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DOE OSTI2015

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Parameters of a Pulsed Reactor for Pumping a Large-Size NPL

The design of a fast pulsed reactor with a large axial cavity that can contain a multichannel laser module to generate high-power pulses of optical radiation is proposed and substantiated by calculations. The results of calculations of the reactor parameters, as well as those of displacements and tensions in its fuel rods, are provided. Preliminary conclusions concerning the procedure of generation of fission pulses in the proposed reactor are formulated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Light Water Reactor Sustainability Program: Integrated Program Plan

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the United States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy’s 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; 3. Develop sustainable nuclear fuel cycles; and 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however, is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

High-Temperature Gas-Cooled Reactor Research Survey and Overview: Preliminary Data Platform Construction for the Nuclear Energy University Program

Since the U.S. Department of Energy Office of Nuclear Energy initiated the Nuclear Energy University Program (NEUP) in 2009, there are 29 NEUP projects focusing on high-temperature gas-cooled reactor (HTGR) research up to July 2022. The resultant research product, either experimental or computational, were published as final NEUP reports, journal articles and conference proceedings. However, these federally funded products have been scattered and sometimes cannot be easily accessed. To improve access to this valuable HTGR validation data and optimize the return on the significant investment made by the Department of Energy, the Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program started a survey of completed and ongoing HTGR NEUP projects to develop a public-access database specific for HTGRs applications that can be used to retrieve computational fluid dynamics and system code validation data. This effort will help guide future NEUP-funded research, define new state of the ART Phenomena Identification and Ranking Table (PIRT), and promote the usage of this data in the codes validation matrices. This report provides an overview of the NEUP-funded HTGR-related research projects from Fiscal Year (FY) 2009–2021 and identifies validation knowledge gaps still existing in HTGR thermal-fluid research. A preliminary data platform has been developed for the 29 NEUP projects investigating HTGR thermal hydraulics, including their final reports as well as the available scientific publications. As an ultimate goal for this work, the ART-GCR program will create a central database at Idaho National Laboratory to identify, organize, and store these datasets generated by experimental investigations or computational models, experimental facility descriptions, and publicly-available academic products from the HTGR-related NEUP projects and provide future guidance for the storage and transmission of important project documentations for later NEUP projects as well.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Feasibility study of Self Powered Neutron Detectors in Fast Reactors for detecting local change in neutron flux distribution

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor system. Diverse possibilities of detector systems installation have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local change in neutron flux distribution. We show that the gamma contribution from fission products decay in the fuel and activation of structural materials is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Finite Volume Method Based Neutronic Solvers for Steady and Transient Analysis of Molten Salt Reactors

In recent years, the coupled neutronics and thermal hydraulics analysis becomes more and more important, because it helps to identify the most relevant safety issues without conservative assumptions. In addition, such coupled simulations are essential for the generation IV advanced nuclear power systems such as molten salt reactor (MSR), because the presence of strong coupling between different phenomena, like the transport of delayed neutron precursors by the fluid flow?. Due to the different discretization methods used in reactor physics analysis and thermal-hydraulics analysis, it is hard to couple different kinds of codes. This work presents an attempt to solve the multigroup neutron diffusion equation with the finite volume method which is quite widely used for the numerical simulation of a variety of applications involving fluid flow and heat and mass transfer. The unstructured finite volume method (FVM) based on open source C++ library OpenFOAM was chosen as the development platform?. The objective of this paper is to concisely present the main features of the developed codes in terms of the theoretical background and numerical approach and as can be seen in the following sections, the calculation results obtained by the NDSFoam which solves the steady-state neutron diffusion equation and NDTFoam which solves the transient-state neutron diffusion equation are discussed and compared to the benchmark problems. In this paper, two new neutronic solvers using finite volume method have been developed to perform the neutron physics analysis. The solvers are based on OpenFOAM, and have been tested with several steady state and transient-state benchmark problems. By verifications of the codes, it is proved that the development of NDSFoam and NDTFoam are correct. In our future work, both of these two solvers will be coupled with the precompiled CFD solvers in OpenFOAM to analyze the molten salt fast reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Characterizing the Thermal Column Beam in the University of Massachusetts Lowell Research Reactor

