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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
NASA NTRS1992

Investigation of advanced counterrotation blade configuration concepts for high speed turboprop systems. Task 4: Advanced fan section aerodynamic analysis

The purpose of this study is the development of a three-dimensional Euler/Navier-Stokes flow analysis for fan section/engine geometries containing multiple blade rows and multiple spanwise flow splitters. An existing procedure developed by Dr. J. J. Adamczyk and associates and the NASA Lewis Research Center was modified to accept multiple spanwise splitter geometries and simulate engine core conditions. The procedure was also modified to allow coarse parallelization of the solution algorithm. This document is a final report outlining the development and techniques used in the procedure. The numerical solution is based upon a finite volume technique with a four stage Runge-Kutta time marching procedure. Numerical dissipation is used to gain solution stability but is reduced in viscous dominated flow regions. Local time stepping and implicit residual smoothing are used to increase the rate of convergence. Multiple blade row solutions are based upon the average-passage system of equations. The numerical solutions are performed on an H-type grid system, with meshes being generated by the system (TIGG3D) developed earlier under this contract. The grid generation scheme meets the average-passage requirement of maintaining a common axisymmetric mesh for each blade row grid. The analysis was run on several geometry configurations ranging from one to five blade rows and from one to four radial flow splitters. Pure internal flow solutions were obtained as well as solutions with flow about the cowl/nacelle and various engine core flow conditions. The efficiency of the solution procedure was shown to be the same as the original analysis.

Crook, Andrew J.↗
NASA NTRS1992

Investigation of advanced counterrotation blade configuration concepts for high speed turboprop systems. Task 4: Advanced fan section aerodynamic analysis computer program user's manual

The computer program user's manual for the ADPACAPES (Advanced Ducted Propfan Analysis Code-Average Passage Engine Simulation) program is included. The objective of the computer program is development of a three-dimensional Euler/Navier-Stokes flow analysis for fan section/engine geometries containing multiple blade rows and multiple spanwise flow splitters. An existing procedure developed by Dr. J. J. Adamczyk and associates at the NASA Lewis Research Center was modified to accept multiple spanwise splitter geometries and simulate engine core conditions. The numerical solution is based upon a finite volume technique with a four stage Runge-Kutta time marching procedure. Multiple blade row solutions are based upon the average-passage system of equations. The numerical solutions are performed on an H-type grid system, with meshes meeting the requirement of maintaining a common axisymmetric mesh for each blade row grid. The analysis was run on several geometry configurations ranging from one to five blade rows and from one to four radial flow splitters. The efficiency of the solution procedure was shown to be the same as the original analysis.

Crook, Andrew J.↗
NASA NTRS2003

On the Determination of Ejecta Structure and Explosion Asymmetry from the X-ray Knots of Cassiopeia A

We present a detailed analysis of Chandra X-ray spectra from individual ejecta knots in the supernova remnant Cassiopeia A. The spectra are fitted to give the electron temperature T(sub e), and (single) ionization age n(sub e)t. These quantities are compared with the predictions of self similar hydrodynamic models incorporating time dependent ionization and radiation losses, and Coulomb electron-ion equilibration behind the reverse shock, for a variety of different ejecta density profiles described by a uniform density core and a power law envelope. We find that the ejecta close to the 'jet' region in the NE, but not actually in the jet itself, have a systematically shallower outer envelope than ejecta elsewhere in the remnant, and we interpret this as being due to more energy of the initial explosion being directed in this polar direction as opposed to equatorially. The degree of asymmetry we infer is at the low end of that generally modelled in asymmetric core-collapse simulations, and may be used to rule out highly asymmetric explosion models.

