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At least 91 records · Page 5

Modeling and Simulation of a Nuclear Fuel Element Test Section

"The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

Moran, Robert P.↗

Heavy Duty Hydrogen: Reference Station, Fueling Performance Test Device Concepts, and Station Capacity Model (CRADA Final Report)

This project will provide valuable information on (1) reference station design, (2) exploration of design concepts for a fueling performance test device, and (3) modeling of station capacity. The work applies to heavy-duty (HD) hydrogen fueling stations with large dispensing capacity and high flowrates servicing HD hydrogen trucks such as class 8 trucks in long haul applications. Additionally, a 4th area will provide near-real-time verification of fuel quality with on-site hydrogen contaminant detectors (HCDs) for use at both light-duty (LD) and HD stations. The work leverages national lab capabilities including staff and equipment at SNL, NLR, and ANL with collaboration and funding cost share from California agencies (CEC, SCAQMD, and GO-Biz). This project will provide tools and information that lead to more efficient design, acceptance, and commissioning of these larger capacity, higher flowrate stations serving HD applications. The HCD work will benefit both LD and HD stations. Note: The 4th area (HCD) mentioned above continued outside of this CRADA and will not be included in this report. There are no tasks included in this CRADA for HCD and the summary text for it should have been removed prior to this CRADA contract execution but remained in error and is included here for completeness.

08 HYDROGEN↗

Test Plan for the Second Reirradiation Heating Test of Fuel Particles from AGR 2 Compact 6-4-1

Post irradiation examination (PIE) and high-temperature heating tests of tristructural isotropic (TRISO)-coated fuel particles and compacts from the second Advanced Gas Reactor (AGR) irradiation experiment, AGR 2, are in progress. This test plan describes the reirradiation and subsequent high-temperature testing of loose fuel particles obtained from AGR 2 compact 6-4-1. Four of these particles will have been cracked, gamma counted, and re-irradiated in the Neutron Radiography (NRAD) reactor prior to the heating test. The heating test will involve heating the particles in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of approximately 1600°C while collecting fission products released from the particles for post-heating test analysis. The intention of the test is to assess the retention of fission products in the fuel kernel. Reirradiation of the particles will allow a determination of the retention of short-lived fission products I-131 (t1/2 = 8.02 days) and Xe-133 (t1/2 = 5.2 days) in the kernel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Affordable Development and Optimization of CERMET Fuels for NTP Ground Testing

CERMET fuel materials for Nuclear Thermal Propulsion (NTP) are currently being developed at NASA's Marshall Space Flight Center. The work is part of NASA's Advanced Space Exploration Systems Nuclear Cryogenic Propulsion Stage (NCPS) Project. The goal of the FY12-14 project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of an NTP system. A key enabling technology for an NCPS system is the fabrication of a stable high temperature nuclear fuel form. Although much of the technology was demonstrated during previous programs, there are currently no qualified fuel materials or processes. The work at MSFC is focused on developing critical materials and process technologies for manufacturing robust, full-scale CERMET fuels. Prototypical samples are being fabricated and tested in flowing hot hydrogen to understand processing and performance relationships. As part of this initial demonstration task, a final full scale element test will be performed to validate robust designs. The next phase of the project will focus on continued development and optimization of the fuel materials to enable future ground testing. The purpose of this paper is to provide a detailed overview of the CERMET fuel materials development plan. The overall CERMET fuel development path is shown in Figure 2. The activities begin prior to ATP for a ground reactor or engine system test and include materials and process optimization, hot hydrogen screening, material property testing, and irradiation testing. The goal of the development is to increase the maturity of the fuel form and reduce risk. One of the main accomplishmens of the current AES FY12-14 project was to develop dedicated laboratories at MSFC for the fabrication and testing of full length fuel elements. This capability will enable affordable, near term development and optimization of the CERMET fuels for future ground testing. Figure 2 provides a timeline of the development and optimization tasks for the AES FY15-17 follow on program.

