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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

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↗
DOE OSTI2016

PGSFR General Arrangement

The Nuclear Island of the PGSFR is constructed upon the seismic isolation structure bearing a common basemat of the Reactor Building and the Auxiliary Building. The Reactor Building is divided into two different functional structures: 1) containment structure comprised of a operating floor, head access area (HAA), annulus area and etc., 2) the other areas accessible to reactor vault area outside of the containment structure via the Auxiliary Building. The arrangement of the Auxiliary Building is composed of symmetric divisional areas where there are two steam generators, heat exchangers of the Decay Heat Removal System (DHRS), facilities of the Intermediate Sodium Purification System (ISPS), the Primary Inert Gas Processing System (PIGPS), a main control room (MCR), electric equipment rooms, and main steam valve houses (MSVH), and etc. The general arrangement of the Reactor Building and the Auxiliary Building of the PGSFR is currently in development. Not only NSSS but supporting systems have not been finalized yet. Moreover, the other facility buildings such as a Turbine Generator Building, a fuel cycle facility, radwaste treatment facilities, and etc. have to be designed with relevant interfaces to the Reactor Building and the Auxiliary Building. We can expect additional modifications to optimize the general arrangement. However, the basic principles such as sodium and non-sodium segregation, convenient accessibility and maintenance, economics of architectural volume would be kept in the design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Development of k{sub eff} in the unfolding process of the LOTUS reactor design

An innovative small thermal reactor concept named LOTUS, which is a pool-type reactor aimed at heat supply. The passive concept applied in the LOTUS and it has a separable core. It means that its core can passively unfold into four parts along its central axis only with the buoyancy and gravity when some accidents happen, especially the wall rupture of the pool caused by earthquake, aircraft impact and so on. This idea can effectively enhance the ability of pool-type reactor for resisting some severe accidents. We have executed some pre-computations in order to check a non-supercritical system for the transient simulation. In this paper, the development of k{sub eff} have been calculated by both stepwise steady-state and transient modes using Serpent, as well as including some sensitivity analyses about the k{sub eff} to unfolding velocity, inserting velocity of control rod (CR) and time step. The result showed that k{sub eff} value decreased definitely as the reactor unfolded and the consistent change exposed between the stepwise steady state and transient result when the unfolding angle varied from 0 to 1 degree. However, the stepwise steady state k{sub eff} negatively declined after that angle, which demonstrated that for such a dynamic unfolding change the transient mode was more suitable, peculiarly for considering delayed neutrons and neutron moderation. For sensitivity calculation, the k{sub eff} decreased more rapidly as the unfolding velocity and inserting velocity of CR increased, and a smaller time step corresponded to a more stable change. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Inductively Coupled Nonthermal Plasma Synthesis of Size-Controlled γ-Al 2 O 3 Nanocrystals

Gamma alumina (γ-Al 2 O 3 ) is widely used as a catalyst and catalytic support due to its high specific surface area and porosity. However, synthesis of γ-Al 2 O 3 nanocrystals is often a complicated process requiring high temperatures or additional post-synthetic steps. Here, we report a single-step synthesis of size-controlled and monodisperse, facetted γ-Al 2 O 3 nanocrystals in an inductively coupled nonthermal plasma reactor using trimethylaluminum and oxygen as precursors. Under optimized conditions, we observed phase-pure, cuboctahedral γ-Al 2 O 3 nanocrystals with defined surface facets. Nuclear magnetic resonance studies revealed that nanocrystal surfaces are populated with AlO 6 , AlO 5 and AlO 4 units with clusters of hydroxyl groups. Nanocrystal size tuning was achieved by varying the total reactor pressure yielding particles as small as 3.5 nm, below the predicted thermodynamic stability limit for γ-Al 2 O 3 .

