Reactor pressure vessel nozzle

DOEpatents

Challberg, R.C.; Upton, H.A.

1994-10-04

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

Reactor pressure vessel nozzle

DOEpatents

Challberg, Roy C.; Upton, Hubert A.

1994-01-01

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-02-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Design of virtual SCADA simulation system for pressurized water reactor

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

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles ofmore » energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.« less

Reactor pressure vessel with forged nozzles

DOEpatents

Desai, Dilip R.

1993-01-01

Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

NASA Astrophysics Data System (ADS)

Katsuyama, Jinya; Uno, Shumpei; Watanabe, Tadashi; Li, Yinsheng

2018-03-01

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

PubMed

Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

2015-02-01

Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

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

Wang, Jy-An John

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regardingmore » Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.« less

ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

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

Walter, Matthew; Yin, Shengjun; Stevens, Gary

2012-01-01

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperaturemore » (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

Influence of operating pressure on the biological hydrogen methanation in trickle-bed reactors.

PubMed

Ullrich, Timo; Lindner, Jonas; Bär, Katharina; Mörs, Friedemann; Graf, Frank; Lemmer, Andreas

2018-01-01

In order to investigate the influence of pressures up to 9bar absolute on the productivity of trickle-bed reactors for biological methanation of hydrogen and carbon dioxide, experiments were carried out in a continuously operated experimental plant with three identical reactors. The pressure increase promises a longer residence time and improved mass transfer of H 2 due to higher gas partial pressures. The study covers effects of different pressures on important parameters like gas hourly space velocity, methane formation rate, conversion rates and product gas quality. The methane content of 64.13±3.81vol-% at 1.5bar could be increased up to 86.51±0.49vol-% by raising the pressure to 9bar. Methane formation rates of up to 4.28±0.26m 3 m -3 d -1 were achieved. Thus, pressure increase could significantly improve reactor performance. Copyright © 2017 Elsevier Ltd. All rights reserved.

Upper internals arrangement for a pressurized water reactor

DOEpatents

Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

2013-07-09

In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, Paul R.

1994-01-01

A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, P.R.

1994-12-27

A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

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

Pahari, S.; Hajela, S.; Rammohan, H. P.

2012-07-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG andmore » PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)« less

Ultrasound pressure distributions generated by high frequency transducers in large reactors.

PubMed

Leong, Thomas; Coventry, Michael; Swiergon, Piotr; Knoerzer, Kai; Juliano, Pablo

2015-11-01

The performance of an ultrasound reactor chamber relies on the sound pressure level achieved throughout the system. The active volume of a high frequency ultrasound chamber can be determined by the sound pressure penetration and distribution provided by the transducers. This work evaluated the sound pressure levels and uniformity achieved in water by selected commercial scale high frequency plate transducers without and with reflector plates. Sound pressure produced by ultrasonic plate transducers vertically operating at frequencies of 400 kHz (120 W) and 2 MHz (128 W) was characterized with hydrophones in a 2 m long chamber and their effective operating distance across the chamber's vertical cross section was determined. The 2 MHz transducer produced the highest pressure amplitude near the transducer surface, with a sharp decline of approximately 40% of the sound pressure occurring in the range between 55 and 155 mm from the transducer. The placement of a reflector plate 500 mm from the surface of the transducer was shown to improve the sound pressure uniformity of 2 MHz ultrasound. Ultrasound at 400 kHz was found to penetrate the fluid up to 2 m without significant losses. Furthermore, 400 kHz ultrasound generated a more uniform sound pressure distribution regardless of the presence or absence of a reflector plate. The choice of the transducer distance to the opposite reactor wall therefore depends on the transducer plate frequency selected. Based on pressure measurements in water, large scale 400 kHz reactor designs can consider larger transducer distance to opposite wall and larger active cross-section, and therefore can reach higher volumes than when using 2 MHz transducer plates. Crown Copyright © 2015. Published by Elsevier B.V. All rights reserved.

Corrosion evaluation of N reactor pressure tube 1756

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

Larrick, A.P.

1967-10-26

N Reactor Zircaloy-2 pressure tube No. 1756 and its associated ASTM A234 steel nozzles were examined for corrosion and hydrogen content after approximately 300 days exposure in-reactor. Visual examination showed tight, adherent, dull black oxides in the pressure tube except for scratching in the bottom due to sliding of fuel and fuel spacers through the tube during charge- discharge operations. Several fretted areas up to $sup 3$/$sub 8$ inch wide by $sup 1$/$sub 2$ inch long by up to 13 mils deep were observed at the downstream end--these pits were caused by vibration of the fuel spacers against the pressuremore » tube. Hydrogen levels were fairly constant along the tube length with an average of about 19 +- 6 ppm except at one location. At approximately 30 inches from the front end of the tube a sharp peak to a maximum of 58 ppm hydrogen occurred. The reason for the peak is unknown. (auth)« less

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

In-reactor oxidation of zircaloy-4 under low water vapor pressures

NASA Astrophysics Data System (ADS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Seaweed intake and blood pressure levels in healthy pre-school Japanese children.

PubMed

Wada, Keiko; Nakamura, Kozue; Tamai, Yuya; Tsuji, Michiko; Sahashi, Yukari; Watanabe, Kaori; Ohtsuchi, Sakiko; Yamamoto, Keiko; Ando, Kyoko; Nagata, Chisato

2011-08-10

Few studies have examined whether dietary factors might affect blood pressure in children. We purposed to investigate whether seaweed intake is associated with blood pressure level among Japanese preschool children. The design of the study was cross-sectional and it was conducted in autumn 2006. Subjects were healthy preschoolers aged 3-6 years in Aichi, Japan. Blood pressure and pulse were measured once by an automated sphygmomanometer, which uses oscillometric methods. Dietary data, including seaweed intake, were assessed using 3-day dietary records covering 2 consecutive weekdays and 1 weekend day. Of a total of 533 children, 459 (86.1 percent) agreed to be enrolled in our study. Finally, blood pressure measurement, complete dietary records and parent-reported height and weight were obtained for 223 boys and 194 girls. When we examined Spearman's correlation coefficients, seaweed intake was significantly negatively related to systolic blood pressure in girls (P = 0.008). In the one-way analysis of covariance for blood pressure and pulse after adjustments for age and BMI, the boys with the lowest, middle and highest tertiles of seaweed intake had diastolic blood pressure readings of 62.8, 59.3 and 59.6 mmHg, respectively (P = 0.11, trend P = 0.038). Girls with higher seaweed intake had significantly lower systolic blood pressure readings (102.4, 99.2 and 96.9 mmHg for girls with the lowest, middle and highest tertiles of seaweed intake, respectively; P = 0.037, trend P = 0.030). Our study showed that seaweed intake was negatively related to diastolic blood pressure in boys and to systolic blood pressure in girls. This suggests that seaweed might have beneficial effects on blood pressure among children.

