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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Geoelectrical Measurement of Multi-Scale Mass Transfer Parameters

Mass transfer affects contaminant transport and is thought to control the efficiency of aquifer remediation at a number of sites within the Department of Energy (DOE) complex. An improved understanding of mass transfer is critical to meeting the enormous scientific and engineering challenges currently facing DOE. Informed design of site remedies and long-term stewardship of radionuclide-contaminated sites will require new cost-effective laboratory and field techniques to measure the parameters controlling mass transfer spatially and across a range of scales. In this project, we sought to capitalize on the geophysical signatures of mass transfer. Previous numerical modeling and pilot-scale field experiments suggested that mass transfer produces a geoelectrical signature—a hysteretic relation between sampled (mobile-domain) fluid conductivity and bulk (mobile + immobile) conductivity—over a range of scales relevant to aquifer remediation. In this work, we investigated the geoelectrical signature of mass transfer during tracer transport in a series of controlled experiments to determine the operation of controlling parameters, and also investigated the use of complex-resistivity (CR) as a means of quantifying mass transfer parameters in situ without tracer experiments. In an add-on component to our grant, we additionally considered nuclear magnetic resonance (NMR) to help parse mobile from immobile porosities. Including the NMR component, our revised study objectives were to: 1. Develop and demonstrate geophysical approaches to measure mass-transfer parameters spatially and over a range of scales, including the combination of electrical resistivity monitoring, tracer tests, complex resistivity, nuclear magnetic resonance, and materials characterization; and 2. Provide mass-transfer estimates for improved understanding of contaminant fate and transport at DOE sites, such as uranium transport at the Hanford 300 Area. To achieve our objectives, we implemented a 3-part research plan involving (1) development of computer codes and techniques to estimate mass-transfer parameters from time-lapse electrical data; (2) bench-scale experiments on synthetic materials and materials from cores from the Hanford 300 Area; and (3) field demonstration experiments at the DOE’s Hanford 300 Area. In a synergistic add-on to our workplan, we analyzed data from field experiments performed at the DOE Naturita Site under a separate DOE SBR grant, on which PI Day-Lewis served as co-PI. Techniques developed for application to Hanford datasets also were applied to data from Naturita.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR-1 Irradiation Experiment

This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Radionuclide Air Emissions Report for 2012

Berkeley Lab operates facilities where radionuclides are produced, handled, store d, and potentially emitted . These facilities are subject to the EPA radioactive air emission regulations in 40 CFR 61, Subpart H (EPA 1989a). Radionuclides may be emitted from stacks or vents on buildings where radionuclide production or use is authorized or they may be emitted as diffuse sources. In 2012, all Berkeley Lab sources were minor sources of radionuclides (sources resulting in a potential dose of less than 0.1 mrem/yr [0.001 mSv/yr]) . These minor sources include d about 140 stack sources and no diffuse sources . T here were no unplanned airborne radionuclide emissions from Berkeley Lab operations . Emissions from minor sources were measured by sampling or monitoring or were calculated based on quantities used, received for use, or produced during the year. Using measured and calculated emissions, and building- specific and common parameters, Laboratory personnel applied the EPA -approved computer code s, CAP88-PC and COMPLY , to calculate doses to the maximally exposed individual (MEI) at any offsite point where there is a residence, school, business, or office. Because radionuclides are used at three noncontiguous locations (the main site, Berkeley West Bio center, and Joint BioEnergy Institute), three different MEIs were identified.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Argonne National Laboratory Site Environmental Report for Calendar Year 2013

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2013. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP-88 Version 3 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Neutron Activation and Thermoluminescent Detector Responses to a Bare Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 11, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

