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NASA NTRS1995

Safety analysis

We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

Knight, John C.↗

NASA NTRS1981

Advanced Low-Emissions Catalytic-Combustor Program, phase 1

Six catalytic combustor concepts were defined, analyzed, and evaluated. Major design considerations included low emissions, performance, safety, durability, installations, operations and development. On the basis of these considerations the two most promising concepts were selected. Refined analysis and preliminary design work was conducted on these two concepts. The selected concepts were required to fit within the combustor chamber dimensions of the reference engine. This is achieved by using a dump diffuser discharging into a plenum chamber between the compressor discharge and the turbine inlet, with the combustors overlaying the prediffuser and the rear of the compressor. To enhance maintainability, the outer combustor case for each concept is designed to translate forward for accessibility to the catalytic reactor, liners and high pressure turbine area. The catalytic reactor is self-contained with air-cooled canning on a resilient mounting. Both selected concepts employed integrated engine-starting approaches to raise the catalytic reactor up to operating conditions. Advanced liner schemes are used to minimize required cooling air. The two selected concepts respectively employ fuel-rich initial thermal reaction followed by rapid quench and subsequent fuel-lean catalytic reaction of carbon monoxide, and, fuel-lean thermal reaction of some fuel in a continuously operating pilot combustor with fuel-lean catalytic reaction of remaining fuel in a radially-staged main combustor.

Sturgess, G. J.↗

DOE OSTI2018

NanoStructured bulk thermoelectric generator for efficient power harvesting for self-powered sensor network

The goal of the Nuclear Energy Enabling Technology Research Project is to develop efficient and reliable thermoelectric (TE) generators for self-powered wireless sensor nodes (WSNs) for nuclear applications. The power harvesting technology has crosscutting significance to all United States Department of Energy Office of Nuclear Energy research and development programs, as it will enable self-powered WSNs in multiple nuclear reactor designs and spent fuel storage facilities using thermal energy available in a nuclear power plant or spent fuel storage facility. This project addresses an important technology gap that exists in truly realizing WSNs due to the need for cables to connect to external power supplies and develop TE power harvesting devices to deliver sufficient power to drive the WSNs. The outcomes of the project will lead to significant advancement in sensors and instrumentation technology, reducing cost, improving monitoring reliability, and therefore enhancing safety. The self-powered WSNs could support the long-term safe and economical operation of all reactor designs and fuel cycle concepts, as well as spent fuel storage and many other nuclear science and engineering applications. The research is based on recent breakthroughs in high-performance nanostructured bulk (nanobulk) half-Heusler and bismuth telluride thermoelectric materials that enable high-efficiency and high-power-density direct heat-to-electricity conversion over a wide temperature range. The nanobulk TE materials that the research team has already developed yield up to a 50% increase in the TE figure of merit, ZT, compared with state-of-the-art bulk counterparts. In addition to increased efficiency, the nanostructured materials are of potentially enhanced radiation tolerance due to the increased interface densities at the nanoscale grain boundaries that serves as efficient sinks to annihilate radiation-induced, freely-migrating structural defects. This report summarizes major technical results the team achieved in this three-year project, with a focus on the results obtained in the third year of the project. The report includes results on thermoelectric materials fabrication and optimization, irradiation effect on thermoelectric materials performance, thermoelectric device simulation, fabrication and testing, wireless sensor node system simulation, and the demonstration of a wireless sensor node powered by thermoelectric generators.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

The mobility of small vacancy/helium complexes in tungsten and its impact on retention in fusion-relevant conditions

Tungsten is a promising plasma facing material for fusion reactors. Despite many favorable properties, helium ions incoming from the plasma are known to dramatically affect the microstructure of tungsten, leading to bubble growth, blistering, and/or to the formation of fuzz. In order to develop mitigation strategies, it is essential to understand the atomistic processes that lead to bubble formation and subsequent microstructural changes. In this work, we use large-scale Accelerated Molecular Dynamics simulations to investigate small (N=1,2) V N He M vacancy/helium complexes, which serve as the nuclei for larger helium bubble growth, over timescales reaching into the milliseconds under conditions typical of the operation of fusion reactors. These complexes can interconvert between different I L V N+L He M variants via Frenkel pair nucleation (leading to the creation of a additional vacancy/interstitial pair) and annihilation events; sequences of these events can lead to net migration of these embryonic bubbles. The competition between nucleation and annihilation produces a very complex dependence of the diffusivity on the number of heliums. Finally, through cluster dynamics simulations, we show that diffusion of these complexes provides an efficient pathway for helium release at fluxes expected in fusion reactors, and hence that accounting for the mobility of these complexes is crucial.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Fast Flux Test Facility Benchmark Status for FY19

