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DOE OSTI2022

Snowmass2021 Cosmic Frontier White Paper: Observational Facilities to Study Dark Matter

We present an overview of future observational facilities that will significantly enhance our understanding of the fundamental nature of dark matter. These facilities span a range of observational techniques including optical/near-infrared imaging and spectroscopy, measurements of the cosmic microwave background, pulsar timing, 21-cm observations of neutral hydrogen at high redshift, and the measurement of gravitational waves. Such facilities are a critical component of a multi-pronged experimental program to uncover the nature of dark matter, while often providing complementary measurements of dark energy, neutrino physics, and inflation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Design of a Test Article and Test Facility to Investigate Flow-Induced Vibration of a Helical Coil Steam Generator

NuScale Power LLC (NuScale) is developing a small modular reactor (SMR) with unique design features and arrangements that require testing to provide quality data for safety analysis codes and design validation, as well as for technology maturation. A multi-phase testing program is underway at NuScale to demonstrate that its unique helical coil steam generator design is not susceptible to flow-induced vibration during operation. The phases of the testing program include modal testing of individual tubes during test article fabrication and modal and flow testing after test article completion. NuScale has contracted SIET S.p.A. in Piacenza, Italy for the design and fabrication of the test article and test facility. In-fabrication testing was performed at the site of test article build in Cremosano, Italy. Modal testing and flow testing of the completed test article will be performed at SIET’s facilities in Piacenza. This paper discusses the design of the test article and test facility, including supporting systems, required to generate the quality data required to support NuScale’s comprehensive vibration assessment program as part of the overall licensing effort for NuScale’s first-of-a-kind small modular light water reactor.

Flow-induced vibration, helical coil, steam genera↗
DOE OSTI2022

Whitepaper submitted to Snowmass21: Advanced accelerator linear collider demonstration facility at intermediate energy

It is widely accepted that the next lepton collider beyond a Higgs factory would require center-of-mass energy of the order of up to 15 TeV. Since, given reasonable space and cost restrictions, conventional accelerator technology reaches its limits near this energy, high-gradient advanced acceleration concepts are attractive. Advanced and novel accelerators (ANAs) are leading candidates due to their ability to produce acceleration gradients on the order of 1--100~GV/m, leading to compact acceleration structures. Over the last 10-15 years significant progress has been achieved in accelerating electron beams by ANAs. For example, the demonstration of several-GeV electron beams from laser-powered capillary discharge waveguides, as well as the proof-of-principle coupling of two accelerating structures powered by different laser pulses, has increased interest in ANAs as a viable technology to be considered for a compact, TeV-class, lepton linear collider. However, intermediate facilities are required to test the technology and demonstrate key subsystems. A 20-100 GeV center-of-mass energy ANA-based lepton collider can be a possible candidate for an intermediate facility. Apart from being a test beam facility for accelerator and detector studies, this collider will provide opportunities to study muon and proton beam acceleration, investigate charged particle interactions with extreme electromagnetic fields (relevant for beam delivery system designs and to study the physics at the interaction point), as well as precision Quantum Chromodynamics and Beyond the Standard Model physics measurements. Possible applications of this collider include the studies of $\gamma\gamma$ and $e$-ion collider designs.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Cryogenic User Facilities for R&D on Noble Liquid Detectors and Low Temperature Devices

Cryogenic test facilities are critical infrastructure for physics experiments in a variety of fields, perhaps most notably for particle detection with noble liquid detectors, low-temperature device development, and quantum information research. However, considerable investment and technical knowledge are required to construct and operate such facilities. This white paper discusses proposals for user facilities aimed at broadening the availability of testing capabilities for the scientific community.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Integration of Nuclear Material Accounting Data and Process Monitoring Data for Improvement on Detection Probability in Safeguarding Electrochemical Processing Facilities (Final Technical Report)

