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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
NASA NTRS1989

Standard Method For Radiation Tests Of Liquids

Simple procedure provides standard method for radiation tests of liquids. Assures identical handling and dosage so data from different test laboratories compared reliably. Pair of glass vials, joined by frangible neck and equipped with valves, used. Third vial also used. Filled with control specimen and not irradiated. Developed for measuring properties of liquids for use in radiation-resistant liquid-dielectric capacitors, procedure used for any of variety of liquids proposed for use in high-radiation environment - in x-ray and particle-accelerating machines and nuclear reactors, for example.

Bouquet, Frank L.↗
NASA NTRS1988

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

Koontz, Steven L.↗
NASA NTRS1950

Preliminary analysis of three cycles for nuclear propulsion of aircraft

A preliminary study was made of the feasibility of three cycles for nuclear propulsion of aircraft: a direct-air-turbojet, a binary liquid-metal turbojet, and a helium compressor jet. All three cycles appeared feasible for flight at a Mach number of 0.9 and altitudes up to 50,000 feet; the liquid-metal cycle appeared feasible for flight at a Mach number of 1.5. The air and helium cycles resulted in heavier aircraft than did the liquid-metal cycle, particularly at a Mach number of 1.5. The relative advantage of the liquid-metal cycle became greater as the flight speed and altitude increased, and as the reactor wall temperature decreased.

Humble, L V↗
DOE OSTI2014

Nuclear Naval Propulsion: A Feasible Proliferation Pathway?

There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navy while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.

export controls↗
DOE OSTI2012

A 1 GeV CW FFAG High Intensity Proton Driver

The drive for high beam power, high duty cycle, and reliable beams at reasonable cost has focused world attention on fixed-field accelerators, notably Fixed-Field Alternating Gradient accelerators (FFAGs). High-intensity GeV proton drivers are of particular interest, as these encounter duty cycle and space-charge limits in the synchrotron and machine size concerns in the weaker-focusing cyclotron. Recently, the concept of isochronous orbits has been explored and developed for non-scaling FFAGs using powerful new methodologies in FFAG accelerator design. These new breeds of FFAGs have been identified by international collaborations for serious study thanks to their potential applications including Accelerator Driven Subcritical Reactors (ADS) a nd Accelerator Transmutation of Waste. The extreme reliability requirements for ADS mandate CW operation capability and the FFAG s strong focusing, particularly in the vertical, will serve to mitigate the effect of space charge (as compared with the weak- focusing cyclotron). This paper reports on these new advances in FFAG accelerator technology and presents a stable, 0.25-1GeV isochronous FFAG for an accelerator driven subcritical reactor.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Design and Analysis of a NPO2 Target for the Production of 238PU

Current production of 238Pu at Oak Ridge National Laboratory uses NpO2–aluminum cermet pellets as the feed material. The advantage of cermet is that its thermal conductivity is much higher than that of the oxide alone, which greatly reduces pellet temperatures. However, the melting point of the material is quite low due to the aluminum in the matrix, and the recovery of the 238Pu results in substantial radioactive liquid waste. There is potentially both a significant production gain and a reduction in post-irradiation processing cost and waste associated with using a pure NpO2 pellet. This work describes ongoing efforts to design a NpO2 pellet to maximize production of 238Pu in the High Flux Isotope Reactor.

McDuffee, Joel↗
DOE OSTI2013

Study on generating of thermal neutron scattering cross sections for LiH

LiH is designated as a promising moderator and shielding material because of its low density, high melting point and large fraction of H atoms. However, lack of the thermal neutron cross sections of LiH makes numerical calculation deviate from experimental data to some extent. As a result, it is necessary to study LiH thermal kernel effect. The phonon property of LiH has been investigated by first-principles calculations using the plane-wave pseudo potential method with CASTEP code. The scattering law and the thermal neutron scattering cross sections for Li and H have been generated using this distribution. The results have been compared with zirconium hydride data. The GASKET and NJOY/LEAPR codes have been used in the calculation of scattering law, whose results have been compared with the reference; the discrepancy mainly comes from phonon spectrums and its expansion. LEAPR had the capability to compute scattering through larger energy and momentum transfers than GASKET did. By studying LiH phonon spectrum and constructing the model of LiH thermal kernel and scattering matrix, the ACE format LiH thermal neutron cross sections for MCNP software could be made and used for reactor Neutronics calculation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermal Conductivity Degradation in High Burnup U-Pu-Zr Fuel