Miller, Thomas Martin↗
DOE OSTI2015

Implementation and Re nement of a Comprehensive Model for Dense Granular Flows

Dense granular ows are ubiquitous in both natural and industrial processes. They manifest three di erent ow regimes, each exhibiting its own dependence on solids volume fraction, shear rate, and particle-level properties. This research project sought to develop continuum rheological models for dense granular ows that bridges multiple regimes of ow, implement them in open-source platforms for gas-particle ows and perform test simulations. The rst phase of the research covered in this project involved implementation of a steady- shear rheological model that bridges quasi-static, intermediate and inertial regimes of ow into MFIX (Multiphase Flow with Interphase eXchanges - a general purpose computer code developed at the National Energy Technology Laboratory). MFIX simulations of dense granular ows in hourglass-shaped hopper were then performed as test examples. The second phase focused on formulation of a modi ed kinetic theory for frictional particles that can be used over a wider range of particle volume fractions and also apply for dynamic, multi- dimensional ow conditions. To guide this work, simulations of simple shear ows of identical mono-disperse spheres were also performed using the discrete element method. The third phase of this project sought to develop and implement a more rigorous treatment of boundary e ects. Towards this end, simulations of simple shear ows of identical mono-disperse spheres con ned between parallel plates were performed and analyzed to formulate compact wall boundary conditions that can be used for dense frictional ows at at frictional boundaries. The fourth phase explored the role of modest levels of cohesive interactions between particles on the dense phase rheology. The nal phase of this project focused on implementation and testing of the modi ed kinetic theory in MFIX and running bin-discharge simulations as test examples.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Development of Improved Graphical Displays for an Advanced Outage Control Center, Employing Human Factors Principles for Outage Schedule Management

The long-term viability of existing nuclear power plants in the United States (U.S.) is dependent upon a number of factors, including maintaining high capacity factors, maintaining nuclear safety, and reducing operating costs, particularly those associated with refueling outages. Refueling outages typically take 20-30 days, and for existing light water NPPs in the U.S., the reactor cannot be in operation during the outage. Furthermore, given that many NPPs generate between $1-1.5 million/day in revenue when in operation, there is considerable interest in shortening the length of refueling outages. Yet refueling outages are highly complex operations, involving multiple concurrent and dependent activities that are somewhat challenging to coordinate; therefore, finding ways to improve refueling outage performance, while maintaining nuclear safety has proven to be difficult. The Advanced Outage Control Center (AOCC) project is a research and development (R&D) demonstration activity under the LWRS Program. LWRS is an R&D program that works closely with industry R&D programs to establish technical foundations for the licensing and managing of long-term, safe, and economical operation of current fleet of NPPs. As such, the LWRS Advanced Outage Control Center project has the goal of improving the management of commercial NPP refueling outages. To accomplish this goal, INL is developing an advanced outage control center (OCC) that is specifically designed to maximize the usefulness of communication and collaboration technologies for outage coordination and problem resolution activities. The overall focus is on developing an AOCC with the following capabilities that enables plant and OCC staff to; Collaborate in real-time to address emergent issues; Effectively communicate outage status to all workers involved in the outage; Effectively communicate discovered conditions in the field to the OCC; Provide real-time work status; Provide automatic pending support notifications; Provide real-time requirements monitoring; Maximize their collective situational awareness to improve decision-making; and Leverage macro data to better support resource allocation. INL has partnered with several commercial NPP utilities to develop a number of advanced outage management technologies. These outage management technologies have focused on both collaborative technologies for control centers and developing mobile technologies for NPP field workers. This report describes recent efforts made in developing a suite of outage technologies to support more effective schedule management. Currently, a master outage schedule is created months in advance using the plant’s existing scheduling software (e.g., Primavera P6). Typically, during the outage, the latest version of the schedule is printed at the beginning of each shift. INL and its partners are developing technologies that will have capabilities such as Automatic Schedule Updating, Automatic Pending Support Notifications, and the ability to allocate and schedule outage support task resources on a sub-hour basis (e.g., outage Micro-Scheduling). The remaining sections of this report describe in more detail the scheduling challenges that occur during outages, how the outage scheduling technologies INL is developing helps address those challenges, and the latest developments on this task (e.g., work accomplished to date and the path forward)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Large Eddy Simulation Modeling of Flashback and Flame Stabilization in Hydrogen-Rich Gas Turbines Using a Hierarchical Validation Approach

