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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Irradiation performance of a U-7Mo in Al-Si matrix dispersion full-size fuel plate assembly

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR’s as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235 U content. Additionally, finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm 2 . No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Optimization and testing results of Zr-bearing ferritic steels

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Software Validation Work With The ZPPR-15 Data

The analysis activities for fast reactors involve using many different pieces of software that are relied upon for their predictive capabilities. For this software to be considered reliable, documented proof that the predictions of the software are accurate is required. In this manuscript, the validation work that covers some of the Argonne software used in fast reactor design activities is discussed and displayed. This validation work includes neutron and gamma flux distributions, reaction rate distributions, and reactivity worth. In an ideal world, a reactor development program would have access to a comprehensive set of experimental facilities to help inform the design aspects of the reactor itself. While thermal-hydraulics experiments, and to a limited degree mechanical experiments, can be carried out today for validation needs, neutronics related experimental facilities are rather impractical because of the lack of experimental facilities. Given the desired time table for construction of new reactors, the reconstitution or creation of new neutronic experimental facilities is untenable and thus those reactor development programs must rely upon any available experimental measurements that are qualitatively similar to the design. While a methodology has been proposed to assess the similarity between the past experimental measurements and the reactor itself, that aspect is beyond the scope of this manuscript. In this manuscript, the focus is entirely placed on the analysis results for a series of experiments carried out at the ZPPR facility in Idaho in the mid-1980s. In this regard, this manuscript only shows the validation of the stated neutronics software for specific loadings of the ZPPR reactor. Because of the fuel form, its proposed enrichment, and the material content of the reactor core, the ZPPR-15 experiments were identified as potential validation data for the reactor. The ZPPR-15 experiments were intended as mockups of a 330 MWe Integral Fast Reactor program which was a follow on program to the Clinch River Breeder Reactor. In the ZPPR-15 series of experiments, measurements of the neutron spectrum, control rod worth, sodium void worth, foil reaction rate distributions, Doppler worth of heated samples, gamma dose, and axial expansion worth were all carried out and published. In many cases, these reactivity coefficients are good candidates to validate the reactivity coefficient calculation scheme used by the analysis software and included in the safety analysis activities of fast reactor development projects today. This manuscript discusses the modeling methodology and accuracy of the calculated experimental results using the LANL software MCNP and the ANL software package ARC (Argonne Reactor Codes). As will be shown, for many of the experimental measurements, the two software packages are found to be good predictive analysis tools for those experiments. In other cases, problems with the analysis methodology or underlying cross section data are exposed which indicates where predictive analysis is not as reliable. Finally, in some of the measurements the conclusion is reached that the experimental measurement cannot be reproduced with the analysis software as it is simply too difficult.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Software Validation Work With The ZPPR-15 Data

The analysis activities for fast reactors involve using many different pieces of software that are relied upon for their predictive capabilities. For this software to be considered reliable, documented proof that the predictions of the software are accurate is required. In this manuscript, the validation work that covers some of the Argonne software used in fast reactor design activities is discussed and displayed. This validation work includes neutron and gamma flux distributions, reaction rate distributions, and reactivity worth. In an ideal world, a reactor development program would have access to a comprehensive set of experimental facilities to help inform the design aspects of the reactor itself. While thermal-hydraulics experiments, and to a limited degree mechanical experiments, can be carried out today for validation needs, neutronics related experimental facilities are rather impractical because of the lack of experimental facilities. Given the desired time table for construction of new reactors, the reconstitution or creation of new neutronic experimental facilities is untenable and thus those reactor development programs must rely upon any available experimental measurements that are qualitatively similar to the design. While a methodology has been proposed to assess the similarity between the past experimental measurements and the reactor itself, that aspect is beyond the scope of this manuscript. In this manuscript, the focus is entirely placed on the analysis results for a series of experiments carried out at the ZPPR facility in Idaho in the mid-1980s. In this regard, this manuscript only shows the validation of the stated neutronics software for specific loadings of the ZPPR reactor. Because of the fuel form, its proposed enrichment, and the material content of the reactor core, the ZPPR-15 experiments were identified as potential validation data for the reactor. The ZPPR-15 experiments were intended as mockups of a 330 MWe Integral Fast Reactor program which was a follow on program to the Clinch River Breeder Reactor. In the ZPPR-15 series of experiments, measurements of the neutron spectrum, control rod worth, sodium void worth, foil reaction rate distributions, Doppler worth of heated samples, gamma dose, and axial expansion worth were all carried out and published. In many cases, these reactivity coefficients are good candidates to validate the reactivity coefficient calculation scheme used by the analysis software and included in the safety analysis activities of fast reactor development projects today. This manuscript discusses the modeling methodology and accuracy of the calculated experimental results using the LANL software MCNP and the ANL software package ARC (Argonne Reactor Codes). As will be shown, for many of the experimental measurements, the two software packages are found to be good predictive analysis tools for those experiments. In other cases, problems with the analysis methodology or underlying cross section data are exposed which indicates where predictive analysis is not as reliable. Finally, in some of the measurements the conclusion is reached that the experimental measurement cannot be reproduced with the analysis software as it is simply too difficult.

