}Fe{sub 0.1}(PO{sub 4}){sub 3}/C under typical working conditions of cathodes of sodium ion batteries has been shown to exceed that of Na{sub 3}V{sub 2}(PO{sub 4}){sub 3}/C. The capacity of the more porous material prepared by the Pechini process (Na{sub 3}V{sub 1.9}Fe{sub 0.1}(PO{sub 4}){sub 3}/C-{II}) approaches the theoretical one at a low charge–discharge rate and retains its high level as the charge rate is raised (its discharge capacity was 117.6, 108.8, and 82.6 mAh/g at a discharge rate of 0.1C, 2C, and 8C, respectively).

25 ENERGY STORAGE↗
DOE OSTI2015

Stationary Liquid Fuel Fast Reactor

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel container is penetrated by twelve hexagonal control assembly (CA) guide tubes, each of which has 3.0 mm thickness and 69.4 mm flat-to-flat outer distance. The distance between two neighboring CA guide tube is selected to be 26 cm to provide an adequate space for CA driving systems. The fuel container has 18181 penetrating coolant tubes of 6.0 mm inner diameter and 2.0 mm thickness. The coolant tubes are arranged in a triangular lattice with a lattice pitch of 1.21 cm. The fuel, structure, and coolant volume fractions inside the fuel container are 0.386, 0.383, and 0.231, respectively. Separate steel reflectors and B4C shields are used outside of the fuel container. Six gas expansion modules (GEMs) of 5.0 cm thickness are introduced in the radial reflector region. Between the radial reflector and the fuel container is a 2.5 cm sodium gap. The TRU inventory at the beginning of equilibrium cycle (BOEC) is 5081 kg, whereas the TRU inventory at the beginning of life (BOL) was 3541 kg. This is because the equilibrium cycle fuel contains a significantly smaller fissile fraction than the LWR TRU feed. The fuel inventory at BOEC is composed of 34.0 a/o TRU, 41.4 a/o Ce, 23.6 a/o Co, and 1.03 a/o solid fission products. Since uranium-free fuel is used, a theoretical maximum TRU consumption rate of 1.011 kg/day is achieved. The semi-continuous fuel cycle based on the 300-batch, 1- day cycle approximation yields a burnup reactivity loss of 26 pcm/day, and requires a daily reprocessing of 32.5 kg of SLFFR fuel. This yields a daily TRU charge rate of 17.45 kg, including a makeup TRU feed of 1.011 kg recovered from the LWR used fuel. The charged TRU-Ce-Co fuel is composed of 34.4 a/o TRU, 40.6 a/o Ce, and 25.0 a/o Co.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Inorganic tin aluminophosphate nanocomposite for reductive separation of pertechnetate

Pertechnetate (TcO4-) is the most abundant chemical form of radioactive contaminant 99Tc present in the legacy nuclear waste streams and in the subsurface of the nuclear waste storage sites.One proposed remediation approach is reductive separation of TcO4-and sequestration in low-temperature waste forms. The development of the relevant technologies has been slow due to the lack of reductive materials that retain their functionalityand are otherwise suitable for application in multicomponent or aggressive media such as highly alkaline, brine-like solutions typifying nuclear tank wastes. This research prepared a tin-based reductive material and demonstrated its potential utility for the separation of TcO4- from the alkaline nuclear wastes. This material consists of Sn(II/IV) phosphate supported by the polycrystalline aluminophosphate matrix. The aluminophosphate matrix is inert to the reaction conditions and offers the benefits of high stability and low solubility in the concentrated alkaline solutions. This Sn(II/IV)-basedmaterial exhibitshigh loading capacity for Tc and selectively removed major fraction of TcO4-from the tank waste supernatant simulants, which contains 7.8 M total sodium and 2.43 M free hydroxide concentrations. The observed Kd values for Tc are about 13,000 and 2,200 mL g-1 for the simulant solutions containing no or 33 mM Cr(VI), respectively, positioning Sn(II/IV) aluminophosphate among the best performing reductive sorbents for TcO4- developed to date. This advanced behaviour is attributed to the synergistic combination of the Sn(II/IV) aluminophosphate functionalities. The presence of Sn(II/IV)-rich fibres facilitates the reduction of TcO4- to Tc(IV), which getsimbedded along the fibre branches. Importantly, the Sn(IV)-containing inert polycrystalline matrix alsoincorporates Tc(IV)which triggers its crystallization to rutile SnO2 phase andstabilizesTc(IV) in the polycrystalline matrix.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2012

An improved energy-collapsing method for core-reflector modelization in SFR core calculations using the PARIS platform

In the framework of the ASTRID project, sodium cooled fast reactor studies are conducted at CEA in compliance with GEN IV reactors criteria, particularly for safety requirements. An improved safety requires better calculation tools to obtain accurate reactivity effects (especially sodium void effect) and power map distributions. The current calculation route lies on the JEFF3.1.1 library and the classical two-step approach performed with the ECCO module of the ERANOS code system at the assembly level and the Sn SNATCH solver - implemented within the PARIS platform - at the core level. 33-group cross sections used by SNATCH are collapsed from 1968-group self-shielded cross-section with a specific flux-current weighting. Recent studies have shown that this collapsing is non-conservative when dealing with core-reflector interface and can lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector. Such a discrepancy is due to the flux anisotropy at the interface, which is not taken into account when cross sections are obtained from separate fuel and reflector assembly calculations. A new approach is proposed in this paper. It consists in separating the self-shielding and the flux calculations. The first one is still performed with ECCO on separate patterns. The second one is done with SNATCH on a 1D traverse, representative of the core-reflector interface. An improved collapsing method using angular flux moments is then carried out to collapse the cross sections onto the 33-group structure. In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms of k{sub eff} and power map are strongly improved for a small increase of the computing time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