Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernander, O.; Haga, I.; Segerberg, F.

BS>From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Although the present status of the boiling water reactor is one of proven technology, design refinements and technical innovations are still being made to further improve reliability, economy and safety. The new standard ASEA- ATOM BWR features a number of such refinements and design improvements involving main circulation punips, containment design, refuelling system and off-gas treatment plant. In some respects the nuclear and hydraulic design of the ASEA- ATOM BWR differs from that adopted by other BWR manufacturers. Since the Oskarshamn I plant was the first nuclear power station havingmore » these features an extensive physics and hydraulics test program was made during the reactor start- up. The results of these tests have fully confirmed the ability of calculation methods to predict the behavior of the reactor. (auth)« less

Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morikawa, Yoshitake

1995-03-01

The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolantmore » system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.« less

Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id

2014-09-30

Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tightmore » concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.« less

Recent developments in BWR fuel design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, S.P.; Noble, L.D.; Wood, J.E.

1991-11-01

Substantial increases in the cost effectiveness and performance capability of boiling water reactor (BWR) fuel designs have been implemented in the past 5 to 7 yr. This increase has been driven by (a) utility desires to lower fuel and operating costs and (b) design innovations that have lowered enrichment requirements, improved thermal-hydraulic performance, and increased discharge exposure. Higher discharge exposures reduce disposal costs for European and Asian utilities and enable US utilities to lengthen operating cycles. A typical BWR reload fuel bundle fabricated today has 25% higher {sup 235}U enrichment and a factor of 2 higher gadolinium loading than onemore » made several years ago. Today's BWR fuel bundles also contain more unheated water reduces the axial water density variation, lowers the void coefficient, and enhances the neutron efficiency of the bundle, reducing both the gadolinium poison and the enrichment requirements. In addition to these general trends, the following unique design innovations have further enhanced the fuel cost efficiency and performance characteristics of BWR fuel: ferrule spacer, part length rods, interactive channel, and bundle enhanced spectral shift. GE's fuel designs offer the flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility and fuel cycle economics.« less

Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arai, Kenji; Ebata, Shigeo

1997-07-01

This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding ofmore » the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.« less

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technicalmore » basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in various locations and at varying degrees during BWR operation based on the core loading pattern. When present during depletion, control blades harden the neutron spectrum locally because they displace the moderator and absorb thermal neutrons. The investigation of the effect of control blades on post operational cask reactivity is documented herein, as is the effect of multiple (continuous and intermittent) exposure periods with control blades inserted. The coupled effects of control blade presence on power density, void profile, or burnup profile will be addressed in future work.« less

High Fidelity BWR Fuel Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Su Jong

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fractionmore » and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.« less

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozlowski, T.; Downar, T.; Xu, Y.

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validationmore » for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)« less

Numerical Simulation of the Emergency Condenser of the SWR-1000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

Radiation field control at the latest BWR plants -- design principle, operational experience and future subjects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchida, Shunsuke; Ohsumi, Katsumi; Takashima, Yoshie

1995-03-01

Improvements of operational procedures to control water chemistry, e.g., nickel/iron control, as well as application of hardware improvements for reducing radioactive corrosion products resulted in an extremely low occupational exposure of less than 0.5 man.Sv/yr without any serious impact on the radwaste system, for BWR plants involved in the Japanese Improvement and Standardization Program. Recently, {sup 60}C radioactively in the reactor water has been increasing due to less crud fixation on the two smooth surfaces of new type high performance fuels and to the pH drop caused by chromium oxide anions released from stainless steel structures and pipings. This increasemore » must be limited by changes in water chemistry, e.g., applications of modified nickel/iron ratio control and weak alkali control. Controlled water chemistry to optimize three points, the plant radiation level and integrities of fuel and structural materials, is the primary future subject for BWR water chemistry.« less

TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

NASA Astrophysics Data System (ADS)

Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

2018-04-01

FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational datamore » available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.« less

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of latticemore » design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.« less

BWR Steam Dryer Alternating Stress Assessment Procedures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morante, R. J.; Hambric, S. A.; Ziada, S.

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

Radiation chemistry related to nuclear power technology

NASA Astrophysics Data System (ADS)

Ishigure, Kenkichi

A brief review is given to the radiation chemical problems, especially with the emphasis on water radiolysis, in the nuclear power technology. Radiation chemistry in aqueous system is pointed out to be closely related to the problems such as corrosion of Zircaloy, the formation of insoluble corrosion products or crud, stress corrosion cracking of stainless steel in BWR and the radioactive waste managements. The results of the constant extention rate tests on sensitized 304 stainless steel under irradiation are shown, and the computer calculations were carried out to simulate the model experiments on the release of crud from the corroding surface under irradiation and also the water radiolysis in core of BWR.

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R J; Gottula, R C; Holcomb, E E

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

Boiling-Water Reactor internals aging degradation study. Phase 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

NASA Astrophysics Data System (ADS)

Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun; Heuser, Brent J.; Mandapaka, Kiran K.; Was, Gary S.

2016-03-01

Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

Report on the BWR owners group radiation protection/ALARA Committee

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aldrich, L.R.

1995-03-01

Radiation protection programs at U.S. boiling water reactor (BWR) stations have evolved during the 1980s and early 1990s from a regulatory adherence-based endeavor to a proactive, risk-based radiation protection and prevention mission. The objectives are no longer to merely monitor and document exposure to radiation and radioactive materials. The focus of the current programs is the optimization of radiation protection of occupational workers consistent with the purpose of producing cost-effective electric power. The newly revised 10 CFR 20 defines the term ALARA (as low as reasonably achievable) to take into account the state of technology, the economics of improvements inmore » relation to the state of the technology, and the benefits to the public health and safety. The BWR Owners Group (BWROG) initially formed the Radiation Protection/ALARA Committee in January 1990 to evaluate methods of reducing occupational radiation exposure during refueling outages. Currently, twenty U.S. BWR owner/operators (representing 36 of the operational 37 domestic BWR units), as well as three foreign BWR operators (associate members), have broadened the scope to promote information exchange between BWR radiation protection professionals and develop good practices which will affect optimization of their radiation protection programs. In search of excellence and the challenge of becoming {open_quotes}World Class{close_quotes} performers in radiation protection, the BWROG Radiation Protection/ALARA Committee has recently accepted a role in assisting the member utilities in improving radiation protection performance in a cost-effective manner. This paper will summarize the recent activities of this Committee undertaken to execute their role of exchanging information in pursuit of optimizing the improvement of their collective radiation protection performance.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J

2016-01-01

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blademore » histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.« less

Recent developments in chemical decontamination technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, C.J.

1995-03-01

Chemical decontamination of parts of reactor coolant systems is a mature technology, used routinely in many BWR plants, but less frequently in PWRs. This paper reviews recent developments in the technology - corrosion minimization, waste processing and full system decontamination, including the fuel. Earlier work was described in an extensive review published in 1990.

Computational Analysis of Splash Occurring in the Deposition Process in Annular-Mist Flow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xie, Heng; Koshizuka, Seiichi; Oka, Yoshiaki

2004-07-01

The deposition process of a single droplet on the film is numerically simulated by the Moving Particle Semi-implicit (MPS) method to analyze the possibility and effect of splash occurring in the deposition process in BWR condition. The model accounts for the presence of inertial, gravitation, viscous and surface tension and is validated by comparison with experiment results. A simple one-dimensional mixture model is developed to calculate the necessary parameters for the simulation of deposition in BWR condition. The deposition process of a single droplet in BWR condition is simulated. The effect of impact angle of droplet and the velocity ofmore » liquid film are analyzed. A film buffer model is developed to fit the simulation results of critical value for splash. A correlation of critical Weber number for splash in BWR condition is obtained and used to analyze the effect of splash. It is found that the splash play important role in the deposition and re-entrainment process in high quality condition in BWR. The mass fraction of re-entrainment caused by splash in different quality condition is also calculated. (authors)« less

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

PubMed

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

Recent GE BWR fuel experience and design evolution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, J.E.; Potts, G.A.; Proebstle, R.A.

1992-01-01

Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su Weinian; Wang, S.-J.; Chiang, S.-C

2005-06-15

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code ismore » chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Majed, M.; Morback, G.; Wiman, P.

1995-09-01

The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give largemore » advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.« less

Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

2015-08-01

Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramicmore » microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, FeCrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.« less

Stability analysis of BWR nuclear-coupled thermal-hyraulics using a simple model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karve, A.A.; Rizwan-uddin; Dorning, J.J.

1995-09-01

A simple mathematical model is developed to describe the dynamics of the nuclear-coupled thermal-hydraulics in a boiling water reactor (BWR) core. The model, which incorporates the essential features of neutron kinetics, and single-phase and two-phase thermal-hydraulics, leads to simple dynamical system comprised of a set of nonlinear ordinary differential equations (ODEs). The stability boundary is determined and plotted in the inlet-subcooling-number (enthalpy)/external-reactivity operating parameter plane. The eigenvalues of the Jacobian matrix of the dynamical system also are calculated at various steady-states (fixed points); the results are consistent with those of the direct stability analysis and indicate that a Hopf bifurcationmore » occurs as the stability boundary in the operating parameter plane is crossed. Numerical simulations of the time-dependent, nonlinear ODEs are carried out for selected points in the operating parameter plane to obtain the actual damped and growing oscillations in the neutron number density, the channel inlet flow velocity, and the other phase variables. These indicate that the Hopf bifurcation is subcritical, hence, density wave oscillations with growing amplitude could result from a finite perturbation of the system even where the steady-state is stable. The power-flow map, frequently used by reactor operators during start-up and shut-down operation of a BWR, is mapped to the inlet-subcooling-number/neutron-density (operating-parameter/phase-variable) plane, and then related to the stability boundaries for different fixed inlet velocities corresponding to selected points on the flow-control line. The stability boundaries for different fixed inlet subcooling numbers corresponding to those selected points, are plotted in the neutron-density/inlet-velocity phase variable plane and then the points on the flow-control line are related to their respective stability boundaries in this plane.« less

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.Y.; Saha, P.

1985-01-01

The Anticipated Transient Without Scram (ATWS) is known to be a dominant accident sequence for possible core melt in a Boiling Water Reactor (BWR). A recent Probabilistic Risk Assessment (PRA) analysis for the Browns Ferry nuclear power plant indicates that ATWS is the second most dominant transient for core melt in BWR/4 with Mark I containment. The most dominant sequence being the failure of long term decay heat removal function of the Residual Heat Removal (RHR) system. Of all the various ATWS scenarios, the Main Steam Isolation Valve (MSIV) closure ATWS sequence was chosen for present analysis because of itsmore » relatively high frequency of occurrence and its challenge to the residual heat removal system and containment integrity. The objective of this paper is to discuss four MSIV closure ATWS calculations using the RAMONA-3B code. The paper is a summary of a report being prepared for the USNRC Severe Accident Sequence Analysis (SASA) program which should be referred to for details. 10 refs., 20 figs., 3 tabs.« less

ALARA Council: Sharing our resources and experiences to reduce doses at Commonwealth Edison Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rescek, F.

1995-03-01

Commonwealth Edison Company is an investor-owned utility company supplying electricity to over three million customers (eight million people) in Chicago and northern Illinois, USA. The company operates 16 generating stations which have the capacity to produce 22,522 megawatts of electricity. Six of these generating stations, containing 12 nuclear units, supply 51% of this capacity. The 12 nuclear units are comprised of four General Electric boiling water (BWR-3) reactors, two General Electric BWR-5 reactors, and six Westinghouse four-loop pressurized water reactors (PWR). In August 1993, Commonwealth Edison created an ALARA Council with the responsibility to provide leadership and guidance that resultsmore » in an effective ALARA Culture within the Nuclear Operations Division. Unlike its predecessor, the Corporate ALARA Committee, the ALARA Council is designed to bring together senior managers from the six nuclear stations and corporate to create a collaborative effort to reduce occupational doses at Commonwealth Edison`s stations.« less

Effects of Thermal Aging on Material Properties, Stress Corrosion Cracking, and Fracture Toughness of AISI 316L Weld Metal

NASA Astrophysics Data System (ADS)

Lucas, Timothy; Forsström, Antti; Saukkonen, Tapio; Ballinger, Ronald; Hänninen, Hannu

2016-08-01

Thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels, as well as duplex, and high-Cr ferritic stainless steels. Spinodal decomposition is largely responsible for the well-known "748 K (475 °C) embrittlement" that results in drastic reductions in ductility and toughness in these materials. This process is also operative in welds of either cast or wrought stainless steels where δ-ferrite is present. While the embrittlement can occur after several hundred hours of aging at 748 K (475 °C), the process is also operative at lower temperatures, at the 561 K (288 °C) operating temperature of a boiling water reactor (BWR), for example, where ductility reductions have been observed after several tens of thousands of hours of exposure. An experimental program was carried out in order to understand how spinodal decomposition may affect changes in material properties in Type 316L BWR piping weld metals. The study included material characterization, nanoindentation hardness, double-loop electrochemical potentiokinetic reactivation (DL-EPR), Charpy-V, tensile, SCC crack growth, and in situ fracture toughness testing as a function of δ-ferrite content, aging time, and temperature. SCC crack growth rates of Type 316L stainless steel weld metal under simulated BWR conditions showed an approximate 2 times increase in crack growth rate over that of the unaged as-welded material. In situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40 pct over the corresponding at-temperature air-tested values. Material characterization results suggest that spinodal decomposition is responsible for the degradation of material properties measured in air, and that degradation of the in situ properties may be a result of hydrogen absorbed during exposure to the high-temperature water environment.

Effects of experimental floods on riparian and aquatic ecosystems: Bill Williams River, Arizona

NASA Astrophysics Data System (ADS)

Shafroth, P. B.; Andersen, D. C.; Wilcox, A. C.; Kui, L.; Stella, J. C.

2013-12-01

Development of flow prescriptions for environmental purposes along rivers is relatively common, but implementation of these 'environmental flows' occurs infrequently. Implementation is critical for testing hypotheses relating flow regime to biotic response, which ultimately can inform adaptive flow management. We describe the development of flow prescriptions and evaluate responses of riparian vegetation, beaver dams, and associated aquatic habitat to experimental floods and intervening base flows associated with an environmental flow program on the Bill Williams River (BWR), in semiarid Arizona. First, we assessed effects of flow releases between 1993 and 2009 designed to favor the establishment and maintenance of native riparian trees (Populus and Salix) and disfavor an invasive, nonnative shrub (Tamarix spp.) downstream of Alamo Dam on the BWR. Our data are multi-scaled and include a several-decade assessment of changes to major vegetation types based on a time series of aerial photography, an assessment of species composition and abundance sampled in permanent vegetation quadrats, and targeted seedling surveys following experimental floods. Between 1993 and 2009, we observed significant increases in Populus and Salix forests and essentially no change in Tamarix. Experimental floods in 2006 and 2007 resulted in higher mortality of Tamarix seedlings than Salix. These results illustrate the potential for managing streamflow to influence riparian vegetation dynamics, including management of nonnative species. Second, we examined the role of beaver as ecosystem engineers in the BWR and linkages to flow releases between 2004 and 2013. Beaver convert lotic stream habitat to lentic through dam construction and maintenance during low flow periods, and the process is reversed when a flood or other event causes dam failure. We estimated the extent of lotic and beaver-created lentic (beaver pond) habitat along the BWR and related the likelihood of damage or destruction of beaver dams to the magnitude and duration of experimental floods. We obtained counts of beaver dams at various times from aerial photographs, aerial videography, and ground surveys. The ratio of lotic to lentic stream length was approximately 6 times greater following a large flood versus a 7 year period with no significant flood releases. Floods of different magnitudes and durations resulted in notably different levels of damage or removal of beaver dams. Finally, we sampled woody vegetation adjacent to the channel to estimate the effect of beaver herbivory, and noted high levels of mature tree mortality in one of our study reaches. Results of our previous and ongoing investigations are reported to land and water managers as part of an adaptive streamflow management process.

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi

2016-06-01

We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

DOE PAGES

Maljovec, D.; Liu, S.; Wang, B.; ...

2015-07-14

Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

Experimental and analytical study of stability characteristics of natural circulation boiling water reactors during startup transient

NASA Astrophysics Data System (ADS)

Woo, Kyoungsuk

Two-phase natural circulation loops are unstable at low pressure operating conditions. New reactor design relying on natural circulation for both normal and abnormal core cooling is susceptible to different types of flow instabilities. In contrast to forced circulation boiling water reactor (BWR), natural circulation BWR is started up without recirculation pumps. The tall chimney placed on the top of the core makes the system susceptible to flashing during low pressure start-up. In addition, the considerable saturation temperature variation may induce complicated dynamic behavior driven by thermal non-equilibrium between the liquid and steam. The thermal-hydraulic problems in two-phase natural circulation systems at low pressure and low power conditions are investigated through experimental methods. Fuel heat conduction, neutron kinetics, flow kinematics, energetics and dynamics that govern the flow behavior at low pressure, are formulated. A dimensionless analysis is introduced to obtain governing dimensionless groups which are groundwork of the system scaling. Based on the robust scaling method and start-up procedures of a typical natural circulation BWR, the simulation strategies for the transient with and without void reactivity feedback is developed. Three different heat-up rates are applied to the transient simulations to study characteristics of the stability during the start-up. Reducing heat-up rate leads to increase in the period of flashing-induced density wave oscillation and decrease in the system pressurization rate. However, reducing the heat-up rate is unable to completely prevent flashing-induced oscillations. Five characteristic regions of stability are discovered at low pressure conditions. They are stable single-phase, flashing near the separator, intermittent oscillation, sinusoidal oscillation and low subcooling stable regions. Stability maps were acquired for system pressures ranging 100 kPa to 400 kPa. According to experimental investigation, the flow becomes stable below a certain heat flux regardless of the inlet subcooling at the core and system pressure. At higher heat flux, unstable phenomena were indentified within a certain range of inlet subcooling. The unstable region diminishes as the system pressure increases. In natural circulation BWRs, the significant gravitational pressure drop over the tall chimney section induces a Type-I instability. The Type-I instability becomes especially important during low power and pressure conditions during reactor start-up. Under these circumstances the effect of pressure variations on the saturation enthalpy becomes significant. An experimental study shows that the flashing phenomenon in the adiabatic chimney section is dominant during the start-up of a natural circulation BWR. Since flashing occurs outside the core, nuclear feedback effects on the stability are small. Furthermore, the thermal-hydraulic oscillation period is much longer than power fluctuation period caused by void reactivity feedback. In the natural circulation system increasing the inlet restriction reduces the natural circulation flow rate, shifting the unstable region to higher inlet subcooling.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walston, S; Rowland, M; Campbell, K

It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting inmore » a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.« less

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

Investigation of two and three parameter equations of state for cryogenic fluids

NASA Technical Reports Server (NTRS)

Jenkins, Susan L.; Majumdar, Alok K.; Hendricks, Robert C.

1990-01-01

Two-phase flows are a common occurrence in cryogenic engines and an accurate evaluation of the heat-transfer coefficient in two-phase flow is of significant importance in their analysis and design. The thermodynamic equation of state plays a key role in calculating the heat transfer coefficient which is a function of thermodynamic and thermophysical properties. An investigation has been performed to study the performance of two- and three-parameter equations of state to calculate the compressibility factor of cryogenic fluids along the saturation loci. The two-parameter equations considered here are van der Waals and Redlich-Kwong equations of state. The three-parameter equation represented here is the generalized Benedict-Webb-Rubin (BWR) equation of Lee and Kesler. Results have been compared with the modified BWR equation of Bender and the extended BWR equations of Stewart. Seven cryogenic fluids have been tested; oxygen, hydrogen, helium, nitrogen, argon, neon, and air. The performance of the generalized BWR equation is poor for hydrogen and helium. The van der Waals equation is found to be inaccurate for air near the critical point. For helium, all three equations of state become inaccurate near the critical point.

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Real time monitoring of environmental crack growth in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, D.; Diehl, C.G.

1988-01-01

A comprehensive field test program was recently completed at several Boiling Water Reactors (BWR) to quantify the effect of coolant impurities on the initiation and growth of stress corrosion cracks. A new technology was utilized which allows for real time monitoring of stress corrosion crack growth rates. The BWR environments were characterized using Ion Chromatography and Electro Chemical Potential (ECP) measurements. The effects of typical water chemistry transients and startups were quantified.

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

NASA Astrophysics Data System (ADS)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAPS model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The 6MCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. (Abstract shortened by UMI.) (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Advanced Neutronics Tools for BWR Design Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Hfaiedh, N.; Letellier, R.

2006-07-01

This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007more » BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

Experimental Study of Two Phase Flow Behavior Past BWR Spacer Grids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ratnayake, Ruwan K.; Hochreiter, L.E.; Ivanov, K.N.

2002-07-01

Performance of best estimate codes used in the nuclear industry can be significantly improved by reducing the empiricism embedded in their constitutive models. Spacer grids have been found to have an important impact on the maximum allowable Critical Heat Flux within the fuel assembly of a nuclear reactor core. Therefore, incorporation of suitable spacer grids models can improve the critical heat flux prediction capability of best estimate codes. Realistic modeling of entrainment behavior of spacer grids requires understanding the different mechanisms that are involved. Since visual information pertaining to the entrainment behavior of spacer grids cannot possibly be obtained frommore » operating nuclear reactors, experiments have to be designed and conducted for this specific purpose. Most of the spacer grid experiments available in literature have been designed in view of obtaining quantitative data for the purpose of developing or modifying empirical formulations for heat transfer, critical heat flux or pressure drop. Very few experiments have been designed to provide fundamental information which can be used to understand spacer grid effects and phenomena involved in two phase flow. Air-water experiments were conducted to obtain visual information on the two-phase flow behavior both upstream and downstream of Boiling Water Reactor (BWR) spacer grids. The test section was designed and constructed using prototypic dimensions such as the channel cross-section, rod diameter and other spacer grid configurations of a typical BWR fuel assembly. The test section models the flow behavior in two adjacent sub channels in the BWR core. A portion of a prototypic BWR spacer grid accounting for two adjacent channels was used with industrial mild steel rods for the purpose of representing the channel internals. Symmetry was preserved in this practice, so that the channel walls could effectively be considered as the channel boundaries. Thin films were established on the rod surfaces by injecting water through a set of perforations at the bottom ends of the rods, ensuring that the flow upstream of the bottom-most spacer grid is predominantly annular. The flow conditions were regulated such that they represent typical BWR operating conditions. Photographs taken during experiments show that the film entrainment increases significantly at the spacer grids, since the points of contact between the rods and the grids result in a peeling off of large portions of the liquid film from the rod surfaces. Decreasing the water flow resulted in eventual drying out, beginning at positions immediately upstream of the spacer grids. (authors)« less

JPRS Report Science & Technology Japan

DTIC Science & Technology

1989-10-25

Testing Slated for New BWR Fuel Assemblies [GENSHIRYOKU SANGYO SHIMBUN, 25 May 89] .... 37 Nuclear Fuel Planning System Developed [GENSHIRYOKU... Development (Debt) 13,272 ((Debt) 3,839) 7,995 (3,610) In addition, the budget has guaranteed that the following programs will proceed according... develop a combined cycle engine that will be capable of attaining high reliability and good fuel consumption at a wide range of speeds from low speed to

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Development of ECT/UT inspection system for bottom mounted instrumentation nozzle of PWR reactor vessels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tanaka, H.; Fukui, S.; Iwahashi, Y.

1994-12-31

The development of inspection technique and tool for Bottom Mounted Instrument (BMI) nozzle of PWR plant was performed for countermeasure of leakage accident at incore instrument nozzle of Hamaoka-1 (BWR). MHI achieved the following development, of which object was PWR Plant R/V: (1) development of ECT/UT Multi-sensored Probe; (2) development of Inspection System (3) development of Data Processing System. The Inspection System had been functionally tested using full scale mock-up. As the result of the functional test, this system was confirmed to be very effective, and assumed to be hopeful for the actual application on site.

Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Knerr, R.; Shoop, U.

1993-01-01

RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

QUAD+ BWR Fuel Assembly demonstration program at Browns Ferry plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doshi, P.K.; Mayhue, L.T.; Robert, J.T.

1984-04-01

The QUAD+ fuel assembly is an improved BWR fuel assembly designed and manufactured by Westinghouse Electric Corporation. The design features a water cross separating four fuel minibundles in an integral channel. A demonstration program for this fuel design is planned for late 1984 in cycle 6 of Browns Ferry 2, a TVA plant. Objectives for the design of the QUAD+ demonstration assemblies are compatibility in performance and transparency in safety analysis with the feed fuel. These objectives are met. Inspections of the QUAD+ demonstration assemblies are planned at each refueling outage.

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides 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. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« less

Divers muscle Fitzpatrick`s mussels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hobbs, B.; Kahabka, J.

1996-01-01

This article describes how an effective manual cleaning technique has helped rid submerged intake structures of this passive-aggressive pest. The New York Power Authority`s (NYPA) James A. Fitzpatrick (JAF) Nuclear Power Plant is located in Lycoming, NY, on the southeast shore of Lake Ontario. An 850-MWe, GE-design boiling water reactor (BWR), the unit has been in service since 1975. Water drawn from the Lake supplies cooling to plant loads via the circulating water system and three service water systems. These share a common intake system consisting of an offshore cap (crib), horizontal intake tunnel, two vertical risers and forebay/screenwell area.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Qualification of APOLLO2 BWR calculation scheme on the BASALA mock-up

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaglio-Gaudard, C.; Santamarina, A.; Sargeni, A.

2006-07-01

A new neutronic APOLLO2/MOC/SHEM/CEA2005 calculation scheme for BWR applications has been developed by the French 'Commissariat a l'Energie Atomique'. This scheme is based on the latest calculation methodology (accurate mutual and self-shielding formalism, MOC treatment of the transport equation) and the recent JEFF3.1 nuclear data library. This paper presents the experimental validation of this new calculation scheme on the BASALA BWR mock-up The BASALA programme is devoted to the measurements of the physical parameters of high moderation 100% MOX BWR cores, in hot and cold conditions. The experimental validation of the calculation scheme deals with core reactivity, fission rate maps,more » reactivity worth of void and absorbers (cruciform control blades and Gd pins), as well as temperature coefficient. Results of the analysis using APOLLO2/MOC/SHEM/CEA2005 show an overestimation of the core reactivity by 600 pcm for BASALA-Hot and 750 pcm for BASALA-Cold. Reactivity worth of gadolinium poison pins and hafnium or B{sub 4}C control blades are predicted by APOLLO2 calculation within 2% accuracy. Furthermore, the radial power map is well predicted for every core configuration, including Void configuration and Hf / B{sub 4}C configurations: fission rates in the central assembly are calculated within the {+-}2% experimental uncertainty for the reference cores. The C/E bias on the isothermal Moderator Temperature Coefficient, using the CEA2005 library based on JEFF3.1 file, amounts to -1.7{+-}03 pcm/ deg. C on the range 10 deg. C-80 deg. C. (authors)« less

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena Arkadievna; Hardin, Ernest

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assumemore » that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.« less

Load Variation Influences on Joint Work During Squat Exercise in Reduced Gravity

NASA Technical Reports Server (NTRS)

DeWitt, John K.; Fincke, Renita S.; Logan, Rachel L.; Guilliams, Mark E.; Ploutz-Snyder, Lori L.

2011-01-01

Resistance exercises that load the axial skeleton, such as the parallel squat, are incorporated as a critical component of a space exercise program designed to maximize the stimuli for bone remodeling and muscle loading. Astronauts on the International Space Station perform regular resistance exercise using the Advanced Resistive Exercise Device (ARED). Squat exercises on Earth entail moving a portion of the body weight plus the added bar load, whereas in microgravity the body weight is 0, so all load must be applied via the bar. Crewmembers exercising in microgravity currently add approx.70% of their body weight to the bar load as compensation for the absence of the body weight. This level of body weight replacement (BWR) was determined by crewmember feedback and personal experience without any quantitative data. The purpose of this evaluation was to utilize computational simulation to determine the appropriate level of BWR in microgravity necessary to replicate lower extremity joint work during squat exercise in normal gravity based on joint work. We hypothesized that joint work would be positively related to BWR load.

Phenomenology of BWR fuel assembly degradation

NASA Astrophysics Data System (ADS)

Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

2018-03-01

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

Status update of the BWR cask simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel G.

2015-09-01

The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximummore » thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on heat load and the effect of simulated wind on a simplified below ground vent configuration.« less

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant.

PubMed

Ródenas, J; Abarca, A; Gallardo, S

2011-08-01

BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool. 2010 Elsevier Ltd. All rights reserved.

Interpretation of the results of the CORA-33 dry core BWR test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.; Hagen, S.

All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ``dry`` core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ``dry`` core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions ofmore » a ``dry`` BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ``dry`` core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed.« less

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Approach to numerical safety guidelines based on a core melt criterion. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azarm, M.A.; Hall, R.E.

1982-01-01

A plausible approach is proposed for translating a single level criterion to a set of numerical guidelines. The criterion for core melt probability is used to set numerical guidelines for various core melt sequences, systems and component unavailabilities. These guidelines can be used as a means for making decisions regarding the necessity for replacing a component or improving part of a safety system. This approach is applied to estimate a set of numerical guidelines for various sequences of core melts that are analyzed in Reactor Safety Study for the Peach Bottom Nuclear Power Plant.

The Japanese utilities` expectations for subchannel analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toba, Akio; Omoto, Akira

1995-12-01

Boiling water reactor (BWR) utilities in Japan began to consider the development of a mechanistic model to describe the critical heat transfer conditions in the BWR fuel subchannel. Such a mechanistic model will not only decrease the necessity of tests, but will also help by removing some overly conservative safety margins in thermal hydraulics. With the use of a postdryout heat transfer correlation, new acceptance criteria may be applicable to evaluate the fuel integrity. Mechanistic subchannel analysis models will certainly back up this approach. This model will also be applicable to the analysis of large-size fuel bundles and examination ofmore » corrosion behavior.« less

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

DOE PAGES

L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

2016-06-06

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

An efficient modeling method for thermal stratification simulation in a BWR suppression pool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haihua Zhao; Ling Zou; Hongbin Zhang

2012-09-01

The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safetymore » analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.« less

KERENA safety concept in the context of the Fukushima accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

Development of crawler type device using new measuring system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maruyama, T.; Sasaki, T.; Yagi, T.

1995-08-01

This paper reports the development and field application of a new device which examine shell to shell weld joints of RPV. In a BWR type nuclear power plant, there is narrow space around the Reactor Pressure Vessel (RPV) because RPV is enclosed by the Reactor Shield Wall (RSW) and thermal insulations. The developed device is characterized by a new position measuring system and magnet wheels for driving. The new position measuring system uses laser beam and ultrasonic wave. The magnet wheels make the device travel freely in the narrow space between RPV and insulation. This device is tested on mock-upsmore » and applied examination of RPVs to verify field applicability.« less

Application of reliability-centered-maintenance to BWR ECCS motor operator valve performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Choi, Y.A.

1993-01-01

This paper describes the application of reliability-centered maintenance (RCM) methods to plant probabilistic risk assessment (PRA) and safety analyses for four boiling water reactor emergency core cooling systems (ECCSs): (1) high-pressure coolant injection (HPCI); (2) reactor core isolation cooling (RCIC); (3) residual heat removal (RHR); and (4) core spray systems. Reliability-centered maintenance is a system function-based technique for improving a preventive maintenance program that is applied on a component basis. Those components that truly affect plant function are identified, and maintenance tasks are focused on preventing their failures. The RCM evaluation establishes the relevant criteria that preserve system function somore » that an RCM-focused approach can be flexible and dynamic.« less

Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J; Williams, Mark L; Wiarda, Dorothea

2015-01-01

Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/B-VII.1 included a more complete set of neutron cross-section covariance data: these data form the basis for a new cross-section covariance library to be released in SCALE 6.2. A range of testing is conducted to investigate the properties of these covariance data and ensure that the data are reasonable. These tests include examination of the uncertainty in critical experiment benchmark model k eff values due to nuclear data uncertainties, asmore » well as similarity assessments of irradiated pressurized water reactor (PWR) and boiling water reactor (BWR) fuel with suites of critical experiments. The contents of the new covariance library, the testing performed, and the behavior of the new covariance data are described in this paper. The neutron cross-section covariances can be combined with a sensitivity data file generated using the TSUNAMI suite of codes within SCALE to determine the uncertainty in system k eff caused by nuclear data uncertainties. The Verified, Archived Library of Inputs and Data (VALID) maintained at Oak Ridge National Laboratory (ORNL) contains over 400 critical experiment benchmark models, and sensitivity data are generated for each of these models. The nuclear data uncertainty in k eff is generated for each experiment, and the resulting uncertainties are tabulated and compared to the differences in measured and calculated results. The magnitude of the uncertainty for categories of nuclides (such as actinides, fission products, and structural materials) is calculated for irradiated PWR and BWR fuel to quantify the effect of covariance library changes between the SCALE 6.1 and 6.2 libraries. One of the primary applications of sensitivity/uncertainty methods within SCALE is the assessment of similarities between benchmark experiments and safety applications. This is described by a c k value for each experiment with each application. Several studies have analyzed typical c k values for a range of critical experiments compared with hypothetical irradiated fuel applications. The c k value is sensitive to the cross-section covariance data because the contribution of each nuclide is influenced by its uncertainty; large uncertainties indicate more likely bias sources and are thus given more weight. Changes in c k values resulting from different covariance data can be used to examine and assess underlying data changes. These comparisons are performed for PWR and BWR fuel in storage and transportation systems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Espinosa-Paredes, Gilberto; Prieto-Guerrero, Alfonso; Nunez-Carrera, Alejandro

This paper introduces a wavelet-based method to analyze instability events in a boiling water reactor (BWR) during transient phenomena. The methodology to analyze BWR signals includes the following: (a) the short-time Fourier transform (STFT) analysis, (b) decomposition using the continuous wavelet transform (CWT), and (c) application of multiresolution analysis (MRA) using discrete wavelet transform (DWT). STFT analysis permits the study, in time, of the spectral content of analyzed signals. The CWT provides information about ruptures, discontinuities, and fractal behavior. To detect these important features in the signal, a mother wavelet has to be chosen and applied at several scales tomore » obtain optimum results. MRA allows fast implementation of the DWT. Features like important frequencies, discontinuities, and transients can be detected with analysis at different levels of detail coefficients. The STFT was used to provide a comparison between a classic method and the wavelet-based method. The damping ratio, which is an important stability parameter, was calculated as a function of time. The transient behavior can be detected by analyzing the maximum contained in detail coefficients at different levels in the signal decomposition. This method allows analysis of both stationary signals and highly nonstationary signals in the timescale plane. This methodology has been tested with the benchmark power instability event of Laguna Verde nuclear power plant (NPP) Unit 1, which is a BWR-5 NPP.« less

BWR Servicing and Refueling Improvement Program: Phase I summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perry, D.R.

