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730 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

As-run physics analysis for the EPRI-2 experiment for cycle 153B

The purpose of this ECAR is to document the as-run physics analysis and source term in Curies and decay heat rate (Watts) for the shipment of the EPRI-2 experiment located in the Center Flux Trap (CFT). EPRI-2 was irradiated during cycle 153B. The experiment irradiation ended on April 13, 2013. The as-run heat rates and flux/fluence results were calculated using the MCNP ATR full core model. The heat rates and flux are calculated based on an average center lobe power of 30.8 MW. The fluence is based on the operating time of 13.45 days. ORIGEN2 Version 2.2 is used to determine the decay heat rate, gamma spectrum, and source term for the EPRI-2 experiment. The source term analysis was performed as requested by the project staff to demonstrate compliance with shipping requirements in the GE-2000 cask following irradiation as well as provide an estimated source term to support Post-Irradiation Examination as needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy

To optimize nuclear waste repository performance, the destruction of minor actinide elements, particularly Np and Am, in a neutron fast spectrum reactor is possible by incorporating these elements into nuclear fuel. Evaluating the performance of minor actinide containing fuel is of paramount importance to enabling this technology. However, such a task is challenging without an available domestic fast spectrum test reactor. A comparison of fuel performance tested in an available domestic thermal reactor at the Idaho National Laboratory, the Advanced Test Reactor, and in a fast spectrum reactor in France (Phénix) is presented here in this study. This study evaluates the capability of using a cadmium shrouded test position to mimic the power profile along the fuel radius present in fast spectrum reactors so that thermally driven phenomenon (e.g., constituent redistribution) can be evaluated in a thermal reactor and determined to be prototypical of a fast reactor. Thus, optical microscopy and scanning electron microscopy has been performed on irradiated 35U-29Pu-4Am-2Np-30Zr fuel samples (where the number preceding the element is the weight percent concentration) from the two mentioned reactors that present similar irradiation temperatures and power conditions. The results indicate that fuel performance phenomena are reproducible in the two irradiation conditions. The redistribution of Zr occurred in the same manner for the two samples. Similar partitioning of U-Pu-Zr phases was observed, and the behavior of Am was similar in the analyzed specimens. Finally, the overall microstructure evolution seems not to be affected by minor actinides addition compared to expected behavior of conventional U-19Pu-10Zr ternary metal fuels for both specimens. Slight differences in fuel cladding chemical interaction were, however, observed. This difference is likely driven by difference in cladding composition rather than irradiation conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sensitivity of ATF Experiments in the Center Flux Trap of the Advanced Test Reactor to Adjacent Experiments

Irradiation experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor are typically assumed to have little effect on one another. This assumption does not hold true for certain experiments in close proximity. To evaluate the impacts on safety and programmatic parameters of experiments in the center flux trap, the contents of the adjacent H and inner-A positions were modeled with a range of possible irradiation targets. First, neutron flux maps with experiments in those positions were compared against a baseline configuration. Next, several safety and programmatic parameters for a generic accident-tolerant fuel test train were calculated. It was shown that these parameters can exhibit considerable sensitivity to the contents of the H and A experiment positions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

ATR Fuel Management

The sixth Core Internals Changeout (CIC) of ATR was completed on March 21, 2022, and Nuclear Testing (NT) commenced shortly thereafter. In support of NT, 235U flux wires were attached to flux wands using electrical tape and installed in water channels 2, 6, 11, 15 and 19 in each of the 40 ATR Mark VII fuel elements that comprise an ATR core to perform low power flux measurements during ATR Cycle 170CIC NT-3. The flux wands were fabricated from Poly (Methyl Methacrylate), or PMMA. Irradiated cobalt capsules were loaded in the ATR Northeast Irradiation Housing Assembly (NEIHA), the East Flux Trap (EFT), South Flux Trap (SFT), inboard A-positions and various H-positions as a 170CIC NT-3 core reactivity hold-down mechanism. The PMMA flux wands received significant gamma dose from the irradiated cobalt during NT-3 which caused discoloration, embrittlement and swelling of the flux wands. The removal of the degraded flux wands from the 0.078 in. wide by 50 in. long fuel element coolant channels has been a unique challenge. This presentation summarizes the cleaning techniques used to remove the flux wand debris from the elements, an aggressive PMMA irradiation experiment performed concomitantly with initial cleaning efforts, and planned inspection and qualification techniques that will be applied in the future to potentially requalify the impacted elements for re-use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evolution of Functional Groups during Pyrolysis Oil Upgrading