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

NASA Astrophysics Data System (ADS)

Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

Comparison of actinide production in traveling wave and pressurized water reactors

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

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

Japanese Experiment Module arrival

NASA Image and Video Library

2007-03-29

The Experiment Logistics Module Pressurized Section for the Japanese Experiment Module arrives at the Space Station Processing Facility. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Japanese Experiment Module arrival

NASA Image and Video Library

2007-03-29

Inside the Space Station Processing Facility, the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module is revealed after the top of the crate is removed. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Japanese Experiment Module arrival

NASA Image and Video Library

2007-03-29

The Experiment Logistics Module Pressurized Section for the Japanese Experiment Module arrives at the Space Station Processing Facility for uncrating. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft..., ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components.'' This draft LR-ISG revises...

77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft...-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components...

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

DOE PAGES

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin; ...

2016-10-26

In this paper, we present thermal-mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress-strain states are significantly higher in case of presence of crack than without crack. In conclusion, the stress-strain state under grid load following condition are more realistic compared to the stress-strainmore » state estimated assuming simplified transients.« less

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

Seaweed intake and blood pressure levels in healthy pre-school Japanese children

PubMed Central

2011-01-01

Background Few studies have examined whether dietary factors might affect blood pressure in children. We purposed to investigate whether seaweed intake is associated with blood pressure level among Japanese preschool children. Methods The design of the study was cross-sectional and it was conducted in autumn 2006. Subjects were healthy preschoolers aged 3-6 years in Aichi, Japan. Blood pressure and pulse were measured once by an automated sphygmomanometer, which uses oscillometric methods. Dietary data, including seaweed intake, were assessed using 3-day dietary records covering 2 consecutive weekdays and 1 weekend day. Of a total of 533 children, 459 (86.1 percent) agreed to be enrolled in our study. Finally, blood pressure measurement, complete dietary records and parent-reported height and weight were obtained for 223 boys and 194 girls. Results When we examined Spearman's correlation coefficients, seaweed intake was significantly negatively related to systolic blood pressure in girls (P = 0.008). In the one-way analysis of covariance for blood pressure and pulse after adjustments for age and BMI, the boys with the lowest, middle and highest tertiles of seaweed intake had diastolic blood pressure readings of 62.8, 59.3 and 59.6 mmHg, respectively (P = 0.11, trend P = 0.038). Girls with higher seaweed intake had significantly lower systolic blood pressure readings (102.4, 99.2 and 96.9 mmHg for girls with the lowest, middle and highest tertiles of seaweed intake, respectively; P = 0.037, trend P = 0.030). Conclusion Our study showed that seaweed intake was negatively related to diastolic blood pressure in boys and to systolic blood pressure in girls. This suggests that seaweed might have beneficial effects on blood pressure among children. PMID:21827710

Anticipatory control of xenon in a pressurized water reactor

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

Impink, A.J. Jr.

1987-02-10

A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals,more » control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.« less

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Efforts to reduce exposure at Japanese PWRs: CVCS improvement

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

Terada, Ryosuke

1995-03-01

Many reports have been focused on the reduction of radiation sources and related occupational exposures. The radiation sources mainly consist of corrosion products. Radiation dose rate is determined by the amount of the activated corrosion products on the surface of the primary loop components of Pressurized Water Reactor (PWR) plants. Therefore, reducing the amount of the corrosion product will contribute to the reduction of occupational exposures. In order to reduce the corrosion products, Chemical and Volume Control System (CVCS) has been improved in Japanese PWRs as follows: (a) Cation Bed Demineralizer Flowrate Control; (b) Hydrogen Peroxide Injection System; (c) Purificationmore » Flowrate During Plant Shutdown; (d) Fine Mesh Filters Upstream of Mixed Bed Demineralizers.« less

STEEL FOR PRESSURE VESSELS FOR POWER REACTORS (in German)

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

Zastrow, E.

1960-11-01

Both gas-cooled and water-cooled reactors place on the steel pressure vessel rigid requirements with respect to the design, radiation stability, gamma -induced internal stresses, and inability to, or difficulty in, repairing the vessel once it is installed. The factors to be considered in the selection of a given steel for a pressure vessel are reviewed, and the properties of steels previously used for this purpose are tabulated. The studies being raade at present to improve the desirable properties of steels for pressure vessels are briefly summarized. The corrosion stability and irradiation stability of steel are discussed. Neutron activation of themore » steel is also briefly reviewed. (J.S.R.)« less

Pressurized water nuclear reactor system with hot leg vortex mitigator

DOEpatents

Lau, Louis K. S.

1990-01-01

A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

Biofilm architecture in a novel pressurized biofilm reactor.

PubMed

Jiang, Wei; Xia, Siqing; Duan, Liang; Hermanowicz, Slawomir W

2015-01-01

A novel pure-oxygen pressurized biofilm reactor was operated at different organic loading, mechanical shear and hydrodynamic conditions to understand the relationships between biofilm architecture and its operation. The ultimate goal was to improve the performance of the biofilm reactor. The biofilm was labeled with seven stains and observed with confocal laser scanning microscopy. Unusual biofilm architecture of a ribbon embedded between two surfaces with very few points of attachment was observed. As organic loading increased, the biofilm morphology changed from a moderately rough layer into a locally smoother biomass with significant bulging protuberances, although the chemical oxygen demand (COD) removal efficiency remained unchanged at about 75%. At higher organic loadings, biofilms contained a larger fraction of active cells distributed uniformly within a proteinaceous matrix with decreasing polysaccharide content. Higher hydrodynamic shear in combination with high organic loading resulted in the collapse of biofilm structure and a substantial decrease in reactor performance (a COD removal of 16%). Moreover, the important role of proteins for the spatial distribution of active cells was demonstrated quantitatively.

Japanese Experiment Module arrival

NASA Image and Video Library

2007-03-29

Inside the Space Station Processing Facility, workers monitor progress as a huge crane is used to remove the top of the crate carrying the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Japanese Experiment Module arrival

NASA Image and Video Library

2007-03-29

Several components for delivery to the International Space Station sit in test stands inside the Space Station Processing Facility highbay. To the right, from back to front, are the Japanese Experiment Module, the Raffaello multi-purpose logistics module, and the European Space Agency's Columbus scientific research module. To the left in front is the starboard truss segment S5. Behind it is the test stand that will hold the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

NASA Astrophysics Data System (ADS)

Fekete, Balazs; Trampus, Peter

2015-09-01

The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

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

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

On-line Analysis of Catalytic Reaction Products Using a High-Pressure Tandem Micro-reactor GC/MS.