Recent advancements in the characterization of irradiated U-Pu-Zr fuels have revealed complexities that challenge existing understanding of constituent redistribution. Traditionally, models have proposed three concentric regions within the fuel, each characterized by distinctive phases and porosity. However, through detailed analysis of high burnup U-Pu-Zr, we discovered the presence of four distinct constituent redistribution regions. Particularly novel is the observation of significant Pu redistribution, a previously unreported phenomenon that necessitates a reevaluation of current models. This work aims to delve deeper into these findings, seeking to correlate mesoscale measurements of thermal diffusivity and respective thermal conductivity with the phases present in each redistribution region. To achieve this objective, we employed mesoscale thermoreflectance methods using the unique, Idaho National Laboratory (INL) developed, Thermal Conductivity Microscope (TCM) at INL’s Irradiated Materials Characterization Laboratory. The TCM employs two tightly focused lasers: one for heating to generate periodic thermal waves in the substrate, and another spatially separated probe laser to detect changes in the optical reflectivity of the gold-coated substrate resulting from thermal wave diffusion. We conducted several thermal diffusivity measurements within each region of constituent redistribution of a U-19Pu-10Zr fuel pin cross section irradiated to 11 at. % burnup. The TCM measurement positions strategically aligned with transmission electron microscopy (TEM) lift-out locations previously collected from the fuel sample. Complementary microstructural analysis techniques such as optical and scanning electron microscopy (OM/SEM), electron probe microanalysis for chemical compositions, and TEM-based selective area electron diffraction (SAED) analysis for crystallographic insights into each phase were also utilized. This comprehensive approach allowed us to correlate local thermal diffusivity data with microstructural characteristics, enabling the computation of local thermal conductivity at each position. The significance of this contribution lies in its pioneering use of the TCM for ternary fuel mesoscale examination, shedding light on the previously overlooked effects of Pu redistribution on local thermal conductivity. By informing current models capturing constituent redistribution and heat transfer, our findings pave the way for more accurate predictions of metallic fuel performance. Moreover, this work sets the stage for future comparisons with similar TCM examinations on U-19Pu-10Zr fuels at ultra-low burnup, facilitating a comprehensive understanding of thermal property changes across different burnup levels. Ultimately, our study not only enriches our understanding of the thermophysical properties of individual redistribution regions within U-Pu-Zr fuel but also offers valuable insights for the design and operational parameters of proposed next-generation fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

U.S. DOE’S RESPONSE TO THE FUKUSHIMA DAIICHI REACTOR ACCIDENT: ANSWERS AND DATA PRODUCTS FOR DECISION MAKERS

The Fukushima Daiichi response posed a plethora of scientific questions to the U.S. Department of Energy’s (DOE) radiological emergency response community. As concerns arose for decision makers, the DOE leveraged a community of scientists well-versed in the tenants of emergency situations to provide answers to time-sensitive questions from different parts of the world. A chronology of the scientific Q and A that occurred is presented along with descriptions of the challenges that were faced and how new methods were employed throughout the course of the response.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Atmospheric pressure ionisation - mass spectrometry for argon online monitoring in ASTRID

Safety, operability, waste transmutation and mastered investment costs are key issues for the development of generation IV reactors, especially sodium cooled fast reactors. To achieve a development scheduled around 2040, the nuclear energy division of CEA, France, is studying a prototype called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) which is in its conceptual design phase. To achieve a high level of maturity, In Service Inspection and Repair is very important, in particular fuel cladding failure detection. If such a failure occurs, fission products Xenon and Krypton are released. Their fast and sensitive detection is a challenge. The use of a commercial API-MS is evaluated in this work. Detection limits below 10 ppt have been measured for Kr 84 and for Xe 132. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Evaluation for Effect of TRACE Spacer Grid Model on RBHT Reflood Tests

