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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Nuclear-Renewable Energy Systems Secondary Product Market Analysis Study

In order to properly create a program surrounding the development of any technological concept it is necessary to fully understand the market in which it is being developed. In the case of Integrated Nuclear-Renewable Hybrid Energy Systems (HES), there are two economic markets in which it must be able to participate in: the electricity market and the secondary product market associated with the specific system. The purpose of the present report is to characterize the secondary product market in the U.S. and to provide recommendations for further developing the HES program. While HESs have been discussed in depth in many other reports, it is helpful to discuss them briefly in the present work [REF]. The concept of the HES can be deduced to a system, featuring a combination of a nuclear power plant, a renewable energy source, and an industrial manufacturing plant . The system is designed in a fashion that allows it either to produce electricity or to manufacture a secondary product as needed. The primary benefit of this concept lies in its ability to maximize economic performance of the integrated system and to manufacture products in a carbon-free manner. A secondary benefit is the enhanced supply-side flexibility gained by allowing the HES to economically provide grid services. A key tenant to nuclear power plant economics in today’s electricity market is their ability to operate at a very high capacity factor. Unfortunately, in regions with a high penetration of renewable energy, the carbon free energy produced by nuclear power may not be needed at all times. This forces the nuclear power plant to find a user for its excess capacity. This may include paying the electric grid to find a user, releasing energy to the environment by ‘dumping steam’, or reducing power. If the plant is unable to economically or safely do any of these actions, the plant is at risk of being shutdown. In order to allow for nuclear power plants to continue to contribute carbon free electricity to the grid in a future with high renewable energy penetration, HESs allow for excess capacity to be diverted to a chemical process. If the chemical products sold on the market replace those sold previously – which would be the case if a currently operating manufacturing plant was modified to be a HES component – then the products would now be produced with reduced emission of carbon and other greenhouse gases. There are several key economic barriers that must be surmounted for HESs to be developed. The two primary barriers are the increased capital cost associated with coupling and controlling the HES components and the decreased utilization of the manufacturing plant capital due to intermittent energy delivery . Because of this, manufacturing plants that are less complex and have smaller non-variable operations and capital costs may be more attractive for integration. A secondary economic barrier for the HES is the market availability for its products. The system must operate a region where there is either an intermittent demand for its electricity, an intermittent demand for its secondary product, or both. In a region with an intermittent demand, product prices should shift accordingly, making it less attractive to produce one of the products. The HES then can shift production in order to maximize profit. Without an intermittent demand for at least one of its products, there would be little need for it to expend the extra capital required for integration as an HES.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

A Business Case for Nuclear Plant Control Room Modernization

This paper presents a generic business case for implementation of technology that supports Control Room Modernization (CRM). The analysis presented in two forms; 1) a standalone technology upgrade, and 2) a technology upgrade that is built upon and incremental to a prior business case created for Mobile Work Packages (MWP). The business case contends that advanced communication and networking and analytical technologies will allow NPP to conduct control room operations with improved focus by reducing human factors and redundant manpower, and therefore operate with fewer errors. While some labor savings can be harvested in terms of overtime, the majority of savings are demonstrated as reduced time to take the plant off line and bring back on line in support of outages. The benefits are quantified to a rough order of magnitude that provides directional guidance to NPPs that are interested in developing a similar business case. This business case focuses on modernization of the operator control room and does not consider a complete overhaul and modernization of a plants instrument and control systems. While operators may be considering such an investment at their plants, the sizable capital investment required is not likely supported by a cost/benefit analysis alone. More likely, it is driven by obsolescence and reliability issues, and requires consideration of mechanical condition of plant systems, capital depreciation, financing, relicensing and overall viability of the plant asset over a 20-year horizon in a competitive market. Prior studies [REF] have indicated that such a modernization of plant I&C systems, alone or as part of a larger modernization effort, can yield very significant reductions in O&M costs. However, the depth of research and analysis required to develop a meaningful business case for a plant modernization effort is well beyond the scope of this study. While CRM as considered in this study can be easily integrated as part of grander plant modernization effort, it can also be considered as a stand-alone project, implemented as a supervisorial layer over and above existing systems. CRM is enabled by a suite of technologies, which are described in further detail in the body of this report. They include: • Modernized control room with interactive displays • High-bandwidth wireless networks • Mobile devices • Component identification technology • Mobile wireless video cameras • Smart or task based operator displays • Computer-Based Procedures (CBP)/Automated Mobile Work Packages (MWP) • Intelligent plant configuration • Advanced data analytics An analysis was conducted to determine how these technologies might impact outage operations at an NPP. The analysis concluded that outage management capabilities would be improved with application of these technologies, and result in annual benefits for the plant. These improved capabilities were identified and described in further detail in this report. • Reduction in need redundant manpower (verification, etc) • Improved critical path shutdown and restart duration during outages • Reduction in human factors and errors • Remote watches • Paperless outage coordination • Bulk work optimization