77 NANOSCIENCE AND NANOTECHNOLOGY↗
NASA NTRS2011

Sustainability of the Catalytic Activity of a Silica-Titania Composite (STC) for Long-Term Indoor Air Quality Control

TiO2-assisted photocatalytic oxidation (PCO) is an emerging technology for indoor air quality control and is also being evaluated as an alternative trace contaminant control technology for crew habitats in space exploration. Though there exists a vast range of literature on the development of photocatalysts and associated reactor systems, including catalyst performance and performance-influencing factors, the critical question of whether photocatalysts can sustain their initial catalytic activity over an extended period of operation has not been adequately addressed. For a catalyst to effectively serve as an air quality control product, it must be rugged enough to withstand exposure to a multitude of low concentration volatile organic compounds (VOCs) over long periods of time with minimal loss of activity. The objective of this study was to determine the functional lifetime of a promising photocatalyst - the silica-titania composite (STC) from Sol Gel Solutions, LLC in a real-world scenario. A bench-scale STC-packed annular reactor under continuous irradiation by a UV-A fluorescent black-light blue lamp ((lambda)max = 365 nm) was exposed to laboratory air continuously at an apparent contact time of 0.27 sand challenged with a known concentration of ethanol periodically to assess any changes in catalytic activity. Laboratory air was also episodically spiked with halocarbons (e.g., octafluoropropane), organosulfur compounds (e.g., sulfur hexafluoride), and organosilicons (e.g., siloxanes) to simulate accidental releases or leaks of such VOCs. Total organic carbon (TOC) loading and contaminant profiles of the laboratory air were also monitored. Changes in STC photocatalytic performance were evaluated using the ethanol mineralization rate, mineralization efficiency, and oxidation intermediate (acetaldehyde) formation. Results provide insights to any potential catalyst poisoning by trace halocarbons and organosulfur compounds.

Coutts, Janelle L.↗
DOE OSTI2023

Digitalization Guiding Principles and Method for Nuclear Industry Work Processes

The commercial U.S. light-water reactor fleet has been operating at historical efficiency, reliability, and safety over the last decade. Nuclear power has the highest capacity factor of any other power generation technology while also serving as the largest baseload source for carbon-free energy. Despite this remarkable achievement, continued operations for many plants are threatened due to fierce electricity market competition and rising operations and maintenance costs of which continued maintenance of obsolete analog equipment is a contributor. The digital age and associated technologies are where the future lies in process control, and nuclear has yet to take full advantage of the capabilities offered therein. The Light Water Reactor Sustainability Program (LWRS) at Idaho National Laboratory (INL), sponsored by the Department of Energy, has a mission to help the light-water reactor fleet manage its foundational capabilities to continue providing safe and reliable carbon-free power. LWRS helps support that mission by providing scientific, technology-based solutions for advanced concepts of operations with a more viable business model that will allow the fleet to continue to operate at peak levels through extended plant operation. The LWRS Digitalization Project at INL seeks to leverage digital technologies to synthesize and transform work processes. We provide a state-of-the-art analysis of digitalized work processes in nuclear power and investigate ways in which researchers at INL and the nuclear industry can work together to identify what data to access, how to access it, what to do with the data, and most importantly, how to use the insights for decision-making across all levels within the business. Borne from these considerations, we present four guiding principles for digitalization: develop a coherent digitalization plan, apply human factors engineering, establish data governance, and anticipate unintended consequences. Together, these principles form a method that plants can use to effectively to digitalize nuclear industry work processes. Our guiding principles are informed by multiple knowledge sources. First, we document activities from the Work Digitalization Initiative, which was conceived as a means for nuclear organizations to help define and standardize the industry’s approach to digitalizing work. Second, we detail primary research conducted with industry professionals regarding drivers and barriers to digitalization adoption. We present survey results that demonstrate what the industry hopes to get out of digitalization and the ways that INL can continue to support the industry’s digital transformation. Third, we present a digitalization use case with industry partners NextAxiom Technology and Xcel Energy. The project objective was to transform the current condition report work process from paper to digital, incorporating digitalized principles. We report the development of the application and lessons learned. The accomplishments achieved by this research and development serve to identify critical needs for plant guidance in support of digitalization implementation and contribute to the knowledge and strategies available for utilities considering or undertaking digitalization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Flexible Siting Criteria and Staff Minimization for Micro-Reactors