The Department of Energy’s Office of Nuclear Energy Advanced Reactor Technologies (ART) Program works on promising Generation-IV R&D. The Program focus is on reactor concepts that could dramatically improve performance in sustainability, safety, economics, security, and proliferation resistance. The near-term ART Program goal is to reduce technical barriers (e.g., by resolving key feasibility and performance challenges) to improve the prospects of advanced technology reactor systems. In the long-term, integrated advanced reactor system designs will be developed that implement the chosen missions and complete the testing and demonstration necessary for licensing. The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The plant description and plant data from one specific passive safety test is now an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP). This report is one in a series documenting the PNNL efforts to support this FFTF loss of flow benchmark for the IAEA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Enhanced performance in fusion plasmas through turbulence suppression by megaelectronvolt ions

Alpha particles with energies on the order of megaelectronvolts will be the main source of plasma heating in future magnetic confinement fusion reactors. Instead of heating fuel ions, most of the energy of alpha particles is transferred to electrons in the plasma. Furthermore, alpha particles can also excite Alfvénic instabilities, which were previously considered to be detrimental to the performance of the fusion device. In this work, we report improved thermal ion confinement in the presence of megaelectronvolts ions and strong fast ion-driven Alfvénic instabilities in recent experiments on the Joint European Torus. Herein, detailed transport analysis of these experiments reveals turbulence suppression through a complex multi-scale mechanism that generates large-scale zonal flows. This holds promise for more economical operation of fusion reactors with dominant alpha particle heating and ultimately cheaper fusion electricity.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Differentiable hybrid neural network approach for enhancing reactor dynamics simulations

Reactor dynamics simulations provide essential insights into the time-dependent behavior of nuclear reactors under various operating conditions. However, high-fidelity simulations can be computationally intensive, requiring significant computational resources. Here, to address this challenge, this study employs a differentiable hybrid model that utilizes neural networks as a corrector to enhance the performance of a low-fidelity simulation, aligning its predictions with those of a high-fidelity simulation. Low-fidelity and high-fidelity simulations were obtained by adjusting the mesh size in the System Dynamics Analysis Tool. The differentiable hybrid model was trained in two approaches: time-step-wise and sequence-wise. It was then applied to simulate various transients in a molten salt reactor. Its performance was evaluated by comparing its responses to transients against those of the high-fidelity simulation. An additional approach was performed using a data-driven model to correct the low-fidelity simulation. In comparison, the differentiable hybrid model showed significant improvements in transient prediction, effectively addressing the limitations of the low-fidelity simulations. The results highlighted the robustness of the differentiable hybrid model in both training approaches. It delivered simulations that were at least 3.8 times faster than high-fidelity models. In the time-step-wise approach, it achieved at least a 39% improvement in accuracy. In the sequence-wise approach, it showed at least an 81% accuracy improvement over the full transient. This approach offers a promising path for improving computational efficiency without compromising accuracy in nuclear reactor simulations, making it suitable for real-time digital twin applications.

42 - ENGINEERING↗

DOE OSTI2017

Supplement to the NUBOW-3D Manual

The NUBOW-3D code (NUBOW) developed in 1978 is a custom FORTRAN analysis code that models the complex mechanical interaction of the fuel assemblies and the core restraint system in response to the effects of thermal expansion and irradiation swelling and creep. NUBOW-3D models the fuel assembly ducts as beams with springs representing the contact between ducts. The core temperature profile and neutron flux profiles are provided at various times during the planned operation of the reactor. NUBOW then calculates the changing shapes of the fuel assemblies and their equilibrium positions within the core restraints. An optional post-processing step calculates the reactivity changes based on new positions of the fuel assemblies using input displacement reactivity WORTHS. Understanding this reactivity change is a key feature to the inherent safety of a fast reactor core. NUBOW was developed to support the design of limited free bow core restraint systems. The NUBOW software was modified in 1987 but the documentation of the upgrades was not completed due to abrupt funding cuts. In the last few years, the software has been ported to a Windows platform using a current FORTRAN compiler and additional changes have been made including consolidating the code to a single executable. While the original NUBOW manual [1] written in 1978 provides an excellent description of the code and input parameters, it lacks the additional features added in the upgrades and in some cases describes features that are no longer operable. This report attempts to bridge the information gap between the original manual and the current state of the software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Assessing Impact of Reactor Scaling on Carbohydrate and Lipid Yields from Saline and Fresh Water Algae