This project was a combined computational and experimental effort to improve predictive capability for boundary layer flashback of premixed swirl flames relevant to gas-turbine power plants operating with high-hydrogen-content fuels. During the course of this project, significant progress in modeling was made on four major fronts: 1) use of direct numerical simulation of turbulent flames to understand the coupling between the flame and the turbulent boundary layer; 2) improved modeling capability for flame propagation in stratified pre-mixtures; 3) improved portability of computer codes using the OpenFOAM platform to facilitate transfer to industry and other researchers; and 4) application of LES to flashback in swirl combustors, and a detailed assessment of its capabilities and limitations for predictive purposes. A major component of the project was an experimental program that focused on developing a rich experimental database of boundary layer flashback in swirl flames. Both methane and high-hydrogen fuels, including effects of elevated pressure (1 to 5 atm), were explored. For this project, a new model swirl combustor was developed. Kilohertz-rate stereoscopic PIV and chemiluminescence imaging were used to investigate the flame propagation dynamics. In addition to the planar measurements, a technique capable of detecting the instantaneous, time-resolved 3D flame front topography was developed and applied successfully to investigate the flow-flame interaction. The UT measurements and legacy data were used in a hierarchical validation approach where flows with increasingly complex physics were used for validation. First component models were validated with DNS and literature data in simplified configurations, and this was followed by validation with the UT 1-atm flashback cases, and then the UT high-pressure flashback cases. The new models and portable code represent a major improvement over what was available before this project was initiated.

42 ENGINEERING↗
DOE OSTI2016

Using Supercomputers to Probe the Early Universe

For decades physicists have been trying to decipher the first moments after the Big Bang. Using very large telescopes, for example, scientists scan the skies and look at how fast galaxies move. Satellites study the relic radiation left from the Big Bang, called the cosmic microwave background radiation. And finally, particle colliders, like the Large Hadron Collider at CERN, allow researchers to smash protons together and analyze the debris left behind by such collisions. Physicists at Los Alamos National Laboratory, however, are taking a different approach: they are using computers. In collaboration with colleagues at University of California San Diego, the Los Alamos researchers developed a computer code, called BURST, that can simulate conditions during the first few minutes of cosmological evolution.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Final Technical Report for Year 5 Early Career Research Project "Viscosity and equation of state of hot and dense QCD matter"

The Section below summarizes research activities and achievements during the fifth (last) year of the PI’s Early Career Research Project (ECRP). Unlike the first four years of the project, the last year was not funded under the American Recovery and Reinvestment Act (ARRA). The ECRP advanced two main areas: i) radiative 3 ↔ 2 radiative transport, via development of a new computer code MPC/Grid that solves the Boltzmann transport equation in full 6+1D (3X+3V+time); and ii) application of relativistic hydrodynamics, via development of a self-consistent framework to convert viscous fluids to particles. In Year 5 we finalized thermalization studies with radiative gg ↔ ggg transport (Sec. 1.1.1) and used nonlinear covariant transport to assess the accuracy of fluid-to-particle conversion models (Sec. 1.1.2), calculated observables with self-consistent fluid-to-particle conversion from realistic viscous hydrodynamic evolution (Secs. 1.2.1 and 1.2.2), extended the covariant energy loss formulation to heavy quarks (Sec. 1.4.1) and studied energy loss in small systems (Sec. 1.4.2), and also investigated how much of the elliptic flow could have non-hydrodynamic origin (Sec 1.3). Years 1-4 of the ECRP were ARRA-funded and, therefore, they have their own report document ’Final Technical Report for Years 1-4 of the Early Career Research Project “Viscosity and equation of state of hot and dense QCD matter”’ (same award number DE-SC0004035). The PI’s group was also part of the DOE JET Topical Collaboration, a multi-institution project that overlapped in time significantly with the ECRP. Purdue achievements as part of the JET Top- ical Collaboration are in a separate report “Final Technical Report summarizing Purdue research activities as part of the DOE JET Topical Collaboration” (award DE-SC0004077).