Hickman, Robert R.↗

FCET Solid Oxide Fuel Cell Testing and Development: CRADA 526 [Abstract only]

The purpose of the proposed project is for PNNL to test the performance of prototype solid oxide fuel cells (SOFCs) created by FCET. Such testing will provide FCET with independent performance data that can be communicated to potential clients and/or investors. Additionally, PNNL will collaborate with FCET on design changes to improve fuel cell performance. Intellectual property developed in this way will benefit FCET with improved products to market, and PNNL through royalty revenue. The key technology held by FCET is a process to deposit extremely thin layers of oxide materials, from 10-50 nm in thickness. The range of possible materials that can be deposited with their method is broad, but this project will focus on the yttria-stabilized zirconia (YSZ) electrolytes for SOFCs. Thin, gas tight YSZ membranes have been a long-sought target in SOFC research. The thinner the YSZ, the lower the cell resistance, and the higher performance of the cell (or the lower the operating temperature). 10-50 nm would be a major step change in YSZ thickness from the state of the art. PNNL can team with FCET on future R&D projects and push this technology forward to improve energy efficiency and reduce carbon emissions.

30 DIRECT ENERGY CONVERSION↗

Soda ash removes sulfur from fuels

Test show that adding soda ash (sodium bicarbonate) to coal/oil/water emulsion reduces 75 to 80% of sulfur dioxide gas emitted during subsequent combustion of emulsion.

Dooher, J.↗

Sister Rod Destructive Examinations (FY22) Appendix E: Mechanical Testing - Spent Fuel and Waste Disposition

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Summary of Factory Fueling and Testing Risks to Support the Transformation of Regulatory Requirements for Microreactors

Microreactors have the potential to open new markets and applications for nuclear energy. New manufacturing and operational regimes, such as factory fueling and testing prior to site installation, are being proposed to meet these new markets. The radiological hazards associated with traditional commercial power reactors are not the same as the hazards of a reactor being constructed in a factory. Currently, a reactor operating license specified under 10 CFR Part 50, Part 52, or (in the near future) Part 53 is required prior to loading fuel in the reactor. This is logical for plants that are immediately to begin full-power operations soon after fuel loading. However, for microreactors, this may not be the case. Some deployment concepts may stage reactors as “ready to go” in a factory prior to site installation and startup. This approach is being considered in the case of diesel generator replacements, mobile or roaming mining operations, and emergency response situations, such as after a major disaster event. Therefore, it is necessary to evaluate hazards assuming low power operation for a short irradiation time. This report investigates the primary hazards unique to microreactors associated with low power reactor and physics tests necessary to confirm core safety assumptions, neutronic and core design performance parameters, and other assumptions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Elastomer Swell Behavior in 1-Propanol, Diisobutylene, Cyclopentanone, and a Furan Mixture Blended in E10 and a Blendstock for Oxygenate Blending (BOB)

The compatibility of four potential bio-derived blendstock molecules with infrastructure elastomers was determined by measuring the volume change following exposure. The blendstock molecules included 1-propanol, diisobutylene, cyclopentanone, and a furan mixture. The elastomers included two fluorocarbons, six nitrile rubbers (NBRs), and one each of fluorosilicone, neoprene, polyurethane, and silicone. The elastomers were exposed to the fuel molecules as blends ranging from 0 to 30 vol.% in both a blendstock for oxygenate blending (BOB) formulation and an E10 fuel. Silicone exhibited excessive swelling in each test fuel, while the other elastomers showed good compatibility (low swell) with diisobutylene, 1-propanol, and the furan mixture when BOB was used as the base fuel. The E10 base fuel produced high (>30%) swell in neoprene, polyurethane, and some nitrile rubbers. In most cases diisobutylene produced the least amount of volume expansion. In contrast, the addition of cyclopentanone produced unacceptably high swelling in each elastomer and is not considered suitable for use with these fuels. Analysis of the results showed that the swelling behavior is predominantly due to the polarity of the elastomer and test fuels.

