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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,765 records · Page 98
All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

ASME Section VIII Recertification of a 33,000 Gallon Vacuum-jacketed LH2 Storage Vessel for Densified Hydrogen Testing at NASA Kennedy Space Center

The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) has been developed at NASA Kennedy Space Center in Florida. GODU-LH2 has three main objectives: zero-loss storage and transfer, liquefaction, and densification of liquid hydrogen. A cryogenic refrigerator has been integrated into an existing, previously certified, 33,000 gallon vacuum-jacketed storage vessel built by Minnesota Valley Engineering in 1991 for the Titan program. The dewar has an inner diameter of 9.5 and a length of 71.5; original design temperature and pressure ranges are -423 F to 100 F and 0 to 95 psig respectively. During densification operations the liquid temperature will be decreased below the normal boiling point by the refrigerator, and consequently the pressure inside the inner vessel will be sub-atmospheric. These new operational conditions rendered the original certification invalid, so an effort was undertaken to recertify the tank to the new pressure and temperature requirements (-12.7 to 95 psig and -433 F to 100 F respectively) per ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. This paper will discuss the unique design, analysis and implementation issues encountered during the vessel recertification process.

Liquefaction↗
NASA NTRS1987

Depth Gauge for Liquids Under High Pressure

Piezoelectric element mounted in hole drilled in high-pressure plug. Transducer used to measure depth of liquid when pressure in vessel high. New configuration transmits ultrasonic vibration directly into liquid, enhancing signal strength, accuracy, and range, yet piezoelectric element protected from high-pressure liquid.

Zuckerwar, A. J.↗
DOE OSTI2016

Thermal-Hydraulic Response of the TREAT Multi-Vessel Static Environment Test Vehicle

As part of the resumption of transient testing in the United States, multiple test vehicles are currently under design, which are ideally suited for achieving various test environments. Of these, the multi- static environment rodlet transient test apparatus (SERTTA) is the most mature design. The test vehicle is designed for testing up to four isolated specimens under a variety of conditions ranging from gas, steam, water, and liquid sodium. During proposed transient events such as reactivity initiated accidents (RIA) and loss of coolant accidents (LOCA) in reactors, prototypic environments dictate a wide range of thermal-hydraulic boundary conditions. The purpose of the this work is to provide the thermal hydraulic response of the Multi-SERTTA test vehicle subjected to fuel specimen energy depositions representative of pressurized water reactor (PWR) conditions. Calculations are performed for a RIA-type power pulse in a 10 cm long UO2-zircaloy rodlet immersed in a water environment in Multi-SERTTA. The relatively small volume available in the Multi-SERTTA causes the vessel pressure to be very sensitive to total energy deposition. The effects of total energy deposition and free gas volume in the pressure vessel are presented. The design provides additional flexibility of environment conditions through use of the expansion tank volume. The results show the influence of the vehicle geometric and safety constraints and provide indication of the expected thermal-hydraulic conditions relative to typical RIA transient events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Application of SPH Factors to PHWR Lattice Homogenization

Pressurized Heavy-Water Reactors (PHWRs) are heavy-water moderated and cooled. They consist of a horizontal, cylindrical, non-pressurized calandria vessel which contains the heavy-water moderator and is penetrated axially by fuel channels. Fuel channels consist of two concentric tubes separated by a gas gap: an inner tube called the pressure tube, and an outer tube called the calandria tube. The pressure tubes hold the fuel bundles, which are cooled by the flow of coolant at high temperature and pressure. The coolant flows in opposite directions in adjacent channels. The gap between the pressure and calandria tubes ensures that the temperature of the moderator remains relatively low, below 80 deg. C. For a CANDU 6 reactor, there are 380 fuel channels, each holding 12 37-element fuel bundles, approximately 10-cm in diameter and 50-cm long each. The fuel pins are arranged in three concentric rings (consisting of 6, 12 and 18 pins respectively), plus a central pin. The distance between channels (lattice pitch) is approximately 28.6 cm. Production core-level neutronics calculations are usually performed in a Cartesian geometry using few group diffusion and a simplified core model, whereby neutronics properties are homogenized over large regions called nodes. For Light Water reactors (LWRs), the two-dimensional axial (x-y) projection of a node corresponds to one fuel assembly. For PHWRs, a node usually consists of one fuel bundle and its associated moderator. For LWRs, advanced core-level calculations based on advanced node-level homogenization methods, such as Generalized Equivalence Theory (GET) or Super-homogenisation (SPH), improve the accuracy of full-core calculations compared to - standard homogenization. For rectangular fuel-pin arrangements, SPH can be used to perform pin-cell level homogenization which allows subsequent core-level diffusion calculations to be performed using fine Cartesian meshes. For PHWRs, the distribution of fuel pins in concentric rings makes the use of Cartesian mesh refinement impractical. However, SPH can still be applied, albeit using a coarser node subdivision, by dividing the heterogeneous node into 3x3 sub-regions in such a way that the central sub-region includes the entire fuel channel. This work investigates the use of SPH homogenization using a 3x3 subdivision of the PHWR node. Preliminary results show that the use of SPH factors does not yield more accurate results than standard homogenization when applied to PHWR nodes using a 3x3 subdivision. The explanation appears to lie in the fact that SPH factors are dependent on node-boundary leakage. Future investigations to confirm the leakage-dependence of SPH factors are recommended. Accuracy improvements may also be achievable by using smaller SPH regions at the periphery of the PHWR node, in an attempt to reduce the SPH factors' dependence on node-boundary leakage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1991

