Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Small Modular Reactor”

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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
NASA NTRS2020

Kilopower Project: The KRUSTY Fission Power Experiment and Potential Missions

The Kilopower Project was initiated by NASA’s Space Technology Mission Directorate/Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Kilopower Project centerpiece is the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) test, which consists of the development and testing of a ground technology demonstrator of a 1-kW(electric)–class fission power system (FPS). The technologies to be developed and validated by KRUSTY are extensible to space FPSs from 1 to 10 kW(electric), which can enable modular surface FPSs for human exploration as well as higher-power future potential deep space science missions. The KRUSTY demonstration is cofunded by NASA and the U.S. Department of Energy National Nuclear Security Administration. The KRUSTY demonstration in the National Critical Experiment Research Center’s Device Assembly Facility was completed in the first quarter of 2018.

Space nuclear reactor↗
NASA NTRS2017

Kilopower: Small and Affordable Fission Power Systems for Space

The Nuclear Systems Kilopower Project was initiated by NASA's Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project centerpiece is the Kilopower Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a fission ground technology demonstrator of a 1 kWe-class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable higher power future potential deep space science missions, as well as modular surface fission power systems for exploration. The Kilopower Project is cofounded by NASA and the Department of Energy National Nuclear Security Administration (NNSA).KRUSTY include the reactor core, heat pipes to transfer the heat from the core to the power conversion system, and the power conversion system. Los Alamos National Laboratory leads the design of the reactor, and the Y-12 National Security Complex is fabricating it. NASA Glenn Research Center (GRC) has designed, built, and demonstrated the balance of plant heat transfer and power conversion portions of the KRUSTY experiment. NASA MSFC developed an electrical reactor simulator for non-nuclear testing, and the design of the reflector and shielding for nuclear testing. In 2016, an electrically heated non-fissionable Depleted Uranium (DU) core was tested at GRC in a configuration identical to the planned nuclear test. Once the reactor core has been fabricated and shipped to the Device Assembly Facility at the NNSAs Nevada National Security Site, the KRUSTY nuclear experiment will be assembled and tested. Completion of the KRUSTY experiment will validate the readiness of 1 to 10 kWe space fission technology for NASAs future requirements for sunlight-independent space power. An early opportunity for demonstration of In-Situ Resource Utilization (ISRU) capability on the surface of Mars is currently being considered for 2026 launch. Since a space fission system is the leading option for power generation for the first Mars human outpost, a smaller version of a planetary surface fission power system could be built to power the ISRU demonstration and ensure its end-to-end validity. Planning is underway to start the hardware development of this subscale flight demonstrator in 2018.

Space nuclear power↗
NASA NTRS1972

Completely modular thermionic reactor ion propulsion system /trips/.

Description of a nuclear reactor powered ion propulsion system which is an advanced completely modularized system lending 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 the thermionic fuel elements, the heat rejection subsystem (head pipes), the power conditioning modules, and the ion thrusters. 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 which allows complete prototype testing. The ion engines are typically matched with one or more p/c modules and are the same size for any power level propulsion system of interest.

Peelgren, M. L.↗
DOE OSTI2023

The Princeton Field-Reversed Configuration for Compact Nuclear Fusion Power Plants

The Princeton Field-Reversed Configuration (PFRC) nuclear fusion reactor concept is an innovative approach to fusion power generation prioritizing low neutron production and small size. Here, a combination of analytical modeling and numerical simulation shows that the novel heating approach generates an FRC with closed field lines. Simulation data from a single-particle Hamiltonian code predicts ms-scale plasma heating in reactor-scale conditions while PIC codes predict formation of warm FRC plasmas from initial mirror fields. The PFRC-1 and PFRC-2 experiments have heated electrons to energies well in excess of 100 eV and plasma durations to 300 ms, more than 10 4 times longer than the predicted tilt instability growth time. From these data, we have created a development plan and anticipated performance metrics for a fusion reactor based on the PFRC concept. The resulting 1–10 MW PFRC reactors would be suitable for diverse applications, from submarines to urban environments to space propulsion. PFRC is a steady-state, driven magnetic confinement device. Plasma, inside a cylindrical array of coils, is confined and heated by external RF antennae. PFRC would be ultra-low radiation due to both its fuel and small size. The choice of advanced fuels, deuterium and helium-3 (D– 3 He), may be enabled by the high-β FRC configuration. The small size of the reactor would enable rapid exhaust of the dangerous tritium ash. Low radiation would make the reactor safer to operate and, in combination with simple geometry and small size, dramatically lowers development and maintenance costs. This review paper gives an introduction to the physics of the PFRC and a summary of the PFRC-2 experiment results to date. It then discusses the future program plan and how PFRC reactors would be commercialized.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Electrogenerative Reactors for Process Intensified Cogeneration of Power and Liquid Fuel from Shale Gas

