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At least 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
NASA NTRS1963

Introduction to Nuclear Propulsion: Lecture 15 - Nuclear Test Operations

The test operation of nuclear power plants, specifically nuclear rockets, bears some interesting similarities to the operation of chemical rocket tests as well as, of course, many differences. A significant feature common to both nuclear and chemical rocket tests is that all the fuel for the entire operation is loaded at the start of the test. As a direct consequence of this fact, the operation of nuclear power plants must be surrounded with adequate safety precautions, as is indeed the case in the operation of chemical rockets, A second direct consequence is that in both types of testing a very thorough and complete checkout is made before starting the test.

TEST METHOD↗
DOE OSTI2019

Development of Waste Disposal Strategy for RH LLW Stored in the ATR Canal - 19453

The Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), is 250 megawatt research reactor designed and operated for the United States Department of Energy for irradiation testing of advanced nuclear fuels and material samples. The pressurized light-water moderated and cooled reactor achieved criticality in 1967. It is designed to allow high neutron flux irradiation of structural materials and nuclear fuels to replicate long-term exposure of materials in traditional lower-flux reactor cores. The ATR also produces rare isotopes for the use in the medical and other industries. The ATR is one of only four reactors in the world with this high-flux irradiation capability. Adjacent to the ATR is a water-filled canal which is used to handle casks, prepare experiments, perform irradiation experiments, and store irradiated components generated from reactor operations. The canal is used to store irradiated (activated) metals that are subsequently determined to be waste. In many instances, these items are too radioactive to remove from the canal and remain in temporary storage until a sufficient decay period has elapsed to where they can be safely removed. The more highly-radioactive remote-handled (RH) waste items requiring hundreds of years to decay have been in storage in the canal since the first ATR core internal changeout (CIC) which was conducted in the 1970's. The INL has recently completed construction and startup of a new Remote-Handled Low-Level Waste (RH LLW) Disposal Facility, which will serve as the final disposal capability for legacy RH (greater than 200 mR/hr on contact) metal waste currently stored in the canal. The ATR will need to develop a method for packaging, characterizing, removing, and shipping this material to the new disposal facility. Based on the configuration of the new disposal cells at the RH LLW Disposal Facility, the ATR must produce an entire new process for handing and packaging the legacy RH waste and preparing it for transport and final disposal. The ATR Canal RH LLW Cleanup Project scope is to develop the processes, tools, and equipment for safely sizing, packaging, and removing RH LLW from the canal. This paper presents highlights and challenges the project faces to develop and implement this capability: - Designing new tools, fabricating new equipment, and implementing new processes at an active Hazard Category-I nuclear facility. - Ensuring new equipment and tools are compatible with the existing equipment and processes at both the generator and the disposal facilities. - Sizing legacy waste under water using remote tools and equipment to minimize void space in waste disposal canisters while ensuring as low as reasonably achievable (ALARA) goals are met. - Loading waste containers underwater and preparing them to meet the liquid-free criteria of the disposal facility waste acceptance criteria (WAC). - Providing a new shielded transport container to transfer waste from ATR to the new on-site RH LLW Disposal Facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix C: Rod Internal Pressure, Void Volume, and Gas Transmission Tests

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 [C-1]. The SNF rods, called the sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) 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. Both fuel rods and empty cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the postirradiation 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 consolidation activities. This appendix documents the status of the ORNL Phase 1 DE activities [C-2, C-3] related to rod internal pressure and void volume measurement techniques, fission gas stack flow measurements applied to selected sister rods, and fission gas release calculations in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix C: Rod Internal Pressure, Void Volume, and Gas Transmission Tests

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 the sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) 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. Both fuel rods and empty cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the postirradiation 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 consolidation activities. This appendix documents the status of the ORNL Phase 1 DE activities related to rod internal pressure and void volume measurement techniques, fission gas stack flow measurements applied to selected sister rods, and fission gas release calculations in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synchrotron micro-computed tomography analysis of neutron-irradiated U-Mo fuel

The three-dimensional (3D) microstructure of neutron-irradiated uranium-10 wt.% molybdenum (U-10Mo) fuel with a burn-up of 9.8 × 10 21 fissions/cm 3 was characterized using a novel, multi-modal synchrotron micro-computed tomography approach combining propagation-based phase-contrast enhanced and absorption contrast techniques. The porosity development, porosity interconnectedness, swelling, composition, local thickness of the zirconium (Zr) diffusion barrier, and the influence of the fuel–cladding interaction on the local composition and pore morphology, were uniquely determined in 3D. Two cuboids were produced using a focused ion beam-scanning electron microscope at the Zr diffusion barrier–fuel interface and in the bulk fuel. The bulk fuel sample swelled by 53.3 [+9.7/−3.1]%, while the fuel near the Zr–fuel interface swelled by 63.3 [+14.7/−7.2]%. The average local thickness of the Zr diffusion barrier decreased by 53 %, compared to the expected pre-irradiated thickness. Four pore morphology regions were identified initiating parallel to the fuel–Zr interaction region: (1) an interaction layer of suppressed porosity, (2) a layer of elongated and interconnected porosity, (3) a transition zone of low porosity, and (4) a layer of unoriented porosity representative of the bulk fuel behavior. The increase in porosity near the diffusion barrier corresponded to a higher U concentration compared to that in the bulk fuel. The interconnected porosity in the fuel near the diffusion barrier was extensive and oriented parallel to the diffusion barrier, while the bulk fuel had more compact and isolated pore networks. The interaction layer, despite having suppressed porosity, was nearly 100 wt.% U. Porosity suppression at the diffusion barrier corresponds to the expected reduction in radiation-driven diffusion of Xe at the interface despite the anticipated increase in fission product nucleation originating from a higher U concentration. In conclusion, the novel 3D insights of the porosity, swelling, and compositional variations characterized herein can improve the fidelity of fuel performance codes for proliferation-resistant fuels for research and test reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