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

High Energy Physics Research with the CMS Experiment at CERN - Energy Frontier Experiment

The Large Hadron Collider (LHC) at the European Organization for Nuclear Research (CERN) near Geneva, Switzerland, is now the highest energy accelerator in the world, colliding protons with protons. On July 4, 2012, the two general-purpose experiments, ATLAS and the Compact Muon Solenoid (CMS) experiment, announced the observation of a particle consistent with the world’s most sought-after particle, the Higgs boson, at a mass of about 125 GeV (approximately 125 times the mass of the proton). The Higgs boson is the final missing ingredient of the standard model, in which it is needed to allow most other particles to acquire mass through the mechanism of electroweak symmetry breaking. We are members of the team in the CMS experiment that found evidence for the Higgs boson through its decay to two photons, the most sensitive channel at the LHC. We are proposing to carry out studies to determine whether the new particle has the properties expected for the standard model Higgs boson or whether it is something else. The new particle can still carry out its role in electroweak symmetry breaking but have other properties as well. Most theorists think that a single standard model Higgs boson cannot be the complete solution – there are other particles needed to answer some of the remaining questions, such as the hierarchy problem. The particle that has been observed could be one of several Higgs bosons, for example, or it could be composite. One model of physics beyond the standard model is supersymmetry, in which every ordinary particle has a superpartner with opposite spin properties. In supersymmetric models, there must be at least five Higgs bosons. In the most popular versions of supersymmetry, the lightest supersymmetric particle does not decay and is a candidate for dark matter. This proposal covers the period from June 1, 2013, to March 31, 2016. During this period the LHC will finally reach its design energy, almost twice the energy at which it now runs. We will be able to study the Higgs boson at the current LHC energy using about three times as much data as were used to make the observation. In 2013 the LHC will shut down to make preparations to run at its design energy in 2015. During the shutdown period, we will be preparing upgrades of the detector to be able to run at the higher rates of proton-proton collisions that will also be possible once the LHC is running at design energy. The upgrade on which we are working, the inner silicon pixel tracker, will be installed in late 2016. Definitive tests of whether the new particle satisfies the properties of the standard model Higgs boson will almost certainly require both the higher energy and the larger amounts of data that can be accumulated using the higher rates. Meanwhile we will use the data taken during 2012 and the higher energy data starting in 2015 to continue to search for beyond-the-standard-model physics such as supersymmetry and heavy neutrinos. We have already made such searches using data since the LHC started running. We are discussing with theorists how a 125-GeV Higgs modifies such models. Finding such particles will probably also require the higher energy and larger amounts of data beginning in 2015. The period of this proposal promises to be very exciting, leading to new knowledge of the matter in the Universe.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2017

Foil Bearing Coating Behavior in CO 2

The Sandia S-CO 2 Recompression Closed Brayton Cycle (RCBC) utilizes a series of gas foil bearings in its turbine-alternator-compressors. At high shaft rotational speed these bearings allow the shaft to ride on a cushion of air. Conversely, during startup and shutdown, the shaft rides along the foil bearing surface. Low-friction coatings are used on bearing surfaces in order to facilitate rotation during these periods. An experimental program was initiated to elucidate the behavior of coated bearing foils in the harsh environments of this system. A test configuration was developed enabling long duration exposure tests, followed by a range of analyses relevant to their performance in a bearing. This report provides a detailed overview of this work. The results contained herein provide valuable information in selecting appropriate coatings for more advanced future bearing-rig tests at the newly established test facility in Sandia-NM.

36 MATERIALS SCIENCE↗
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↗
DOE OSTI2017

Final Technical Report for "Nuclear Technologies for Near Term Fusion Devices"

Over approximately 18 years, this project evolved to focus on a number of related topics, all tied to the nuclear analysis of fusion energy systems. For the earliest years, the University of Wisconsin (UW)’s effort was in support of the Advanced Power Extraction (APEX) study to investigate high power density first wall and blanket systems. A variety of design concepts were studied before this study gave way to a design effort for a US Test Blanket Module (TBM) to be installed in ITER. Simultaneous to this TBM project, nuclear analysis supported the conceptual design of a number of fusion nuclear science facilities that might fill a role in the path to fusion energy. Beginning in approximately 2005, this project added a component focused on the development of novel radiation transport software capability in support of the above nuclear analysis needs. Specifically, a clear need was identified to support neutron and photon transport on the complex geometries associated with Computer-Aided Design (CAD). Following the initial development of the Direct Accelerated Geoemtry Monte Carlo (DAGMC) capability, additional features were added, including unstructured mesh tallies and multi-physics analysis such as the Rigorous 2-Step (R2S) methodology for Shutdown Dose Rate (SDR) prediction. Throughout the project, there were also smaller tasks in support of the fusion materials community and for the testing of changes to the nuclear data that is fundamental to this kind of nuclear analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.