With the abundance of natural gas and its primary components of methane, ethane and propane, there is the opportunity to convert these light alkanes to higher-value products. The growth in natural gas supply in the U.S. has increased substantially since the mid-2000s due to advances in horizontal drilling and hydraulic fracturing. As petroleum prices are volatile, natural gas resources are progressively attractive as alternative fuel sources. However, many of these shale gas wells are in regions of the U.S. that are remote and costly to transport to petrochemical and refining centers. In addition, much of this shale gas is “wet gas” that contains condensable amounts of ethane. While dedicated ethane gas pipelines have been built to get a small fraction of this gas to market, they are very capital intensive and compete for existing land use. As a result of this Phase I project, we developed a process to convert ethane to a transportable liquid fuel. This was achieved by implementing process intensification for this gas-to-liquids approach; we use novel eletrogenerative reactors in a two-step process rather than multi-step well-known technologies such as Fisher-Tropsch synthesis. These reactors borrow on well proven solid oxide fuel cell technology. Therefore, when produced in mass will be a very low-cost modular chemical reactor. This project demonstrated that EcoCatlaytic’s catalysts are highly selective towards ethylene and are compatible with anode materials. These integrated catalysts with the anode were able to maintain stable activity under ethane. Process simulations demonstrate an energy savings while producing gasoline and electricity for a modular system that produces 1,000 bbl/day gasoline. The system, not only produces gasoline, but also generates electricity which is a very valuable commodity in the remote shale gas facilities. This process could potentially allow remote natural gas producers to upgrade their feedstock to valuable gasoline on-site rather than flaring.

03 NATURAL GAS↗
DOE OSTI2019

Multiphysics Simulation of Moisture-Graphite Oxidation in MHTGR

A small amount of moisture on the ppm level could present in the primary helium coolant under normal operations of High-Temperature Gas-cooled Reactors (HTGRs). One potential safety issue related to HTGRs is the chronic moisture-graphite oxidation, which could affect the integrity of the fuel blocks and reflectors in the reactor core, and the support columns in the hot plenum. However, it is infeasible to perform chronic moisture-graphite oxidation tests under the prototypic high-pressure, high-temperature conditions over a long period of time that is comparable to the full service time, i.e., 36 months for Modular High Temperature Gas-cooled Reactor (MHTGR). As an alternate, accelerated moisture-graphite oxidation tests at the atmospheric pressure and high moisture concentrations have been performed in the literature. Through these accelerated tests, global reaction rate equations have been proposed, which makes it possible to study the moisture-graphite oxidation numerically. This paper is aimed to establish a multiphysics model that can evaluate the moisture-graphite oxidation under MHTGR normal operation conditions. COMSOL Multiphysics was applied to couple the fluid flow, heat and mass transfer, chemical reaction and material structural changes. To reduce the calculation time, the prototypic three-dimensional (3-D) structures were simplified into a two-dimensional (2-D) simulation domain. The performance of four nuclear grades of graphite (IG-110, 2114, PCEA, and NBG-17) over a 36-month service period was then investigated under the most likely prototypic conditions of MHTGR. The simulation results indicate that most oxidation occurs in the three or four bottom fuel blocks due to their higher temperatures. In general, however, only a thin layer of 1 mm into the graphite will be considerably oxidized even for those three or four bottom fuel blocks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗
DOE OSTI2016

Materials Irradiation Facility (MIF) at the High Flux Isotope Reactor (HFIR)

