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This project seeks to develop a single vessel for carbon dioxide (CO2) capture and electrolysis for in situ Mars propellant production by eliminating several steps of CO2 processing, two cryocoolers, a high temperature reactor, a recycle pump, and a water condenser; thus greatly reducing mass, volume, and power.
Transient coaxial helicity injection (CHI) current start-up in a spherical tokamak (ST) has been demonstrated for the first time using a reactor-relevant floating single biased (FSB) electrode configuration on QUEST. Unlike previous applications of CHI on HIT-II and on NSTX in which the entire ST vessel was split in two using two toroidal ceramic rings as vacuum breaks, in the FSB electrode configuration on QUEST, only the lower outer divertor plate is electrically insulated from the rest of the vessel. This is the CHI injector region. A CHI discharge is generated by driving current, using an external power supply, along magnetic field lines that connect the inner and outer divertor plates in the CHI injector region. The CHI plasma evolution and closed flux formation on QUEST are very similar to those on NSTX and have generated toroidal currents of 150 kA with 100 kA of current flowing on closed flux surfaces. An alternate, force-free model approach has been used to derive the CHI scaling relations. The resulting CHI scaling relations from the model are consistent with the conventional CHI scaling relations, which state that the CHI generated toroidal current is proportional to the CHI injector flux and the CHI injector current is inversely proportional to the toroidal field. These results from QUEST bode well for the application of CHI in a new generation of compact high field STs and low aspect ratio tokamaks in which the space for the central solenoid is very restricted.
The recent identification of a substantial lunar resource of the fusion energy fuel He-3 may provide the first terrestrial market for a lunar commodity and, therefore, a major impetus to lunar development. The impact of this resource-when burned in D-He-3 fusion reactors for space power and propulsion-may be even more significant as an enabling technology for safe, efficient exploration and development of space. One possible reactor configuration among several options, the tandem mirror, illustrates the potential advantages of fusion propulsion. The most important advantage is the ability to provide either fast, piloted vessels or high-payload-fraction cargo vessels due to a range of specific impulses from 50 sec to 1,000,000 sec at thrust-to-weight ratios from 0.1 to 5x10(exp -5). Fusion power research has made steady, impressive progress. It is plausible, and even probable, that fusion rockets similar to the designs presented here will be available in the early part of the twenty-first century, enabling a major expansion of human presence into the solar system.
Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.
For the temperature range of 900°C − 950°C, Alloy 617 is a candidate IHX structural material for high temperature gas reactors (HTGRs) because of its high temperature creep properties. Also, its superior strength over a broad temperature range also offers advantages for certain component applications. In order for the designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the ASME (American Society of Mechanical Engineers) Boiler and Pressure Vessel Code. A plan has been developed to propose a Code Case for use of Alloy 617 at elevated temperature in Section III of the ASME Code by September 2015. There has not been a new high temperature material approved for use in Section III for almost 20 years. The Alloy 617 Code Case effort would lead the way to establish a path for Code qualification of new high temperature materials of interest to other advanced SMRs. Creep-fatigue at elevated temperatures is the most damaging structural failure mode. In the past 40 years significant efforts have been devoted to the elevated temperature Code rule development in Section III, Subsection NH* of the ASME Boiler and Pressure Vessel Code, to ascertain conservative structural designs to prevent creep-fatigue failure.
This paper presents the structural design and supporting analysis for the Microreactor Applications Research Validation and Evaluation (MARVEL) primary coolant system (PCS) using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Division 5, rules. MARVEL is a liquid metal–cooled microreactor intended to provide experimental capabilities for the rapid testing and development of microreactor technologies. The PCS utilizes high-temperature sodium-potassium liquid metal as the primary coolant and operates at a design temperature of 570°C, necessitating the consideration of creep-related failure mechanisms. The base metal for the PCS is 316H stainless steel, and the weldments are made with a 16-8-2 filler. The design approach incorporates the current base code rules along with ASME code cases N-924, N-861, and N-862 to address primary load, ratcheting, and creep-fatigue evaluations, respectively. The reactor’s operation involves complex thermal and mechanical interactions due to natural convective flow and differential thermal expansion between components. In conclusion, this paper discusses the structural engineering challenges encountered, such as managing thermal stresses in the distribution plenum and guard vessel, and outlines the strategies implemented to meet the code requirements, including design modifications and operational constraints.
The Mechanisms Engineering Test Loop (METL) was built to streamline and accelerate the in-sodium testing of systems and components under conditions that simulate a sodium-cooled fast reactor pool environment. The METL team at Argonne National Laboratory (ANL) can assist experimenters in achieving their technical goals by providing liquid-metal expertise and access to infrastructure required for most alkali metal related research. This document offers a brief overview of METL and provides a basic design guide for researchers interested in conducting research at the facility. Additional information regarding the history and operations of METL can be found in §6.1. Furthermore, high resolution images found in this document as well as CAD files of aforementioned vessels can be provided upon request.