Usually, the spacer grid decreases the flow area and then enhances the convective heat transfer due to the flow acceleration and the turbulence increases. Their phenomena is shown in both egg-crate style spacer grids and mixing vane grids. Also, the mixing vane swirls the flow and enhances the convective heat transfer. In this study, the effect of spacer grid model of TRACE was performed on RBHT (Rod Bundle Heat Transfer) reflood tests. The RBHT tests were designed to conduct several separate-effects tests such as single phase steam cooling, low flow boiling and dispersed flow film boiling heat transfer, which is very important during a large break loss of coolant accident (LOCA). This facility has a heated length of 3.66 m (12.0 ft), which is typical of a pressurized water reactor (PWR) fuel assembly. The test section consists of a 7 x 7 rod bundle with 45 heated electrical rods, and seven mixing vane grids. The electrical rods have a diameter of 9.5 mm and are arranged with a 12.6 mm pitch with 45 heater rods and four unheated rods. Tests 1096, 1108, 1170, 1196, 1285 and 1383, which are in the TRACE assessment manual , were selected to evaluate the effect of the spacer grid model of TRACE. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Uncertainties in Predictions of Material Performance Using Experimental Data That Is Only Distantly Related to the System of Interest

There is a need for predictive material “aging” models in the nuclear energy industry, where applications include life extension of existing reactors, the development of high burnup fuels, and dry cask storage of used nuclear fuel. These problems require extrapolating from the validation domain, where there is available experimental data, to the application domain, where there is little or no experimental data. The need for predictive material aging models will drive the need for associated assessments of the uncertainties in the predictions. Methods to quantify uncertainties in model predictions, using experimental data that is only distantly related to the application domain, are discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary Tritium Management Design Activities at ORNL

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Preserving physics knowledge at the fast flux test facility

One of the goals of the Dept. of Energy's Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Evaluation of accuracy of calculations of VVER-1000 core states with incomplete covering of fuel by the absorber

An additional verification of bundled software (BS) SAPFIR-95 and amp;RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (K{sub eff}) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K{sub eff} on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Single-reactor process for producing liquid-phase organic compounds from biomass

Disclosed is a method for preparing liquid fuel and chemical intermediates from biomass-derived oxygenated hydrocarbons. The method includes the steps of reacting in a single reactor an aqueous solution of a biomass-derived, water-soluble oxygenated hydrocarbon reactant, in the presence of a catalyst comprising a metal selected from the group consisting of Cr, Mn, Fe, Co, Ni, Cu, Mo, Tc, Ru, Rh, Pd, Ag, W, Re, Os, Ir, Pt, and Au, at a temperature, and a pressure, and for a time sufficient to yield a self-separating, three-phase product stream comprising a vapor phase, an organic phase containing linear and/or cyclic mono-oxygenated hydrocarbons, and an aqueous phase.

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↗
DOE OSTI2019

Status Report on the NuScale Module Developed in the Modelica Framework

This report has been prepared to give a status update on the current state of the NuScale reactor model in the Modelica framework. Current model development has led to the creation of a dynamic NuScale module in the Modelica language that operates under natural circulation and is consistent with design parameters set forth in the design certification documentation. Controllers have been developed that are consistent with controllers mentioned by NuScale and similar pressurized water reactor systems. Simulation results demonstrate the capability of said controllers to properly maintain their target parameters while preserving system values within reasonable bounds. Through the commencement of this work, the first generation of NuScale model in the Modelica M&S framework has been completed. This first generation is capable of recreating nominal design values laid forth in the public literature and accommodating transients down to 60% power. Future generations of the model will incorporate finer tuning on controllers to reduce oscillations seen with steam offtake and will include a more rigorous representation of components laid forth in the future work section.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