36 MATERIALS SCIENCE↗
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

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Final Project Report on Computational Modeling and Analysis of Air-Based NSTF

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). The air-cooled RCCS uses natural convection to drive the flow of air from outside the power plant to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a ½ - scale model of the primary features of one conceptual air-cooled RCCS design. The NSTF program performs ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Well Integrity for Natural Gas Storage in Depleted Reservoirs and Aquifers

Natural gas storage facilities are a critical component of our energy supply and distribution chain, allowing elasticity in gas supply to accommodate daily to seasonal demand fluctuations. As has been made evident by the recent Aliso Canyon Gas Storage facility incident, a loss of well integrity may result in significant consequences, including the prolonged shutdown of an entire facility. The Aliso Canyon gas well blowout emitted approximately 100,000 tonnes of natural gas (mostly methane) over 4 months and displaced thousands of nearby residents from their homes. The high visibility of the event has led to increased scrutiny of the safety of natural gas storage at the Aliso Canyon facility, led to questions about energy reliability, and raised broader concerns for natural gas storage integrity throughout the country.

03 NATURAL GAS↗
DOE OSTI2018

Visual Inspection of Defense Waste Processing Facility Melter 2 Interior after End of Service

Defense Waste Processing Facility (DWPF) Melter 2 reached its end of service life after nearly14 years. A visual inspection of the interior of the melter was performed using a remotely operated camera to provide insight into melter performance and potential failure mechanisms. The condition of the interior of the melter was generally good considering its relatively long service life. The plenum refractory appeared to be in good condition. Spalling and wear of the glass contact refractory was evident at the melt line. There appeared to be some degree of degradation and deposition of material at the dome heater penetrations. Deformation, or sagging of the dome heaters was visible. Only the top surfaces of the upper electrodes were visible due to the full inventory of glass remaining after melter shutdown. The electrodes exhibited small cracks as well as minor deformation (creep or slumping). The lower electrode failure mechanism could not be determined since it was not visible.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Solid-State Ultra-High Power Amplifier Modules Based on 100 V RF GaN Technology for LINACS Operating at 325 and 650 MHz

Purpose / objectives. This research project explores the groundwork for developing a new high efficiency, high reliability and cost effective radio frequency technology to be used in power sources for future linear accelerators. The frequencies of operation are 650 and 325 MHz, in use at Fermi National Laboratory. The technology can be extended to other frequencies. Brief Description. In this research work ultra-high power solid-state radio frequency sources are proposed operating at the stated frequencies of interest. The approach employs high voltage Gallium Nitride radio frequency transistors, operating at 100 V or higher, to achieve 5-kilowatts continuous-wave and 10 kilowatts pulse power level in a small form factor. The 5 kilowatts building block reduces the required number of combiners and related losses to achieve megawatts power levels. The design employs harmonic tuning techniques as used in class E, F and inverse F power amplifiers to achieve >70% efficiency and assembly techniques that overcome heat dissipation in such high power density systems. Solid-state technology achieves million hour operating lifetime compared to 10 – 100 thousand hours for incumbent technology based on vacuum tubes. Additionally, the modular approach afforded by a solid-state solution avoids complete system shutdown when one module fails. Overall, the approach reduces capital and operating cost. Research Results. During Phase I radio frequency Gallium Nitride building block transistors with 50 mm gate periphery were designed and built which achieve ~600 W continuous-wave when operating at 100 V bias and 1-kW pulsed when operating at 150V. This level of performance has been demonstrated with operation at 325 and 650 MHz, and in both cases measured efficiency exceeded 70%. Potential Applications. The technology identified as suitable for high efficiency power sources in particle accelerators can also be applied in megawatt class radars and several industrial, scientific and medical applications such as microwave sintering, radio frequency and microwave ablation, oil percolation, MRI scanners and Nuclear Magnetic Resonance spectrometers, etc… The high efficiency of these devices also translates into reduced electrical power consumption and lower cost of operation.