Full-scale altitude engine test of a turbofan exhaust-gas-forced mixer to reduce thrust specific fuel consumption

The specific fuel consumption of a low-bypass-ratio, confluent-flow, turbofan engine was measured with and without a mixer installed. Tests were conducted for flight Mach numbers from 0.3 to 1.4 and altitudes from 10,670 to 14,630 meters (35,000 to 48,000 ft) for core-stream-to-fan-stream temperature ratios of 2.0 and 2.5 and mixing-length-to-diameter ratios of 0.95 and 1.74. For these test conditions, the reduction in specific fuel consumption varied from 2.5 percent to 4.0 percent. Pressure loss measurements as well as temperature and pressure surveys at the mixer inlet, the mixer exit, and the nozzle inlet were made.

Cullom, R. R.↗
DOE OSTI2022

Advanced Stirling Power Generation System for Combined Heat and Power

The goal of this project was the development of an advanced Free-Piston Stirling Engine (FPSE) power generation for combined heat and power application. The innovative approach used in this project was the use of Additively Manufactured (AM) for critical components of an FPSE. A Stirling convertor consists of three major components/subassemblies, 1) Stirling engine, 2) linear alternator, and 3) gas fuel burner. The primary focus of this project was the design of the FPSE. This engine was integrated with an ITC/Qnergy 1 kW linear alternator. The integrated convertor was then coupled to a gas burner.

09 BIOMASS FUELS↗
NASA NTRS1946

The Knock-Limited Performance of Fuel Blends Containing Aromatics V : N-Propylbenzene, N-Butylbenzene, Isobutylbenzene, M-Xylene, and 1-Isopropyl-4-Methylbenzene

Results are reported of knock-limited tests of five aromatics, each individually blended with selected base fuels and tested with and without TEL, using 17.6, F-4, and F-3 small-scale engines. The five aromatics rated in the following order of decreasing antiknock effectiveness at fuel/air ratio 0.10: m-xylene, 1-isopropyl-4-methylbenzene, n-propylbenzene, isobutylbenzene, and n-butylbenzene.

FUELS - ISOPROPYLMETHYLBENZENE↗
NASA NTRS2004

HSCT Sector Combustor Evaluations for Demonstration Engine

In LET Task 10, critical development issues of the HSCT lean-burn low emissions combustor were addressed with a range of engineering tools. Laser diagnostics and CFD analysis were applied to develop a clearer understanding of the fuel-air premixing process and premixed combustion. Subcomponent tests evaluated the emissions and operability performance of the fuel-air premixers. Sector combustor tests evaluated the performance of the integrated combustor system. A 3-cup sector was designed and procured for laser diagnostics studies at NASA Glenn. The results of these efforts supported the earlier selection of the Cyclone Swirler as the pilot stage premixer and the IMFH (Integrated Mixer Flame Holder) tube as the main stage premixer of the LPP combustor. In the combustor system preliminary design subtask, initial efforts to transform the sector combustor design into a practical subscale engine combustor met with significant challenges. Concerns about the durability of a stepped combustor dome and the need for a removable fuel injection system resulted in the invention and refinement of the MRA (Multistage Radial Axial) combustor system in 1994. The MRA combustor was selected for the HSR Phase II LPP subscale combustor testing in the CPC Program.

Greenfield, Stuart↗
DOE OSTI2016

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1993

Thermal-fluid analysis of the fill and drain operations of a cryrogenic fuel tank

The Generic Research Cryogenic Tank was designed to establish techniques for testing and analyzing the behavior of reusable fuel tank structures subjected to cryogenic fuels and aerodynamic heating. The Generic Research Cryogenic Tank tests will consist of filling a pressure vessel to a prescribed fill level, waiting for steady-state conditions, then draining the liquid while heating the external surface to simulate the thermal environment associated with hypersonic flight. Initial tests of the Generic Research Cryogenic Tank will use liquid nitrogen with future tests requiring liquid hydrogen. Two-dimensional finite-difference thermal-fluid models were developed for analyzing the behavior of the Generic Research Cryogenic Tank during fill and drain operations. The development and results of the two-dimensional fill and drain models, using liquid nitrogen, are provided, along with results and discussion on extrapolating the model results to the operation of the full-size Generic Research Cryogenic Tank. These numerical models provided a means to predict the behavior of the Generic Research Cryogenic Tank during testing and to define the requirements for the Generic Research Cryogenic Tank support systems such as vent, drain, pressurization, and instrumentation systems. In addition, the fill model provided insight into the unexpected role of circumferential conduction in cooling the Generic Research Cryogenic Tank pressure vessel during fill operations.

