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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.

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Use of freeze-casting in advanced burner reactor fuel design

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

U.S. HIGH PERFORMANCE RESEARCH REACTOR PRELIMINARY DESIGN MILESTONE FOR CONVERSION TO LOW ENRICHED URANIUM FUEL

The U.S. reactors that regularly refuel with highly enriched uranium (HEU) have undertaken conversion to low-enriched uranium (LEU) using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo) that is under development for the U.S. High Performance Research Reactors (USHPRR). The U.S. National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program is pursuing USHPRR conversion. The USHPRR consists of nine national laboratories and three universities working with a uranium processing plant and commercial fuel fabricator to convert five reactors and an associated critical assembly. Feasibility designs were completed in 2009 for all USHPRR. Since then, the USHPRR have completed key milestones leading to preliminary designs. Between 2007 and 2011, the Fuel Qualification (FQ) Pillar conducted seven irradiation campaigns consisting of 14 large size plates and over 60 mini plates with U-10Mo fuel with a zirconium interlayer. The Preliminary Report on U-Mo Monolithic Fuel for Research Reactors submitted for review to the U.S. Nuclear Regulatory Commission (NRC) in 2017 summarizes these irradiation tests. The Fuel Fabrication (FF) Pillar has installed and is operating a fabrication pilot line at BWX Technologies, Inc. In tandem, the Reactor Conversion (RC) Pillar, has been working with each USHPRR to mature conversion designs and to include recent data. As of 2018, all five USHPRR have completed LEU designs. This includes three Preliminary Safety Analysis Reports (PSARs) submitted between 2014 and 2017 to the NRC by the National Institute of Standards and Technology, University of Missouri, and Massachusetts Institute of Technology for their reactors NBSR, MURR, and MITR, respectively. In 2018, the two USHPRR reactors regulated by the U.S. Department of Energy (DOE) at Idaho National Laboratory (ATR) and Oak Ridge National Laboratory (HFIR) completed optimized conceptual or preliminary designs, with ATR confirming a ‘base fuel’ design of the same fabrication complexity as required for MURR, MITR and NBSR. Instead of available U-10Mo designs, HFIR is pursuing a longer silicide (U3Si2) fuel design with a fuel meat density of 4.8 gU/cm3. However, HFIR conversion will require additional fuel qualification due to the complex fuel design containing boron and since HFIR exceeds existing silicide qualification limits. These preliminary designs allow for a commercial fabrication demonstration, and subsequent irradiation qualification of these fuel element designs in order to complete final SARs that allow for USHPRR conversion to LEU fuel.

Wilson, E. H.↗
DOE OSTI2012

Generic small modular reactor plant design.

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Cipiti, Benjamin B.↗
DOE OSTI2018

Design of a Mixed-Spectrum Reactor With Improved Proliferation Resistance for Long-Lived Applications

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are then shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Griffin Capability Improvements in Support of Ex-core Deep-Penetration Problems

Advanced reactor designs, especially portable reactors that are designed to be located closer to humans and operate autonomously, require the ability to accurately compute the ex-core neutron and gamma flux solutions in terms of shielding design optimization to reduce dose rates at the vessel boundary and detector signal prediction to drive the reactor control system. The Nuclear Energy Advanced Modeling and Simulation program has prioritized improvements to the Griffin discrete ordinates (SN) solver for deep-penetration problems in fiscal year 2025. Significant advancements have been made to the Griffin methodologies for solving ex-core deep-penetration problems for steady-state, fixed-source and transient calculations. This work presents the methodology improvements as well as a comprehensive demonstration with a Transient Test Reactor model and measurements.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Supply Security in Future Nuclear Fuel Markets

Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Supply Security in Future Nuclear Fuel Markets

Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Preliminary design of reactor power systems for the manned space base.

The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

Mckhann, G. G.↗
DOE OSTI2018

Manufactured chemistry: Rethinking unit operation design in the age of additive manufacturing

Advanced manufacturing enables new approaches to chemical reactor design due to the ability to build complex geometries and topologies unaccessible to traditional manufacturing technologies. These benefits have been demonstrated in the design and manufacture of novel heat exchangers and fluidic devices. However, due to prevailing heuristics in chemical reactor design, many of the benefits of advanced manufacturing have yet to be realized by the chemical engineer. As such, in order to reap these benefits, a formalized design methodology such as topology optimization should be employed for the development of high-efficiency small scale modular reactors capable of being manufactured at large volumes. In this way we may be able to start to descend fast manufacturing learning curves and unleash a new wave of innovation in the chemical engineering sciences.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

Preliminary results of AGC-4 irradiation in the advanced test reactor and design of AGC-5 - 18469

