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At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

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

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

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Semi-Annual Report for Horizontal Compact High Temperature Gas Reactor (HC-HTGR) Development during Performance Period October 2022-March 2023

Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. Additionally, Argonne is providing analysis of the primary heat transport system to ensure temperatures in the reactor systems, structures and components with significant safety margins during normal operation and design basis accident scenarios. This third semi-annual report summarized the progress made at Argonne on the two tasks during the first half of FY23. As a part of the RCCS design task, a design process for the water panel and the system configuration was performed to improve the thermal performance of the RCCS. The updated water panel design had achieved enhanced thermal performance with 0.96 MWth capability with major design updates made in structural interfaces with the RPV and initial configuration of the water tanks. The loop configuration of the RCCS has been proposed to have two independently working loops for system redundancy. A water panel material study was performed focusing on the use of carbon steel in water systems. Additional modeling strategies of primary system thermal fluids analyses were developed to meet modeling needs that are not well suited for the 1D-3D assembly level approach. The first of these is a reduced order assembly model, in which fuel centered unit cells are used to represent a fuel assembly. The 2D approach used in this model is much more computationally efficient, making this model useful for full core transient scenarios where fuel to coolant heat transfer is still the dominant flow path. The next model is the 3D core conduction model to be used to analyze decay heat removal in loss of primary system flow scenarios. Because these scenarios require a large domain to be modeled, a homogenized core model is being pursued to reduce the required computational costs. To accurately model decay heat scenarios it is necessary to couple a RCCS model to the 3D core conduction model. A simplified case is presented to demonstrate how the coupling methodology will be applied to the full core 3D conduction model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2023

Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense (Final)

In the world of ever-advancing technology, Autonomous Systems (AS) find extensive application, bolstering functionalities of critical infrastructures such as nuclear power plants. These systems, however, are increasingly becoming a target for nefarious activities, namely through inference attacks, trojan attacks, and adversarial reprogramming. This paper delves into a comprehensive exploration of machine learning (ML)-driven autonomous control systems within advanced nuclear reactor designs, revealing the vulnerabilities and proposing strategies for defense against potential cyber-attacks. Advanced cyber-attacks against critical infrastructure and the energy sector are becoming more common. With the invention of autonomous control systems (ACS) within advanced nuclear reactor designs, system designers, reactor operators, and regulators must consider cybersecurity during the design and operational phases. This article provides a cyber threat assessment of machine learning (ML)- based digital twinning (DT) technologies in the context of advanced reactor ACS. A cyber-physical testbed was created to emulate nuclear reactor digital instrumentation and controls (I&C) and act as a basis for the ACS. The ACS was designed as two plant-level DTs predicting reactor malfunctions and determining control actions and two component-level DTs responsible for classifying component states and forecasting component inputs and outputs (I/O). Two duplicate ACS designs– one using a traditional ML framework and one using an automated ML (AutoML) framework– were created and tested against cyber-attacks on training data, real-time process data, and ML model architectures to determine their respective qualitative cyber-risk in terms of likelihood and impact. Both frameworks showed similar cyber-resilience against training, real-time, and ML architecture attacks, proving that neither is inherently more secure. Recommended safeguard and security measures are posed to system designers, reactor operators, and regulators to maintain the cybersecurity of ML-based DT technologies such as ACS, prompting a holistic view of shared responsibility for maintaining cyber-secure ML-based systems. As global reliance on generation III reactors begins to be critically assessed, the evolution towards advanced reactor systems utilizing digital instrumentation and controls (I&C) becomes not merely preferable, but essential. The integration of semi and fully autonomous control systems (ACS), powered by digital I&C and machine learning (ML)-based digital twinning (DT) technologies, emerges as a potent strategy to mitigate operations and maintenance costs, thereby enhancing the economic feasibility of novel reactor designs. However, with a staggering 500% and 380% increase in cyber-attacks reported against the energy sector by the United States Department of Energy (DoE) and the European Union respectively, a surge in cyber vulnerabilities specifically targeting the nuclear industry has been 2 markedly observed. Notable incidents, such as the W32.Ramnit spyware infiltration at the Gundremmingen nuclear power plant in Germany and the Dtrack spyware intrusion at the Kudankulam nuclear power plant in India, while not directly compromising core industrial control systems (ICS), underscore a compelling necessity to fortify cybersecurity protocols in safeguarding reactor systems against increasingly adept digital adversaries. In light of this, our investigation extends beyond conventional cybersecurity parameters, diving into the intricate web of potential vulnerabilities woven into ML-based DTs and ACS in advanced reactor systems. A crafted cyber-physical testbed and preliminary ACS were devised to act as a mirror, reflecting potential configurations of advanced reactor control designs. Moreover, this study is intertwined with a scrutinization of ML models, developed either through conventional, manually tuned methodologies or via automated means through AutoML, probing into their cyber-risk profiles within operational technology (OT) environments. Expanding on this, two distinct ACS blueprints were forged – one navigating through the corridors of traditional ML and the other traversing the path of AutoML – in an effort to holistically encapsulate the considerations pivotal to ML-based DT control system design. Employing the SANS Institute Industrial Control System (ICS) Kill Chain and the MITRE ATT&CK Tactics, Techniques, and Procedures (TTP) framework, a structured analysis was conducted, launching three targeted attacks against the training dataset, real-time dataset, and ML models, therein dissecting the potential cyber-attack implications against both ML frameworks within an ACS milieu. It is essential to note that three distinct categories of attacks were conducted against both ACS configurations, each encompassing three distinct ML-based DTs, cumulating in a total of 18 varied attacks. This exploration extends into the realms of Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense, unraveling vulnerabilities, and opportunities for fortified defenses against such intrusions, particularly where ML-driven technologies, and by extension, ACS, are deployed. Final recommendations, articulated through a lens of security, safeguard, and implementation considerations, are presented for both traditional and AutoML models, anchoring upon the existing knowledge landscape and ML-based DT modeling for ACS, and are offered as a beacon to guide the nuclear industry through the intricate cybersecurity challenges that lie ahead.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Finally, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pilot Plant Reliability Metrics for Grinding and Fast Pyrolysis of Woody Residues

