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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Study of natural circulation in a VHTR after a LOFA using different turbulence models

Natural convection currents in the core are anticipated in the event of the failure of the gas circulator in a prismatic gas-cooled very high temperature reactor (VHTR). The paths that the helium coolant takes in forming natural circulation loops and the effective heat transport are of interest. The heated flow in the reactor core is turbulent during normal operating conditions and at the beginning of the LOFA with forced convection, but the flow may significantly be slowed down after the event and laminarized with mixed convection. In the present study, the potential occurrence and effective heat transport of natural circulation are demonstrated using computational fluid dynamic (CFD) calculations with different turbulence models as well as laminar flow. Validations and recommendation on turbulence model selection are conducted. The study concludes that large loop natural convection is formed due to the enhanced turbulence levels by the buoyancy effect and the turbulent regime near the interface of upper plenum and flow channels increases the flow resistance for channel flows entering upper plenum and thus less heat can be removed from the core than the prediction by laminar flow assumption.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Fuel performance evaluation of annular metallic fuels for an advanced fast reactor concept

The annular metallic fuel design has been considered as a promising solution to relieve the fuel-cladding mechanical interaction (FCMI) caused by significant swelling of metallic fuels at high burnup. In this paper, the fuel performance of annular metallic fuel with HT9 cladding in a novel advanced fast reactor design optimized for high-burnup performance was evaluated. The BISON code was utilized to perform fuel performance simulations using a modified fuel swelling correlation for conservative evaluation. Three representative fuel pins, which feature highest peak burnup (26.8 % FIMA, fission per initial metal atom), highest fuel temperature (645 degrees C), and highest fast neutron flux (2.68 x 10 15 n/cm 2 ∙s), were investigated. A series of key fuel performance parameters, such as peak fuel temperature, reduction of fuel central void, and cumulative damage fraction (CDF) of the cladding, were predicted for both upper and lower plenum configurations. It is evident that annular fuel design with a lower smeared density is capable of accommodating the significant swelling of metallic fuel at high burnup. Finally, the adoption of a lower plenum configuration effectively lowers the gas pressure inside the rod and reduces the creep damage accumulation in the HT9 cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Porous Flow Modeling of Axial Gas Redistribution in Fragmented LWR Fuel Rods using MOOSE

Understanding how gas axially redistributes within fragmented fuel pellets is crucial for predicting the behavior of Light Water Reactor (LWR) fuel rods, particularly during transient and accidental scenarios. The time scale of this phenomenon plays a fundamental role in determining the progression and hazard of a Loss Of Coolant Accident (LOCA), especially when high burn-up fuel in a severe state of fragmentation is involved. Here, this study presents a Computational Fluid Dynamics (CFD) model developed within the Multiphysics Object-Oriented Simulation Environment (MOOSE) to predict the time-scale of plenum depressurization in Light-Water Reactor (LWR) fuel rods driven by axial gas transport through fragmented pellets. The model examines the effects of incorporating non-linearities in the friction term by comparing the results with experimental data. The experiment employed surrogate fuel rods containing pellets subjected to mechanical and/or thermal loadings to simulate various severity of cracking, and aimed at studying the influence of fuel conditions on axial gas redistribution. The results of this analysis indicate that under certain flow regime conditions - determined by the value of an equivalent Reynolds number - accounting for the non-linear friction term in Navier-Stokes equations guarantees better predictions for the time-scale of plenum depressurization. Also, the model enabled the simulation of the pressure decay by assigning distinct permeability values to each pellet instead of a single uniform value. Multiple simulations were run across all possible pellet position combinations, having each pellet assigned with values of permeability extracted from the experimental data. This allows to quantify the impact of the considering various non-uniform distributions of permeability on the dynamics of axial gas redistribution. The present work findings enhance the understanding of axial gas transport, and provide valuable insights for the integration of a model for predicting the axial gas redistribution during a LOCA scenario into the BISON fuel performance code.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Design and full core fuel performance assessment of high burnup cores for 4-loop PWRs