Seal testing of large diameter rocket motors

This investigative program examines leakage testing of elastomeric O-ring seals for a solid rocket casing and provides direction towards an improved nondestructive postassembly test. It also details test equipment for the Space Shuttle systems solid rocket boosters (SRB). The results are useful to designers of hardware for pressure containment vessels which use O-ring seals. Using several subscale seal and groove configuration test fixtures equipped with either two or three O-ring seals in series, seal integrity is investigated with both a pressure decay and flowmeter methods. Both types of test equipment adequately detect the practical range of expected seal leak rates of 1 to 0.0001 sccs. The flowmeter leak test equipment appears to reduce testing time substantially. Limited seal leakage testing is performed on full-sized rocket motor segment seals, a pre-Challenger short stack, providing comparison of bore seals to test specimen bore and face seals. The conclusions are that seal reliability, verified via a performance pressure test, can be affected by temperature, quantity of grease, test pressure, and seal pressure load direction. Potential seal failure scenarios including contamination, seal damage, and sealing surface damage are discussed. Recommendations are made for an improved test procedure.

Moore, N. B.↗
NASA NTRS1986

Ultrasonic level sensors for liquids under high pressure

An ultrasonic level sensor of novel design continuously measures the level of a liquid subjected to a high pressure (up to about 40 MPa), as is sometimes required for the effective transfer of the liquid. The sensor operates as a composite resonator fabricated from a standard high-pressure plug. A flat-bottom hole is machined into the plug along its center line. An ultrasonic transducer is bonded rigidly to the interior surface of the bottom wall, while the exterior surface is in contact with the liquid. Although the bottom wall is designed to satisfy the pressure code, it is still sufficiently thin to permit ready excitation of the axisymmetric plate modes of vibration. The liquid level is measured by a conventional pulse-echo technique. A prototype sensor was tested successfully in a 2300-l water vessel at pressures up to about 37 MPa. A spectral analysis of the transmitted pulse reveals that the flexural, extensional, thickness-shear, and radial plate modes are excited into vibration, but none of these appears to be significantly affected by the pressurization of the liquid.

Zuckerwar, A. J.↗
DOE OSTI2017

UT-GOM2-1 Hydrate Pressure Coring Expedition Volume

From 2-May-2017 to 22-May-2017, the UT-GOM2-1 Hydrate Pressure Coring Expedition drilled two wells in Green Canyon Block 955 (GC 955) in the deepwater Gulf of Mexico: Hole GC 955 H002 (H002) and Hole GC 955 H005 (H005). 21 10 ft (3.05 m) pressure cores were attempted in and near the methane hydrate reservoir. In the first hole, H002, 1 of the 8 cores were recovered under pressure and there was 34% recovery of sediment (both pressurized and depressurized). In the second hole, H005, 12 of the 13 cores were recovered under pressure and there was 72% recovery of sediment. The pressure cores were imaged and logged under pressure. Samples were quantitatively degassed either on-board or on-shore to determine the hydrate concentration and the gas composition. Pore water analyses were performed on depressurized samples, and sediment samples were collected to enable characterization of the microbial community. 21 3.3 ft (1 m) vessels containing pressure core sections were returned to the University of Texas for storage, distribution, and further analysis. These cores will provide a foundation for scientific exploration by the greater hydrate research community.

03 NATURAL GAS↗
DOE OSTI2018

UT-GOM2-1 Hydrate Pressure Coring Expedition Summary (Chapter 1. Expedition Summary)

The UT-GOM2-1 Hydrate Pressure Coring Expedition drilled two wells in Green Canyon Block 955 (GC 955) in the deepwater Gulf of Mexico: Hole GC 955 H002 (H002) and Hole GC 955 H005 (H005). 21 10 ft (3.05 m) pressure cores were attempted in and near the methane hydrate reservoir. In the first hole, H002, 1 of the 8 cores were recovered under pressure and there was 34% recovery of sediment (both pressurized and depressurized). In the second hole, H005, 12 of the 13 cores were recovered under pressure and there was 72% recovery of sediment. The pressure cores were imaged and logged under pressure. Samples were quantitatively degassed either on-board or on-shore to determine the hydrate concentration and the gas composition. Pore water analyses were performed on depressurized samples, and sediment samples were collected to enable characterization of the microbial community. 21 3.3 ft (1 m) vessels containing pressure core sections were returned to the University of Texas for storage, distribution, and further analysis. These cores will provide a foundation for scientific exploration by the greater hydrate research community. This is the initial expedition summary (preliminary expedition report).