PubMed

Watanabe, Atsushi; Kim, Young-Min; Hosaka, Akihiko; Watanabe, Chuichi; Teramae, Norio; Ohtani, Hajime; Kim, Seungdo; Park, Young-Kwon; Wang, Kaige; Freeman, Robert R

2017-01-01

When a GC/MS system is coupled with a pressurized reactor, the separation efficiency and the retention time are directly affected by the reactor pressure. To keep the GC column flow rate constant irrespective of the reaction pressure, a restrictor capillary tube and an open split interface are attached between the GC injection port and the head of a GC separation column. The capability of the attached modules is demonstrated for the on-line GC/MS analysis of catalytic reaction products of a bio-oil model sample (guaiacol), produced under a pressure of 1 to 3 MPa.

IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

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

Brumovsky, M.; Steele, L.E.

1997-02-01

In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracturemore » mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.« less

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

NASA Astrophysics Data System (ADS)

Bury, Tomasz

2011-12-01

Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

CONVECTION REACTOR

DOEpatents

Hammond, R.P.; King, L.D.P.

1960-03-22

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

NASA Astrophysics Data System (ADS)

Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

2014-12-01

Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

NASA Astrophysics Data System (ADS)

Zabusov, O.

2016-04-01

In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

Proceedings: 2002 Workshop on Pressurized Water Reactor Elevated Feedwater Iron Transport

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

None

2002-11-01

Some pressurized water reactor (PWR) stations have experienced difficulty with maintaining feedwater (FW) iron concentrations below recommended concentration on a regular basis. A workshop held on September 17-18 in Dana Point, California, addressed the challenge of elevated feedwater iron transport in PWRs.

Flux effect analysis in WWER-440 reactor pressure vessel steels

NASA Astrophysics Data System (ADS)

Kryukov, A.; Blagoeva, D.; Debarberis, L.

2013-11-01

The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

Achievement of Target Blood Pressure Levels among Japanese Workers with Hypertension and Healthy Lifestyle Characteristics Associated with Therapeutic Failure.

PubMed

Kudo, Nagako; Yokokawa, Hirohide; Fukuda, Hiroshi; Sanada, Hironobu; Miwa, Yuichi; Hisaoka, Teruhiko; Isonuma, Hiroshi

2015-01-01

Few studies have examined Japanese with regard to the achievement rates for target blood pressure levels, or the relationship between these rates and healthy lifestyle characteristics in patients with hypertension as defined by the newly established hypertension management guidelines (JSH2014). The aim of this study was to elucidate achievement rates and examine healthy lifestyle characteristics associated with achievement status among Japanese. This cross-sectional study, conducted in January-December 2012, examined blood pressure control and healthy lifestyle characteristics in 8,001 Japanese workers with hypertension (mean age, 57.0 years; 78.8% were men) who participated in a workplace health checkup. Data were collected from workplace medical checkup records and participants' self-administered questionnaires. We divided into 5 groups [G1; young, middle-aged, and early-phase elderly patients (65-74 years old) without diabetes mellitus or chronic kidney disease (CKD) (<140/90 mmHg), G2; late-phase elderly patients (≥75 years old) without diabetes mellitus or CKD (<150/90 mmHg), G3; diabetic patients (<130/80 mmHg), G4; patients with CKD (<130/80 mmHg), and G5; patients with cerebrovascular and/or coronary artery diseases (<140/90 mmHg)] according to JSH2014. And then, achievement rates were calculated in each group. Multivariate analysis identified healthy lifestyle characteristics associated with "therapeutic failure" of target blood pressure. Target blood pressures were achieved by 60.2% of young, middle-aged, and early-phase elderly patients (G1), 71.4% of late-phase elderly patients (G2), 30.5% of diabetic patients (G3), 33.4% of those with chronic kidney disease (G4), and 66.0% of those with cerebrovascular and/or coronary artery diseases (G5). A body mass index of 18.5-24.9 and non-daily alcohol consumption were protective factors, and adequate sleep was found to contribute to therapeutic success. We found low achievement rates for treatment goals among

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

NASA Technical Reports Server (NTRS)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

Apparatus and process to eliminate diffusional limitations in a membrane biological reactor by pressure cycling

DOEpatents

Efthymiou, George S.; Shuler, Michael L.

1989-08-29

An improved multilayer continuous biological membrane reactor and a process to eliminate diffusional limitations in membrane reactors in achieved by causing a convective flux of nutrient to move into and out of an immobilized biocatalyst cell layer. In a pressure cycled mode, by increasing and decreasing the pressure in the respective layers, the differential pressure between the gaseous layer and the nutrient layer is alternately changed from positive to negative. The intermittent change in pressure differential accelerates the transfer of nutrient from the nutrient layers to the biocatalyst cell layer, the transfer of product from the cell layer to the nutrient layer and the transfer of byproduct gas from the cell layer to the gaseous layer. Such intermittent cycling substantially eliminates mass transfer gradients in diffusion inhibited systems and greatly increases product yield and throughput in both inhibited and noninhibited systems.

Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

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

Kulesza, J.A.; Fero, A.H.; Rouden, J.

2011-07-01

A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

Blood pressure control with cilnidipine treatment in Japanese post-stroke hypertensive patients: The CA-ATTEND study.

PubMed

Aoki, Shiro; Hosomi, Naohisa; Nezu, Tomohisa; Teshima, Tsukasa; Sugii, Hitoshi; Nagahama, Shinobu; Kurose, Yoshiki; Maruyama, Hirofumi; Matsumoto, Masayasu

2017-01-01

Blood pressure control is important in post-stroke hypertensive patients and antihypertensive treatment is recommended for such patients. Ca-channel blockers are recommended as the medications of choice for the treatment of post-stroke patients. Here, we report the results of a large-scale prospective post-marketing surveillance study of post-stroke hypertensive patients (n = 2667, male 60.4%, 69.0 ± 10.9 years) treated with cilnidipine, with regard to blood pressure control and adverse reactions. Cilnidipine treatment caused a decrease in both clinic and home blood pressures 2 months after the beginning of treatment, and the decreased blood pressure was maintained until the end of 12 months' observation. The proportion of patients in whom clinic blood pressure was well controlled (<140/90 mmHg) increased from 21.5% to 65.3% in cilnidipine treatment, with no differences in effectiveness among the various clinical subtypes of stroke. In total, 346 adverse events occurred, with an overall incidence of 8.9% (238 of 2667 patients). In the elderly group, specifically, a fall and a hip fracture each occurred in 1 (0.1%) patient. These results indicate that cilnidipine was effective in treating uncontrolled blood pressure and was well tolerated in Japanese post-stroke hypertensive patients in a real-world clinical setting.