Laming, J. Martin↗
NASA NTRS2007

Initial Testing of the Stainless Steel NaK-Cooled Circuit (SNaKC)

An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK) was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around the 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. This presentation addresses the construction, fill and initial testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Garber, Anne↗
NASA NTRS2007

Stainless Steel NaK-Cooled Circuit (SNaKC) Fabrication and Assembly

An actively pumped Stainless Steel NaK Circuit (SNaKC) has been designed and fabricated by the Early Flight Fission Test Facility (EFF-TF) team at NASA's Marshall Space Flight Center. This circuit uses the eutectic mixture of sodium and potassium (NaK) as the working fluid building upon the experience and accomplishments of the SNAP reactor program from the late 1960's The SNaKC enables valuable experience and liquid metal test capability to be gained toward the goal of designing and building an affordable surface power reactor. The basic circuit components include a simulated reactor core a NaK to gas heat exchanger, an electromagnetic (EM) liquid metal pump, a liquid metal flow meter, an expansion reservoir and a drain/fill reservoir To maintain an oxygen free environment in the presence of NaK, an argon system is utilized. A helium and nitrogen system are utilized for core, pump, and heat exchanger operation. An additional rest section is available to enable special component testing m an elevated temperature actively pumped liquid metal environment. This paper summarizes the physical build of the SNaKC the gas and pressurization systems, vacuum systems, as well as instrumentation and control methods.

Godfroy, Thomas J.↗
NASA NTRS2021

Dynamics associated with the Corer on M2020 Perseverance Rover

This paper will report on the development of the Dynamics specifications for the Corer on Mars 2020. The corer is a critical piece of the M2020 mission, leading into the Mars Sample Return campaign. However, the configuration of the rover places it on the turret of the Robotic Arm assembly, and near to several science sensors (PIXL, SHERLOC) which are also mounted to the Turret. This paper will provide some similarities and differences to the MSL drill. Details in the development, evolution, and verification of the Dynamics specifications will be described in detail, including the hardware which had to meet these challenging specifications. In addition, the special testing venues used throughout the development & verification of M2020 (EDU corer, EM actuators, HALT testing [simulation of coring dynamics], and QMDT) will be described.

Chrystal, Kyle↗
NASA NTRS2021

Simulation of Landing and Take-Off Noise for Supersonic Transport Aircraft at a Conceptual Design Fidelity Level

The German Aerospace Center has launched an internal project to assess the noise impact associated with supersonic transport aircraft during approach and departure. A dedicated simulation process is established to cover all relevant disciplines, i.e., aircraft and engine design, engine installation effects, flight simulation, and system noise prediction. The core of the simulation process is comprised of methods at the complexity and fidelity level of conceptual aircraft design, i.e., typical overall aircraft design methods and a semi-empirical approach for the noise modeling. Dedicated interfaces allow to process data from high fidelity simulation that will support or even replace initial low fidelity results in the long run. All of the results shown and discussed in this study are limited to the fidelity level of conceptual design. The application of the simulation process to the NASA 55t Supersonic Technology Concept Aeroplane, i.e., based on non-proprietary data for this vehicle, yields similar noise level predictions when compared to the published NASA results. This is used as an initial feasibility check of the new process and confirms the underlying methods and models. Such an initial verification of the process is understood as an essential step due to the lack of available noise data for supersonic transport aircraft in general. The advantageous effect of engine noise shielding on the resulting system noise is demonstrated based on predicted level time histories and certification noise levels. After this initial verification, the process is applied to evaluate a conceptual supersonic transport design based on a PhD thesis with two engines mounted under the wing, which is referred to as aircraft TWO. Full access to this vehicle’s design and performance data allows to investigate the influence of flight procedures on the resulting noise impact along approach and departure. These noise results are then assembled according to proposed Federal Aviation Agency regulations in their Notice of Proposed Rulemaking, e.g., speed limitations, for Supersonic transport noise certification and the regulations from Noise Chapters of the Annex 16 from the International Civil Aviation Organization in order to evaluate the resulting levels as a function of the flight procedure.

noise↗
NASA NTRS2021

I’ll have the Porter: Interactively Visualizing the Results of Statistical Maneuver Analysis

Mission design and navigation relies on statistical maneuver analysis and Monte Carlo simulations when evaluating candidate mission trajectories. Engineers must analyze large quantities of data to optimize mission safety and propellant margins, but currently rely on static text files and dense PDF slide decks to review simulation results. This approach is time-intensive, non-interactive, and difficult to share or coordinate with other mission designers. To improve this process, we present Porter: a web-based interactive mission analysis tool. This work describes Porter’s user-centered design process, its processing pipeline for importing and processing LAMBIC simulation data, core interactive features, and preliminary user feedback.

