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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 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
NASA NTRS1972

Completely modular Thermionic Reactor Ion Propulsion System (TRIPS)

The nuclear reactor powered ion propulsion system described is an advanced completely modularized system which lends itself to development of prototype and/or flight type components without the need for complete system tests until late in the development program. This modularity is achieved in all of the subsystems and components of the electric propulsion system including (1) the thermionic fuel elements, (2) the heat rejection subsystem (heat pipes), (3) the power conditioning modules, and (4) the ion thrusters. Both flashlight and external fuel type in-core thermionic reactors are considered as the power source. The thermionic fuel elements would be useful over a range of reactor power levels. Electrical heated acceptance testing in their flight configuration is possible for the external fuel case. Nuclear heated testing by sampling methods could be used for acceptance testing of flashlight fuel elements. The use of heat pipes for cooling the collectors and as a means of heat transport to the radiator allows early prototype or flight configuration testing of a small module of the heat rejection subsystem as opposed to full scale liquid metal pumps and radiators in a large vacuum chamber. The power conditioner (p/c) is arranged in modules with passive cooling.

Peelgren, M. L.↗
DOE OSTI2023

SCALE Modeling of the Sodium Cooled Fast-Spectrum Advanced Burner Test Reactor

This report documents the modeling and simulation of a sodium-cooled fast reactor (SFR) as part of a U.S. Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with the Oak Ridge National Laboratory code SCALE and the Sandia National Laboratories (SNL) code MELCOR. Based on publicly available benchmark specifications, a fully heterogeneous 3D SCALE model of the 250 MWth Advanced Burner Test Reactor (ABTR) was developed to demonstrate SCALE’s capabilities for full-core reactivity analysis, fuel inventory prediction, and decay heat analysis of an SFR. The benchmark specifications contain modeling details for the ABTR core at the beginning of equilibrium cycle (BOEC) at operating conditions; they were derived from a 2006 preconceptual design report produced by Argonne National Laboratory. The ABTR was designed to demonstrate reactor-based transmutation of transuranics, that is, to “burn” transuranics recovered from light-water reactor (LWR) spent fuel. The ABTR’s fuel is designed to operate in 4 month cycles using uranium/transuranic (U/TRU) metallic fuel, with a TRU content of approximately 20%, at a conversion ratio of approximately 0.6. Various reactivity calculations were performed with SCALE for the ABTR and, where possible, compared with results available in the open literature. Additionally, SCALE was used to perform a full-core depletion calculation over the 4 month cycle to obtain the nuclide inventory at the end of equilibrium cycle (EOEC). These nuclide inventories, decay heat, power profiles, and reactivity feedback coefficients at EOEC represent the initial conditions for analyzing severe accident scenarios with MELCOR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly

A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235 U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134 Cs/ 137 Cs 134 Cs/ 154 Eu, and 154 Eu/ 137 Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235 U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining models to be reported in Parts 2 and 3.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

AGC-2 Graphite Preirradiation Data Analysis Report

This report describes the specimen loading order and documents all preirradiation examination material property measurement data for graphite specimens contained within the Second Advanced Graphite Capsule (AGC 2) irradiation capsule. The AGC 2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens in order to garner quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear grade graphites. This testing will ascertain the in service behavior of the graphite for pebble bed and prismatic very high temperature reactor designs. Similar to the First Advanced Graphite Capsule (AGC 1) preirradiation examination report, material property tests were conducted on specimens from 18 nuclear grade graphite types. However, AGC 2 tested an increased number of specimens (i.e., 512) prior to loading them into the AGC 2 irradiation assembly. All AGC 2 specimen testing was conducted at Idaho National Laboratory from July 2009 to August 2010. This report also details the specimen loading methodology for graphite specimens inside the AGC 2 irradiation capsule. The AGC 2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid plane. This provides similar specimens with and without an applied load. Analysis in this document utilizes the neutron flux profile calculated for the AGC 2 capsule design, the capsule dimensions, and the size (i.e., length) of the selected graphite specimens to create a stacking order that produces “matched pairs” of graphite specimens above and below the AGC 2 capsule elevation mid point, thus providing specimens with similar neutron dose levels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Initial exploration of a novel transient arrest system involving fuel heating