The U.S. Department of Energy Advanced Reactor Technologies Program will irradiate up to six nuclear graphite creep experiments in Idaho National Laboratory Advanced Test Reactor (ATR). The graphite experiments are being irradiated over an approximate 8-year period to support development of a graphite irradiation performance database on new nuclear grade graphites now available for use in high-temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the very high-temperature gas reactor and other future gas reactors. Each of the experiments consist of a single capsule that contains six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load and the other half of specimens not subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of the diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with online temperature and compressive load monitoring and control. Also, samples are taken of sweep gas effluent to measure any oxidation or off-gassing that may occur during initial startup of the experiment. The first experiment Advanced Graphite Capsule (AGCJ-I started irradiation in September 2009 and completed in January 2011; the second experiment AGC-2 started irradiation in April 2011 and completed in May 2012; the third experiment AGC-3 started irradiation in late November 2012 and completed in April 2014; and AGC-4 is currently being irradiated in ATR. This paper will briefly discuss preliminary irradiation results from the AGC-4 experiment and design of AGC-3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Design and analysis of embedded I and C for loop-scale magnetically suspended pump - 261

Improving nuclear reactor designs and fuel processing technologies for safer and more efficient operation require the development of new component designs. In particular, many of the advanced reactor design such as the molten salt reactors and high-temperature gas cooled reactors have operating environments beyond the ability of many currently commercially available components. To address this, new cross-cutting technologies need to be developed that will enable the design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operating in these extreme environments. Under this initiative Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump testbed with submerged magnetic bearings. The eventual goal is to apply these techniques to a high temperature (700 deg. C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper we will detail the design and analysis of the embedded I and C testbed with submerged magnetic bearings. Then we will analyze the fluid forces on the shaft. Finally, we will develop the radial and thrust bearing geometries needed to meet the force requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Safety analysis of the FASt TEst reactor (FASTER) preconceptual design

Here, the FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Simulation of Isotopic Concentrations and Gamma Spectra from Dynamic Fission Sources

A tool was developed to rapidly generate synthetic gamma-ray spectra to evaluate safeguards material control and accounting methods for liquid-fueled molten salt reactors. Molten salt reactor operations pose unique challenges to nuclear safeguards methods and protocols compared to deployed reactors designs (e.g., light water reactors). This research evaluates the use of gamma-ray spectroscopy to monitor fission product isotopic flow through a reactor model to understand expected operations and investigate changes to the spectra with material diversion scenarios. The large design space of molten salt reactors (e.g., liquid-fueled, liquid-cooled, online separations) could potentially lead to many measurement points within the reactor system. The developed analytical tool generates and evaluates synthetic gamma-ray spectra from dynamic reactor simulations by extracting isotopic inventory to generate source terms. An implementation in the Gamma Detector Response and Analysis Software (GADRAS) Application Program Interface (API) uses the source terms, a model of the reactor component, a detector response function, and measurement plan to quickly generate and analyze spectra. Prospective measurements are then evaluated in the more accurate but slower Geant4 simulations. Utilization of the developed modeling tool and analysis of the subsequent spectra enables optimization of collimation, shielding requirements, and expected count rates that are used to determine key measurement points in the modeled reactor design.

O'Brien, Sean↗
DOE OSTI2022

Conceptual design of inverted core lead bismuth eutectic fast reactor for marine applications

The development of an inverted core fast reactor aims to generate 60 MWth for about 30 Effective Full Power Years without refueling. The reactor design is a transportable reactor using UO{sub 2} fuel and lead-bismuth-eutectic cooled designed for marine applications and is intended to improve the reactor performances compared to the normal core design: better condition for passive cooling system capability by lower core pressure drop, taking advantage of potential power uprate from the lower maximum fuel temperature. Systematic design processes are presented in this work: fuel pin geometry selection, fuel assembly (FA) design, and core design. A relationship between pressure drops, coolant velocity, maximum fuel temperature, coolant channel diameter, and fuel volume fraction was introduced in a single graph used as a tool to select fuel pin geometry. Fuel fabrication capability also took place in consideration of FA design which led to 7 holes per FA, and two-dimensional temperature distribution studies were also carried out. Core design processes including radial zoning, axial zoning, and core optimization were conducted using Monte Carlo code MCS, which is UNIST CORE laboratory in-house code. The current core design uses 3 fuel enrichment levels and 3 FA types to control the local power distribution and power shift during its lifetime. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Efficient low-temperature, catalyst-free dehydrogenation of alkanes

Steam cracking of ethane, a non-catalytic thermochemical process, remains the dominant means of ethylene production. The severe reaction conditions and energy expenditure involved in this process incentivize the search for alternative reaction pathways and reactor designs which maximize ethylene yield while minimizing cost and energy input. According to the present invention, ethylene yields as high as 68% were obtained with a quartz open tube reactor without the use of a catalyst or a cofed stream of oxidizing agents. The open tube reactor design promotes simplicity, low cost, and negligible coke formation. Reactor designs can be optimized to improve the conversion of ethane to ethylene via non-oxidative dehydrogenation, an approach which shows promise for decentralized production of ethylene from natural gas deposits.