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Radial pressure profiles in a cold‐flow gas‐solid vortex reactor

PubMed Central

Pantzali, Maria N.; Kovacevic, Jelena Z.; Marin, Guy B.; Shtern, Vladimir N.

2015-01-01

A unique normalized radial pressure profile characterizes the bed of a gas‐solid vortex reactor over a range of particle densities and sizes, solid capacities, and gas flow rates: 950–1240 kg/m3, 1–2 mm, 2 kg to maximum solids capacity, and 0.4–0.8 Nm3/s (corresponding to gas injection velocities of 55–110 m/s), respectively. The combined momentum conservation equations of both gas and solid phases predict this pressure profile when accounting for the corresponding measured particle velocities. The pressure profiles for a given type of particles and a given solids loading but for different gas injection velocities merge into a single curve when normalizing the pressures with the pressure value downstream of the bed. The normalized—with respect to the overall pressure drop—pressure profiles for different gas injection velocities in particle‐free flow merge in a unique profile. © 2015 The Authors AIChE Journal published by Wiley Periodicals, Inc. on behalf of American Institute of Chemical Engineers AIChE J, 61: 4114–4125, 2015 PMID:27667827

Japanese experiment module (JEM)

NASA Technical Reports Server (NTRS)

Kato, T.

1986-01-01

Japanese hardware elements studied during the definition phase of phase B are described. The hardware is called JEM (Japanese Experiment Module) and will be attached to the Space Station core. JEM consists of a pressurized module, an exposed facility, a scientific/equipment airlock, a local remote manipulator, and experimental logistic module. With all those hardware elements JEM will accommodate general scientific and technology development research (some of the elements are to utilize the advantage of the microgravity environment), and also accommodate control panels for the Space Station Mobile Remote Manipulator System and attached payloads.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Assemblies and methods for mitigating effects of reactor pressure vessel expansion

DOEpatents

Challberg, Roy C.; Gou, Perng-Fei; Chu, Cherk Lam; Oliver, Robert P.

1999-01-01

Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

Assemblies and methods for mitigating effects of reactor pressure vessel expansion

DOEpatents

Challberg, R.C.; Gou, P.F.; Chu, C.L.; Oliver, R.P.

1999-07-27

Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block. 6 figs.

Iron Catalyst Chemistry in High Pressure Carbon Monoxide Nanotube Reactor

NASA Technical Reports Server (NTRS)

Scott, Carl D.; Povitsky, Alexander; Dateo, Christopher; Gokcen, Tahir; Smalley, Richard E.

2001-01-01

The high-pressure carbon monoxide (HiPco) technique for producing single wall carbon nanotubes (SWNT) is analyzed using a chemical reaction model coupled with properties calculated along streamlines. Streamline properties for mixing jets are calculated by the FLUENT code using the k-e turbulent model for pure carbon monixide. The HiPco process introduces cold iron pentacarbonyl diluted in CO, or alternatively nitrogen, at high pressure, ca. 30 atmospheres into a conical mixing zone. Hot CO is also introduced via three jets at angles with respect to the axis of the reactor. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Cluster reaction rates are from Krestinin, et aI., based on shock tube measurements. Another model is from classical cluster theory given by Girshick's team. The calculations are performed on streamlines that assume that a cold mixture of Fe(CO)5 in CO is introduced along the reactor axis. Then iron forms clusters that catalyze the formation of SWNTs from the Boudouard reaction on Fe-containing clusters by reaction with CO. To simulate the chemical process along streamlines that were calculated by the fluid dynamics code FLUENT, a time history of temperature and dilution are determined along streamlines. Alternative catalyst injection schemes are also evaluated.

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

Officials welcome the arrival of the Japanese Experiment Module

NASA Image and Video Library

2007-04-17

In the Space Station Processing Facility, NASA and Japanese Aerospace and Exploration Agency (JAXA) officials welcome the arrival of the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module, or JEM, to the Kennedy Space Center. Seen here at right are JAXA representatives, including Japanese astronaut Takao Doi (center of front row), who is a crew member for mission STS-123 that will deliver the module to the space station. The new International Space Station component arrived at Kennedy March 12 to begin preparations for its future launch on mission STS-123. It will serve as an on-orbit storage area for materials, tools and supplies. It can hold up to eight experiment racks and will attach to the top of another larger pressurized module.

Skyshine radiation from a pressurized water reactor containment dome

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

Peng, W.H.

1986-06-01

The radiation dose rates resulting from airborne activities inside a postaccident pressurized water reactor containment are calculated by a discrete ordinates/Monte Carlo combined method. The calculated total dose rates and the skyshine component are presented as a function of distance from the containment at three different elevations for various gamma-ray source energies. The one-dimensional (ANISN code) is used to approximate the skyshine dose rates from the hemisphere dome, and the results are compared favorably to more rigorous results calculated by a three-dimensional Monte Carlo code.

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

NASA Astrophysics Data System (ADS)

Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

2016-04-01

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

3D magnetohydrodynamic modelling of a dc low-current plasma arc batch reactor at very high pressure in helium

NASA Astrophysics Data System (ADS)

Lebouvier, A.; Iwarere, S. A.; Ramjugernath, D.; Fulcheri, L.

2013-04-01

This paper deals with a three-dimensional (3D) time-dependent magnetohydrodynamic (MHD) model under peculiar conditions of very high pressures (from 2 MPa up to 10 MPa) and low currents (<1 A). Studies on plasma arc working under these unusual conditions remain almost unexplored because of the technical and technological challenges to develop a reactor able to sustain a plasma at very high pressures. The combined effect of plasma reactivity and high pressure would probably open the way towards new promising applications in various fields: chemistry, lightning, materials or nanomaterial synthesis. A MHD model helps one to understand the complex and coupled phenomena surrounding the plasma which cannot be understood by simply experimentation. The model also provides data which are difficult to directly determine experimentally. The model simulates an experimental-based batch reactor working with helium. The particular reactor in question was used to investigate the Fischer-Tropsch application, fluorocarbon production and CO2 retro-conversion. However, as a first approach in terms of MHD, the model considers the case for helium as a non-reactive working gas. After a detailed presentation of the model, a reference case has been fully analysed (P = 8 MPa, I = 0.35 A) in terms of physical properties. The results show a bending of the arc and displacement of the anodic arc root towards the top of the reactor, due to the combined effects of convection, gravity and electromagnetic forces. A parametric study on the pressure (2-10 MPa) and current (0.25-0.4 A) was then investigated. The operating pressure does not show an influence on the contraction of the arc but higher pressures involve a higher natural convection in the reactor, driven by the density gradients between the cold and hot gas.