Ramaswamy, Basak Alper↗
DOE OSTI2018

IAEA CRP on HTGR UAM: Propagation of Phase I cross section uncertainties to Phase II neutronics steady state using SCALE/SAMPLER and PHISICS/RELAP5-3D

The IAEA Coordinated Research Project (CRP) on High Temperature Gas Cooled Reactor (HTGR) Uncertainties in Modeling (UAM) has been actively supported by the HTGR Methods group at Idaho National Laboratory since 2013. The INL team is responsible for the development of the prismatic reactor design exercises, based on the 350 MW General Atomics Modular HTGR (MHTGR). The focus of this paper is the reporting of the INL results obtained for the propagation of the lattice cross section uncertainties to the stand-alone neutronics simulation of the full MHTGR-350 core. For the nominal (i.e. best-estimate) model, a series of fresh and depleted single block and super cell SCALE/KENO and SCALE/NEWT models were developed at INL to assess spectral effects in the MHTGR-350, and a combination of these lattice cells were used to construct 8-group cross section libraries for use in the INL-developed PHISICS/RELAP5-3D code suite. The block-level homogenized libraries were assigned to the 220 fuel zones utilized in the MHTGR-350 PHISICS core model, according to the fresh and mixed fresh/depleted core loading pattern. A set of 1,000 perturbed NEWT lattice and PHISICS/RELAP5-3D core calculations were performed using the stochastic sampling module SAMPLER in SCALE 6.2. The paper discusses the mean and standard deviation values obtained for the eigenvalue, power offset and peak axial power density uncertainty results for this sequence of 1,000 lattice-to-core MHTGR-350 simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Development of iteration strategies for a practical implementation of a higher order transverse leakage approximation

Transverse integrated nodal diffusion methods currently represent the standard in full core neutronic simulation. The primary shortcoming in this approach is the utilization of the quadratic transverse leakage approximation. This approach, although proven to work well for typical LWR problems, is not consistent with the formulation of nodal methods and can cause accuracy and convergence problems. In previous work, an improved, consistent quadratic leakage approximation was formulated, which derived from the class of higher order nodal methods developed some years ago. In this paper a number of iteration schemes are developed around this consistent quadratic leakage approximation which yield accurate node average results in much improved calculational times. The developed consistent leakage approximation is extended in this work via a number of numerical schemes, the most promising of which results from utilizing the consistent leakage approximation as a correction method to the standard quadratic leakage approximation. Numerical results are demonstrated on a set of benchmark problems, such as the 3D IAEA LWR and MOX C5 problems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

PHISICS toolkit: Multi-reactor transmutation analysis utility - MRTAU

The principal idea of this paper is to present the new capabilities available in the PHISICS toolkit, connected with the implementation of the depletion code MRTAU, a generic depletion/decay/burn-up code developed at the Idaho National Laboratory. It is programmed in a modular structure and modern FORTRAN 95/2003. The code tracks the time evolution of the isotopic concentration of a given material accounting for nuclear reactions happening in presence of neutron flux and also due to natural decay. MRTAU has two different methods to perform the depletion calculation, in order to let the user choose the best on with respect to his needs. Both the methodologies and some significant results are reported in this paper. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Whole-core comet solutions to a 3-dimensional PWR benchmark problem with gadolinium

A pressurized water reactor (PWR) benchmark problem with gadolinium was used to determine the accuracy and computational efficiency of the coarse mesh radiation transport method COMET. The benchmark problem contains 193 square fuel assemblies. The COMET solution (eigenvalue, assembly averaged and fuel pin averaged fission density distributions) was compared with those obtained from the corresponding Monte Carlo reference solution using the same 2-group material cross section library. The comparison showed that both the core eigenvalue and fission density distribution averaged over each assembly and fuel pin predicated by COMET agree very well with the corresponding MCNP reference solution if the incident flux response expansion used in COMET is truncated at 2nd order in the two spatial and the two angular variables. The benchmark calculations indicate that COMET has Monte Carlo accuracy. In, particular, the eigenvalue difference between the codes ranged from 17 pcm to 35 pcm, being within 2 standard deviations of the calculational uncertainty. The mean flux weighted relative differences in the assembly and fuel pin fission densities were 0.47% and 0.65%, respectively. It was also found that COMET's full (whole) core computational speed is 30,000 times faster than MCNP in which only 1/8 of the core is modeled. It is estimated that COMET would have been about over 6 orders of magnitude faster than MCNP if the full core were also modeled in MCNP. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Comet solutions to a stylized BWR benchmark problem