Effective deconstruction of algal biomass is a major challenge associated with algal biofuels/bioproduct production. Improving biomass conversion requires understanding how reaction parameters affect the scale-up of bench scale processes to move to cost-effective conversion. Pretreatment of both aquatic and terrestrial feedstocks, using similar reaction conditions, identical feedstock and pretreatment chemistries have reported wide ranges in total sugar yields. This has been attributed to differences in reactor size or scale, type of operation (batch vs continuous), method of heating, reactor geometry and solids loading (w/w). These factors can impact both mass and heat transfer affecting both yield and reaction kinetics. To understand how total sugar yields in algal biomass are affected by different dilute acid pretreatment configurations we evaluated total sugar and lipid release at three reactor scales ranging from 4 ml to 300 ml, operated under mixed and non-mixed regimes.

09 BIOMASS FUELS↗

DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using openmc

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using OpenMC

61 - RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2017

Energy recovery from one- and two-stage anaerobic digestion of food waste

Highlights: • One- and two-stage anaerobic digestion of food waste were compared. • The first stage of the two-stage process was properly operated to allow H{sub 2} recovery. • The two-stage process was characterised by a higher CH{sub 4} production than the single-stage one. • The enhanced CH{sub 4} production was due to improved fermentation in the first stage. • The two-stage process displayed a 20% comparatively higher energy recovery yield. - Abstract: One- and two-stage anaerobic digestion of food waste aimed at recovering methane (CH{sub 4}) and hydrogen and methane (H{sub 2} + CH{sub 4}), respectively, were compared in order to assess the potential benefits from the two-stage process in terms of overall energy recovery. Results suggest that a two-stage process where the first reactor is properly operated in order to achieve a significant net hydrogen production, may display a 20% comparatively higher energy recovery yield as a result, mainly, of enhanced methane production as well as of the associated hydrogen production. The highest methane production of the two-stage process was due to improved hydrolysis and fermentation of food waste, with increased amounts of volatile fatty acids being readily available to methanogenesis.

09 BIOMASS FUELS↗

DOE OSTI2021

Experiment and computational fluid dynamics investigation of biochar elutriation in fluidized bed

Here, in fluidized bed biomass fast pyrolysis, the biomass is converted to biochar and elutriated. The elutriation rate is a key parameter in reactor designs and operations. This research presents a video-based continuous measurement of biochar elutriation rate in a fluidized bed with sands and biomass as bed materials. The fluidized bed is simulated with the Computational Fluid Dynamics - Coarse-Grained Discrete Element Method (CFD-CGDEM) in MFiX. The fluidization behavior of non-spherical sands can be more accurately captured when a rolling friction model is used. The predicted elutriation rate is close to the experimental measurement when the particle size distributions are considered and the filtered drag with a shape correction is used. These results validated the accuracy of the MFiX based CFD framework for the prediction of biochar elutriations in the fluidized bed biomass fast pyrolysis reactor.

09 BIOMASS FUELS↗

DOE OSTI2024

Computing the solubility of argon and xenon in molten sodium chloride and potassium chloride salts

Molten salt reactors (MSRs) offer significant advancements in nuclear reactor safety and efficiency by operating at higher temperatures and lower pressures compared to traditional reactors. A critical aspect of MSR operation involves understanding the solubility of fission byproducts, particularly noble gases, in the molten salts used. This study employs molecular dynamics (MD) simulations to compute Henry’s law constants and enthalpies of solvation for argon and xenon in molten sodium chloride (NaCl) and potassium chloride (KCl). We developed a new pairwise potential for the noble gas and salt interactions based on first principles calculations. We then used this potential to calculate Henry’s law constants of the two gases in the molten salts, which were modeled using both a rigid ion model (RIM) and a polarizable ion model (PIM). The solubility calculations, performed using the Widom insertion method, show qualitative agreement with limited experimental data, highlighting the temperature dependence and greater solubility of both gases in KCl compared to NaCl. Additionally, free volume analysis elucidated the role of available space within the molten salts in governing solubility trends. Our findings suggest that PIM trajectories provide more reliable predictions for noble gas solubility than RIM due to their accurate density representation. Furthermore, these results enhance understanding of gas solubility in MSR environments, and the methods can be readily extended to other systems.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2017

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Engineering Scale Models and Analysis

As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \usi\ fuel and iron-chromium-aluminum (FeCrAl) cladding under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report summarizes the latest recommended material models to be used for engineering scale simulations based upon lower length scale analyses described elsewhere. Comparative analyses between traditional fuel rods (U 3 Si 2 /Zircaloy-4) and the proposed accident tolerant fuel rods (UO 2 /FeCrAl and U 3 Si 2 /Zircaloy-4) under normal operation, loss of coolant, and reactivity insertion conditions have been revisited using the latest state-of-the-art models implemented into the Bison fuel performance code. Sensitivity analyses have also been redone to take into account the latest knowledge on uncertain parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1995