03 NATURAL GAS↗
NASA NTRS1982

Effect of prolonged hypokinesia on resistance of resistive vessels in rats

Under the effect of prolonged hypokinesia, the perfusion pressure in resistive vessels, measured under conditions of deep anesthesia and complete denervation, increased by approximately the same degree as arterial pressure in non-anesthetized animals. The increase in arterial, perfusion pressure and the resistance of resistive vessels in animals subjected to prolonged hypokinesia was accompanied by an increase in adrenoreactivity. During prolonged hypokinesia, partial obliteration of the vascular bed of the skeletal muscles plays a significant role in the observed increase in resistance of vessels of the extremities. The increase in adrenoreactivity of the vessels during hypokinesia may be realized as a partial case of an increase in the adrenoreactivity of structures whose innervation is disturbed.

Saltykova, V. A.↗
DOE OSTI2016

Status Report on Ex-Vessel Coolability and Water Management

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Benchmarking of Probabilistic Fracture Mechanics Models in Grizzly

In 2020, as part of contract/task order 31310019N0006/31310020F0060, the Nuclear Regulatory Commission tasked Idaho National Laboratory (INL) with benchmarking the MOOSE/GRIZZLY code against the NRC’s Fracture Analysis of Vessels - Oak Ridge (FAVOR), version 16.1 code. Both codes can predict large light-water reactor vessel integrity deterministically and probabilistically. In fact, such predictions are the primary purpose of the FAVOR code, and are a main focus of the GRIZZLY code, although GRIZZLY is capable of modeling a wider range of geometries and problems. The vessel integrity modeling capabilities of the GRIZZLY code are still under development and being gradually expanded, and the capability to model vessel fracture by crack growth has not yet been integrated into GRIZZLY (although it does exist in pre-release versions). Thus, the present benchmark study focused on comparing predictions for the conditional probability of crack growth initiation (CPI) between GRIZZLY and FAVOR v16.1. The present benchmarking study expands on a previous study completed in 2020, which focused on single plate regions, various flaw types, and full vessel studies. The previous study was limited in scope and revealed some relatively large discrepancies, particularly in the case of weld regions. In the present study, detailed investigations into the sources of discrepancies between GRIZZLY and FAVOR were performed, with a focus on weld regions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Quantifying liquid boundary and vapor distributions in a fuel spray by rainbow schlieren deflectometry

The rainbow schlieren deflectometry (RSD) technique is used to determine the liquid boundary and the fuel volume fraction distributions in the vapor region of a high-pressure fuel spray. Experiments were conducted in a constant pressure flow vessel, whereby a customized single-hole common-rail diesel injector is used to introduce n-heptane fuel into a coflow of low-speed ambient air at two different test conditions. Only the quasi-steady period of the fuel spray is considered, and multiple injections are performed to acquire statistically significant data at an image acquisition rate of 20 kHz. An algorithm to identify the liquid boundary using intensity recorded by the RSD images is presented. The results are compared against measurements obtained by the Mie scattering technique. Our results demonstrate that the RSD can be a powerful optical diagnostics technique to simultaneously quantify both the vapor and liquid regions in the high-pressure fuel sprays.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Benchmarking of ENDF/B-VIII.0 cross sections for LWR ex-core transport applications

ENDF/B-VIII.0 cross section data is benchmarked on a simple Monte-Carlo model representative of the ex-core geometry of a PWR used for reactor vessel neutron fluence calculations. The benchmarking indicates that the ENDF/B-VIII.0 results show large differences as compared to previous evaluations, which have been extensively compared to measured data. This exercise suggests that ENDF/B-VIII.0 nuclear data is not suitable for ex-core LWR radiation transport applications. Evaluators are aware of the problems with the ENDF/B-VIII.0 evaluation of iron and work is underway to improve the evaluations for an upcoming release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Monte Carlo N-Particle Tracking of Ultrafine Particle Flow in Bent Micro-Tubes