The Material Irradiation Facility (MIF) at the Oak Ridge National Laboratory (ORNIL) High Flux Isotope Reactor (HIFR) controls and monitors instrumented irradiation experiments for materials and fuel research. In 2013 ORNIL founded a Laboratory Directed Research and Development project to reinvigorate the use of instrumented experiments by redesigning MIF to enhance its capabilities. MIF improvements include an updated and more sophisticated experiment monitoring system, a modular design allowing easier addition of simultaneous experiments, and features that allow rapid incorporation of additional instrumentation Interest in materials and fuel irradiation has increased significantly over the last decade. Much of this interest is related to investigating the performance of advanced materials for fusion environments, designing and developing improved fuel and cladding materials, and examining material characteristics and behavior under irradiation conditions. At ORNL, the irradiation engineering team has more than 60 years of experience designing and fabricating various types of irradiation experiments. Rabbit capsule experiments are small scale and relatively inexpensive. These capsules are typically designed to be irradiated in the flux-trap region of the core. Full-length target capsules are larger scale and could be either instrumented or non-instrumented. These more complicated capsules can be irradiated in any of the experiment facilities at HFIR. Instrumented experiment irradiation provides significant advantages for researchers. The unique advantage is that the researcher can monitor and acquire in situ data and change capsule operating conditions in real time during the irradiation. This capability enables tight control of desired experimental conditions so that they match the design specifications throughout the irradiation. Also, the experiment conditions (e.g., temperature and pressure) can be altered during the test, enabling significant flexibility and the ability to collect considerably more data from a single experimental capsule. This capability allows measurement of a variety of material properties (conductivity, fission product composition, etc.) that otherwise would only be available upon post-irradiation examination, where non-equilibrium conditions brought on by irradiation do not exist. The data generated by these experiments contribute significantly to the understanding of material behavior under irradiation. MIF is an experimental platform that can host fully instrumented and monitored experiments in a high neutron flux environment. Its primary functions are to provide clean inert gas (helium, neon, argon) to experiments at controlled pressures and flows, monitor irradiation conditions (typically temperatures and pressures) during irradiation, and monitor effluent gases from experiments. This facility allows monitoring of chemical processes during a fueled irradiation, where this becomes important for detecting fuel failures. This paper reports the efforts towards reestablishing the value of HFIR for instrumented irradiation experiments and the improvements to MIF that ultimately reduce the costs of instrumented irradiation. This project advances ORNL materials capabilities to new levels by providing a state-of-the art, robust and flexible facility that can be expanded as customer requirements change, while supporting the goals of the fusion, advanced fuels, and structural materials programs. (authors)

42 ENGINEERING↗
DOE OSTI2020

DEM and MELT In-Can Vitrification Process for Fukushima Daiichi Water Treatment Secondary Waste - 20034

Fukushima Daiichi Nuclear Power Station was severely damaged by a tsunami induced by the earthquake on March 11, 2011. A large amount of water treatment secondary waste has been generated from the purification work of contaminated water that consists of the water flowing into the reactor building. The Mid and Long-Term Road map specifies that solid waste at Fukushima Daiichi Nuclear Power Station should be stored safely and reasonably according to its characteristics. In order to meet this goal, a versatile treatment solution is needed, and the In-Can vitrification seems to be a suitable solution. The DEM and MELT In-Can vitrification process is being developed and qualified in France to treat waste arising from decommissioning and dismantling operations. This qualification is carried out through the consortium of the DEM and MELT project coordinated by CEA which gathers Orano, ECM technologies and Andra. The DEM and MELT In-Can vitrification process is a robust, simple and versatile in situ vitrification process. It is designed for high or intermediate level waste and is compact enough to be implemented in an existing facility or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The DEM and MELT In-Can vitrification process is also developed with a modular design in order to be adapted to nuclear operators needs and requirements. The In-Can melting process has already been developed and fully qualified by the CEA for Alpha liquid waste and can be rated TRL 7 for this scope. The In-Can vitrification technology is considered as a qualified solution that could be applied for the treatment of waste resulting from the D and D work in Fukushima Daiichi Nuclear Power Station. The feasibility evaluation of the DEM and MELT In-Can vitrification process for Fukushima Daiichi water treatment secondary waste is a project led by Orano, CEA and ANADEC. This R and D project is performed by the funds from the Ministry of Economy, Trade and Industry as The Subsidy Program 'Project of Decommissioning and Contaminated Water Management'. This paper presents the first results of this feasibility evaluation, from the scenarios and glass formulation definition to the industrial implementation feasibility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2020

