Heat Pipe Modeling with Sockeye - PSU Seminar
An overview presentation of the heat pipe code Sockeye to a seminar at The Pennsylvania State University.
97 MATHEMATICS AND COMPUTING↗
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High-Fidelity Neutronics Model of a Realistic Heat Pipe Microreactor Report
The creation of a digital twin requires two parallel products: a virtual model and a physical asset. This report details the development the high-fidelity reactor physics model of a realistic heat pipe micro reactor for the virtual model. The realistic heat pipe micro reactor provides a generic model that research can utilize to make safeguards considerations. The model will be leveraged for capturing diversion and misuse scenarios expected to take place in a microreactor. A python wrapper has been developed to rapidly explore various acquisition pathways by manipulating the reactor physics model on-the-fly.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
SCALE Analyses of Scenarios in the TRISO-based Heat Pipe Microreactor Fuel Cycle
This report documents the application of the SCALE code to the analysis of a TRistructural-ISOtropic (TRISO)-based heat pipe microreactor (HPMR) within the context of its nuclear fuel cycle stages. The evaluation was conducted in support of the US Nuclear Regulatory Commission’s ongoing efforts to assess modeling capabilities for advanced non–light-water reactor technologies. The generic HPMR selected as a representative microreactor concept features a compact core design that incorporates TRISO fuel compacts, passive heat removal via heat pipes, and a transportable configuration intended for deployment in remote environments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Heat Pipe Modeling with Sockeye
This presentation gives an overview of the models in the heat pipe code Sockeye, as well as the current status of its validation.
42 ENGINEERING↗
SCALE Modeling of the Fast Spectrum Heat Pipe Reactor
As part of the severe accident analysis collaboration with Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC), SCALE models were developed for a fast-spectrum heat pipe reactor. These models were based on the Idaho National Laboratory (INL) Design A concept, which is an alternative design to the Los Alamos National Laboratory (LANL) Special Purpose Reactor (SPR), also known as the Megapower reactor. The model contains 1,134 heat pipes, surrounded by hexagonal fuel elements, with a potassium working fluid; the fuel is UO 2 with 19.75 wt% 235 U enrichment. The model contains axial beryllium oxide (BeO) reflectors above and below the active fuel region along with a radial alumina reflector containing 12 B 4 C control drums. The center of the core is left unfueled to make room for two shutdown control rods, one annular and one solid. The active region of the core was discretized into twenty axial and five radial zones to analyze spatial variations in power and burnup. Infinite lattice unit cell sensitivity studies were used to perform verification between the SCALE and INL models. The eigenvalue results agreed well with the reported results to within roughly 50 percent mille (pcm). Full-core model verification was performed by analyzing system eigenvalues with differing configurations of control drum and shutdown rod positions. These full core results all had eigenvalue differences less than 310 pcm. Control drum and shutdown rod worths were also compared, with differences of 3.2% or less. Using the verified model, the isotopic inventory and decay heat, as well as temperature feedback coefficients, were calculated and provided to SNL as input to the MELCOR severe accident code to analyze potential releases from this class of reactor. The results of the MELCOR analysis are provided in a different report.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor
MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Compliant heat exchangers, heat pipes and methods for making same
A method for fabricating heat exchangers using additive manufacturing technologies. Additive manufacturing enables the manufacture of heat exchangers with complex geometries and/or with internal and external integral surface features. Additive manufacture also facilitates the manufacture of heat exchangers with regional variations, such as changes in size, shape and surface features. In one embodiment, the present invention provides a heat exchanger with a helicoidal shape that provides axial elastic compliance. In one embodiment, the internal channel of the heat exchanger varies along its length. The internal channel may have a cross-sectional area that increases progressively from one end to the other. In one embodiment, the external shape of the tubular structure may be non-circular to optimize heat transfer with an external heat transfer fluid. In one embodiment, the present invention provides a heat pipe with an internal wicking structure formed as an integral part of the additive manufacturing process.
Lara-Curzio, Edgar↗
Metal wick crimping method for heat pipe internals
A wick assembly for use with a heat pipe is disclosed. The wick assembly includes an end plug including a wick receiving area, a wick, and a crimp. A portion of the wick is positioned about the wick receiving area. The crimp is positioned about the portion of the wick and the wick receiving area. The end plug, the portion of the wick, and the crimp are diffusion bonded.
Gross, David M.↗
A Sockeye Model of a Test at the Michigan Single Sodium Heat Pipe Facility
This conference summary describes a Sockeye model of an experiment at a single heat pipe test facility called MISOH1 at the University of Michigan.