1978-09-01

Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was tomore » identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.« less

A Pumping Algorithm for Ergodic Stochastic Mean Payoff Games with Perfect Information

NASA Astrophysics Data System (ADS)

Boros, Endre; Elbassioni, Khaled; Gurvich, Vladimir; Makino, Kazuhisa

In this paper, we consider two-person zero-sum stochastic mean payoff games with perfect information, or BWR-games, given by a digraph G = (V = V B ∪ V W ∪ V R , E), with local rewards r: E to { R}, and three types of vertices: black V B , white V W , and random V R . The game is played by two players, White and Black: When the play is at a white (black) vertex v, White (Black) selects an outgoing arc (v,u). When the play is at a random vertex v, a vertex u is picked with the given probability p(v,u). In all cases, Black pays White the value r(v,u). The play continues forever, and White aims to maximize (Black aims to minimize) the limiting mean (that is, average) payoff. It was recently shown in [7] that BWR-games are polynomially equivalent with the classical Gillette games, which include many well-known subclasses, such as cyclic games, simple stochastic games (SSG's), stochastic parity games, and Markov decision processes. In this paper, we give a new algorithm for solving BWR-games in the ergodic case, that is when the optimal values do not depend on the initial position. Our algorithm solves a BWR-game by reducing it, using a potential transformation, to a canonical form in which the optimal strategies of both players and the value for every initial position are obvious, since a locally optimal move in it is optimal in the whole game. We show that this algorithm is pseudo-polynomial when the number of random nodes is constant. We also provide an almost matching lower bound on its running time, and show that this bound holds for a wider class of algorithms. Let us add that the general (non-ergodic) case is at least as hard as SSG's, for which no pseudo-polynomial algorithm is known.

Sediment Budgeting in Dam-Affected Rivers: Assessing the Influence of Damming, Tributaries, and Alluvial Valley Sediment Storage on Sediment Regimes

NASA Astrophysics Data System (ADS)

Wilcox, A. C.; Dekker, F. J.; Riebe, C. S.

2014-12-01

Although sediment supply is recognized as a fundamental driver of fluvial processes, measuring how dams affect sediment regimes and incorporating such knowledge into management strategies remains challenging. To determine the influences of damming, tributary supply, and valley morphology and sediment storage on downstream sediment supply in a dryland river, the Bill Williams River (BWR) in western Arizona, we measured basin erosion rates using cosmogenic nuclide analysis of beryllium-10 (10Be) at sites upstream and downstream of a dam along the BWR, as well as from tributaries downstream of the dam. Riverbed sediment mixing calculations were used to test if the dam, which blocks sediment supply from the upper 85% of the basin's drainage area, increases the proportion of tributary sediment to residual upstream sediment in mainstem samples downstream of the dam. Erosion rates in the BWR watershed are more than twice as large in the upper catchment (136 t km-2 yr-1) than in tributaries downstream of Alamo Dam (61 t km-2 yr-1). Tributaries downstream of the dam have little influence on mainstem sediment dynamics. The effect of the dam on reducing sediment supply is limited, however, because of the presence of large alluvial valleys along the mainstem BWR downstream of the dam that store substantial sediment and mitigate supply reductions from the upper watershed. These inferences, from our 10Be derived erosion rates and mixing calculations, are consistent with field observations of downstream changes in bed material size, which suggest that sediment-deficit conditions are restricted to a 10 km reach downstream of the dam, and limited reservoir bathymetry data. Many studies have suggested that tributary sediment inputs downstream of dams play a key role in mitigating dam-induced sediment deficits, but here we show that in a dryland river with ephemeral tributaries, sediment stored in alluvial valleys can also play a key role and in some cases trumps the role of tributaries.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hznnera, K.; Hetzler, F.; Hyden, L.

From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Some features of ASEA-ATOM's BWR fuel design and fabrication processes are given. The in pile fuel performance experience to date is reviewed. (auth)

Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2002-10-15

A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies suchmore » as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

3D modeling of missing pellet surface defects in BWR fuel

DOE PAGES

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.; ...

2016-07-26

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

NASA Astrophysics Data System (ADS)

Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, Larry J.; Howell, Michael; Robb, Kevin R.

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 characteristicsmore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.

The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

Level-2 IPE for the Laguna Verde NPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arellano, J.; De Loera, M.A.; Rea, R.

1996-12-31

In response to generic letter GL 88-20, Comision Federal de Electricidad and Instituto de Investigaciones Electricas have jointly developed the individual plant examination (IPE) for the Laguna Verde nuclear power station unit I (LVNPS). This plant is a 675-MW(electric) boiling water reactor (BWR/5) with a reinforced concrete Mark-II containment. The approach used to fulfill the IPE requirements was to make a level-1 probabilistic risk assessment (IPE level 1) plus a containment performance analysis including the behavior and release of the fission products to the environment (IPE level 2). This paper describes the level-2 portion of the LVNPS IPE, paying specialmore » attention to both some improvements to the traditional analytical methods and to the main results.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable ofmore » propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.« less

Characterization of 14C in Swedish light water reactors.

PubMed

Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

2008-08-01

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

A study of the effect of space-dependent neutronics on stochastically-induced bifurcations in BWR dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Analytis, G.T.

1995-09-01

A non-linear one-group space-dependent neutronic model for a finite one-dimensional core is coupled with a simple BWR feed-back model. In agreement with results obtained by the authors who originally developed the point-kinetics version of this model, we shall show numerically that stochastic reactivity excitations may result in limit-cycles and eventually in a chaotic behaviour, depending on the magnitude of the feed-back coefficient K. In the framework of this simple space-dependent model, the effect of the non-linearities on the different spatial harmonics is studied and the importance of the space-dependent effects is exemplified and assessed in terms of the importance ofmore » the higher harmonics. It is shown that under certain conditions, when the limit-cycle-type develop, the neutron spectra may exhibit strong space-dependent effects.« less

Fatigue crack growth in SA508-CL2 steel in a high temperature, high purity water environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerber, T.L.; Heald, J.D.; Kiss, E.

1974-10-01

Fatigue crack growth tests were conducted with 1 in. plate specimens of SA508-CL 2 steel in room temperature air, 550$sup 0$F air and in a 550$sup 0$F, high purity, water environment. Zero-tension load controlled tests were run at cyclic frequencies as low as 0.037 CPM. Results show that growth rates in the simulated Boiling Water Reactor (BWR) water environment are faster than growth rates observed in 550$sup 0$F air and these rates are faster than the room temperature rate. In the BWR water environment, lowering the cyclic frequency from 0.37 to 0.037 CPM caused only a slight increase in themore » fatigue crack growth rate. All growth rates measured in these tests were below the upper bound design curve presented in Section XI of the ASME Code. (auth)« less

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toman, G.; Gazdzinski, R.; O`Hearn, E.

1994-02-01

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specificmore » aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.« less

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandiamore » National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.« less

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, Douglas M.; Nesbitt, Loyd B.

1997-01-01

A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

Enhancing BWR proliferation resistance fuel with minor actinides

NASA Astrophysics Data System (ADS)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

RELAP-7 Development Updates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

75 FR 21046 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on May 6-8, 2010, 11545 Rockville Pike, Rockville....: Boiling Water Reactor (BWR) Owners Group (BWROG) Topical Report NEDC-33347P, ``Containment Overpressure...

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Singh, Gyanender P.; Gorton, Jacob

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperaturesmore » on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Reactor shutdown experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

An overview of zinc addition for BWR dose rate control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marble, W.J.

1995-03-01

This paper presents an overview of the BWRs employing feedwater zinc addition to reduce primary system dose rates. It identifies which BWRs are using zinc addition and reviews the mechanical injection and passive addition hardware currently being employed. The impact that zinc has on plant chemistry, including the factor of two to four reduction in reactor water Co-60 concentrations, is discussed. Dose rate results, showing the benefits of implementing zinc on either fresh piping surfaces or on pipes with existing films are reviewed. The advantages of using zinc that is isotopically enhanced by the depletion of the Zn-64 precursor tomore » Zn-65 are identified.« less

Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Weber, C.F.; Hodge, S.A.

1984-01-01

This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris ontomore » the drywell floor.« less

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

The startup of the Dodewaard natural circulation boiling water reactor -- Experiences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nissen, W.H.M.; Van Der Voet, J.; Karuza, J.

1994-07-01

Because of its similarity to the simplified boiling water reactor (SBWR), the Dodewaard natural circulation boiling water reactor (BWR) is of special interest to further development of the SBWR design. It has become especially important to gain more insight into the Dodewaard BWR behavior during startup, paying special attention to its stability. Therefore, special instrumentation was used by means of which a series of measurements were taken during the two startups in February and June 1992. The results obtained from these measurements are used to deepen insight into the recirculation flow and the stability of the reactor during startup undermore » conditions with a normal pressure/power trajectory. They have already shown a very early recirculation flow onset during low-power operation and no indication of reactor instability. Furthermore, they will be used as a basis for the research program investigating the reactor behavior under different pressure/power conditions, which is scheduled for next year.« less

Removal of Sb-125 and Tc-99 from Liquid Radwaste by Novel Adsorbents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harjula, R.O.; Koivula, R.; Paajanen, A.

2006-07-01

Novel proprietary metal oxide materials (MOM) have been tested for the removal of Sb-125 from simulated Floor Drain Waters of BWR. Antimony was present in the solutions as oxidized anionic form. Long term column experiment with simulated liquid that showed high Sb-125 removal at least up to 8000 bed volumes. One column experiments was carried out using nonradioactive Sb to exhaust the column. Leaching tests with 1000 ppm boric acid showed that 100 % of absorbed Sb remains in the sorbent material. Column experiments with real Fuel Pond Water from Olkiluoto NPP (BWR) showed reduction of Sb-125 (feed level 400more » Bq/L, 1.10{sup -5} {mu}Ci/mL) below detection limit (MDA = 1.7 Bq/L, 5.10{sup -8},{mu}Ci/mL). Additional experiments have also been carried out with pertechnetate (Tc-99) ions. Results indicate that MOM materials are efficient also for the removal of Tc-99 from concentrated NaNO{sub 3} solution. (authors)« less

BWR zero pressure containment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dillmann, C.W.; Townsend, H.E.; Nesbitt, L.B.

1992-02-25

This patent describes the operation of a nuclear reactor system, the system including a containment defining a drywall space wherein a nuclear reactor is disposed, there being a suppression pool in the containment with the suppression pool having a wetwell space above a level of the pool to which an non-condensable gases entering the suppression pool can vent. It comprises: continuously exhausting the wetwell space to remove gas mixture therefrom while admitting inflow of air from an atmospheric source thereof to the wetwell during normal operation by blocking off the inflow during a loss-of-coolant-accident whenever a pressure in the wetwellmore » space is above a predetermined value, and subjecting the gas subsequent to its removal from the wetwell to a treatment operation to separate any particulate material entrained therein from the gas mixture.« less

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J.; Marshall, William B. J.; Ilas, Germina

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidancemore » in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.« less

Trace Assessment for BWR ATWS Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

2010-04-22

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1991-05-21

This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

Posttest Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hessheimer, M.F.; Ludwigsen, J.S.

A high pressure test of a scale model of a steel containment vessel (SCV) was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. This testis part of a program to investigate the response of representative models of nuclear containment structures to pressure loads beyond the design basis accident. The posttest analyses of this test focused on three areas where the pretest analysis effort did not adequately predict the model behavior duringmore » the test. These areas are the onset of global yielding, the strain concentrations around the equipment hatch and the strain concentrations that led to a small tear near a weld relief opening that was not modeled in the pretest analysis.« less

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, Timothy A.; Liao, Huafei

2014-12-01

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constitutedmore » a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Development and Assessment of CTF for Pin-resolved BWR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

DOEpatents

Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

1980-06-06

An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, D.M.; Nesbitt, L.B.

1997-01-21

A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

77 FR 69507 - Proposed Model Safety Evaluation for Plant-Specific Adoption of Technical Specifications Task...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-19

..., ``Revise Shutdown Margin Definition To Address Advanced Fuel Designs'' AGENCY: Nuclear Regulatory... Shutdown Margin Definition to Address Advanced Fuel Designs.'' DATES: Comment period expires on December 19... address newer BWR fuel designs, which may be more reactive at shutdown temperatures above 68[emsp14][deg]F...

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Geist, William H.; Root, Margaret A.

The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

Development of new UV-I. I. Cerenkov Viewing Device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

Application of the IBERDROLA RETRAN Licensing Methodology to the Confrentes BWR-6 110% Extended Power Uprate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fuente, Rafael de la; Iglesias, Javier; Sedano, Pablo G.

IBERDROLA (Spanish utility) and IBERDROLA INGENIERIA (engineering branch) have been developing during the last 2 yr the 110% Extended Power Uprate Project for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved in advance by the Spanish Nuclear Regulatory Authority. This methodology has been applied to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 and 13 and to develop a significant number of safety analyses of the Cofrentes Extended Power.Because the scope of the licensing process of the Cofrentes Extended Power Uprate exceeds the range of analysis includedmore » in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients. This is the case of the total loss of feedwater (TLFW) transient.The content of this paper shows the benefits of having an in-house design and licensing methodology and describes the process to extend the applicability of the Cofrentes RETRAN model to the analysis of new transients, particularly in this paper the TLFW transient.« less

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bovalini, R.; D`Auria, F.; Galassi, G.M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool ofmore » ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions.« less

Systematic void fraction studies with RELAP5, FRANCESCA and HECHAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stosic, Z.; Preusche, G.

1996-08-01

In enhancing the scope of standard thermal-hydraulic codes applications beyond its capabilities, i.e. coupling with a one and/or three-dimensional kinetics core model, the void fraction, transferred from thermal-hydraulics to the core model, plays a determining role in normal operating range and high core flow, as the generated heat and axial power profiles are direct functions of void distribution in the core. Hence, it is very important to know if the void quality models in the programs which have to be coupled are compatible to allow the interactive exchange of data which are based on these constitutive void-quality relations. The presentedmore » void fraction study is performed in order to give the basis for the conclusion whether a transient core simulation using the RELAP5 void fractions can calculate the axial power shapes adequately. Because of that, the void fractions calculated with RELAP5 are compared with those calculated by BWR safety code for licensing--FRANCESCA and the best estimate model for pre- and post-dryout calculation in BWR heated channel--HECHAN. In addition, a comparison with standard experimental void-quality benchmark tube data is performed for the HECHAN code.« less

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

BWR station blackout: A RISMC analysis using RAVEN and RELAP5-3D

DOE PAGES

Mandelli, D.; Smith, C.; Riley, T.; ...

2016-01-01

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates and improved operations. In order to evaluate the impact of these factors on the safety of the plant, the Risk-Informed Safety Margin Characterization (RISMC) project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions and accident scenarios. This paper presents a case study in order to show the capabilities of the RISMC methodology to assess impact of power uprate of a Boiling Watermore » Reactor system during a Station Black-Out accident scenario. We employ a system simulator code, RELAP5-3D, coupled with RAVEN which perform the stochastic analysis. Furthermore, our analysis is performed by: 1) sampling values from a set of parameters from the uncertainty space of interest, 2) simulating the system behavior for that specific set of parameter values and 3) analyzing the outcomes from the set of simulation runs.« less

Optimum Water Chemistry in radiation field buildup control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Chien, C.

1995-03-01

Nuclear utilities continue to face the challenGE of reducing exposure of plant maintenance personnel. GE Nuclear Energy has developed the concept of Optimum Water Chemistry (OWC) to reduce the radiation field buildup and minimize the radioactive waste production. It is believed that reduction of radioactive sources and improvement of the water chemistry quality should significantly reduce both the radiation exposure and radwaste production. The most important source of radioactivity is cobalt and replacement of cobalt containing alloy in the core region as well as in the entire primary system is considered the first priority to achieve the goal of lowmore » exposure and minimized waste production. A plant specific computerized cobalt transport model has been developed to evaluate various options in a BWR system under specific conditions. Reduction of iron input and maintaining low ionic impurities in the coolant have been identified as two major tasks for operators. Addition of depleted zinc is a proven technique to reduce Co-60 in reactor water and on out-of-core piping surfaces. The effect of HWC on Co-60 transport in the primary system will also be discussed.« less

M3FT-16OR020202112 - Report on viability of hydrothermal corrosion resistant SiC/SiC Joint development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Koyanagi, Takaaki; Kiggans Jr, James O.

2016-06-30

Hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under PWR and BWR relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti-Si-C system, and SiC nanopowder sintering. Most of the formed joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession rate of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing activity environments. The SiC nanopowder sinteredmore » joint was the most corrosion tolerant under the oxidizing activity environment among the four joints.« less

10 CFR 50.75 - Reporting and recordkeeping for decommissioning planning.

Code of Federal Regulations, 2010 CFR

2010-01-01

... investing or otherwise, that a prudent investor would use in the same circumstances. The term “prudent... than or equal to 3400 MWt $105 between 1200 MWt and 3400 MWt (For a PWR of less than 1200 MWt, use P... 3400 MWt (For a BWR of less than 1200 MWt, use P=1200 MWt) $(104+0.009P) (2) An adjustment factor at...

Comparison of new generation low-complexity flood inundation mapping tools with a hydrodynamic model

NASA Astrophysics Data System (ADS)

Afshari, Shahab; Tavakoly, Ahmad A.; Rajib, Mohammad Adnan; Zheng, Xing; Follum, Michael L.; Omranian, Ehsan; Fekete, Balázs M.

2018-01-01

The objective of this study is to compare two new generation low-complexity tools, AutoRoute and Height Above the Nearest Drainage (HAND), with a two-dimensional hydrodynamic model (Hydrologic Engineering Center-River Analysis System, HEC-RAS 2D). The assessment was conducted on two hydrologically different and geographically distant test-cases in the United States, including the 16,900 km2 Cedar River (CR) watershed in Iowa and a 62 km2 domain along the Black Warrior River (BWR) in Alabama. For BWR, twelve different configurations were set up for each of the models, including four different terrain setups (e.g. with and without channel bathymetry and a levee), and three flooding conditions representing moderate to extreme hazards at 10-, 100-, and 500-year return periods. For the CR watershed, models were compared with a simplistic terrain setup (without bathymetry and any form of hydraulic controls) and one flooding condition (100-year return period). Input streamflow forcing data representing these hypothetical events were constructed by applying a new fusion approach on National Water Model outputs. Simulated inundation extent and depth from AutoRoute, HAND, and HEC-RAS 2D were compared with one another and with the corresponding FEMA reference estimates. Irrespective of the configurations, the low-complexity models were able to produce inundation extents similar to HEC-RAS 2D, with AutoRoute showing slightly higher accuracy than the HAND model. Among four terrain setups, the one including both levee and channel bathymetry showed lowest fitness score on the spatial agreement of inundation extent, due to the weak physical representation of low-complexity models compared to a hydrodynamic model. For inundation depth, the low-complexity models showed an overestimating tendency, especially in the deeper segments of the channel. Based on such reasonably good prediction skills, low-complexity flood models can be considered as a suitable alternative for fast predictions in large-scale hyper-resolution operational frameworks, without completely overriding hydrodynamic models' efficacy.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.

A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11--12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented.

Statistical evaluation of the metallurgical test data in the ORR-PSF-PVS irradiation experiment. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stallmann, F.W.

1984-08-01

A statistical analysis of Charpy test results of the two-year Pressure Vessel Simulation metallurgical irradiation experiment was performed. Determination of transition temperature and upper shelf energy derived from computer fits compare well with eyeball fits. Uncertainties for all results can be obtained with computer fits. The results were compared with predictions in Regulatory Guide 1.99 and other irradiation damage models.

TRIGA MARK-II source term

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Analysis of loss of decay-heat-removal sequences at Browns Ferry Unit One

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.

1983-01-01

This paper summarizes the Oak Ridge National Laboratory (ORNL) report Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA). The purpose of the SASA studies is to predetermine the probable course of postulated severe accidents so as to establish the timing andmore » the sequence of events. The SASA studies also produce recommendations concerning the implementation of better system design and better emergency operating instructions and operator training. The ORNL studies also include a detailed, best-estimate calculation of the release and transport of radioactive fission products following postulated severe accidents.« less

Main steam-line break core shroud loading calculations for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1995-12-31

In July 1994, the U.S. Nuclear regulatory Commission sent out Generic Letter 94-03 to all boiling water reactors in the United States, informing them of intergranular stress corrosion cracking of core shrouds found in 2 reactors. The letter directed all to perform safety analysis of the BWR units. Penn State performed scoping calculations to determine the forces experienced by the core shroud during a main-stream line break transient.

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

ESBWR response to an extended station blackout/loss of all AC power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barrett, A. J.; Marquino, W.

2012-07-01

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackouts for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by internationalmore » regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event. Beyond 7 days, only a few simple actions are needed to cope with the SBO for an indefinite amount of time. The operation of the ICS as the primary success path for mitigation of an SBO, allows for near immediate plant restart once power is restored. (authors)« less

TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1996-12-31

The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient.

Interim reliability evaluation program, Browns Ferry 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.

1981-01-01

Probabilistic risk analysis techniques, i.e., event tree and fault tree analysis, were utilized to provide a risk assessment of the Browns Ferry Nuclear Plant Unit 1. Browns Ferry 1 is a General Electric boiling water reactor of the BWR 4 product line with a Mark 1 (drywell and torus) containment. Within the guidelines of the IREP Procedure and Schedule Guide, dominant accident sequences that contribute to public health and safety risks were identified and grouped according to release categories.

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

Test Plan for the Boiling Water Reactor Dry Cask Simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing themore » internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.« less

Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

BNL severe-accident sequence experiments and analysis program. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, G.A.; Ginsberg, T.; Tutu, N.K.

1983-01-01

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

Disposition of feedwater nozzle UT indications in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leshnoff, S.D.; Orski, M.A.

A technical logic is developed, which justifies the disposition of feedwater nozzle ultrasonic testing (UT) indications in order to return to operation without visual inspection of the vessel inside surface. Present regulatory guidance is to inspect the inside surface from the inside if a reportable indication is found. A highly sensitive, tomographic UT technique, developed by Kraftwerk Union, is used to detect and size machined notches in the blend radius and bore regions of a full-sized feedwater nozzle mock-up.

Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.C. Ryman

This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operatingmore » conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 7.28). The performance of the calculation and development of this document are carried out in accordance with AP-3.124, ''Design Calculation and Analyses'' (Ref. 7.29).« less

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

Vascularity as assessed by Doppler intraoral ultrasound around the invasion front of tongue cancer is a predictor of pathological grade of malignancy and cervical lymph node metastasis.

PubMed

Yamamoto, Chika; Yuasa, Kenji; Okamura, Kazuhiko; Shiraishi, Tomoko; Miwa, Kunihiro

2016-01-01

To quantitatively evaluate the relationship of vascularity of tongue cancer as demonstrated on intraoral ultrasonography images and tumour thickness with pathological grade of malignancy and the presence of cervical lymph node metastases. 18 patients with tongue cancer were enrolled in this retrospective study. Using Doppler ultrasonography images of the invasion front of the cancers along the length of their tumour boundaries, three vascular indexes were analysed quantitatively, namely ratio of blood flow signal area within the cancer to whole tumour area (BAR), blood flow signal number ratio (BNR) and blood flow signal width ratio (BWR). The associations between these three indexes and occurrence of cervical lymph node metastasis and pathological grade of malignancy [Yamamoto-Kohama (YK) classification] were assessed. Furthermore, the relationship between tumour thickness and occurrence of cervical lymph node metastasis was evaluated on B-mode intraoral ultrasonography images. There was no significant association between BAR and tumour thickness or occurrence of cervical lymph node metastasis. The BNRs and BWRs of patients with cervical lymph node metastasis were significantly higher than those of patients without nodal involvement. The BWRs of patients with high-grade malignancy (YK-4C) were significantly higher than those of patients with low-grade malignancy (YK-2 or 3). BNR and BWR on the invasion front of the tongue cancer are predictors of pathological grade of malignancy and cervical lymph node metastasis.

Application Of The Iberdrola Licensing Methodology To The Cofrentes BWR-6 110% Extended Power Up-rate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mata, Pedro; Fuente, Rafael de la; Iglesias, Javier

Iberdrola (spanish utility) and Iberdrola Ingenieria (engineering branch) have been developing during the last two years the 110% Extended Power Up-rate Project (EPU 110%) for Cofrentes BWR-6. IBERDROLA has available an in-house design and licensing reload methodology that has been approved by the Spanish Nuclear Regulatory Authority. This methodology has been already used to perform the nuclear design and the reload licensing analysis for Cofrentes cycles 12 to 14. The methodology has been also applied to develop a significant number of safety analysis of the Cofrentes Extended Power Up-rate including: Reactor Heat Balance, Core and Fuel performance, Thermal Hydraulic Stability,more » ECCS LOCA Evaluation, Transient Analysis, Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) Since the scope of the licensing process of the Cofrentes Extended Power Up-rate exceeds the range of analysis included in the Cofrentes generic reload licensing process, it has been required to extend the applicability of the Cofrentes licensing methodology to the analysis of new transients. This is the case of the TLFW transient. The content of this paper shows the benefits of having an in-house design and licensing methodology, and describes the process to extend the applicability of the methodology to the analysis of new transients. The case of analysis of Total Loss of Feedwater with the Cofrentes Retran Model is included as an example of this process. (authors)« less

Measurement of liver function using hepatobiliary scintigraphy improves risk assessment in patients undergoing major liver resection.

PubMed

Cieslak, Kasia P; Bennink, Roelof J; de Graaf, Wilmar; van Lienden, Krijn P; Besselink, Marc G; Busch, Olivier R C; Gouma, Dirk J; van Gulik, Thomas M

2016-09-01

(99m)Tc-mebrofenin-hepatobiliary-scintigraphy (HBS) enables measurement of future remnant liver (FRL)-function and was implemented in our preoperative routine after calculation of the cut-off value for prediction of postoperative liver failure (LF). This study evaluates our results since the implementation of HBS. Additionally, CT-volumetric methods of FRL-assessment, standardized liver volumetry and FRL/body-weight ratio (FRL-BWR), were evaluated. 163 patients who underwent major liver resection were included. Insufficient FRL-volume and/or FRL-function <2.7%/min/m(2) were indications for portal vein embolization (PVE). Non-PVE patients were compared with a historical cohort (n = 55). Primary endpoints were postoperative LF and LF related mortality. Secondary endpoint was preoperative identification of patients at risk for LF using the CT-volumetric methods. 29/163 patients underwent PVE; 8/29 patients because of insufficient FRL-function despite sufficient FRL-volume. According to FRL-BWR and standardized liver volumetry, 16/29 and 11/29 patients, respectively, would not have undergone PVE. LF and LF related mortality were significantly reduced compared to the historical cohort. HBS appeared superior in the identification of patients with increased surgical risk compared to the CT-volumetric methods. Implementation of HBS in the preoperative work-up led to a function oriented use of PVE and was associated with a significant decrease in postoperative LF and LF related mortality. Copyright © 2016 International Hepato-Pancreato-Biliary Association Inc. Published by Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

1995-09-01

The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

Seismic margin assessment of the Edwin I. Hatch Nuclear Plant, Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barr, W.T.; Moore, D.P.; Smith, J.E.

1991-06-01

This summary presents the results and lessons learned from the seismic margin assessment (SMA) of Unit 1 of the Hatch Nuclear Plant. The primary purpose of this SMA was to assess the practicality of the EPRI SMA methodology on a BWR on a soil site such as Hatch. The major findings from the Hatch SMA are briefly described along with the lessons learned during the project implementation. The experience gained on the Hatch SMA is expected to benefit others in the performance of future SMAs. 12 refs.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

The effect of zinc injection on the increasing of Inconel 600 TT corrosion resistances

NASA Astrophysics Data System (ADS)

Febrianto; Sriyono; Widodo, Surip; Sunaryo, Geni Rina

2018-02-01

Many failures were found in reactor pressure vessel head penetration (RPV) head material. Those failures caused by boric acid corrosion, and from visual examination were found a big hole and white deposit crystal of boric acid during shutdown maintenance at David Besse reactor. Zinc Oxide addition in BWR reactor known as Zinc Injection that has purposed to reduce radiation exposure cause of Hydrogen addition. Beside reducing the radiation exposure, Zinc injection also has an effect in reducing material corrosion. The purpose of study is to determine the effect of zinc addition, boric acid, temperature also the effects of Cobalt Nitrate and Zinc Oxide addition to Inconel 600 TT as RPV head penetration material. The result in the BWR reactor experience will be implementated at PWR reactor, weather zinc oxide addition also has an effect in reducing the corrosion of Inconel 600. The method that used in this research is to observe the corrosion rates for Inconel 600 material using Potentiostat. Examination were conducted in 30, 40, 60, 70, 80 and 80 °C using 1000, 1500, 2000, 2500 and 3000 ppm boric acid concentration. The results showed that the corrosion rate for the material were very small, but the highest corrosion rate occurred in 3000 ppm boric acid concentration at 90 °C with Cobalt Nitrate addition, around 5.210 x 10-1 mpy. In the same condition at 3000 ppm boric acid concentration for temperature at 90 °C, Inconel 600 TT corrosion rate is smaller with Zinc oxide addition, around 4.631 x 10-1 mpy.

Recent plant studies using Victoria 2.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

BIXLER,NATHAN E.; GASSER,RONALD D.

2000-03-08

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised andmore » expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment.« less

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Was, Gary

This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

Spent fuel data base: commercial light water reactors. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Use of eddy current mixes to solve a weld examination application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, R.C.; LaBoissonniere, A.

1995-12-31

The augmentation of typical nondestructive (i.e., ultrasound) weld inspection techniques by the use of eddy current tools may significantly enhance the quality and reliability of weld inspections. One recent example is the development of an eddy current technique for use in the examination of BWR core shroud welds, where multi-frequency mixes are used to eliminate signals coming from the weld material so that the examination of the heat affected zone is enhanced. An analysis tool most commonly associated with ultrasound examinations, the C-Scan based on gated information, may be implemented with eddy current data to enhance analysis.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Preliminary Thermal Modeling of Hi-Storm 100S-218 Version B Storage Modules at Hope Creek Nuclear Power Station ISFSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuta, Judith M.; Adkins, Harold E.

2013-08-30

This report fulfills the M3 milestone M3FT-13PN0810022, “Report on Inspection 1”, under Work Package FT-13PN081002. Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for four modules at the Hope Creek Nuclear Generating Station ISFSI that have been identified as candidates for inspection in late summer or early fall/winter of 2013. Thesemore » are HI-STORM 100S-218 Version B modules storing BWR 8x8 fuel in MPC-68 canisters. The temperature predictions reported in this document were obtained with detailed COBRA-SFS models of these four storage systems, with the following boundary conditions and assumptions.« less

In-Containment Signal Conditioning and Transmission via Power Lines within High Dose Rate Areas of Nuclear Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, Steffen; Weigel, Robert; Koelpin, Alexander

2015-07-01

Signal conditioning and transmission for sensor systems and networks within the containment of nuclear power plants (NPPs) still poses a challenge to engineers, particularly in the case of equipment upgrades for existing plants, temporary measurements, decommissioning of plants, but also for new builds. This paper presents an innovative method for efficient and cost-effective instrumentation within high dose rate areas inside the containment. A transmitter-receiver topology is proposed that allows simultaneous, unidirectional point-to-point transmission of multiple sensor signals by superimposing them on existing AC or DC power supply cables using power line communication (PLC) technology. Thereby the need for costly installationmore » of additional cables and containment penetrations is eliminated. Based on commercial off-the-shelf (COTS) electronic parts, a radiation hard transmitter is designed to operate in harsh environment within the containment during full plant operation. Hardware modularity of the transmitter allows application specific tradeoffs between redundancy and channel bandwidth. At receiver side in non-radiated areas, signals are extracted from the power line, demodulated, and provided either in analog or digital output format. Laboratory qualification tests and field test results within a boiling water reactor (BWR) are validating the proof of concept of the proposed system. (authors)« less

Dynamic interaction between fetal adversity and a genetic score reflecting dopamine function on developmental outcomes at 36 months.

PubMed

Bischoff, Adrianne R; Pokhvisneva, Irina; Léger, Étienne; Gaudreau, Hélène; Steiner, Meir; Kennedy, James L; O'Donnell, Kieran J; Diorio, Josie; Meaney, Michael J; Silveira, Patrícia P

2017-01-01

Fetal adversity, evidenced by poor fetal growth for instance, is associated with increased risk for several diseases later in life. Classical cut-offs to characterize small (SGA) and large for gestational age (LGA) newborns are used to define long term vulnerability. We aimed at exploring the possible dynamism of different birth weight cut-offs in defining vulnerability in developmental outcomes (through the Bayley Scales of Infant and Toddler Development), using the example of a gene vs. fetal adversity interaction considering gene choices based on functional relevance to the studied outcome. 36-month-old children from an established prospective birth cohort (Maternal Adversity, Vulnerability, and Neurodevelopment) were classified according to birth weight ratio (BWR) (SGA ≤0.85, LGA >1.15, exploring a wide range of other cut-offs) and genotyped for polymorphisms associated with dopamine signaling (TaqIA-A1 allele, DRD2-141C Ins/Ins, DRD4 7-repeat, DAT1-10- repeat, Met/Met-COMT), composing a score based on the described function, in which hypofunctional variants received lower scores. There were 251 children (123 girls and 128 boys). Using the classic cut-offs (0.85 and 1.15), there were no statistically significant interactions between the neonatal groups and the dopamine genetic score. However, when changing the cut-offs, it is possible to see ranges of BWR that could be associated with vulnerability to poorer development according to the variation in the dopamine function. The classic birth weight cut-offs to define SGA and LGA newborns should be seen with caution, as depending on the outcome in question, the protocols for long-term follow up could be either too inclusive-therefore most costly, or unable to screen true vulnerabilities-and therefore ineffective to establish early interventions and primary prevention.