Small Lunar Base Camp and in Situ Resource Utilization Oxygen Production Facility Power System Comparison

This report examines the power requirements for operating an in situ resource utilization (ISRU) oxygen production system on the lunar surface and a small six-person base camp. The baseline ISRU system produced 1.63 kg/h for a total day and night production rate of 1,154 kg. It was estimated that this plant would require 25.83 kW of power to operate. The base camp power includes auxiliary equipment as well as a communications system. The required power estimate for the base camp was 28.05 kW. This estimation was used to size a power system and determine its mass for meeting these requirements. Three types of power systems were considered: a solar photovoltaic (PV) array system using batteries for energy storage, a PV array system using a regenerative fuel cell (RFC) for energy storage, and a modular 10-kW electrical output power Kilopower reactor system. Three separate cases were examined: a stand-alone ISRU oxygen production system, a base camp, and a combined ISRU oxygen production system and base camp. For the PV array-based system, the RFC energy storage method had a mass advantage over a battery- based energy storage system. For higher power nighttime power operation for all three cases, the RFC system’s specific energy was just over 830 Wh/kg. For the lower power nighttime “keep-alive” level used as part of the Case 1 analysis, the specific energy for the RFC was 456 Wh/kg. Both of these levels are significantly above the specific energy of 200 Wh/kg for the battery. Because of this higher specific energy, the RFC-based system provided significant mass advantages over the battery-based energy storage system. The baseline reactor system utilized shielding and separation distance to meet the desired maximum radiation dose level of 5 rem/yr for personnel operating within the vicinity of the power loads, base camp, and oxygen production facility. There are methods that could potentially be utilized to reduce the shielding requirements and separation distance. Implementing these would reduce the overall system mass for the reactor. Also, optimizing the reactor output to a specific mission would provide benefits in mass at the expense of modularity. The results of the power system comparison between a solar PV array-based system and a Kilopower reactor-based system has shown that for missions required to operate throughout the lunar night at power levels comparable to those used during the day, the reactor-based system provides a significant mass advantage. However, for applications that can meet their mission requirements while only having to operate during the daytime with minimal power required to survive the nighttime, the PV array-based system provides a mass advantage.

Colozza, Anthony J.↗
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 OSTI2017

Improvements to PHISICS/RELAP5-3D{sup C} Capabilities for Simulating HTGRs

Accurate, high-resolution full core simulations of prismatic high temperature reactors for burnup and transient analysis remain a challenge. The cores are geometrically large, but some features are quite small (tri-structural isotropic (TRISO) particles, burnable poison pellets). The treatment of both neutron scattering in graphite and resonance capture are complex and not adequately captured using the methods traditionally used in High Temperature Gas-cooled Reactors (HTGR) and commonly used in Light Water Reactors (LWR). For transient analysis, temperature feedback is a function of the TRISO fuel form, but full core models cannot resolve phenomena at this scale without averaging over space and energy. For burnup calculations, a second level of heterogeneity must be resolved to capture the local effects of fuel and burnable poisons, while accurately propagating their effects through and between blocks. The HTGR Methods and Simulation group has been utilizing the INL-developed PHISICS/RELAP5-3D{sup C} (P/R), code suite for participation in the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Gas Cooled Reactor (MHTGR) Benchmark, funded by the US Department of Energy (DOE) under the Advanced Reactor Technologies (ART) program. The MHTGR-350 benchmark consists of several lattice, steady-state and transient exercises, and the P/R code has been successfully applied to most of the cases defined. In addition to this code-to-code verification activity, it is planned that P/R will be validated using data from several Loss of Forced Cooling (LOFC) tests performed at the High Temperature Test Reactor (HTTR) in a joint effort with Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) project. These verification and validation simulations will identify code and model discrepancies, sensitivities, and establish the current state-of-the-art in HTGR simulation. The Department of Astronautical, Electrical and Energy Engineering (DIAEE), 'Sapienza' University has been involved in the development of P/R since 2014 through various graduate and PhD student projects. In the framework of this collaboration the P/R optimization for application on the OECD MHTGR-350 and HTTR models will be carried out.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1989