The KAERI advanced spent fuel conditioning process (ACP) process is a critical component of the US- South Korean nuclear cooperation and the following “123 Agreement.” Its development has received considerable attention in both countries. The ACP is an electrochemical processing (pyroprocessing) that recycles over 96% of the used nuclear fuel (UNF). It is also intrinsically proliferation-resistant in theory. In normal operation, the U/TRU product is very hot radiologically. In addition, the Cm provides a high level of spontaneous neutrons, making the product unsuitable for weapon use. However, as pointed in some study, “the need for safeguards to protect against the diversion and misuse of separated plutonium applies essentially equally to all grades of plutonium.” As pointed by many studies, the well-established traditional Nuclear Material Accounting (NMA) approach cannot be directly applied to electrochemical processing because of the lack of an input accountability tank, the non-continuous material flow, and the unsatisfactory level of confidence in sampling methods. Therefore, nuclear safeguards remain a grand challenge in the developing of commercial electrochemical separations facilities, especially around the heart of such facilities, the electrorefiner (ER) systems. In contrast to NMA data, process monitoring (PM) data is normally an indirect measurement of the SNM and is acquired much more frequently. In a broad sense, PM includes monitoring by various types of equipment, e.g. radiation detectors, cameras, voltage, current sensors. Because it is already being collected by the operator, the additional cost to safeguards is low. It has long been believed that PM data can supplement NMA data and help improve safeguards, although the benefits are hard to quantify. The U.S. DOE’s Material Protection, Accounting, and Control Technology (MPACT) campaign has made substantial investments into innovative PM sensor technology and predictive model development for real- or near real-time measurement and prediction of molten salt density and level, salt composition and actinide concentration especially Pu, the cell voltage, and the cell current to supplement traditional NMA. For aqueous-based reprocessing facilities, it is reported that PM, integrated with traditional NMA, have a high detection probability for specific diversions. For electrochemical reprocessing, preliminary studies have shown that PM data can support traditional NMA in various ways by providing a basis to estimate some of the in-processing nuclear material inventories. Despite early success, further studies on fusion of PM data and NMA data are still needed, which is the goal of this proposed work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Planning Tool for Strategic Evaluation of Facility Plans - 13570

Savannah River National Laboratory (SRNL) has developed a strategic planning tool for the evaluation of the utilization of its unique resources for processing and research and development of nuclear materials. The Planning Tool is a strategic level tool for assessing multiple missions that could be conducted utilizing the SRNL facilities and showcasing the plan. Traditional approaches using standard scheduling tools and laying out a strategy on paper tended to be labor intensive and offered either a limited or cluttered view for visualizing and communicating results. A tool that can assess the process throughput, duration, and utilization of the facility was needed. SRNL teamed with Newport News Shipbuilding (NNS), a division of Huntington Ingalls Industries, to create the next generation Planning Tool. The goal of this collaboration was to create a simulation based tool that allows for quick evaluation of strategies with respect to new or changing missions, and clearly communicates results to the decision makers. This tool has been built upon a mature modeling and simulation software previously developed by NNS. The Planning Tool provides a forum for capturing dependencies, constraints, activity flows, and variable factors. It is also a platform for quickly evaluating multiple mission scenarios, dynamically adding/updating scenarios, generating multiple views for evaluating/communicating results, and understanding where there are areas of risks and opportunities with respect to capacity. The Planning Tool that has been developed is useful in that it presents a clear visual plan for the missions at the Savannah River Site (SRS). It not only assists in communicating the plans to SRS corporate management, but also allows the area stakeholders a visual look at the future plans for SRS. The design of this tool makes it easily deployable to other facility and mission planning endeavors. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