The economic potential of micro-reactors is vast and underestimated. Commonly-emphasized applications include niche markets such as remote communities, mines and military bases. However, micro-reactors could be used as flexible energy generators also for larger markets, such as mobile and containerized agriculture and manufacturing facilities, district heating, micro-grids for data centers, sea ports, airports and hospitals. The implication is that micro-reactors may have to be deployed also in non-remote locations. Successful implementation of micro-reactors needs a navigable and predictable licensing process, technology-appropriate siting restrictions, risk-informed emergency and safety requirements, and practical operating and maintenance requirements. The primary goal of this project was to develop siting criteria that are tailored to micro-reactors deployable in densely-populated areas, e.g., urban environments. To achieve that goal, we compared the characteristics of the MIT research reactor (MITR) with those of leading micro-reactor concepts (e.g., eVinci, USNC, Aurora), and evaluated whether and how the MITR design basis (e.g., inherent safety features, engineered safety systems, source term, emergency planning and emergency operating procedures) and associated regulations may be applicable to these new micro-reactors as well. What makes MITR a unique analogue in this context is its small power rating (6 MWt) and physical size, mode of operations (24/7 with a somewhat more commercial flavor than typical university reactors), and especially its urban location. Of course significant differences exist, such as mission (power production vs. research) and the reactor design itself. Leveraging the MITR experience, this project was able to generate criteria that will allow micro-reactors to realize their full economic potential as flexible heat and electricity generators for a diverse portfolio of applications in non-remote locations. As such, the outcome of this project might encourage investment in and use of micro-reactors. A second goal of the project was to conceptualize a model of operations for micro-reactors that would minimize the staffing requirements, and thus reduce the cost of electricity and heat generated by these systems. Here too our approach was to systematically review the MITR experience and requirements, as well as survey the innovations in autonomous control technologies and monitoring (e.g., advanced sensors, drones, robotics, AI) that would permit a dramatic reduction in staffing at future micro-reactor installations. The scope of work was expanded after the start date to include also an evaluation of micro-reactor security, using the so-called consequence-based analysis, and the development of a methodology to perform dynamic risk assessment for micro-reactors, using system theory and modeling and simulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Vertical control of DIII-D discharges with strong negative triangularity

Through predictive modeling validated by a series of experiments on DIII-D, the vertical stability of low β diverted plasmas with strong negative triangularity (NT) ($\delta\sim-0.6$) is assessed. As a result of their unique magnetic geometry, NT plasmas feature larger Shafranov shifts and more elongated inner flux surfaces than positive triangularity counterparts, typically leading to enhanced vertical instability growth rates. However, coupling with the non-conformal vessel DIII-D wall reduces these growth rates to controllable values, providing a path forward for stabilizing strongly NT plasma with a diverted geometry. These discharges are used to validate GSdesign (part of the TokSys code suite) stability calculations in strong NT plasmas on DIII-D, with errors of no more than ~20% observed between modeled and experimentally measured growth rates. Additions of diagnostic noise and power supply tuning to the TokSys model are needed to accurately capture the time dependence of DIII-D NT discharges, assisting with the design of control schemes specific to the DIII-D poloidal field coils. Lastly, implications of these results on a future NT reactor are briefly described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Determining Reactor Operating Conditions to Prevent Xenon Walkaway Accident

The xenon walk-away is a postulated uncontrolled criticality that could occur if operators are forced to evacuate the control room in certain circumstances. This poster shows how previously held assumptions are not entirely valid and that less restrictions should be placed on operators and engineers than initially assumed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Demonstration of NEAMS Multiphysics Tools for Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 and FY20 expanded to additional fast reactor applications including lead cooled fast reactors (LFR) and sodium cooled fast reactors (SFR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, and (2) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR). First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study which began in FY19 and continued in FY20. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice. Along with the application on SFR and LFR, the capability of the tools has also been matured to deal with different reactor types and designs. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, in FY19 a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress for PROTEUS-SN and completed with MCNP. The MCNP power distributions were transferred to Nek5000 to perform thermal hydraulic calculations of the control and fuel assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

MARVEL Lessons Learned – 2026 Edition Fabrication, Fuel Procurement, Quality Assurance, and Safety Basis

