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DOE OSTI2019

3D Porous Graphitic Nanocarbon for Enhancing Performance and Durability of Pt Catalysts: Balance between Graphitization and Hierarchical Porosity

Carbon supports used in oxygen-reduction cathode catalysts for proton exchange membrane fuel cells (PEMFCs) are vulnerable to corrosion under harsh operating conditions, leading to poor performance durability. To address this issue, we have developed a highly stable porous graphitic carbon (PGC) produced through pyrolysis of a 3D polymer hydrogel in combination with Mn. The resulting PGC features multilayer carbon sheets assembled in porous and flower-like morphologies. In-situ high-temperature electron microscopy was employed to dynamically monitor the carbonization process up to 1100°C, suggesting that 3D polymer hydrogel provides high porosity at multiple scales, and that Mn catalyzes the graphitization process more effectively than other metals. Compared to conventional carbons such as Vulcan, Ketjenblack, and graphitized carbon, the PGC provides an improved balance between high graphitization and hierachical porosity, which is favorable for uniform Pt nanoparticle dispersion and enhanced corrosion resistance. As a result, Pt supported on PGC exhibits remarkably enhanced stability. In addition to thorough testing in aqueous electrolytes, we also conducted fuel cell testing using durability protocols recommended by the U.S. Department of Energy (DOE). After 5000 voltage cycles from 1.0 to 1.5 V, the Pt/PGC catalyst only lost 9 mV at a current density of 1.5 A/cm 2 , dramatically exceeding the DOE support durability target (<30 mV), and surpassing commercial Pt/C catalysts. Along with the enhanced carbon corrosion resistance of the PGC support, the enhanced catalyst-support interactions are beneficial for stability improvement, likely due to nitrogen doping into carbon, which was further elucidated through X-ray absorption spectroscopy and density functional theory (DFT) calculations. Thus, the high stability and activity of PGC-based Pt catalysts are attributed to the combination of high graphitization degree, favorable surface areas and porosity, and nitrogen doping, which effectively stabilize highly dispersed Pt nanoparticles.

25 ENERGY STORAGE↗

NASA NTRS1986

Mass flow measurement of liquid cryogens using the triboelectric effect

A cross correlator technique using triboelectric technology has been shown to be a feasible method to measure liquid flow rate for liquid nitrogen and JP4 jet fuel. This technology, invented and pioneered by Auburn International, Inc., is also expected to be suitable for use with all other insulating liquids and cryogens. The technology described is particularly well suited for cryogenic use, since the sensor is non-contacting and non-intrusive, and therefore, causes no additional pressure drop within the flow stream. Further development of the in-line sensor is required to produce a prototypical version for the test purposes under SSME fuel flow conditions. However, with the knowledge gained from this feasibility study, it is very likely that an acceptable sensor design for a full test bed evaluation could be produced.

Dechene, Ronald L.↗

DOE OSTI2017

Biomass Energy Generation Project

The Municipal Electric Utility of the City of Cedar Falls (dba Cedar Fals Utilities or CFU) received a congressionally directed grant funded through DOE-EERE to run three short (4 hour) duration test burns and one long (10 days) duration test burn to test the viability of renewable fuels in Streeter Station Boiler #6, a stoker coal fired electric generation unit. The long test burn was intended to test supply chain assumptions, optimize boiler combustion and assess the effects of a longer duration burn of biomass on the boiler.

09 BIOMASS FUELS↗

DOE OSTI2017

ON-GOING STATUS OF KJRR FUEL (U-7MO) QUALIFICATION

In order to cope with global shortage of Mo-99 supplies and with growing demand of neutron transmutation doping, KJRR construction plan has been launched since April 2012 to provide self-sufficiency of domestic RI demand, and to extend Si doping capacity for power device market growth. Through comprehensive surveillance of the fuels in-reactor behavior, KAERI has selected the fuel meat of U-7%Mo dispersion in an aluminum matrix with 5wt%Si for the KJRR fuel. As part of the efforts for fuel licensing and qualification of the KJRR fuel, an LTA irradiation test at the ATR started from November 2015 was successfully completed by reaching at 219 EFPD in the end of February 2017. Together with the results of HAMP-1 already completed irradiation and PIE, the successful irradiation of the LTA also demonstrates the fuel integrity under more rigorous conditions than the KJRR operation conditions. This paper updates the current status of the KJRR U7Mo (8 g-U/cm3) LTA irradiation and PIE plan up to date as of February 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRSDate not supplied

The Use of Immersion Rigs for High Temperature Hydrogen Exposure Testing Within the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) was designed to perform non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and materials. NTREES can simulate the thermal hydraulic environment within an NTR and has been leveraged to provide critical information on component performance and material compatibility. NTREES has the ability to expose test articles to pressurized Hydrogen gas up to 1000 psig and 225 g/s. NTREES tests are operated in a deoxygenated environment of Nitrogen or Argon pressurized up to 1000 psig. NTREES simulates fission heat through induction heating powered by a 1.2 MW AC power supply, and NTREES has recorded temperatures exceeding 3700K. To perform hot Hydrogen exposure tests on materials more rapidly and earlier in the fuel development process, NTREES must be able to test small material samples on the order of 1 square inch in volume, individual fuel assembly subcomponents, and particle materials. To meet these needs, multiple testing rigs have been developed to immerse test specimens in hot Hydrogen, with each "immersion rig" varying in its capability of sample type and performance range in order to meet the broadest range of testing needs. Herein is a discussion of a selection of these “immersion rigs.”

