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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Benchmark Calculation for the Quad Cities Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

In this study, benchmark calculations were performed for the Quad Cities Unit 1 cycles 1–3 to validate the SCALE 6.3/Polaris–PARCS v3.4.2 code package with the ENDF/B-VII.1 AMPX 56-group library by comparing the simulated results with the measured data. The benchmark results will be used in evaluating uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis for key nuclear parameters such as reactivity and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS; additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results are summarized herein so that they can be used to evaluate uncertainties with other benchmark results for key nuclear parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Benchmark Calculation for the Hatch Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

This study was the performance of the benchmark calculation for the Hatch Unit 1 cycles 1–3, to validate the SCALE 6.3/Polaris–PARCS v3.4.2 with the ENDF/B-VII.1 AMPX 56-group library by comparing the simulated results with the measured data. The benchmark results will be used in evaluating uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor (BWR) physics analysis for key nuclear parameters such as reactivity and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS and PARCS, and additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used in evaluating uncertainties for key nuclear parameters with other BWR benchmark results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Benchmark Calculation for the Peach Bottom Unit 2 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

In this study, benchmark calculations were performed for Peach Bottom Unit 2 cycles 1–3 to validate the SCALE 6.3/Polaris–PARCS v3.4.2 with the ENDF/B-VII.1 AMPX 56-group library by comparing the simulated results with the measured data. The benchmark results will be used to evaluate uncertainties of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis for key nuclear parameters such as reactivity and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS and PARCS. Additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used to evaluate uncertainties with other benchmark results for key nuclear parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Benchmark Calculation for the Watts Bar Unit 1 Cycles 1-3 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

Benchmark calculations for the Watts Bar Unit 1 cycles 1–3 were performed to validate the SCALE 6.3/Polaris–PARCS v3.4.2 with the ENDF/B-VII.1 56-group library by comparing the simulated results with the measured data. The benchmark results are used in evaluating uncertainties of the SCALE/Polaris–PARCS code package for pressurized water reactor physics analysis for key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris, GenPMAXS, and PARCS. Additional details are provided for the input and output files produced for the benchmark calculations. The benchmark results were summarized such that they can be used in evaluating uncertainties for key nuclear parameters with other benchmark results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Benchmark Calculation for BEAVRS Cycles 1 and 2 Using the SCALE 6.3/Polaris–PARCS v3.4.2 Code Package

This study performed the benchmark calculation for BEAVRS to validate the SCALE 6.3/Polaris–PARCS v3.4.2 code package with the ENDF/B-VII.1 56-group library, comparing the simulated results with the measured data. The benchmark results will be used in evaluating uncertainties of the SCALE/Polaris–PARCS code package for pressurized water reactor physics analysis for key nuclear parameters such as reactivity, control bank worth, temperature coefficients, and pin and assembly power peaking factors. This report details plant and fuel design specifications and input data for SCALE/Polaris as well as GenPMAXS and PARCS. Additionally, detailed information is provided for all the input and output files produced for the benchmark calculations. The benchmark results are summarized herein so that they can be used in evaluating uncertainties for key nuclear parameters with other benchmark results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1977

Nuclear thermionic converter

Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

Phillips, W. M.↗
NASA NTRS1991

Constitutive and damage material modeling in a high pressure hydrogen environment

Numerous components in reusable space propulsion systems such as the SSME are exposed to high pressure gaseous hydrogen environments. Flow areas and passages in the fuel turbopump, fuel and oxidizer preburners, main combustion chamber, and injector assembly contain high pressure hydrogen either high in purity or as hydrogen rich steam. Accurate constitutive and damage material models applicable to high pressure hydrogen environments are therefore needed for engine design and analysis. Existing constitutive and cyclic crack initiation models were evaluated only for conditions of oxidizing environments. The main objective is to evaluate these models for applicability to high pressure hydrogen environments.