Increasing the fuel discharge burnup of current light water reactors (LWRs) promises reductions in fuel cycle and/or operations costs. By assuming a constant core power density, the economic gain is enabled by better fuel utilization and/or an increased capacity factor. In this effort to investigate greater than 62 MWd/kgU maximum rod average burnup for 110+ kW/l core power density, two core designs have been developed for a standard 17x17, 193 fuel assemblies pressurized water reactor (PWR). The levelized unit cost methodology is employed to evaluate fuel cycle, operation and maintenance, and capital cost impacts and to examine the economic viability of both core design pathways. Core design and optimization are performed using the commercial STUDSVIK code package. Fuel performance analysis is realized in full core configuration via auditing FRAPCON4.1, FAST1.2, and the high-fidelity code BISON. To provide a realistic assessment, the core design process takes into consideration best practices in current PWR core design. It features acceptable performance in terms of various core design constraints on maximum allowable peaking and boron concentration. Gadolinia (Gd2O3) is used as a burnable poison with a maximum of 9 wt% concentration while feeding 89 or 77 fuel assemblies in a 3-batch refueling scheme. Full core fuel performance simulation, which allows for characterization of relevant fuel temperatures, plenum pressures, stresses, and strains, is performed with respect to two bounding burnup levels. Such performance is potentially licensable for the 18-month high burnup core (<68 MWd/kgU peak pin), while it is more challenging for the 24-month high burnup core design pathway (<75 MWd/kgU peak pin). Maximum rod plenum pressure is identified as the most limiting fuel performance parameter. Here, while the scope of the present study focuses on the steady-state plus overpower conditions, the acceptability of the new discharge burnup has to be further assessed by considering uncertainties and impacts under accident scenarios in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Enhancing the x-ray output of a single-wire explosion with a gas-puff based plasma opening switch

Here, we present experiments performed on the 1 MA COBRA generator using a low density, annular, gas-puff z-pinch implosion as an opening switch to rapidly transfer a current pulse into a single metal wire on axis. This gas-puff on axial wire configuration was studied for its promise as an opening switch and as a means of enhancing the x-ray output of the wire. We demonstrate that current can be switched from the gas-puff plasma into the wire, and that the timing of the switch can be controlled by the gas-puff plenum backing pressure. X-ray detector measurements indicate that for low plenum pressure Kr or Xe shots with a copper wire, this configuration can offer a significant enhancement in the peak intensity and temporal distribution of radiation in the 1–10 keV range.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Study of Triple Ar Gas Puff Z-Pinches on 0.9-MA, 200-ns COBRA

Ar gas puff z-pinch experiments were conducted on the 0.9 MA, 200 ns COBRA generator at Cornell University. In the experiments, a triple-nozzle gas valve was used to produce the zpinch load with concentric outer and inner gas puffs and a central gas jet. We used Planer Laser Induced Fluorescence (PLIF) to measure the initial gas profile of the load, two 4-frame 4-ns gated XUV cameras to image the imploding plasma, and filtered Photo- Conducting Diamond Detectors (PCDs) to measure the x-ray output. We have observed shock heating and/or photoionization of the dense center jet by the imploding plasma. With an outer to inner plenum pressure ratio of 1:3, de-acceleration of the imploding plasma occurred when the outer plasma shell imploded on the inner shell resulting in a stable implosion. By varying the plenum pressure for the center jet, the x-ray emission was optimized. In conclusion, an Ar K-shell x-ray (3-4 keV) yield of 77±17 J was produced at a current level of 0.87±0.2 MA and an implosion time of 200-ns.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Numerical Analysis of Combustion Dynamics in a Full-Scale Rotating Detonation Rocket Engine using Large Eddy Simulations

Large eddy simulations (LES) using detailed chemistry and leveraging adaptive mesh refinement (AMR) are performed to gain insights into the combustion dynamics within a full-scale methane-oxygen non-premixed rotating detonation rocket engine (RDRE) employing impinging discrete injection schemes. In particular, a comparative analysis of two operating conditions corresponding to the same global equivalence ratio but different mass flow rates is carried out to investigate the resultant impact on detonation wave characteristics and RDRE global performance. Multiple co-rotating detonation waves with spatially-distributed wave structure and preferential alignment with the inner wall of the annulus (due to asymmetry in fuel distribution) are encountered under both conditions. Both cases exhibit pre-detonation deflagrative burning in the fill region, while one of the cases shows higher susceptibility to backflow into the feed plenums due to lower plenum pressures. Furthermore, heat release analysis shows that the thrust obtained from the RDRE is closely linked to the distribution of total heat release between detonative and deflagrative combustion. On the other hand, combustion efficiency is associated with the fraction of heat release occurring in fuel-rich versus fuel-lean regions within the RDRE.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2018