A novel approach to criticality accident detection for a legacy facility

For operations on the Sellafield site involving fissile material a Criticality Accident Alarm System (CAAS) has traditionally been installed unless an omission case can be made. An omission case would examine if, in the absence of all controls, a criticality could reasonably be expected or whether the potential dose to an operator would remain below a maximum acceptable dose. It is a UK legal requirement for adequate emergency arrangements to be in place for hazards on a nuclear licensed site. For facilities where criticality is credible there is a regulatory expectation that a -Criticality Warning System- should be in place. A -Criticality Warning System- is not necessarily a full CAAS and alternative means of criticality detection may be part of an adequate emergency plan. The Sellafield site has a Magnox fuel reprocessing facility which has been operated since the 1960's. Evaporation of the Medium Active (MA) liquid waste from this process is undertaken in an associated building. Following extensive review of the plant and process a number of maloperations which can route fissile material to the MA evaporator were identified. Adequate controls to prevent these scenarios were shown to be in place. In the unlikely event of a failure of controls leading to sustained loss of fissile material a criticality incident may occur and hence an emergency plan was determined to be necessary. The novel emergency arrangements which were developed utilized the existing gamma monitor network located across multiple building floors. These gamma monitors are present for radiological safety. This emergency arrangement would require rapid interpretation of the gamma alarms by an operator and the manual triggering of a criticality evacuation alarm. The gamma monitors were demonstrated to be able to detect the minimum incident of concern (10{sup 15} fissions) as well as being able to alarm during the ramp up in power of a large criticality incident (10{sup 18} fissions). It was shown that a criticality accident could readily be distinguished from a radiological event (e.g. loss of containment). To support this work, criticality excursion analysis was used to examine the likely characteristics of a criticality in the MA evaporator. This novel arrangement was justified due to a number of factors including: the limited remaining operational lifetime of this facility (<5 years), the low frequency of maloperation coupled with a failure of controls and the crucial risk reduction provided by the continued reprocessing of Magnox fuel. Furthermore, the time and cost involved with design and installation of a standard CAAS would be substantially greater than a plan based upon existing instrumentation. These arguments were accepted by the UK regulator the Office for Nuclear Regulation (ONR). (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Legacy Facility Risk Reduction - Challenges, Strategies, Partnerships, and Benefits Lawrence Livermore National Laboratory - 18463

As Lawrence Livermore National Laboratory (LLNL) carries out its national security missions, we strive to ensure our facility portfolio meets the latest technologies and addresses challenges presented by its aging infrastructure. Established in 1942 as a Naval Air Station, repurposed in 1952 by the Atomic Energy Commission (AEC), LLNL continues to host a robust National Nuclear Security Administration (NNSA) Research and Development (R and D) mission. As an aging site, many of its facilities have met their life cycle contribution. As process contaminated facilities are decommissioned, new challenges arise in their near-term stewardship and long-term disposition actions, presenting risk reduction opportunities. The challenge has been to determine approaches to steward such risks, stabilize, advocate for long-term mitigation, and take actions to eliminate risks. Encumbered within is the balance of priorities in supporting ongoing and emerging mission and those of past legacy activities. This paper will endeavor to highlight Challenges and Benefits presented by such a paradigm, where we identified Strategies and Partnerships working towards mutual Risk Reduction activities. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental Activities with the IELLLO Lithium-Lead Facility

Fusion power offers the prospect of an almost inexhaustible source of energy for future generations. Although this may be true, it also presents so far serious scientific and engineering challenges. Breeding Blanket is one of key components of future fusion reactors, since it directly involves tritium breeding and energy extraction, both of them critical to development of fusion power. To deal with the development of Breeding Blankets, ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) designed and built a lead lithium loop (IELLLO: Integrated European Lead Lithium Loop): the experiment aims to support R and D activities for the conceptual and preliminary design of the European test blanket systems. IELLLO was completed, installed and operated at ENEA Brasimone Research Centre. The purpose of this work is to describe the experimental activities that have been carried out concerning the qualification of the facility, and relative instrumentations, for lead lithium applications. A series of thermo-hydraulic investigations on IELLLO facility has been elaborated: lead lithium pressure transducers, heat exchanger, permanent magnet pumping system, mass flow meters and air cooler have been tested. The absolute pressure meter in contact with Pb-16Li was qualified in pressure range up to 10 bar and temperature up to 400 deg. C. The installation of new kinds of Pb-16Li differential pressure transducers will allow to check the pump characteristic curve and accurately calculate valves pressure drops. These factors could give a better information for the thermo-hydraulic qualification of the facility. Vortex Mass Flow Meter has shown good results in terms of accuracy. The performance has been validated by thermal balance with a dedicated heating system. Nevertheless, the installations of different types of mass flow meters will enable to test IELLLO at ITER relevant conditions, and at lower mass flow rates than those tested up to now. The new pumping system with a PM Pump, was qualified for the first time in Pb-16Li, and the procedure for the operation and maintenance of the pump was developed. The pump is able to supply a mass flow rate in the range from 1.2 to 2.4 kg/s with a temperature of Pb-16Li up to 550 deg. C in the test section. Significant results have been obtained also for the heat exchanger, which has shown high efficiencies, even more than expected. Likewise, the Air Cooler has been verified to be able to cool Pb-16Li in a relatively short time, even at the worst temperature and mass flow conditions. (authors)