Creep-fatigue (CF) interaction damage is the primary damage mode for high-temperature structural components subjected to cyclic loading. Over the past several decades, researchers within the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5, have focused on developing elevated temperature code rules to ensure conservative structural designs that mitigate CF failure in high-temperature reactors. The existing CF evaluation methodologies in the Code are based on the creep and fatigue damage diagram approach, which is complex and often excessively conservative. The alternative CF evaluation approach proposed here is intended to significantly simplify the evaluation procedure while reducing conservatism in high-temperature component design analysis. This alternative CF evaluation method integrates the elastic–perfectly plastic (EPP) analysis approach with the simplified model test (SMT) CF design concept, leveraging the advantages of both methods. This report presents the preliminary analysis and the approach for developing CF design curves for Alloy 709, utilizing fatigue and CF data generated for the 100,000-hr Code Case to support its qualification to ASME Section III, Division 5 for Class A construction of high temperature reactors. This study is to support the incorporation of Alloy 709 in this alternative CF evaluation method. Recommendations for the remaining work needed to complete the effort are also provided.
The SNAP 8 developmental reactor lithium hydride shield was examined after being irradiated for over 7000 hours at relatively low temperature. A crack was located in the seam weld of the containment vessel, probably the result of hot short cracking under thermal stress. The LiH was visually examined at two locations and its appearance was typical of low temperature irradiated LiH. The adherence of the chrome oxide emittance coating was found to be excellent.
Hydrogen is produced by the solar photolysis of water in a first photooxidation vessel with a transparent wall in the presence of a water soluble photooxidizable reagent and an insoluble hydrogen recombination catalyst. Simultaneously oxygen is produced in a second photoreduction reactor with a transparent wall in the presence of an insoluble photoreduction reagent catalyst. When spent, the solution from the first reactor is fed into the second reactor. A reaction occurs in the dark in which the redox reagents are regenerated, and the regenerated photooxidation reagent solution is recycled to the first reactor. The photoreduction-catalyst is a bifunctional reagent catalyst including a transition metal salt together with a hydroxyl or chlorohydroxyl decomposition catalyst of high area.
A new technology for developing fusion energy is to use hydrogen isotopes i.e., deuterium (D) and tritium (T) which is a combine effort for building up of International Thermonuclear Reactor (ITER) named as Tokamak, which will come into operation in 2020. Handsome amount of work has already been done by many researchers contingent with plasma shape, halo current and plasma equilibrium properties by numerical techniques. Halo current calculation is the most important work in fusion reactors. In this paper, we have calculated halo current and plasma cross-section energy. The research covers calculation of plasma energy and determines the force on the vessel by magnetic field transformation method, which is the essential part of calculating halo current. Further, we have developed VDE up/down halo current models. These models give total magnetic field, asymmetric forces, conducting points and poloidal halo current. Confined plasma by magnetic field is considered to be an abundant and durable source of energy for future aspects. In this paper, plasma magnetic field of each cross section was calculated by using energy integral. Transformation method has been successfully applied which provides the connection between major sideways forces acting on the plasma. Theoretical investigation and analyses of halo current have successfully been done. Two observations were analyzed, large ratio of halo current during VDE down and cancellation of specific forces at conducting point. The results obtained will be useful for equalizing the simulation and experimental results. The research will be extended to developed new models for future fusion research. (authors)
A new technology for developing fusion energy is to use hydrogen isotopes i.e., deuterium (D) and tritium (T) which is a combine effort for building up of International Thermonuclear Reactor (ITER) named as Tokamak, which will come into operation in 2020. Handsome amount of work has already been performed by many scientists contingent with plasma shape, halo current and plasma equilibrium properties by numerical techniques. Halo current calculation is the most important work in fusion reactors. In this paper, halo current and plasma cross-section energy have been calculated. The research covers calculation of plasma energy and determines the force on the vessel by magnetic field transformation method, which is the essential part of calculating halo current. Further, vertical displacement event (VDE) up/down halo current models have been calculated. These models give total magnetic field, asymmetric forces, conducting points and poloidal halo current.