Passive gamma analysis of the boiling-water-reactor assemblies

NASA Astrophysics Data System (ADS)

Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

Passive gamma analysis of the boiling-water-reactor assemblies

DOE PAGES

Vo, D.; Favalli, A.; Grogan, B.; ...

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

Method for depleting BWRs using optimal control rod patterns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1991-01-01

Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonicsmore » calculations.« less

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Clarno, Kevin T.; Gentry, Cole

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Coupling Schemes for Multiphysics Reactor Simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vijay Mahadeven; Jean Ragusa

2007-11-01

This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of amore » BWR core.« less

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Current and anticipated uses of thermal-hydraulic codes in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

NASA Astrophysics Data System (ADS)

Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

1990-12-01

Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation,more » boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.« less

Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1990-02-01

The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of anymore » cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs.« less

Dividing phases in two-phase flow and modeling of interfacial drag

DOE Office of Scientific and Technical Information (OSTI.GOV)

Narumo, T.; Rajamaeki, M.

1997-07-01

Different models intended to describe one-dimensional two-phase flow are considered in this paper. The following models are introduced: conventional six-equation model, conventional model equipped with terms taking into account nonuniform transverse velocity distribution of the phases, several virtual mass models and a model in which the momentum equations have been derived by using the principles of Separation of the Flow According to Velocity (SFAV). The dynamics of the models have been tested by comparing their characteristic velocities to each other and against experimental data. The results show that the SFAV-model makes a hyperbolic system and predicts the propagation velocities ofmore » disturbances with the same order of accuracy as the best tested virtual mass models. Furthermore, the momentum interaction terms for the SFAV-model are considered. These consist of the wall friction terms and the interfacial friction term. The authors model wall friction with two independent terms describing the effect of each fluid on the wall separately. In the steady state, a relationship between the slip velocity and friction coefficients can be derived. Hence, the friction coefficients for the SFAV-model can be calculated from existing correlations, viz. from a drift-flux correlation and a wall friction correlation. The friction model was tested by searching steady-state distributions in a partial BWR fuel channel and comparing the relaxed values with the drift-flux correlation, which agreed very well with each other. In addition, response of the flow to a sine-wave disturbance in the water inlet flux was calculated as function of frequency. The results of the models differed from each other already with frequency of order 5 Hz, while the time constant for the relaxation, obtained from steady-state distribution calculation, would have implied significant differences appear not until with frequency of order 50 Hz.« less

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, M; Blink, J A; Greenberg, H R

2012-04-25

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of wastemore » forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, B. W.; Williamson, R. L.; Stafford, D. S.

One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can bemore » used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed in this paper. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Finally, parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding adjacent to the defect.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilman, J

2005-03-15

To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials.

Characterization of carbon-14 generated by the nuclear power industry. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eabry, S.; Vance, J.N.; Cline, J.E.

1995-11-01

This report describes an evaluation of C-14 production rates in light-water reactors (LWRs) and characterization of its chemical speciation and environmental behavior. The study estimated the total production rate of the nuclide in operating PWRs and BWRs along with the assessment of the C-14 content of solid radwaste. The major source of production of C-14 in both PWR`s and BWRs was the activation of 0-17 in the water molecule and of N-14 dissolved in reactor coolant. The production of C-14 was estimated to range from 7 Ci/GW(e)-year to 11 Ci/GW(e)-year. The estimated range of the quantity of C-14 in LLWmore » was 1-2 Ci/ reactor-year which compares favorably with data obtained from shipping manifests. The environmental behavior of C-14 associated with low-level waste (LLW) disposal is greatly dependent upon its chemical speciation. This scoping study was performed to help identify the occurrence of inorganic and organic forms of C-14 in reactor coolant water and in primary coolant demineralization resins. These represent the major source for C-14 in LLW from nuclear power stations. Also, the behavior of inorganic and two of the organic forms of C-14 on soil uptake was determined by measuring distribution coefficients (Kd`s) on two soil types and a cement, using two different groundwater types. This study confirms that C-14 concentrations are significantly higher in the primary coolant from PWR stations compared to BWR stations. The C-14 followed trends of Co-60 generation during primary coolant demineralization at all but one of the stations examined. However, the C-14/Co-60 activity ratios measured by this study in resin samples through which samples of coolant were drawn were about 8 to 42 times higher than those reported for waste samples in the industry data base for PWR stations, and 15 to 730 times lower for the BWR stations.« less

New core-reflector boundary conditions for transient nodal reactor calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, E.K.; Kim, C.H.; Joo, H.K.

1995-09-01

New core-reflector boundary conditions designed for the exclusion of the reflector region in transient nodal reactor calculations are formulated. Spatially flat frequency approximations for the temporal neutron behavior and two types of transverse leakage approximations in the reflector region are introduced to solve the transverse-integrated time-dependent one-dimensional diffusion equation and then to obtain relationships between net current and flux at the core-reflector interfaces. To examine the effectiveness of new core-reflector boundary conditions in transient nodal reactor computations, nodal expansion method (NEM) computations with and without explicit representation of the reflector are performed for Laboratorium fuer Reaktorregelung und Anlagen (LRA) boilingmore » water reactor (BWR) and Nuclear Energy Agency Committee on Reactor Physics (NEACRP) pressurized water reactor (PWR) rod ejection kinetics benchmark problems. Good agreement between two NEM computations is demonstrated in all the important transient parameters of two benchmark problems. A significant amount of CPU time saving is also demonstrated with the boundary condition model with transverse leakage (BCMTL) approximations in the reflector region. In the three-dimensional LRA BWR, the BCMTL and the explicit reflector model computations differ by {approximately}4% in transient peak power density while the BCMTL results in >40% of CPU time saving by excluding both the axial and the radial reflector regions from explicit computational nodes. In the NEACRP PWR problem, which includes six different transient cases, the largest difference is 24.4% in the transient maximum power in the one-node-per-assembly B1 transient results. This difference in the transient maximum power of the B1 case is shown to reduce to 11.7% in the four-node-per-assembly computations. As for the computing time, BCMTL is shown to reduce the CPU time >20% in all six transient cases of the NEACRP PWR.« less

Dynamic interaction between fetal adversity and a genetic score reflecting dopamine function on developmental outcomes at 36 months

PubMed Central

Pokhvisneva, Irina; Léger, Étienne; Gaudreau, Hélène; Steiner, Meir; Kennedy, James L.; O’Donnell, Kieran J.; Diorio, Josie; Meaney, Michael J.; Silveira, Patrícia P.

2017-01-01

Background Fetal adversity, evidenced by poor fetal growth for instance, is associated with increased risk for several diseases later in life. Classical cut-offs to characterize small (SGA) and large for gestational age (LGA) newborns are used to define long term vulnerability. We aimed at exploring the possible dynamism of different birth weight cut-offs in defining vulnerability in developmental outcomes (through the Bayley Scales of Infant and Toddler Development), using the example of a gene vs. fetal adversity interaction considering gene choices based on functional relevance to the studied outcome. Methods 36-month-old children from an established prospective birth cohort (Maternal Adversity, Vulnerability, and Neurodevelopment) were classified according to birth weight ratio (BWR) (SGA ≤0.85, LGA >1.15, exploring a wide range of other cut-offs) and genotyped for polymorphisms associated with dopamine signaling (TaqIA-A1 allele, DRD2-141C Ins/Ins, DRD4 7-repeat, DAT1-10- repeat, Met/Met-COMT), composing a score based on the described function, in which hypofunctional variants received lower scores. Results There were 251 children (123 girls and 128 boys). Using the classic cut-offs (0.85 and 1.15), there were no statistically significant interactions between the neonatal groups and the dopamine genetic score. However, when changing the cut-offs, it is possible to see ranges of BWR that could be associated with vulnerability to poorer development according to the variation in the dopamine function. Conclusion The classic birth weight cut-offs to define SGA and LGA newborns should be seen with caution, as depending on the outcome in question, the protocols for long-term follow up could be either too inclusive—therefore most costly, or unable to screen true vulnerabilities—and therefore ineffective to establish early interventions and primary prevention. PMID:28505190

Condensate polisher prefiltration study for Laguna Verde Station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garcia, A.; Oyen, L.C.; Nelson, R.A.

1995-05-01

This paper describes an analysis of the iron and copper in the condensate and the technical and economic assessment of the installation of condensate polisher prefilters in Comision Federal de Electricidad`s Laguna Verde Nuclear Generating Station (LVNGS) north of Veracruz, Mexico. LVNGS is a 654 MWe General Electric BWR plant; Unit 1 has been in commercial operation since July, 1990, and Unit 2 is scheduled to become operational in June, 1995. The primary purpose of this study was to (1) analyze the high iron and copper concentrations in the condensate and feedwater, (2) identify, assess, and evaluate techniques to reducemore » the iron and copper concentrations, and (3) perform a cost-benefit analysis of the installation of implementing the appropriate techniques.« less

Data summary report for fission product release Test VI-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, M.F.; Lorentz, R.A.; Travis, J.R.

Test VI-7 was the final test in the VI series conducted in the vertical furnace. The fuel specimen was a 15.2-cm-long section of a fuel rod from the Monticello boiling water reactor (BWR). The fuel had experienced a burnup of {approximately}-40 Mwd/kg U. It was heated in an induction furnace for successive 20-min periods at 2000 and 2300 K in a moist air-helium atmosphere. Integral releases were 69% for {sup 85}Kr, 52% for {sup 125}Sb, 71% for both {sup 134}Cs and {sup 137}Cs, and 0.04% for {sup 154}Eu. For the non-gamma-emitting species, release values for 42% for I, 4.1% formore » Ba, 5.3% for Mo, and 1.2% for Sr were determined. The total mass released from the furnace to the collection system, including fission products, fuel, and structural materials, was 0.89 g, with 37% being collected on the thermal gradient tubes and 63% downstream on filters. Posttest examination of the fuel specimen indicated that most of the cladding was completely oxidized to ZrO{sub 2}, but that oxidation was not quite complete at the upper end. The release behaviors for the most volatile elements, Kr and Cs, were in good agreement with the ORNL-Booth Model.« less

Emergency cooling system and method

DOEpatents

Oosterkamp, W.J.; Cheung, Y.K.

1994-01-04

An improved emergency cooling system and method are disclosed that may be adapted for incorporation into or use with a nuclear BWR wherein a reactor pressure vessel (RPV) containing a nuclear core and a heat transfer fluid for circulation in a heat transfer relationship with the core is housed within an annular sealed drywell and is fluid communicable therewith for passage thereto in an emergency situation the heat transfer fluid in a gaseous phase and any noncondensibles present in the RPV, an annular sealed wetwell houses the drywell, and a pressure suppression pool of liquid is disposed in the wetwell and is connected to the drywell by submerged vents. The improved emergency cooling system and method has a containment condenser for receiving condensible heat transfer fluid in a gaseous phase and noncondensibles for condensing at least a portion of the heat transfer fluid. The containment condenser has an inlet in fluid communication with the drywell for receiving heat transfer fluid and noncondensibles, a first outlet in fluid communication with the RPV for the return to the RPV of the condensed portion of the heat transfer fluid and a second outlet in fluid communication with the drywell for passage of the noncondensed balance of the heat transfer fluid and the noncondensibles. The noncondensed balance of the heat transfer fluid and the noncondensibles passed to the drywell from the containment condenser are mixed with the heat transfer fluid and the noncondensibles from the RPV for passage into the containment condenser. A water pool is provided in heat transfer relationship with the containment condenser and is thermally communicable in an emergency situation with an environment outside of the drywell and the wetwell for conducting heat transferred from the containment condenser away from the wetwell and the drywell. 5 figs.

Round Robin Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Klamerus, E.W.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. Several organizations from the US, Europe, and Asia were invited to participate in a Round Robin analysis to perform independent pretest predictions and posttest evaluations of the behavior of the SCV model during the high pressure test. Both pretest and posttest analysis results from all Round Robin participants were compared tomore » the high pressure test data. This paper summarizes the Round Robin analysis activities and discusses the lessons learned from the collective effort.« less

Characterization of cracking behavior using posttest fractographic analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, T.; Shockey, D.A.

A determination of time to initiation of stress corrosion cracking in structures and test specimens is important for performing structural failure analysis and for setting inspection intervals. Yet it is seldom possible to establish how much of a component's lifetime represents the time to initiation of fracture and how much represents postinitiation crack growth. This exploratory research project was undertaken to examine the feasibility of determining crack initiation times and crack growth rates from posttest examination of fracture surfaces of constant-extension-rate-test (CERT) specimens by using the fracture reconstruction applying surface topography analysis (FRASTA) technique. The specimens used in this studymore » were Type 304 stainless steel fractured in several boiling water reactor (BWR) aqueous environments. 2 refs., 25 figs., 2 tabs.« less

Fuel Performance Calculations for FeCrAl Cladding in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

Optimizations of geothermal cycle shell and tube exchangers of various configurations with variable fluid properties and site specific fouling. [SIZEHX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, W.L.; Pines, H.S.; Silvester, L.F.

1978-03-01

A new heat exchanger program, SIZEHX, is described. This program allows single step multiparameter cost optimizations on single phase or supercritical exchanger arrays with variable properties and arbitrary fouling for a multitude of matrix configurations and fluids. SIZEHX uses a simplified form of Tinker's method for characterization of shell side performance; the Starling modified BWR equation for thermodynamic properties of hydrocarbons; and transport properties developed by NBS. Results of four parameter cost optimizations on exchangers for specific geothermal applications are included. The relative mix of capital cost, pumping cost, and brine cost ($/Btu) is determined for geothermal exchangers illustrating themore » invariant nature of the optimal cost distribution for fixed unit costs.« less

Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagrman, D. L.

1979-02-01

An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on amore » more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.« less

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.; Cazzoli, E.

1984-01-01

This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve,more » and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.« less

Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

DOE PAGES

Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; ...

2015-11-13

Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, large uncertainty in a particular aspect of the model may or may not have a large impact on the final results. For example, modeling of a relief valve may result in large uncertainty, however, the actual effects on final peakmore » clad temperature in a reactor transient may be small and the large uncertainty with respect to valve modeling may be considered acceptable. Additionally, the ability to determine the adequacy of a model and the validation supporting it should be considered within a risk informed framework. Low fidelity modeling with large uncertainty may be considered adequate if the uncertainty is considered acceptable with respect to risk. In other words, models that are used to evaluate the probability of failure should be evaluated more rigorously with the intent of increasing safety margin. Probabilistic risk assessment (PRA) techniques have traditionally been used to identify accident conditions and transients. Traditional classical event tree methods utilize analysts’ knowledge and experience to identify the important timing of events in coordination with thermal-hydraulic modeling. These methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This study presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. Finally, in order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.« less

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.« less

Corrosion and hydrogen pick-up behaviors of cladding and structural components in BWR high burnup 9x9 lead use assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miyashita, Toshiyasu; Nakae, Nobuo; Ogata, Keizo

The high burnup BWR 9x9 lead use fuel assemblies, which have been designed for maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiations to confirm the reliability of the current safety evaluation methodology, and to accumulate data to judge the adequacy to apply it to the future higher burnup fuel. After 3 and 5 cycle irradiations, post irradiation examinations were performed for both 9x9 Type-A and Type-B fuel assemblies. Both Type LUAs utilize Zry-2 claddings, while there are deviation in the contents of impurity and alloying elements between Type-A and Type-B, especially in Fe and Simore » concentration. Measured oxide thicknesses of fuel rods showed no significant difference between after 3 and 5 cycle irradiation except for some rods at corner position in Type B LUA. The axial profile of hydrogen concentration and oxide thickness for the corner rods in Type B LUA after 5 cycle irradiation had peaks at the second lowest span from the bottom. The maximum oxide thickness is about 50 {mu}m on the surface facing the bundle outside at the second lowest span and dense hydrides layer (Hydride rim) is observed in peripheral region of cladding showing unexpected high hydrogen concentration. The results of calculated thermal-hydraulic conditions show that the thermal neutron flux at the corner position was higher than the other position. On the other hand, the void fraction and the mass flux were relatively lower at the corner position. The oxide thickness on spacer band and spacer cell of Zry-2 increases from 3 to 5 cycle irradiations. Spacer band of Zry-4 showed significantly thick oxide after 5 cycle irradiations but Hydrogen concentration was relatively small in contrast its obviously thick oxide in comparison with Zry-2 spacer bands. The large increase in hydrogen concentration was measured in Zry-2 spacers after 5 cycle irradiations and the evaluated hydrogen pick-up rate also increased remarkably. (authors)« less

Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Munson, L.F.; Divine, J.R.; Martin, J.B.

1983-06-01

Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show highmore » decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masuda, Y.; Chiba, N.; Matsuo, Y.

This research proposes to investigate the impact behavior of the steel plate of BWR containment vessels against missiles, caused by the postulated catastrophic failure of components with a high kinetic energy. Although the probability of the occurrence of missiles inside and outside of containment vessels is extremely low, the following items are required to maintain the integrity of containment vessels: the probability of the occurrence of missiles, the weight and energy of missiles, and the impact behavior of containment vessel steel plate against postulated missiles. In connection with the third item, an actualscale missile test was conducted. In addition, amore » computation analysis was performed to confirm the impact behavior against the missiles, in order to search for wide applicability to the various kinds of postulated missiles. This research tries to derive a new empirical formula which carries out the assessment of the integrity of containment vessels.« less

CPR methodology with new steady-state criterion and more accurate statistical treatment of channel bow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baumgartner, S.; Bieli, R.; Bergmann, U. C.

2012-07-01

An overview is given of existing CPR design criteria and the methods used in BWR reload analysis to evaluate the impact of channel bow on CPR margins. Potential weaknesses in today's methodologies are discussed. Westinghouse in collaboration with KKL and Axpo - operator and owner of the Leibstadt NPP - has developed an optimized CPR methodology based on a new criterion to protect against dryout during normal operation and with a more rigorous treatment of channel bow. The new steady-state criterion is expressed in terms of an upper limit of 0.01 for the dryout failure probability per year. This ismore » considered a meaningful and appropriate criterion that can be directly related to the probabilistic criteria set-up for the analyses of Anticipated Operation Occurrences (AOOs) and accidents. In the Monte Carlo approach a statistical modeling of channel bow and an accurate evaluation of CPR response functions allow the associated CPR penalties to be included directly in the plant SLMCPR and OLMCPR in a best-estimate manner. In this way, the treatment of channel bow is equivalent to all other uncertainties affecting CPR. Emphasis is put on quantifying the statistical distribution of channel bow throughout the core using measurement data. The optimized CPR methodology has been implemented in the Westinghouse Monte Carlo code, McSLAP. The methodology improves the quality of dryout safety assessments by supplying more valuable information and better control of conservatisms in establishing operational limits for CPR. The methodology is demonstrated with application examples from the introduction at KKL. (authors)« less

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Flow-induced vibration and fretting-wear damage in a moisture separator reheater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pettigrew, M.J.; Taylor, C.E.; Fisher, N.J.

1996-12-01

Tube failures due to excessive flow-induced vibration were experienced in the tube bundles of moisture separator reheaters in a BWR nuclear station. This paper presents the results of a root cause analysis and covers recommendations for continued operation and for replacement tube bundles. The following tasks are discussed: tube failure analysis; flow velocity distribution calculations; flow-induced vibration analysis with particular emphasis on finned-tubes; fretting-wear testing of a tube and tube-support material combination under simulated operating conditions; field measurements of flow-induced vibration; and development of vibration specifications for replacement tube bundles. The effect of transient operating conditions and of other operationalmore » changes such as tube fouling were considered in the analysis. This paper outlines a typical field problem and illustrates the application of flow-induced vibration technology for the solution of a practical problem.« less

Benefits of barrier fuel on fuel cycle economics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Kunz, C.L.

1988-01-01

Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect ofmore » fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs.« less

Steel Containment Vessel Model Test: Results and Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Hessheimer, M.F.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformationmore » behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.« less

Aging of electronics with application to nuclear power plant instrumentation. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Jr, R T; Thome, F V; Craft, C M

1983-01-01

A survey to identify areas of needed research to understand aging mechanisms for electronics in nuclear power plant instrumentation has been completed. The emphasis was on electronic components such as semiconductors, capacitors, and resistors used in safety-related instrumentation in the reactor containment area. The environmental and operational stress factors which may produce degradation during long-term operation were identified. Some attention was also given to humidity effects as related to seals and encapsulants, and failures in printed circuit boards and bonds and solder joints. Results suggest that neutron as well as gamma irradiations should be considered in simulating the aging environmentmore » for electronic components. Radiation dose-rate effects in semiconductor devices and organic capacitors need to be further investigated, as well as radiation-voltage bias synergistic effects in semiconductor devices and leakage and permeation of moisture through seals in electronics packages.« less

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

NASA Astrophysics Data System (ADS)

Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

2017-05-01

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhegang Ma

The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significantmore » damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.« less

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

High burn-up spent nuclear fuel transport reliability investigation

DOE PAGES

Wang, Jy-An; Wang, Hong; Jiang, Hao; ...

2018-04-15

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

High burn-up spent nuclear fuel transport reliability investigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An; Wang, Hong; Jiang, Hao

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

Investigation of natural circulation instability and transients in passively safe novel modular reactor

NASA Astrophysics Data System (ADS)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal-hydraulic and nuclear coupled startup transients are performed to investigate the flow instabilities at low pressure and low power conditions. Two different power ramps are chosen to study the effect of power density on the flow instability. The experimental startup transient tests show the existence of three different flow instability mechanisms during the low pressure startup transients, i.e., flashing instability, condensation induced instability, and density wave oscillations. Flashing instability in the chimney section of the test loop and density wave oscillation are the main flow instabilities observed when the system pressure is below 0.5 MPa. They show completely different type of oscillations, i.e., intermittent oscillation and sinusoidal oscillation, in void fraction profile during the startup transients. In order to perform nuclear-coupled startup transients with void reactivity feedback, the Point Kinetics model is utilized to calculate the transient power during the startup transients. In addition, the differences between the electric resistance heaters and typical fuel element are taken into account. The reactor power calculated shows some oscillations due to flashing instability during the transients. However, the void reactivity feedback does not have significant influence on the flow instability during the startup procedure for the NMR-50. Further investigation of very small power ramp on the startup transients is carried out for the thermal-hydraulic startup transients. It is found that very small power density can eliminate the flashing oscillation in the single phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. Furthermore, initially pressurized startup procedure is investigated to eliminate the main flow instabilities. The results show that the pressurized startup procedure can suppress the flashing instability at low pressure and low power conditions. In order to have a deep understanding of natural circulation flow instability, the quasi-steady tests are performed using the test facility installed with preheater and subcooler. The effects of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback are investigated in the quasi-steady state tests. The stability boundaries are determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. In order to predict the stability boundary theoretically, linear stability analysis in the frequency domain is performed at four sections of the loop. The flashing in the chimney is considered as an axially uniform heat source. The dimensionless characteristic equation of the pressure drop perturbation is obtained by considering the void fraction effect and outlet flow resistance in the chimney section. The flashing boundary shows some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium is recommended to improve the accuracy of flashing instability boundary.

Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.

1982-12-01

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency andmore » R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Job analysis of maintenance-mechanic position for the nuclear power plant maintenance personnel reliability model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siegel, A.I.; Bartter, W.D.; Kopstein, F.F.

1982-06-01

The task list method of job survey was used. In collaboration with BWR and PWR personnel, a list of 107 tasks performed by maintenance mechanics was developed, grouped into: remove and install, test and repair, inspect and perform preventive maintenance, miscellaneous, communication, and report preparation. For each listed task, the questionnaire form inquired into: frequency of performance, task completion time, safety consequences of improper performance, and the amount of training required to perform the task proficiently. Scaled information was requested about seven abilities: (1) visual speed, accuracy, and recognition; (2) gross motor coordination; (3) fine manual dexterity; (4) strength andmore » stamina; (5) cognition; (6) memory; and (7) problem solving required for function completion. Survey forms were distributed to 27 nuclear power plants. Thirty-one maintenance mechanics representing 17 plants returned the completed forms. Frequency of performing tasks was bimodally distributed: (1) between once a year and once every six months, and (2) about once a week. More than half of the tasks have potential risk consequences if improperly performed. The five tasks with the greatest risk implications in the case of inadequate performance were: (1) remove and install reactor and dry-well heads, (2) test and repair reactor system components, (3) remove and install pressurizer mechanical relief valves, (4) test and repair pressurizer relief valves, (5) remove and install core spray pumps, seals, and valves. Hierarchically, the public risk associated with the various functions was: (1) remove and install, (2) test and repair, (3) preventive maintenance, (4) miscellaneous tasks, (5) communication, and (6) report preparation.« less

Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel; Lindgren, Eric R.

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internalmore » convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.« less

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Waste Package Component Design Methodology Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.C. Mecham

2004-07-12

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and usemore » of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.« less

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.

2016-08-30

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling themore » integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.« less

Development and Application of Laser Peening System for PWR Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)« less

Status Report on Ex-Vessel Coolability and Water Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.; Robb, K. R.

Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell ventmore » path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.« less

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, R. L.; Pastore, G.; Gamble, K. A.

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to designmore » a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.« less

Corrosion of pre-oxidized nickel alloy X-750 in simulated BWR environment

NASA Astrophysics Data System (ADS)

Tuzi, Silvia; Lai, Haiping; Göransson, Kenneth; Thuvander, Mattias; Stiller, Krystyna

2017-04-01

Samples of pre-oxidized Alloy X-750 were exposed to a simulated boiling water reactor environment in an autoclave at a temperature of 286 °C and a pressure of 80 bar for four weeks. The effect of alloy iron content on corrosion was investigated by comparing samples with 5 and 8 wt% Fe, respectively. In addition, the effect of two different surface pre-treatments was investigated. The microstructure of the formed oxide scales was studied using mainly electron microscopy. The results showed positive effects of an increased Fe content and of removing the deformed surface layer by pickling. After four weeks of exposure the oxide scale consists of oxides formed in three different ways. The oxide formed during pre-oxidization at 700 °C, mainly consisting of chromia, is partly still present. There is also an outer oxide consisting of NiFe2O4 crystals, reaching a maximum size of 3 μm, which has formed by precipitation of dissolved metal ions. Finally, there is an inner nanocrystalline and porous oxide, with a metallic content reflecting the alloy composition, which has formed by corrosion.

Clicker Implementation Models

ERIC Educational Resources Information Center

White, Peter J. T.; Delaney, David G.; Syncox, David; Akerberg, Oscar Avila; Alters, Brian

2011-01-01

Student response systems can help instructors integrate active learning into their classrooms. Such technology is known by a variety of names, including classroom response systems, student response systems, audience response systems, electronic response systems, personal response systems, zappers, and clickers. The "system" consists of three…

Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Triplett, C.E.

1996-12-01

This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the formmore » Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.« less

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H. M.

Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest claddingmore » were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.« less

A two-step method for developing a control rod program for boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1992-01-01

This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in amore » computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.« less

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baversten, B.; Linden, M.J.

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclearmore » overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.« less

Computational Analyses of Pressurization in Cryogenic Tanks

NASA Technical Reports Server (NTRS)

Ahuja, Vineet; Hosangadi, Ashvin; Mattick, Stephen; Lee, Chun P.; Field, Robert E.; Ryan, Harry

2008-01-01

A) Advanced Gas/Liquid Framework with Real Fluids Property Routines: I. A multi-fluid formulation in the preconditioned CRUNCH CFD(Registered TradeMark) code developed where a mixture of liquid and gases can be specified: a) Various options for Equation of state specification available (from simplified ideal fluid mixtures, to real fluid EOS such as SRK or BWR models). b) Vaporization of liquids driven by pressure value relative to vapor pressure and combustion of vapors allowed. c) Extensive validation has been undertaken. II. Currently working on developing primary break-up models and surface tension effects for more rigorous phase-change modeling and interfacial dynamics B) Framework Applied to Run-time Tanks at Ground Test Facilities C) Framework Used For J-2 Upper Stage Tank Modeling: 1) NASA MSFC tank pressurization: a) Hydrogen and oxygen tank pre-press, repress and draining being modeled at NASA MSFC. 2) NASA AMES tank safety effort a) liquid hydrogen and oxygen are separated by a baffle in the J-2 tank. We are modeling pressure rise and possible combustion if a hole develops in the baffle and liquid hydrogen leaks into the oxygen tank. Tank pressure rise rates simulated and risk of combustion evaluated.

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durbin, Samuel G.; Lindgren, Eric Richard

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing themore » internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.« less

What is health systems responsiveness? Review of existing knowledge and proposed conceptual framework.

PubMed

Mirzoev, Tolib; Kane, Sumit

2017-01-01

Responsiveness is a key objective of national health systems. Responsive health systems anticipate and adapt to existing and future health needs, thus contributing to better health outcomes. Of all the health systems objectives, responsiveness is the least studied, which perhaps reflects lack of comprehensive frameworks that go beyond the normative characteristics of responsive services. This paper contributes to a growing, yet limited, knowledge on this topic. Herewith, we review the current frameworks for understanding health systems responsiveness and drawing on these, as well as key frameworks from the wider public services literature, propose a comprehensive conceptual framework for health systems responsiveness. This paper should be of interest to different stakeholders who are engaged in analysing and improving health systems responsiveness. Our review shows that existing knowledge on health systems responsiveness can be extended along the three areas. First, responsiveness entails an actual experience of people's interaction with their health system, which confirms or disconfirms their initial expectations of the system. Second, the experience of interaction is shaped by both the people and the health systems sides of this interaction. Third, different influences shape people's interaction with their health system, ultimately affecting their resultant experiences. Therefore, recognition of both people and health systems sides of interaction and their key determinants would enhance the conceptualisations of responsiveness. Our proposed framework builds on, and advances, the core frameworks in the health systems literature. It positions the experience of interaction between people and health system as the centrepiece and recognises the determinants of responsiveness experience both from the health systems (eg, actors, processes) and the people (eg, initial expectations) sides. While we hope to trigger further thinking on the conceptualisation of health system responsiveness, the proposed framework can guide assessments of, and interventions to strengthen, health systems responsiveness.

What is health systems responsiveness? Review of existing knowledge and proposed conceptual framework

PubMed Central

Mirzoev, Tolib; Kane, Sumit

2017-01-01

Responsiveness is a key objective of national health systems. Responsive health systems anticipate and adapt to existing and future health needs, thus contributing to better health outcomes. Of all the health systems objectives, responsiveness is the least studied, which perhaps reflects lack of comprehensive frameworks that go beyond the normative characteristics of responsive services. This paper contributes to a growing, yet limited, knowledge on this topic. Herewith, we review the current frameworks for understanding health systems responsiveness and drawing on these, as well as key frameworks from the wider public services literature, propose a comprehensive conceptual framework for health systems responsiveness. This paper should be of interest to different stakeholders who are engaged in analysing and improving health systems responsiveness. Our review shows that existing knowledge on health systems responsiveness can be extended along the three areas. First, responsiveness entails an actual experience of people’s interaction with their health system, which confirms or disconfirms their initial expectations of the system. Second, the experience of interaction is shaped by both the people and the health systems sides of this interaction. Third, different influences shape people’s interaction with their health system, ultimately affecting their resultant experiences. Therefore, recognition of both people and health systems sides of interaction and their key determinants would enhance the conceptualisations of responsiveness. Our proposed framework builds on, and advances, the core frameworks in the health systems literature. It positions the experience of interaction between people and health system as the centrepiece and recognises the determinants of responsiveness experience both from the health systems (eg, actors, processes) and the people (eg, initial expectations) sides. While we hope to trigger further thinking on the conceptualisation of health system responsiveness, the proposed framework can guide assessments of, and interventions to strengthen, health systems responsiveness. PMID:29225953

Effectiveness Analysis of a Part-Time Rapid Response System During Operation Versus Nonoperation.

PubMed

Kim, Youlim; Lee, Dong Seon; Min, Hyunju; Choi, Yun Young; Lee, Eun Young; Song, Inae; Park, Jong Sun; Cho, Young-Jae; Jo, You Hwan; Yoon, Ho Il; Lee, Jae Ho; Lee, Choon-Taek; Do, Sang Hwan; Lee, Yeon Joo

2017-06-01

To evaluate the effect of a part-time rapid response system on the occurrence rate of cardiopulmonary arrest by comparing the times of rapid response system operation versus nonoperation. Retrospective cohort study. A 1,360-bed tertiary care hospital. Adult patients admitted to the general ward were screened. Data were collected over 36 months from rapid response system implementation (October 2012 to September 2015) and more than 45 months before rapid response system implementation (January 2009 to September 2012). None. The rapid response system operates from 7 AM to 10 PM on weekdays and from 7 AM to 12 PM on Saturdays. Primary outcomes were the difference of cardiopulmonary arrest incidence between pre-rapid response system and post-rapid response system periods and whether the rapid response system operating time affects the cardiopulmonary arrest incidence. The overall cardiopulmonary arrest incidence (per 1,000 admissions) was 1.43. Although the number of admissions per month and case-mix index were increased (3,555.18 vs 4,564.72, p < 0.001; 1.09 vs 1.13, p = 0.001, respectively), the cardiopulmonary arrest incidence was significantly decreased after rapid response system (1.60 vs 1.23; p = 0.021), and mortality (%) was unchanged (1.38 vs 1.33; p = 0.322). After rapid response system implementation, the cardiopulmonary arrest incidence significantly decreased by 40% during rapid response system operating times (0.82 vs 0.49/1,000 admissions; p = 0.001) but remained similar during rapid response system nonoperating times (0.77 vs 0.73/1,000 admissions; p = 0.729). The implementation of a part-time rapid response system reduced the cardiopulmonary arrest incidence based on the reduction of cardiopulmonary arrest during rapid response system operating times. Further analysis of the cost effectiveness of part-time rapid response system is needed.

Responsive systems - The challenge for the nineties

NASA Technical Reports Server (NTRS)

Malek, Miroslaw

1990-01-01

A concept of responsive computer systems will be introduced. The emerging responsive systems demand fault-tolerant and real-time performance in parallel and distributed computing environments. The design methodologies for fault-tolerant, real time and responsive systems will be presented. Novel techniques of introducing redundancy for improved performance and dependability will be illustrated. The methods of system responsiveness evaluation will be proposed. The issues of determinism, closed and open systems will also be discussed from the perspective of responsive systems design.