42 ENGINEERING↗
DOE OSTI2016

MIT reactor (MITR) driven subcritical facility for the fluoride-salt-cooled high-temperature reactor (FHR) - 18623

The Fluoride-salt-cooled High temperature Reactor (FHR) uses clean salt coolants, carbon-matrix coated-particle fuel similar to that used in High-temperature Gas-cooled Reactors (HTGRs) and can be coupled to a Nuclear Air-Brayton Combined Cycle (NACC). Recent assessments indicate favorable economics and safety characteristics, but no FHR has been built. The question is what experimental facilities should be constructed to reduce technical uncertainties before a decision to build a test/demonstration reactor? The MIT Reactor (MITR) design and license would allow the construction and operation of a subcritical facility with 700 deg. C salt circulating through multiple full-width partial-height fuel assemblies operating with a power density up to 30% of a commercial FHR. This option would allow hot systems testing as a major step toward building the test/demonstration reactor. Preliminary system design, power control options, testing capabilities, and key nuclear characteristics of such a reactor-driven subcritical facility are described. A method of deriving subcritical multiplicity using surface source has been proposed and verified in this study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Test facility for fission chamber assemblies - 239

The Institution -Project Center ITER- with participation of the National Research Nuclear University MEPhI are developing the neutron diagnostics -Divertor Neutron Flux Monitor- (DNFM) for the International Thermonuclear Experimental Reactor (ITER). This diagnostics shall ensure measurement of the total neutron yield and ITER fusion power. Assemblies of fission chambers (FC) with different fissile materials and sensitivities are planned to be used for the neutron flux measurement. Study of fission chambers characteristics and acceptance testing of the diagnostic system will be carried out at the testing facility of the neutronic laboratory of the 'TRINITI' Institute. The facility includes: compact powerful neutron generator with the power, control and monitoring system; remote handling system for FCs positioning and movement; novel three channels data acquisition and processing system. Each acquisition channel contains some measuring circuits with signal processing. This set of circuits provides neutron flux measurement in the range of at least 10 decades for each single FC. Data acquisition and processing system features are: signal digitizing with sample rate up to 125 MHz; real time data processing with computation and data update cycle from 0.1 ms. This paper examines the structure, composition and specification of the test facility. Application software, hardware solutions and the results of fission chambers prototypes tests are discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Development of Fusion Neutronics Methodology, Software, and Testing Facility in FDS Team

Fusion neutronics including methodology, software, experiments, etc. is studied concerning the generation, utilizing and protection of neutrons in fusion facilities. Achievements have been made in the past decades to solve the new problems fusion neutronics. However there is still some way to go such as following. Automatic modeling method based on CAD has been developed to avoid the heavy burden of manual geometry description. The pre-processing costs much time to simplify the CAD model for converting to Constructive Solid Geometry (CSG), 2D or 3D mesh based geometry. This problem has become a new focus. Deep penetration of radiation shielding is notable problem for fusion. Adaptive variance reduction methods such as Consistent Adjoint Driven Importance Sampling (CADIS) were developed to accelerate convergence. However the gap still exists between engineering requirements and current methodology. The geometry simplification, homogenizing based tally, error in nuclear data and error propagation in coupled calculation, etc., causes considerable uncertainty, which leads to 'safety factor' applied in simulation results to achieve required safety margin. It can significantly raise the cost and even makes it impractical to initiate the fusion facilities. Uncertainty qualification, error analysis and error propagation problem should be studied. New methodology and software should be validated. For example the (n,p), (n,a), (n,t) cross section of Cu, Fe, W, Ni which is important for evaluation of material radiation damage is absent. The validation of neutronics properties of special components such as first wall, blanket and coil are needed. The D-T fusion neutron generator can directly represent the neutron environment in fusion reactors. However the neutron yield is several orders of magnitudes lower than requirements. To resolve obstacles in fusion neutronics, broad improvements of methodology, software and testing facility have been achieved by FDS Team in recent years and the main progress was introduced in this paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Post Irradiation Examination of Fast Neutron Irradiated 14YWT Tubes at Nuclear Science User Facilities

