Airfoil self-noise and prediction

A prediction method is developed for the self-generated noise of an airfoil blade encountering smooth flow. The prediction methods for the individual self-noise mechanisms are semiempirical and are based on previous theoretical studies and data obtained from tests of two- and three-dimensional airfoil blade sections. The self-noise mechanisms are due to specific boundary-layer phenomena, that is, the boundary-layer turbulence passing the trailing edge, separated-boundary-layer and stalled flow over an airfoil, vortex shedding due to laminar boundary layer instabilities, vortex shedding from blunt trailing edges, and the turbulent vortex flow existing near the tip of lifting blades. The predictions are compared successfully with published data from three self-noise studies of different airfoil shapes. An application of the prediction method is reported for a large scale-model helicopter rotor, and the predictions compared well with experimental broadband noise measurements. A computer code of the method is given

Habitat Use by Reintroduced Mountain Quail

Mountain quail (Oreortyx pictus) have declined in much of the Intermountain Region of the western United States. Many areas that once supported these birds now seemingly lack necessary food and cover, especially in critical riparian zones. Additionally, mountain quail appear to need periodic disturbance (fire, moderate grazing, etc.) to provide adequate forage and nesting areas. If mountain quail do not readily occupy suitable habitats, either because of restricted movements or because of habitat discontinuities, it may be necessary to stock birds in order to restore populations. In September 1995, we began a restoration program with the objective of reintroducing mountain quail into former ranges in eastern Oregon and Washington. In the winter of 1996--1997, we released 17 radio-marked birds into a drainage in Hell\u27s Canyon as a pilot study to determine habitat use, survival estimates, and movement patterns. An additional 40 radio-marked birds were released during spring 1998 to determine habitat use, nesting success, and brood survival

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels

Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a particular batch scheme). The volume of test space in IPTs is larger in MATRIX than in ATR with comparable magnitude of neutron flux. In addition to the IPTs, the Cylindrical MATRIX concept features test spaces at the centers of fuel assemblies where very high fast flux can be achieved. This magnitude of fast flux is similar to that achieved in the ATR A-positions, however, the available volume having these conditions is greater in the MATRIX design than in the ATR. From the analyses performed in this work, it appears that the Cylindrical MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this statement must be qualified by acknowledging that this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design matures. Also, some of the requirements were not strictly met, but are believed to be achievable once features to be added later are designed

Dual energy X-ray absorptiometry positioning protocols in assessing body composition: A systematic review of the literature:A systematic review of the literature

OBJECTIVES: To systematically identify and assess methods and protocols used to reduce technical and biological errors in published studies that have investigated reliability of dual energy X-ray absorptiometry (DXA) for assessing body composition. DESIGN: Systematic review. METHODS: Systematic searches of five databases were used to identify studies of DXA reliability. Two independent reviewers used a modified critical appraisal tool to assess their methodological quality. Data was extracted and synthesised using a level of evidence approach. Further analysis was then undertaken of methods used to decrease DXA errors (technical and biological) and so enhance DXA reliability. RESULTS: Twelve studies met eligibility criteria. Four of the articles were deemed high quality. Quality articles considered biological and technical errors when preparing participants for DXA scanning. The Nana positioning protocol was assessed to have a strong level of evidence. The studies providing this evidence indicated very high test–retest reliability (ICC 0.90–1.00 or less than 1% change in mean) of the Nana positioning protocol. The National Health and Nutrition Examination Survey (NHANES) positioning protocol was deemed to have a moderate level of evidence due to lack of high quality studies. However, the available studies found the NHANES positioning protocol had very high test–retest reliability. Evidence is limited and reported reliability has varied in papers where no specific positioning protocol was used or reported. CONCLUSIONS: Due to the strong level of evidence of excellent test–retest reliability that supports use of the Nana positioning protocol, it is recommended as the first choice for clinicians when using DXA to assess body composition

Thermal hydraulic design of a 2400 MW t̳h̳ direct supercritical CO₂-cooled fast reactor