100 VDC↗
DOE OSTI2018

Evaluation of flow and heat transfer inside lean pre-mixed combustor systems under reacting flow conditions

The objective of the present work was to develop methodologies to measure heat transfer within a reacting gas turbine combustor. To accomplish this, an optically accessible research combustor system was designed and constructed at Virginia Tech, capable of operating at 650 K inlet temperature, maximum air mass flow rates of 1.3 kg/s, and flame temperatures over 1800 K. Flow and heat transfer measurements at non-reacting and reacting conditions were carried out for Reynolds numbers (Re) with respect to the combustor diameter ranging from ~11 500 to ~140 000 (depending on the condition). Particle Image Velocimetry (PIV) was used to measure the non-reacting flow field within the burner, leading to the identification of coherent structures in the flow that accounted for over 30% of the flow fluctuation kinetic energy along the swirling jet shear layers. The capability of infrared (IR) thermography to image surface temperatures through a fused silica (quartz) glass was demonstrated at non-reacting conditions. IR thermography was then used to measure the non-reacting steady state heat transfer along the combustor liner. A peak in heat transfer was identified at ~1 nozzle diameter downstream of the combustor dome plate. The peak Nusselt number along the liner was over 18 times higher than that predicted from fully developed turbulent pipe flow correlations, which have traditionally been used to estimate flame side combustor heat transfer. For the reacting measurements, a novel time-dependent heat transfer methodology was developed that allowed for the investigation of transient heat loads, including those occurring during engine ignition and shutdown. The methodology was validated at non-reacting conditions, by comparing results from an experiment with changing flow temperature, to the results obtained at steady state. The difference between the time-dependent and the steady state measurements were between 3% and 17.3% for different mass flow conditions. The time-dependent methodology was applied to reacting conditions for combustor Reynolds numbers of ~12 000 and ~24 000. At an equivalence ratio of ~0.5 and a combustor Reynolds number of ~12 000, the peak heat load location in reaction was shifted downstream by 0.2 nozzle diameters compared to the non-reacting cases. At higher equivalence ratios, and more visibly at a Reynolds number of ~24 000, the heat transfer distribution along the combustor liner exhibited two peaks, upstream and downstream of the impingement location (X/DN = 0.8-1.0 and X/DN ≈ 2.5). Reacting PIV was performed at Re = 12 000 showing the presence of a strong corner recirculation, which could potentially convect reactants upstream of the impingement point, leading to the double peak structure observed. The methodologies developed have provided insight into heat transfer within gas turbine combustors. The methods can be used to explore additional conditions and expand the dataset beyond what is presented, to fully characterize reacting combustor heat transfer.

42 ENGINEERING↗
DOE OSTI2018

Baseline Radiological Conditions for the Transient Reactor Test (TREAT) Restart

The Transient Reactor Test (TREAT) facility, located on the Idaho National Laboratory (INL), northwest of the Materials and Fuels Complex (MFC), has been in programmatic shutdown for more than twenty-three years. Construction of TREAT by the Teller Construction Co., Portland, Oregon, was started in February, 1958, and completed in early November, 1958. The reactor first achieved criticality on February 23, 1959. Operations continued until April 28, 1994, when the reactor was shut down due to the lack of a programmatic mission. Recently, the U.S. Department of Energy (DOE) identified a need to resume transient testing of new reactor fuels. The Resumption of Transient Testing Program (RTTP) has readied the TREAT facility to resume transient testing operations. The first reactor criticality since the shut down in 1994 was achieved on November 14, 2017. This report details the radiological measurements taken in the TREAT Reactor Building and the surrounding environment prior to and just after its restart in order to characterize the radiological conditions that existed prior to and during the initial restart operations. The authors would like to acknowledge the support from RTTP Management and the TREAT Operations staff for allowing the extensive radiological surveys, as well as the health physics technicians from the INL Radiological Control organization who performed many of the measurements referenced in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

PRA BASICS FOR REGULATORY APPLICATIONS P-105

This report includes the following: Course Overview •Separated into two sections –Section I - PRA methodology •First 11 modules of Section I •Overview of basic terms and concepts •Designed to provide information about the development of a probabilistic risk assessment model (all of the pieces) •Provide background information on different types of PRAs (external events, low power/ shutdown, Level 2 & 3) –Section II - PRA applications •Last 6 modules of Section II •Risk Informed Regulation and Risk Informed Decision-Making •NRC uses of PRA –Configuration risk management –Significance Determination –Mitigating System Performance Indicator

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Analysis of Loss-of-Offsite-Power Events: 1987-2016

Loss of offsite power (LOOP) can have a major negative impact on a power plant’s ability to achieve and maintain safe shutdown conditions. LOOP event frequencies and times required for subsequent restoration of offsite power are important inputs to plant probabilistic risk assessments. This report presents a statistical and engineering analysis of LOOP frequencies and durations at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience during calendar years 1987 through 2016. LOOP events during critical operation that do not result in a reactor trip are not included. Frequencies and durations were determined for four event categories: plant-centered, switchyard-centered, grid-related, and weather-related. Emergency diesel generator reliability is also considered (failure to start, failure to load and run, and failure to run more than 1 hour). An adverse trend in overall LOOP frequency is identified. An adverse trend in LOOP durations is identified. Plant-centered and weather-related LOOP events do not show statistically significant seasonality. The engineering analysis of LOOP data shows that human errors have been much less frequent since 1997 than in the 1986–1996 time period.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