The Microreactor Applications Research Validation and Evaluation (MARVEL) project is intended to be among the first U.S. advanced reactor demonstrations in four decades. By virtue of being sponsored directly by the U.S. Department of Energy (DOE), the reactor is expected to benefit the broader nuclear community by exercising design processes, safety reviews, and supply chains. This project is committed to publicly documenting key lessons learned along the way toward demonstration. Building on a 2025 report that focused on lessons learned from guard vessel fabrication and Primary Coolant Apparatus Test (PCAT) testing (Abou-Jaoude et al. 2025), this 2026 edition captures findings spanning structural fabrication, quality assurance (QA), reactivity control system (RCS) assembly, fuel fabrication and shipment, safety basis development, balance-of-plant design, and overall reactor design.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2013

Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated

Worrall, Andrew↗
NASA NTRS1991

Development of a proton-exchange membrane electrochemical reclaimed water post-treatment system

A single-cell electrochemical reactor that utilizes a proton exchange membrane (PEM) as a solid electrolyte is being investigated for posttreatment of reclaimed waste waters with low or negligible electrolyte content. Posttreatment is a final 'polishing' of reclaimed waste waters prior to reuse, and involves removing organic impurities at levels as high as 100 ppm to below 500 ppb total organic carbon (TOC) content to provide disinfection. The system does not utilize or produce either expendable hardware components or chemicals and has no moving parts. Test data and kinetic analysis are presented. The feasibility and application for water reclamation processes in controlled ecological environments (e.g., lunar/Mars habitats) are also presented. Test results show that the electrochemical single cell reactor provides effective posttreatment.

Kaba, Lamine↗
DOE OSTI2016

SHARP Multiphysics Tutorials

SHARP, developed under the NEAMS Reactor Product Line, is an advanced modeling and simulation toolkit for the analysis of advanced nuclear reactors. SHARP is comprised of three physics modules currently including neutronics, thermal hydraulics, and structural mechanics. SHARP empowers designers to produce accurate results for modeling physical phenomena that have been identified as important for nuclear reactor analysis. SHARP can use existing physics codes and take advantage of existing infrastructure capabilities in the MOAB framework and the coupling driver/solver library, the Coupled Physics Environment (CouPE), which utilizes the widely used, scalable PETSc library. This report aims at identifying the coupled-physics simulation capability of SHARP by introducing the demonstration example called sahex in advance of the SHARP release expected by Mar 2016. sahex consists of 6 fuel pins with cladding, 1 control rod, sodium coolant and an outer duct wall that encloses all the other components. This example is carefully chosen to demonstrate the proof of concept for solving more complex demonstration examples such as EBR II assembly and ABTR full core. The workflow of preparing the input files, running the case and analyzing the results is demonstrated in this report. Moreover, an extension of the sahex model called sahex_core, which adds six homogenized neighboring assemblies to the full heterogeneous sahex model, is presented to test homogenization capabilities in both Nek5000 and PROTEUS. Some primary information on the configuration and build aspects for the SHARP toolkit, which includes capability to auto-download dependencies and configure/install with optimal flags in an architecture-aware fashion, is also covered by this report. A step-by-step instruction is provided to help users to create their cases. Details on these processes will be provided in the SHARP user manual that will accompany the first release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Molten Salt Reactor Experiment (MSRE) Lay-Up - 19187

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. 2019 is the 50. anniversary of the shutdown of the Molten Salt Reactor Experiment (MSRE). MSRE ran a brief four years in the 1960's but earned an enduring legacy as an innovative reactor technology concept which has enjoyed a renewal of global interest in the past several years. In 2018, it was given elevated status among advanced reactor programs within DoE's Nuclear Energy office. The MSRE was a graphite-moderated, liquid fueled reactor that operated from June 1965 through December 1969. Although it was shut down nearly 50 years ago, certain systems within the reactor have continued to operate to maintain the facility in a safe and stable condition until it can be demolished as part of the comprehensive cleanup of the Oak Ridge Reservation. These systems are composed of a complex network of pipes, valves, gauges and associated equipment that control air filtration and pressure buildup resulting from fluorine gas generated through radiolysis of the fluoride salt which once fueled the reactor. In 2016, an MSRE Engineering Evaluation Team was chartered to assess current facility conditions and develop recommendations to maintain the facility in a safe state until deactivation and demolition commences. Plans are now underway to 'lay up' the MSRE facility until deactivation and demolition can occur. The proposed MSRE Facility Layup will minimize maintenance costs, reduce personnel risk of injury and exposure, provide reliable electric service to the facility and require zero personnel occupancy at MSRE. In the proposed layup approach a new source of electrical power to required loads will be routed through the facility, and new gas processing and ventilation systems will be installed that will be optimized for safe processing of the constituent hazardous gasses that exist in the legacy vessels at MSRE. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Control And Optimization Modular Modeling Application For Nuclear Deployment

