Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “nuclear reactor systems”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

Cryogenic Fluid Management Technology Development for Nuclear Thermal Propulsion

Cryogenic fluid management technology is critical to the success of future nuclear thermal propulsion powered vehicles and long duration missions. This paper discusses current capabilities in key technologies and their development path. The thermal environment, complicated from the radiation escaping a reactor of a nuclear thermal propulsion system, is examined and analysis presented. The technology development path required for maintaining cryogenic propellants in this environment is reviewed. This paper is intended to encourage and bring attention to the cryogenic fluid management technologies needed to enable nuclear thermal propulsion powered deep space missions.

Taylor, B. D.↗
DOE OSTI2022

Refueling water storage tank (RWST) with tailored passive emergency core cooling (ECC) flow

A nuclear reactor comprises a pressure vessel containing a nuclear reactor core. A reactor core cooling system comprises a standpipe including a plurality of orifices in fluid communication with a refueling water storage tank (RWST) to drain water from the RWST into the standpipe, and an injection line configured to drain water from the standpipe to the pressure vessel. In some embodiments the standpipe is disposed in the RWST, while in other embodiments the standpipe is disposed outside of the RWST and cross-connection pipes connect the plurality of orifices with the RWST. The reactor core cooling system may further comprise a valve configured to control flow through one orifice of the plurality of orifices in fluid communication with the RWST based on water level in the standpipe. The valve may comprise a float valve having its float disposed in the standpipe.

Watson, Ronald C.↗
DOE OSTI2022

Impacts of processing decisions on TNSL cross sections and their applications

Thermal neutron scattering law (TNSL) data describe low-energy neutrons scattering off of bound materials, and can have a significant impact on modeling any system with slow neutrons, including nuclear reactors. Previous work to introduce TNSL data to neutron transport codes at LLNL focused on COG and TART, with the limitation that these codes require highly specialized data processing and formatting. We have recently increased efforts to process TNSL data with the LLNL nuclear data processing code FUDGE. FUDGE reads and writes the evaluated and processed files using the generalized nuclear database structure (GNDS). This process uncovered some significant difficulties in processing TNSL data, and unearthed assumptions made in current TNSL data processing that we have found inadequate. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2022

Nuclear Power Concepts and Development Strategies for High-Power Electric Propulsion Missions to Mars

Under the Mars Transportation Assessment Study (MTAS), NASA and the Department of Energy are performing analyses and generating concepts for crewed nuclear electric propulsion (NEP) missions to Mars. This report presents the results of trade studies and concept development for the nuclear electric power system, consisting of the fission reactor, radiation shielding, power conversion, heat rejection, and power management and distribution (PMAD). The nuclear power team completed trade studies to evaluate different reactor and power conversion technologies and developed preliminary concepts for the crew shielding, waste heat radiators, and PMAD. The initial results suggest that a modified terrestrial microreactor combined with supercritical CO2 Brayton conversion could be used to perform the crew and cargo missions with satisfactory performance and modest risk. The report includes preliminary development strategies that could bring the NEP technology to fruition for Mars missions in the late 2030s or early 2040s.

Lee S. Mason↗
DOE OSTI2020

Study of Waste Generated by Retrieval of Fuel Debris at Fukushima Daiichi Nuclear Power Station - 20178

As a member of International Research Institute for Nuclear Decommissioning (IRID), we organized information on waste generated from the fuel debris retrieval, and considered containment, transport and storage processes. An important safety function to consider for these wastes is not only the handling of ordinary high-dose waste, but also the prevention of diffusion of fuel component contamination and the countermeasures against hydrogen gas generation. In order to satisfy these safety functions, we considered guideline for waste generated from the fuel debris retrieval, and proposed an example of handling process that includes generation of waste, container storage, on-site transfer and storage in a storage building. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Comparison of Nuclear Thermal Propulsion Reactor Cores Using Different Fuel Forms, Neutron Spectra, and Uranium Enrichments