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Hybrid Enrichment Verification Array: Module Characterization Studies

The work presented in this report is focused on the characterization and refinement of the Hybrid Enrichment Verification Array (HEVA) approach, which combines the traditional 186-keV 235 U signature with high-energy prompt gamma rays from neutron capture in the detector and surrounding collimator material, to determine the relative enrichment and 235 U mass of the cylinder. The design of the HEVA modules (hardware and software) deployed in the current field trial builds on over seven years of study and evolution by PNNL, and consists of a ø3''×3'' NaI(Tl) scintillator coupled to an Osprey digital multi-channel analyzer tube base from Canberra. The core of the HEVA methodology, the high-energy prompt gamma-ray signature, serves as an indirect method for the measurement of total neutron emission from the cylinder. A method for measuring the intrinsic efficiency of this “non-traditional” neutron signature and the results from a benchmark experiment are presented. Also discussed are potential perturbing effects on the non-traditional signature, including short-lived activation of materials in the HEVA module. Modeling and empirical results are presented to demonstrate that such effects are expected to be negligible for the envisioned implementation scenario. In comparison to previous versions, the new design boosts the high-energy prompt gamma-ray signature, provides more flexible and effective collimation, and improves count-rate management via commercially available pulse-processing electronics with a special modification prompted by PNNL.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Passive Safety Features Evaluation of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Advanced Output Coupling for High Power Gyrotrons

The Phase II program developed an internal RF coupler that transforms the whispering gallery RF mode produced in gyrotron cavities to an HE11 waveguide mode propagating in corrugated waveguide. This power is extracted from the vacuum using a broadband, chemical vapor deposited (CVD) diamond, Brewster angle window capable of transmitting more than 1.5 MW CW of RF power over a broad range of frequencies. This coupling system eliminates the Mirror Optical Units now required to externally couple Gaussian output power into corrugated waveguide, significantly reducing system cost and increasing efficiency. The program simulated the performance using a broad range of advanced computer codes to optimize the design. Both a direct coupler and Brewster angle window were built and tested at low and high power. Test results confirmed the performance of both devices and demonstrated they are capable of achieving the required performance for scientific, defense, industrial, and medical applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Fission Meter Information Barrier Attribute Measurement System: Task 1 Report: Document existing Fission Meter neutron IB system

An SNM attribute Information Barrier (IB) system was developed for a 2011 US/UK Exercise. The system was modified and extensively tested in a 2013-2014 US-UK Measurement Campaign. This work demonstrated rapid deployment of an IB system for potential treaty use. The system utilizes an Ortec Fission Meter neutron multiplicity counter and custom computer code. The system demonstrates a proof-of-principle automated Pu-240 mass determination with an information barrier. After a software start command is issued, the system automatically acquires and downloads data, performs an analysis, and displays the results. This system conveys the results of a Pu mass threshold measurements in a way the does not reveal sensitive information. In full IB mode, only red/green ‘lights’ are displayed in the software. In test mode, more detailed information is displayed. The code can also read in, analyze, and display results from previously acquired or simulated data. Because the equipment is commercial-off-the-shelf (COTS), the system demonstrates a low-cost short-lead-time technology for treaty SNM attribute measurements. A deployed system will likely require integration of additional authentication and tamper-indicating technologies. This will be discussed for the project in this and future progress reports.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package were developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Status of the Flooding Fragility Testing Development

This report provides an update on research addressing nuclear power plant component reliability under flooding conditions. The research includes use of the Component Flooding Evaluation Laboratory (CFEL) where individual components and component subassemblies will be tested to failure under various flooding conditions. The resulting component reliability data can then be incorporated with risk simulation strategies to provide a more thorough representation of overall plant risk. The CFEL development strategy consists of four interleaved phases. Phase 1 addresses design and application of CFEL with water rise and water spray capabilities allowing testing of passive and active components including fully electrified components. Phase 2 addresses research into wave generation techniques followed by the design and addition of the wave generation capability to CFEL. Phase 3 addresses methodology development activities including small scale component testing, development of full scale component testing protocol, and simulation techniques including Smoothed Particle Hydrodynamic (SPH) based computer codes. Phase 4 involves full scale component testing including work on full scale component testing in a surrogate CFEL testing apparatus.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Water NSTF Design, Instrumentation, and Test Planning

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released for the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric variations and off-normal configurations. The facility design follows, including as-built dimensions and specifications of the various mechanical and liquid systems, design choices for the test section, water storage tank, and network piping. Specifications of the instrumentation suite are then presented, along with specific information on performance windows, measurement uncertainties, and installation locations. Finally, descriptions of the control systems and heat removal networks are provided, which have been engineered to support precise quantification of energy balances and facilitate well-controlled test operations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