Coded-aperture imaging in nuclear medicine

Coded-aperture imaging is a technique for imaging sources that emit high-energy radiation. This type of imaging involves shadow casting and not reflection or refraction. High-energy sources exist in x ray and gamma-ray astronomy, nuclear reactor fuel-rod imaging, and nuclear medicine. Of these three areas nuclear medicine is perhaps the most challenging because of the limited amount of radiation available and because a three-dimensional source distribution is to be determined. In nuclear medicine a radioactive pharmaceutical is administered to a patient. The pharmaceutical is designed to be taken up by a particular organ of interest, and its distribution provides clinical information about the function of the organ, or the presence of lesions within the organ. This distribution is determined from spatial measurements of the radiation emitted by the radiopharmaceutical. The principles of imaging radiopharmaceutical distributions with coded apertures are reviewed. Included is a discussion of linear shift-variant projection operators and the associated inverse problem. A system developed at the University of Arizona in Tucson consisting of small modular gamma-ray cameras fitted with coded apertures is described.

Smith, Warren E.↗
NASA NTRS1989

M.I.N.G., Mars Investment for a New Generation: Robotic construction of a permanently manned Mars base

A basic procedure for robotically constructing a manned Mars base is outlined. The research procedure was divided into three areas: environment, robotics, and habitat. The base as designed will consist of these components: two power plants, communication facilities, a habitat complex, and a hangar, a garage, recreation and manufacturing facilities. The power plants will be self-contained nuclear fission reactors placed approx. 1 km from the base for safety considerations. The base communication system will use a combination of orbiting satellites and surface relay stations. This system is necessary for robotic contact with Phobos and any future communication requirements. The habitat complex will consist of six self-contained modules: core, biosphere, science, living quarters, galley/storage, and a sick bay which will be brought from Phobos. The complex will be set into an excavated hole and covered with approximately 0.5 m of sandbags to provide radiation protection for the astronauts. The recreation, hangar, garage, and manufacturing facilities will each be transformed from the four one-way landers. The complete complex will be built by autonomous, artificially intelligent robots. Robots incorporated into the design are as follows: Large Modular Construction Robots with detachable arms capable of large scale construction activities; Small Maneuverable Robotic Servicers capable of performing delicate tasks normally requiring a suited astronaut; and a trailer vehicle with modular type attachments to complete specific tasks; and finally, Mobile Autonomous Rechargeable Transporters capable of transferring air and water from the manufacturing facility to the habitat complex.

Amos, Jeff↗
DOE OSTI2017

Estimation of LOCA break size using cascaded support vector regression - 334

A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Status Report on Development of MARCH Module Designs

The Transient Reactor Test (TREAT) facility was constructed in the late 1950’s, provided thousands of transient irradiations before being placed in standby in 1994, and was restarted in 2017 in order to resume its crucial role in nuclear heated safety research. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modern science-based and separate effects test capabilities at TREAT. An innovative approach to this type of testing leverages the relatively low radioisotope accumulation during brief TREAT irradiations by arranging small fresh fuel specimens in low activation hardware “modules” so that they can be easily extracted and shipped for examination within weeks. The concept was termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, and its maturation began under a Lab Directed Research and Development (LDRD) project starting in fiscal year 2017. Within the first year of this project, enough design and predictive modeling had been performed to prove the principle and ascertain support from major direct-funded DOE fuel development programs including the Accident Tolerant Fuels (ATF) program that sponsored the first modern fueled test to be performed in TREAT using the MARCH system. Since the design was modular in nature, it was straightforward for other sources to adopt and mature viable test modules while the LDRD project continued concurrently for two more years in conceptualizing and demonstrating the viability of other innovative modules. Many of these modules were adopted for continued development and irradiation under the Advanced Fuels Campaign and other direct sources. This report summarizes the MARCH modules that were conceptualized, designed, and deployed to date with the primary aim of providing reference information to future experiment designers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Characterizing the negative triangularity reactor core operating space with integrated modeling

