Gaseous core reactors
Gaseous core reactors for nuclear powered space propulsion
SPACECRAFT POWER SUPPLY↗
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Gapless core reactor
A gapless core reactor includes a saturable magnetic core having reactor legs without air gaps and multiple windings. The windings are each wound around a common leg, spaced apart from each other and connected in counter series. The windings are configured such that a magnetic flux generated from an alternating current flowing through the windings generates a plurality of substantially equal and counter magnetic fluxes flowing through two or more separate magnetic circuits.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis
Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12-18 months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.
Brown, Nicholas R.↗
Parallel simulated annealing with embedded machine learning and multifidelity models for reactor core design
This paper presents extensions to a penalty-free, parallel simulated annealing (SA) algorithm for multi-constrained combinatorial optimization with the aim of embedding multi-fidelity physics models into the annealing procedure. The method uses a low-fidelity, quickly executing model for rapid design space exploration and a high-fidelity model for detailed constraint resolution and on-the-fly bias correction. Machine learning models updated within the annealing procedure were used to bridge the gap between the multi-fidelity models, which led to accurate rapid exploration and efficient detailed constraint resolution. A software implementation of the new multi-fidelity optimization methods, called ML-PSA, was demonstrated on a continuous multi-fidelity optimization problem and a constrained combinatorial PWR lattice design problem. These problems demonstrate some of the features, parallel performance characteristics, and extensible nature of the multi-fidelity SA methods. This paper shows that the developed software and procedure are a general optimization tool that can be applied to a wide variety of scientific and engineering design optimization applications. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores
A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.
Litchford, Ron J.↗
A flow experiment on a curved-porous-wall gas- core reactor geometry.
Gas core reactor curved porous walls geometry tested with clear and smoky air injection, noting mass flow ratio in reactor cavity
Lanzo, C. D.↗
Micro-reactor core mechanical support
A nuclear reactor core mechanical support bracket is disclosed. The support bracket includes a housing, a spring disposed internally within the housing, a shaft slidingly disposed within the housing, a shaft travel pin, and a flange. The shaft is configured to engage the spring to compress and decompress the spring as the shaft travels in and out of the housing. The shaft travel pin controls the travel of the shaft. The flange is configured to mount the nuclear reactor core mechanical support bracket to a canister of a nuclear reactor. The shaft includes an inset configured to interface with a nuclear reactor core component.
Aleshin, Yuriy↗
Applications of plasma core reactors to terrestrial energy systems
Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-
Latham, T. S.↗
Heat pipe reactor core and heat exchangers formation and deployment
A heat pipe reactor may include a reactor core and one or more heat exchangers positioned on one or both sides of the reactor core. The heat pipe reactor may also include a plurality of heat pipes extending from the reactor core and out through the one or more heat exchangers. The reactor core may be composed of a plurality of monolithic blocks.
Reid, Robert Stowers↗
GASEOUS CORE REACTORS
Problems in the development of gas core reactors for heating propellants, considering the vortex reactor and plasma core reactor concepts
GAS REACTOR↗
Preliminary analysis of effect of Vanadium Self-Powered Neutron Detectors on AP1000 reactor core physical parameters - 360
For the AP1000 reactor, the Vanadium Self-Powered Neutron Detectors are installed in the instrumentation thimbles of the fuel assembly. The (n,? ) reaction of Vanadium SPNDs will impact on the neutron spectrum of the location, and finally impact on the reactor core physical parameters. But, this effect is not considered for the AP1000 reactor Nuclear Design. So, it is necessary to analyze the effect of the presence of Vanadium SPNDs on the AP1000 reactor core physical parameters. CASMO5 is used to calculate physical parameters of four typical assemblies of AP1000 reactor, and SIMULATE5 is used to calculate physical parameters of the initial cycle. The results show that the maximum absolute deviations of K-infinite, pin power, thermal flux for the assemblies are 181 pcm, 6.3% and 20.1%, separately. The maximum absolute deviations of critical boron concentration, heat flux hot channel factor, nuclear enthalpy rise hot channel factor, axial offset, quadrant power tilt ratio and axial power distribution of the initial cycle are 5 ppm, 0.7%, 0.5%, 0.1%,0.09% and 0.19%,separately. So, the presence of Vanadium SPNDs has a significant impact on the power distribution of pin cells around the instrumentation thimble and the thermal flux in the location of instrumentation thimble, but doesn't affect core physical parameters of the initial cycle very much. As is shown, it is very important to accurately consider the effect of presence of Vanadium SPNDs when calculating the Vanadium SPNDs current response. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Post impact behavior of mobile reactor core containment systems
The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.