78 FR 34995 - Response Systems to Adult Sexual Assault Crimes Panel (Response Systems Panel); Notice of Federal...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... DEPARTMENT OF DEFENSE Office of the Secretary Response Systems to Adult Sexual Assault Crimes Panel (Response Systems Panel); Notice of Federal Advisory Committee Meeting AGENCY: Department of... committee meeting of the Response Systems to Adult Sexual Assault Crimes Panel. DATES: A meeting of the...

Development and Implementation of Mechanistic Terry Turbine Models in RELAP-7 to Simulate RCIC Normal Operation Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zou, Ling; Zhang, Hongbin

As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia’s original work [1], have been developed and implemented in the RELAP-7 code to simulate the RCIC system. In 2016, our effort has been focused on normal working conditions of the RCIC system. More complex off-design conditions will be pursued in later years when more data are available. In the Sandia model, the turbine stator inletmore » velocity is provided according to a reduced-order model which was obtained from a large number of CFD (computational fluid dynamics) simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine stator inlet. The models include both an adiabatic expansion process inside the nozzle and a free expansion process outside of the nozzle to ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input information for the Terry turbine rotor model. The analytical models for the nozzle were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The newly developed nozzle models and modified turbine rotor model according to the Sandia’s original work have been implemented into RELAP-7, along with the original Sandia Terry turbine model. A new pump model has also been developed and implemented to couple with the Terry turbine model. An input model was developed to test the Terry turbine RCIC system, which generates reasonable results. Both the INL RCIC model and the Sandia RCIC model produce results matching major rated parameters such as the rotational speed, pump torque, and the turbine shaft work for the normal operation condition. The Sandia model is more sensitive to the turbine outlet pressure than the INL model. The next step will be further refining the Terry turbine models by including two-phase flow cases so that off-design conditions can be simulated. The pump model could also be enhanced with the use of the homologous curves.« less

Lecture-Free High School Biology Using an Audience Response System

ERIC Educational Resources Information Center

Barnes, Larry J.

2008-01-01

Audience Response Systems (ARS) represent a powerful new tool for increasing student engagement. ARS technology (known variously as electronic voting systems, personal response systems, interactive student response systems, and classroom performance systems) includes one hand-held remote per student, a receiver (infrared or radio frequency,…

44 CFR 208.3 - Authority for the National US&R Response System.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Order 12148. (b) Implementing plan. The National Response Plan identifies DHS as the primary Federal...&R Response System. 208.3 Section 208.3 Emergency Management and Assistance FEDERAL EMERGENCY... RESPONSE SYSTEM General § 208.3 Authority for the National US&R Response System. (a) Enabling legislation...

44 CFR 208.3 - Authority for the National US&R Response System.

Code of Federal Regulations, 2012 CFR

2012-10-01

... Order 12148. (b) Implementing plan. The National Response Plan identifies DHS as the primary Federal...&R Response System. 208.3 Section 208.3 Emergency Management and Assistance FEDERAL EMERGENCY... RESPONSE SYSTEM General § 208.3 Authority for the National US&R Response System. (a) Enabling legislation...

44 CFR 208.3 - Authority for the National US&R Response System.

Code of Federal Regulations, 2013 CFR

2013-10-01

... Order 12148. (b) Implementing plan. The National Response Plan identifies DHS as the primary Federal...&R Response System. 208.3 Section 208.3 Emergency Management and Assistance FEDERAL EMERGENCY... RESPONSE SYSTEM General § 208.3 Authority for the National US&R Response System. (a) Enabling legislation...

44 CFR 208.3 - Authority for the National US&R Response System.

Code of Federal Regulations, 2011 CFR

2011-10-01

... Order 12148. (b) Implementing plan. The National Response Plan identifies DHS as the primary Federal...&R Response System. 208.3 Section 208.3 Emergency Management and Assistance FEDERAL EMERGENCY... RESPONSE SYSTEM General § 208.3 Authority for the National US&R Response System. (a) Enabling legislation...

44 CFR 208.3 - Authority for the National US&R Response System.

Code of Federal Regulations, 2014 CFR

2014-10-01

... Order 12148. (b) Implementing plan. The National Response Plan identifies DHS as the primary Federal...&R Response System. 208.3 Section 208.3 Emergency Management and Assistance FEDERAL EMERGENCY... RESPONSE SYSTEM General § 208.3 Authority for the National US&R Response System. (a) Enabling legislation...

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

ALARA efforts in nordic BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ingemansson, T.; Lundgren, K.; Elkert, J.

1995-03-01

Some ALARA-related ABB Atom projects are currently under investigation. One of the projects has been ordered by the Swedish Radiation Protection Institute, and two others by the Nordic BWR utilities. The ultimate objective of the projects is to identify and develop methods to significantly decrease the future exposure levels in the Nordic BWRS. As 85% to 90% of the gamma radiation field in the Nordic BWRs originates from Co-60, the only way to significantly decrease the radiation doses is to effect Co and Co-60. The strategy to do this is to map the Co sources and estimate the source strengthmore » of Co from these sources, and to study the possibility to affect the release of Co-60 from the core surfaces and the uptake on system surfaces. Preliminary results indicate that corrosion/erosion of a relatively small number of Stellite-coated valves and/or dust from grinding of Stellite valves may significantly contribute to the Co input to the reactors. This can be seen from a high measured Co/Ni ratio in the feedwater and in the reactor water. If stainless steel is the only source of Co, the Co/Ni ratio would be less than 0.02 as the Co content in the steel is less than 0.2%. The Co/Ni ratio in the reactor water, however, is higher than 0.1, indicating that the major fraction of the Co originates from Stellite-coated valves. There are also other possible explanations for an increase of the radiation fields. The Co-60 inventory on the core surfaces increases approximately as the square of the burn-up level. If the burn-up is increased from 35 to 5 MWd/kgU, the Co-60 inventory on the core surfaces will be doubled. Also the effect on the behavior of Co-60 of different water chemistry and materials conditions is being investigated. Examples of areas studied are Fe and Zn injection, pH-control, and different forms of surface pre-treatments.« less

78 FR 53429 - Response Systems to Adult Sexual Assault Crimes Panel (Response Systems Panel); Notice of Federal...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... recommendations regarding how to improve the effectiveness of such systems. The Panel is interested in written and... DEPARTMENT OF DEFENSE Office of the Secretary Response Systems to Adult Sexual Assault Crimes Panel (Response Systems Panel); Notice of Federal Advisory Committee Meeting AGENCY: Department of...

Ecosystem and immune systems: Hierarchial response provides resilience against invasions

USGS Publications Warehouse

Allen, Craig R.

2001-01-01

Janssen (2001) provides the stimulus for thoughtful comparison and consideration of the ranges of responses exhibited by immune systems and ecological systems in the face of perturbations such as biological invasions. It may indeed be informative to consider the similarities of the responses to invasions exhibited by immune systems and ecological systems. Clearly, both types of systems share a general organizational structure with all other complex hierarchical systems. Their organization provides these systems with resilience. However, when describing the response of ecological-economic systems to invasions, Janssen emphasizes the human-economic response. I would like to expand on his comparison by focusing on how resilience is maintained in complex systems under the threat of invasion.

Error response test system and method using test mask variable

NASA Technical Reports Server (NTRS)

Gender, Thomas K. (Inventor)

2006-01-01

An error response test system and method with increased functionality and improved performance is provided. The error response test system provides the ability to inject errors into the application under test to test the error response of the application under test in an automated and efficient manner. The error response system injects errors into the application through a test mask variable. The test mask variable is added to the application under test. During normal operation, the test mask variable is set to allow the application under test to operate normally. During testing, the error response test system can change the test mask variable to introduce an error into the application under test. The error response system can then monitor the application under test to determine whether the application has the correct response to the error.

Comparison of Models for Spacer Grid Pressure Loss in Nuclear Fuel Bundles for One and Two-Phase Flows

NASA Astrophysics Data System (ADS)

Maskal, Alan B.

Spacer grids maintain the structural integrity of the fuel rods within fuel bundles of nuclear power plants. They can also improve flow characteristics within the nuclear reactor core. However, spacer grids add reactor coolant pressure losses, which require estimation and engineering into the design. Several mathematical models and computer codes were developed over decades to predict spacer grid pressure loss. Most models use generalized characteristics, measured by older, less precise equipment. The study of OECD/US-NRC BWR Full-Size Fine Mesh Bundle Tests (BFBT) provides updated and detailed experimental single and two-phase results, using technically advanced flow measurements for a wide range of boundary conditions. This thesis compares the predictions from the mathematical models to the BFBT experimental data by utilizing statistical formulae for accuracy and precision. This thesis also analyzes the effects of BFBT flow characteristics on spacer grids. No single model has been identified as valid for all flow conditions. However, some models' predictions perform better than others within a range of flow conditions, based on the accuracy and precision of the models' predictions. This study also demonstrates that pressure and flow quality have a significant effect on two-phase flow spacer grid models' biases.

Computer program for optimal BWR congtrol rod programming

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Carmody, J.M.

1995-12-31

A fully automated computer program has been developed for designing optimal control rod (CR) patterns for boiling water reactors (BWRs). The new program, called OCTOPUS-3, is based on the OCTOPUS code and employs SIMULATE-3 (Ref. 2) for the analysis. There are three aspects of OCTOPUS-3 that make it successful for use at PECO Energy. It incorporates a new feasibility algorithm that makes the CR design meet all constraints, it has been coupled to a Bourne Shell program 3 to allow the user to run the code interactively without the need for a manual, and it develops a low axial peakmore » to extend the cycle. For PECO Energy Co.`s limericks it increased the energy output by 1 to 2% over the traditional PECO Energy design. The objective of the optimization in OCTOPUS-3 is to approximate a very low axial peaked target power distribution while maintaining criticality, keeping the nodal and assembly peaks below the allowed maximum, and meeting the other constraints. The user-specified input for each exposure point includes: CR groups allowed-to-move, target k{sub eff}, and amount of core flow. The OCTOPUS-3 code uses the CR pattern from the previous step as the initial guess unless indicated otherwise.« less

Irradiation-induced sensitization and stress corrosion cracking of Type 304 stainless steel core-internal components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H.M.; Ruther, W.E.; Sanecki, J.E.

1991-08-01

High- and commercial-purity heats of Type 304 stainless steel, obtained from neutron absorber tubes after irradiation to fluence levels of up to 2 {times} 10{sup 21} n{center dot}cm{sup {minus}2} (E > 1 MeV) in two boiling water reactors, were examined by Auger electron spectroscopy to characterize irradiation-induced grain- boundary segregation and depletion of alloying and impurity elements. Segregation of Si, P, Ni, and an unidentified element or compound that gives rise to an Auger energy peak at 59 eV was observed in the commercial-purity heat. Such segregation was negligible in high-purity material, except for Ni. No evidence of S segregationmore » was observed in either material. Cr depletion was more pronounced in the high-purity material than in the commercial-purity material. These observations suggest a synergism between the significant level of impurities and Cr depletion in the commercial-purity heat. In the absence of such synergism, Cr depletion appears more pronounced in the high-purity heat. Initial results of constant-extension-rate tests conducted on the two heats in air an in simulated BWR water were correlated with the results from analysis by Auger electron spectroscopy. 15 refs., 10 figs.« less

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

NASA Astrophysics Data System (ADS)

Rochman, Dimitri A.; Vasiliev, Alexander; Dokhane, Abdelhamid; Ferroukhi, Hakim

2018-05-01

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

[The legal and ethical aspects of nerve tissue transplantation].

PubMed

Sramka, M; Rattaj, M

1992-01-01

The authors have specified the following criteria for the withdrawal of embryonal tissue at their department: 1) only tissue from dead fetus is allowed to be used in neurotransplantation; 2) embryonal tissue is to be obtained after spontaneous abortions from volunteers or from women asking for artificial abortion; 3) the women should be informed about the curative purposes of embryonal tissue voluntary donorship and they must give a written consent; 4) decision on abortion should be separated from the use of embryonal tissue; 5) women should not know recipients; no payments should be made for tissue; 6) the donor is not permitted to impregnate in order to use embryos for research or clinical purposes; 7) sampling of BWR, HBsAG, anti-HIV, cytomegalovirus, herpes I and II is to be made for serologic examinations and that from the cervix for cultivation and sensitivity, as well as ultrasound verification of a germinal age is done in potential donors; 8) consent should be signed to embryonal brain transplantation by recipient or his legitimate deputy if the recipient is certifiable. The above criteria should protect both the donor and the recipient. The use of embryonal tissue cultures seems to be promising. In addition to legal and ethic problems, immunological problems and problems concerning the aseptic withdrawal of embryonal tissue are falling off.

Development of Self Fire Retardant Melamine-Animal Glue Formaldehyde (MGF) Resin for the Manufacture of BWR Ply Board

NASA Astrophysics Data System (ADS)

Khatua, Pijus Kanti; Dubey, Rajib Kumar; Roymahapatra, Gourisankar; Mishra, Anjan; Shahoo, Shadhu Charan; Kalawate, Aparna

2017-10-01

Wood is one of the most sustainable, naturally growing materials that consist mainly of combustible organic carbon compounds. Since plywood are widely used nowadays especially in buildings, furniture and cabinets. Too often the fire behavior of ply-board may be viewed as a drawback. Amino-plastic based thermosetting resin adhesives are the important and most widely used in the plywood panel industries. The fire retardant property of wood panel products by adding animal glue as an additive in the form of MGF resin and used as substitute of melamine for manufacture of plywood. Environment concerns and higher cost of petroleum based resins have resulted in the development of technologies to replace melamine partially by biomaterials for the manufacturing of resin adhesive. Natural bio-based materials such as tannin, CNSL (cardanol), lignin, soya etc. are used as partial substitution of melamine. This article presents the development of melamine-animal glue formaldehyde resin as plywood binder. About 30 % melamine was substituted by animal glue and optimized. The different physico-mechanical and fire retardant property properties tested as per IS: 1734-1983 and IS: 5509-2000 respectively are quite satisfactory. The production of adhesive from melamine with compatible natural proteinous material is cost effective, eco-friendly and enhance the fire retardant property.

78 FR 25972 - Establishment of the Response Systems to Adult Sexual Crimes Panel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-03

.... 920 (Article 120 of the UCMJ). The Response Systems Panel's review shall include the following: a... recommendations and advice to the Response Systems Panel for full deliberation and discussion. Subcommittees, task... DEPARTMENT OF DEFENSE Office of the Secretary Establishment of the Response Systems to Adult...

Benefits of Using Online Student Response Systems in Japanese EFL Classrooms

ERIC Educational Resources Information Center

Mork, Cathrine-Mette

2014-01-01

Online student response systems (OSRSs) are fast replacing classroom response systems (CRSs), also known as personal or audience response systems or "clickers". OSRSs can more easily be implemented in the classroom because they are web-based and allow students to use any browser and device to do the "clicking" required to…

An analytical study and wind tunnel tests of an aeromechanical gust-alleviation system for a light airplane

NASA Technical Reports Server (NTRS)

Stewart, E. C.

1976-01-01

The results of an analytical study of a system using stability derivatives determined in static wind tunnel tests of a 1/6 scale model of a popular, high wing, light airplane equipped with the gust alleviation system are reported. The longitudinal short period mode dynamics of the system are analyzed, and include the following: (1) root loci, (2) airplane frequency responses to vertical gusts, (3) power spectra of the airplane responses in a gust spectrum, (4) time history responses to vertical gusts, and (5) handling characteristics. The system reduces the airplane's normal acceleration response to vertical gusts while simultaneously increasing the pitching response and reducing the damping of the longitudinal short period mode. The normal acceleration response can be minimized by using the proper amount of static alleviation and a fast response system with a moderate amount of damping. The addition of a flap elevator interconnect or a pitch damper system further increases the alleviation while moderating the simultaneous increase in pitching response. The system provides direct lift control and may reduce the stick fixed longitudinal static stability.

A Lattice Boltzmann Framework for the simulation of boiling hydrodynamics in BWRs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jain, P. K.; Tentner, A.; Uddin, R.

2008-01-01

Multi phase and multi component flows are ubiquitous in nature as well as in many man-made processes. A specific example is the Boiling Water Reactor (BWR) core, in which the coolant enters the core as liquid, undergoes a phase change as it traverses the core and exits as a high quality two-phase mixture. Two-phase flows in BWRs typically manifest a wide variety of geometrical patterns of the co-existing phases depending on the local system conditions. Modeling of such flows currently relies on empirical correlations (for example, in the simulation of bubble nucleation, bubble growth and coalescence, and inter-phase surface topologymore » transitions) that hinder the accurate simulation of two-phase phenomena using Computational Fluid Dynamics (CFD) approaches. The Lattice Boltzmann Method (LBM) is in rapid development as a modeling tool to understand these macro-phenomena by coupling them with their underlying micro-dynamics. This paper presents a consistent LBM formulation for the simulation of a two-phase water-steam system. Results of initial model validation in a range of thermodynamic conditions typical for BWRs are also shown. The interface between the two coexisting phases is captured from the dynamics of the model itself, i.e., no interface tracking is needed. The model is based on the Peng-Robinson (P-R) non-ideal equation of state and can quantitatively approximate the phase-coexistence curve for water at different temperatures ranging from 125 to 325 oC. Consequently, coexisting phases with large density ratios (up to {approx}1000) may be simulated. Two-phase models in the 200-300 C temperature range are of significant importance to nuclear engineers since most BWRs operate under similar thermodynamic conditions. Simulation of bubbles and droplets in a gravity-free environment of the corresponding coexisting phase until steady state is reached satisfies Laplace law at different temperatures and thus, yield the surface tension of the fluid. Comparing the LBM surface tension thus calculated using the LBM to the corresponding experimental values for water, the LBM lattice unit (lu) can be scaled to the physical units. Using this approach, spatial scaling of the LBM emerges from the model itself and is not imposed externally.« less

Transformative Learning: Patterns of Psychophysiologic Response and Technology-Enabled Learning and Intervention Systems

DTIC Science & Technology

2008-09-01

Psychophysiologic Response and Technology -Enabled Learning and Intervention Systems PRINCIPAL INVESTIGATOR: Leigh W. Jerome, Ph.D...NUMBER Transformative Learning : Patterns of Psychophysiologic Response and Technology - Enabled Learning and Intervention Systems 5b. GRANT NUMBER...project entitled “Transformative Learning : Patterns of Psychophysiologic Response in Technology Enabled Learning and Intervention Systems.” The

78 FR 63454 - Response Systems to Adult Sexual Assault Crimes Panel; Notice of Federal Advisory Committee Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... DEPARTMENT OF DEFENSE Office of the Secretary Response Systems to Adult Sexual Assault Crimes... Advisory Committee meeting of the Response Systems to Adult Sexual Assault Crimes Panel. DATES: A meeting of the Response Systems to Adult Sexual Assault Crimes Panel (``the Panel'') will be held November 7...

FY2012 summary of tasks completed on PROTEUS-thermal work.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Aviation System Analysis Capability Quick Response System Report for Fiscal Year 1998

NASA Technical Reports Server (NTRS)

Ege, Russell; Villani, James; Ritter, Paul

1999-01-01

This document presents the additions and modifications made to the Quick Response System (QRS) in FY 1998 in support of the ASAC QRS development effort. this Document builds upon the Aviation System Analysis Capability Quick Responses System Report for Fiscal Year 1997.

Method of calibrating a fluid-level measurement system

NASA Technical Reports Server (NTRS)

Woodard, Stanley E. (Inventor); Taylor, Bryant D. (Inventor)

2010-01-01

A method of calibrating a fluid-level measurement system is provided. A first response of the system is recorded when the system's sensor(s) is (are) not in contact with a fluid of interest. A second response of the system is recorded when the system's sensor(s) is (are) fully immersed in the fluid of interest. Using the first and second responses, a plurality of expected responses of the system's sensor(s) is (are) generated for a corresponding plurality of levels of immersion of the sensor(s) in the fluid of interest.

48 CFR 234.003 - Responsibilities.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Responsibilities. 234.003 Section 234.003 Federal Acquisition Regulations System DEFENSE ACQUISITION REGULATIONS SYSTEM, DEPARTMENT OF DEFENSE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION 234.003 Responsibilities. DoDD...

Rapid response systems.

PubMed

Lyons, Patrick G; Edelson, Dana P; Churpek, Matthew M

2018-07-01

Rapid response systems are commonly employed by hospitals to identify and respond to deteriorating patients outside of the intensive care unit. Controversy exists about the benefits of rapid response systems. We aimed to review the current state of the rapid response literature, including evolving aspects of afferent (risk detection) and efferent (intervention) arms, outcome measurement, process improvement, and implementation. Articles written in English and published in PubMed. Rapid response systems are heterogeneous, with important differences among afferent and efferent arms. Clinically meaningful outcomes may include unexpected mortality, in-hospital cardiac arrest, length of stay, cost, and processes of care at end of life. Both positive and negative interventional studies have been published, although the two largest randomized trials involving rapid response systems - the Medical Early Response and Intervention Trial (MERIT) and the Effect of a Pediatric Early Warning System on All-Cause Mortality in Hospitalized Pediatric Patients (EPOCH) trial - did not find a mortality benefit with these systems, albeit with important limitations. Advances in monitoring technologies, risk assessment strategies, and behavioral ergonomics may offer opportunities for improvement. Rapid responses may improve some meaningful outcomes, although these findings remain controversial. These systems may also improve care for patients at the end of life. Rapid response systems are expected to continue evolving with novel developments in monitoring technologies, risk prediction informatics, and work in human factors. Copyright © 2018 Elsevier B.V. All rights reserved.

Community Environmental Response Facilitation Act (CERFA) Report, Sacramento Army Depot, Sacramento, California

DTIC Science & Technology

1994-04-01

Response, Compensation, and Liability Information System CERFA Community Environmental Response Facilitation Act CORTESE State-designated hazardous...waste cleanup sites DESCOM U.S. Army Depot Systems Command DTSC Department of Toxic Substance Control EMD Environmental Management Division EPA U.S...Environmental Protection Agency ERNS Emergency Response Notification system FFA Federal Facility Agreement FINDS Facility index system HWCSA Hazardous

Improved Academic Performance and Student Perceptions of Learning through Use of a Cell Phone-Based Personal Response System

ERIC Educational Resources Information Center

Ma, Sihui; Steger, Daniel G.; Doolittle, Peter E.; Stewart, Amanda C.

2018-01-01

Personal response systems, such as clickers, have been widely used to improve the effectiveness of teaching in various classroom settings. Although hand-held clicker response systems have been the subject of multiple prior studies, few studies have focused on the use of cell phone-based personal response system (CPPRS) specifically. This study…

Stimulus Responsive Nanoparticles

NASA Technical Reports Server (NTRS)

Sierros, Konstantinos A. (Inventor); Cairns, Darran Robert (Inventor); Huebsch, Wade W. (Inventor); Shafran, Matthew S. (Inventor)

2017-01-01

Disclosed are various embodiments of methods and systems related to stimulus responsive nanoparticles. In one embodiment including a stimulus responsive nanoparticle system, the system includes a first electrode, a second electrode, and a plurality of elongated electro-responsive nanoparticles dispersed between the first and second electrodes, the plurality of electro-responsive nanorods configured to respond to an electric field established between the first and second electrodes.

Stimulus responsive nanoparticles

NASA Technical Reports Server (NTRS)

Cairns, Darren Robert (Inventor); Shafran, Matthew S. (Inventor); Huebsch, Wade W. (Inventor); Sierros, Konstantinos A. (Inventor)

2013-01-01

Disclosed are various embodiments of methods and systems related to stimulus responsive nanoparticles. In one embodiment includes a stimulus responsive nanoparticle system, the system includes a first electrode, a second electrode, and a plurality of elongated electro-responsive nanoparticles dispersed between the first and second electrodes, the plurality of electro-responsive nanorods configured to respond to an electric field established between the first and second electrodes.

Stimulus Responsive Nanoparticles

NASA Technical Reports Server (NTRS)

Cairns, Darran Robert (Inventor); Huebsch, Wade W. (Inventor); Sierros, Konstantinos A. (Inventor); Shafran, Matthew S. (Inventor)

2015-01-01

Disclosed are various embodiments of methods and systems related to stimulus responsive nanoparticles. In one embodiment includes a stimulus responsive nanoparticle system, the system includes a first electrode, a second electrode, and a plurality of elongated electro-responsive nanoparticles dispersed between the first and second electrodes, the plurality of electro-responsive nanorods configured to respond to an electric field established between the first and second electrodes.

Wireless Damage Location Sensing System

NASA Technical Reports Server (NTRS)

Woodard, Stanley E. (Inventor); Taylor, Bryant Douglas (Inventor)

2012-01-01

A wireless damage location sensing system uses a geometric-patterned wireless sensor that resonates in the presence of a time-varying magnetic field to generate a harmonic response that will experience a change when the sensor experiences a change in its geometric pattern. The sensing system also includes a magnetic field response recorder for wirelessly transmitting the time-varying magnetic field and for wirelessly detecting the harmonic response. The sensing system compares the actual harmonic response to a plurality of predetermined harmonic responses. Each predetermined harmonic response is associated with a severing of the sensor at a corresponding known location thereof so that a match between the actual harmonic response and one of the predetermined harmonic responses defines the known location of the severing that is associated therewith.

Method and system for an on-chip AC self-test controller

DOEpatents

Flanagan, John D [Rhinebeck, NY; Herring, Jay R [Poughkeepsie, NY; Lo, Tin-Chee [Fishkill, NY

2008-09-30

A method and system for performing AC self-test on an integrated circuit that includes a system clock for use during normal operation are provided. The method includes applying a long data capture pulse to a first test register in response to the system clock, applying an at speed data launch pulse to the first test register in response to the system clock, inputting the data from the first register to a logic path in response to applying the at speed data launch pulse to the first test register, applying an at speed data capture pulse to a second test register in response to the system clock, inputting the logic path output to the second test register in response to applying the at speed data capture pulse to the second test register, and applying a long data launch pulse to the second test register in response to the system clock.

75 FR 30845 - Request Voucher for Grant Payment and Line of Credit Control System (LOCCS) Voice Response System...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-02

... request vouchers for distribution of grant funds using the automated Voice Response System (VRS). An... Payment and Line of Credit Control System (LOCCS) Voice Response System Access Authorization AGENCY... subject proposal. Payment request vouchers for distribution of grant funds using the automated Voice...

77 FR 17462 - Notice of Submission for OMB Review; Institute of Education Sciences; Quick Response Information...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-26

... DEPARTMENT OF EDUCATION Notice of Submission for OMB Review; Institute of Education Sciences; Quick Response Information System (QRIS) 2012-2015 System Clearance SUMMARY: The National Center for Education Statistics (NCES) Quick Response Information System (QRIS) consists of the Fast Response Survey...

40 CFR 281.37 - Financial responsibility for UST systems containing petroleum.

Code of Federal Regulations, 2011 CFR

2011-07-01

... systems containing petroleum. 281.37 Section 281.37 Protection of Environment ENVIRONMENTAL PROTECTION... for No-Less-Stringent § 281.37 Financial responsibility for UST systems containing petroleum. (a) In... UST systems containing petroleum, the state requirements for financial responsibility for petroleum...

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

A HO-IRT Based Diagnostic Assessment System with Constructed Response Items

ERIC Educational Resources Information Center

Yang, Chih-Wei; Kuo, Bor-Chen; Liao, Chen-Huei

2011-01-01

The aim of the present study was to develop an on-line assessment system with constructed response items in the context of elementary mathematics curriculum. The system recorded the problem solving process of constructed response items and transfered the process to response codes for further analyses. An inference mechanism based on artificial…

Implementing a vector surveillance-response system for chagas disease control: a 4-year field trial in Nicaragua.

PubMed

Yoshioka, Kota; Tercero, Doribel; Pérez, Byron; Nakamura, Jiro; Pérez, Lenin

2017-03-06

Chagas disease is one of the neglected tropical diseases (NTDs). International goals for its control involve elimination of vector-borne transmission. Central American countries face challenges in establishing sustainable vector control programmes, since the main vector, Triatoma dimidiata, cannot be eliminated. In 2012, the Ministry of Health in Nicaragua started a field test of a vector surveillance-response system to control domestic vector infestation. This paper reports the main findings from this pilot study. This study was carried out from 2012 to 2015 in the Municipality of Totogalpa. The Japan International Cooperation Agency provided technical cooperation in designing and monitoring the surveillance-response system until 2014. This system involved 1) vector reports by householders to health facilities, 2) data analysis and planning of responses at the municipal health centre and 3) house visits or insecticide spraying by health personnel as a response. We registered all vector reports and responses in a digital database. The collected data were used to describe and analyse the system performance in terms of amount of vector reports as well as rates and timeliness of responses. During the study period, T. dimidiata was reported 396 times. Spatiotemporal analysis identified some high-risk clusters. All houses reported to be infested were visited by health personnel in 2013 and this response rate dropped to 39% in 2015. Rates of insecticide spraying rose above 80% in 2013 but no spraying was carried out in the following 2 years. The timeliness of house visits improved significantly after the responsibility was transferred from a vector control technician to primary health care staff. We argue that the proposed vector surveillance-response system is workable within the resource-constrained health system in Nicaragua. Integration to the primary health care services was a key to improve the system performance. Continual efforts are necessary to keep adapting the surveillance-response system to the dynamic health systems. We also discuss that the goal of eliminating vector-borne transmission remains unachievable. This paper provides lessons not only for Chagas disease control in Central America, but also for control efforts for other NTDs that need a sustainable surveillance-response system to support elimination.

Experiences of faculty and students using an audience response system in the classroom.

PubMed

Thomas, Christine M; Monturo, Cheryl; Conroy, Katherine

2011-07-01

The advent of innovative technologies, such as the audience response system, provides an opportunity to engage students and enhance learning. Based on their experiences, three nursing faculty evaluated the use of an audience response system in four distinct nursing courses through the use of informal survey results. When using the audience response system, the faculty experienced an increased perception of student attentiveness and engagement, high level of class attendance, and enhanced learning. Faculty feelings were mixed concerning the burden in adapting to increased classroom time and increased preparation time. Students' perception of the value of audience response system use was mostly positive, except when responses were included as part of the grade. The majority of the students indicated that use of the audience response system enhanced learning and was a helpful learning method when used with NCLEX-style questions. Overall, faculty believed that the benefits of student engagement and enhanced learning outweighed the burdens of incorporating this new technology in the classroom.

Experimental investigation of stress-inducing properties of system response times.

PubMed

Kuhmann, W

1989-03-01

System response times are regarded as a major stressor in human-computer interaction. In two earlier studies short (2s) and long (8s) response times were found to have differential effects on psychological, subjective, and performance variables, but results did not favour either response time. Therefore, in another laboratory study with 48 subjects in four independent groups working at a stimulated computer workplace, system response times of 2, 4, 6 and 8s were introduced in the same error detection task as used before, during 3 training and 5 working trials of 20 min each, and the same physiological, subjective and performance measures were obtained. There were no global effects on physiological variables, possibly due to low work load as a result of missing time pressure, but subjective and performance variables clearly favoured the longer system response times. When task periods and response time-periods were analysed separately, a shift of electrodermal activity could be observed from task- to response time-periods during the course of trials in the 8s condition. This did not appear in any other condition, which points to psychophysiological excitement that develops when system response times are too long, thus providing support for the concept of optimal system response times.

Evaluating the Reliability of Emergency Response Systems for Large-Scale Incident Operations

PubMed Central

Jackson, Brian A.; Faith, Kay Sullivan; Willis, Henry H.

2012-01-01

Abstract The ability to measure emergency preparedness—to predict the likely performance of emergency response systems in future events—is critical for policy analysis in homeland security. Yet it remains difficult to know how prepared a response system is to deal with large-scale incidents, whether it be a natural disaster, terrorist attack, or industrial or transportation accident. This research draws on the fields of systems analysis and engineering to apply the concept of system reliability to the evaluation of emergency response systems. The authors describe a method for modeling an emergency response system; identifying how individual parts of the system might fail; and assessing the likelihood of each failure and the severity of its effects on the overall response effort. The authors walk the reader through two applications of this method: a simplified example in which responders must deliver medical treatment to a certain number of people in a specified time window, and a more complex scenario involving the release of chlorine gas. The authors also describe an exploratory analysis in which they parsed a set of after-action reports describing real-world incidents, to demonstrate how this method can be used to quantitatively analyze data on past response performance. The authors conclude with a discussion of how this method of measuring emergency response system reliability could inform policy discussion of emergency preparedness, how system reliability might be improved, and the costs of doing so. PMID:28083267

Interactive Response Systems (IRS) Socrative Application Sample

ERIC Educational Resources Information Center

Aslan, Bilge; Seker, Hasan

2017-01-01

In globally developing education system, technology has made instructional improved in many ways. One of these improvements is the Interactive Response Systems (IRS) that are applied in classroom activities. Therefore, it is "smart" to focus on interactive response systems in learning environment. This study was conducted aiming to focus…

41 CFR 102-118.350 - Does establishing a prepayment audit system or program change the responsibilities of the...

Code of Federal Regulations, 2010 CFR

2010-07-01

... prepayment audit system or program change the responsibilities of the certifying officers? 102-118.350... System (Continued) FEDERAL MANAGEMENT REGULATION TRANSPORTATION 118-TRANSPORTATION PAYMENT AND AUDIT... Does establishing a prepayment audit system or program change the responsibilities of the certifying...

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Demand Response For Power System Reliability: FAQ

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirby, Brendan J

2006-12-01

Demand response is the most underutilized power system reliability resource in North America. Technological advances now make it possible to tap this resource to both reduce costs and improve. Misconceptions concerning response capabilities tend to force loads to provide responses that they are less able to provide and often prohibit them from providing the most valuable reliability services. Fortunately this is beginning to change with some ISOs making more extensive use of load response. This report is structured as a series of short questions and answers that address load response capabilities and power system reliability needs. Its objective is tomore » further the use of responsive load as a bulk power system reliability resource in providing the fastest and most valuable ancillary services.« less

Method and system for an on-chip AC self-test controller

DOEpatents

Flanagan, John D.; Herring, Jay R.; Lo, Tin-Chee

2006-06-06

A method for performing AC self-test on an integrated circuit, including a system clock for use during normal operation. The method includes applying a long data capture pulse to a first test register in response to the system clock, and further applying at an speed data launch pulse to the first test register in response to the system clock. Inputting the data from the first register to a logic path in response to applying the at speed data launch pulse to the first test register. Applying at speed data capture pulse to a second test register in response to the system clock. Inputting the output from the logic path to the second test register in response to applying the at speed data capture pulse to the second register. Applying a long data launch pulse to the second test register in response to the system clock.