A preliminary analysis on a novel accident response system to diminish the severity of super- critical transients was conducted. The novel accident response system, called the instant shock arrest system, involves using electricity to heat the nuclear fuel at the onset of a large accidental reactivity insertion. This system is specifically designed for reactors with metallic fuel, such that the fuel is capable of conducting electricity, and being resistively heated. A reactor dynamics model of the advanced test reactor was created using the point kinetics equations and a linear reactivity feedback model to simulate how the system would effect the maximum fuel temperatures experienced during the transient. Transients with the instant shock arrest system were compared to those without it. It was found that the instant shock arrest system initially heated the fuel more than the unaffected transient but the negative reactivity inserted from such heating was enough to lower the maximum fuel temperature experienced during the transient. After simulating six different accident scenarios with reactivity insertions ranging from 0.5 to 1.3 dollar, it was found that an optimal system response could reduce peak fuel temperatures during the transient by 3.5% to 5%. Furthermore, discussion was given on how the optimal system response could be obtained using relatively simple numerical optimization algorithms due to the smoothness of the optimization problem. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

TREAT neutronics analysis of water-loop concept accommodating LWR 9-rod bundle

The Transient Reactor Test (TREAT) facility that was commissioned in 1959 and that was put in stand-by in 1994, is now facing an upgrade to provide irradiation services to a variety of experimental programs. The neutronic studies presented in this paper show that the TREAT facility is capable of performing transient tests on fresh 9-rod PWR bundles in a water loop design constructed from Inconel piping using only fuel assemblies from the original core. High burnup rods in the same proposed design configuration approach, but do not meet, energy deposition targets for narrow-pulse reactivity insertion accident testing. The option to use a Zircaloy-based water loop was evaluated and found to increase the power coupling factor (PCF) adequately, but several engineering considerations show that the Inconel loop should be considered the baseline design option until detailed design work is performed. TREAT's capabilities for steam-based LOCA testing were also evaluated and found to allow for higher specimen energy deposition than RIA tests primarily because the reactor can release more energy over longer shaped transients. The same test configurations were evaluated in the presence of a partial TREAT-upgrade (TU) core with the notable observation that TU elements can enable high burnup rods to achieve RIA energy targets with the Inconel water loop. Further observations show that the partial TU core can help recover excess reactivity as needed and flatten pin-to-pin power profiles for more prototypic test conditions. These studies demonstrate that a simple U/C atom ratio gradient implementation results in more energy deposition into the test specimen and works to flatten the rod-to-rod power profile. It is expected that the actual, specially designed TREAT upgrade core will be much more effective than this study of a modified version demonstrates and deserves a comprehensive investigation to identify true capability. Overall, the complex interplay between fuel, geometry, neutron absorbers, and moderators emphasize that the facility's capabilities are flexible and further enabled by the ability to use TU elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Conceptual Spacer Design for the ATR GEN I Target for Pu-238 Production in the Advanced Test Reactor at Idaho National Laboratory

The initial target design used for Pu-238 production at Idaho National Laboratory was designed by Oak Ridge National Laboratory to optimize the production of Pu-238 in the High Flux Isotope Reactor (HFIR) and are referred to as HFIR GEN II targets. To take advantage of the Advanced Test Reactor’s (ATR) taller active core region a redesign of the HFIR GEN II targets was needed. It was proposed to stack two HFIR GEN II targets nose to nose about the core center line; however, this resulted in excessive neutron and photon heating in the pellets located in the center. This peak heating was not desirable so three alternative designs were investigated for the ATR GEN I targets. The python-based code, MCNP to ORIGEN2 in Python (MOPY), was used to calculate the heating rates after 40 days of irradiation to capture the effects of each configuration. The purpose of this paper is to document the details of these conceptual design calculations and comparisons for the ATR GEN I targets.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Materials and Fuels Complex FY-22 – FY-26 Five-Year Mission Strategy

The Idaho National Laboratory (INL), through its designated mission of advancing innovative nuclear energy solutions, is actively engaged in the research, development, demonstration and deployment of advanced nuclear technology. Key to the success of INL’s mission is the Materials and Fuels Complex (MFC), the only complex in the U.S. that hosts a world-class assemblage of facilities, capabilities and instruments for handling, testing, and characterizing nuclear fuel and radioactive materials. Driven by its mission/vision of “Engineering and Experiments that Drive the World’s Nuclear Energy Future,” MFC is at the center of INL’s – and indeed the Department of Energy’s – advanced nuclear technology development initiatives, providing essential capabilities such as research-scale high-assay low-enriched uranium (HALEU) fuel production, reactor demonstration facilities, post-irradiation examination, and transient irradiation testing. Furthermore, MFC provides an ideal environment for test beds that are utilized for research, development and demonstration (RD&D) activities on used fuel treatment, nuclear non-proliferation, forensics, and nuclear power sources used for space exploration missions conducted by the National Aeronautics and Space Administration (NASA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensor design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of measurements made before and after the irradiation, at the same temperature, allowed us to measure the loss in reflectivity as well as the Bragg wavelength drift. The results are quite promising for some of the investigated gratings. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Thick Concrete Specimen Construction, Testing, and Preliminary Analysis