The University of Massachusetts Lowell operates a 1 MW, light water cooled and moderated Research Reactor (UMLRR) fueled by low-enriched uranium to generate neutron flux that is used at a variety of in-core and beam port facilities within the reactor to perform irradiation experiments for academic and industrial research. One of the irradiation sites is a graphite-moderated thermal column designed to deliver a collimated beam of thermal neutrons to a neutron radiography imaging system. A detailed characterization of the neutron beam is essential for the design and interpretation of experiments at the facility and for radiation safety. This work aimed to characterize the neutron flux and energy spectrum in the UMLRR thermal column beam by employing experimental and computational methods known as the foil activation and spectrum unfolding technique. A previously existing Monte Carlo N-Particle transport (MCNP) model of the UMLRR reactor core has been extended to the thermal column where the neutron flux and energy spectrum have been computed and experimentally validated. This work employed the foil activation and spectrum unfolding technique along with neutron transport computation to extend the knowledge of the neutron fluence and energy spectrum from the UMLRR core to the adjacent thermal column beam. Radiography systems are used for materials characterization, studying dynamic behavior of engineering systems, and for nuclear forensics applications. Foil activation is an experimental method where thin foils of highly pure materials are placed in the beam where interactions between the neutrons and the foil material activate radioisotopes in the foils whose activities can be measured after the irradiation. Downstream computation with MCNP and the spectrum unfolding code SAND can deconvolute the neutron fluence and energy spectrum that must have been necessary to induce the production rates of radioisotopes that were obtained through foil activation. The necessary ingredients for successful spectrum unfolding are the experimentally measured production rates and some a priori (initial guess) knowledge of the flux spectrum. The thermal column is equipped with a neutron radiography imaging system, so any experimenters who desire to use neutron radiography at UMass Lowell will benefit significantly from knowing the characteristics of the neutron beam. Comprehensive knowledge of these characteristics of the beam is essential to compare and benchmark results obtained for similar experiments at other facilities. The Integrated Nuclear Security and Safeguards Laboratory (INSSL) is pursuing the investigation of the performance of self-healing materials under stress-strain environments and the characterization of radiation detectors and special containers for storing and transporting nuclear material. This work contributes to the improvement of the current radiography system, expands the utility of the system for nuclear safeguards and security application and provides information that will enhance the general safety of the UMLRR experimental facilities. The main objective of this work has been achieved; an appropriate neutron flux spectrum for the UMLRR thermal column beam has been determined using the foil activation method followed by spectrum unfolding. There now exists an experimentally validated computational model of the neutron flux spectrum and the thermal column's neutron environment is characterized. We can conclude that the thermal column is well characterized; an appropriate neutron flux spectrum has been determined through an experimentally validated computational model. Future experimenters interested in the thermal column may now expand on the computational model or use the spectral information for their own purposes. The knowledge gained herein will be helpful for the INSSL's future work using neutron radiography to analyze the effectiveness of self-healing materials and special containers. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Mechanistic source terms for the XE-100 reactor XSTERM Code Development Approach - 18587

X-Energy is developing the Xe-100 pebble bed modular High Temperature Gas Reactor (HTGR) as an advanced nuclear power source. To achieve its safety objectives, the Xe-100 reactor relies on materials inherent characteristics and passive safety features. Key features include the inherent high temperature characteristics of TRISO-coated fuel particles, graphite moderator, and helium coolant, along with passive heat removal capability of a low-power-density core. This paper will describe the approach adopted by X-Energy to develop and validate mechanistic source terms for the Xe-100. The Mechanistic Source Term (MST) approach for the Xe-100 reactor will build upon that of the prior DOE Next Generation Nuclear Program (NGNP) (Refs. 1, 2, 3 and 4). MST development is closely related to Fuel Qualification (FQ), Licensing Basis Event (LBE) selection and Probabilistic Risk Assessment (PRA). This paper provides a brief overview of the MST road map process. Software plays a crucial role in establishing and quantifying MSTs required for licensing the Xe-100. An insight into X-Energy s comprehensive, integrated software development program for calculating time-dependent radionuclide releases from the Xe-100 over a wide range of events is provided. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Primary coolant flow measurement for integral pressurized water reactors using ultrasonic technique - 182