Russell, D. A.↗
NASA NTRS2010

Nondestructive Crack Detection in a Fuel System Component

The paper discusses development of various NDE techniques to detect cracks in A40 steel poppets used in a valve of the fuel system of the Space Shuttle Orbiter. The valve assembly experiences a severe high cycle fatigue environment during its operation. Cracks were discovered at the radius of the poppet flange. Experience shows that very small cracks or material anomalies do not cause failure in a single operation event. While the design is being modified to eliminate the issue, NDE has been used to screen the poppets for cracks before every use. Several surface flaw detection techniques were considered and a few NDE techniques were developed to provide NDE screening for the flaw detection. The primary method used was the eddy current testing. In the eddy current technique, the X-Y channel test data from the eddy current instrument was recorded as computer files. A Matlab data review and plotting application was developed to analyze the data files. The Matlab application provides much higher resolution than the eddy current instrument that was used to acquire the data. Other techniques that were used included ultrasonic surface wave and magnetic particle testing. A probability of detection (POD) study was undertaken to determine the 90/95 size for the eddy current technique. This study used specimens with same geometry and material as the poppet. Fatigue cracks were grown in these specimens. Information on results of the NDE techniques and results of the POD study are provided.

Koshti, Ajay↗
DOE OSTI2022

Diode assembly and method of forming a diode assembly for pulsed fusion events

A diode assembly for producing a pulsed fusion event in a z-pinch driver. The diode assembly includes an inner core formed of a fusionable fuel source material including a lithium compound formed of one or more lithium isotopes and one or more hydrogen isotopes. A lithium metal outer sheath is integrally formed around the inner core by decomposing a surface of the fusionable fuel source material.

Brown, Sam W.↗
DOE OSTI2021

MEDE Project

Presentation on the Melt-Drain-Evaporation Testing associated with a non-irradiated Fermi-1 reactor assembly

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1974

Linear Test Bed. Volume 2: Test Bed No. 2

Test bed No. 2 consists of 10 combustors welded in banks of 5 to 2 symmetrical tubular nozzle assemblies, an upper stationary thrust frame, a lower thrust frame which can be hinged, a power package, a triaxial combustion wave ignition system, a pneumatic control system, pneumatically actuated propellant valves, a purge and drain system, and an electrical control system. The power package consists of the Mark 29-F fuel turbopump, the Mark 29-0 oxidizer turbopump, a gas generator assembly, and propellant ducting. The system, designated as a linear aerospike system, was designed to demonstrate the feasibility of the concept and to explore technology related to thrust vector control, thrust vector optimization, improved sequencing and control, and advanced ignition systems. The propellants are liquid oxygen/liquid hydrogen. The system was designed to operate at 1200-psia chamber pressure at an engine mixture ratio of 5.5. With 10 combustors, the sea level thrust is 95,000 pounds.

Source record↗
DOE OSTI2022

Current Progress of the Final Design of a Subcritical Assembly at the Oak Ridge National Laboratory

During the past decade, the US Department of Energy (DOE)/National Nuclear Security Administration (NNSA) Nuclear Criticality Safety Program (NCSP) has been coordinating nuclear criticality safety (NCS) courses to provide training and qualification to new staff, supervisors, managers, regulators, and other professionals involved with NCS. To address the growing need for more NCS training possibilities, a subcritical assembly was designed at Oak Ridge National Laboratory (ORNL). The Oak Ridge Subcritical Assembly (ORSA), operated at ORNL, could be used in a new course for fissile material handlers/operators or as a backup training facility for currently existing NCSP hands-on NCS training. This paper describes some of the design steps necessary for obtaining a final ORSA design, beginning with a successful feasibility study performed in 2020. Numerous parameters within each main element of the assembly were considered, such as the reflector used, the type of fuel coating or plating used, and the number of fuel plates used. The design choices related to each parametric study described in this paper are detailed and justified at the end of each subsection. An overview of the current final design characteristics of ORSA and the experiments’ possibilities is also provided. ORSA will be built with fuel plates from legacy Aerojet General Nucleonics (AGN)-201M research reactors currently stored at the Y-12 National Security Complex (Y-12). The cylindrical fuel plates comprise UO 2 particles embedded in polyethylene and have a nominal enrichment of 19.5 ± 0.5 wt % 235 U. The basis for the final design was the feasibility study from 2020. In the feasibility study, ORSA achieved k eff = 0.95, corresponding to a multiplication factor of 20, with approximately 620 g of 235 U (four 4 cm plates, three 2 cm plates, and two 1 cm plates) and an 8.5 cm thick graphite reflector radially and on top of the assembly. The bottom of the assembly has a graphite thickness of 5 cm, and the chosen graphite density was 2.3 g/cm 3 . All design calculations were performed using SCALE 6.2.4 KENO-VI and with the ENDF/B VII.1 Continuous Energy cross-section library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