The purpose of the U.S. Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the operating lifetimes of nuclear power plants (NPPs) beyond 60 years. Since many important safety structures in an NPP are constructed of concrete, inspection techniques must be developed and tested to evaluate the internal condition. In-service containment structures generally do not allow for the destructive measures necessary to validate the accuracy of these inspection techniques. This creates a need for comparative testing of the various nondestructive evaluation (NDE) measurement techniques on concrete specimens with known material properties, voids, internal microstructure flaws, and reinforcement locations. A preliminary report detailed some of the challenges associated with thick reinforced concrete sections and prioritized conceptual designs of specimens that could be fabricated to represent NPP concrete structures for using in NDE evaluation comparisons. This led to the construction of the concrete specimen presented in this report, which has sufficient reinforcement density and cross-sectional size to represent an NPP containment wall. Details on how a suitably thick concrete specimen was constructed are presented, including the construction materials, final nominal design schematic, as well as formwork and rigging required to safely meet the desired dimensions of the concrete structure. The report also details the type and methods of forming the concrete specimen as well as information on how the rebar and simulated defects were embedded. Details on how the resulting specimen was transported, safely anchored, and marked to allow access for systematic comparative NDE testing of defects in a representative NPP containment wall concrete specimen are also given. Data collection using the MIRA Ultrasonic NDE equipment and initial results are also presented along with a discussion of the preliminary findings. Comparative NDE of various defects in reinforced concrete specimens is a key component in identifying the most promising techniques and directing the research and development efforts needed to characterize concrete degradation in commercial NPPs. This requires access to the specimens for data collection using state-of-the-art technology. The construction of the specimen detailed in this report allows for an evaluation of how different NDE techniques may interact with the size and complexities of NPP concrete structures. These factors were taken into account when determining specimen size and features to ensure a realistic design. The lateral dimensions of the specimen were also chosen to mitigate unrealistic boundary effects that would not affect the results of field NPP concrete testing. Preliminary results show that, while the current methods are able to identify some of the deeper defects, improvements in data processing or hardware are necessary to be able to achieve the precision and reliability achieved in evaluating thinner and less heavily reinforced concrete structures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

In-situ small-scale mechanical testing of fast reactor mixed oxide pins: completion report

The present work represents a pioneering study in the application of advanced techniques to measure micro-mechanical properties of highly radioactive advanced nuclear fuels. First data of Young’s modulus and hardness of fast reactor MOX at burnup exceeding 150 GWd/tHM have been obtained. Despite the challenges encountered in the current work due to equipment malfunction, the experimental approach sought in this work has been proved successful in obtaining new data of high relevance for fuel performance on system with high Pu content and high radiotoxicity with extremely limited exposure to personnel. Further experiments will be conducted in the future based on the present approach, leveraging the experience gained in this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary irradiation effect on corrosion resistance of zirconium alloys - 2016-0067

In order to improve the composition and structure of zirconium alloys, it is important to know how changes in their structure-phase state under irradiation affect various properties, including corrosion. In this work, autoclave corrosion tests were carried out on fuel cladding samples made of zirconium-niobium and zirconium-niobium-tin-iron alloy systems. All samples were preliminarily irradiated in a BOR-60 nuclear reactor at a fluence of 2 X10{sup 26} m{sup -2} (E>0.1 MeV). The autoclave tests of irradiated and unirradiated samples were performed at 350 deg. C for 240 days in distilled water containing 10 ppm lithium and 1,600 ppm boron. The irradiation of all alloys resulted in a significant change in their initial structure-phase state. The content of niobium in b-niobium-phase precipitates was reduced from 80 % to - 90 % to 50 % to - 60 %. Iron atoms from the Laves-phase particles were diffused into a matrix, and their lattice transformed from hexagonal close-packed to body-centered cubic. An average size of the transformed second-phase precipitates in the irradiated alloys increased relatively to an average size of the initial precipitates. Elongated radiation-induced precipitates (RIPs) with an average length up to 7 nm were formed in all alloys under irradiation. In comparison with the unirradiated state, irradiation reduced the corrosion of zirconium-niobium alloys, and this effect became more apparent with increased niobium content in the alloy. For zirconium-niobium-tin-iron alloys, irradiation led to more severe corrosion, mostly because of increasing tin content. Niobium content increased up to 2.37 %, and iron content decreased up to 0.18 % in the alloys, contributing to higher corrosion resistance. An interconnection between the niobium/(tin + iron) alloying parameter, quantity of nanometric precipitates formed therein under irradiation, and their corrosion performance has been suggested. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Irradiation test about oxidation-resistant graphite in WWR-K research reactor - 18533