Nuclear Thermal Propulsion↗

DOE OSTI2018

Test Plan for the 1200°C Heating Test of AGR-3/4 Compact 10-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 10-2. The test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1200°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating would most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose a number of AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Test Plan for the 1400°C Heating Test of AGR-3/4 Compact 10-4

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 10-4. The test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1400°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating would most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose a number of AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Test Plan for the 1400°C Heating Test of AGR-3/4 Compact 8-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). This test plan describes the heating test of AGR-3/4 Compact 8-2. The test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1400°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating would most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose a number of AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Preliminary Project Execution Plan for the TREAT LEU Fuel Conversion Project

This project execution plan for the Transient Reactor Test (TREAT) facility Low Enriched Uranium (LEU) Fuel Conversion Project informs the reader of a high-level path forward for conversion of TREAT facilities at Idaho National Laboratory and provides guidance for INL personnel regarding the TREAT LEU Fuel Conversion Project. This project is part of the National Nuclear Security Administration’s Global Threat Reduction Initiative Reactor Conversion Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

SE2-ANL Subchannel Analysis with New Heating Calculation Based on VARIANT Transport Code

The multi-assembly, steady-state sub-channel analysis code SE2-ANL has been widely used in thermal-hydraulic analysis of sodium cooled fast reactors (SFR). It is used to compute the core-wide temperature profiles for a given flow distribution and to determine the assembly flow rates yielding a desired (assembly-wise) peak temperature distribution. It is a modified version of SUPERENERGY-2 coupled with the DIF3D/GAMSOR/DIF3D procedure for coupled neutron and gamma heating calculations, developed at Argonne National Laboratory (ANL). In the current version of SE2-ANL, the neutron and gamma fluxes are calculated using the finite-difference diffusion theory option of the DIF3D code with six triangular meshes per hexagonal assembly. In addition, the pin power distributions are estimated by assuming that the power distribution within an assembly is separable in the radial and axial directions. Furthermore, the total heat generation rate in an assembly duct wall is simply determined by a user-specified fraction of the total gamma heating within that assembly. In order to determine the pin power distributions more accurately by eliminating these limitations, a new coupled heating calculation procedure has recently developed based on the neutron and gamma transport calculations with the VARIANT code. For this, the GAMSOR code has been extended to generate the intra-nodal gamma source distribution to be consistent with the trial functions of VARIANT. Using the computed neutron and gamma flux distributions, the power distribution in each fuel pin and duct wall is determined using a new utility code CURVE, which has been developed to evaluate the VARIANT flux solution at the center of each fuel pin and along the duct mid-wall. For each pin segment contained in a VARIANT node, the axial power shape is represented as a quadratic profile. The SE2-ANL code has been modified to interface with the new, coupled neutron and gamma heating calculation scheme. In addition, several improvements have been made on the computational methods and models of SE2-ANL. To eliminate the limit on the axial mesh size due to the numerical instability problem of the current explicit scheme, the numerical scheme for discretizing the energy equation in the axial direction has been changed to the {theta}-method of the SLTHEN code. An automated flow allocation scheme has also been implemented in SE2-ANL to determine assembly flow rates such that the peak cladding mid-wall temperatures of individual orifice zones are equalized over the burn cycle. The impacts of the new heating calculation scheme based on VARIANT transport solutions on the cladding and fuel temperatures have been tested against the existing method of SE2-ANL. Using the existing and new heating calculation schemes, full-core subchannel analyses were performed for the metal core design of the 1000 MWt Advanced Burner Reactor (ABR-1000), and the resulting coolant, cladding and fuel temperatures were compared. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Non-Destructive Postirradiation Examination of the AFC-4C Capsule/Rodlet 5 Experiment

The AFC-4C irradiation test was irradiated at the Idaho National Laboratory Advanced Test Reactor to investigate several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and coating/barrier were tested in ferritic-martensitic HT-9 cladding. The AFC-4C capsule 5 was removed from the reactor at around 8.7 at %HM burnup to begin postirradiation examination. This report presents and discusses the non-destructive PIE results of the AFC-4C-capsule / rodlet 5, a U-10Zr solid fuel, sodium bonded with HT9 with internal Cr-coating, and its performance is evaluated against the historical fuel performance of previously irradiated fuel from literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2015

