MARVEL Dry Criticality Assembly [Slides]
This slideshow presents a high-level overview of the assembly sequence for the MARVEL dry criticality configuration.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
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CRAB/MELCOR Code to Code Comparison
Overview of the development of a CRAB/MELCOR workflow for mechanistic source term analysis
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Development of Digital Twin-Informed Predictive Maintenance for Critical Components in Advanced Reactors
Small modular reactors (SMRs) and microreactors, along with other advanced reactor (AR) technologies, are key to the future of nuclear energy. For these systems to achieve low operating costs, high reliability, and flexibility across applications, their operation and maintenance must be optimized. Digital twin (DT) technology is one of the technologies that enables real-time (or faster than real-time) monitoring and prognosis of critical components which are vital for operational efficiency, low costs, and enhanced safety of ARs, accelerating their deployment. DT technology provides dynamic virtual representation of physical assets by integrating real-time data, physics-based models, and advanced analytics, which is critical to optimizing the performance of the entire energy system throughout the life cycle. DTs empower engineers and operators to virtually explore different scenarios, configurations, and control strategies, allowing for the identification of optimal solutions that maximize reactor efficiency, safety, and economics.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Leveraging MARVEL and SPHERE to Demonstrate NEAMS Thermal Hydraulics Codes
The Single Primary Heat Extraction and Removal Emulator (SPHERE) and Microreactor Applications Research Validation and Evaluation (MARVEL) were utilized to demonstrate thermal hydraulics codes of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. A recent long-duration test of a sodium heat pipe conducted at the SPHERE facility was utilized for the validation of the heat pipe code Sockeye. The first 10 hours of operation featured a startup of the heat pipe from room temperature; this portion was modeled with Sockeye and compared to the experimental data. Measurements included several thermocouple measurements across the length of the heat pipe surface. Two of Sockeye’s heat pipe models were applied to experiment, as well as several models of the cooling at the condenser end of the heat pipe. Sockeye results gave a decent match to the experimental data, with notable discrepancy in the condenser section, where experimental data indicated a significant inactive length of the heat pipe. The MARVEL microreactor was modeled through a coupled simulation of System Analysis Module (SAM) and SubChannel Module (SCM). This multiscale coupled simulation was performed as an exercise to demonstrate the compatibility and applicability of these codes for thermal-hydraulic analysis of microreactors. The coupling between the SCM and SAM codes for multiscale modeling of MARVEL, was achieved through a domain overlapping approach. In this example a transient SAM simulation provided boundary conditions to SCM, and SCM provided pressure drop to SAM.
97 - MATHEMATICS AND COMPUTING↗
Assessing Several Modeling Approaches in Depletion Studies of a TRISO-Fueled Microreactor [Slides]
Abstract not provided.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Recent Developments and Findings of Heat Pipe Experiments for Microreactor Applications
Microreactor technologies are required to provide reliable carbon-free power generation in remote applications. The heat pipe–cooled microreactor concept, in particular, offers notable advantages due to the passive operation of heat pipes, enabling increased reliability and simplicity in a more compact form factor. There is a significant need for experimental work to aid and expedite the deployment of heat pipe microreactors due to their unique technological characteristics. Thus, there has been increased interest in heat pipe experiments by numerous institutions in order to support these efforts. Finally, the present work is a comprehensive review of recent heat pipe experiments from six major institutions, describing their designs, instruments, methods, and results. In addition, this paper also presents a background on heat pipe experiments along with discussions on instrumentation, accident scenarios, wick enhancement, and proposed future directions.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Heat Pipe Cooled Microreactors