The reliability of biomass size reduction, feeding, and conversion processes remains a critical technical hurdle and important cost driver for biomass conversion systems. While Biorefinery projects have typically focused on scaling up conversion technologies that maximize yield of desired products, e.g. fuels and chemicals, solids handling and reactor feeding systems are often considered secondarily. These are frequently scaled down or adapted from other industries, without a full appreciation of how biomass heterogeneity and variability impact process reliability. Here we present the effects of feedstock variability on relative throughput, mass loss, plant uptime, operator intervention, product yield, and product quality for grinding and fast pyrolysis operations at the pilot scale. Loblolly pine residues were sourced and collected using commercial harvesting techniques. Grinding and pyrolysis experiments were performed at INL's Biomass Feedstock National User Facility and NREL's Thermal and Catalytic Process Development Unit, respectively. Preprocessing throughput using a hammer mill varied between 31-48% of nameplate capacity (5 tons/hr), and fast pyrolysis throughput varied between 57-72% of nameplate capacity (20 kg/hr) for an overall system throughput of 18-35% of nameplate capacity across biomass preprocessing and conversion to pyrolysis oil. Pyrolysis oil yields varied between 46-53% (feedstock carbon to oil, dry basis). Grinder overloads in the size reduction step were more prevalent for lower ash and higher moisture materials. During fast pyrolysis operations, downtime was caused by bridging in the feed and char removal systems and plugging in the condensation system, and significant variation in the type and frequency of operator intervention was observed between feedstocks. Cohesion of solids (feedstock and char) leading to system plugging appeared less problematic for higher ash feedstocks, and differences in condenser plugging behavior between high and low ash feedstocks indicates the composition and physical nature of the deposited material is important for process reliability and is feedstock dependent.