Stephens, Craig A.↗
DOE OSTI2024

TEM Characterization of Neutron Irradiated HfAl3-Al Composite Specimens

Particles comprised of a thermal neutron absorbing material (HfAl3) are dispersed in a metal matrix material with high thermal conductivity (aluminum) to conduct the heat generated by neutron capture away from the fuel and materials. This metal matrix composite is very promising for use as a conduction-cooled neutron absorber and has the potential to be useful as a shroud or heat sink for testing advanced fast reactor fuels and materials in an existing thermal reactor. To design and optimize the absorber block system for advanced reactor designs, fundamental understanding of the irradiation effect on material properties is necessary. This dataset contains TEM characterization results of neutron irradiated HfAl3-Al composite specimens. It is focused on the irradiation induced defects (dislocation lines and loops) characterization using the on-zone axis bright field STEM technique. This data was collected using a FEI Tecnai G2 F30 S/TEM at the Microscopy and Characterization Suite (MaCS), Center for Advanced Energy Studies (CAES).

Guillen, Donna↗
DOE OSTI2024

TEM Characterization of Neutron Irradiated HfAl3-Al Composite Specimens

Particles comprised of a thermal neutron absorbing material (HfAl3) are dispersed in a metal matrix material with high thermal conductivity (aluminum) to conduct the heat generated by neutron capture away from the fuel and materials. This metal matrix composite is very promising for use as a conduction-cooled neutron absorber and has the potential to be useful as a shroud or heat sink for testing advanced fast reactor fuels and materials in an existing thermal reactor. To design and optimize the absorber block system for advanced reactor designs, fundamental understanding of the irradiation effect on material properties is necessary. This dataset contains TEM characterization results of neutron irradiated HfAl3-Al composite specimens. It is focused on the irradiation induced defects (dislocation lines and loops) characterization using the on-zone axis bright field STEM technique. This data was collected using a FEI Tecnai G2 F30 S/TEM at the Microscopy and Characterization Suite (MaCS), Center for Advanced Energy Studies (CAES).

Guillen, Donna↗
NASA NTRS1931

The Design and Development of an Automatic Injection Valve with an Annular Orifice of Varying Area

The purpose of this investigation was to provide an automatic injection valve of simple construction which would produce a finely atomized oil spray of broad cone angle and would fulfill the requirements of fuel injection in aircraft oil engines. The injection valve designed has only six parts - i. e., two concentric nozzle tubes flared at one end, two body parts, and two nuts. Analysis and engine tests indicate that the fuel spray from this type of injection valve has characteristics which reduce the time lag of autoignition and promote efficient combustion in high-speed oil engines.

Joachim, William F↗
DOE OSTI2016

The PROMETRA program: plane strain tests on fresh and highly irradiated zircaloy-4, ZIRLO and M5 fuel claddings

In the framework of the PROMETRA program, clad hardening constitutive laws (stress-strain curves) have been already implemented in the SCANAIR code. These laws were derived from uniaxial hoop tensile tests performed on narrow machined ring samples. However, they do not reproduce the conditions of strain multi-axiality acting on cladding during a RIA (Reactivity-Initiated Accident) in a PWR. As a consequence, a new specimen called Plane Strain Tensile (PST) similar to the annular sample developed by Pennsylvania State University has been developed at CEA on a wider notched ring. It aims to accurately predict the cladding failure in more prototypical conditions, close to plane strain hoop tension in the middle of the gage section. This paper details the PST tests implementation and results in the framework of the PROMETRA program on fresh and irradiated claddings. The development of this test has led also to the development of a technique to measure local deformations in the gage section of the PST specimen using 2-D Digital Image Correlation (DIC). Tests performed on fresh Zircaloy-4 and on irradiated Zircaloy-4, M5 and ZIRLO specimens are analyzed relying of post-test examinations and finite element simulations. In the case of tests performed on irradiated Zry-4 and ZIRLO at 280 and 350 Celsius degrees, the presence of a hydride rim favors the crack nucleation on the outer surface. In this case, rupture nucleation occurs preferentially in the vicinity of notches. Such tests are not representative of the rupture of a cladding subjected to plane strain loading. Nevertheless, they give a good representation of the initiation of rupture observed during RIA on irradiated cladding, resulting in a hydride damage. Zry-4 specimens tested at 480 Celsius degrees and M5TM at 280 and 350 C. degrees result in a ductile fracture nucleation in the center of the specimen. This rupture is prototypical of a ductile rupture of an irradiated specimen subjected to a plane strain loading. However, the nucleation on inner surface by shear at the mandrel contact is not prototypical of a clad rupture during the PCMI (Pellet Clad Mechanical Interaction) phase of a RIA (membrane stresses in the clad thickness in RIA). An important lesson learned from this experimental approach is that the PST test is a particularly complex test, both in its implementation and interpretation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