Assessment of Radiation Embrittlement in Nuclear Reactor Pressure Vessel Surrogate Materials

NASA Astrophysics Data System (ADS)

Balzar, Davor

2010-10-01

The radiation-enhanced formation of small (1-2 nm) copper-rich precipitates (CRPs) is critical for the occurrence of embrittlement in nuclear-reactor pressure vessels. Small CRPs are coherent with the bcc matrix, which causes local matrix strain and interaction with the dislocation strain fields, thus impeding dislocation mobility. As CRPs grow, there is a critical size at which a phase transformation occurs, whereby the CRPs are no longer coherent with the matrix, and the strain is relieved. Diffraction-line-broadening analysis (DLBA) and small-angle neutron scattering (SANS) were used to characterize the precipitate formation in surrogate ferritic reactor-pressure vessel steels. The materials were aged for different times at elevated temperature to produce a series of specimens with different degrees of copper precipitation. SANS measurements showed that the precipitate size distribution broadens and shifts toward larger sizes as a function of ageing time. Mechanical hardness showed an increase with ageing time, followed by a decrease, which can be associated with the reduction in the number density as well as the loss of coherency at larger sizes. Inhomogeneous strain correlated with mechanical hardness.

Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

NASA Astrophysics Data System (ADS)

Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

2014-02-01

Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

REACTOR CONTROL

DOEpatents

Fortescue, P.; Nicoll, D.

1962-04-24

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Socioeconomic status is significantly associated with dietary salt intakes and blood pressure in Japanese workers (J-HOPE Study).

PubMed

Miyaki, Koichi; Song, Yixuan; Taneichi, Setsuko; Tsutsumi, Akizumi; Hashimoto, Hideki; Kawakami, Norito; Takahashi, Masaya; Shimazu, Akihito; Inoue, Akiomi; Kurioka, Sumiko; Shimbo, Takuro

2013-03-11

The association of socioeconomic status (SES) with nutrients intakes attracts public attention worldwide. In the current study, we examined the associations of SES with dietary salt intake and health outcomes in general Japanese workers (2,266) who participated in this Japanese occupational cohort. SES was assessed by a self-administered questionnaire. Dietary intakes were assessed with a validated, brief, self-administered diet history questionnaire (BDHQ). Multiple linear regression and stratified analysis were used to evaluate the associations of salt intake with the confounding factors. Education levels and household incomes were significantly associated with salt intake, as well as blood pressures (P < 0.05). After adjusting for age, sex and total energy intake, both years of education and household income significantly affect the salt intake (for education, β = -0.031, P = 0.040; for household income, β = -0.046, P = 0.003). SES factors also affect the risk of hypertension, those subjects with higher levels of education or income had lower risk to become hypertensive (ORs for education was 0.904, P < 0.001; ORs for income was 0.956, P = 0.032). Our results show that SES is an independent determinant of salt intake and blood pressure, in order to lower the risk of hypertension, the efforts to narrow the social status gaps should be considered by the health policy-makers.

Development and evaluation of a self-efficacy instrument for Japanese sleep apnea patients receiving continuous positive airway pressure treatment

PubMed Central

Saito, Ayako; Kojima, Shigeko; Sasaki, Fumihiko; Hayashi, Masamichi; Mieno, Yuki; Sakakibara, Hiroki; Hashimoto, Shuji

2015-01-01

The purpose of this study was to develop and evaluate a self-efficacy instrument for Japanese obstructive sleep apnea (OSA) patients treated with continuous positive airway pressure (CPAP). Analyzed subjects were 653 Japanese OSA patients (619 males and 34 females) treated with CPAP at a sleep laboratory in a respiratory clinic in a Japanese city. Based on Bandura’s social cognitive theory, the CPAP Self-Efficacy Questionnaire for Sleep Apnea in Japanese (CSESA-J) was developed by a focus group of experts, using a group interview of OSA patients for the items of two previous self-efficacy scales for Western sleep apnea patients receiving CPAP treatment. CSESA-J has two subscales, one for self-efficacy and the other for outcome expectancy, and consists of a total of 15 items. Content validity was confirmed by the focus group. Confirmatory factor analysis showed that the factor loadings of self-efficacy and outcome expectancy were 0.47–0.76 and 0.41–0.92, respectively, for the corresponding items. CSESA-J had a significant but weak positive association with the General Self-Efficacy Scale, and a strong positive association with “Self-efficacy scale on health behavior in patients with chronic disease.” Cronbach’s alpha coefficient was 0.85 for the self-efficacy subscale and 0.89 for the outcome expectancy subscale. The intraclass correlation coefficient using data from the first and second measurements with CSESA-J for a subset of 130 subjects was 0.93 for the self-efficacy and outcome expectancy subscales. These results support CSESA-J as a reliable and valid instrument for measuring the self-efficacy of Japanese OSA patients treated with CPAP. Further studies are warranted to confirm validity for female OSA patients and generalizability. PMID:25678832

Association of High Pulse Pressure With Proteinuria in Subjects With Diabetes, Prediabetes, or Normal Glucose Tolerance in a Large Japanese General Population Sample

PubMed Central

Yano, Yuichiro; Sato, Yuji; Fujimoto, Shouichi; Konta, Tsuneo; Iseki, Kunitoshi; Moriyama, Toshiki; Yamagata, Kunihiro; Tsuruya, Kazuhiko; Yoshida, Hideaki; Asahi, Koichi; Kurahashi, Issei; Ohashi, Yasuo; Watanabe, Tsuyoshi

2012-01-01

OBJECTIVE To examine whether there is a difference in the association between high pulse pressure and proteinuria, independent of other blood pressure (BP) indices, such as systolic or diastolic BP, among subjects with diabetes, prediabetes, or normal glucose tolerance. RESEARCH DESIGN AND METHODS Using a nationwide health checkup database of 228,778 Japanese aged ≥20 years (mean 63.2 years; 39.3% men; none had pre-existing cardiovascular disease), we examined the association between high pulse pressure, defined as the highest quintile of pulse pressure (≥63 mmHg, n = 40,511), and proteinuria (≥1+ on dipstick, n = 12,090) separately in subjects with diabetes (n = 27,913), prediabetes (n = 100,214), and normal glucose tolerance (n = 100,651). RESULTS The prevalence of proteinuria was different among subjects with diabetes, prediabetes, and normal glucose tolerance (11.3 vs. 5.0 vs. 3.9%, respectively; P < 0.001). In subjects with diabetes, but not those with prediabetes or normal glucose tolerance, high pulse pressure was associated with proteinuria independently of significant covariates, including systolic BP (odds ratio 1.15 [95% CI 1.04–1.28]) or diastolic or mean BP (all P < 0.01). In patients with diabetes, a +1 SD increase of pulse pressure (+13 mmHg) was associated with proteinuria, even after adjustment for systolic BP (1.07 [1.00–1.13]) or diastolic or mean BP (all P < 0.05). CONCLUSIONS Among the Japanese general population, there was a significant difference in the association between high pulse pressure and proteinuria among subjects with diabetes, prediabetes, and normal glucose tolerance. Only in diabetes was high pulse pressure associated with proteinuria independent of systolic, diastolic, or mean BP levels. PMID:22474041

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

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

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

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

Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin nextmore » year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.« less

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum ofmore » the individual components equaling the measured values.« less

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

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

Spencer, Benjamin; Backman, Marie; Williams, Paul

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decisionmore » making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.« less

Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

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

Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng

2015-07-01

The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures themore » effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.« less

Vibration monitoring of Kraftwerk Union pressurized water reactors - Review, present status, and further development

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

Stolben, H.; Wehling, H.J.