Microreactor technologies are required to provide reliable carbon-free power generation in remote applications. The heat pipe cooled microreactor concept in particular offers notable advantages due to the passive operation of heat pipes enabling increased reliability and simplicity in a more compact form factor. There is a significant need for experimental work to aid and expedite the deployment of heat pipe microreactors due to their unique technological characteristics. Thus, there has been increased interest in heat pipe experiments by numerous institutions in order to support these efforts. The present work is a comprehensive review of heat pipe experiments, describing instrumentation, methods, phenomena of interest, and recent developments. In addition, legacy work on the operation of high-temperature heat pipes under irradiation is reviewed and discussed. Furthermore, the verification and validation efforts for the flagship heat pipe simulation code Sockeye are reviewed and requirements for future experiments are outlined. Lastly, future directions are proposed for heat pipe experimentation.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Dynamic Modeling and Characterization of Nuclear-grade Graphite
Idaho National Labs serves as the spearhead for many innovative energy solutions to the world's energy crisis. One such solution is the INL's Microreactor which is designed to deploy to extreme/remote environments where other sources of power are either unavailable or unreliable. In order to best design these energy solutions for their operational environments, it is crucial to understand how the design, components, and materials will respond to the environmental conditions. One key material in these innovative designs is a nuclear-grade graphite known as PCEA. This study examines the behavior of PCEA graphite under dynamic loading, similar to that which may occur in extreme environments. The objective is to characterize the dynamic behavior and produce an accurate, reliable constitutive model suitable for use in simulation tools such as INL's MOOSE. Graphite specimens were tested using a Split Hopkinson Pressure Bar (SHPB) to administer the dynamic compressive load. The SHPB was charged at various pressures to produce a range of strain rates on the material in compression. Data was acquired via strain gauges on the SHPB setup, from which stress, strain, and time data were collected. Analysis revealed the stress-strain behavior of the material as well as insights into the material behavior's relationship to strain rate. Further work must continue to characterize the various other dynamic behaviors of the material which will combine to create a substantially trustworthy constitutive model for this grade of nuclear-grade graphite. Ultimately, this will allow for realistic simulation of the material in reactor designs, allowing for prediction of design weaknesses and leading to improved designs for increased resilience, security, and reliability.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Verification of the Serpent-Griffin Workflow using the SNAP 8 Experimental Reactor
The Systems for Nuclear Auxiliary Power (SNAP) program accumulated extensive experimental measurements over a span of 15 years. This work builds upon previous studies which validated Serpent against experimental data for various criticality configurations of the SNAP 8 Experimental Reactor (S8ER). A 2-stage sequence is applied here with Serpent used for the generation of few-group cross sections and Griffin as the transport eigenvalue solver. Sensitivity studies are performed to qualify the effect and uncertainty of various parameters, comparing against reference models for the S8ER. The results from this work will then be expanded to create a generalized methodology for the Serpent-Griffin 2-stage approach for microreactor applications.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor
The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The Elevated-Temperature Cyclic Properties of Alloy 316L Manufactured Using Powder Metallurgy Hot Isostatic Pressing
According to developers at the GAIN 2019 microreactors workshop, advanced manufacturing (AM) is needed to fabricate microreactor components. A methodology has not been established to qualify AM for elevated-temperature nuclear construction. A potential approach is to use the wrought design methodology by establishing the AM properties as equivalent to or better than wrought. This investigation aims to understand the relationship between powder metallurgy hot isostatic pressing (PM HIP) and material characteristics for Alloy 316L as they relate to elevated-temperature component design and construction. The elevated-temperature cyclic behavior of Alloy 316L manufactured by PM HIP differs from wrought. Of particular significance is the reduced number of cycles to failure for the PM HIP Alloy 316L compared to wrought. Additional work to identify the microstructural feature that control the elevated-temperature cyclic behavior is needed.