IER 516: Zirconium Test Assembly CED Phase-1: Preliminary Experiment Design

IER-516, Zirconium Test Assembly (ZTA), is a campaign to design, execute, and document a series of high-fidelity critical benchmark experiments to validate current and future zirconium (Zr) and zirconium hydride (ZrH x ) nuclear data evaluations. ZTA is a collaborative project between Los Alamos National Laboratory and the French Autoritè de Sûretè Nuclèaire et de Radioprotection. The experiments will be fueled with highly enriched uranium (HEU), and utilize the Comet critical assembly machine at the National Criticality Experiments Research Center. A total of ten preliminary critical experiments were designed and optimized for Zr and ZrHx nuclear data sensitivities using the MCNP-Particle Swarm Optimization methodology in the thermal, epithermal, intermediate, and fast neutron energy regions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A neutron fluence map of the Los Alamos National Laboratory Godiva IV critical assembly

A neutron fluence map and a total ionizing dose map of the Los Alamos National Laboratory Godiva IV fast burst critical assembly was generated using passive reactor dosimetry, comprised of sulfur pellets and thermoluminescent dosimeters. Godiva IV is an unmoderated, fast burst, critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5 % molybdenum for strength. The mapping was performed during a single 75.6 ºC temperature rise burst operation, with the top and sides of the cylindrical Godiva-IV Top Hat covered in passive dosimetry. Dosimetry was placed in a symmetric pattern around the Top Hat, with higher concentrations near the control rods and burst rod. A specific portion of the lower quadrant of the burst rod was mapped to confirm a testing region where the neutron fluence varied by no more than ± 5%. The results will be used to assess the neutron, gamma, and total ionizing dose environment in three-dimensional space around the assembly for higher fidelity experiment placement, active dosimetry positioning, and radiation field characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

IHS-60 Stored Energy Analysis

The IHS-60 consists of a General Purpose Heat Source (GPHS) Fueled Clad (FC) that has been hot isostatically pressed to eliminate the gap between the iridium clad and the fuel pellet, thus preventing differential motion between the pellet and the clad in shock and vibration environments. The vented FC is enclosed within a Molybdenum barrier comprising the liner body and the lower support member. The slip- t Molybdenum liner members serve as a mechanical barrier to plutonia contact of the tantalum alloy (T-111) strength member in the event the iridium clad should fail. The slip- t liner is not a barrier to gaseous transport, so a tantalum getter is installed near the vent region of the FC to preclude potential oxygen transport to the T-111 interior surface. The T-111 strength member encloses the Molybdenum liner assembly and provides for both helium gas retention and isotope fuel containment under impact and external pressure environments. It is surrounded by an outer clad of Hastelloy-S which protects the strength member from external oxidation and corrosive sources during nominal operation and in the event of off-nominal operation or accidents. Both the strength member and outer clad parts are closed with seal welds. The strength member and outer clad are the primary containment members of the 60-Watt IHS. The iridium clad contains the fuel pellet during nominal operation and mild accident environments, but it is not considered to be a primary containment member.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1992

Integral edge seals for phosphoric acid fuel cells

A phosphoric acid fuel cell having integral edge seals formed by an elastomer permeating an outer peripheral band contiguous with the outer peripheral edges of the cathode and anode assemblies and the matrix to form an integral edge seal which is reliable, easy to manufacture and has creep characteristics similar to the anode, cathode and matrix assemblies inboard of the seals to assure good electrical contact throughout the life of the fuel cell.

Granata, Jr., Samuel J.↗
DOE OSTI2021

An overview of bipolar plates in proton exchange membrane fuel cells

Bipolar plates are a crucial component of proton exchange membrane fuel cells. They are responsible for transporting reactant gases, carrying the current from the membrane electrode assembly to the end plates, providing heat and water management, and separating the individual cells. However, these plates also contribute to 80% of the fuel cell’s weight, 50% of its volume, and 40% of its cost, posing a barrier to the commercialization of fuel cells. This paper provides a comprehensive review of the materials and manufacturing processes used in the fabrication of bipolar plates as well as recent research conducted on the improvement of bipolar plate weight, volume, and cost through material selection and manufacturing methods. Additive manufacturing is highlighted in this work as an innovative manufacturing method to produce bipolar plates. Novel contributions in this paper include a detailed explanation of traditional manufacturing processes for metallic and graphitic-polymer bipolar plates as well as a cost comparison between additive and traditional manufacturing processes.

08 HYDROGEN↗