The problem of large pressure-differential driven laminar convective-diffusive ultrafine aerosol flow through bent micro-tubes is of interest in several contemporary research areas including; release of contents from pressurized containment vessels, aerosol sampling equipment, advanced scientific instruments, gas-phase micro-heat exchangers, and microfluidic devices. In each of these areas, the predominant problem is the determination of the fraction of particles entering the micro-tube that is deposited within the tube and the fraction that is transmitted through. Due to the extensive parameter restrictions of this class of problems, a Lagrangian particle tracking method making use of the coupling of the analytical stream line solutions of Dean and the simplified Langevin equation is quite a useful tool in problem characterization. This method is a direct analog to the Monte Carlo N-Particle method of particle transport extensively used in nuclear physics and engineering. In this work, 10 nm diameter particles with a density of 1 g/cm3 are tracked within micro-tubes with toroidal bends with pressure differentials ranging between 0.2175 and 0.87 atmospheres. The tubes have radii of 25 microns and 50 microns and the radius of curvature is between 1 m and 0.3183 cm. The carrier gas is helium, and temperatures of 298 K and 558 K are considered. Numerical convergence is considered as a function of time step size and of the number of particles per simulation. Particle transmission rates and deposition patterns within the bent micro-tubes are calculated.

Casella, Andrew M.↗
DOE OSTI2016

Production Facility Prototype Blower Installation Report with 1000 Hr Test Results

The roots blower in use at ANL for in-beam experiments and also at LANL for flow tests was sized for 12 mm diameter disks and significantly less beam heating. Currently, the disks are 29 mm in diameter, with a 12 mm FWHM Gaussian beam spot at 42 MeV and 2.86 μA on each side of the target, 5.72 μA total. The target design itself is reported elsewhere. With the increased beam heating, the helium flow requirement increased so that a larger blower was needed for a mass flow rate of 400 g/s at 2.76 MPa (400 psig). An Aerzen GM 12.4 blower was selected, and is currently being installed at the LANL facility for target and component flow testing. This report describes this blower/motor/pressure vessel package and the status of the facility preparations. The blower has been operated for 1000 hours as a preliminary investigation of long-term performance, operation and possible maintenance issues. The blower performed well, with no significant change in blower head or mass flow rate developed under the operating conditions. Upon inspection, some oil had leaked out of the shaft seal of the blower. The shaft seal and bearing race have been replaced.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Production Facility Prototype Blower Installation Report with 1000 Hour Test Results

The roots blower in use at ANL for in-beam experiments and also at LANL for flow tests was sized for 12 mm diameter disks and significantly less beam heating. Currently, the disks are 29 mm in diameter, with a 12 mm FWHM Gaussian beam spot at 42 MeV and 2.86 μA on each side of the target, 5.72 μA total. The target design itself is reported elsewhere. With the increased beam heating, the helium flow requirement increased so that a larger blower was needed for a mass flow rate of 400 g/s at 2.76 MPa (400 psig). An Aerzen GM 12.4 blower was selected, and is currently being installed at the LANL facility for target and component flow testing. This report describes this blower/motor/pressure vessel package and the status of the facility preparations. The blower has been operated for 1000 hours as a preliminary investigation of long term performance, operation and possible maintenance issues. The blower performed well, with no significant change in blower head or mass flow rate developed under the operating conditions. Upon inspection, some oil had leaked out of the shaft seal of the blower. The shaft seal and bearing race have been replaced. Test results and conclusions are reported.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

First Step in Validating the Radiological Inventories of the Current French Nuclear Fleet - 19000

To anticipate the evolution of the mechanical characteristics of its PWR vessels, EDF developed a vessels fluence monitoring program. This program was based on vessel samples and neutron dosimeters deployed between the fuel assemblies and vessels. These samples and dosimeters are regularly analyzed in laboratories. To calculate the radioactive inventories, we have developed a calculation scheme which includes a 3D mapping of the neutron flux, the calculation of the radiological inventories of the structures of interest, and finally the validation of the results. The validation needs samples to propose calculation/measurement ratios. Unfortunately before the final shut down, the internal and vessel samplings are not possible. To anticipate the decommissioning of the current French nuclear fleet, this paper shows how the samples of the vessels fluence monitoring program can be used to validate the calculated radiological inventories. Using state-of-the-art simulations (i.e. 3D geometric model, 3D neutron sources, 3D neutron propagation, operational fuel loading coupled with effective power distributions) we calculated the radiological inventories of some vessel fluence monitoring program samples. The comparisons between calculated and measured results were done on 'standard chemical elements'. To qualify a chemical element as 'standard', many criteria must be satisfied such as a high concentration level, a low uncertainty about concentration, activation into an expanded cross section range, production of radionuclides having significant half-lives and finally no difficulties measuring the radionuclides produced. The retained 'standard' is Fe which produces, under neutrons flux, Fe-55. The comparisons show that the calculations slightly overestimate the measurements. (authors)

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

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