Cladding materials for next generation nuclear reactors require resistance to high doses of fast neutron flux. Work through the Department of Energy's (DOE) Nuclear Technology Research and Development (NTR and D) Program's Advanced Fuel Campaign has identified 14YWT, a nano-ferritic oxide dispersion strengthened alloy as a promising alloy and has made a 50 kg batch of material, FCRD-NFA1, with nominal composition Fe-14Cr-3W0.4Ti-0.2Y in weight percentage. Plates were extruded from milled powder at Oak Ridge National Laboratory and then electro-discharge machined (EDM) into two sizes of tubes at Los Alamos National Laboratory (LANL). The OD of the tubes were polished to remove the EDM recast layer. Twenty-four of these tubes were placed into the BOR60 reactor, a sodium fast reactor, at the RIAR lab in Dimitrovgrad Russia. Irradiation conditions depended on exact location within the reactor, but the two samples in this study (L108 and L202) were irradiated to 7 dpa at 360-370 deg. C (Sample L108), and 6 dpa at 385-430 deg. C (L202). More detailed temperature models are pending for sample L202. Post irradiation examination took place at both LANL and the University of California, Berkeley (UCB). Facilities used at LANL (CMR Wing 9 Hot Cells, Electron Microscopy Lab (EML)) are part of a pending NSUF partner facility application, while the work at UCB was completed under a Nuclear Science User Facility, Rapid Turnaround Experiment Proposal (NSUF-RTE)

36 MATERIALS SCIENCE↗
DOE OSTI2018

Activation analysis of the concrete in the Fleurus cyclotron facility - 25604

The purpose of this paper is to discuss the approach and the results of a recent activation analysis concerning a cyclotron facility located in Fleurus, Belgium, where isotopes of medical applications were produced by irradiating either solid (Tl) or gaseous targets (Xe) with 20-30 MeV proton beams. This paper discusses the choice of the intra-nuclear cascade model implemented in MCNP/X to correctly simulate the spallation reactions of the protons in the target, since no nuclear data are available for the isotopes of interest. TENDL nuclear data library, which is based on the models of the nuclear code system TALYS, plays a key role in the selection of the model. The neutron flux distribution throughout the installation and the activation rates in the concrete walls are then calculated with the help of a comprehensive MCNPX model of the facility. The estimated activities are in good agreement with the measurements taken in 2017 in the frame of the ongoing decommissioning of this facility. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Atalante research facility new mass limits in a unit containing plutonium and minor actinides - 14402