In this paper, a stylized 3-D BWR benchmark problem was used to evaluate the performance of the coarse mesh radiation transport method COMET. The benchmark problem consists of 560 fuel bundles at 3 different burnups and 3 coolant void states. The COMET solution was compared with the corresponding Monte Carlo reference solution using the same 2-group material cross section library for three control blade (rod) configurations, namely, all rods out (ARO), all rods in (ARI) and some rods in (SRJ). The differences in the COMET and MCNP eigenvalues were 43 pcm, 66 pcm and 32 pcm for the ARO, ARI and SRI cases, respectively. These differences are all within 3 standard deviations of the COMET uncertainty. The average relative differences in the bundle averaged fission densities for these three cases were 0.89%, 1.24%, and 1.05%, respectively. The corresponding differences in the fuel pin averaged fission densities were 1.24%, 1.84% and 1.29%, respectively. It was found that COMET is 3,000 times faster than Monte Carlo, while its statistical uncertainty in the fuel pin fission density is much lower than that of Monte Carlo (i.e., {approx}40 times lower). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Coupling procedure for TRANSURANUS and KTF codes

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned tools can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Modeling of melt retention in EU-APR1400 ex-vessel core catcher

A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Heat Transfer in a Molten Natural Uranium Nuclear Reactor With Thermal Storage

Previous papers describe a method for controlling a thermal nuclear reactor by moderator displacement between the reactor core and the reflector. The low neutron loss and high control range of this method enable a design for a molten natural uranium fueled and cooled heavy water moderated reactor. The design is geometrically similar to CANDU (Canadian Deuterium Uranium) reactors but with fewer, larger, vertically oriented fuel tubes. The large fuel tubes have a mixed spectrum which is largely thermal at the periphery of the tubes with a very hard spectrum in the large central region of the tube. This configuration provides a fast fission rate of around 15% (mostly of U-238) and an initial conversion ratio of around 1.15. It is a hypothetical reactor because it is unclear what materials could actually be used to fabricate the fuel tubes and other primary coolant plumbing. The molten uranium serves both as the fuel and the primary coolant and as such is continuously circulated through the fuel tubes in the reactor core, the primary heat exchange, and back to the the reactor core. This means that at any point in time all the fuel has the same isotopic configuration no matter where it is in the reactor or heat exchanger. Over time the isotopic configuration changes, but it changes uniformly for all the fuel. Uranium melts at 1135 deg. C, so a proposed temperature range for the circulating molten uranium is 1200 deg. C at the exit of the heat exchanger and entry to the core fuel tubes and 1400 deg. C at the exit from the core fuel tubes and the entry into the heat exchanger. The advantages of this design include: 1. Reduced nuclear proliferation concern because of natural uranium fuel. 2. High conversion ratio and large fuel mass provides 30 to 100 year fuel life. 3. No heavy metal waste because fissile content increases over the fuel life. 4. High output temperatures are suitable for high efficiency turbine generators. This paper discusses the heat transfer in the molten natural uranium fueled reactor described above and in the cited paper in the simple case and when the reactor is combined with a thermal storage unit which is composed of ordinary salt (NaCl). The proposed molten uranium fueled and cooled nuclear reactor works particularly well when coupled with a molten salt heat storage unit, a process heat user, and high Efficiency air-Brayton cycle turbine generators which can also add energy to the hot air output of the tin/air heat exchanges by burning some natural gas or hydrogen. The combination can provide highly variable electrical output to the grid to accommodate large daily and seasonal mismatches between electrical demand and supply from variable sources such as wind and solar. Future work includes modeling of the temperatures throughout the molten salt heat storage unit under varying charge and discharge cycles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