Aliberti, Gerardo↗
DOE OSTI2024

Electrodialysis: An effective methodology to purify the leachate of spent Li-ion batteries

The electrification of transportation and the transition of society towards low or net-zero carbon emissions has led to a skyrocketing global demand for Li-ion batteries. After a service life of three to ten years, Li-ion batteries have less than 80 % of their initial capacities and draw near to the end of their lives for practical utilization. Due to potential supply chain shortages and the value embodied in Li-ion batteries, it is imperative to recycle them, to recover the materials, and to improve the circularity and the sustinability of the industry. How to cost-effectively purify spent batteries while reducing time, energy, and waste emissions is a challenge faced by Li-ion battery recyclers. The first electrochemical membrane reactor reported in our group hasa high selectivity towards lower Cu 2+ , Al 3+ and Fe 3+ ions (<5 ppm) and retains >95% of the Ni 2+ , Co 2+ and Mn 2+ ions in the leachate. An advanced electrochemical membrane reactor was developed in this study. Further, not only does the new reactor have the same selectivity as the original reactor, but other advantages including a faster leachate processing rate (up to 10X faster). The advanced reactor can also directly generate acid at the anode side; eliminating the reactor restoration step. The prominent advantages that this electrodialysis technology has over chemical-precipitation methods include: (1) ion recovery efficiencies do not diminish after removing the impurities, Ni 2+ , Co 2+ and Mn 2+ , even at a higher initial Ni 2+ ion concentrations; in comparison, chemical precipitation has Ni 2+ , Co 2+ and Mn 2+ ion recovery efficiencies reduced significantly when the initial Ni 2+ , Co 2+ and Mn 2+ ion concentrations increase. (2) electrodialysis does not change the concentrations of Ni 2+ , Co 2+ and Mn 2+ ions significantly, but chemical precipitation could reduce Ni 2+ and Co 2+ ions to less than half of their initial values. Through electro-dialyzing the leachate, the H 2 evolution reaction mechanism was found to switch from the Volmer-acid Heyrovsky mechanism to the Volmer-alkaline Heyrovsky mechanism at a pH of around 3.7.

25 ENERGY STORAGE↗
DOE OSTI2013

Progress report on the mechanical testing and thermal aging of Zr-bearing ferritic steels

The mission of the Nuclear Energy Enabling Technology (NEET) program is to develop cross-cutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) primarily designed for management of high-level wastes, but also the extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization.