03 NATURAL GAS↗

Dissolution of Material Test Reactor Fuel in an H-Canyon Dissolver

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO 3 flowsheet. Since the development of the existing flow sheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Operation of the 25kW NASA Lewis Research Center Solar Regenerative Fuel Cell Tested Facility

Assembly of the NASA Lewis Research Center(LeRC)Solar Regenerative Fuel Cell (RFC) Testbed Facility has been completed and system testing has proceeded. This facility includes the integration of two 25kW photovoltaic solar cell arrays, a 25kW proton exchange membrane (PEM) electrolysis unit, four 5kW PEM fuel cells, high pressure hydrogen and oxygen storage vessels, high purity water storage containers, and computer monitoring, control and data acquisition.

Solar Regenerative Fuel Cell electrolyzer↗

HIGH BURNUP FUEL-COOLANT INTERACTION ANALYSIS SUPPORTING FUEL SAFETY TESTING AT IDAHO NATIONAL LABORATORY

In the near future, experiments on HBu fuel under loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) conditions will be performed within the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL). These experiments will be performed using the Transient Water Irradiation System for TREAT (TWIST) experiment vehicle. To support these experiments, analysis of fuel-coolant interaction (FCI) energetics is underway. This paper discusses FCIs in the context of light water reactor (LWR) safety, differentiating between the severe accident focus of commercial reactors and experimental RIA test programs where FCIs have occurred. However, it is highlighted that as the nuclear industry aims for increased burnup limits, the FCI events observed in RIA test programs may become relevant to commercial LWR safety analysis. The paper then presents developments to the UW-FCI computer program to enable simulation of FCIs initiated by solid fuel particles dispersing into the coolant during RIAs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Influence of NOx Chemistry on the Prediction of Natural Gas End-Gas Autoignition

Research on the influence of NOx (NO/NO2) chemistry on HC oxidation has shown that NOx plays an important role in the promotion/inhibition of HC autoignition [1-6]. Current data for n-heptane, iso-octane, ethanol, and toluene [3, 4] have shown that NO chemistry inhibits autoignition at low temperatures and concentrations. However, at medium and high temperatures and low concentrations, autoignition is promoted, decreasing the Negative Temperature Coefficient (NTC) effect. This autoignition promotion, however, becomes weaker as NO concentrations are increased. Additionally, fundamental kinetics studies have suggested that NOx chemistry also promotes the autoignition of smaller HC, such as CH4 [7-10], C2H6 [11-14], and C3H8 [15]. Additionally to the chemical kinetics studies, a growing body of evidence supports the importance of NOx chemistry on SI engine simulations [16, 17]. Foong et al. [18] suggested that the addition of NO in their models significantly improved the agreement between simulations and experiments where the onset of knock was advanced due to the presence of NO, and Morganti et al. [19] showed that NO addition in the residual-gas was necessary for their models to obtain good agreement with experimental data. Lastly, Mohr et al. [20, 21] investigated the effect of Natural Gas (NG) reactivity and EGR substitution rate and composition on homogeneous ignition delay, flame speeds, and EGAI for stoichiometric NG/oxidizer/EGR blends. They observed that the addition of EGR composed of only inert species (CO2 and Ar) suppressed homogeneous autoignition in all NG fuels tested under all conditions. However, for all NG fuels tested, increasing EGR rates with reactive species (Ar, CO2, CO, and NO) promoted homogeneous autoignition under all conditions, thus reproducing the same trends for various NG fuel compositions that were observed in other studies and fuels. For this reason, a study is being conducted to analyze the effects of NOx chemistry on the prediction of NG/air/EGR homogeneous autoignition. This study has been divided into two tasks. First, the addition of NOx chemistry to the ARIES82 mechanism [20] is being validated using homogeneous ignition delay data collected by Mohr [20, 21]. Second, once the new modified mechanism with NOx reactions has been validated, it will be employed on engine multi-dimensional modeling presented in [22, 23] to analyze the influence of NOx chemistry on NG EGAI. Results thus far have shown that homogeneous ignition delay calculations are sensitive to NOx chemistry, where the modified mechanisms captured well all trends and closely matched the homogeneous ignition delays observed by Mohr et al. Also, the results in this work suggest that NOx chemistry is necessary to correctly capture the trends in NG homogeneous autoignition when EGR with reactive species is used since the opposite behavior is observed when only inert species are considered for the EGR composition.