A comparison of 300 MWt HEU and HALEU NTP cores using different fuel forms is presented to demonstrate trends and potentially inform fuel development. Compared to HEU, HALEU cores require additional in-core moderation. Beryllium was chosen as the reference in-core moderator material; metal hydrides may also be considered. Typical NTP fuel forms such as carbide and carbonitride solid solutions, carbide-graphite composites, CERMETs and CERCERs were considered together with less typical fuel forms such as UC and U metal. Be-moderated HALEU cores are heavier than HEU cores (1.9-2.7 and 0.9-2.0 metric tons, respectively). Furthermore, HALEU loading is very sensitive to the amount of in-core moderation which translates into tight fabrication tolerances. HALEU cores requires more reactivity control mechanisms than HEU cores to compensate for temperature feed-back. Using enrichments only slightly above HALEU (e.g., 25%) may provide a compromise between the administrative preference for using HALEU and the engineering preference for simplifying the reactor systems by, for example, not using in-core moderators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Manned space flight nuclear system safety. Voluem 5: Nuclear system safety guidelines. Part 2: Space shuttle/nuclear payloads safety

The design and operations guidelines and requirements developed in the study of space shuttle nuclear system transportation are presented. Guidelines and requirements are presented for the shuttle, nuclear payloads (reactor, isotope-Brayton and small isotope sources), ground support systems and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

Source record↗
DOE OSTI2021

Performance evaluation of a Terry GS-2 steam impulse turbine with air-water mixtures

Terry steam turbines are widely used in various industries because of their robust design. Within the nuclear power generation industry, they are used in the Reactor Core Isolation Cooling System to remove decay heat during reactor isolation events. During the Fukushima Daiichi nuclear power station disaster in Japan in 2011, the Reactor Core Isolation Cooling System and associated Terry turbine operated for over 70 hours in Unit 2; this runtime is well beyond the expected operating duration. Theories suggest the turbine was subjected to a two-phase inlet flow, which could degrade the turbine performance. In this work, an experimental test rig was constructed to test a full-scale Terry model GS-2 steam turbine under two-phase air/water flows. Steady-state efficiency and torque performance maps of the turbine were developed over a range of turbine inlet pressures (1.38–4.83 bar or 20–70 psia), air mass fractions (0.05–1.0) and rotational speeds up to 4000 RPM. Furthermore, turbine performance followed expected trends with torque varying linearly and efficiency varying quadratically with rotational speed. In addition, high-speed images of the two-phase flow entering the turbine were also analyzed to understand how changes in inlet pressure and air mass fraction affect the flow regime and homogenization. The present tests with air–water two-phase mixtures are an important step towards providing an understanding of the full-scale Terry turbine’s behavior and performance curves under two-phase conditions. The results of this work will be combined with air/water and steam/water data gathered using a small-scale Terry ZS-1 steam turbine in order to understand the scaling relationship between large and small size Terry turbines and fluid pairs. The combined data set will enable further development of analytical models over a wide range of conditions and may be used to provide technical justification for expanded use of the Terry turbines in nuclear power plant safety systems and other systems.

42 ENGINEERING↗
DOE OSTI2024

Development of Supervisory Control System for Thermal Energy Distribution System

The integrated energy system (IES) refers to the combination of nuclear energy generation with other energy sources to enable the efficiency and reliability of power generations. To create technologically viable and economically competitive systems, supervisory control strategies are critical for optimizing performance and ensuring stability across different energy generation, transportation, and utilization. This work focuses on the control strategies for the thermal energy distribution system (TEDS), which is a cornerstone of the Dynamic Energy Transport and Integration Laboratory at Idaho National Laboratory. TEDS currently relies on operators to coordinate across different components to manage energy storage and ensure efficiency. This work demonstrates the use of model predictive control (MPC) with surrogate models in determining optimal setpoints for major TEDS components. The capability of MPC-based supervisory control system is evaluated by autonomously matching the power outputs from a Dymola-based TEDS with target heat demands.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Extension of Argonne’s Plant Dynamics Code (PDC) Capabilities to TerraPower Direct-Cycle Supercritical CO2 (“Pascal”) Nuclear Reactor Concept. Final CRADA report