Detailed Chemical Kinetics Based Simulation of Detonation-Containing Flows

Reliable and robust simulations of detonations in inhomogeneous and turbulent environments are of direct importance in the design of rotating detonation engines (RDEs). In particular, computational models will be especially useful in designing and optimizing discrete injectors that introduce fuel and air separately into the detonation chamber, but ensure appropriate level of mixing to sustain detonations but minimize backflow of detonation products and pressure waves into the feed plenums. Since the structure of detonations itself is non-ideal, models have to include a detailed description of this reacting flow in order to be predictive in nature. Here, a highly-scalable open source based solver has been developed for complex detonating flows such that a) the detonation processes are described using detailed chemical kinetics, b) the method is computationally efficient through the introduction of adaptive mesh refinement, and c) the solver can handle complex geometries of relevance to RDE design. Grid convergence of key metrics for detonations is evaluated using canonical flows. Further, the importance of the use of detailed chemical kinetics is illustrated by extracting the composition structure behind a two-dimensional detonation front. Finally, simulations of a practical RDE configuration are used to demonstrate the applicability of this solver to analyzing geometries. The simulation captures the general trends of the experiment well. It is found that the detonation occurs under partially-premixed conditions. Propagation of pressure waves to the injection system is observed which could influence flow behavior in the oxidizer plenum.

03 NATURAL GAS↗
DOE OSTI2017

Rod Internal Pressure Distribution and Uncertainty Analysis Using FRAPCON

The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data, and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed that tracks intercycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB 2 ) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rod–specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB 2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rods without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/tonne U is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceeds a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed that is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Identification and Characterization of Thermal Fluid Phenomena Associated with Selected Operating/Accident Scenarios in Modular High Temperature Gas-cooled Reactors

Dominant phenomena are identified and characterized for the loss-of-forced-convection (LOFC) scenario, the depressurized loss-of-forced-convection (DLOFC) scenario, and the moisture ingress scenario with relief valves available scenario. These scenarios are considered for the modular high temperature gas-cooled reactor (MHTGR) baseline design—with both a prismatic core and a pebble-bed core. The objectives of this effort, described in the above paragraph, were achieved by following a methodology identified by the U.S. Nuclear Regulatory Commission (NRC) and recommended in their Regulatory Guide 1.203. The methodology employed is described. For the above scenarios, baseline phenomena identification and ranking tables (PIRTs) generated by a NRC-sponsored effort in 2008 for the LOFC and DLOFC scenarios and a baseline PIRT generated by a Department of Energy-sponsored effort in 2011 by their Next Generation Nuclear Plant (NGNP) Moisture Ingress Assessment Committee were used to identify the dominant phenomena and existing levels-of-knowledge. Ten dominant phenomena (with only medium or low levels-of-knowledge) were identified for the LOFC and DLOFC scenarios including: (1) Inlet plenum stratification and plumes, (2) core coolant flow distribution, (3) radiant heat transfer from the top of the core to the upper vessel head, (4) reactor cavity air circulation and heat transfer, (5) reactor cavity cooling system behavior, (6) decay heat distributions, (7) core effective thermal conductivity, (8) molecular diffusion in air ingress DLOFC, (9) duct exchange flow, (10) phenomena that affect cavity gas composition for air-ingress. Five dominant phenomena, with high levels of knowledge, were identified for the moisture-ingress scenario with relief valves available including: (1) flow through break, (2) moisture level in the primary, (3) transport to moisture monitor, (4) moisture monitor instrument response, and (5) heat removal by shutdown cooling system. These phenomena are discussed and characterized. A listing of experiments—both completed and ongoing—is given with a view toward including data useful for further more detailed examination of the dominant phenomena listed in the previous paragraph. Note that only data for the LOFC and DLOFC scenarios are listed—since the moisture ingress scenario with relief valves available –have been identified as having adequate levels-of-knowledge by the PIRT committee. Finally, a preliminary computational fluid dynamics (CFD) calculation of inlet plenum stratification and plumes is described. Techniques employed for this calculation are typical for preliminary scoping calculations performed to validate numeric tools.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Thermally Effective and Efficient Cooling Technologies for Advanced Gas Turbine Systems

This report presents the research which has been undertaken under DOE grant DE-FE-0011875 as part of NETL’s UTSR (University Turbine System Research) Program. The experimental portion of this work was undertaken at the University of North Dakota (UND) under the direction of Dr. Forrest Ames. The computational work was undertaken at LSU (Louisiana State University), UM (University of Memphis), and at IIT (Illinois Institute of Technology) under the direction of Dr. Sumanta Acharya. The research program was organized into a four phase program which included: 1) development of a FDA/FEA (finite difference analysis/finite element analysis) cooling model, 2) experimental/computational investigation of appropriate internal cooling methods, 3) experimental/computational investigation of external heat transfer and film cooling, 4) development of an enhanced FDA/FEA model along with the development of a warm cascade test for comparison purposes. The experimental work at UND included research on both internal cooling methods and external heat transfer. The internal cooling work included the design and testing of multiple variable hole-size “incremental impingement” configurations along with testing of two converging rounded diamond pedestal arrays for the trailing edge region. The external heat transfer work included the building of a large scale cascade facility to test a vane design with a large leading edge and an aft loaded suction surface. UND has acquired both midspan and full suction surface heat transfer measurements as well as prepared for film cooling measurements under the grant. LSU/UM/IIT has developed and conducted wall resolved LES computations for the incremental impingement geometries and provided rational understanding of the effect of hole sizes and their distributions. Additionally they have provided design correlations for the incremental impingement heat transfer coefficients. They have also developed LES methods to predict flow through film cooling plenums, which have pedestals along with the development of slot film cooling downstream from the plenums. Additionally, they have developed wall resolved LES to predict vane surface heat transfer including bypass transition without the use of a transition model. Through these simulations they provide understanding of the aerodynamic and surface heat transfer behavior under a range of influences such as freestream turbulence. The present study has produced some very useful research outcomes.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2018