Incipient damage to mechanical structure may be detected early in time by deviations from normal dynamic behavior. For vibration monitoring of coupled systems, only a small number of transducers are necessary, in general. On the basis, Kraftwerk Union has been involved in the development and construction of vibration monitoring systems for pressurized water reactors over the last 20 yr. The current state of the art permits vibration monitoring during normal operation by reactor personnel without expert assistance. The new SUS-86 microprocessor-based system allows further expansion toward an expert system.

Motor-driven screwing and transporting tool for reactor pressure vessel head retaining fastenings

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

Scholz, M.

1977-09-13

The invention concerns a motor-driven screwing and transporting tool for tightening or loosening the threaded studs and associated tightening nuts of the head bolting of pressure vessels. After the tightening nuts are loosened or before they are tightened, the weight of the studs is taken over by rotating bearings that can be lifted, so that the studs with their tightening nuts can be screwed in or out, the screw threads of the studs being thus weight-relieved. The invention is intended primarily for nuclear reactor pressure vessels. 21 claims, 6 figures.

Reactor pressure vessel embrittlement: Insights from neural network modelling

NASA Astrophysics Data System (ADS)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Advanced Computational Modeling of Vapor Deposition in a High-Pressure Reactor

NASA Technical Reports Server (NTRS)

Cardelino, Beatriz H.; Moore, Craig E.; McCall, Sonya D.; Cardelino, Carlos A.; Dietz, Nikolaus; Bachmann, Klaus

2004-01-01

In search of novel approaches to produce new materials for electro-optic technologies, advances have been achieved in the development of computer models for vapor deposition reactors in space. Numerical simulations are invaluable tools for costly and difficult processes, such as those experiments designed for high pressures and microgravity conditions. Indium nitride is a candidate compound for high-speed laser and photo diodes for optical communication system, as well as for semiconductor lasers operating into the blue and ultraviolet regions. But InN and other nitride compounds exhibit large thermal decomposition at its optimum growth temperature. In addition, epitaxy at lower temperatures and subatmospheric pressures incorporates indium droplets into the InN films. However, surface stabilization data indicate that InN could be grown at 900 K in high nitrogen pressures, and microgravity could provide laminar flow conditions. Numerical models for chemical vapor deposition have been developed, coupling complex chemical kinetics with fluid dynamic properties.

Advanced Computational Modeling of Vapor Deposition in a High-pressure Reactor

NASA Technical Reports Server (NTRS)

Cardelino, Beatriz H.; Moore, Craig E.; McCall, Sonya D.; Cardelino, Carlos A.; Dietz, Nikolaus; Bachmann, Klaus

2004-01-01

In search of novel approaches to produce new materials for electro-optic technologies, advances have been achieved in the development of computer models for vapor deposition reactors in space. Numerical simulations are invaluable tools for costly and difficult processes, such as those experiments designed for high pressures and microgravity conditions. Indium nitride is a candidate compound for high-speed laser and photo diodes for optical communication system, as well as for semiconductor lasers operating into the blue and ultraviolet regions. But InN and other nitride compounds exhibit large thermal decomposition at its optimum growth temperature. In addition, epitaxy at lower temperatures and subatmospheric pressures incorporates indium droplets into the InN films. However, surface stabilization data indicate that InN could be grown at 900 K in high nitrogen pressures, and microgravity could provide laminar flow conditions. Numerical models for chemical vapor deposition have been developed, coupling complex chemical kinetics with fluid dynamic properties.

Effects of age, ethnicity and menopause on ambulatory blood pressure: Japanese-American and Caucasian school teachers in Hawaii.

PubMed

Brown, D E; Sievert, L L; Aki, S L; Mills, P S; Etrata, M B; Paopao, R N; James, G D

2001-01-01

Ambulatory blood pressure (BP) measurements of 120 female teachers of Japanese-American or Caucasian ethnicity working in public schools located in Hilo, Hawaii, were recorded. BP was measured at 15-min intervals during waking hours and 30-min intervals during sleep over a 24-hr period that included a full work day. These measurements were averaged during three daily settings: at work, at home while awake ("home"), and during sleep. ANCOVAs using ethnicity as a predictor variable of BP, with age and the body mass index (BMI) as covariates, show a significant interaction effect between age and ethnicity in some daily settings. Among Japanese-Americans partial correlations between age and systolic BP controlling for the BMI are significant in these settings, while among Caucasians none of the correlations are significant. Menopausal status is not significantly related to BP when age is controlled in analyses. There was no significant ethnic difference in number of symptoms reported, including frequency of "hot flushes/flashes," within the past two weeks. Those who reported hot flushes had significantly elevated BP in waking settings but not during sleep. The greater increase in BP with age in Japanese-Americans may be related to their elevated risk for development of hypertension. The lack of a significant relationship between menopausal status and BP may be due to the high rate of usage of hormonal replacement therapy in this sample, as well as an unusually high rate of hysterectomy.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Pressurized-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 a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

Japanese suppliers in transition from domestic nuclear reactor vendors to international suppliers

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

Forsberg, C.W.; Reich, W.J.; Rowan, W.J.

1994-06-27

Japan is emerging as a major leader and exporter of nuclear power technology. In the 1990s, Japan has the largest and strongest nuclear power supply industry worldwide as a result of the largest domestic nuclear power plant construction program. The Japanese nuclear power supply industry has moved from dependence on foreign technology to developing, design, building, and operating its own power plants. This report describes the Japanese nuclear power supply industry and examines one supplier--the Mitsubishi group--to develop an understanding of the supply industry and its relationship to the utilities, government, and other organizations.

Current status of achieving blood pressure target and its clinical correlates in Japanese type 2 diabetes.