The US fusion community has coalesced around the goal of building an FPP as described by the National Academies of Science, Engineering, and Medicine (NASEM). In addition to demonstrating the viability of the technologies necessary to operate such a plant, including demonstration of net energy and electricity production, NASEM found that a “fusion pilot plant will need to demonstrate the ability to efficiently perform remote maintenance and replacement in support of the design of a power plant, taking into account details of the consequences of the fusion environment, such as material activation and tritium retention in components.” Current designs of fusion demonstration reactors do usually foresee a regular exchange of their first wall modules, including the tritium breeding blankets. In the European Power Plant Conceptual Studies, it is assumed that a fusion reactor will need to change its divertor every 2 years and its first wall blanket module every 5 to 6 years to reach acceptable availability. Underlying this capability are remote-handling technologies to keep the outage for the exchange of these components short. There are many uncertainties in the remote-handling schemes, and most schemes are at a preconceptual level at best. In addition, the exchange of these components would either produce an enormous rad-waste stream or would require an enormous refurbishment activity with huge cost-prohibitive hot-cells. Past Fusion Nuclear Science Facility (FNSF) preconceptual studies have led to hot cell dimensions of an unbelievable size, likely costing tens of billions of dollars. Already at The Way (previously International Thermonuclear Experimental Reactor, ITER), hot-cells have become cost-prohibitive, demanding redesigns of the ITER first wall to reduce the toxic rad-waste/inventory. In this in-situ PFC repair project, a concept for a long-life, maintainable first wall module concept is developed and tested. This first wall concept relies on innovative remote handling to repair the first wall modules in-situ, avoiding costly refurbishments outside of the tokamak vessel. This approach was highlighted in the Fusion Energy Sciences Advisory Committee (FESAC) report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy. In general, the damage of the first wall armor is due to particle and radiation exposures. Load conditions vary from one fusion reactor design to another. In tokamaks, first wall Plasma Facing Components (PFCs) are exposed to far-Scrape-Off-Layer plasma fluxes, electromagnetic radiation, energetic CX neutrals, and potentially runaway electron beams. Protecting the first wall to the worst-case load conditions would require the design of a very thick first wall armor. Transient heat and particle fluxes due to disruptions or edge localized modes can lead to excessive heat loads resulting potentially in melting PFCs down to the cooling channel. Catastrophic events like these need to be avoided by appropriate disruption mitigation systems. However, failure of these systems will still put a first wall at an unacceptable risk. Hence, a first wall design needs to accommodate the occasional transient heat loads by introducing sacrificial limiters, which will absorb these transients.
Steel-plate composite (SC) structures have been used to expedite construction of the third generation of nuclear power plants, for example, consider AP1000® and US-APWR®, both of which use SC walls for the primary and secondary shield walls within the containment internal structures (CIS). Additionally, AP1000® uses an SC shield building outside of the steel containment vessel to provide radiation shielding, earthquake resistance, and aircraft impact resistance.
Presentation on American Society of Mechanical Engineers (ASME) Section III, Division 5 design rules for metallic components at the ASME Boiler and Pressure Vessel (BPV) III International Meeting, September 25 to 28, 2023
It was very hard to estimate how the accident had progressed at each unit of Fukushima Daiichi NPS at the early stage of post-accident analytical activities, because the knowledge relevant to BWR severe accident phenomena and progression was insufficient and the collected data was very limited since complete station blackout (loss of AC and DC power) was occurred at Fukushima Daiichi NPS. However, because of the progression of the measured data analysis, system behavior or response of reactors before the core melt has already been clarified. System behaviors after the core melt have been clarified, as well. Measured data and circumstantial evidence clearly indicated that the fuel temperature went up to melting point and the RPV was damaged by the molten core, and the Fukushima accident is known as a more serious accident than TMI-2 because of including ex-vessel phenomena. But, the information regarding detailed accident progression (for example, amount of melted fuel and timing of RPV breach, etc.) is still unclear and we are struggling to solve such a detailed accident progression. At present, there are two challenges in front of us. One of them is to improve nuclear reactor safety by applying the insights from the Fukushima accident and the other is to solve unclear issues before conducting the first RPV internal investigation. Tokyo Electric Power Company is willing to stand up against these two challenges as the company concerned and appreciate supports of the world nuclear community. (authors)
A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF{sub 6} instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.
Vertical position stability plays a crucial role in maintaining safe and reliable plasma operation for long-pulse fusion devices. In general, the vertical position is measured by using inductive magnetic coils installed inside the vacuum vessel; however, the integration drift effects are inherent for steady-state or long-pulse plasma operation. Further, developing a non-magnetic approach provides a fusion reactor-relevant steady-state solution that avoids the negative impact of integration drift. In this paper, we compare the non-inductively determined vertical position achieved by line-integrated interferometer and polarimeter measurements to that employing an inductive flux loop for a 1056 s discharge recently achieved on EAST (Experimental Advanced Superconducting Tokamak). Experimental results show that the non-inductive measurement is more robust than flux loops after 300 s if the integrator is not reset to suppress integrator drift. Real-time vertical position control using the non-inductive system is proposed for the next EAST experimental campaign.