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006."September 2006." In title on t.p.,double-underscored letters "t" and "h" appear as subscript.Includes bibliographical references (p. 229-233).The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled direct cycle plant with core outlet temperatures in the 850"C to 10000C range. Pursued in the interest of high cycle efficiency and the provision of process heat for hydrogen production by thermochemical water cracking, these high temperatures present materials challenges which may prove difficult to overcome in the near future. By taking advantage of the low compressibility of CO2 near its critical point, the supercritical CO2 (S-CO2) recompression cycle can achieve an efficiency of 48% with a relatively low core outlet temperature of 650'C. The 4-loop 2400 MWth direct S-CO2 cooled fast reactor under investigation at MIT is thus a lower-temperature alternative to the mainstream helium cooled GFR design. A steady state core design was developed which utilizes an innovative, high fuel volume fraction, vented Tube-In-Duct (TID) fuel assembly. Through an extensive series of iterative calculations, RELAP5-3D was then used to evaluate the natural circulation performance of an active/passive hybrid Shutdown/Emergency Cooling System (SCS/ECS). Routes were identified by which significant post-LOCA core bypass could occur and degrade the decay heat removal performance. Moderately-sized blowers were shown to be capable of overcoming even extreme core bypass routes. An active SCS/ECS was thus adopted for the reference design.(cont.) The loss of external load (LOEL) event is analyzed and a bypass valve scheme is recommended which prevents shaft overspeed and excessive core coolant mass flow rate. A large dry pressurized water reactor (PWR) containment building having a free volume of 70,000 m3 and a peak design pressure of 6 bar is selected for this design based on a 100 in2 cold duct break. During this same loss of coolant accident (LOCA), the depressurization time is shown to be in excess of 10 minutes. No action need be taken by the SCS/ECS blowers before this time in order to prevent core damage. After this time, a total blower power less than 90 kW is sufficient to cool the core out to 10,000 seconds. A loss of flow (LOF) transient in which a PCS loop is instantaneously isolated and no mitigating action is taken (i.e. no reactor scram) is also shown not to cause core damage. It is concluded that a large S-CO2 cooled GFR coupled to a supercritical Brayton power conversion system can withstand the thermal hydraulic challenges posed by the usual menu of severe accident scenarios.by Michael A. Pope.Ph.D

Reactor physics design of supercritical CO₂-cooled fast reactors

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.Includes bibliographical references (p. 109-113).Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO₂ (S-CO₂) as a Brayton cycle working fluid in a direct cycle is evaluated. By using S- CO₂ at turbine inlet conditions of 20 MPa and 550⁰C - 700⁰C, efficiencies between 45% and 50% can be achieved with extremely compact components. Neutronic evaluation of candidate core materials was performed for potential use in block-type matrix fueled GFRs with particular concentration on lowering coolant void reactivity to less than $1. SiC cercer fuel was found to have relatively low coolant void worth (+22 cents upon complete depressurization of S-CO₂ coolant) and tolerable reactivity- limited burnup at matrix volume fractions of 60$2.00 in ODS MA956-clad case having H/D of 1).(cont.) It was shown, however, that S-CO₂ coolant void reactivity could be lowered significantly - to less than $1 - in pin cores by increasing neutron leakage (e.g. lowering the core H/D ratio to 0.625 in a pin core with ODS MA956 cladding), an effect not observed in cores using helium coolant at 8 MPa and 500⁰C. An innovative "block"-geometry tube-in-duct fuel consisting of canisters of vibrationally compacted (VIPAC) oxide fuel was introduced and some preliminary calculations were performed. A reference tube-in-duct core was shown to exhibit favorable neutron economy with a conversion ratio (CR) at beginning of life (BOL) of 1.37, but had a coolant void reactivity of +$ 1.4. The high CR should allow designers to lower coolant void worth by increasing leakage while preserving the ability of the core to reach high burnup. Titanium, vanadium and scandium were found to be excellent reflector materials from the standpoint of ... and coolant void reactivity due to their unique elastic scattering cross-section profiles: for example, the SiC cercer core having void reactivity of +$0.22 with titanium was shown to have +$0.57 with Zr₃Si₂.(cont.) Overall, present work confirmed that the S-CO₂-cooled GFR concept has promising characteristics and a sufficiently broad opion space such that a safe and competitive design could be developed in future work with considerably less than $1 void reactivity and a controllable [delta]k due to burnup.by Michael A. Pope.S.M

The Domain Name System—Past, Present, and Future

The Domain Name System (DNS) is a critical component of the global Internet infrastructure. Throughout its history, its design and administration has experienced significant dynamic changes as the Internet itself has evolved. The history of the DNS is divided into six eras, based on underlying technological and administrative themes within each era. Developments in its governance, its application, and in other factors are discussed. Future directions for DNS use and abuse are explored, along with challenges in its future governance. Finally, a proposed research model is included to guide future study of the DNS evolution and its influences from political, legal, psychological, sociological, and technological perspectives