The purpose of the COMMAND code is to provide a flexible, scalable tool for use in developing, integrating, and testing the technologies necessary for achieving autonomous operations of advanced nuclear reactors. The code enables users to efficiently implement custom simulations and experiments by combining key methods from different software modules. These modules are focused on: modeling and simulation tools, such as nuclear simulation tools used for high-fidelity modeling (e.g., Reactor Excursion and Leak Analysis Program [RELAP5-3D] and Monte Carlo N-Particle [MCNP]); machine learning and optimization tools (e.g., anomaly detection and data-driven modeling techniques); advanced control in its digital, high-performance, and supervisory control forms (e.g., proportional integral derivative (PID) control and model predictive control (MPC); and integration with hardware through industrial communication protocols. To ensure flexibility and scalability, COMMAND was designed to be both modular—the software “pieces” all inherit from generic building blocks and can be combined and connected to create complicated simulations—and high performing—designed for parallel processing, enabling simulations and experiments to take advantage of multi-core computers, servers, and nodes. The code is written in the Python programming language due to the language's popularity, active community, and open-source and cross-platform nature. Maintaining consistency with other simulation tools used within the nuclear energy community, users implement simulations and experiments through text input files, which define components, parameters, connections, etc., through lines of text. Given that COMMAND is written in Python, these input files are native Python scripts, and so use the standard Python structure and formatting. This also enables users to take advantage of Python's extensive package library to develop custom capabilities for their specific use cases.

Faber, Jacob [Idaho National Laboratory (INL), Ida↗
DOE OSTI2021

Ring Model Development and Validation for Prismatic HTGR Core Thermal-hydraulics and Safety Analysis

Because of the complex core geometry, prismatic high temperature gas-cooled reactors (prismatic HTGRs) often exhibit complex thermal fluid behaviors during both normal operating and transient conditions. Most HTGR designs rely on passive safety system for decay heat removal, such as the reactor cavity cooling system (RCCS). During postulated accidents like Pressurized Conduction Cooldown (PCC) event, the decay heat is first radially transferred from the core region to the reactor vessel outer surface, then to the RCCS cooling panels. The peak fuel temperature is controlled by heat transfer mechanisms with two distinctive characteristic length scales, i.e., the core-wise effective heat conduction and local heat conduction in the fuel pellet scale. As both length scales are essential to determine the fuel temperature, from the modeling perspective, computer codes must be able to capture heat transfer in both scales. This is challenging for both computational fluid dynamics (CFD) tools because of extremely large amount of computation resources required, and for system analysis codes because of the challenge to model the complex core geometry. Under the support of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, efforts have been pursued to support HTGR technology development and its modeling and simulation needs. There is a particular need for advanced modeling and simulation tools to predict thermal-fluid behavior during safety-related transients. In our previous studies, the ring model was adopted in SAM and further developed to simulate the normal operating condition and a PCC event using the MHTGR-350 design of General Atomics as the reference design. This current work represents a continuation of these previous efforts, and the focus is to critically review and examine simplifications and assumptions made to develop the ring model, and to perform code validation using experimental data from an integral-effect test facility, the High Temperature Test Facility (HTTF) at the Oregon State University. In this study, the test PG-27 from the HTTF test suite was selected for code validation purpose. The test PG-27 is a transient test designed to simulate the PCC event of the MHTGR design. Very good agreements between SAM prediction and experimental measurements were found in both coolant and solid structure temperatures during the transient.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