PubMed

Yokoyama, Hiroki; Araki, Shin-Ichi; Kawai, Koichi; Hirao, Koichi; Kurihara, Yoshio; Seino, Hiroaki; Takamura, Hiroshi; Sugimoto, Hidekatsu; Okada, Akira; Maegawa, Hiroshi

2017-07-21

To investigate the current status of achieved blood pressure levels in association with the number of antihypertensive drug classes as of 2013, and to explore the clinical correlates with achievement of target blood pressure in a large-scale cohort of Japanese subjects with type 2 diabetes. A nationwide survey was conducted including 12,811 subjects with type 2 diabetes. Subjects were divided by achieved blood pressure, <130/80 or 140/90 mmHg, and the number of drug classes taken. The percentages achieving a blood pressure of <130/80 or 140/90 mmHg were 52.0% and 86.1%, respectively. The prevalence of hypertension, if defined as ≥130/80 mmHg or treated, became 67.9%. Among subjects taking antihypertensive drugs, a blood pressure of <130/80 or <140/90 mmHg was 46.7% and 83.2%, respectively. The percentages of <130/80 mmHg were 55.9% without drugs, 47.1% on 1, 42.5% on 2, 47.2% on 3, and 56.8% on ≥4 drugs, respectively. The most prescribed drugs were renin-angiotensin system inhibitors, followed by calcium channel blockers, diuretics, and β-blockers. The multiple logistic regression analysis indicated that a blood pressure <130/80 mmHg was associated with lower values in age, body mass index, albuminuria, and glomerular filtration rate, higher proportions on targets for HbA 1C and lipids, and less retinopathy. In type 2 diabetes, hypertension is common and only 52% achieved <130/80 mmHg, indicating a difficulty in blood pressure lowering. This was correlated with difficulties in glycemic and lipid management, obesity, and vascular complications, implying these clustering to be a serious problem. This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

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

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

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

Dickson, T.L.

1993-01-01

This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracturemore » Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.« less

FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

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

Dickson, T.L.

1993-04-01

This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracturemore » Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.« less

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

Calculation of Internal Pressures in the Fuel Tube of a Nuclear Reactor

NASA Technical Reports Server (NTRS)

Rosenbaum, B. M.; Allen, G.

1952-01-01

General procedures for computing internal pressures in fuel tubes of nuclear reactors are described and the effects on the pressure of varying neutron flux, fissioning material, and operating temperatures are discussed. A general proof is given that during pile operation each fission product is monotonically increasing and therefore a maximum amount of all elements is present at the time of shit down. The post-shutdown build-up of elements that are held in check during pile operation because of their inordinately high capture cross sections is calculated quantitatively. An account of chemical interactions between the many fission-product elements and the resulting effect on the total pressure completes the discussion. The general methods are illustrated by calculations applied to a system consisting of 90 percent enriched U235 in the form of UO2 packed into a hollow metal cylinder or "pin", operating at a flux of 8 x 10(exp 14) at 2000 F. Calculations of the pressure inside a pin are made with and without a sodium metal heat-transfer additive. The bulk of the pressure is shown to depend on the four elements, xenon, krypton, rubidium, and cesium; the amount of free oxygen, however, was also significant. For a shutdown time of 10(exp 6) seconds, the pressure was about 100 atmospheres.

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

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

Pope, M. A.; Sen, R. S.; Ougouag, A. M.

2012-07-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space availablemore » for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)« less

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

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

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag

2012-04-01

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel,more » the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.« less

Pressurized thermal shock: TEMPEST computer code simulation of thermal mixing in the cold leg and downcomer of a pressurized water reactor. [Creare 61 and 64

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

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

The TEMPEST computer program was used to simulate fluid and thermal mixing in the cold leg and downcomer of a pressurized water reactor under emergency core cooling high-pressure injection (HPI), which is of concern to the pressurized thermal shock (PTS) problem. Application of the code was made in performing an analysis simulation of a full-scale Westinghouse three-loop plant design cold leg and downcomer. Verification/assessment of the code was performed and analysis procedures developed using data from Creare 1/5-scale experimental tests. Results of three simulations are presented. The first is a no-loop-flow case with high-velocity, low-negative-buoyancy HPI in a 1/5-scale modelmore » of a cold leg and downcomer. The second is a no-loop-flow case with low-velocity, high-negative density (modeled with salt water) injection in a 1/5-scale model. Comparison of TEMPEST code predictions with experimental data for these two cases show good agreement. The third simulation is a three-dimensional model of one loop of a full size Westinghouse three-loop plant design. Included in this latter simulation are loop components extending from the steam generator to the reactor vessel and a one-third sector of the vessel downcomer and lower plenum. No data were available for this case. For the Westinghouse plant simulation, thermally coupled conduction heat transfer in structural materials is included. The cold leg pipe and fluid mixing volumes of the primary pump, the stillwell, and the riser to the steam generator are included in the model. In the reactor vessel, the thermal shield, pressure vessel cladding, and pressure vessel wall are thermally coupled to the fluid and thermal mixing in the downcomer. The inlet plenum mixing volume is included in the model. A 10-min (real time) transient beginning at the initiation of HPI is computed to determine temperatures at the beltline of the pressure vessel wall.« less

Japanese Competitiveness and Japanese Management.

ERIC Educational Resources Information Center

Minabe, Shigeo

1986-01-01

Analyzes and compares Japanese and American industrial policy and labor practices. Proposes that certain aspects of the Japanese system be adapted by American businesses for purpose of increasing international competitiveness. Proposes specific actions and plans for both the Japanese and American systems. (ML)

Ethnic differences in the degree of morning blood pressure surge and in its determinants between Japanese and European hypertensive subjects: data from the ARTEMIS study.

PubMed

Hoshide, Satoshi; Kario, Kazuomi; de la Sierra, Alejandro; Bilo, Grzegorz; Schillaci, Giuseppe; Banegas, José Ramón; Gorostidi, Manuel; Segura, Julian; Lombardi, Carolina; Omboni, Stefano; Ruilope, Luis; Mancia, Giuseppe; Parati, Gianfranco

2015-10-01

Morning blood pressure (BP) surge has been reported to be a prognostic factor for cardiovascular events. Its determinants are still poorly defined, however. In particular, it is not clear whether ethnic differences play a role in determining morning surge (MS) size. Aim of our study was to explore whether differences exist in the size of MS between Japanese and Western European hypertensive patients. We included 2887 untreated hypertensive patients (age 62.3±8.8 years) from a European ambulatory BP monitoring database and 811 hypertensive patients from a Japanese database (Jichi Medical School Ambulatory Blood Pressure Monitoring WAVE1, age 72.3±9.8 years) following the same inclusion criteria. Their 24-hour ambulatory BP monitoring recordings were analyzed focusing on MS. Sleep-trough MS was defined as the difference between mean systolic BP during the 2 hours after awakening and mean systolic BP during the 1-hour night period that included the lowest sleep BP level. The sleep-trough MS was higher in Japanese than in European hypertensive patients after adjusting for age and 24-hour mean BP levels (40.1 [95% confidence interval 39.0-41.2] versus 23.0 [22.4-23.5] mm Hg; P<0.001). This difference remained significant after accounting for differences in night-time BP dipping. Age was independently associated with MS in the Japanese database, but not in the European subjects. Our results for the first time show the occurrence of substantial ethnic differences in the degree of MS. These findings may help in understanding the role of ethnic factors in cardiovascular risk assessment and in identifying possible ethnicity-related differences in the most effective measures to be implemented for prevention of BP-related cardiovascular events. © 2015 American Heart Association, Inc.