Effects of Burning Alternative Fuel in a 5-Cup Combustor Sector

A goal of NASA's Environmentally Responsible Aviation (ERA) program is to develop a combustor that will reduce the NOx emissions and that can burn both standard and alternative fuels. To meet this goal, NASA partnered with General Electric Aviation to develop a 5-cup combustor sector; this sector was tested in NASA Glenn's Advanced Subsonic Combustion Rig (ASCR). To verify that the combustor sector was fuel-flexible, it was tested with a 50-50 blend of JP-8 and a biofuel made from the camelina sativa plant. Results from this test were compared to results from tests where the fuel was neat JP-8. Testing was done at three combustor inlet conditions: cruise, 30% power, and 7% power. When compared to burning JP-8, burning the 50-50 blend did not significantly affect emissions of NOx, CO, or total hydrocarbons. Furthermore, it did not significantly affect the magnitude and frequency of the dynamic pressure fluctuations.

alternative fuel↗

NASA NTRS1997

Gravitational Influences on Flame Propagation Through Non-Uniform Premixed Gas Systems

We have built an apparatus for measuring flame spread rates through non-homogeneous fuel-air mixtures as a function of layer thickness and concentration. The layer thickness is adjusted by controlling the diffusion time above a fuel-saturated porous media, while the concentration is controlled by the fuel temperature. Normal gravity tests with methanol have so far explored largely the effect of temperature, as well as the effects of various aspects of the apparatus. Good agreement with previous research has been obtained. We have also demonstrated the ability of a rainbow schlieren system to quantitatively measure fuel vapor concentrations in the static case.

Miller, Fletcher J.↗

NASA NTRS1973

Effect of increased fuel temperature on emissions of oxides of nitrogen from a gas turbine combustor burning natural gas

An annular gas turbine combustor was tested with heated natural gas fuel to determine the effect of increasing fuel temperature on the formation of oxides of nitrogen. Fuel temperatures ranged from ambient to 800 K (980 F). Combustor pressure was 6 atmospheres and the inlet air temperature ranged from 589 to 894 K (600 to 1150 F). The NOx emission index increased with fuel temperature at a rate of 4 to 9 percent per 100 K (180 F), depending on the inlet air temperature. The rate of increase in NOx was lowest at the highest inlet air temperature tested.

Marchionna, N. R.↗

NASA NTRS1973

Enriched aluminide coatings for dispersion strengthened nickel materials

Improved aluminide/barrier coating combinations for dispersion strengthened nickel materials were investigated. The barrier materials involved alloys with refractory metal content to limit interdiffusion between the coating and the substrate, thereby minimizing void formation. Improved aluminide coatings involved the dispersion of aluminum-rich compounds. Coatings were tested in argon at 1533 K (2300 F) for 100 hours and in cyclic oxidation at 1422 K (2100 F). Two coatings on TDNiCr completed 300 hours of oxidation testing, none on TDNi. Selected coating combinations were evaluated in Mach 1 burner rig testing using JP-4 fuel and air at 1422 K (2100 F) and 1477 K (2200 F) for 350 and 100 hours, respectively. Static oxidation in 1-hour cycles was conducted at 1533 K (2300 F) for 100 hours. For comparison purposes a physical vapor deposition (PVD) NiCrAlY coating was tested concurrently. Only the NiCrA1Y coating survived the 1477 K (2200 F)/100-hour burner rig test and 275 hours of the 350-hour 1422 K (2100 F) test. Elevated temperature exposure reduced room temperature tensile properties but had little effect on elevated temperature properties.

Levinstein, M. A.↗

NASA NTRS1937

Fuel Spray and Flame Formation in a Compression-Ignition Engine Employing Air Flow

The effects of air flow on fuel spray and flame formation in a high-speed compression-ignition engine have been investigated by means of the NACA combustion apparatus. The process was studied by examining high-speed motion pictures taken at the rate of 2,200 frames a second. The combustion chamber was of the flat-disk type used in previous experiments with this apparatus. The air flow was produced by a rectangular displacer mounted on top of the engine piston. Three fuel-injection nozzles were tested: a 0.020-inch single-orifice nozzle, a 6-orifice nozzle, and a slit nozzle. The air velocity within the combustion chamber was estimated to reach a value of 425 feet a second. The results show that in no case was the form of the fuel spray completely destroyed by the air jet although in some cases the direction of the spray was changed and the spray envelope was carried away by the moving air. The distribution of the fuel in the combustion chamber of a compression-ignition engine can be regulated to some extent by the design of the combustion chamber, by the design of the fuel-injection nozzle, and by the use of air flow.

Rothrock, A M↗

DOE OSTI2017

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