Control of Initialized Fractional-Order Systems. Revised

NASA Technical Reports Server (NTRS)

Hartley, Tom T.; Lorenzo, Carl F.

2002-01-01

Due to the importance of historical effects in fractional-order systems, this paper presents a general fractional-order control theory that includes the time-varying initialization response. Previous studies have not properly accounted for these historical effects. The initialization response, along with the forced response, for fractional-order systems is determined. Stability properties of fractional-order systems are presented in the complex w-plane, which is a transformation of the s-plane. Time responses are discussed with respect to pole positions in the complex w-plane and frequency response behavior is included. A fractional-order vector space representation, which is a generalization of the state space concept, is presented including the initialization response. Control methods for vector representations of initialized fractional-order systems are shown. Nyquist, root-locus, and other input-output control methods are adapted to the control of fractional-order systems. Finally, the fractional-order differintegral is generalized to continuous order-distributions that have the possibility of including a continuum of fractional orders in a system element.

Non-linear dynamic compensation system

NASA Technical Reports Server (NTRS)

Lin, Yu-Hwan (Inventor); Lurie, Boris J. (Inventor)

1992-01-01

A non-linear dynamic compensation subsystem is added in the feedback loop of a high precision optical mirror positioning control system to smoothly alter the control system response bandwidth from a relatively wide response bandwidth optimized for speed of control system response to a bandwidth sufficiently narrow to reduce position errors resulting from the quantization noise inherent in the inductosyn used to measure mirror position. The non-linear dynamic compensation system includes a limiter for limiting the error signal within preselected limits, a compensator for modifying the limiter output to achieve the reduced bandwidth response, and an adder for combining the modified error signal with the difference between the limited and unlimited error signals. The adder output is applied to control system motor so that the system response is optimized for accuracy when the error signal is within the preselected limits, optimized for speed of response when the error signal is substantially beyond the preselected limits and smoothly varied therebetween as the error signal approaches the preselected limits.

Control of Initialized Fractional-Order Systems

NASA Technical Reports Server (NTRS)

Hartly, Tom T.; Lorenzo, Carl F.

2002-01-01

Due to the importance of historical effects in fractional-order systems, this paper presents a general fractional-order control theory that includes the time-varying initialization response. Previous studies have not properly accounted for these historical effects. The initialization response, along with the forced response, for fractional-order systems is determined. Stability properties of fractional-order systems are presented in the complex Airplane, which is a transformation of the s-plane. Time responses are discussed with respect to pole positions in the complex Airplane and frequency response behavior is included. A fractional-order vector space representation, which is a generalization of the state space concept, is presented including the initialization response. Control methods for vector representations of initialized fractional-order systems are shown. Nyquist, root-locus, and other input-output control methods are adapted to the control of fractional-order systems. Finally, the fractional-order differintegral is generalized to continuous order-distributions that have the possibility of including a continuum of fractional orders in a system element.

Optimal Linear Responses for Markov Chains and Stochastically Perturbed Dynamical Systems

NASA Astrophysics Data System (ADS)

Antown, Fadi; Dragičević, Davor; Froyland, Gary

2018-03-01

The linear response of a dynamical system refers to changes to properties of the system when small external perturbations are applied. We consider the little-studied question of selecting an optimal perturbation so as to (i) maximise the linear response of the equilibrium distribution of the system, (ii) maximise the linear response of the expectation of a specified observable, and (iii) maximise the linear response of the rate of convergence of the system to the equilibrium distribution. We also consider the inhomogeneous, sequential, or time-dependent situation where the governing dynamics is not stationary and one wishes to select a sequence of small perturbations so as to maximise the overall linear response at some terminal time. We develop the theory for finite-state Markov chains, provide explicit solutions for some illustrative examples, and numerically apply our theory to stochastically perturbed dynamical systems, where the Markov chain is replaced by a matrix representation of an approximate annealed transfer operator for the random dynamical system.

[Characteristics of the sympathoadrenal system response to psychoemotional stress under hypoxic conditions in aged people with physiological and accelerated aging of the respiratory system].

PubMed

Asanov, E O; Os'mak, Ie D; Kuz'mins'ka, L A

2013-01-01

The peculiarities of the response of the sympathoadrenal system to psychoemotional and hypoxic stress in healthy young people and in aged people with physiological and accelerated aging of respiratory system were studied. It was shown that in aging a more pronounced response of the sympathoadrenal system to psychoemotional stress. At the same time, elderly people with different types of aging of the respiratory system did not demonstrate a difference in the response of the sympathoadrenal system to psychoemotional stress. Unlike in young people, in aged people, combination of psychoemotional and hypoxic stresses resulted in further activation of the sympathoadrenal system. The reaction of the sympathoadrenal system was more expressed in elderly people with accelerated ageing of the respiratory system.

Pilot study to examine use of transverse vibration nondestructive evaluation for assessing floor systems

Treesearch

Zhiyong Cai; Robert J. Ross; Michael O. Hunt; Lawrence A. Soltis

2002-01-01

Evaluation of existing timber structures requires procedures to evaluate in situ structural members and components. This report evaluates the transverse vibration response of laboratory-built floor systems with new and salvaged joists. The objectives were to 1) compare floor system response to individual member response; 2) examine response sensitivity to location of...

49 CFR 37.77 - Purchase or lease of new non-rail vehicles by public entities operating a demand responsive...

Code of Federal Regulations, 2014 CFR

2014-10-01

... public entities operating a demand responsive system for the general public. 37.77 Section 37.77...-rail vehicles by public entities operating a demand responsive system for the general public. (a) Except as provided in this section, a public entity operating a demand responsive system for the general...

49 CFR 37.77 - Purchase or lease of new non-rail vehicles by public entities operating a demand responsive...

Code of Federal Regulations, 2010 CFR

2010-10-01

... public entities operating a demand responsive system for the general public. 37.77 Section 37.77...-rail vehicles by public entities operating a demand responsive system for the general public. (a) Except as provided in this section, a public entity operating a demand responsive system for the general...

49 CFR 37.77 - Purchase or lease of new non-rail vehicles by public entities operating a demand responsive...

Code of Federal Regulations, 2011 CFR

2011-10-01

... public entities operating a demand responsive system for the general public. 37.77 Section 37.77...-rail vehicles by public entities operating a demand responsive system for the general public. (a) Except as provided in this section, a public entity operating a demand responsive system for the general...

49 CFR 37.77 - Purchase or lease of new non-rail vehicles by public entities operating a demand responsive...

Code of Federal Regulations, 2012 CFR

2012-10-01

... public entities operating a demand responsive system for the general public. 37.77 Section 37.77...-rail vehicles by public entities operating a demand responsive system for the general public. (a) Except as provided in this section, a public entity operating a demand responsive system for the general...

After-action review of the 2009-10 H1N1 Influenza Outbreak Response: Ohio's Public Health System's performance.

PubMed

Mase, William A; Bickford, Beth; Thomas, Casey L; Jones, Shamika D; Bisesi, Michael

In early 2009, H1N1 influenza was identified within the human population. Centers for Disease Control and Prevention (CDC) officials responded with focused assessment, policy development, and assurances. The response was mobilized through efforts including procurement of adequate vaccine supply, local area span of control, materials acquisition, and facilities and resource identification. Qualitative evaluation of the assurance functions specific to the system's ability to assure safe and healthy conditions are reported. The methodology mirrors the Homeland Security Exercise and Evaluation Program used to assess system capability. Findings demonstrate the effectiveness of community responsive disease prevention efforts in partnership with the public health systems mission to unify traditional public sector systems, for-profit systems, and local area systems was accomplished. As a result of this response pharmaceutical industries, healthcare providers, healthcare agencies, police/safety, colleges, and health and human service agencies were united. Findings demonstrate the effectiveness of community response strategies utilizing feedback from system stakeholders. After-action review processes are critical in all-hazards preparedness. This analysis of local health district response to the H1N1 influenza outbreak informs future public health service delivery. Results provide a synthesis of local health department's emergency response strategies, challenges encountered, and future-focused emergency response strategy implementation. A synthesis is provided as to policy and practice developments which have emerged over the past seven years with regard to lessons learned from the 2009-10 H1N1 influenza outbreak and response.

Evaluation of Mucosal and Systemic Immune Responses Elicited by GPI-0100- Adjuvanted Influenza Vaccine Delivered by Different Immunization Strategies

PubMed Central

Liu, Heng; Patil, Harshad P.; de Vries-Idema, Jacqueline; Wilschut, Jan; Huckriede, Anke

2013-01-01

Vaccines for protection against respiratory infections should optimally induce a mucosal immune response in the respiratory tract in addition to a systemic immune response. However, current parenteral immunization modalities generally fail to induce mucosal immunity, while mucosal vaccine delivery often results in poor systemic immunity. In order to find an immunization strategy which satisfies the need for induction of both mucosal and systemic immunity, we compared local and systemic immune responses elicited by two mucosal immunizations, given either by the intranasal (IN) or the intrapulmonary (IPL) route, with responses elicited by a mucosal prime followed by a systemic boost immunization. The study was conducted in BALB/c mice and the vaccine formulation was an influenza subunit vaccine supplemented with GPI-0100, a saponin-derived adjuvant. While optimal mucosal antibody titers were obtained after two intrapulmonary vaccinations, optimal systemic antibody responses were achieved by intranasal prime followed by intramuscular boost. The latter strategy also resulted in the best T cell response, yet, it was ineffective in inducing nose or lung IgA. Successful induction of secretory IgA, IgG and T cell responses was only achieved with prime-boost strategies involving intrapulmonary immunization and was optimal when both immunizations were given via the intrapulmonary route. Our results underline that immunization via the lungs is particularly effective for priming as well as boosting of local and systemic immune responses. PMID:23936066

48 CFR 2434.003 - Responsibilities.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 6 2012-10-01 2012-10-01 false Responsibilities. 2434.003 Section 2434.003 Federal Acquisition Regulations System DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GENERAL CONTRACTING REQUIREMENTS MAJOR SYSTEM ACQUISITIONS 2434.003 Responsibilities. (a) The Senior...

48 CFR 2434.003 - Responsibilities.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 6 2011-10-01 2011-10-01 false Responsibilities. 2434.003 Section 2434.003 Federal Acquisition Regulations System DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GENERAL CONTRACTING REQUIREMENTS MAJOR SYSTEM ACQUISITIONS 2434.003 Responsibilities. (a) The Senior...

48 CFR 2434.003 - Responsibilities.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 48 Federal Acquisition Regulations System 6 2013-10-01 2013-10-01 false Responsibilities. 2434.003 Section 2434.003 Federal Acquisition Regulations System DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GENERAL CONTRACTING REQUIREMENTS MAJOR SYSTEM ACQUISITIONS 2434.003 Responsibilities. (a) The Senior...

48 CFR 2434.003 - Responsibilities.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 48 Federal Acquisition Regulations System 6 2014-10-01 2014-10-01 false Responsibilities. 2434.003 Section 2434.003 Federal Acquisition Regulations System DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GENERAL CONTRACTING REQUIREMENTS MAJOR SYSTEM ACQUISITIONS 2434.003 Responsibilities. (a) The Senior...

48 CFR 2434.003 - Responsibilities.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Responsibilities. 2434.003 Section 2434.003 Federal Acquisition Regulations System DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GENERAL CONTRACTING REQUIREMENTS MAJOR SYSTEM ACQUISITIONS 2434.003 Responsibilities. (a) The Senior...

Input-output characterization of an ultrasonic testing system by digital signal analysis

NASA Technical Reports Server (NTRS)

Williams, J. H., Jr.; Lee, S. S.; Karagulle, H.

1986-01-01

Ultrasonic test system input-output characteristics were investigated by directly coupling the transmitting and receiving transducers face to face without a test specimen. Some of the fundamentals of digital signal processing were summarized. Input and output signals were digitized by using a digital oscilloscope, and the digitized data were processed in a microcomputer by using digital signal-processing techniques. The continuous-time test system was modeled as a discrete-time, linear, shift-invariant system. In estimating the unit-sample response and frequency response of the discrete-time system, it was necessary to use digital filtering to remove low-amplitude noise, which interfered with deconvolution calculations. A digital bandpass filter constructed with the assistance of a Blackman window and a rectangular time window were used. Approximations of the impulse response and the frequency response of the continuous-time test system were obtained by linearly interpolating the defining points of the unit-sample response and the frequency response of the discrete-time system. The test system behaved as a linear-phase bandpass filter in the frequency range 0.6 to 2.3 MHz. These frequencies were selected in accordance with the criterion that they were 6 dB below the maximum peak of the amplitude of the frequency response. The output of the system to various inputs was predicted and the results were compared with the corresponding measurements on the system.

Device for rapid quantification of human carotid baroreceptor-cardiac reflex responses

NASA Technical Reports Server (NTRS)

Sprenkle, J. M.; Eckberg, D. L.; Goble, R. L.; Schelhorn, J. J.; Halliday, H. C.

1986-01-01

A new device has been designed, constructed, and evaluated to characterize the human carotid baroreceptor-cardiac reflex response relation rapidly. This system was designed for study of reflex responses of astronauts before, during, and after space travel. The system comprises a new tightly sealing silicon rubber neck chamber, a stepping motor-driven electrodeposited nickel bellows pressure system, capable of delivering sequential R-wave-triggered neck chamber pressure changes between +40 and -65 mmHg, and a microprocessor-based electronics system for control of pressure steps and analysis and display of responses. This new system provokes classic sigmoid baroreceptor-cardiac reflex responses with threshold, linear, and saturation ranges in most human volunteers during one held expiration.

78 FR 70023 - Response Systems to Adult Sexual Assault Crimes Panel; Notice of Federal Advisory Committee Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-22

... developing recommendations regarding how to improve the effectiveness of such systems. The Panel is... DEPARTMENT OF DEFENSE Office of the Secretary Response Systems to Adult Sexual Assault Crimes... Advisory Committee meeting of the Response Systems to Adult Sexual Assault Crimes Panel. This meeting is...

Inverse full state hybrid projective synchronization for chaotic maps with different dimensions

NASA Astrophysics Data System (ADS)

Ouannas, Adel; Grassi, Giuseppe

2016-09-01

A new synchronization scheme for chaotic (hyperchaotic) maps with different dimensions is presented. Specifically, given a drive system map with dimension n and a response system with dimension m, the proposed approach enables each drive system state to be synchronized with a linear response combination of the response system states. The method, based on the Lyapunov stability theory and the pole placement technique, presents some useful features: (i) it enables synchronization to be achieved for both cases of n < m and n > m; (ii) it is rigorous, being based on theorems; (iii) it can be readily applied to any chaotic (hyperchaotic) maps defined to date. Finally, the capability of the approach is illustrated by synchronization examples between the two-dimensional Hénon map (as the drive system) and the three-dimensional hyperchaotic Wang map (as the response system), and the three-dimensional Hénon-like map (as the drive system) and the two-dimensional Lorenz discrete-time system (as the response system).

Step-control of electromechanical systems

DOEpatents

Lewis, Robert N.

1979-01-01

The response of an automatic control system to a general input signal is improved by applying a test input signal, observing the response to the test input signal and determining correctional constants necessary to provide a modified input signal to be added to the input to the system. A method is disclosed for determining correctional constants. The modified input signal, when applied in conjunction with an operating signal, provides a total system output exhibiting an improved response. This method is applicable to open-loop or closed-loop control systems. The method is also applicable to unstable systems, thus allowing controlled shut-down before dangerous or destructive response is achieved and to systems whose characteristics vary with time, thus resulting in improved adaptive systems.

48 CFR 1334.003 - Responsibilities.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 5 2012-10-01 2012-10-01 false Responsibilities. 1334.003 Section 1334.003 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION General 1334.003 Responsibilities. (a) The designee authorized to carry...

48 CFR 1334.003 - Responsibilities.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 48 Federal Acquisition Regulations System 5 2013-10-01 2013-10-01 false Responsibilities. 1334.003 Section 1334.003 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION General 1334.003 Responsibilities. (a) The designee authorized to carry...

48 CFR 1334.003 - Responsibilities.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Responsibilities. 1334.003 Section 1334.003 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION General 1334.003 Responsibilities. (a) The designee authorized to carry...

48 CFR 1334.003 - Responsibilities.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Responsibilities. 1334.003 Section 1334.003 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION General 1334.003 Responsibilities. (a) The designee authorized to carry...

48 CFR 1334.003 - Responsibilities.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 48 Federal Acquisition Regulations System 5 2014-10-01 2014-10-01 false Responsibilities. 1334.003 Section 1334.003 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE SPECIAL CATEGORIES OF CONTRACTING MAJOR SYSTEM ACQUISITION General 1334.003 Responsibilities. (a) The designee authorized to carry...

Healthcare system responsiveness in Jiangsu Province, China.

PubMed

Chao, Jianqian; Lu, Boyang; Zhang, Hua; Zhu, Liguo; Jin, Hui; Liu, Pei

2017-01-13

The perceived responsiveness of a healthcare system reflects its ability to satisfy reasonable expectations of the public with respect to non-medical services. Recently, there has been increasing attention paid to responsiveness in evaluating the performance of a healthcare system in a variety of service settings. However, the factors that affect the responsiveness have been inconclusive so far and measures of improved responsiveness have not always thoroughly considered the factors. The aim of this study was to evaluate both the responsiveness of the healthcare system in Jiangsu Province, China, the factors that influence responsiveness and the measures of improved responsiveness considering it, as determined by a responsiveness survey. A multistage, stratified random sampling method was used to select 1938 adult residents of Jiangsu Province in 2011. Face-to-face interviews were conducted using a self-designed questionnaire modeled on the World Health Organization proposal. The final analysis was based on 1783 (92%) valid questionnaires. Canonical correlation analysis was used to assess the factors that affect responsiveness. The average score of all responsiveness-related domains in the surveyed healthcare system was satisfactory (7.50 out of a maximum 10.0). The two highest scoring domains were dignity and confidentiality, and the two lowest scoring domains choice and prompt attention. The factors affecting responsiveness were age, regional economic development level, and geographic area (urban vs. rural). The responsiveness regarding basic amenities was rated worse by the elderly than by younger respondents. Responsiveness ranked better by those with a poorer economic status. Choice in cities was better than in rural regions. The responsiveness of the Jiangsu healthcare system was considered to be satisfactory but could be improved by offering greater choice and providing more prompt attention. Perceptions of healthcare system responsiveness differ with age, regional economic development level, and geographic area (urban vs. rural). Measures to increase the perceived level of responsiveness include better service at higher level hospitals, shorter waiting time, more hospitals in rural regions, an improved medical environment, and provision of infrastructures that makes the medical environments more comfortable.

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1992-05-12

This patent describes a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies, an improvement to a fuel bundle assembly for placement in the reactor. It comprises a fuel channel having vertically extending walls forming a continuous channel around a fuel assembly volume, the channel being open at the bottom end for engagement to a lower tie plate and open at the upper end for engagement to an upper tie plate; rods for placement within the chamber, each the rod containing fissile material for producing nuclear reaction when in the presence of sufficient moderatedmore » neutron flux; a lower tie plate for supporting the bundle of rods within the channel, the lower tie plate for supporting the bundle of rods within the channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water in the channel between the rods for the generating of steam during the nuclear reaction; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein a two phase region of the water and steam in the bundle is defined during nuclear steam generating reaction in the fuel bundle.« less

Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

Task related doses in Spanish pressurized water reactors over the period 1988-1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Donnell, P.; Labarta, T.; Amor, I.

1995-03-01

In order to evaluate in depth the collective dose trend and its correlation with the effectiveness of the practical application of the ALARA principle in Spanish nuclear facilities, and base the different policy lines to promote this criteria, the CSN has fullfilled an analysis of the task related doses data over the period 1988-1992. Previously, the CSN had required to the utilities the compilation of their refuelling outage collective dose from 1988 according with a predeterminate number of tasks, in order to have available a representative and retrospective set of data in an homogeneous way and coherent with the internationalmore » data banks on occupational exposure in NPP, as the CEC and the NEA ones. The scope of this analysis was the following: first, the collective dose summaries for outage tasks and departments for PWR and for BWR, including the minimum, maximum and average dose (and statistics data) for 18 different refuelling outage tasks and 12 personal departments for each generation of each type of rector, the task and department related collective dose trends in each plant and in each generation, and second, the dose reduction techniques having been used during that period in each plant and the relative level of adoption. In this presentation the main results and conclusions of the first part of the study are reviewed for PWR.« less

Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Shack, W. J.

2008-01-21

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1more » MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.« less

Risk-informed regulation and safety management of nuclear power plants--on the prevention of severe accidents.

PubMed

Himanen, Risto; Julin, Ari; Jänkälä, Kalle; Holmberg, Jan-Erik; Virolainen, Reino

2012-11-01

There are four operating nuclear power plant (NPP) units in Finland. The Teollisuuden Voima (TVO) power company has two 840 MWe BWR units supplied by Asea-Atom at the Olkiluoto site. The Fortum corporation (formerly IVO) has two 500 MWe VVER 440/213 units at the Loviisa site. In addition, a 1600 MWe European Pressurized Water Reactor supplied by AREVA NP (formerly the Framatome ANP--Siemens AG Consortium) is under construction at the Olkiluoto site. Recently, the Finnish Parliament ratified the government Decision in Principle that the utilities' applications to build two new NPP units are in line with the total good of the society. The Finnish utilities, Fenno power company, and TVO company are in progress of qualifying the type of the new nuclear builds. In Finland, risk-informed applications are formally integrated in the regulatory process of NPPs that are already in the early design phase and these are to run through the construction and operation phases all through the entire plant service time. A plant-specific full-scope probabilistic risk assessment (PRA) is required for each NPP. PRAs shall cover internal events, area events (fires, floods), and external events such as harsh weather conditions and seismic events in all operating modes. Special attention is devoted to the use of various risk-informed PRA applications in the licensing of Olkiluoto 3 NPP. © 2012 Society for Risk Analysis.

Design of energy storage system to improve inertial response for large scale PV generation

DOE PAGES

Wang, Xiaoyu; Yue, Meng

2016-07-01

With high-penetration levels of renewable generating sources being integrated into the existing electric power grid, conventional generators are being replaced and grid inertial response is deteriorating. This technical challenge is more severe with photovoltaic (PV) generation than with wind generation because PV generation systems cannot provide inertial response unless special countermeasures are adopted. To enhance the inertial response, this paper proposes to use battery energy storage systems (BESS) as the remediation approach to accommodate the degrading inertial response when high penetrations of PV generation are integrated into the existing power grid. A sample power system was adopted and simulated usingmore » PSS/E software. Here, impacts of different penetration levels of PV generation on the system inertial response were investigated and then BESS was incorporated to improve the frequency dynamics.« less

I Like the Way This Feels: Using Classroom Response System Technology to Enhance Tactile Learners' Introductory American Government Experience

ERIC Educational Resources Information Center

Ulbig, Stacy G.

2016-01-01

Do individual-level student learning styles affect appreciation for and benefit from the use of classroom response system technology? This research investigates the benefit of in-class electronic classroom response systems ("classroom clickers"). With these systems, students answer questions posed to them in a PowerPoint presentation…

Qualitative and quantitative outcomes of audience response systems as an educational tool in a plastic surgery residency program.

PubMed

Arneja, Jugpal S; Narasimhan, Kailash; Bouwman, David; Bridge, Patrick D

2009-12-01

In-training evaluations in graduate medical education have typically been challenging. Although the majority of standardized examination delivery methods have become computer-based, in-training examinations generally remain pencil-paper-based, if they are performed at all. Audience response systems present a novel way to stimulate and evaluate the resident-learner. The purpose of this study was to assess the outcomes of audience response systems testing as compared with traditional testing in a plastic surgery residency program. A prospective 1-year pilot study of 10 plastic surgery residents was performed using audience response systems-delivered testing for the first half of the academic year and traditional pencil-paper testing for the second half. Examination content was based on monthly "Core Quest" curriculum conferences. Quantitative outcome measures included comparison of pretest and posttest and cumulative test scores of both formats. Qualitative outcomes from the individual participants were obtained by questionnaire. When using the audience response systems format, pretest and posttest mean scores were 67.5 and 82.5 percent, respectively; using traditional pencil-paper format, scores were 56.5 percent and 79.5 percent. A comparison of the cumulative mean audience response systems score (85.0 percent) and traditional pencil-paper score (75.0 percent) revealed statistically significantly higher scores with audience response systems (p = 0.01). Qualitative outcomes revealed increased conference enthusiasm, greater enjoyment of testing, and no user difficulties with the audience response systems technology. The audience response systems modality of in-training evaluation captures participant interest and reinforces material more effectively than traditional pencil-paper testing does. The advantages include a more interactive learning environment, stimulation of class participation, immediate feedback to residents, and immediate tabulation of results for the educator. Disadvantages include start-up costs and lead-time preparation.

Drug-induced and genetic alterations in stress-responsive systems: Implications for specific addictive diseases.

PubMed

Zhou, Yan; Proudnikov, Dmitri; Yuferov, Vadim; Kreek, Mary Jeanne

2010-02-16

From the earliest work in our laboratory, we hypothesized, and with studies conducted in both clinical research and animal models, we have shown that drugs of abuse, administered or self-administered, on a chronic basis, profoundly alter stress-responsive systems. Alterations of expression of specific genes involved in stress responsivity, with increases or decreases in mRNA levels, receptor, and neuropeptide levels, and resultant changes in hormone levels, have been documented to occur after chronic intermittent exposure to heroin, morphine, other opiates, cocaine, other stimulants, and alcohol in animal models and in human molecular genetics. The best studied of the stress-responsive systems in humans and mammalian species in general is undoubtedly the HPA axis. In addition, there are stress-responsive systems in other parts in the brain itself, and some of these include components of the HPA axis, such as CRF and CRF receptors, along with POMC gene and gene products. Several other stress-responsive systems are known to influence the HPA axis, such as the vasopressin-vasopressin receptor system. Orexin-hypocretin, acting at its receptors, may effect changes which suggest that it should be properly categorized as a stress-responsive system. However, less is known about the interactions and connectivity of some of these different neuropeptide and receptor systems, and in particular, about the possible connectivity of fast-acting (e.g., glutamate and GABA) and slow-acting (including dopamine, serotonin, and norepinephrine) neurotransmitters with each of these stress-responsive components and the resultant impact, especially in the setting of chronic exposure to drugs of abuse. Several of these stress-responsive systems and components, primarily based on our laboratory-based and human molecular genetics research of addictive diseases, will be briefly discussed in this review. Copyright 2009 Elsevier B.V. All rights reserved.

Creating a Team Archive During Fast-Paced Anomaly Response Activities in Space Missions

NASA Technical Reports Server (NTRS)

Malin, Jane T.; Hicks, LaDessa; Overland, David; Thronesbery, Carroll; Christofferesen, Klaus; Chow, Renee

2002-01-01

This paper describes a Web-based system to support the temporary Anomaly Response Team formed from distributed subteams in Space Shuttle and International Space Station missions. The system was designed for easy and flexible creation of small collections of files and links associated with work on a particular anomaly. The system supports privacy and levels of formality for the subteams. First we describe the supported groups and an anomaly response scenario. Then we describe the support system prototype, the Anomaly Response Tracking and Integration System (ARTIS). Finally, we describe our evaluation approach and the results of the evaluation.

Development of an analysis for the determination of coupled helicopter rotor/control system dynamic response. Part 2: Program listing

NASA Technical Reports Server (NTRS)

Sutton, L. R.

1975-01-01

A theoretical analysis is developed for a coupled helicopter rotor system to allow determination of the loads and dynamic response behavior of helicopter rotor systems in both steady-state forward flight and maneuvers. The effects of an anisotropically supported swashplate or gyroscope control system and a deformed free wake on the rotor system dynamic response behavior are included.

Systemic acquired tolerance to virulent bacterial pathogens in tomato.

PubMed

Block, Anna; Schmelz, Eric; O'Donnell, Phillip J; Jones, Jeffrey B; Klee, Harry J

2005-07-01

Recent studies on the interactions between plants and pathogenic microorganisms indicate that the processes of disease symptom development and pathogen growth can be uncoupled. Thus, in many instances, the symptoms associated with disease represent an active host response to the presence of a pathogen. These host responses are frequently mediated by phytohormones. For example, ethylene and salicylic acid (SA) mediate symptom development but do not influence bacterial growth in the interaction between tomato (Lycopersicon esculentum) and virulent Xanthomonas campestris pv vesicatoria (Xcv). It is not apparent why extensive tissue death is integral to a defense response if it does not have the effect of limiting pathogen proliferation. One possible function for this hormone-mediated response is to induce a systemic defense response. We therefore assessed the systemic responses of tomato to Xcv. SA- and ethylene-deficient transgenic lines were used to investigate the roles of these phytohormones in systemic signaling. Virulent and avirulent Xcv did induce a systemic response as evidenced by expression of defense-associated pathogenesis-related genes in an ethylene- and SA-dependent manner. This systemic response reduced cell death but not bacterial growth during subsequent challenge with virulent Xcv. This systemic acquired tolerance (SAT) consists of reduced tissue damage in response to secondary challenge with a virulent pathogen with no effect upon pathogen growth. SAT was associated with a rapid ethylene and pathogenesis-related gene induction upon challenge. SAT was also induced by infection with Pseudomonas syringae pv tomato. These data show that SAT resembles systemic acquired resistance without inhibition of pathogen growth.

Infrasound-array-element frequency response: in-situ measurement and modeling

NASA Astrophysics Data System (ADS)

Gabrielson, T.

2011-12-01

Most array elements at the infrasound stations of the International Monitoring System use some variant of a multiple-inlet pipe system for wind-noise suppression. These pipe systems have a significant impact on the overall frequency response of the element. The spatial distribution of acoustic inlets introduces a response dependence that is a function of frequency and of vertical and horizontal arrival angle; the system of inlets, pipes, and summing junctions further shapes that response as the signal is ducted to the transducer. In-situ measurements, using a co-located reference microphone, can determine the overall frequency response and diagnose problems with the system. As of July 2011, the in-situ frequency responses for 25 individual elements at 6 operational stations (I10, I53, I55, I56, I57, and I99) have been measured. In support of these measurements, a fully thermo-viscous model for the acoustics of these multiple-inlet pipe systems has been developed. In addition to measurements at operational stations, comparative analyses have been done on experimental systems: a multiple-inlet radial-pipe system with varying inlet hole size; a one-quarter scale model of a 70-meter rosette system; and vertical directionality of a small rosette system using aircraft flyovers. [Funded by the US Army Space and Missile Defense Command

Mathematical modeling of human cardiovascular system for simulation of orthostatic response

NASA Technical Reports Server (NTRS)

Melchior, F. M.; Srinivasan, R. S.; Charles, J. B.

1992-01-01

This paper deals with the short-term response of the human cardiovascular system to orthostatic stresses in the context of developing a mathematical model of the overall system. It discusses the physiological issues involved and how these issues have been handled in published cardiovascular models for simulation of orthostatic response. Most of the models are stimulus specific with no demonstrated capability for simulating the responses to orthostatic stimuli of different types. A comprehensive model incorporating all known phenomena related to cardiovascular regulation would greatly help to interpret the various orthostatic responses of the system in a consistent manner and to understand the interactions among its elements. This paper provides a framework for future efforts in mathematical modeling of the entire cardiovascular system.

Unmanned and Unattended Response Capability for Homeland Defense

DOE Office of Scientific and Technical Information (OSTI.GOV)

BENNETT, PHIL C.

2002-11-01

An analysis was conducted of the potential for unmanned and unattended robotic technologies for forward-based, immediate response capabilities that enables access and controlled task performance. The authors analyze high-impact response scenarios in conjunction with homeland security organizations, such as the NNSA Office of Emergency Response, the FBI, the National Guard, and the Army Technical Escort Unit, to cover a range of radiological, chemical and biological threats. They conducted an analysis of the potential of forward-based, unmanned and unattended robotic technologies to accelerate and enhance emergency and crisis response by Homeland Defense organizations. Response systems concepts were developed utilizing new technologiesmore » supported by existing emerging threats base technologies to meet the defined response scenarios. These systems will pre-position robotic and remote sensing capabilities stationed close to multiple sites for immediate action. Analysis of assembled systems included experimental activities to determine potential efficacy in the response scenarios, and iteration on systems concepts and remote sensing and robotic technologies, creating new immediate response capabilities for Homeland Defense.« less

The implementation and operation of a variable-response electronic throttle control system for a TF-104G aircraft

NASA Technical Reports Server (NTRS)

Neal, Bradford; Sengupta, Upal

1989-01-01

During some flight programs, researchers have encountered problems in the throttle response characteristics of high-performance aircraft. To study and to help solve these problems, the National Aeronautics and Space Administration Ames Research Center's Dryden Flight Research Facility (Ames-Dryden) conducted a study using a TF-104G airplane modified with a variable-response electronic throttle control system. Ames-Dryden investigated the effects of different variables on engine response and handling qualities. The system provided transport delay, lead and lag filters, second-order lags, command rate and position limits, and variable gain between the pilot's throttle command and the engine fuel controller. These variables could be tested individually or in combination. Ten research flights were flown to gather data on engine response and to obtain pilot ratings of the various system configurations. The results should provide design criteria for engine-response characteristics. The variable-response throttle components and how they were installed in the TF-104G aircraft are described. How the variable-response throttle was used in flight and some of the results of using this system are discussed.

Innate immune memory in plants.

PubMed

Reimer-Michalski, Eva-Maria; Conrath, Uwe

2016-08-01

The plant innate immune system comprises local and systemic immune responses. Systemic plant immunity develops after foliar infection by microbial pathogens, upon root colonization by certain microbes, or in response to physical injury. The systemic plant immune response to localized foliar infection is associated with elevated levels of pattern-recognition receptors, accumulation of dormant signaling enzymes, and alterations in chromatin state. Together, these systemic responses provide a memory to the initial infection by priming the remote leaves for enhanced defense and immunity to reinfection. The plant innate immune system thus builds immunological memory by utilizing mechanisms and components that are similar to those employed in the trained innate immune response of jawed vertebrates. Therefore, there seems to be conservation, or convergence, in the evolution of innate immune memory in plants and vertebrates. Copyright © 2016 Elsevier Ltd. All rights reserved.