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Fluid Stratification Separate Effects Analysis, Testing and Benchmarking

High Temperature Gas Reactors (HTGRs) are positioned to disrupt local and global markets via their unique ability to produce carbon-free process heat, high efficiency power generation, and passively safe operational features. However, significant impediments still exist to delay deployment of this particular technology, including a lack of experimental data, verified code application, and lack of consensus with regards to severe accident progression. In particular, air ingress accidents represent a particular challenge to designers and engineers, as they represent low probability, but highly complex, accident scenarios. Including phenomena such as molecular diffusion, free convection, and complex heat and mass transfer paths, experimental and traceable data is essential to maturing the state of the industry. Therefore, this work presents an experimental investigation of the transition to natural convection in HTGR applications using the Stratified Flow Separate Effects Test Facility, housed at Oregon State University. In particular, this work will present data that challenges the assumption that molecular diffusion is a significant factor in this severe accident in the reference facility of the General Atomic 600 MW th Gas Turbine-Modular Helium Reactor (GT-MHR). Rather, experimentally produced data shows a statistically suggestive retardant effect of cross duct orientation, indicating that ingress mechanics may be fundamentally altered via facility geometry. Experimentally, this is achieved using a simplified cross duct that may be positioned in one of two ways so as to provide either horizontal or vertical access to the lower plenum area. Onset of natural convection (ONC) is measured using an oxygen sensor probe, immersed in the helium working fluid, so as to provide direct indication of air presence in the upper plenum. This document will present the doctoral dissertation produced by this work, along with its analysis, experimental data, observations, and conclusions, in addition to select documents that will highlight critical changes during the development process. Those documents will include the quality assurance plan, all subsequent change requests, and the engineering transfer from Harris Thermal in order to establish the most complete quality record possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

MCNP calculations concerning the UW 137 Cs release event (Final Report)

This document supersedes earlier reports (LA-UR-19-32358 and LA-UR-20-22216) that discussed MCNP-calculated conversion factors (count rate per unit activity per unit area) for collimated 2”x2” NaI detectors (Ludlum model 44-10 and Thermo model SPA3) in contact with duct and plenum walls. Specifically, for duct and plenum dimensions and construction details of interest to the University of Washington ¹³⁷Cs release event. This document also corrects some of the data presented in the earlier LA-UR reports.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Room Return Path Sensitivity

Standard practice for HVAC return design has evolved from running a dedicated return to each room with a supply, to systems with returns in more centrally located areas of the home with jump ducts, transfer grilles, or simply door undercuts used as return air pathways from isolated rooms. In these latter systems, hallways and stairwells act as large open ducts for conveying air back to a single (or sometimes multiple) central return. When partition doors to an isolated room are closed, the return airflow resistance goes up, significantly impacting airflow balance if an adequate relief pathway is not installed. Airflow imbalances can lead to comfort and building durability issues and increased envelope leakage. To combat this, some jurisdictions have requirements for return air pathways. The supply system topology–or layout–impacts the airflow balance stability in response to adjustments of return pathway resistances. Branching supply topologies typically have reduced static pressure after each split. The static pressure at the final split will be lower than the primary supply plenum. Because of this, if there is a restriction in a room’s return path, supply airflow will tend to redistribute to adjacent ducts at the end of the branch. A properly designed trunk and branch supply system can effectively equalize static pressure by reducing the cross sectional area after each takeoff . Maintaining static pressure within the trunk will reduce the system’s sensitivity to changes in return paths. However, in practice, it is difficult to design and time consuming to install a complex supply plenum. Velocity effects and poor takeoff placement also impact airflow balance. This report considers return systems in three main categories: distributed, with a return duct to each room; multiple central, with one return grille on each floor of a home; and single central, with a single return grille located near the air handling unit. Three supply categories are also considered: radial splitter box, trunk and branch, and home-run with all ducts connecting directly to a central manifold.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