In this paper, we examine the evolution of functional groups (carbonyl, carboxyl, phenol, and hydroxyl) during stabilization at 100–200 °C of two typical wood derived pyrolysis oils from BTG and Amaron in a batch reactor over Ru/C catalyst for 4h. An aqueous and an oily phase were obtained. The content of functional groups in both phases were analyzed by GC/MS, 31P-NMR, 1H-NMR, elemental analysis, KF titration, carbonyl groups by Faix, Folin – Ciocalteu method and UV-Fluorescence. The consumption of hydrogen was between 0.007 and 0.016 g/g oil, and 0.001-0.020 g of CH4/g of oil, 0.005-0.016 g of CO2/g oil and 0.03-0.10 g H2O/g oil were formed. The content of carbonyl, hydroxyl, and carboxyl groups in the volatile GC-MS detectable fraction decreased (80, 65, and ~70% respectively), while their behavior in the total oil and hence in the non-volatile fraction was more complex. The carbonyl groups initially decreased having minimum at ~125-150°C and then increased, while the hydroxyl groups had reversed trend. This might be explained by initial hydrogenation of the carbonyl groups to form hydroxyls, followed by continued dehydration reactions at higher temperatures that may increase their content. The 31P-NMR was on the limit of its sensitivity for the carboxylic groups to precisely detect changes in the non-volatile fraction, however the more precise titration method showed that the concentration of carboxylic groups in the non-volatile fraction remains constant with increased stabilization temperature. The UV-Fluorescence results show that repolymerization increases with temperature. ATR-FTIR method coupled with deconvolution of the region between 1490 and 1850 cm-1 showed to be a good tool for following the changes in carbonyl groups and phenols of the stabilized pyrolysis oils. The deconvolution of the IR bands around 1050 and 1260 cm-1 correlated very well with the changes in the 31P-NMR silent O groups (likely ethers). Most of the H2O formation could be explained from the significant reduction of these silent O groups (from 12% in the fresh oils, to 6 to 2% in the stabilized oils) most probably belonging to ethers.

Stankovikj, Filip↗
DOE OSTI2024

Boiling-Water Reactor Testing Capability in the Advanced Test Reactor

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Nano-Engineered Catalyst Supported on Ceramic Hollow Fibers for the Utilization of CO 2 in Dry Reforming to Produce Syngas

The objective of this project was to develop a novel catalytic reactor containing nano-engineered catalysts for the utilization of CO 2 (captured from coal-fired power plants and other CO 2 emitting sources) in dry methane reforming (DMR) (CO 2 + CH 4 → 2 H 2 + 2 CO) to produce synthesis gas (syngas). The technology aims to reduce CO2 emissions by developing beneficial uses for CO 2 from coal-fired power plants. It also offers an alternative to mitigate CO 2 emissions in areas where geologic storage may not be an optimal solution and/or utilization could significantly offset the costs of carbon capture and sequestration. The nano-engineered Ni-based catalyst was prepared by atomic layer deposition (ALD). The Ni particles were as small as ~2-4 nm. The nano-engineered catalyst showed CH 4 conversion >95%, H 2 /CO ratio in the range of 0.7-1.0, and CH 4 reforming rate as high as 2,500 L/h/gNi at 850 ºC and pressure of 15-25 psia. The Ni-based ALD catalyst also showed good stability in DMR reaction during a 200-h continuous operation at 850 °C. This is due to strong bonding between the nanoparticles and substrates since the Ni nanoparticles were chemically bonded to the substrate during the ALD process. The high thermal stability maintains the high dispersion of Ni nanoparticles, which can inhibit coke formation because their step edges are small enough to limit carbon nucleation and growth. Technoeconomic analysis (TEA) indicates the levelized cost of syngas (LCOS) is $172/ton with our technology, which is lower than the equivalent (molar) cost of hydrogen produced by steam methane reforming (SMR) or autothermal reforming (ATR). The major operating cost is natural gas feed and fuel, and the levelized cost is highly sensitive to the price of natural gas and relatively insensitive to the CAPEX. Revenues from syngas could have a significant impact on the net cost of electricity (COE), depending on the cost of natural gas and the selling price of syngas, estimated at $36 per MWh if the syngas were sold at $195 per ton. Following DOE NETL’s guidance, a lifecycle analysis (LCA) was conducted to compare with SMR. The functional unit for the basis of comparison was defined as 1kg carbon monoxide in the product stream. The global warming potential (GWP) of our process was found to be 40% lower than the state-of-the-art SMR process. The sensitivity analysis confirms the emissions are most sensitive to the natural gas fuel requirements to deliver heat to the process.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2014

Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2018

Dose Uncertainty Quantification in AGC Experiment

This memorandum formally documents completion of the Advanced Reactor Technologies Level 3 Milestone (M3NT-18IN060504037), “Interoffice Memorandum detailing status of the AGC irradiation uncertainty analysis” due SeptemberAugust 30, 2018. This memorandum details the progress made toward the issuance of a final report detailing the AGC uncertainty analysis including the proposed methodology, progress on activities, and explanation of how the data will be used. Radiation damage estimation is an important component of the post-irradiation analysis of the AGC experiments. It depends primarily on the fast fluence, which is determined using well-established methods of spectral adjustments. These are based on best estimates from models (e.g. MCNP). Input cross-sections, and measured activities from flux wires in the experiment, propagate uncertainties in each of these. While the methods of propagating uncertainty are well-established, the final uncertainty estimates they provide are only as good as the estimates of uncertainty in the inputs on which they are based. The purpose of this work is to implement a method that eliminates some deficiencies in the ways input uncertainties are presently estimated. The rRadiation damage estimates for the AGC experiments are ostensibly based on experimental data in the form of flux wire activation measurements; each AGC experiment has been instrumented with flux wires at various locations along the axial length to provide this data. However, the estimatesy also depend on the results of MCNP calculations. To understand why, consider that the objective of the analysis is to estimate some number of parameters (the energy group fluences ?_i) that best reproduce the observations (the measured activities of the flux wires a_i).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