Riley, Christopher Ryan↗
DOE OSTI2019

Canadian Nuclear Laboratories Waste Management: Challenges and the Integrated Waste Strategy - 19160

Canadian Nuclear Laboratories (CNL) currently faces various waste management challenges and CNL's Integrated Waste Strategy is adapting to manage them. As Canada's premier nuclear science and technology organization, CNL is a global leader in developing peaceful and innovative applications from nuclear technology through its expertise in the scientific field, including physics, metallurgy, chemistry, biology and engineering. As demands change, CNL is shifting its focus to alpha therapies for cancer treatments and sustainable energy solutions. The next decade will see CNL decommission the sites it manages across Canada, whilst transforming Chalk River Laboratories (CRL) into a new, world-class facility which will have the capability to demonstrate the commercial viability of advanced reactor designs including Small Modular Reactors (SMRs). This is being enabled through an accelerated decommissioning program for all CNL operated sites, while renovating CRL, which includes revitalizing essential site infrastructure, and investment in new, world-class science facilities in a modern campus environment. With decommissioning and revitalization occurring simultaneously, the long-term strategy and vision for CNL presents some unique challenges to waste management. Therefore, integrated, collaborative and coordinated waste management is a vital component to ensuring a successful future. The Integrated Waste Strategy coordinates, and provides alignment of, waste management across CNL and: - Includes all CNL operated (Atomic Energy of Canada Limited (AECL) owned) sites; - Considers all CNL managed waste (including radioactive and non-radioactive operational, decommissioning, legacy and historic waste) for the planned lifetime of the sites; - Defines waste routes using a life cycle approach (from the generation of waste to final disposal) A baseline Integrated Waste Strategy and associated action plan to mitigate identified gaps and issues have been produced. A program of work is underway to close those gaps identified in a consistent, transparent, prioritized and justified manner. One of the key challenges is to ensure that waste management planning is completed effectively and in a timely manner. This will enable the provision of suitable and sufficient waste capability to meet CNL's long-term strategy. The Integrated Waste Strategy has been developed and maintained through engagement at all levels across CNL, ensuring buy-in of proposed solutions. Key process stakeholders include generators of CNL managed waste to ensure integrated consideration of legacy, current and future waste generation; waste management personnel to identify requirements, including processing, storage, transport and disposal; AECL as the client and waste liability owner; and the public to ensure they understand the 'big picture' regarding waste strategy. (author)

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

Catalytic Oxygen Reduction for Deep O 2 Removal from CO 2 Streams

Streams of CO 2 from various capture processes may contain several impurities of which O 2 is often overlooked as a problematic impurity. Because of its reactivity, the National Energy Technology Laboratory has recommended a stringent limit, less than 10 ppm, for the O 2 level in treated CO 2 products for safe transport, storage, and utilization. Here, in this study, a variety of catalytic oxygen reduction approaches using commercial automotive exhaust emission catalysts are evaluated for O 2 removal from CO 2 streams using different reducing agents, including H 2 , CO, CH 3 OH, and CH 4 . When the amount of reductant added into the feed is carefully controlled, H 2 , CO, or CH 3 OH can effectively remove O 2 from 1.5% to below 10 ppm with a single-reactor design (>99.93% removal efficiency), while simultaneously meeting the impurity limits for other species. With CH 4 as a reductant, it is challenging to simultaneously meet the the O 2 and CO specifications in a single-reactor design because CH 4 also reacts with CO 2 to produce high levels of CO in the stream (at several hundred ppm). A dual-reactor design is developed to enable the use of CH 4 , a readily available and low-cost reductant, for the purification of CO 2 streams to meet the specifications for O 2 and other impurities.

CO2 purification↗
DOE OSTI2025

Basis for Dose and Reactor Safety Design Criteria for Army Regulation AR 50–7 and DA Pamphlet

This report describes the basis used to develop the radiological dose acceptance and design criteria contained in the draft updates to Army Regulation 50–7 (AR 50–7) Army Reactor Program and its accompanying draft Department of the Army (DA) Pamphlet (PAM), Army Reactor Program Procedures. These criteria will apply to Army nuclear reactors that fall under AR 50–7 and its accompanying DA PAM and ensure alignment with the overall objectives of the Army Reactor Program. The development basis for the radiological dose and design criteria supports a modern, technology-neutral, risk-informed, and performance-based approach to Army regulation of reactors and the demonstration of “adequate protection of the public.” To establish these criteria that support the Army’s unique operational requirements, multiple well-known and well-established standards and their supporting documentation were reviewed to ensure consistency with existing regulatory safety levels, including guidance from U.S. and international sources. These include the U.S. Nuclear Regulatory Commission’s (NRC’s) regulations and policy, the Canadian Nuclear Safety Commission’s (CNSC’s) regulatory documents, the International Atomic Energy Agency’s (IAEA’s) safety standards, as well as industry input that is tailored specifically to advanced microreactors. This report walks through the key definitions and associated references used for these criteria, which are outlined in Section 2.0. Based on these definitions, the dose acceptance criteria were established for various receptors for routine reactor operations (Section 3.2), design basis accidents (Section 3.3), and beyond design basis accidents (Section 3.4). Comparisons of multiple national and international dose limits are provided in these sections. Lastly, Section 4.0 outlines the reactor safety design criteria contained in the draft DA PAM and their associated bases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