Tan, Lizhen↗
DOE OSTI2017

Coupled CFD-DEM Analysis of Molten Salt-Cooled Pebble-Bed Reactor Experiment

The development of new nuclear reactor designs is an area of ongoing investigation, and a focus of dedicated work for various laboratories around the world. Despite the steadfast and reliable service of Light Water Reactors (LWR), the ideal power reactor, with a combination of high operational temperature, passive safety, proliferation resistance, economic build and waste minimization remains elusive. One promising design that might address these diverse, and at times contradictory requirements, is the Molten Salt Reactor (MSR). In its most radical form, a liquid fueled (LF) MSR has the nuclear fuel dissolved in the molten-salt primary coolant itself, as demonstrated by Oak Ridge National Lab (ORNL) in 1965 to 1969. An alternative fuel arrangement is to retain conventional, solid fuel elements. A proposed method is to implement pebble fuel elements, as have been used in previous gas-cooled reactors, with the gas coolant being replaced with a molten salt. This is referred to as pebble-bed, molten-salt reactor (PB-MSR). The spherical fuel pebbles used within PB-MSRs contain TRISO particles dispersed throughout the pebble's graphite matrix. Many pebbles are constrained to form a porous bed within the core region of a reactor. Coolant molten-salt passes through the pebble bed to remove heat produced due to the nuclear fission taking place within the fuel pebbles' TRISO particles. Various benefits are apparent with the PB-MSR design. When considering the choice of fuel elements, they are a geometry that has been produced and utilized in a similar way in the past, within gas-cooled reactors. The fuel pebbles are also able to be added to, and removed from, the core during operation, enabling on-line refueling. When considering the coolant choice, molten-salts have much higher boiling points (1703 K at 1 atm. for FLiBe) than conventional coolants such as water (630 K at 150 atm.), and are capable of reaching these temperatures at atmospheric pressure. This in turn removes the need for high pressure containment that is present in light water reactors and gas-cooled reactors. The future implementation of new reactors, and the search for improvements in efficiency and safety in current designs, require the advancement of simulation tools, which this work ultimately aims to address.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Adjusted Graphite Thermal Scattering Data Applied to Fluoride-Salt-Cooled Reactors

Thermal scattering data is a key component of fission reactor simulation. While most materials can be reasonably modeled with a free gas approximation, the moderators in thermal systems require a more rigorous treatment to sufficiently capture their specific neutron interaction physics. Graphite does not typically feature as a principle moderator in power reactors in service today, but remains an attractive prospect for a number of new reactor concepts currently in development. In comparison to light water, a material nearly ubiquitous in thermal reactors, graphite offers a better neutron economy by way of a small absorption cross section, high thermodynamic stability and can be incorporated into the structure of the core itself. Graphite in its purest form is a crystalline material composed of stacked layers of graphene, a hexagonal planar lattice of carbon atoms. In-plane, each carbon atom is covalently bonded to 3 neighbouring carbon atoms, creating a strong bond, while between planes only a weak van der Waals force exists. Pure graphite has a density of 2.26 g/cm{sup 3}. However, reactor-grade contains a significant amount of carbon binder, leading to a porous bulk medium with randomly oriented micro crystals. Densities of 1.5 - 1.9 g/cm{sup 3} are typical for reactor-grade graphite, corresponding to porosities on the order of 10-30%. This deviation from pure crystalline graphite should be accounted for in the thermal scattering data. When an incident neutron is in or near the thermal region, its de Broglie wavelength is the same order as the molecular bond lengths, leading to elastic scattering. At similar energies, the dynamics of the moderator lattice (in the case of solid moderators) can cause inelastic scattering. For these moderators, special cross sections should be calculated and used in reactor simulations. To generate this thermal scattering data, parameters pertaining to the properties and structure of the material in question are fed into a nuclear data processing code, such as NJOY. Within NJOY is a module called LEAPR which is used to generate the thermal scattering law S(α,β), which is in turn is fed into another module THERMR which is used to generate thermal cross sections that factor in the molecular-scale scattering phenomena. The core of the thermal scattering law is the phonon spectrum ρ(β) which describes the modes of excitation of the material lattice. An investigation into the phonon frequency spectrum was conducted at North Carolina State University (NCSU) for reactor-grade graphite which accounts for porosity. This phonon spectrum was calculated using the ab initio density functional theory (DFT) code VASP. The inelastic cross section calculated using this spectrum was found to have good agreement with relevant experimental data. The resulting spectra differed from the benchmark spectrum produced by Young and Koppel, the same data used in current nuclear data evaluations. The latter uses the lattice dynamics of perfect crystalline graphite with four force constants fitted to the model from thermodynamics data from reactor-grade (porous) graphite, which yields a reasonable approximation to the phonon spectrum of reactor-grade graphite.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Modeling for a Digital Twin-Based Remote Operation System Framework