A meta-analysis of response-time tests of the sequential two-systems model of moral judgment.

PubMed

Baron, Jonathan; Gürçay, Burcu

2017-05-01

The (generalized) sequential two-system ("default interventionist") model of utilitarian moral judgment predicts that utilitarian responses often arise from a system-two correction of system-one deontological intuitions. Response-time (RT) results that seem to support this model are usually explained by the fact that low-probability responses have longer RTs. Following earlier results, we predicted response probability from each subject's tendency to make utilitarian responses (A, "Ability") and each dilemma's tendency to elicit deontological responses (D, "Difficulty"), estimated from a Rasch model. At the point where A = D, the two responses are equally likely, so probability effects cannot account for any RT differences between them. The sequential two-system model still predicts that many of the utilitarian responses made at this point will result from system-two corrections of system-one intuitions, hence should take longer. However, when A = D, RT for the two responses was the same, contradicting the sequential model. Here we report a meta-analysis of 26 data sets, which replicated the earlier results of no RT difference overall at the point where A = D. The data sets used three different kinds of moral judgment items, and the RT equality at the point where A = D held for all three. In addition, we found that RT increased with A-D. This result holds for subjects (characterized by Ability) but not for items (characterized by Difficulty). We explain the main features of this unanticipated effect, and of the main results, with a drift-diffusion model.

Recent Advances in Cyclodextrin-Based Light-Responsive Supramolecular Systems.

PubMed

Zhang, Xiaojin; Ma, Xin; Wang, Kang; Lin, Shijun; Zhu, Shitai; Dai, Yu; Xia, Fan

2018-06-01

Cyclodextrins (CDs), one of the host molecules in supramolecular chemistry, can host guest molecules to form inclusion complexes via non-covalent and reversible host-guest interactions. CD-based light-responsive supramolecular systems are typically constructed using CDs and guest molecules with light-responsive moieties, including azobenzene, arylazopyrazole, o-nitrobenzyl ester, pyrenylmethyl ester, coumarin, and anthracene. To date, numerous efforts have been reported on the topic of CD-based light-responsive supramolecular systems, but these have not yet been highlighted in a separated review. This review summarizes the efforts reported over the past ten years. The main text of this review is divided into five sections (vesicles, micelles, gels, capturers, and nanovalves) according to the formation of self-assemblies. This feature article aims to afford a comprehensive understanding of the light-responsive moieties used in the construction of CD-based light-responsive supramolecular systems and to provide a helpful guide for the further design of CD-based light-responsive supramolecular systems. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Delayed system response times affect immediate physiology and the dynamics of subsequent button press behavior.

PubMed

Kohrs, Christin; Hrabal, David; Angenstein, Nicole; Brechmann, André

2014-11-01

System response time research is an important issue in human-computer interactions. Experience with technical devices and general rules of human-human interactions determine the user's expectation, and any delay in system response time may lead to immediate physiological, emotional, and behavioral consequences. We investigated such effects on a trial-by-trial basis during a human-computer interaction by measuring changes in skin conductance (SC), heart rate (HR), and the dynamics of button press responses. We found an increase in SC and a deceleration of HR for all three delayed system response times (0.5, 1, 2 s). Moreover, the data on button press dynamics was highly informative since subjects repeated a button press with more force in response to delayed system response times. Furthermore, the button press dynamics could distinguish between correct and incorrect decisions and may thus even be used to infer the uncertainty of a user's decision. Copyright © 2014 Society for Psychophysiological Research.

High Torque-to-Inertia Servo System for Stabilizing Sensor Systems. Candidate Systems Include Missile Guidance, Surveillance, and Tracking

DTIC Science & Technology

1980-04-01

specifications ... 3-10 25. Typical isolation curve ... 3-12 26. Servo amp/motor/load frequency response (inner gimbal) ... 4-3 27. Slave loop ( open loop...slave loop ( open loop) frequency response (inner gimbal) . . . 4-4 30. Slave loop (closed loop) frequency response (inner gimbal) ... 4-5 3 . Slave...loop inner gimbal time response ... 4-5 32. Servo amp/motor/load frequency response (outer gimbal) ... 4-6 33. Slave loop ( open loop) uncompensated

Responsibility for Curriculum Evaluation in Centralized Systems.

ERIC Educational Resources Information Center

Deschamp, Phil; McGaw, Barry

1979-01-01

While responsibility for curriculum development in Australia is devolving to the local level, the state systems, in the name of accountability, are retaining responsibility for curriculum evaluation. This article examines the curriculum centralization/decentralization patterns in Australia's states, points out the paradoxes in such systems, and…

48 CFR 253.209-1 - Responsible prospective contractors.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 48 Federal Acquisition Regulations System 3 2013-10-01 2013-10-01 false Responsible prospective contractors. 253.209-1 Section 253.209-1 Federal Acquisition Regulations System DEFENSE ACQUISITION REGULATIONS SYSTEM, DEPARTMENT OF DEFENSE CLAUSES AND FORMS FORMS Prescription of Forms 253.209-1 Responsible...

48 CFR 253.209-1 - Responsible prospective contractors.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 3 2012-10-01 2012-10-01 false Responsible prospective contractors. 253.209-1 Section 253.209-1 Federal Acquisition Regulations System DEFENSE ACQUISITION REGULATIONS SYSTEM, DEPARTMENT OF DEFENSE CLAUSES AND FORMS FORMS Prescription of Forms 253.209-1 Responsible...

The influence of an audience response system on knowledge retention: an application to resident education.

PubMed

Pradhan, Archana; Sparano, Dina; Ananth, Cande V

2005-11-01

The purpose of the study was to compare delivery methods of lecture material regarding contraceptive options by either traditional or interactive lecture style with the use of an audience response system with obstetrics and gynecology residents. A prospective, randomized controlled trial that included 17 obstetrics and gynecology residents was conducted. Group differences and comparison of pre/posttest scores to evaluate efficacy of lecture styles were performed with the Student t test. Each participant completed an evaluation to assess usefulness of the audience response system. Residents who received audience response system interactive lectures showed a 21% improvement between pretest and posttest scores; residents who received the standard lecture demonstrated a 2% improvement (P = .018). The evaluation survey showed that 82% of residents thought that the audience response system was a helpful learning aid. The results of this randomized controlled trial demonstrate the effectiveness of audience response system for knowledge retention, which suggests that it may be an efficient teaching tool for residency education.

A Comprehensive Evaluation System for Military Hospitals' Response Capability to Bio-terrorism.

PubMed

Wang, Hui; Jiang, Nan; Shao, Sicong; Zheng, Tao; Sun, Jianzhong

2015-05-01

The objective of this study is to establish a comprehensive evaluation system for military hospitals' response capacity to bio-terrorism. Literature research and Delphi method were utilized to establish the comprehensive evaluation system for military hospitals' response capacity to bio-terrorism. Questionnaires were designed and used to survey the status quo of 134 military hospitals' response capability to bio-terrorism. Survey indicated that factor analysis method was suitable to for analyzing the comprehensive evaluation system for military hospitals' response capacity to bio-terrorism. The constructed evaluation system was consisted of five first-class and 16 second-class indexes. Among them, medical response factor was considered as the most important factor with weight coefficient of 0.660, followed in turn by the emergency management factor with weight coefficient of 0.109, emergency management consciousness factor with weight coefficient of 0.093, hardware support factor with weight coefficient of 0.078, and improvement factor with weight coefficient of 0.059. The constructed comprehensive assessment model and system are scientific and practical.

Meeting national response time targets for priority 1 incidents in an urban emergency medical services system in South Africa: More ambulances won't help.

PubMed

Stein, Christopher; Wallis, Lee; Adetunji, Olufemi

2015-09-19

Response time is viewed as a key performance indicator in most emergency medical services (EMS) systems. To determine the effect of increased emergency vehicle numbers on response time performance for priority 1 incidents in an urban EMS system in Cape Town, South Africa, using discrete-event computer simulation. A simulation model was created, based on input data from part of the EMS operations. Two different versions of the model were used, one with primary response vehicles and ambulances and one with only ambulances. In both cases the models were run in seven different scenarios. The first scenario used the actual number of emergency vehicles in the real system, and in each subsequent scenario vehicle numbers were increased by adding the baseline number to the cumulative total. The model using only ambulances had shorter response times and a greater number of responses meeting national response time targets than models using primary response vehicles and ambulances. In both cases an improvement in response times and the number of responses meeting national response time targets was observed with the first incremental addition of vehicles. After this the improvements rapidly diminished and eventually became negligible with each successive increase in vehicle numbers. The national response time target for urban areas was never met, even with a seven-fold increase in vehicle numbers. The addition of emergency vehicles to an urban EMS system improves response times in priority 1 incidents, but alone is not capable of the magnitude of response time improvement needed to meet the national response time targets.

MIMO system identification using frequency response data

NASA Technical Reports Server (NTRS)

Medina, Enrique A.; Irwin, R. D.; Mitchell, Jerrel R.; Bukley, Angelia P.

1992-01-01

A solution to the problem of obtaining a multi-input, multi-output statespace model of a system from its individual input/output frequency responses is presented. The Residue Identification Algorithm (RID) identifies the system poles from a transfer function model of the determinant of the frequency response data matrix. Next, the residue matrices of the modes are computed guaranteeing that each input/output frequency response is fitted in the least squares sense. Finally, a realization of the system is computed. Results of the application of RID to experimental frequency responses of a large space structure ground test facility are presented and compared to those obtained via the Eigensystem Realization Algorithm.

Coherent motion of chaotic attractors

NASA Astrophysics Data System (ADS)

Louodop, Patrick; Saha, Suman; Tchitnga, Robert; Muruganandam, Paulsamy; Dana, Syamal K.; Cerdeira, Hilda A.

2017-10-01

We report a simple model of two drive-response-type coupled chaotic oscillators, where the response system copies the nonlinearity of the driver system. It leads to a coherent motion of the trajectories of the coupled systems that establishes a constant separating distance in time between the driver and the response attractors, and their distance depends upon the initial state. The coupled system responds to external obstacles, modeled by short-duration pulses acting either on the driver or the response system, by a coherent shifting of the distance, and it is able to readjust their distance as and when necessary via mutual exchange of feedback information. We confirm these behaviors with examples of a jerk system, the paradigmatic Rössler system, a tunnel diode system and a Josephson junction-based jerk system, analytically, to an extent, and mostly numerically.

Exploring Health System Responsiveness in Ambulatory Care and Disease Management and its Relation to Other Dimensions of Health System Performance (RAC) – Study Design and Methodology

PubMed Central

Röttger, Julia; Blümel, Miriam; Engel, Susanne; Grenz-Farenholtz, Brigitte; Fuchs, Sabine; Linder, Roland; Verheyen, Frank; Busse, Reinhard

2015-01-01

Background: The responsiveness of a health system is considered to be an intrinsic goal of health systems and an essential aspect in performance assessment. Numerous studies have analysed health system responsiveness and related concepts, especially across different countries and health systems. However, fewer studies have applied the concept for the evaluation of specific healthcare delivery structures and thoroughly analysed its determinants within one country. The aims of this study are to assess the level of perceived health system responsiveness to patients with chronic diseases in ambulatory care in Germany and to analyse the determinants of health system responsiveness as well as its distribution across different population groups. Methods and Analysis: The target population consists of chronically ill people in Germany, with a focus on patients suffering from type 2 diabetes and/or from coronary heart disease (CHD). Data comes from two different sources: (i) cross-sectional survey data from a postal survey and (ii) claims data from a German sickness fund. Data from both sources will be linked at an individual-level. The postal survey has the purpose of measuring perceived health system responsiveness, health related quality of life, experiences with disease management programmes (DMPs) and (subjective) socioeconomic background. The claims data consists of information on (co)morbidities, service utilization, enrolment within a DMP and sociodemographic characteristics, including the type of residential area. Discussion: RAC is one of the first projects linking survey data on health system responsiveness at individual level with claims data. With this unique database, it will be possible to comprehensively analyse determinants of health system responsiveness and its relation to other aspects of health system performance assessment. The results of the project will allow German health system decision-makers to assess the performance of nonclinical aspects of healthcare delivery and their determinants in two important areas of health policy: in ambulatory and chronic disease care. PMID:26188807

Exploring Health System Responsiveness in Ambulatory Care and Disease Management and its Relation to Other Dimensions of Health System Performance (RAC) - Study Design and Methodology.

PubMed

Röttger, Julia; Blümel, Miriam; Engel, Susanne; Grenz-Farenholtz, Brigitte; Fuchs, Sabine; Linder, Roland; Verheyen, Frank; Busse, Reinhard

2015-05-20

The responsiveness of a health system is considered to be an intrinsic goal of health systems and an essential aspect in performance assessment. Numerous studies have analysed health system responsiveness and related concepts, especially across different countries and health systems. However, fewer studies have applied the concept for the evaluation of specific healthcare delivery structures and thoroughly analysed its determinants within one country. The aims of this study are to assess the level of perceived health system responsiveness to patients with chronic diseases in ambulatory care in Germany and to analyse the determinants of health system responsiveness as well as its distribution across different population groups. The target population consists of chronically ill people in Germany, with a focus on patients suffering from type 2 diabetes and/or from coronary heart disease (CHD). Data comes from two different sources: (i) cross-sectional survey data from a postal survey and (ii) claims data from a German sickness fund. Data from both sources will be linked at an individual-level. The postal survey has the purpose of measuring perceived health system responsiveness, health related quality of life, experiences with disease management programmes (DMPs) and (subjective) socioeconomic background. The claims data consists of information on (co)morbidities, service utilization, enrolment within a DMP and sociodemographic characteristics, including the type of residential area. RAC is one of the first projects linking survey data on health system responsiveness at individual level with claims data. With this unique database, it will be possible to comprehensively analyse determinants of health system responsiveness and its relation to other aspects of health system performance assessment. The results of the project will allow German health system decision-makers to assess the performance of nonclinical aspects of healthcare delivery and their determinants in two important areas of health policy: in ambulatory and chronic disease care. © 2015 by Kerman University of Medical Sciences.

Understanding the immune response to seasonal influenza vaccination in older adults: a systems biology approach.

PubMed

Lambert, Nathaniel D; Ovsyannikova, Inna G; Pankratz, V Shane; Jacobson, Robert M; Poland, Gregory A

2012-08-01

Annual vaccination against seasonal influenza is recommended to decrease disease-related mortality and morbidity. However, one population that responds suboptimally to influenza vaccine is adults over the age of 65 years. The natural aging process is associated with a complex deterioration of multiple components of the host immune system. Research into this phenomenon, known as immunosenescence, has shown that aging alters both the innate and adaptive branches of the immune system. The intricate mechanisms involved in immune response to influenza vaccine, and how these responses are altered with age, have led us to adopt a more encompassing systems biology approach to understand exactly why the response to vaccination diminishes with age. Here, the authors review what changes occur with immunosenescence, and some immunogenetic factors that influence response, and outline the systems biology approach to understand the immune response to seasonal influenza vaccination in older adults.

DR5 as a reporter system to study auxin response in Populus.

PubMed

Chen, Yiru; Yordanov, Yordan S; Ma, Cathleen; Strauss, Steven; Busov, Victor B

2013-03-01

KEY MESSAGE : Auxin responsive promoter DR5 reporter system is functional in Populus to monitor auxin response in tissues including leaves, roots, and stems. We described the behavior of the DR5::GUS reporter system in stably transformed Populus plants. We found several similarities with Arabidopsis, including sensitivity to native and synthetic auxins, rapid induction after treatment in a variety of tissues, and maximal responses in root tissues. There were also several important differences from Arabidopsis, including slower time to maximum response and lower induction amplitude. Young leaves and stem sections below the apex showed much higher DR5 activity than did older leaves and stems undergoing secondary growth. DR5 activity was highest in cortex, suggesting high levels of auxin concentration and/or sensitivity in this tissue. Our study shows that the DR5 reporter system is a sensitive and facile system for monitoring auxin responses and distribution at cellular resolution in poplar.

Multiswitching compound antisynchronization of four chaotic systems

NASA Astrophysics Data System (ADS)

Khan, Ayub; Khattar, Dinesh; Prajapati, Nitish

2017-12-01

Based on three drive-one response system, in this article, the authors investigate a novel synchronization scheme for a class of chaotic systems. The new scheme, multiswitching compound antisynchronization (MSCoAS), is a notable extension of the earlier multiswitching schemes concerning only one drive-one response system model. The concept of multiswitching synchronization is extended to compound synchronization scheme such that the state variables of three drive systems antisynchronize with different state variables of the response system, simultaneously. The study involving multiswitching of three drive systems and one response system is first of its kind. Various switched modified function projective antisynchronization schemes are obtained as special cases of MSCoAS, for a suitable choice of scaling factors. Using suitable controllers and Lyapunov stability theory, sufficient condition is obtained to achieve MSCoAS between four chaotic systems and the corresponding theoretical proof is given. Numerical simulations are performed using Lorenz system in MATLAB to demonstrate the validity of the presented method.

Response formulae for n-point correlations in statistical mechanical systems and application to a problem of coarse graining

NASA Astrophysics Data System (ADS)

Lucarini, Valerio; Wouters, Jeroen

2017-09-01

Predicting the response of a system to perturbations is a key challenge in mathematical and natural sciences. Under suitable conditions on the nature of the system, of the perturbation, and of the observables of interest, response theories allow to construct operators describing the smooth change of the invariant measure of the system of interest as a function of the small parameter controlling the intensity of the perturbation. In particular, response theories can be developed both for stochastic and chaotic deterministic dynamical systems, where in the latter case stricter conditions imposing some degree of structural stability are required. In this paper we extend previous findings and derive general response formulae describing how n- point correlations are affected by perturbations to the vector flow. We also show how to compute the response of the spectral properties of the system to perturbations. We then apply our results to the seemingly unrelated problem of coarse graining in multiscale systems: we find explicit formulae describing the change in the terms describing the parameterisation of the neglected degrees of freedom resulting from applying perturbations to the full system. All the terms envisioned by the Mori-Zwanzig theory—the deterministic, stochastic, and non-Markovian terms—are affected at first order in the perturbation. The obtained results provide a more comprehensive understanding of the response of statistical mechanical systems to perturbations. They also contribute to the goal of constructing accurate and robust parameterisations and are of potential relevance for fields like molecular dynamics, condensed matter, and geophysical fluid dynamics. We envision possible applications of our general results to the study of the response of climate variability to anthropogenic and natural forcing and to the study of the equivalence of thermostatted statistical mechanical systems.

Development of a model protection and dynamic response monitoring system for the national transonic facility

NASA Technical Reports Server (NTRS)

Young, Clarence P., Jr.; Balakrishna, S.; Kilgore, W. Allen

1995-01-01

A state-of-the-art, computerized mode protection and dynamic response monitoring system has been developed for the NASA Langley Research Center National Transonic Facility (NTF). This report describes the development of the model protection and shutdown system (MPSS). A technical description of the system is given along with discussions on operation and capabilities of the system. Applications of the system to vibration problems are presented to demonstrate the system capabilities, typical applications, versatility, and investment research return derived from the system to date. The system was custom designed for the NTF but can be used at other facilities or for other dynamic measurement/diagnostic applications. Potential commercial uses of the system are described. System capability has been demonstrated for forced response testing and for characterizing and quantifying bias errors for onboard inertial model attitude measurement devices. The system is installed in the NTF control room and has been used successfully for monitoring, recording and analyzing the dynamic response of several model systems tested in the NTF.

The Stress Response Systems: Universality and Adaptive Individual Differences

ERIC Educational Resources Information Center

Ellis, Bruce J.; Jackson, Jenee James; Boyce, W. Thomas

2006-01-01

Biological reactivity to psychological stressors comprises a complex, integrated system of central neural and peripheral neuroendocrine responses designed to prepare the organism for challenge or threat. Developmental experience plays a role, along with heritable variation, in calibrating the response dynamics of this system. This calibration…

Recursive Inversion By Finite-Impulse-Response Filters

NASA Technical Reports Server (NTRS)

Bach, Ralph E., Jr.; Baram, Yoram

1991-01-01

Recursive approximation gives least-squares best fit to exact response. Algorithm yields finite-impulse-response approximation of unknown single-input/single-output, causal, time-invariant, linear, real system, response of which is sequence of impulses. Applicable to such system-inversion problems as suppression of echoes and identification of target from its scatter response to incident impulse.

Health Systems' Responsiveness and Its Characteristics: A Cross-Country Comparative Analysis

PubMed Central

Robone, Silvana; Rice, Nigel; Smith, Peter C

2011-01-01

Objectives Responsiveness has been identified as one of the intrinsic goals of health care systems. Little is known, however, about its determinants. Our objective is to investigate the potential country-level drivers of health system responsiveness. Data Source Data on responsiveness are taken from the World Health Survey. Information on country-level characteristics is obtained from a variety of sources including the United Nations Development Program (UNDP). Study Design A two-step procedure. First, using survey data we derive a country-level measure of system responsiveness purged of differences in individual reporting behavior. Secondly, we run cross-sectional country-level regressions of responsiveness on potential drivers. Principal Findings Health care expenditures per capita are positively associated with responsiveness, after controlling for the influence of potential confounding factors. Aspects of responsiveness are also associated with public sector spending (negatively) and educational development (positively). Conclusions From a policy perspective, improvements in responsiveness may require higher spending levels. The expansion of nonpublic sector provision, perhaps in the form of increased patient choice, may also serve to improve responsiveness. However, these inferences are tentative and require further study. PMID:21762144

21 CFR 820.20 - Management responsibility.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Management responsibility. 820.20 Section 820.20...) MEDICAL DEVICES QUALITY SYSTEM REGULATION Quality System Requirements § 820.20 Management responsibility. (a) Quality policy. Management with executive responsibility shall establish its policy and...

21 CFR 820.20 - Management responsibility.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Management responsibility. 820.20 Section 820.20...) MEDICAL DEVICES QUALITY SYSTEM REGULATION Quality System Requirements § 820.20 Management responsibility. (a) Quality policy. Management with executive responsibility shall establish its policy and...

21 CFR 820.20 - Management responsibility.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Management responsibility. 820.20 Section 820.20...) MEDICAL DEVICES QUALITY SYSTEM REGULATION Quality System Requirements § 820.20 Management responsibility. (a) Quality policy. Management with executive responsibility shall establish its policy and...

21 CFR 820.20 - Management responsibility.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Management responsibility. 820.20 Section 820.20...) MEDICAL DEVICES QUALITY SYSTEM REGULATION Quality System Requirements § 820.20 Management responsibility. (a) Quality policy. Management with executive responsibility shall establish its policy and...

21 CFR 820.20 - Management responsibility.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Management responsibility. 820.20 Section 820.20...) MEDICAL DEVICES QUALITY SYSTEM REGULATION Quality System Requirements § 820.20 Management responsibility. (a) Quality policy. Management with executive responsibility shall establish its policy and...

33 CFR 154.1020 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... personnel identified to staff the organizational structure identified in a response plan to manage response... identifying response systems and equipment in a response plan for the applicable operating environment... visibility, and currents within the COTP zone in which the systems or equipment are intended to function...

Analysis of DISMS (Defense Integrated Subsistence Management System) Increment 4

DTIC Science & Technology

1988-12-01

response data entry; and rationale supporting an on-line system based on real time management information needs. Keywords: Automated systems; Subsistence; Workload capacity; Bid response; Contract administration; Computer systems.

40 CFR 1065.309 - Continuous gas analyzer system-response and updating-recording verification-for gas analyzers...

Code of Federal Regulations, 2010 CFR

2010-07-01

... change the system response. (b) Measurement principles. This procedure verifies that the updating and... gas detectors used to generate a continuously combined/compensated concentration measurement signal... verifies that the measurement system meets a minimum response time. For this procedure, ensure that all...

78 FR 38341 - Federal Acquisition Regulation; Information Collection; Preaward Survey Forms (Standard Forms...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-26

... Contractor Accounting System Respondents: 708. Responses annually: 1. Total Responses: 708. Hours per... contractor responsibility in each case. B. Annual Reporting Burden There are no Governmentwide systems for... Procurement Data System (FPDS) ad hoc report was completed identifying that in Fiscal Year (FY) 2012 an...

48 CFR 301.604-71 - HCA authorities and responsibilities.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 4 2012-10-01 2012-10-01 false HCA authorities and responsibilities. 301.604-71 Section 301.604-71 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES GENERAL HHS ACQUISITION REGULATION SYSTEM Career Development, Contracting Authority, and Responsibilities 301.604-71 HCA authorities and...

48 CFR 301.604-71 - HCA authorities and responsibilities.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 4 2011-10-01 2011-10-01 false HCA authorities and responsibilities. 301.604-71 Section 301.604-71 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES GENERAL HHS ACQUISITION REGULATION SYSTEM Career Development, Contracting Authority, and Responsibilities 301.604-71 HCA authorities and...

48 CFR 301.604-71 - HCA authorities and responsibilities.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 48 Federal Acquisition Regulations System 4 2014-10-01 2014-10-01 false HCA authorities and responsibilities. 301.604-71 Section 301.604-71 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES GENERAL HHS ACQUISITION REGULATION SYSTEM Career Development, Contracting Authority, and Responsibilities 301.604-71 HCA authorities and...

48 CFR 301.604-71 - HCA authorities and responsibilities.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false HCA authorities and responsibilities. 301.604-71 Section 301.604-71 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES GENERAL HHS ACQUISITION REGULATION SYSTEM Career Development, Contracting Authority, and Responsibilities 301.604-71 HCA authorities and...

48 CFR 301.604-71 - HCA authorities and responsibilities.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 48 Federal Acquisition Regulations System 4 2013-10-01 2013-10-01 false HCA authorities and responsibilities. 301.604-71 Section 301.604-71 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES GENERAL HHS ACQUISITION REGULATION SYSTEM Career Development, Contracting Authority, and Responsibilities 301.604-71 HCA authorities and...

The association between attributions of responsibility for motor vehicle accidents and patient satisfaction: a study within a no-fault injury compensation system.

PubMed

Thompson, Jason; Berk, Michael; O'Donnell, Meaghan; Stafford, Lesley; Nordfjaern, Trond

2015-05-01

This study set out to test the relationship between attributions of responsibility for motor vehicle accidents and satisfaction with personal injury compensation systems. The study analysed survey data from 1394 people injured in a motor vehicle accident who were compensated under a no-fault personal injury compensation system. Patients' ratings of satisfaction with the compensation system across five domains (resolves your issues, keeps you up-to-date, treats you as an individual, cares about you, and overall satisfaction) were analysed alongside patient attributions of responsibility for their accident (not responsible, partly responsible, totally responsible). Postaccident physical and mental health status, age, gender, and duration of compensation claim were controlled for in the analysis. A multivariate analysis of covariance indicated attributions of responsibility for accidents were significantly associated with levels of patient satisfaction across all five domains under study (F (10, 2084) = 3.7, p<0.001, η(2)  =0.02). Despite access to virtually indistinguishable services, patients who attributed responsibility for their accidents to others were significantly less satisfied with the injury compensation system than those who attributed responsibility to themselves. Satisfaction with no-fault motor vehicle injury compensation services are associated with patients' attributions of responsibility for their accident. Compensation systems and other rehabilitation services monitoring patient satisfaction should adjust for attributions of responsibility when assessing levels of patient satisfaction between time periods, services, or injured populations. Differences in levels of patient satisfaction observed between compensation or rehabilitation populations may reflect differences in attributions of responsibility for accidents rather than objective service quality. © The Author(s) 2014.

Sensor/Response Coordination In A Tactical Self-Protection System

NASA Astrophysics Data System (ADS)

Steinberg, Alan N.

1988-08-01

This paper describes a model for integrating information acquisition functions into a response planner within a tactical self-defense system. This model may be used in defining requirements in such applications for sensor systems and for associated processing and control functions. The goal of information acquisition in a self-defense system is generally not that of achieving the best possible estimate of the threat environment; but rather to provide resolution of that environment sufficient to support response decisions. We model the information acquisition problem as that of achieving a partition among possible world states such that the final partition maps into the system's repertoire of possible responses.

Opposing and following responses in sensorimotor speech control: Why responses go both ways.

PubMed

Franken, Matthias K; Acheson, Daniel J; McQueen, James M; Hagoort, Peter; Eisner, Frank

2018-06-04

When talking, speakers continuously monitor and use the auditory feedback of their own voice to control and inform speech production processes. When speakers are provided with auditory feedback that is perturbed in real time, most of them compensate for this by opposing the feedback perturbation. But some responses follow the perturbation. In the present study, we investigated whether the state of the speech production system at perturbation onset may determine what type of response (opposing or following) is made. The results suggest that whether a perturbation-related response is opposing or following depends on ongoing fluctuations of the production system: The system initially responds by doing the opposite of what it was doing. This effect and the nontrivial proportion of following responses suggest that current production models are inadequate: They need to account for why responses to unexpected sensory feedback depend on the production system's state at the time of perturbation.

Evaluation of TLR Agonists as Potential Mucosal Adjuvants for HIV gp140 and Tetanus Toxoid in Mice

PubMed Central

Buffa, Viviana; Klein, Katja; Fischetti, Lucia; Shattock, Robin J.

2012-01-01

In the present study we investigate the impact of a range of TLR ligands and chitosan as potential adjuvants for different routes of mucosal immunisation (sublingual (SL), intranasal (IN), intravaginal (IVag) and a parenteral route (subcutaneous (SC)) in the murine model. We assess their ability to enhance antibody responses to HIV-1 CN54gp140 (gp140) and Tetanus toxoid (TT) in systemic and vaginal compartments. A number of trends were observed by route of administration. For non-adjuvanted antigen, SC>SL>IN immunisation with respect to systemic IgG responses, where endpoint titres were greater for TT than for gp140. In general, co-administration with adjuvants increased specific IgG responses where IN = SC>SL, while in the vaginal compartment IN>SL>SC for specific IgA. In contrast, for systemic and mucosal IgA responses to antigen alone SL>IN = SC. A number of adjuvants increased specific systemic IgA responses where in general IN>SL>SC immunisation, while for mucosal responses IN = SL>SC. In contrast, direct intravaginal immunisation failed to induce any detectable systemic or mucosal responses to gp140 even in the presence of adjuvant. However, significant systemic IgG responses to TT were induced by intravaginal immunisation with or without adjuvant, and detectable mucosal responses IgG and IgA were observed when TT was administered with FSL-1 or Poly I∶C. Interestingly some TLRs displayed differential activity dependent upon the route of administration. MPLA (TLR4) suppressed systemic responses to SL immunisation while enhancing responses to IN or SC immunisation. CpG B enhanced SL and IN responses, while having little or no impact on SC immunisation. These data demonstrate important route, antigen and adjuvant effects that need to be considered in the design of mucosal vaccine strategies. PMID:23272062

48 CFR 47.305-15 - Loading responsibilities of contractors.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Loading responsibilities of contractors. 47.305-15 Section 47.305-15 Federal Acquisition Regulations System FEDERAL... responsibilities of contractors. (a)(1) Contractors are responsible for loading, blocking, and bracing carload...

Comparison of digital controllers used in magnetic suspension and balance systems

NASA Technical Reports Server (NTRS)

Kilgore, William A.

1990-01-01

Dynamic systems that were once controlled by analog circuits are now controlled by digital computers. Presented is a comparison of the digital controllers presently used with magnetic suspension and balance systems. The overall responses of the systems are compared using a computer simulation of the magnetic suspension and balance system and the digital controllers. The comparisons include responses to both simulated force and position inputs. A preferred digital controller is determined from the simulated responses.

Response diversity to land use occurs but does not consistently stabilise ecosystem services provided by native pollinators.

PubMed

Cariveau, Daniel P; Williams, Neal M; Benjamin, Faye E; Winfree, Rachael

2013-07-01

More diverse biological communities may provide ecosystem services that are less variable over space or time. However, the mechanisms underlying this relationship are rarely investigated empirically in real-world ecosystems. Here, we investigate how a potentially important stabilising mechanism, response diversity, the differential response to environmental change among species, stabilises pollination services against land-use change. We measured crop pollination services provided by native bees across land-use gradients in three crop systems. We found that bee species responded differentially to increasing agricultural land cover in all three systems, demonstrating that response diversity occurs. Similarly, we found response diversity in pollination services in two of the systems. However, there was no evidence that response diversity, in general, stabilised ecosystem services. Our results suggest that either response diversity is not the primary stabilising mechanism in our system, or that new measures of response diversity are needed that better capture the stabilising effects it provides. © 2013 John Wiley & Sons Ltd/CNRS.

Development of a Graphics Based Automated Emergency Response System (AERS) for Rail Transit Systems

DOT National Transportation Integrated Search

1989-05-01

This report presents an overview of the second generation Automated Emergency Response System (AERS2). Developed to assist transit systems in responding effectively to emergency situations, AERS2 is a microcomputer-based information retrieval system ...

Effects of mistuning and matrix structure on the topology of frequency response curves

NASA Technical Reports Server (NTRS)

Afolabi, Dare

1989-01-01

The stability of a frequency response curve under mild perturbations of the system's matrix is investigated. Using recent developments in the theory of singularities of differentiable maps, it is shown that the stability of a response curve depends on the structure of the system's matrix. In particular, the frequency response curves of a cylic system are shown to be unstable. Consequently, slight parameter variations engendered by mistuning will induce a significant difference in the topology of the forced response curves, if the mistuning transformation crosses the bifurcation set.

Shallow End Response from ATEM

NASA Astrophysics Data System (ADS)

Vetrov, A.

2014-12-01

Different geological, hydrological, environmental and engineering targets are located shallow underground. The information collected with ATEM systems might be very useful for their study; although there are many deeper targets that the ATEM systems are traditionally used for. The idea to raise magnetic moment output and get deeper penetration response was one of the goals of ATEM systems development during the last decade. The shallow geology response was a trade for such systems, which sometimes were almost blind in the first hundred meter under surface. The possibility to achieve shallow end response from ATEM systems has become significant subject in last years. Several airborne TDEM systems got second higher frequency and lower magnetic moment signal to pick up shallow response together with deep one. Having a potential advantage such implementation raises complication and cost of the system. There's no need to receive 500 meter deep response when exploring shallow geology. P-THEM system having a compact size transmitter and relatively light weight is working on one base frequency at a time, but this frequency can be preset before a flight considering survey goals. A study of shallow geology response of the P-THEM system working on different base frequency has been conducted in 2014 in Ontario. The Alliston test area located in Southern Ontario has been flown with the P-THEM system working on base frequencies 30Hz and 90Hz. Results of the observations will be discussed in the presentation. The shallow end data can be used for mineral exploration applications and also for hydrological and environmental studies.