Puthoff, R. L.↗
Nuclear reactor core support system providing radial and axial support
A nuclear reactor core mechanical support bracket is disclosed. The support bracket includes a housing, a spring disposed internally within the housing, a shaft slidingly disposed within the housing, a shaft travel pin, and a flange. The shaft is configured to engage the spring to compress and decompress the spring as the shaft travels in and out of the housing. The shaft travel pin controls the travel of the shaft. The flange is configured to mount the nuclear reactor core mechanical support bracket to a canister of a nuclear reactor. The shaft includes an inset configured to interface with a nuclear reactor core component.
Aleshin, Yuriy↗
Wheel-flow gaseous-core reactor concept
Wheel-flow gaseous-core reactor concept for space vehicle propulsion
SPACECRAFT PROPULSION↗
Successes and Lessons Learned in the Design, Construction and Use of a Tool for Sampling the WR-1 Reactor Core - 1900z
Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on an organically cooled, heavy water moderated nuclear reactor in the world, Whiteshell Reactor 1 (WR-1). WR-1 was safely shut down in 1985, de-fueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed In Situ Decommissioning of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, pressure tubes) was necessary to verify inventory estimates and neutron activation models produced circa 1992. The Fuel Channels of WR-1 (pressure tubes) were expected to be the most activated components of the reactor core, and to account for about 60% of the reactor core inventory. The interior surface of the fuel channel is also the most readily accessible surface within the reactor core. As such the decision was made to collect a sample of one of each of 3 Fuel Channel material types (Stainless Steel, Zr-2.5Nb, Ozhenite) for comparison to modelled estimates of neutron activation of various core materials. The expected dose rates from the samples were between 5 and 100 mSv (500-10,000 mrem) on contact based on a 200 mg sample. Sampling the Fuel Channels required verification the Fuel Channel was drained of organic coolant, removal of residual organic coolant from the Fuel Channel wall, removal of material from the interior surface of the Fuel Channel at the centerline of flux (∼25' below the operator position), and extraction of that material from the core. Several options for existing tooling and technologies, including hydraulic scrapers, and horizontal drill technologies, were looked at for their applicability to WR-1. While suitable approaches to higher radiation field sample collection, they were overly complex, costly and time consuming for the configuration at WR-1. In an effort to reduce costs, and to simplify the sampling process, a specialized and simple tool was developed to mechanically scrape a small sample of the interior wall of the Fuel Channels, collect the sample into a vial, and remove the sample from the core for analysis. The tool was designed, constructed, tested and implemented in the field entirely by Canadian Nuclear Laboratories Staff. The tool required no powered components, or electronic guidance or measurements; and was designed and constructed for a fraction of the cost of other commercially available technologies for sample collection within reactor cores. The samples were successfully collected without any injuries, contamination events or equipment failures. The tool remains operational in the event additional sampling is required. This paper provides additional details of the design, construction, testing and use of the tool and the successes and lessons learned during the various stages of design and implementation. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Experimental study of effect of simulated reactor core position on nozzle heat transfer
Reactor core position influence on conical nozzle heat transfer and boundary layer characteristics
HEAT TRANSFER↗
Energetic closed-cycle gas core reactors for orbit raising
Closed-cycle gas core reactor power plants can be of two types. In the 'mixed flow' type, the gaseous nuclear fuel is intimately mixed with the working gas in the cavity. In the 'light bulb' type the fissioning plasma is enclosed in a transparent tube, and energy transfer to the separate working gas occurs by thermal radiation. The potentials of high temperature gas core reactors in terrestrial electric power generator applications have been considered, and a number of civilian power-beaming applications for gaseous fuel nuclear-MHD power plants in space have been suggested. Major conclusions of investigations related to the design of space power systems are discussed. Attention is given to options for conversion cycles, the power system specific mass, and research and technology issues.
Rosa, R. J.↗