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). It is desirable to enhance oxidation resistance of graphite to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies, Toyo Tanso, Ibiden, Tokai Carbon and Nippon Techno-Carbon are studying to develop oxidation-resistant graphite for the in-core components. Neutron irradiation test with two capsules was carried out by WWR-K research reactor of Institute of Nuclear Physics of the Republic of Kazakhstan (INP) through the International Science and Technology Center (ISTC) partner project. The total irradiation cycles of WWR-K operation was 10 cycles and the total duration was 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E>0.18 MeV) for the gas-swept capsule irradiated at a central irradiation hole was preliminary calculated as 1.2x10{sup 25} m{sup -2}, and for the sealed capsule at a peripheral irradiation hole as 4.2x10{sup 24} m{sup -2}. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Advances in Integral and Separate Effects Experiments for Water Cooled Small Modular Reactors

This paper focuses on the progress being made in water-cooled small modular reactor (SMR) advanced integral effects and separate effects experiments for reactor licensing. SMRs, considered modular in design, are mostly factory-built and then shipped to the reactor site. Of the several types of SMR designs available, water-cooled SMRs are likely to receive regulatory approval faster than others, as most of the technologies involved (e.g., the fuel and coolant technologies) are matured. However, the unique safety systems of SMRs, which depend on specific SMR design features, require integral and separate effects experiments to achieve reactor licensing. Thus, of the many SMR designs being proposed, only a few have successfully undergone licensing and reached the final development and demonstration stage. Many nuclear vendors and newcomer companies are investing millions of dollars to develop integral and separate effects testing facilities for preparing final safety analysis reports to include in licensing applications. Development and analysis of these experimental facilities is costly and takes about four to five years. The unique challenges involved can be reduced when stakeholders synergistically apply lessons learned, knowing the critical role played by advancements in experiments that support the licensing of SMR safety systems. Identification of knowledge/research gaps with the phenomena identification and ranking table (PIRT) and designing experimental facilities focusing on the phenomena of interest (POI) and figures of merit (FOMs) are pivotal to select the critical path to successful design demonstration and licensing application.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Developing the test conditions for the first fueled transients in the TREAT reactor

The restart of the Transient Reactor Test (TREAT) facility at Idaho National Laboratory, re-establishes a national in-pile testing capability for nuclear fuels in accident scenarios. The purpose of the first tests in TREAT are to study the response of nuclear fuel durring a reactivity initiated accident (RIA). The initial objective is to demonstrate the ability to create a desired enthalpy increase in a representative fuel test specimen over a desired time step and to better understand the uncertainty. The five test rods initially available for these irradiations are UO2-Zircaloy with standard Pressurized Water Reactor (PWR) geometry. The rods are 15 cm long with 9 cm of active fuel. The initial testing environment is dry helium. BISON fuel performance calculations were run to model the heat transfer during postulated transients. Based on those results energy deposition targets were selected around the range of observable phenomena in the testing environment. Reactor / experiment power coupling factors, step insertion size and transient clipping response times were then calculated using results from a RELAP point kinetics model. Calculated energy deposition targets are 150, 350, 650, 960, and 1300 J/gUO2. The energy coupling factor for the experiments was calculated to be ~1.5 J/gUO2-MJ. Two step insertion sizes were selected: a 0.6% ?k/k step for the 150 J/g transient and a 1.8% ?k/k step for the other transients. Clipping response times ranged from 1.3 seconds to unclipped transients. The experiments will be irradiated in the TREAT facility in the fall of 2018. The first transient will be the 150 J/g pulse with a 0.6% ?k/k step that is unclipped. Follow on tests may modify the reactor operating parameters based on results from previous transients. Any preliminary data available on the results of the transients at the time of the NUMAT Conference will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