ATALANTE, located in Marcoule, is one of the main Nuclear Facilities of the French Alternative Energies and Atomic Energy Commission (CEA). This facility is mainly dedicated to research and development for nuclear energy related to the back end of the fuel cycle: spent fuel and final waste management. It is a large research complex housing chemistry and analytical laboratories and shielded cells (240 glove boxes and 11 shielded cells), equipped for basic studies through to demonstration experiments. Within the facility, one unit can contain plutonium and curium at the same time. In this case, nuclear criticality safety is achieved by exercising control over mass. A subcritical mass limit is defined for plutonium (280 g for {sup 239}Pu) and an exemption mass limit (3 g for {sup 245}Cm) is used for the actinide. In order to increase the maximum mass (in particular {sup 245}Cm) which can be handled in the unit, new mass control limits were studied, based on control over the sum of the mass ratio to the upper subcritical mass, noted m{sub lim}, for each nuclide (350 g for {sup 239}Pu; 34.8 g or 23 g for {sup 245}Cm) as shown in equation (1) below. m({sup 239}Pu)/ m{sub lim}({sup 239}Pu)+ m({sup 245}Cm)/m{sub lim}({sup 245}Cm)≤1 (1) To validate this criticality safety control method, a study was conducted to demonstrate that, for different masses of {sup 239}Pu and {sup 245}Cm meeting equation (1), the calculated effective multiplication factor is always lower than the effective multiplication factor corresponding to a sphere of 350 g of pure {sup 239}Pu, moderated, with 20 cm water reflection. The two pseudo fissile media are: - metallic plutonium (100% {sup 239}Pu, density 19.86), moderated with water, - metallic curium (100% {sup 245}Cm, density 13.566), moderated with water. The calculation model is a sphere of {sup 245}Cm metal positioned in a sphere of {sup 239}Pu metal, with optimum moderation for each fissile material, and a full water reflection (20 cm). The results show that, under certain conditions, the effective multiplication factor of the mixture can be higher than the effective multiplication factor of each pure fissile material when using the upper subcritical mass of 34.8 g for {sup 245}Cm. When using 23 g of {sup 245}Cm, the k{sub eff} of the mixture is always lower than the k{sub eff} of pure {sup 239}Pu. These calculations enable the safety documentation to be changed by introducing a new process operational procedure, based on equation (1). This change in procedure allows the operator to work with a larger quantity of curium in the unit. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

High Performance Heat Pipe Power Transient Testing at SPHERE Facility

Microreactors are being researched, designed, and built at Idaho National Laboratory (INL). Microreactors are small reactors defined at less than 20MW of power. These reactor concepts are also being looked at throughout industry for various applications. An important aspect of these reactor designs is economic feasibility i.e. lower overnight capital cost. The driving factors for implementing microreactors are quick setup and takedown, minimal operators, and the ability to manufacture them readily and to fit in mid-sized containers for transport. A specific area of research to aid in successful integration of these factors within the designs is passive heat removal of the core’s thermal power. Interest in heat pipes to achieve this passive heat removal has been shown across multiple industry partners. Because of this interest, INL has developed a test facility to facilitate experimental tests for sodium filled heat pipes. INL has developed the Single Primary Heat Extraction and Removal Emulator (SPHERE) facility to run experiments on high performance, sodium filled heat pipes. As mentioned above, heat pipes are passive heat transfer devices. Radially, heat pipes are broken up into an outer wall, a small annular gap, a wick structure, and a centerline gap. They function by utilizing latent heat transfer. Heat pipes are traditionally separated into three regions, an evaporator (heat input), an adiabatic region, and finally a condenser region (heat removal). As heat is being applied to the evaporator, the working fluid undergoes a phase change to a vapor. This phase change causes a differential pressure across the axial length of the pipe driving flow down the center gap of the heat pipe. The vapor flows down past the adiabatic region to the condenser where the heat is removed. This heat removal forces the working fluid to phase change back to a liquid. The wick structure is then utilized to drive the flow back towards the evaporator by capillary forces. This backflow is aided by the annular gap. Because this heat transfer mechanism functions with latent heat transfer, the heat pipe is close to isothermal down the axial length. Heat pipes can operate under a wide range of working fluids. Considerations for these working fluids are primarily driven by operating temperatures amongst other important factors based around overall performance. Sodium filled heat pipes operate from 450°C up to 900°C. This temperature range works well for the current microreactor designs. In conjunction with this experimental capability, INL has developed a modeling software to simulate heat pipe physics within reactor cores. This modeling software is called Sockeye and functions under the established INL Multiphysics Object Oriented Simulation Environment (MOOSE). SPHERE also supports Sockeye development by providing the modeling team with experimental data on an array of setups and operating parameters to support validation efforts. A power transient experiment was performed utilizing the SPHERE facility to continue to aid with Sockeye development. The testing followed a proposed test plan to ramp up and down the temperature of the heat pipe. Sockeye models steady state heat pipe operation with high accuracy, the data provided by the power transient testing aims to assist with the validation efforts and further enhance transient modeling capability of the tool [2].