Computational Simulation Of A Nuclear Engine

Nuclear Engine System Simulation (NESS) computer program developed to satisfy need for accurate, stand-alone software tool for preliminary design analysis of nuclear thermal propulsion (NTP) engine system. Used for rapid, preliminary detailed design analysis of both reactor and key engine systems. Principal outputs include operating characteristics and weights of reactors, engine-system parameters as performances, weights, dimensions, pressures, temperatures, and mass flows, and operating characteristics of turbopumps under both design and off-design operating conditions. Written in FORTRAN 77.

Walton, J. T.↗

DOE OSTI2023

A New Proposal Generalized Predictive Control Algorithm With Polynomial Reference Tracking Applied for Sodium Fast Reactors

This paper proposes a generalized predictive control (GPC) with constraints and orthonormal Laguerre functions using the simplified model of the primary system (reactor core and intermediate heat exchanger (IHX)) of a prototypical sodium fast reactor (SFR). This paper develops a multiple-input multiple-output (MIMO) GPC with input constraints able to track polynomial references of any degree applied in coolant temperature difference across the core and fractional power. The manipulated variables of the GPC-SFR are the reactivity and the sodium flow rate of the primary and secondary pipes. Moreover, orthonormal Laguerre functions and step down condition number techniques were also applied to avoid the numerical ill-conditioning issue in quadratic programming of large systems. Thus, a GPC type-2 was designed to control fractional power, coolant temperature difference across the core and sodium tank temperature of the SFR primary system when temperature references change according to a linear ramp after reaching their steady-state operation, sustaining 100% power operation on the reactor. In order to analyze the load tracking capability of the GPC-SFR type-2, the load following from 100% fractional power (FP) to 60% FP at 0.8% FP/min rate is simulated. Constraints on the rate of coolant temperature difference across the core and reactivity were applied for the design safety. For comparison criteria, this paper compares the GPC-SFR type-2 with the GPC-SFR type-1, i.e., standard model predictive control (MPC), to verify the viability and superior performance of the proposal regarding: (a) ramp-tracking capability of temperature and load; (b) the rejections of a reactivity disturbance of -1 cent and a secondary sodium inlet temperature disturbance of +10°F; and (c) a simulation with uncertainty in reactor design. The simulations show that the GPC-SFR type-2 overcome the GPC-SFR type-1 robustness and performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI1988

Nuclear Safety [Vol. 29, No. 3, July-September 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 259 Fifteenth Water Reactor Safety Information Meeting by E. G. Silver; CONTROL AND INSTRUMENTATION: 284 Reliability Technology to Improve and/or Maintain Emergency Diesel Generator Performance by S. Karimian and J. H. Taylor, 293 A Noise Diagnostics System for Operator Advice by G. Hessel, P. Liewers, P. Schumann, W. Schmitt, and F.-P. Weiss; PLANT SAFETY FEATURES: 307 A Passive Containment System for Advanced Light-Water Reactors by O. B. Falls, Jr., and F. W. Kleimola; ENVIRONMENTAL EFFECTS: 318 Data Base Construction for a Computerized Radiological Risk Investigation System by L. M. Hively, J. E. Nyquist, J. L. Bledsoe, and A. L. Sjoreen, 326 Erratum to "Radiation Hormesis and Nuclear Safety," Vol. 29, No. 1; OPERATING EXPERIENCES: 327 Operational Safety Experience and Passive Safety Testing at the Fast Flux Text Facility by Q. L. Baird, J. L. Rathbun, D. D. Stepnewski, R. L. Stover, and A. E. Waltar, 344 Backfilling of Independent Residual Heat Removal Systems in West Germany and Switzerland by G. Eckert and Y. Salomon, 353 Reactor Shutdown Experience Compiled by J. W. Cletcher, 356 Selected Safety-Related Events Compiled by G. A. Murphy, 363 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 384 General Administrative Activities Compiled by E. G. Silver, 390 Reports, Standards, and Safety Guides by D. S. Queener, 395 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 400 Proposed Rule Changes as of Mar. 31,1988; ANNOUNCEMENTS: 283 International ENS/ANS Conference on Thermal Reactor Safety "NUCSAFE 88", 317 Northwestern University Short Course on Radiation Safety, 407 Second International Seminar on Small- and Medium-Sized Nuclear Reactors, 407 International Workshop on New Developments in Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, 408 Fourth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-4), 404 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