97 MATHEMATICS AND COMPUTING↗
Data-Driven Model Predictive Control for Temperature Management of Heat Pipe Microreactor
To enable the self-regulating capability of heat pipe (HP) microreactors, an anticipatory control strategy through model predictive control (MPC) could proactively respond to potential disturbances and deviations in operating setpoints. However, a key factor prohibiting the widespread adoption of MPCs in nuclear applications is the effort and computational costs associated with learning and calibrating first-principles-based process models when the target system is complex and when there are gaps between modeled and target reactor systems. In this paper, we demonstrate data-driven MPC using three approaches for modeling the system dynamics, including a linear state-space model, feedforward neural network, and recurrent neural networks long short-term memory. We present the development and validation process of each model and compare the performance of data-driven MPCs in controlling the temperatures of selected HPs at the evaporator and condenser regions in a 37-HP-monolith system. Our results show that, qualitatively, all data-driven MPCs are producing similar control actions, while quantitatively, with artificial neural nets (especially feedforward neural nets), MPC can better follow drastic changes in setpoints with smallest errors.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Preliminary studies on diversion and misuse for TRISO-fueled heat-pipe-cooled microreactors
Microreactors are being developed by multiple reactor designers to mass-produce nuclear facilities for worldwide deployment. Including microreactors in a country’s energy portfolio introduces new concerns for international safeguards regarding the number of reactor locations. Microreactors have smaller quantities of nuclear material compared to current light-water reactors, which makes diverting enough material for clandestine purposes difficult for a single reactor. Diversion or misuse of multiple microreactors to obtain a significant quantity introduces a new acquisition pathway. This work examines diversion and misuse scenarios for a realistic heat-pipe-cooled microreactor to understand how these scenarios would effect reactor operations to determine if monitoring key parameters could reduce the burden of inspection on monitoring agencies. Here, we determined misuse caused too drastic of an effect on core operations to be a valid acquisition pathway for the realistic heat-pipe-cooled microreactor. Diversion was explored across varying levels of severity, where potential diversion scenarios could yield one significant quantity of material from between six and eleven microreactors. Through examining the critical control drum angle, excess reactivity, control drum worth, and power distribution changes from nominal were detected over the operational lifetime, which could indicate a divergence from normal operations.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Method of installing a heat pipe wick into a container of differing thermal expansion coefficient
A heat pipe is disclosed includes a container, a container lid including a groove defined therein, a wick, and an end plug operably coupled to the wick. The end plug includes a pin extending therefrom. The groove of the container lid is configured to receive the pin.
Swartz, Matthew M.↗
Heat pipe microreactor assembly VTB submission
This will be a submission to the virtual test bed (VTB). It features a multiphysics simulation of a fictitious heat-pipe-cooled microreactor assembly.
97 MATHEMATICS AND COMPUTING↗
The Monolithic Heat Pipe Microreactor Reference Plant Model Updates
This work presents the latest improvements to, and investigations performed with, the generic monolithic heat-pipe-cooled microreactor reference plant model for the United States Nuclear Regulatory Commission. This model serves as the foundation for the future detailed design evaluation models based on license applications. This model has been developed with the Comprehensive Reactor Analysis Bundle (BlueCRAB) and its specifications are based on open literature publications for the eVinci™ design . BlueCRAB is the U.S. Nuclear Regulatory Commission non-light-water reactor analysis system based on MOOSE, the Multiphysics Object-Oriented Simulation Environment framework, which can couple the Griffin, BISON, and Sockeye applications to resolve the various physics that are essential for the safety analysis of this type of reactor system. The core specifications include tristructural isotropic fuel, graphite monolith, graphite reflectors, and drums composed of graphite and B 4 C.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Pulsating Heat Pipes for Accelerator Magnets. Phase I Final Report
The primary objective of the Phase 1 award was to investigate the performance of cryogenic pulsating heat pipes (PHPs) and their use in practical large superconducting magnet systems. PHPs exhibit an effective thermal conductivity an order of magnitude and more over high purity metals such as copper and aluminum. They can therefore be utilized to reduce cold mass and effectively transfer heat from the magnet system components such as the coil pack, shields, or current leads to cryocoolers or cryogen reservoirs.
43 PARTICLE ACCELERATORS↗
Heat Pipe Enhanced Thermal Management System for GaN Switches in an Integrated Modular Motor Drive for Aircraft Propulsion
This paper presents the process of sizing a heat pipe enhanced thermal management system (TMS) for GaN switches in an integrated modular motor drive (IMMD) powering a 250 kW aircraft propulsion electric motor. A test setup mimicking the TMS and GaN switches assembly is constructed. The condenser chamber of the TMS is modeled in ANSYS Fluent while the GaN switches assembly is modeled in ANSYS Thermal allowing us to verify heat spreader temperature under test conditions and estimate GaN temperatures under test and actual application operating conditions.
DIRECT ENERGY CONVERSION↗
Heat Pipe Modeling Capabilities in Sockeye
Presentation to be given in a meeting with stakeholders of the Nuclear Energy Advanced Modeling and Simulation program for modeling related to microreactors. This presentation gives an overview of the capabilities and status of the heat pipe code Sockeye.
97 MATHEMATICS AND COMPUTING↗
Oscillating heat pipe integrated thermal management system for power electronics
Disclosed is a thermal management system for removing heat from a power electronic heat source, the system comprising: a base plate having a plurality of fluid passages there through and extending between and inlet side of the base plate and an outlet side of the base plate; and a plurality of heat transfer pipe segments respectively attached to one or more of the plurality of fluid passages at the inlet side of the base plate and the outlet side of the base plate, the plurality of heat transfer pipe segments arranged adjacent one another, the plurality of heat transfer pipe segments containing a two-phase working fluid, and the plurality of heat transfer pipe segments forming a continuous flow path through and back into the base plate for the two-phase working fluid.
Wang, Jim↗