Argonne National Laboratory (the Contractor) and TerraPower, LLC, (the Participant), entered into a Cooperative Research and Development Agreement (CRADA) to extend the Plant Dynamics Code (PDC) to model the Pascal reactor plant. PDC is the system-level computer code developed at Argonne, primarily for design and transient analysis of supercritical carbon dioxide (sCO2) energy conversion systems for various power plants. Pascal is a supercritical carbon dioxide (sCO2)-cooled direct-cycle nuclear power reactor under development by TerraPower. New features were added to the code to model components and systems specific to the Pascal plant and to make the code readily available as a tool for design, control, and safety analysis for this and other direct-cycle power plants. TerraPower provided design information of the Pascal reactor and the requirements for the development of the new PDC features. Argonne modified the PDC capabilities to introduce the new features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

System Study: Auxiliary Feedwater 1998-2022

This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 62 U.S. commercial operating nuclear reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from calendar year 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the industry-wide estimates of AFW system start-only unreliability, but a highly statistically significant decreasing trend was identified in the industry-wide estimates of AFW system 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Preliminary Conceptual Design of Nuclear Thermal Rocket Reactor Cores Using Ceramic Fuels with Beryllium or Composite Neutron Moderators

Herein several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO 2 )] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH 1.6 or YH 1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235 U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Nuclear Thermal Propulsion Scoping Analysis of Fuel Plate Core Configurations

Nuclear thermal propulsion (NTP) engines use a nuclear reactor to heat up hydrogen to high temperatures. Thrust is generated as the hydrogen is ejected at high velocity through a nozzle. NTP has the potential of reducing travel times for deep space missions, such as to Mars. The United States studied NTP early in the space age and the National Aeronautics and Space Administration (NASA) is currently revisiting the potential of the technology to enable its missions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2004

High Temperature Heat Rejection System for Large Heat Loads; Architecture and Trade Study Results

To investigate space nuclear reactor technologies, NASA has awarded several contracts under Project Prometheus, the Nuclear Systems Program. The effort described in this paper was performed under one of those contracts (the Brayton NRA) . Like all power conversion systems, nuclear power conversion systems operate at efficiencies less than 100% resulting in the need to reject waste heat to space. Several different HRSs (Heat Rejection Systems) potential designs have been identified for rejecting NEP (Nuclear Electric Power) waste heat and several of them for a CBC (Closed Brayton Cycle) power conversion system are described herein and the results of their initial analyses presented. The analyses presented were performed as part of an initial trade study to recommend a promising HRS for advancement of its TRL.

Nilitkin, Michael N.↗
NASA NTRS2010

Testing of Liquid Metal Components for Nuclear Surface Power Systems

The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

Polzin, Kurt A.↗
DOE OSTI2023

Modeling and control of nuclear–renewable integrated energy systems: Dynamic system model for green electricity and hydrogen production

The need for decarbonization and diversification of energy resources has led to the development of integrated energy systems (IESs), where multiple resources supply more than one energy sector. Here, one such IES with small modular nuclear reactors and renewables (wind and solar) as generating resources, catering to the demand of the electric grid while producing hydrogen for industries, is modeled in this paper. The physics-based component models are represented using the Modelica language and interconnected to form the IES. The control and coordination of the overall system are ensured by designing a suitable control architecture composed of individual subsystem-level controls and supervisory control. The dynamic performance and the load-following capability of the IES are evaluated, while satisfying the safe operational limits of the components. Different configurations and modes of IES operation are considered, where the adaptability of the control system in the presence of varying demands and renewable generations is validated. The simulation results indicate that hydrogen as a flexible load facilitates the supply of varying grid demand. Additionally, the renewables are also accommodated into the IES owing to the flexibility of the balance of plant associated with the nuclear reactors.

24 POWER TRANSMISSION AND DISTRIBUTION↗
NASA NTRS2003

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Godfroy, Thomas↗