New reactor designs and technologies are being developed to grow and advance the nuclear industry. Microreactors are one of the many new concepts for advancing the industry. Microreactors are very small reactors generally designed to have an operating power of 20 MWth or less. They are ideal for many applications in which it would not be feasible to have a large-scale reactor, such as powering remote communities, military bases, and mining sites. Many of these applications currently rely on diesel generators for power, and replacing those generators with the carbon-free energy of a microreactor is a major driving factor for microreactor development. However, most of the use cases for microreactors are in isolated locations where construction and labor costs are much higher. Microreactors will need to be comparable to other energy production methods for their deployment to be successful. Remotely operating the microreactor has a great potential to benefit the economics and make it more cost competitive. With remote operations, the operation facility location could be strategically chosen based on factors such as construction costs, workforce size, etc. The benefits of remote operations could be leveraged even more if the remote operation system is semi-autonomous. If the remote operation system is semi-autonomous, more microreactors could be operated and monitored from one remote operation facility. Additionally, with the system handling some tasks for the human operator, it could reduce the number of operating staff necessary for the microreactor. Digital twins can be used to introduce a level of automation to the remote operation system. Digital twins are capable of component monitoring, system operation and control, and predictive performance [1]. All of those features are important for a successful semi-autonomous remote operation system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

U.S. Department of Energy Office of Nuclear Technology Research and Eevelopment ((NTRD) comprehensive summary of QA assessments for FY17

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) has developed a research and development (R&D) roadmap for its research, development, and demonstration (RD&D) activities to ensure nuclear energy remains a compelling and viable energy option for the U.S. The roadmap defines NE RD&D activities and objectives that address the challenges to research, develop and demonstrate options to the current U.S commercial fuel cycle to enable the safe, secure, economic, and sustainable expansion of nuclear energy, while minimizing proliferation and terrorism risks expanding the use of nuclear power. The roadmap enables the development of technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. In addition, it will help to develop improvements in the affordability of the new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

NBSR reactor control room upgrade - 215

The maintenance and modernization of the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) reactor, known as the NBSR, is an ongoing process to endure until the eventual decommissioning of the reactor. As part of the renovation effort, the reactor control system is to be upgraded. Following some initial outsourcing efforts, the NBSR reactor control system upgrade has been restructured to be implemented by in-house means. The beginning stage of the upgrade involved the development of a simulator environment based on a PC system and digitization of the majority of nuclear reactor monitoring data. The NBSR reactor surveillance data consists of approximately thirteen hundred input/output points, including analog/digital sensor signals, signal conditioners, and relay logic. The aim of the current phase is to implement a new reactor data acquisition and display system (RDS) and lay out the infrastructure for the upgrade of the reactor control system (RCS). Key design criteria are to implement a reliable, redundant, diverse, modular and candid system where any complex digital manipulations are taken outside of the control room. Basically, the data analytics and custom codes are separated from reactor operation control and decision making logic. Furthermore, the new RDS is designed such that it could be maintained internally by NCNR personnel. The RDS is unique in design and modularity, minimizing custom codes and testing, making it ideal for research reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Radiation Resistance of XLPE Nano-dielectrics for Advanced Reactor Applications