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Development of a water source heat pump hardware-in-the-loop (HIL) testing facility for smart building applications

Over the last decade, the global fight against climate change through electrification has led to an increase in research on building heating, ventilation, and air conditioning (HVAC) systems that utilize intelligent control algorithms to provide demand-side grid service while maintaining the thermal comfort of building occupants. As the pivotalpoint between building electricity consumption and indoor thermal comfort, high-efficiency electrical heatpumps are at the center of these emerging studies, and various grid-interactive and occupant-comfort control algorithms have been developed for them. The impact of these algorithms on the heatpump operation andperformance under different weather, building load, and grid requests calls for investigation and verification via experimental tests with actual heat pumps integrated with real-time building and grid responses. This study presents a Water-Source Heat Pump Hardware-in-The-Loop (HIL) Test Facility developed with the capability to perform such tests. The hardware configuration for this testfacility introduces a hydronic system that emulates the conditions for the heat pump water-side, and a duct system that emulates conditions for the heat pump airside. Both data acquisition and emulator control are implemented through the National Instruments (NI) LabVIEW software running on an NI PXIplatform. The HIL mechanism based on the hardware-software integration that allows the testbed to communicate with a generic simulation environment is also discussed. Currently, the test facility setup includes a single heatpump and virtual building model in EnergyPlus coupled with an occupant behavioral model in MATLAB. Preliminary test results of the current setup demonstrate the building load emulator's ability to track the simulated gone temperature with a Root Mean Square Deviation (RSME) below 0.12°C (0.216°F). An uncertainty analysis based on sensor accuracies shows that the heat pump coefficient of performance (COP) can be measured with a relative uncertainty of 10.4% in cooling and 3.7% in heating. Apartfrom the current testing on a single heat pump, the test facility also provides the flexibility to include additional heat pumps to form a heat pump cluster, as well as coupling the heat pump with active thermal storage to provide enhanced demandflexibility.

Calfa, Caleb↗
DOE OSTI2025

Calculating the Effects of Solids Input and Removal as a Temperature Control in the Advanced Scale Up Reactor Experiment (ASURE) Facility at NETL Using Aspen

The Advanced Scale Up Reactor Experiment (ASURE) facility at NETL is being designed to be a fuel-flexible multi-purpose reactor that can be used for pyrolysis/gasification or evaluation of other high pressure “circulating fluidized bed” (CFB) chemical processes. The initial system design calculations for pyrolysis/gasification are presented in this work showing the expected performance of the ASURE facility when used as a biomass conversion reactor. Several other areas of application include gasification of any carbonaceous fuel including biomass, coal, plastics, and other waste materials. The reactor can therefore be used to produce SYNGAS of various compositions and hydrogen as well as other high value chemicals resulting from a typical tuned gasification process. This paper discusses an ASPEN model of the facility, focusing on the riser of the CFB reactor and the solids recirculation loop. The ASPEN model divides the riser into two sections. A bottom section which receives ash, char and sand which have been recirculated from a return loop. In this section an inert fluidization gas, (N2 or CO2), is introduced which acts as the primary mover of the solids through the system. The bottom section is equipped with a restricted air feed so that the recirculated char can be partially oxidized. This oxidation process along with the inventory of recirculating sand are used to effectively control the temperature in the following two chemical conversion sections of the reactor which are the pyrolysis zone followed by a tar cracking zone. Fresh fuel is added to the pyrolysis zone and undergoes drying and devolatilization. The products ash, char, volatile matter, and water vapor exit the pyrolysis zone and enter the reaction block for tar cracking. Steam and CO2 gasification reactions will be incorporated into the tar cracking zone, however at the design operating temperature, conversion from these reactions is expected to be essentially zero. The unit when completed in 2026 will test mixtures of biomass, plastics, and waste coal. This paper presents the basic ASPEN engineering design model for this project and provides preliminary sensitivity studies to determine how the various parts of the reactor will perform.

ASSURE↗

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