09 BIOMASS FUELS↗
DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Preliminary Design of Engineering-Scale Salt Accident Analysis Facility to Support Molten Salt Reactor Licensing

Systems-level nuclear accident analysis codes for reactor licensing must be validated using experimental data that represent behaviors expected during actual full-scale accidents. Some behaviors may arise from coupled processes and only manifest at large scales. This report presents the preliminary design of the Salt Accident Analysis Facility (SAAF, pronounced “safe”), which is an experimental test facility to be constructed at Argonne that can be used to conduct integrated salt accident tests at an engineering scale. The measurement capabilities of the SAAF are based on previously developed methods and will provide the representative datasets that are needed to support molten salt reactor (MSR) licensing. Details of the design, the processes to be quantified, the measurement techniques for quantifying the processes, the variables that can be adjusted to simulate different accident scenarios, and operational considerations are presented herein. This report provides stakeholders the opportunity to give feedback on the test facility capabilities and planned analyses before it is constructed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Design of an epithermal neutron velocity selection system for the Penn State Breazeale Reactor

A series of mechanical neutron choppers to operate as a velocity selection system have been developed for the Pennsylvania State Breazeal Reactor (PSBR). This chopper system will provide pulsed epithermal neutrons in the energy range of 0.5–40 eV with 2% or better energy resolution, and with a transmission of 1E-6 or better. Four different chopper geometries have been evaluated for their utility as mechanical neutron choppers. Specifically, Fermi, ring, piston, and disc choppers have been evaluated to assess their potential neutronics performance and mechanical constructability. A series of high-speed disc choppers were selected for the final design, and optimization work was performed to maximize the neutron pulse intensity. It is estimated that the optimized system will produce an epithermal neutron intensity of approximately 96 n/s. This system can also be operated in a time of flight (TOF) configuration such that the neutrons arriving at the source have a white spectrum. This operation can be accomplished by leaving the second stage of choppers in the open position, or by removing it from the beam completely. Such unique source of epithermal neutrons produced by this chopper system will have applications in both prompt and delayed epithermal neutron activation analysis (ENAA), as well as in neutron resonance transmission analysis (NRTA).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Solar Heat for Industrial Processes: Integration with Chemical Reactors

The integration of solar thermal systems with chemical reactors has been proposed as part of a larger effort to develop and deploy solar heat for industrial processes (SHIP) technologies. A strong motivation for SHIP processes and technologies is the potential for high thermal efficiency coupled with low-cost thermal energy storage (TES) which can enable commercial deployment of such systems. While there are different ways to categorize SHIP technologies, one important such distinction is between directly irradiated systems and indirect off-sun process driven by a SHIP system. While directly irradiated systems can provide high thermal efficiencies and high fluxes, they usually require complex engineering solutions due to the need for redesigning the established processes and unit operations. In most cases, it is also more challenging to couple such a process to a TES system, losing some of the benefits of SHIP. On the other hand, using a SHIP system to drive an industrial process off-sun can allow better integration with existing process chains, easier TES capabilities, and potential for more applications fitting a specific SHIP technology. However, the integration of SHIP systems with the industrial processes is not fully explored in detail, especially in the case of high-temperature processes such as reforming, cracking, cement manufacturing, and iron/steelmaking. Many of these systems require heating fluxes of >50 kW/m^2, supplied via combustion of hydrocarbons in a fire box and benefitting from radiative heat transfer between the flue gases and the reaction zones. As such, using SHIP systems for such processes is more complex than providing the same thermal input in the form of a heat transfer medium (HTM) entering the reactor, kiln, or furnace. Moreover, in case convective heat transfer using SHIP is envisioned, for example using supercritical CO2 as the HTM from a particle receiver, the thermal integration might be more challenging than initially envisioned: lower heat transfer coefficients and limited approach temperature might require large flow rates, causing a mismatch between the process thermal requirements and the thermal capacity of the SHIP system. In addition, even if the heat exchange between SHIP and reactor is effective, there is still a cold leg HTM at the reaction temperature or slightly below it. Chemical plants usually include a set of heat exchangers, heat recovery steam generators, and even power generation units - in a tightly integrated design - to recover the flue gases which are eventually vented. With SHIP systems mostly operating on a closed HTM loop, bottoming the cold leg is crucial. In this talk, we will present different modeling results for a variety of syngas production reactions, using catalytic and chemical looping processes, and discuss some of the challenges and design considerations for off-sun chemical reactors using SHIP systems.