Recently there has been renewed interest in nuclear reactor safety, particularly as commercial reactors are approaching 40 years service and lifetime extensions are considered, as well as for new reactor building projects around the world. The materials that are currently used in cabling for instrumentation, reactor control, and communications include cross-linked polyethylene (XLPE), ethylene propylene rubber (EPR), polyvinyl chloride (PVC), neoprene, and chlorosulfonated polyethylene. While these materials show suitable radiation tolerance in laboratory tests, failures before their useful lifetime occur due to the combined environmental effects of radiation, temperature and moisture, or operation under abnormal conditions. In addition, the extended use of commercial reactors beyond their original service life places a greater demand on insulating materials to perform beyond their current ratings in these nuclear environments. Nanocomposite materials that are based on XLPE and other epoxy resins incorporating TiO2, MgO, SiO2, and Al2O3 nanoparticles are being fabricated using a novel in-situ method established at ORNL to demonstrate materials with increased resistance to radiation. As novel nanocomposite dielectric materials are developed, characterization of the non-irradiated and irradiated nanodielectrics will lead to a knowledge base that allow for dielectric materials to be engineered with specific nanoparticle additions for maximum benefit to wide-variety of radiation environments found in nuclear reactors. This paper presents the initial findings on the development of XLPE-based SiO2 nano-composite dielectrics in the context of electrical performance and radiation degradation.

dielectrics↗
DOE OSTI2024

DIII-D research to provide solutions for ITER and fusion energy

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ≥ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in Te and radiation which need control. High-β P scenarios with q min > 2, q 95 –7.9, β N > 4, β T –3.3% and H 98y2 > 1.5 are sustained with high density ($\overline{n}$ = 7E19 m -3 , f G –1) for 6 τ E , improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Web-Based Computational Model Builder For Nuclear Energy Advanced Modeling and Simulation (Final Report)

According to the World Nuclear Association, the U.S. is the largest producer of nuclear power worldwide. In fact, the U.S. provides over 30 percent of the world's nuclear generation of electricity. Although few new reactors have been built in the past 30 years, the association anticipates that four to six reactors may be built by 2020. The first of the reactors will be built in response to 16 license applications, which have been completed since 2007, to build 24 additional nuclear reactors. Currently, both government and industry are working to accelerate approval for the construction and design of new nuclear energy plants. The U.S. nuclear industry has already achieved significant developments in the implementation of nuclear power plants due to advancements in refueling, maintenance, and safety systems at current power plants. Meanwhile, changes in government policy, which have occurred since the late 1990s, have provided for noteworthy expansion in nuclear capacity. For example, the Energy Policy Act of 2005 evoked investment in electricity infrastructure such as nuclear power. To design future systems and continue operational improvements at existing U.S. plants, advanced modeling and simulation of nuclear reactors are crucial. We developed a Cloud/Web-based tool, HPCCloud, that enables nuclear energy firms, or other manufacturing and engineering companies, to innovate using advanced modeling and simulation. Through the Web-Based Computational Model Builder for Nuclear Energy Advanced Modeling and Simulation grant, we develop HPCCloud to simplify and democratize advanced modeling and simulation in the nuclear energy industry on a range of nuclear engineering applications. It leverages millions of investment dollars from the Department of Energy's Consortium for Advanced Simulation of Light Water Reactors (CASL) and the Office of Nuclear Energy's research and development in Nuclear Energy Advanced Modeling and Simulation (NEAMS). Our approach simplifies the workflow, alleviates the need for in-house computational science and engineering experts, and lowers the capital investments required for advanced modeling and simulation. The Cloud/Web-based simulation environment guides end-users and developers through the advanced modeling and simulation lifecycle. In addition to providing significantly improved, intuitive software, the environment offers reproducible workflows where one can save, share, and even publish the full path of data from input to final analyzed results. HPCCloud addresses identified challenges to the nuclear energy industry's use of advanced modeling and simulation. With this unique Cloud/Web-based environment, we enable an ecosystem of synergistic activities, fueling new research and development for nuclear energy. HPCCloud lays the groundwork for commercialization, through customization, technology integration, and advanced research and development services.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A Proof of Concept: Grizzly, the LWRS Program Materials Aging and Degradation Pathway Main Simulation Tool

Nuclear power currently provides a significant fraction of the United States’ non-carbon emitting power generation. In future years, nuclear power must continue to generate a significant portion of the nation’s electricity to meet the growing electricity demand, clean energy goals, and ensure energy independence. New reactors will be an essential part of the expansion of nuclear power. However, given limits on new builds imposed by economics and industrial capacity, the extended service of the existing fleet will also be required.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

BDDR, a new CEA technological and operating reactor database

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