36 MATERIALS SCIENCE↗
Microreactor Applications in U.S. Markets: Evaluation of State-Level Legal, Regulatory, Economic and Technology Implications
This report provides an evaluation of state-level legal, regulatory, economic and technology implications for microreactor applications in U.S. markets. The initial focus is on Alaska (AK) and Wyoming (WY) energy markets serving location-specific energy needs for electricity and heat. A state-by-state evaluation of current carbon/carbon-related policies and nuclear policies is conducted to assess broader market applicability in states undergoing energy transitions.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Detection of Diversion in a Realistic Heat Pipe Microreactor Using Supervised Machine Learning
Microreactors (MRs) pose new challenges for international safeguards. Here, their small size and mass reproducibility make them ideal for deployment in greater numbers and in remote locations, making the job of safeguards inspectors more challenging. Machine learning (ML) is currently being applied to many fields to augment human performance and increase automation; in particular, ML could be used to provide insight for international inspectors to help detect the diversion of nuclear fuel from MR cores. Four ML model types (k-nearest neighbors, decision tree, random forest, and histogram-based gradient boosted ensemble) were trained on integrated flux and critical control drum angle data generated with Serpent 2 for a realistic heat pipe MR design, achieving nearly 100% binary classification accuracy of nominal and diversion core configurations by the end of 1 full power year for three of the four model types. Regression model variants were also trained, using the same input data, for predicting the number of fuel pins diverted. Root-mean-square errors below 5% of the total number of fuel pins were achieved by the 1 full power year mark for all models.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Numerical Simulations of Passive Heat Removal from Mobile Microreactors
Not provided.
Nuclear Science & Technology↗
Alternatives to MARVEL Power Conversion -- Comparison of Stirling Engine Thermal Efficiency and Design to Other Power Conversion Cycles
PRESENTATION for CONFERENCE PAPER: The Microreactor Applications, Research, Validation, and Evaluation (MARVEL) Reactor is a small liquid-metal thermal reactor that will be built at the Idaho National Laboratory to demonstrate design and operating processes for microreactors, microgrid integration, and process heat applications. Power conversion in the MARVEL design is provided by Stirling engines, which have disadvantages in nuclear environments. Compared to Stirling engine performance, some alternative power cycles can increase power production when coupled to a liquid-metal thermal reactor. In this paper, the thermal efficiency of MARVEL’s power production with Stirling engines is compared to the thermal efficiency of power production with MARVEL and alternative power cycles. Those cycles include a superheated Rankine cycle, open and closed Brayton cycles, and a supercritical carbon dioxide cycle. All cycles (except the Stirling engines) were modeled with an intermediate helium loop to meet MARVEL’s principal design criteria. All models are simple designs with conservative assumptions for consistent comparison. Detailed optimization will depend largely on reactor location and application, and the relative merit of each cycle is discussed for different environmental conditions. The study informs significant early decisions on power cycle design and economics for deployment of advanced microreactors as they move from theory and concept to execution.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
MARVEL Guard Vessel System FEA and ASME Analysis
The INL is seeking to build the first microreactor in the United States, and the first advanced reactor in over 50 years. With the Microreactor Applications Research, Validation & EvaLuation (MARVEL) Project, this microreactor should be operational within a significantly faster timescale than has ever been seen in the nuclear industry. It uses two passive-convection loops, primarily Sodium Potassium eutectic (NaK), and Ga-In-Sn eutectic as an intermediate fluid. Due to the high reactivity of NaK, the primary coolant loop requires a secondary confinement boundary to prevent uncovering of the core in the event of a NaK leak. This secondary boundary is mostly accomplished through the addition of the Guard Vessel, which is a bucket-like metal containment, which is welded onto the primary boundary to maintain an inert gas environment and pressure
99 GENERAL AND MISCELLANEOUS↗
Nuclear Technology Selection and Preliminary Integration at St. Charles and Carrollton Chemical Plants
Nuclear energy is identified as a valuable candidate for decarbonizing industrial facilities as a non-emitting baseload energy source capable of dispatching energy as needed for an integrated facility. As industrial facilities are unique, analyzing what nuclear technology is the best candidate should be evaluated on a case-by-case basis. For two identified facilities of the Dow Chemical Company, an ideal mix of nuclear energy resources is calculated, and integration strategies are put forward. This research was funded through a Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher, specifically seeking near-term engineering solutions for the technical issue of introducing nuclear generated energy into the Dow chemical facilities. The research is split into four main phases: background on reactors and chemical facilities, preliminary engineering analysis determining how reactor technologies may be integrated with specific facilities, site characterization and discrete hazards analysis, and finally an optimization of reactor-facility selection through the use of Framework for Optimization of ResourCes and Economics (FORCE) tools.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Utilization of MARVEL: Advanced Sensors and Controls Deployment
Utilization of MARVEL: Advanced Sensors and Controls Deployment
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