Several pressurized water reactor (PWR) systems under development employ integral primary coolant loop designs, where the primary coolant does not leave the reactor pressure vessel during operation (I{sup 2}S-LWR, mPower, NuScale). The absence of primary coolant piping inhibits traditional pressurized water reactor primary coolant flow measurement, in which the differential pressure across the steam generator is correlated to flow rate. To address this, a transit time, reflection mode, ultrasonic flow meter is proposed. An experimental flow loop utilizing a test section representative of a thin azimuthal slice of the down-comer region of an integral PWR has been constructed to test the efficacy of the General Electric AT600 ultrasonic flow meter (UFM) in such an application. A Great Plains Industries TM300 turbine flow meter (TFM), reporting volumetric flow rate and installed in-line with the flow loop piping, is used as the reference meter. UFM performance is affected by several factors, including flow development and symmetry. The UFM was installed on three locations along the flow path of water through the test section. The meter performed the best at the downstream end of the test section, where the fluid flow is the most developed. Performance, measured by percent error in UFM average reading with respect to the TFM average reading, generally improved with increasing flow rate at all locations, though absolute difference between meter readings increased with flow rate. This research suggests that ultrasonic approaches hold promise for non-invasive measurement of primary coolant flow in integral PWRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Overview of simulation and control tools for extended operability of nuclear reactors - 183

The increasing penetration of Renewable Energy Sources (RES) requires the current power grid to be suitably improved. Because of the unsteady nature of RES, as long as energy storage facilities are not deployed and commercialized, an increased effort for the frequency regulation in order to ensure high reliability and performance is necessary. In such a scenario, nuclear units (traditionally deployed as baseload generating units) will be required to cooperate with intermittent RES-based power plants. Therefore, new reactors will need larger operational flexibility with respect to the capabilities of the currently operated units. Over the last few years, a comprehensive research program oriented at developing a suite of simulation tools and control system development capabilities was pursued at Argonne National Laboratory. In addition to the study of advanced reactor concepts (e.g. metal-fueled Sodium-cooled Fast Reactors) and technological innovations (such as S-CO{sub 2} Energy Conversion Cycles), new control approaches were developed to improve the operability of nuclear power plants. In particular, Model-based Predictive Control algorithm for Multiple Input Multiple Output control strategies and set-point definition algorithm based on Reference Governor were studied and applied to specific scenarios, providing with a range of novel capabilities able to meet the expected greater sets of requirements mentioned above. In this work, an overview of such tools and case studies is presented with particular focus on their impact on operability and economic competitiveness of nuclear energy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

VASIMR{sup R} Interfaced High Temperature Gas-Cooled Fast Reactor for Space Applications

This project was undertaken to assess the feasibility and advantages of utilizing a gas-cooled fast reactor to supply power to an electric propulsion system for deep space exploration to Europa, one of Jupiter's moons. Specifically, the primary objective of this project was to design a 9 MWe gas cooled fast reactor with an accompanying power conversion system to supply electricity to multiple electrically-powered thrusters for a four year mission duration. The thruster design chosen for this analysis is the Variable Specific Impulse Magnetoplasma Rocket (VASIMR). The VASIMR propulsion system, currently being developed by Ad Astra Rocket Company (AARC), is an electric propulsion rocket that uses electric power to accelerate neutral plasma propellant. Advanced electric space propulsion systems such as the VASIMR offer unique capabilities that allow for higher exhaust velocities that are up to ten times that of traditional chemical rockets. Furthermore, the thrust and specific impulse of the VASIMR system can be varied in order to achieve optimal travel times to given destinations for a desired payload. As a result, electric propulsion systems characteristically have lower ratios of system mass to power output than chemical rockets. This ratio is commonly known as the alpha parameter. Alpha parameters of approximately 10 kg/kWe or less make electric propulsion more attractive for deep space exploration missions in terms of resource efficiency and cost. However, the primary challenge with electric propulsion is to develop a high power density source of electricity in order to minimize the total system mass. The alpha parameter is the main benchmark used in this project to test whether or not the proposed coupled gas cooled reactor and power conversion system design is more competitive than other proposed systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