Negative triangularity (NT) has received renewed interest as a fusion reactor regime due to its beneficial power-handling properties, including low scrape-off layer power and a larger divertor wetted area that facilitates simple divertor integration. NT experiments have also demonstrated core performance on par with positive triangularity (PT) high confinement mode (H-mode) without edge-localized modes (ELMs), encouraging further study of an NT reactor core. In this work, we use integrated modeling to scope the operating space around two NT reactor strategies. The first is the high-field, compact fusion pilot plant concept Modular, Adjustable, NT ARC (MANTA) and the second is a low field, high aspect ratio concept based on work by Medvedev et al. By integrating equilibrium, core transport, and edge ballooning instability models, we establish a range of operating points with less than 50 MW scrape-off layer power and fusion power comparable to PT H-mode reactor concepts. Heating and seeded impurities are leveraged to accomplish the same fusion performance and scrape-off layer exhaust power for various pressure edge boundary conditions. Scans over these pressure edge conditions accommodate any current uncertainty of the properties of the NT edge and show that the performance of an NT reactor will be extremely dependent on the edge pressure. The high-field case is found to enable lower scrape-off layer power because it is capable of reaching high fusion powers at a relatively compact size, which allows increased separatrix density without exceeding the Greenwald density limit. Adjustments in NT shaping exhibit small changes in fusion power, with an increase in fusion power density seen at weaker NT. Infinite-n ballooning instability models indicate that an NT reactor core can reach fusion powers comparable to leading PT H-mode reactor concepts while remaining ballooning-stable. Seeded krypton is leveraged to further lower scrape-off layer power since NT does not have a requirement to remain in H-mode while still maintaining high confinement. We contextualize the NT reactor operating space by comparing to popular PT H-mode reactor concepts, and find that NT exhibits competitive ELM-free performance with these concepts for a variety of edge conditions while maintaining relatively low scrape-off layer power.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Analysis of AP1000 Small-Break Loss-of-Coolant Accident Using Reactor Transient Simulator

The Westinghouse Electric Company’s Advanced Passive Reactor (AP1000) is characterized by the incorporation of passive safety systems (PSSs) designed to ensure core cooling during transient events. The assessment of PSSs requires evaluation of their performance through a combination of experiments and simulations employing various thermal-hydraulic codes. In addition, detailed evaluation of PSSs for a specific reactor system transient analysis such as loss-of-coolant-accident analysis supports understanding representative integral effects test facility development and the further evolution model development and assessment process. Developing a reactor system code is a complex and time-consuming process that requires significant engineering expertise and effort. It can take several months to even years to complete in the early stages of reactor system design and analysis. However, this process can be expedited through the use of transient simulator models for similar reactor systems, which can be used for lesson learning and training purposes. This study uses the Personal Computer Transient Analyzer (PCTRAN) code. The main advantage of PCTRAN is its ease of use and ability to run faster than real time. This study presents the results obtained for a small-break loss-of-coolant accident (SBLOCA) for two breaks using the full version (licensed) of PCTRAN. The purpose of this investigation is to evaluate the overall system behavior during the postulated SBLOCA event as well as assess the capability of the PCTRAN code to reproduce the system response during transient events. The obtained results were compared with the Westinghouse NOTRUMP system code. The PCTRAN code proved to be reliable in predicting the qualitative behavior of the system in both transient cases. As for the system response, it was found that it is contingent on the activation time of the PSSs. The differences in reactor coolant system pressure between the two codes were attributed to the critical flow model and simplification of mass and energy balance. Despite PCTRAN’s limitations, it can still provide a reasonable prediction of various reactor parameters such as pressure, mass flow rate, and void fraction during a SBLOCA scenario. It is worth noting that PCTRAN currently employs a bulk approach similar to that of the Modular Accident Analysis Program (MAAP) and MELCOR codes. However, the upcoming version of PCTRAN will include an artificial intelligence–based detection and accident prevention system, as well as different models for different reactor components. Consequently, PCTRAN has the potential to be upgraded to match the system thermal-hydraulic codes of the U.S. Nuclear Regulatory Commission and become more widely used in cybersecurity to safeguard nuclear power plants from cyberattacks.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