14 SOLAR ENERGY↗
DOE OSTI2021

Behavior of an air-cooled thermosiphon during a non-air gas ingress event

Here, this study investigates the transient response of an air-cooled thermosiphon subjected to a sudden non-air gas ingress event. Results presented are from experimental testing conducted on a large-scale thermal hydraulic facility that examines heat removal performance of the Reactor Cavity Cooling System (RCCS) concept, a passive safety system for advanced nuclear reactors. Relying on buoyancy driven natural circulation flow, these systems feature a series of riser standpipes and coupled chimney network that provide a pathway for heat rejection to the atmosphere. In this work, the facility was modified to allow a transition of the draft intake from normal ambient air, to draft intake from an open-top volume containing high purity argon. Testing began by establishing normal operation and allowing the facility to reach steady-state thermal hydraulic flow conditions with natural air draft. The transitioned sequence was then initiated which simultaneously changed the inlet boundary from open air to a pathway from the quiescent argon volume. This event caused 1,200 cu-feet of the heavy gas, twice the internal volume of total facility flow path, to be drawn into the inlet plenum and ingress into the heated riser standpipes. With the upper chimney network still containing residual air, there was insufficient density difference to maintain buoyancy driven natural circulation. After a period of only 90 s past the ingress event, the facility experienced complete flow stagnation before entering an extended period of severely degraded system flow. Due to the cessation of bulk fluid movement and subsequent failure of its heat removal function, fluid and structural temperatures began to rise sharply. Re-circulation patterns developed within the multiple parallel riser standpipes, where hot gas near the outlets was observed to travel downward and re-enter the inlet of adjacent channels. After approximately 18-minutes, fluid temperatures and their associated density difference rose to a level sufficient to allow reestablishment of buoyancy driven system flow, and ultimately, recovery of facility operation to normal behavior.

42 ENGINEERING↗
DOE OSTI2023

Assessing parallel path cooling tower performance via artificial neural networks

Real-time monitoring of a research nuclear reactor, a system in which all generated power is dissipated to the environment, can be performed via analysis of the heat rejection from the cooling system. Given an inlet water temperature and flow rate, the reactor power can be well-approximated from the outlet water temperature; however, the instrumentation to measure outlet conditions may not be robust or accurate. If we know how a cooling tower performs from historical data, but cannot measure the outlet temperature, a mathematical representation of the system can be inverted to obtain the outlet water temperature that describes the cooling capacity. Unfortunately, model inversion processes are computationally expensive. To address this, an artificial neural network (ANN) is implemented to assess the performance of a multi-cell cooling tower for a nuclear reactor. This approach leverages the Merkel model to obtain an extensive data set describing performance of the cooling tower cells throughout a wide array of potential operating conditions. The Merkel model is expressed as a function of four parameters: the inlet and outlet water temperatures, inlet air wet bulb temperature, and ratio of liquid-to-gas mass flow rates (L/G), which together provide a non-dimensional number indicative of cooling tower performance, called the Merkel integral. Computing a 4-dimensional data structure that describes finite combinations of the Merkel integral, an inverse model is then generated using an ANN to determine the cell outlet water temperature from the other three model parameters along with the computed Merkel integral. Compared to traditional model inversion methods, the ANN reduces the computational time by approximately 4 orders of magnitude, with effectively no sacrifice to solution accuracy, and could be applied for different cooling towers in the event the performance curve is known. Finally, three use cases of the ANN are then reviewed: (1) determining the cell outlet water temperatures when gas flow at rated conditions (GFRC) is known, (2) performing the prior case without knowledge of the GRFC, and (3) assessing performance differences between the individual tower cells.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

KRUSTY Reactor Design

The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a reactor design, development, and test program to demonstrate the nuclear operation of a Kilopower reactor. Kilopower systems are intended to provide between 1 and 10 kW(electric) in space, or on the surface of planets or moons, with a clear evolution to substantially higher power systems. KRUSTY was a prototype of a 1-kW(electric) highly enriched uranium–fueled Kilopower system. In March of 2018, KRUSTY successfully operated as a fission power system and was the first nuclear-powered operation of any truly new reactor concept in the United States in over 40 years. This paper discusses the design of the KRUSTY reactor along with the philosophy, goals, and engineering work that ultimately led to KRUSTY’s success.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