03 NATURAL GAS↗

Chemical Kinetics Underlying the Sooting Tendency and Auto-Ignition Characteristics of Linear, Branched, and Cyclic Ether Compounds

Biofuels present opportunities for improving the performance and reducing emissions from internal combustion engines by incorporating oxygenated functional groups to the fuels. Among various oxygenates, ethers have been recognized as promising candidates for an alternative to conventional diesel fuel owing to their higher reactivity and lower sooting tendency. The detailed guidelines for designing ethers, however, have not been fully discussed, even though their combustion characteristics are sensitive to the molecular structure. This study was devoted to exploring the structure-property relationships, particularly focusing on the cetane number and yield sooting index, using five linear, branched, and cyclic ethers: di-amyl-ether, 4-butoxy-heptane, 3,3-dimethyl-oxetane, 2-ethyl-4-methyl-1,3-dioxolane, and 2-isopropyl-4-methyl-1,3-dioxolane. First, we examined the chemical kinetics underlying the sooting tendency of the test fuels. The combustion product distribution was measured from flow reactor experiments at 750-1100 K, F=3, at atmospheric pressure. As a result, it was revealed that the sooting tendency is closely related to the size of hydrocarbon intermediates in the high-temperature regime (>1000 K); that is, larger hydrocarbons lead to more soot precursor formation. The underlying chemistry determining the size of the hydrocarbons from the tested fuels was analyzed using reaction pathway analysis and quantum mechanics calculations, which showed that the branched and cyclic ether structures form abundant C3-C4 compounds. Moreover, the auto-ignition characteristics of the test fuels were studied using the flow reactor at low-temperature (400-700 K) and F=1. We found a clear difference in the combustion-product distribution from high and low reactivity fuels, which was then correlated to the systematic analysis of the key reaction energy barriers with the varying molecular structure.

BIOMASS FUELS,INORGANIC, ORGANIC, PHYSICAL, AND AN↗

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗

Numerical investigation of the pressure-dependence of yield sooting indices for n-alkane and aromatic species

The pressure-dependence of the Yield Sooting Index (YSI), as a measure of sooting tendency, is numerically investigated for pressures ranging from 1 atm to 15 atm. Straight-chain alkanes, including ethane, propane, n-hexane, n-heptane, and n-dodecane, and aromatic species, including benzene, toluene, xylene, and ethylbenzene, are selected as test fuels. A detailed chemical mechanism is used for all test fuels considered, and a flamelet-based YSI simulation methodology is employed, where the YSI values are computed based on the Polycyclic Aromatic Hydrocarbon (PAH) dimer production rate at each pressure. It is found that the maximum flame centerline soot volume fraction increases with increasing pressure for all test compounds. However, the pressure dependence of the maximum soot volume fraction is similar for the hydrocarbons within the same class (n-alkane or aromatics). As a consequence, the YSI values computed at pressures up to 15 atm show only small deviations from their values computed and measured at atmospheric pressure. These results suggest that YSI is robust also at elevated pressures, close to those found in practical combustion devices. This study is entirely based on numerical simulations without any experimental measurements available for direct validation. Experimental YSI measurements at high pressures would be required to fully confirm the conclusions reported in this manuscript. Still, our numerical results provide some guidelines for the design of such experiments in the future.

42 ENGINEERING↗

Thermal Modelling of Advanced Test Reactor Fuel in a Generalized Dry Storage System

Star-CCM+, a computational fluid dynamics (CFD) software was used to conduct modeling and simulation of the thermal performance of a dry storage configuration consisting of Department of Energy Standardized Canisters (DOESCs) loaded with aluminum-clad spent nuclear fuel (ASNF.) The configuration includes nine DOESCs loaded with Advanced Test Reactor ASNF contained within a stainless steel overcanister centered in a ventilated, concrete overpack. The simulations were used to estimate the maximum temperatures reached by backfill gases inside the overcanister and DOESCs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