Student Response (Clicker) Systems: Preferences of Biomedical Physiology Students in Asian Classes

ERIC Educational Resources Information Center

Hwang, Isabel; Wong, Kevin; Lam, Shun Leung; Lam, Paul

2015-01-01

Student response systems (commonly called "clickers") are valuable tools for engaging students in classroom interactions. In this study, we investigated the use of two types of response systems (a traditional clicker and a mobile device) by students in human physiology courses. Our results showed high student satisfaction with the use of…

Using crosscorrelation techniques to determine the impulse response of linear systems

NASA Technical Reports Server (NTRS)

Dallabetta, Michael J.; Li, Harry W.; Demuth, Howard B.

1993-01-01

A crosscorrelation method of measuring the impulse response of linear systems is presented. The technique, implementation, and limitations of this method are discussed. A simple system is designed and built using discrete components and the impulse response of a linear circuit is measured. Theoretical and software simulation results are presented.

Student Response Systems and Learner Engagement in Large Classes

ERIC Educational Resources Information Center

Heaslip, Graham; Donovan, Paul; Cullen, John G.

2014-01-01

The use of student response systems is becoming more prevalent in higher level education. Evidence on the effectiveness of this technology can be an important resource for tutors seeking to engage with learners and raise the quality of learning experiences. Student response systems have been found to increase student engagement and participation…

"Clickers" and Metacognition: How Do Electronic Response Devices ("Clickers") Influence Student Metacognition?

ERIC Educational Resources Information Center

Manke-Brady, Melanie

2012-01-01

The purpose of this study was to examine whether electronic response systems influence student metacognitions in large lecture settings, and how metacognitive processes are influenced. Moreover, this study compared electronic response systems with a low technology system and sought to establish whether differences exist in how the two response…

48 CFR 1845.502 - Contractor responsibility.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Contractor responsibility. 1845.502 Section 1845.502 Federal Acquisition Regulations System NATIONAL AERONAUTICS AND SPACE... Contractors 1845.502 Contractor responsibility. ...

Nonvocal language acquisition in adolescents with severe physical disabilities: Bliss symbol versus iconic stimulus formats.

PubMed Central

Hurlbut, B I; Iwata, B A; Green, J D

1982-01-01

This study compared training in two language systems for three severely handicapped, nonvocal adolescents: the Bliss symbol system and an iconic picture system. Following baseline, training and review trials were implemented using an alternating treatments design. Daily probes were conducted to assess maintenance, stimulus generalization, and response generalization, and data were collected on spontaneous usage of either language system throughout the school day. Results showed that students required approximately four times as many trials to acquire Bliss symbols as iconic pictures, and that students maintained a higher percentage of iconic pictures. Stimulus generalization occurred in both language systems, while the number of correct responses during responses generalization probes was much greater for the iconic system. Finally, students almost always showed more iconic responses than Bliss responses in daily spontaneous usage. These results suggest that an iconic system might be more readily spontaneous usage. These results suggest than an iconic system might be more readily acquired, maintained, and generalized to daily situations. Implications of these findings for the newly verbal person were discussed. PMID:6181049

14 CFR 1212.704 - System manager.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 14 Aeronautics and Space 5 2012-01-01 2012-01-01 false System manager. 1212.704 Section 1212.704... Authority and Responsibilities § 1212.704 System manager. (a) Each system manager is responsible for the following with regard to the system of records over which the system manager has cognizance: (1) Overall...

14 CFR 1212.705 - System manager.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 14 Aeronautics and Space 5 2013-01-01 2013-01-01 false System manager. 1212.705 Section 1212.705... Authority and Responsibilities § 1212.705 System manager. (a) Each system manager is responsible for the following with regard to the system of records over which the system manager has cognizance: (1) Overall...

14 CFR 1212.704 - System manager.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 14 Aeronautics and Space 5 2011-01-01 2010-01-01 true System manager. 1212.704 Section 1212.704... Authority and Responsibilities § 1212.704 System manager. (a) Each system manager is responsible for the following with regard to the system of records over which the system manager has cognizance: (1) Overall...

14 CFR 1212.704 - System manager.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 14 Aeronautics and Space 5 2010-01-01 2010-01-01 false System manager. 1212.704 Section 1212.704... Authority and Responsibilities § 1212.704 System manager. (a) Each system manager is responsible for the following with regard to the system of records over which the system manager has cognizance: (1) Overall...

Occupational Safety and Health System for Workers Engaged in Emergency Response Operations in the USA.

PubMed

Toyoda, Hiroyuki; Kubo, Tatsuhiko; Mori, Koji

2016-12-03

To study the occupational safety and health systems used for emergency response workers in the USA, we performed interviews with related federal agencies and conducted research on related studies. We visited the Federal Emergency Management Agency (FEMA) and National Institute for Occupational Safety and Health (NIOSH) in the USA and performed interviews with their managers on the agencies' roles in the national emergency response system. We also obtained information prepared for our visit from the USA's Occupational Safety and Health Administration (OSHA). In addition, we conducted research on related studies and information on the website of the agencies. We found that the USA had an established emergency response system based on their National Incident Management System (NIMS). This enabled several organizations to respond to emergencies cooperatively using a National Response Framework (NRF) that clarifies the roles and cooperative functions of each federal agency. The core system in NIMS was the Incident Command System (ICS), within which a Safety Officer was positioned as one of the command staff supporting the commander. All ICS staff were required to complete a training program specific to their position; in addition, the Safety Officer was required to have experience. The All-Hazards model was commonly used in the emergency response system. We found that FEMA coordinated support functions, and OSHA and NIOSH, which had specific functions to protect workers, worked cooperatively under NRF. These agencies employed certified industrial hygienists that play a professional role in safety and health. NIOSH recently executed support activities during disasters and other emergencies. The USA's emergency response system is characterized by functions that protect the lives and health of emergency response workers. Trained and experienced human resources support system effectiveness. The findings provided valuable information that could be used to improve the occupational safety and health function in the Japanese system.

Apparatus and Methods for Manipulation and Optimization of Biological Systems

NASA Technical Reports Server (NTRS)

Sun, Ren (Inventor); Ho, Chih-Ming (Inventor); Wong, Pak Kin (Inventor); Yu, Fuqu (Inventor)

2014-01-01

The invention provides systems and methods for manipulating biological systems, for example to elicit a more desired biological response from a biological sample, such as a tissue, organ, and/or a cell. In one aspect, the invention operates by efficiently searching through a large parametric space of stimuli and system parameters to manipulate, control, and optimize the response of biological samples sustained in the system. In one aspect, the systems and methods of the invention use at least one optimization algorithm to modify the actuator's control inputs for stimulation, responsive to the sensor's output of response signals. The invention can be used, e.g., to optimize any biological system, e.g., bioreactors for proteins, and the like, small molecules, polysaccharides, lipids, and the like. Another use of the apparatus and methods includes is for the discovery of key parameters in complex biological systems.

Health systems' responsiveness and its characteristics: a cross-country comparative analysis.

PubMed

Robone, Silvana; Rice, Nigel; Smith, Peter C

2011-12-01

OBJECTIVES. Responsiveness has been identified as one of the intrinsic goals of health care systems. Little is known, however, about its determinants. Our objective is to investigate the potential country-level drivers of health system responsiveness. DATA SOURCE. Data on responsiveness are taken from the World Health Survey. Information on country-level characteristics is obtained from a variety of sources including the United Nations Development Program (UNDP). STUDY DESIGN. A two-step procedure. First, using survey data we derive a country-level measure of system responsiveness purged of differences in individual reporting behavior. Secondly, we run cross-sectional country-level regressions of responsiveness on potential drivers. PRINCIPAL FINDINGS. Health care expenditures per capita are positively associated with responsiveness, after controlling for the influence of potential confounding factors. Aspects of responsiveness are also associated with public sector spending (negatively) and educational development (positively). CONCLUSIONS. From a policy perspective, improvements in responsiveness may require higher spending levels. The expansion of nonpublic sector provision, perhaps in the form of increased patient choice, may also serve to improve responsiveness. However, these inferences are tentative and require further study. © Health Research and Educational Trust.

Systemic inflammatory response following acute myocardial infarction

PubMed Central

Fang, Lu; Moore, Xiao-Lei; Dart, Anthony M; Wang, Le-Min

2015-01-01

Acute cardiomyocyte necrosis in the infarcted heart generates damage-associated molecular patterns, activating complement and toll-like receptor/interleukin-1 signaling, and triggering an intense inflammatory response. Inflammasomes also recognize danger signals and mediate sterile inflammatory response following acute myocardial infarction (AMI). Inflammatory response serves to repair the heart, but excessive inflammation leads to adverse left ventricular remodeling and heart failure. In addition to local inflammation, profound systemic inflammation response has been documented in patients with AMI, which includes elevation of circulating inflammatory cytokines, chemokines and cell adhesion molecules, and activation of peripheral leukocytes and platelets. The excessive inflammatory response could be caused by a deregulated immune system. AMI is also associated with bone marrow activation and spleen monocytopoiesis, which sustains a continuous supply of monocytes at the site of inflammation. Accumulating evidence has shown that systemic inflammation aggravates atherosclerosis and markers for systemic inflammation are predictors of adverse clinical outcomes (such as death, recurrent myocardial infarction, and heart failure) in patients with AMI. PMID:26089856

Spinning Reserve From Hotel Load Response: Initial Progress

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kueck, John D; Kirby, Brendan J

2008-11-01

This project was motivated by the fundamental match between hotel space conditioning load response capability and power system contingency response needs. As power system costs rise and capacity is strained demand response can provide a significant system reliability benefit at a potentially attractive cost. At ORNL s suggestion, Digital Solutions Inc. adapted its hotel air conditioning control technology to supply power system spinning reserve. This energy saving technology is primarily designed to provide the hotel operator with the ability to control individual room temperature set-points based upon occupancy (25% to 50% energy savings based on an earlier study [Kirby andmore » Ally, 2002]). DSI added instantaneous local load shedding capability in response to power system frequency and centrally dispatched load shedding capability in response to power system operator command. The 162 room Music Road Hotel in Pigeon Forge Tennessee agreed to host the spinning reserve test. The Tennessee Valley Authority supplied real-time metering equipment in the form of an internet connected Dranetz-BMI power quality meter and monitoring expertise to record total hotel load during both normal operations and test results. The Sevier County Electric System installed the metering. Preliminary testing showed that hotel load can be curtailed by 22% to 37% depending on the outdoor temperature and the time of day. These results are prior to implementing control over the common area air conditioning loads. Testing was also not at times of highest system or hotel loading. Full response occurred in 12 to 60 seconds from when the system operator s command to shed load was issued. The load drop was very rapid, essentially as fast as the 2 second metering could detect, with all units responding essentially simultaneously. Load restoration was ramped back in over several minutes. The restoration ramp can be adjusted to the power system needs. Frequency response testing was not completed. Initial testing showed that the units respond very quickly. Problems with local power quality generated false low frequency signals which required testing to be stopped. This should not be a problem in actual operation since the frequency trip points will be staggered to generate a droop curve which mimics generator governor response. The actual trip frequencies will also be low enough to avoid power quality problems. The actual trip frequencies are too low to generate test events with sufficient regularity to complete testing in a reasonable amount of time. Frequency response testing will resume once the local power quality problem is fully understood and reasonable test frequency settings can be determined. Overall the preliminary testing was extremely successful. The hotel response capability matches the power system reliability need, being faster than generation response and inherently available when the power system is under the most stress (times of high system and hotel load). Periodic testing is scheduled throughout the winter and spring to characterize hotel response capability under a full range of conditions. More extensive testing will resume when summer outdoor temperatures are again high enough to fully test hotel response.« less

Dynamic Docking Test System (DDTS) active table frequency response test results. [Apollo Soyuz Test Project

NASA Technical Reports Server (NTRS)

Gates, R. M.

1974-01-01

Results are presented of the frequency response test performed on the dynamic docking test system (DDTS) active table. Sinusoidal displacement commands were applied to the table and the dynamic response determined from measured actuator responses and accelerometers mounted to the table and one actuator.

40 CFR 281.37 - Financial responsibility for UST systems containing petroleum.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Financial responsibility for UST... for No-Less-Stringent § 281.37 Financial responsibility for UST systems containing petroleum. (a) In order to be considered no less stringent than the federal requirements for financial responsibility for...

Analysis of dynamic system response to product random processes

NASA Technical Reports Server (NTRS)

Sidwell, K.

1978-01-01

The response of dynamic systems to the product of two independent Gaussian random processes is developed by use of the Fokker-Planck and associated moment equations. The development is applied to the amplitude modulated process which is used to model atmospheric turbulence in aeronautical applications. The exact solution for the system response is compared with the solution obtained by the quasi-steady approximation which omits the dynamic properties of the random amplitude modulation. The quasi-steady approximation is valid as a limiting case of the exact solution for the dynamic response of linear systems to amplitude modulated processes. In the nonlimiting case the quasi-steady approximation can be invalid for dynamic systems with low damping.

Visual examination apparatus

NASA Technical Reports Server (NTRS)

Haines, R. F.; Fitzgerald, J. W.; Rositano, S. A. (Inventor)

1976-01-01

An automated visual examination apparatus for measuring visual sensitivity and mapping blind spot location including a projection system for displaying to a patient a series of visual stimuli. A response switch enables him to indicate his reaction to the stimuli, and a recording system responsive to both the visual stimuli per se and the patient's response. The recording system thereby provides a correlated permanent record of both stimuli and response from which a substantive and readily apparent visual evaluation can be made.

Joint Planning Of Energy Storage and Transmission Considering Wind-Storage Combined System and Demand Side Response

NASA Astrophysics Data System (ADS)

Huang, Y.; Liu, B. Z.; Wang, K. Y.; Ai, X.

2017-12-01

In response to the new requirements of the operation mode of wind-storage combined system and demand side response for transmission network planning, this paper presents a joint planning of energy storage and transmission considering wind-storage combined system and demand side response. Firstly, the charge-discharge strategy of energy storage system equipped at the outlet of wind farm and demand side response strategy are analysed to achieve the best comprehensive benefits through the coordination of the two. Secondly, in the general transmission network planning model with wind power, both energy storage cost and demand side response cost are added to the objective function. Not only energy storage operation constraints and but also demand side response constraints are introduced into the constraint condition. Based on the classical formulation of TEP, a new formulation is developed considering the simultaneous addition of the charge-discharge strategy of energy storage system equipped at the outlet of the wind farm and demand side response strategy, which belongs to a typical mixed integer linear programming model that can be solved by mature optimization software. The case study based on the Garver-6 bus system shows that the validity of the proposed model is verified by comparison with general transmission network planning model. Furthermore, the results demonstrate that the joint planning model can gain more economic benefits through setting up different cases.

Consumer participation in early detection of the deteriorating patient and call activation to rapid response systems: a literature review.

PubMed

Vorwerk, Jane; King, Lindy

2016-01-01

This review investigated the impact of consumer participation in recognition of patient deterioration and response through call activation in rapid response systems. Nurses and doctors have taken the main role in recognition and response to patient deterioration through hospital rapid response systems. Yet patients and visitors (consumers) have appeared well placed to notice early signs of deterioration. In response, many hospitals have sought to partner health professionals with consumers in detection and response to early deterioration. However, to date, there have been no published research-based reviews to establish the impact of introducing consumer involvement into rapid response systems. A critical research-based review was undertaken. A comprehensive search of databases from 2006-2014 identified 11 studies. Critical appraisal of these studies was undertaken and thematic analysis of the findings revealed four major themes. Following implementation of the consumer activation programmes, the number of calls made by the consumers following detection of deterioration increased. Interestingly, the number of staff calls also increased. Importantly, mortality numbers were found to decrease in one major study following the introduction of consumer call activation. Consumer and staff knowledge and satisfaction with the new programmes indicated mixed results. Initial concerns of the staff over consumer involvement overwhelming the rapid response systems did not eventuate. Evaluation of successful consumer-activated programmes indicated the importance of: effective staff education and training; ongoing consumer education by nurses and clear educational materials. Findings indicated positive patient outcomes following introduction of consumer call activation programmes within rapid response systems. Effective consumer programmes included information that was readily accessible, easy-to-understand and available in a range of multimedia materials accompanied by the explanation and support of health professionals. Introduction of consumer-activated programmes within rapid response systems appears likely to improve outcomes for patients experiencing deterioration. © 2015 John Wiley & Sons Ltd.

System identification methods for aircraft flight control development and validation

NASA Technical Reports Server (NTRS)

Tischler, Mark B.

1995-01-01

System-identification methods compose a mathematical model, or series of models, from measurements of inputs and outputs of dynamic systems. The extracted models allow the characterization of the response of the overall aircraft or component subsystem behavior, such as actuators and on-board signal processing algorithms. This paper discusses the use of frequency-domain system-identification methods for the development and integration of aircraft flight-control systems. The extraction and analysis of models of varying complexity from nonparametric frequency-responses to transfer-functions and high-order state-space representations is illustrated using the Comprehensive Identification from FrEquency Responses (CIFER) system-identification facility. Results are presented for test data of numerous flight and simulation programs at the Ames Research Center including rotorcraft, fixed-wing aircraft, advanced short takeoff and vertical landing (ASTOVL), vertical/short takeoff and landing (V/STOL), tiltrotor aircraft, and rotor experiments in the wind tunnel. Excellent system characterization and dynamic response prediction is achieved for this wide class of systems. Examples illustrate the role of system-identification technology in providing an integrated flow of dynamic response data around the entire life-cycle of aircraft development from initial specifications, through simulation and bench testing, and into flight-test optimization.

Assessing the responsiveness of chronic disease care - is the World Health Organization's concept of health system responsiveness applicable?

PubMed

Röttger, Julia; Blümel, Miriam; Fuchs, Sabine; Busse, Reinhard

2014-07-01

The concept of health system responsiveness is an important dimension of health system performance assessment. Further efforts have been made in recent years to improve the analysis of responsiveness measurements, yet few studies have applied the responsiveness concept to the evaluation of specific health care delivery structures. The objective of this study was to test the World Health Organization's (WHO's) responsiveness concept for an application in the evaluation of chronic disease care. In September and October 2012 we conducted four focus groups of chronically ill people (n = 38) in Germany, in which participants discussed their experiences and expectations regarding health care. The data was analyzed deductively (on the basis of the WHO responsiveness concept) and inductively using directed content analysis. Ten themes related to health system responsiveness and one theme (finances) not directly related to health system responsiveness, but of high importance to the focus group participants, could be identified. Eight of the ten responsiveness themes are consistent with the WHO concept. Additionally, two new themes were identified: trust (consultation and treatment are not led by any motive other than the patients' wellbeing) and coordination (treatment involving different providers is coordinated and different actors communicate with each other). These findings indicate the suitability of the WHO responsiveness concept for the evaluation of chronic disease care. However, some amendments, in particular an extension of the concept to include the two domains trust and coordination, are necessary for a thorough assessment of the responsiveness of chronic disease care. Copyright © 2014 Elsevier Ltd. All rights reserved.

Reinterpreting Responsiveness for Health Systems Research in Low and Middle-Income Countries.

PubMed

Pratt, Bridget; Hyder, Adnan A

2015-07-01

The ethical concept of responsiveness has largely been interpreted in the context of international clinical research. In light of the increasing conduct of externally funded health systems research (HSR) in low- and middle-income countries (LMICs), this article examines how responsiveness might be understood for such research and how it can be applied. It contends that four features (amongst others) set HSR in LMICs apart from international clinical research: a focus on systems; being context-driven; being policy-driven; and being closely linked to development objectives. These features support reinterpreting responsiveness for HSR in LMICs as responsiveness to systems needs, where health system performance assessments can be relied upon to identify systems needs, and/or responsiveness to systems priorities, which entails aligning research with HSR priorities set through country-owned processes involving national and sub-national policymakers from host countries. Both concepts may be difficult to achieve in practice. Country ownership is not an established fact for many countries and alignment to their priorities may be meaningless without it. It is argued that more work is, therefore, needed to identify strategies for how the responsiveness requirement can be ethically fulfilled for HSR in LMICs under non-ideal conditions such as where host countries have not set HSR priorities via country-owned processes. Embeddedness is proposed as one approach that could be the focus of further development. © 2014 John Wiley & Sons Ltd.

A microcomputer-based emergency response system*.

PubMed

Belardo, S; Howell, A; Ryan, R; Wallace, W A

1983-09-01

A microcomputer-based system was developed to provide local officials responsible for disaster management with assistance during the crucial period immediately following a disaster, a period when incorrect decisions could have an adverse impact on the surrounding community. While the paper focuses on a potential disaster resulting from an accident at a commercial nuclear power generating facility, the system can be applied to other disastrous situations. Decisions involving evacuation, shelter and the deployment of resources must be made in response to floods, earthquakes, accidents in the transportation of hazardous materials, and hurricanes to name a few examples. As a decision aid, the system was designed to enhance data display by presenting the data in the form of representations (i.e. road maps, evacuation routes, etc.) as well as in list or tabular form. The potential impact of the event (i.e. the release of radioactive material) was displayed in the form of a cloud, representing the dispersion of the radioactive material. In addition, an algorithm was developed to assist the manager in assigning response resources to demands. The capability for modelling the impact of a disaster is discussed briefly, with reference to a system installed in the communities surrounding the Indian Point nuclear power plant in New York State. Results demonstrate both the technical feasibility of incorporating microcomputers indecision support systems for radiological emergency response, and the acceptance of such systems by those public officials responsible for implementing the response plans.

Investigations on response time of magnetorheological elastomer under compression mode

NASA Astrophysics Data System (ADS)

Zhu, Mi; Yu, Miao; Qi, Song; Fu, Jie

2018-05-01

For efficient fast control of vibration system with magnetorheological elastomer (MRE)-based smart device, the response time of MRE material is the key parameter which directly affects the control performance of the vibration system. For a step coil current excitation, this paper proposed a Maxwell behavior model with time constant λ to describe the normal force response of MRE, and the response time of MRE was extracted through the separation of coil response time. Besides, the transient responses of MRE under compression mode were experimentally investigated, and the effects of (i) applied current, (ii) particle distribution and (iii) compressive strain on the response time of MRE were addressed. The results revealed that the three factors can affect the response characteristic of MRE quite significantly. Besides the intrinsic importance for contributing to the response evaluation and effective design of MRE device, this study may conduce to the optimal design of controller for MRE control system.

5 CFR 9701.410 - DHS responsibilities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... RESOURCES MANAGEMENT SYSTEM Performance Management § 9701.410 DHS responsibilities. In carrying out its performance management system(s), DHS must— (a) Transfer ratings between subordinate organizations and to... 9701.410 Administrative Personnel DEPARTMENT OF HOMELAND SECURITY HUMAN RESOURCES MANAGEMENT SYSTEM...

5 CFR 9701.410 - DHS responsibilities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... RESOURCES MANAGEMENT SYSTEM Performance Management § 9701.410 DHS responsibilities. In carrying out its performance management system(s), DHS must— (a) Transfer ratings between subordinate organizations and to... 9701.410 Administrative Personnel DEPARTMENT OF HOMELAND SECURITY HUMAN RESOURCES MANAGEMENT SYSTEM...

5 CFR 9701.410 - DHS responsibilities.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RESOURCES MANAGEMENT SYSTEM Performance Management § 9701.410 DHS responsibilities. In carrying out its performance management system(s), DHS must— (a) Transfer ratings between subordinate organizations and to... 9701.410 Administrative Personnel DEPARTMENT OF HOMELAND SECURITY HUMAN RESOURCES MANAGEMENT SYSTEM...

Spill response system configuration study

DOT National Transportation Integrated Search

1996-05-01

This report describes the development of a prototype decision support system for oil spill response configuration plannig that will help U.S. Coast Guard planners to determine the appropriate response equipment and personnel for major spills. The rep...

Development of water environment information management and water pollution accident response system

NASA Astrophysics Data System (ADS)

Zhang, J.; Ruan, H.

2009-12-01

In recent years, many water pollution accidents occurred with the rapid economical development. In this study, water environment information management and water pollution accident response system are developed based on geographic information system (GIS) techniques. The system integrated spatial database, attribute database, hydraulic model, and water quality model under a user-friendly interface in a GIS environment. System ran in both Client/Server (C/S) and Browser/Server (B/S) platform which focused on model and inquiry respectively. System provided spatial and attribute data inquiry, water quality evaluation, statics, water pollution accident response case management (opening reservoir etc) and 2D and 3D visualization function, and gave assistant information to make decision on water pollution accident response. Polluted plume in Huaihe River were selected to simulate the transport of pollutes.

The Impact of Student Response Systems on the Learning Experience of Undergraduate Psychology Students

ERIC Educational Resources Information Center

Walklet, Elaine; Davis, Sarah; Farrelly, Daniel; Muse, Kate

2016-01-01

Student response systems (SRS) are hand-held devices or mobile phone polling systems which collate real-time, individual responses to on-screen questions. Previous research examining their role in higher education has highlighted both advantages and disadvantages of their use. This paper explores how different SRS influence the learning experience…

Integrating the Rights of the Child with the Responsibility of the Parent

ERIC Educational Resources Information Center

Scholz, Carolyn L.

2011-01-01

This paper will explore the balance between children's rights and parental responsibility from a family systems perspective. Children do not grow up in a vacuum; they are part of a biological, psychological and social system. The interaction of the child and parent within this system must include the development of responsibilities by the parent…

78 FR 71676 - Submission for Review: 3206-0201, Federal Employees Health Benefits (FEHB) Open Season Express...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... (FEHB) Open Season Express Interactive Voice Response (IVR) System and Open Season Web site AGENCY: U.S... Benefits (FEHB) Open Season Express Interactive Voice Response (IVR) System and the Open Season Web site... Season Express Interactive Voice Response (IVR) System, and the Open Season Web site, Open Season Online...

14 CFR § 1212.705 - System manager.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 14 Aeronautics and Space 5 2014-01-01 2014-01-01 false System manager. § 1212.705 Section § 1212... NASA Authority and Responsibilities § 1212.705 System manager. (a) Each system manager is responsible for the following with regard to the system of records over which the system manager has cognizance...

Development of a Tailored Methodology and Forensic Toolkit for Industrial Control Systems Incident Response

DTIC Science & Technology

2014-06-01

for industrial control systems ,” in Proceedings of the VDE Kongress, 2004. [15] K. Stouffer et al., “Special publication 800-82: Guide to industrial...TAILORED METHODOLOGY AND FORENSIC TOOLKIT FOR INDUSTRIAL CONTROL SYSTEMS INCIDENT RESPONSE by Nicholas B. Carr June 2014 Thesis Co...CONTROL SYSTEMS INCIDENT RESPONSE 5. FUNDING NUMBERS 6. AUTHOR(S) Nicholas B. Carr 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) Naval

Oncolytic Viral Therapy and the Immune System: A Double-Edged Sword Against Cancer.

PubMed

Marelli, Giulia; Howells, Anwen; Lemoine, Nicholas R; Wang, Yaohe

2018-01-01

Oncolytic viral therapy is a new promising strategy against cancer. Oncolytic viruses (OVs) can replicate in cancer cells but not in normal cells, leading to lysis of the tumor mass. Beside this primary effect, OVs can also stimulate the immune system. Tumors are an immuno-suppressive environment in which the immune system is silenced in order to avoid the immune response against cancer cells. The delivery of OVs into the tumor wakes up the immune system so that it can facilitate a strong and durable response against the tumor itself. Both innate and adaptive immune responses contribute to this process, producing an immune response against tumor antigens and facilitating immunological memory. However, viruses are recognized by the immune system as pathogens and the consequent anti-viral response could represent a big hurdle for OVs. Finding a balance between anti-tumor and anti-viral immunity is, under this new light, a priority for researchers. In this review, we provide an overview of the various ways in which different components of the immune system can be allied with OVs. We have analyzed the different immune responses in order to highlight the new and promising perspectives leading to increased anti-tumor response and decreased immune reaction to the OVs.

General response formula and application to topological insulator in quantum open system.

PubMed

Shen, H Z; Qin, M; Shao, X Q; Yi, X X

2015-11-01

It is well-known that the quantum linear response theory is based on the first-order perturbation theory for a system in thermal equilibrium. Hence, this theory breaks down when the system is in a steady state far from thermal equilibrium and the response up to higher order in perturbation is not negligible. In this paper, we develop a nonlinear response theory for such quantum open system. We first formulate this theory in terms of general susceptibility, after which we apply it to the derivation of Hall conductance for open system at finite temperature. As an example, the Hall conductance of the two-band model is derived. Then we calculate the Hall conductance for a two-dimensional ferromagnetic electron gas and a two-dimensional lattice model. The calculations show that the transition points of topological phase are robust against the environment. Our results provide a promising platform for the coherent manipulation of the nonlinear response in quantum open system, which has potential applications for quantum information processing and statistical physics.

Nonlinear dynamics in ecosystem response to climatic change: Case studies and policy implications

USGS Publications Warehouse

Burkett, Virginia R.; Wilcox, Douglas A.; Stottlemyer, Robert; Barrow, Wylie; Fagre, Dan; Baron, Jill S.; Price, Jeff; Nielsen, Jennifer L.; Allen, Craig D.; Peterson, David L.; Ruggerone, Greg; Doyle, Thomas

2005-01-01

Many biological, hydrological, and geological processes are interactively linked in ecosystems. These ecological phenomena normally vary within bounded ranges, but rapid, nonlinear changes to markedly different conditions can be triggered by even small differences if threshold values are exceeded. Intrinsic and extrinsic ecological thresholds can lead to effects that cascade among systems, precluding accurate modeling and prediction of system response to climate change. Ten case studies from North America illustrate how changes in climate can lead to rapid, threshold-type responses within ecological communities; the case studies also highlight the role of human activities that alter the rate or direction of system response to climate change. Understanding and anticipating nonlinear dynamics are important aspects of adaptation planning since responses of biological resources to changes in the physical climate system are not necessarily proportional and sometimes, as in the case of complex ecological systems, inherently nonlinear.

Visual examination apparatus

NASA Technical Reports Server (NTRS)

Haines, R. F.; Fitzgerald, J. W.; Rositano, S. A. (Inventor)

1973-01-01

An automated visual examination apparatus for measuring visual sensitivity and mapping blind spot location is described. The apparatus includes a projection system for displaying to a patient a series of visual stimuli, a response switch enabling him to indicate his reaction to the stimuli, and a recording system responsive to both the visual stimuli per se and the patient's response. The recording system provides a correlated permanent record of both stimuli and response from which a substantive and readily apparent visual evaluation can be made.

Bacteria-Triggered Systemic Immunity in Barley Is Associated with WRKY and ETHYLENE RESPONSIVE FACTORs But Not with Salicylic Acid1[C][W

PubMed Central

Dey, Sanjukta; Wenig, Marion; Langen, Gregor; Sharma, Sapna; Kugler, Karl G.; Knappe, Claudia; Hause, Bettina; Bichlmeier, Marlies; Babaeizad, Valiollah; Imani, Jafargholi; Janzik, Ingar; Stempfl, Thomas; Hückelhoven, Ralph; Kogel, Karl-Heinz; Mayer, Klaus F.X.

2014-01-01

Leaf-to-leaf systemic immune signaling known as systemic acquired resistance is poorly understood in monocotyledonous plants. Here, we characterize systemic immunity in barley (Hordeum vulgare) triggered after primary leaf infection with either Pseudomonas syringae pathovar japonica (Psj) or Xanthomonas translucens pathovar cerealis (Xtc). Both pathogens induced resistance in systemic, uninfected leaves against a subsequent challenge infection with Xtc. In contrast to systemic acquired resistance in Arabidopsis (Arabidopsis thaliana), systemic immunity in barley was not associated with NONEXPRESSOR OF PATHOGENESIS-RELATED GENES1 or the local or systemic accumulation of salicylic acid. Instead, we documented a moderate local but not systemic induction of abscisic acid after infection of leaves with Psj. In contrast to salicylic acid or its functional analog benzothiadiazole, local applications of the jasmonic acid methyl ester or abscisic acid triggered systemic immunity to Xtc. RNA sequencing analysis of local and systemic transcript accumulation revealed unique gene expression changes in response to both Psj and Xtc and a clear separation of local from systemic responses. The systemic response appeared relatively modest, and quantitative reverse transcription-polymerase chain reaction associated systemic immunity with the local and systemic induction of two WRKY and two ETHYLENE RESPONSIVE FACTOR (ERF)-like transcription factors. Systemic immunity against Xtc was further associated with transcriptional changes after a secondary/systemic Xtc challenge infection; these changes were dependent on the primary treatment. Taken together, bacteria-induced systemic immunity in barley may be mediated in part by WRKY and ERF-like transcription factors, possibly facilitating transcriptional reprogramming to potentiate immunity. PMID:25332505

Input-output characterization of an ultrasonic testing system by digital signal analysis

NASA Technical Reports Server (NTRS)

Karaguelle, H.; Lee, S. S.; Williams, J., Jr.

1984-01-01

The input/output characteristics of an ultrasonic testing system used for stress wave factor measurements were studied. The fundamentals of digital signal processing are summarized. The inputs and outputs are digitized and processed in a microcomputer using digital signal processing techniques. The entire ultrasonic test system, including transducers and all electronic components, is modeled as a discrete-time linear shift-invariant system. Then the impulse response and frequency response of the continuous time ultrasonic test system are estimated by interpolating the defining points in the unit sample response and frequency response of the discrete time system. It is found that the ultrasonic test system behaves as a linear phase bandpass filter. Good results were obtained for rectangular pulse inputs of various amplitudes and durations and for tone burst inputs whose center frequencies are within the passband of the test system and for single cycle inputs of various amplitudes. The input/output limits on the linearity of the system are determined.

Impaired Memory Retrieval Correlates with Individual Differences in Cortisol Response but Not Autonomic Response

ERIC Educational Resources Information Center

Tranel, Daniel; Adolphs, Ralph; Buchanan, Tony W.

2006-01-01

Stress can enhance or impair memory performance. Both cortisol release and sympathetic nervous system responses have been implicated in these differential effects. Here we investigated how memory retrieval might be affected by stress-induced cortisol release, independently of sympathetic nervous system stress responses. Thirty-two healthy…

The systemic inflammatory response syndrome.

PubMed

Robertson, Charles M; Coopersmith, Craig M

2006-04-01

The systemic inflammatory response syndrome (SIRS) is the body's response to an infectious or noninfectious insult. Although the definition of SIRS refers to it as an "inflammatory" response, it actually has pro- and anti-inflammatory components. This review outlines the pathophysiology of SIRS and highlights potential targets for future therapeutic intervention in patients with this complex entity.

48 CFR 6103.303 - Responses to claims [Rule 303].

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 7 2010-10-01 2010-10-01 false Responses to claims [Rule 303]. 6103.303 Section 6103.303 Federal Acquisition Regulations System CIVILIAN BOARD OF CONTRACT APPEALS, GENERAL SERVICES ADMINISTRATION TRANSPORTATION RATE CASES 6103.303 Responses to claims [Rule 303]. (a) Content of responses. Within 30 calenda...

Using Classroom Response Technology to Create an Active Learning Environment in Marketing Classes

ERIC Educational Resources Information Center

Muncy, James A.; Eastman, Jacqueline K.

2012-01-01

Classroom response systems (CRS), also called student/audience response systems or clickers, have been used by business instructors, particularly in larger classes, to allow instructors to ask students questions in class and have their responses immediately tabulated and reported electronically. While clickers have typically been used to measure…

Single-channel voice-response-system program documentation volume I : system description

DOT National Transportation Integrated Search

1977-01-01

This report documents the design and implementation of a Voice Response System (VRS) using Adaptive Differential Pulse Code Modulation (ADPCM) voice coding. Implemented on a Digital Equipment Corporation PDP-11/20,R this VRS system supports a single ...

48 CFR 715.303 - Responsibilities.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Responsibilities. 715.303 Section 715.303 Federal Acquisition Regulations System AGENCY FOR INTERNATIONAL DEVELOPMENT CONTRACTING METHODS AND CONTRACT TYPES CONTRACTING BY NEGOTIATION Source Selection 715.303 Responsibilities. ...

48 CFR 715.303 - Responsibilities.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Responsibilities. 715.303 Section 715.303 Federal Acquisition Regulations System AGENCY FOR INTERNATIONAL DEVELOPMENT CONTRACTING METHODS AND CONTRACT TYPES CONTRACTING BY NEGOTIATION Source Selection 715.303 Responsibilities. ...

System level analysis and control of manufacturing process variation

DOEpatents

Hamada, Michael S.; Martz, Harry F.; Eleswarpu, Jay K.; Preissler, Michael J.

2005-05-31

A computer-implemented method is implemented for determining the variability of a manufacturing system having a plurality of subsystems. Each subsystem of the plurality of subsystems is characterized by signal factors, noise factors, control factors, and an output response, all having mean and variance values. Response models are then fitted to each subsystem to determine unknown coefficients for use in the response models that characterize the relationship between the signal factors, noise factors, control factors, and the corresponding output response having mean and variance values that are related to the signal factors, noise factors, and control factors. The response models for each subsystem are coupled to model the output of the manufacturing system as a whole. The coefficients of the fitted response models are randomly varied to propagate variances through the plurality of subsystems and values of signal factors and control factors are found to optimize the output of the manufacturing system to meet a specified criterion.

Dissociating Medial Temporal and Striatal Memory Systems With a Same/Different Matching Task: Evidence for Two Neural Systems in Human Recognition.

PubMed

Sinha, Neha; Glass, Arnold Lewis

2017-01-01

The medial temporal lobe and striatum have both been implicated as brain substrates of memory and learning. Here, we show dissociation between these two memory systems using a same/different matching task, in which subjects judged whether four-letter strings were the same or different. Different RT was determined by the left-to-right location of the first letter different between the study and test string, consistent with a left-to-right comparison of the study and test strings, terminating when a difference was found. This comparison process results in same responses being slower than different responses. Nevertheless, same responses were faster than different responses. Same responses were associated with hippocampus activation. Different responses were associated with both caudate and hippocampus activation. These findings are consistent with the dual-system hypothesis of mammalian memory and extend the model to human visual recognition.

Colombeau algebra as a mathematical tool for investigating step load and step deformation of systems of nonlinear springs and dashpots

NASA Astrophysics Data System (ADS)

Průša, Vít; Řehoř, Martin; Tůma, Karel

2017-02-01

The response of mechanical systems composed of springs and dashpots to a step input is of eminent interest in the applications. If the system is formed by linear elements, then its response is governed by a system of linear ordinary differential equations. In the linear case, the mathematical method of choice for the analysis of the response is the classical theory of distributions. However, if the system contains nonlinear elements, then the classical theory of distributions is of no use, since it is strictly limited to the linear setting. Consequently, a question arises whether it is even possible or reasonable to study the response of nonlinear systems to step inputs. The answer is positive. A mathematical theory that can handle the challenge is the so-called Colombeau algebra. Building on the abstract result by Průša and Rajagopal (Int J Non-Linear Mech 81:207-221, 2016), we show how to use the theory in the analysis of response of nonlinear spring-dashpot and spring-dashpot-mass systems.

Simultaneous approach using systemic, mucosal and transcutaneous routes of immunization for development of protective HIV-1 vaccines.

PubMed

Belyakov, I M; Ahlers, J D

2011-01-01

Mucosal tissues are major sites of HIV entry and initial infection. Induction of a local mucosal cytotoxic T lymphocyte response is considered an important goal in developing an effective HIV vaccine. In addition, activation and recruitment of memory CD4(+) and CD8(+) T cells in systemic lymphoid circulation to mucosal effector sites might provide the firewall needed to prevent virus spread. Therefore a vaccine that generates CD4(+) and CD8(+) responses in both mucosal and systemic tissues might be required for protection against HIV. However, optimal routes and number of vaccinations required for the generation of long lasting CD4(+) and CD8(+) CTL effector and memory responses are not well understood especially for mucosal T cells. A number of studies looking at protective immune responses against diverse mucosal pathogens have shown that mucosal vaccination is necessary to induce a compartmentalized immune response including maximum levels of mucosal high-avidity CD8(+) CTL, antigen specific mucosal antibodies titers (especially sIgA), as well as induction of innate anti-viral factors in mucosa tissue. Immune responses are detectable at mucosal sites after systemic delivery of vaccine, and prime boost regimens can amplify the magnitude of immune responses in mucosal sites and in systemic lymphoid tissues. We believe that the most optimal mucosal and systemic HIV/SIV specific protective immune responses and innate factors might best be achieved by simultaneous mucosal and systemic prime and boost vaccinations. Similar principals of vaccination may be applied for vaccine development against cancer and highly invasive pathogens that lead to chronic infection.

Long-term culture change related to rapid response system implementation.

PubMed

Stevens, Jennifer; Johansson, Anna; Lennes, Inga; Hsu, Douglas; Tess, Anjala; Howell, Michael

2014-12-01

Increasing attention to patient safety in training hospitals may come at the expense of trainee autonomy and professional growth. This study sought to examine changes in medical trainees' self-reported behaviour after the institution-wide implementation of a rapid response system. We conducted a two-point cross-sectional survey of medical trainees in 2006, during the implementation of a rapid response system, and in 2010, in a single academic medical centre. A novel instrument was used to measure trainee likelihood of calling for supervisory assistance, perception of autonomy, and comfort in managing decompensating patients. Non-parametric tests to assess for change were used and year of training was evaluated as an effect modifier. Response rates were 38% in 2006 and 70% in 2010. After 5 years of the full implementation of the rapid response system, residents were significantly more likely to report calling their attending physicians for assistance (rising from 40% to 65% of relevant situations; p < 0.0001). Year of training was a significant effect modifier. Interns felt significantly more comfortable in managing acutely ill patients; juniors and seniors felt significantly less concerned about their autonomy at 5 years after the implementation of the rapid response system. These changes were mirrored in the actual use of the rapid response system, which increased by 41% during the 5-year period after adjustment for patient volume (p < 0.0001). A primary team-focused implementation of a rapid response system was associated with durable changes in resident physicians' reported behaviour, including increased comfort with involving more experienced physicians and managing unstable patients. © 2014 John Wiley & Sons Ltd.

Hypoxia Responsive Drug Delivery Systems in Tumor Therapy.

PubMed

Alimoradi, Houman; Matikonda, Siddharth S; Gamble, Allan B; Giles, Gregory I; Greish, Khaled

2016-01-01

Hypoxia is a common characteristic of solid tumors. It is mainly determined by low levels of oxygen resulting from imperfect vascular networks supplying most tumors. In an attempt to improve the present chemotherapeutic treatment and reduce associated side effects, several prodrug strategies have been introduced to achieve hypoxia-specific delivery of cytotoxic anticancer agents. With the advances in nanotechnology, novel delivery systems activated by the consequent outcomes of hypoxia have been developed. However, developing hypoxia responsive drug delivery systems (which only depend on low oxygen levels) is currently naïve. This review discusses four main hypoxia responsive delivery systems: polymeric based drug delivery systems, oxygen delivery systems combined with radiotherapy and chemotherapy, anaerobic bacteria which are used for delivery of genes to express anticancer proteins such as tumor necrosis alpha (TNF-α) and hypoxia-inducible transcription factors 1 alpha (HIF1α) responsive gene delivery systems.

Responsiveness of the health insurance and private systems in Alexandria, Egypt.

PubMed

Mosallam, Rasha A; Aly, Mahmoud M; Moharram, Ahmed M

2013-04-01

Responsiveness to patients is a key indicator for measuring the health system performance with respect to nonhealth aspects. This study aimed to compare responsiveness of the Health Insurance Organization (HIO) with the private healthcare system and also to assess the importance of the different responsiveness domains according to the study population's perspective. Patients attending both inpatient and outpatient settings of both organizations were interviewed (200 outpatients and 200 inpatients from each selected hospital) using the WHO questionnaire. The questionnaire elicits the ratings of the respondents on their experiences with the healthcare system over the past 12 months in terms of responsiveness domains, respondents' inability to access medical care because of financial barriers, and their ranking of the relative importance of responsiveness domains. Almost twice the number of HIO participants reported poor responsiveness compared with the private organization participants (27.8 vs. 56.8%, respectively). The outpatient setting scored much favorably compared with the inpatient setting at the HIO (52.3% of respondents reported poor responsiveness in the outpatient setting compared with 76.3% in the inpatient setting); however, they were comparable in the private setting. Communication, prompt attention, and dignity were the domains most frequently rated as the most important (36.0, 32.0, and 14.7%, respectively). The type of organization (HIO vs. private organization) and setting of care (inpatient vs. outpatient) were significant predictors of responsiveness score (P<0.001). The overall rating of the patients on responsiveness of the HIO system is low, especially when compared with the private sector. The results emphasize the importance of establishment of systems for monitoring the performance of the providers and discontinuation of the services for the nonperformers.

Combined local and systemic bleomycin administration in electrochemotherapy to reduce the number of treatment sessions

PubMed Central

Tellado, Matias; Olaiz, Nahuel; Michinski, Sebastian; Marshall, Guillermo

2016-01-01

Background Electrochemotherapy (ECT), a medical treatment widely used in human patients for tumor treatment, increases bleomycin toxicity by 1000 fold in the treated area with an objective response rate of around 80%. Despite its high response rate, there are still 20% of cases in which the patients are not responding. This could be ascribed to the fact that bleomycin, when administered systemically, is not reaching the whole tumor mass properly because of the characteristics of tumor vascularization, in which case local administration could cover areas that are unreachable by systemic administration. Patients and methods We propose combined bleomycin administration, both systemic and local, using companion animals as models. We selected 22 canine patients which failed to achieve a complete response after an ECT treatment session. Eleven underwent another standard ECT session (control group), while 11 received a combined local and systemic administration of bleomycin in the second treatment session. Results According to the WHO criteria, the response rates in the combined administration group were: complete response (CR) 54% (6), partial response (PR) 36% (4), stable disease (SD) 10% (1). In the control group, these were: CR 0% (0), PR 19% (2), SD 63% (7), progressive disease (PD) 18% (2). In the combined group 91% objective responses (CR+PR) were obtained. In the control group 19% objective responses were obtained. The difference in the response rate between the treatment groups was significant (p < 0.01). Conclusions Combined local and systemic bleomycin administration was effective in previously to ECT non responding canine patients. The results indicate that this approach could be useful and effective in specific population of patients and reduce the number of treatment sessions needed to obtain an objective response. PMID:27069450

Sex differences in physiological reactivity to acute psychosocial stress in adolescence.

PubMed

Ordaz, Sarah; Luna, Beatriz

2012-08-01

Females begin to demonstrate greater negative affective responses to stress than males in adolescence. This may reflect the concurrent emergence of underlying differences in physiological response systems, including corticolimbic circuitries, the hypothalamic-pituitary-adrenal axis (HPAA), and the autonomic nervous system (ANS). This review examines when sex differences in physiological reactivity to acute psychosocial stress emerge and the directionality of these differences over development. Indeed, the literature indicates that sex differences emerge during adolescence and persist into adulthood for all three physiological response systems. However, the directionality of the differences varies by system. The emerging corticolimbic reactivity literature suggests greater female reactivity, particularly in limbic regions densely innervated by gonadal hormone receptors. In contrast, males generally show higher levels of HPAA and ANS reactivity. We argue that the contrasting directionality of corticolimbic and peripheral physiological responses may reflect specific effects of gonadal hormones on distinct systems and also sex differences in evolved behavioral responses that demand different levels of peripheral physiological activation. Studies that examine both subjective reports of negative affect and physiological responses indicate that beginning in adolescence, females respond to acute stressors with more intense negative affect than males despite their comparatively lower peripheral physiological responses. This dissociation is not clearly explained by sex differences in the strength of the relationship between physiological and subjective responses. We suggest that females' greater subjective responsivity may instead arise from a greater activity in brain regions that translate stress responses to subjective awareness in adolescence. Future research directions include investigations of the role of pubertal hormones in physiological reactivity across all systems, examining the relationship of corticolimbic reactivity and negative affect, and sex differences in emotion regulation processes. Copyright © 2012 Elsevier Ltd. All rights reserved.

Sex differences in physiological reactivity to acute psychosocial stress in adolescence

PubMed Central

Ordaz, Sarah; Luna, Beatriz

2012-01-01

Summary Females begin to demonstrate greater negative affective responses to stress than males in adolescence. This may reflect the concurrent emergence of underlying differences in physiological response systems, including corticolimbic circuitries, the hypothalamic—pituitary— adrenal axis (HPAA), and the autonomic nervous system (ANS). This review examines when sex differences in physiological reactivity to acute psychosocial stress emerge and the directionality of these differences over development. Indeed, the literature indicates that sex differences emerge during adolescence and persist into adulthood for all three physiological response systems. However, the directionality of the differences varies by system. The emerging corti-colimbic reactivity literature suggests greater female reactivity, particularly in limbic regions densely innervated by gonadal hormone receptors. In contrast, males generally show higher levels of HPAA and ANS reactivity. We argue that the contrasting directionality of corticolimbic and peripheral physiological responses may reflect specific effects of gonadal hormones on distinct systems and also sex differences in evolved behavioral responses that demand different levels of peripheral physiological activation. Studies that examine both subjective reports of negative affect and physiological responses indicate that beginning in adolescence, females respond to acute stressors with more intense negative affect than males despite their comparatively lower peripheral physiological responses. This dissociation is not clearly explained by sex differences in the strength of the relationship between physiological and subjective responses. We suggest that females' greater subjective responsivity may instead arise from a greater activity in brain regions that translate stress responses to subjective awareness in adolescence. Future research directions include investigations of the role of pubertal hormones in physiological reactivity across all systems, examining the relationship of corticolimbic reactivity and negative affect, and sex differences in emotion regulation processes. PMID:22281210

Brain reward system's alterations in response to food and monetary stimuli in overweight and obese individuals.

PubMed

Verdejo-Román, Juan; Vilar-López, Raquel; Navas, Juan F; Soriano-Mas, Carles; Verdejo-García, Antonio

2017-02-01

The brain's reward system is crucial to understand obesity in modern society, as increased neural responsivity to reward can fuel the unhealthy food choices that are driving the growing obesity epidemic. Brain's reward system responsivity to food and monetary rewards in individuals with excessive weight (overweight and obese) versus normal weight controls, along with the relationship between this responsivity and body mass index (BMI) were tested. The sample comprised 21 adults with obesity (BMI > 30), 21 with overweight (BMI between 25 and 30), and 39 with normal weight (BMI < 25). Participants underwent a functional magnetic resonance imaging (fMRI) session while performing two tasks that involve the processing of food (Willing to Pay) and monetary rewards (Monetary Incentive Delay). Neural activations within the brain reward system were compared across the three groups. Curve fit analyses were conducted to establish the association between BMI and brain reward system's response. Individuals with obesity had greater food-evoked responsivity in the dorsal and ventral striatum compared with overweight and normal weight groups. There was an inverted U-shape association between BMI and monetary-evoked responsivity in the ventral striatum, medial frontal cortex, and amygdala; that is, individuals with BMIs between 27 and 32 had greater responsivity to monetary stimuli. Obesity is associated with greater food-evoked responsivity in the ventral and dorsal striatum, and overweight is associated with greater monetary-evoked responsivity in the ventral striatum, the amygdala, and the medial frontal cortex. Findings suggest differential reactivity of the brain's reward system to food versus monetary rewards in obesity and overweight. Hum Brain Mapp 38:666-677, 2017. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

Response Diversity and Resilience in Social-Ecological Systems

PubMed Central

Leslie, Paul; McCabe, J. Terrence

2013-01-01

Recent work in ecology suggests that the diversity of responses to environmental change among species contributing to the same ecosystem function can strongly influence ecosystem resilience. To render this important realization more useful for understanding coupled human-natural systems, we broaden the concept of response diversity to include heterogeneity in human decisions and action. Simply put, not all actors respond the same way to challenges, opportunities, and risks. The range, prevalence, and spatial and temporal distributions of different responses may be crucial to the resilience or the transformation of a social-ecological system, and thus have a bearing on human vulnerability and well-being in the face of environmental, socioeconomic, and political change. Response diversity can be seen at multiple scales (e.g., household, village, region) and response diversity at one scale may act synergistically with or contrary to the effects of diversity at another scale. Although considerable research on the sources of response diversity has been done, our argument is that the consequences of response diversity warrant closer attention. We illustrate this argument with examples drawn from our studies of two East African pastoral populations and discuss the relationship of response diversity to characteristics of social-ecological systems that can promote or diminish resilience. PMID:24855324

Investigation of absolute and relative response for three different liquid chromatography/tandem mass spectrometry systems; the impact of ionization and detection saturation.

PubMed

Nilsson, Lars B; Skansen, Patrik

2012-06-30

The investigations in this article were triggered by two observations in the laboratory; for some liquid chromatography/tandem mass spectrometry (LC/MS/MS) systems it was possible to obtain linear calibration curves for extreme concentration ranges and for some systems seemingly linear calibration curves gave good accuracy at low concentrations only when using a quadratic regression function. The absolute and relative responses were tested for three different LC/MS/MS systems by injecting solutions of a model compound and a stable isotope labeled internal standard. The analyte concentration range for the solutions was 0.00391 to 500 μM (128,000×), giving overload of the chromatographic column at the highest concentrations. The stable isotope labeled internal standard concentration was 0.667 μM in all samples. The absolute response per concentration unit decreased rapidly as higher concentrations were injected. The relative response, the ratio for the analyte peak area to the internal standard peak area, per concentration unit was calculated. For system 1, the ionization process was found to limit the response and the relative response per concentration unit was constant. For systems 2 and 3, the ion detection process was the limiting factor resulting in decreasing relative response at increasing concentrations. For systems behaving like system 1, simple linear regression can be used for any concentration range while, for systems behaving like systems 2 and 3, non-linear regression is recommended for all concentration ranges. Another consequence is that the ionization capacity limited systems will be insensitive to matrix ion suppression when an ideal internal standard is used while the detection capacity limited systems are at risk of giving erroneous results at high concentrations if the matrix ion suppression varies for different samples in a run. Copyright © 2012 John Wiley & Sons, Ltd.

Improving Systems Engineering Effectiveness in Rapid Response Development Environments

DTIC Science & Technology

2012-06-02

environments where large, complex, brownfield systems of systems are evolved through parallel development of new capabilities in response to external, time...license 14. ABSTRACT Systems engineering is often ineffective in development environments where large, complex, brownfield systems of systems are...IEEE Press, Hoboken, NJ, 2008 [18] Boehm, B.: Applying the Incremental Commitment Model to Brownfield Systems Development, Proceedings, CSER 2009

Long-Range Activation of Systemic Immunity through Peptidoglycan Diffusion in Drosophila

PubMed Central

Gendrin, Mathilde; Welchman, David P.; Poidevin, Mickael; Hervé, Mireille; Lemaitre, Bruno

2009-01-01

The systemic immune response of Drosophila is known to be induced both by septic injury and by oral infection with certain bacteria, and is characterized by the secretion of antimicrobial peptides (AMPs) into the haemolymph. To investigate other possible routes of bacterial infection, we deposited Erwinia carotovora (Ecc15) on various sites of the cuticle and monitored the immune response via expression of the AMP gene Diptericin. A strong response was observed to deposition on the genital plate of males (up to 20% of a septic injury response), but not females. We show that the principal response to genital infection is systemic, but that some AMPs, particularly Defensin, are induced locally in the genital tract. At late time points we detected bacteria in the haemolymph of immune deficient RelishE20 flies, indicating that the genital plate can be a route of entry for pathogens, and that the immune response protects flies against the progression of genital infection. The protective role of the immune response is further illustrated by our observation that RelishE20 flies exhibit significant lethality in response to genital Ecc15 infections. We next show that a systemic immune response can be induced by deposition of the bacterial elicitor peptidoglycan (PGN), or its terminal monomer tracheal cytotoxin (TCT), on the genital plate. This immune response is downregulated by PGRP-LB and Pirk, known regulators of the Imd pathway, and can be suppressed by the overexpression of PGRP-LB in the haemolymph compartment. Finally, we provide strong evidence that TCT can activate a systemic response by crossing epithelia, by showing that radiolabelled TCT deposited on the genital plate can subsequently be detected in the haemolymph. Genital infection is thus an intriguing new model for studying the systemic immune response to local epithelial infections and a potential route of entry for naturally occurring pathogens of Drosophila. PMID:20019799

Long-range activation of systemic immunity through peptidoglycan diffusion in Drosophila.

PubMed

Gendrin, Mathilde; Welchman, David P; Poidevin, Mickael; Hervé, Mireille; Lemaitre, Bruno

2009-12-01

The systemic immune response of Drosophila is known to be induced both by septic injury and by oral infection with certain bacteria, and is characterized by the secretion of antimicrobial peptides (AMPs) into the haemolymph. To investigate other possible routes of bacterial infection, we deposited Erwinia carotovora (Ecc15) on various sites of the cuticle and monitored the immune response via expression of the AMP gene Diptericin. A strong response was observed to deposition on the genital plate of males (up to 20% of a septic injury response), but not females. We show that the principal response to genital infection is systemic, but that some AMPs, particularly Defensin, are induced locally in the genital tract. At late time points we detected bacteria in the haemolymph of immune deficient Relish(E20) flies, indicating that the genital plate can be a route of entry for pathogens, and that the immune response protects flies against the progression of genital infection. The protective role of the immune response is further illustrated by our observation that Relish(E20) flies exhibit significant lethality in response to genital Ecc15 infections. We next show that a systemic immune response can be induced by deposition of the bacterial elicitor peptidoglycan (PGN), or its terminal monomer tracheal cytotoxin (TCT), on the genital plate. This immune response is downregulated by PGRP-LB and Pirk, known regulators of the Imd pathway, and can be suppressed by the overexpression of PGRP-LB in the haemolymph compartment. Finally, we provide strong evidence that TCT can activate a systemic response by crossing epithelia, by showing that radiolabelled TCT deposited on the genital plate can subsequently be detected in the haemolymph. Genital infection is thus an intriguing new model for studying the systemic immune response to local epithelial infections and a potential route of entry for naturally occurring pathogens of Drosophila.

Assessment of the effects of student response systems on student learning and attitudes over a broad range of biology courses.

PubMed

Preszler, Ralph W; Dawe, Angus; Shuster, Charles B; Shuster, Michèle

2007-01-01

With the advent of wireless technology, new tools are available that are intended to enhance students' learning and attitudes. To assess the effectiveness of wireless student response systems in the biology curriculum at New Mexico State University, a combined study of student attitudes and performance was undertaken. A survey of students in six biology courses showed that strong majorities of students had favorable overall impressions of the use of student response systems and also thought that the technology improved their interest in the course, attendance, and understanding of course content. Students in lower-division courses had more strongly positive overall impressions than did students in upper-division courses. To assess the effects of the response systems on student learning, the number of in-class questions was varied within each course throughout the semester. Students' performance was compared on exam questions derived from lectures with low, medium, or high numbers of in-class questions. Increased use of the response systems in lecture had a positive influence on students' performance on exam questions across all six biology courses. Students not only have favorable opinions about the use of student response systems, increased use of these systems increases student learning.

State deadbeat response and observability in multi-modal systems

NASA Technical Reports Server (NTRS)

Conner, L. T., Jr.; Stanford, D. P.

1984-01-01

Two aspects of multimodal systems are examined. It is shown that any completely controllable system with state dimension n not exceeding three allows a choice of feedback matrices resulting in a state deadbeat response. Some of the results presented here are valid for arbitrary n, and it is suggested that for all n the state deadbeat response can be obtained under the hypothesis of complete controllability. The controllability canonical form for a multimodal system is refined by introducing a notion of observability which is dual to controllability for these systems.

Field camera measurements of gradient and shim impulse responses using frequency sweeps.

PubMed

Vannesjo, S Johanna; Dietrich, Benjamin E; Pavan, Matteo; Brunner, David O; Wilm, Bertram J; Barmet, Christoph; Pruessmann, Klaas P

2014-08-01

Applications of dynamic shimming require high field fidelity, and characterizing the shim field dynamics is therefore necessary. Modeling the system as linear and time-invariant, the purpose of this work was to measure the impulse response function with optimal sensitivity. Frequency-swept pulses as inputs are analyzed theoretically, showing that the sweep speed is a key factor for the measurement sensitivity. By adjusting the sweep speed it is possible to achieve any prescribed noise profile in the measured system response. Impulse response functions were obtained for the third-order shim system of a 7 Tesla whole-body MR scanner. Measurements of the shim fields were done with a dynamic field camera, yielding also cross-term responses. The measured shim impulse response functions revealed system characteristics such as response bandwidth, eddy currents and specific resonances, possibly of mechanical origin. Field predictions based on the shim characterization were shown to agree well with directly measured fields, also in the cross-terms. Frequency sweeps provide a flexible tool for shim or gradient system characterization. This may prove useful for applications involving dynamic shimming by yielding accurate estimates of the shim fields and a basis for setting shim pre-emphasis. Copyright © 2013 Wiley Periodicals, Inc.

Preparedness and Emergency Response Research Centers: Using a Public Health Systems Approach to Improve All-Hazards Preparedness and Response

PubMed Central

Leinhos, Mary; Williams-Johnson, Mildred

2014-01-01

In 2008, at the request of the Centers for Disease Control and Prevention (CDC), the Institute of Medicine (IOM) prepared a report identifying knowledge gaps in public health systems preparedness and emergency response and recommending near-term priority research areas. In accordance with the Pandemic and All-Hazards Preparedness Act mandating new public health systems research for preparedness and emergency response, CDC provided competitive awards establishing nine Preparedness and Emergency Response Research Centers (PERRCs) in accredited U.S. schools of public health. The PERRCs conducted research in four IOM-recommended priority areas: (1) enhancing the usefulness of public health preparedness and response (PHPR) training, (2) creating and maintaining sustainable preparedness and response systems, (3) improving PHPR communications, and (4) identifying evaluation criteria and metrics to improve PHPR for all hazards. The PERRCs worked closely with state and local public health, community partners, and advisory committees to produce practice-relevant research findings. PERRC research has generated more than 130 peer-reviewed publications and nearly 80 practice and policy tools and recommendations with the potential to significantly enhance our nation's PHPR to all hazards and that highlight the need for further improvements in public health systems. PMID:25355970

Narrow band noise response of a Belleville spring resonator.

PubMed

Lyon, Richard H

2013-09-01

This study of nonlinear dynamics includes (i) an identification of quasi-steady states of response using equivalent linearization, (ii) the temporal simulation of the system using Heun's time step procedure on time domain analytic signals, and (iii) a laboratory experiment. An attempt has been made to select material and measurement parameters so that nearly the same systems are used and analyzed for all three parts of the study. This study illustrates important features of nonlinear response to narrow band excitation: (a) states of response that the system can acquire with transitions of the system between those states, (b) the interaction between the noise source and the vibrating load in which the source transmits energy to or draws energy from the load as transitions occur; (c) the lag or lead of the system response relative to the source as transitions occur that causes the average frequencies of source and response to differ; and (d) the determination of the state of response (mass or stiffness controlled) by observation of the instantaneous phase of the influence function. These analyses take advantage of the use of time domain analytic signals that have a complementary role to functions that are analytic in the frequency domain.

The stress response and immune system share, borrow, and reconfigure their physiological network elements: Evidence from the insects.

PubMed

Adamo, Shelley A

2017-02-01

The classic biomedical view is that stress hormone effects on the immune system are largely pathological, especially if the stress is chronic. However, more recent interpretations have focused on the potential adaptive function of these effects. This paper examines stress response-immune system interactions from a physiological network perspective, using insects because of their simpler physiology. For example, stress hormones can reduce disease resistance, yet activating an immune response results in the release of stress hormones in both vertebrates and invertebrates. From a network perspective, this phenomenon is consistent with the 'sharing' of the energy-releasing ability of stress hormones by both the stress response and the immune system. Stress-induced immunosuppression is consistent with the stress response 'borrowing' molecular components from the immune system to increase the capacity of stress-relevant physiological processes (i.e. a trade off). The insect stress hormones octopamine and adipokinetic hormone can also 'reconfigure' the immune system to help compensate for the loss of some of the immune system's molecular resources (e.g. apolipophorin III). This view helps explain seemingly maladaptive interactions between the stress response and immune system. The adaptiveness of stress hormone effects on individual immune components may be apparent only from the perspective of the whole organism. These broad principles will apply to both vertebrates and invertebrates. Copyright © 2016 Elsevier Inc. All rights reserved.

Linking Health System Responsiveness to Political Rights and Civil Liberties: A Multilevel Analysis Using Data From 44 Countries.

PubMed

Witvliet, Margot I; Stronks, Karien; Kunst, Anton E; Mahapatra, Tanmay; Arah, Onyebuchi A

2015-01-01

Responsiveness is a dimension of health system functioning and might be dependent upon contextual factors related to politics. Given this, we performed cross-national comparisons with the aim of investigating: 1) the associations of political factors with patients' reports of health system responsiveness and 2) the extent to which health input and output might explain these associations. World Health Survey data were analyzed for 44 countries (n = 103 541). Main outcomes included, respectively, 8 and 7 responsiveness domains for inpatient and outpatient care. Linear multilevel regressions were used to assess the associations of politics (namely, civil liberties and political rights), socioeconomic development, health system input, and health system output (measured by maternal mortality) with responsiveness domains, adjusted for demographic factors. Political rights showed positive associations with dignity (regression coefficient = 0.086 [standard error = 0.039]), quality (0.092 [0.049]), and support (0.113 [0.048]) for inpatient care and with dignity (0.075 [0.040]), confidentiality (0.089 [0.043]), and quality (0.124 [0.053]) for outpatient care. Positive associations were observed for civil liberties as well. Health system input and output reduced observed associations. Results tentatively suggest that strengthening political rights and, to a certain extent, civil liberties might improve health system responsiveness, in part through their effect on health system input and output. © The Author(s) 2015.

Propagating Molecular Recognition Events through Highly Integrated Sense-Response Chemical Systems

DTIC Science & Technology

2017-08-01

Propagating Molecular Recognition Events through Highly Integrated Sense-Response Chemical Systems The views, opinions and/or findings contained in...University of California - San Diego Title: Propagating Molecular Recognition Events through Highly Integrated Sense-Response Chemical Systems Report Term...including enzymatic reactions , occurring at the aqueous interfaces of thermotropic LCs show promise as the basis of biomolecular triggers of LC

Dynamic Docking Test System (DDTS) active table computer program NASA Advanced Docking System (NADS)

NASA Technical Reports Server (NTRS)

Gates, R. M.; Jantz, R. E.

1974-01-01

A computer program was developed to describe the three-dimensional motion of the Dynamic Docking Test System active table. The input consists of inertia and geometry data, actuator structural data, forcing function data, hydraulics data, servo electronics data, and integration control data. The output consists of table responses, actuator bending responses, and actuator responses.

Soil life in reconstructed ecosystems: initial soil food web responses after rebuilding a forest soil profile for a climate change experiment

Treesearch

Paul T. Rygiewicz; Vicente J. Monleon; Elaine R. Ingham; Kendall J. Martin; Mark G. Johnson

2010-01-01

Disrupting ecosystem components, while transferring and reconstructing them for experiments can produce myriad responses. Establishing the extent of these biological responses as the system approaches a new equilibrium allows us more reliably to emulate comparable native systems. That is, the sensitivity of analyzing ecosystem processes in a reconstructed system is...

A new approach for the calculation of response spectral density of a linear stationary random multidegree of freedom system

NASA Astrophysics Data System (ADS)

Sharan, A. M.; Sankar, S.; Sankar, T. S.

1982-08-01

A new approach for the calculation of response spectral density for a linear stationary random multidegree of freedom system is presented. The method is based on modifying the stochastic dynamic equations of the system by using a set of auxiliary variables. The response spectral density matrix obtained by using this new approach contains the spectral densities and the cross-spectral densities of the system generalized displacements and velocities. The new method requires significantly less computation time as compared to the conventional method for calculating response spectral densities. Two numerical examples are presented to compare quantitatively the computation time.