Metabolic syndrome in overweight and obese Japanese children.

PubMed

Yoshinaga, Masao; Tanaka, Satoru; Shimago, Atsushi; Sameshima, Koji; Nishi, Junichiro; Nomura, Yuichi; Kawano, Yoshifumi; Hashiguchi, Jun; Ichiki, Takeo; Shimizu, Shinichiro

2005-07-01

To determine the prevalence of and sex differences related to the metabolic syndrome among obese and overweight elementary school children. Subjects were 471 overweight or obese Japanese children. Children meeting at least three of the following five criteria qualified as having the metabolic syndrome: abdominal obesity, elevated blood pressure, low high-density lipoprotein-cholesterol levels, high triglyceride levels, and high fasting glucose levels. Fasting insulin levels were also examined. Japanese obese children were found to have a significantly lower prevalence (17.7%) of the metabolic syndrome than U.S. obese adolescents (28.7%, p = 0.0014). However, Japanese overweight children had a similar incidence (8.7%) of the metabolic syndrome compared with U.S. overweight adolescents (6.8%). Hyperinsulinemia in girls and abdominal obesity in boys are characteristic features of individual metabolic syndrome factors in Japanese children. The prevalence of the metabolic syndrome is not lower in preteen Japanese overweight children than in U.S. overweight adolescents, although it is significantly lower in Japanese obese preteen children than in U.S. obese adolescents. Primary and secondary interventions are needed for overweight preteen children in Japan.

Simplified failure sequence evaluation of reactor pressure vessel head corroding in-core instrumentation assembly

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

McVicker, J.P.; Conner, J.T.; Hasrouni, P.N.

1995-11-01

In-Core Instrumentation (ICI) assemblies located on a Reactor Pressure Vessel Head have a history of boric acid leakage. The acid tends to corrode the nuts and studs which fasten the flanges of the assembly, thereby compromising the assembly`s structural integrity. This paper provides a simplified practical approach in determining the likelihood of an undetected progressing assembly stud deterioration, which would lead to a catastrophic loss of reactor coolant. The structural behavior of the In-Core Instrumentation flanged assembly is modeled using an elastic composite section assumption, with the studs transmitting tension and the pressure sealing gasket experiencing compression. Using the abovemore » technique, one can calculate the flange relative deflection and the consequential coolant loss flow rate, as well as the stress in any stud. A solved real life example develops the expected failure sequence and discusses the exigency of leak detection for safe shutdown. In the particular case of Calvert Cliffs Nuclear Power Plant (CCNPP) it is concluded that leak detection occurs before catastrophic failure of the ICI flange assembly.« less

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Officials welcome the arrival of the Japanese Experiment Module

NASA Image and Video Library

2007-04-17

In the Space Station Processing Facility, Scott Higginbotham, payload manager for the International Space Station, discusses the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module (JEM), with Dr. Hidetaka Tanaka, the JEM Project Team resident manager at KSC for the Japanese Aerospace and Exploration Agency (JAXA). Earlier, NASA and JAXA officials welcomed the arrival of the module. The new International Space Station component arrived at Kennedy March 12 to begin preparations for its future launch on mission STS-123. It will serve as an on-orbit storage area for materials, tools and supplies. It can hold up to eight experiment racks and will attach to the top of another larger pressurized module.

Pressure regulator

DOEpatents

Ebeling, Jr., Robert W.; Weaver, Robert B.

1979-01-01

The pressure within a pressurized flow reactor operated under harsh environmental conditions is controlled by establishing and maintaining a fluidized bed of uniformly sized granular material of selected density by passing the gas from the reactor upwardly therethrough at a rate sufficient to fluidize the bed and varying the height of the bed by adding granular material thereto or removing granular material therefrom to adjust the backpressure on the flow reactor.

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

NASA Technical Reports Server (NTRS)

Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

1986-01-01

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Serum glucose, cholesterol and blood pressure levels in Japanese type 1 and 2 diabetic patients: BioBank Japan.

PubMed

Yokomichi, Hiroshi; Nagai, Akiko; Hirata, Makoto; Kiyohara, Yutaka; Muto, Kaori; Ninomiya, Toshiharu; Matsuda, Koichi; Kamatani, Yoichiro; Tamakoshi, Akiko; Kubo, Michiaki; Nakamura, Yusuke; Yamagata, Zentaro

2017-03-01

Evidence of characteristics of Japanese patients with diabetes from a large-scale population is necessary. Few studies have compared glycaemic controls, complications and comorbidities between type 1 and 2 diabetic patients. This paper focuses on illustrating a clinical picture of Japanese diabetic patients and comparing glycaemic control and prognoses between type 1 and 2 diabetes using multi-institutional data. The BioBank Japan Project enrolled adult type 1 and 2 diabetic patients between fiscal years 2003 and 2007. We have presented characteristics, controls of serum glucose, cholesterol and blood pressure, prevalence of complications and comorbidities and survival curves. We have also shown glycaemic controls according to various individual profiles of diabetic patients. A total of 558 type 1 diabetic patients and 30,834 type 2 diabetic patients participated in this study. The mean glycated haemoglobin A1c was higher in type 1 diabetes than in type 2 diabetes. In the type 1 diabetic patients, the glycated haemoglobin A1c had no consistent trend according to age and body mass index. The Kaplan-Meier estimates represented a longer survival time from baseline with type 1 diabetes than with type 2 diabetes. Compared with type 1 diabetic patients, type 2 diabetic patients had double the prevalence of macrovascular complications. This work has revealed detailed plasma glucose levels of type 1 and 2 diabetic patients according to age, body mass index, blood pressure, serum cholesterol levels and smoking and drinking habits. Our data have also shown that the prognosis is worse for type 2 diabetes than for type 1 diabetes in Japan. Copyright © 2017 The Authors. Production and hosting by Elsevier B.V. All rights reserved.

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

NASA Technical Reports Server (NTRS)

Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

1973-01-01

A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

A cross-sectional study of workplace social capital and blood pressure: a multilevel analysis at Japanese manufacturing companies

PubMed Central

Fujino, Yoshihisa; Kubo, Tatsuhiko; Kunimoto, Masamizu; Tabata, Hidetoshi; Tsuchiya, Takuto; Kadowaki, Koji; Nakamura, Takehiro; Oyama, Ichiro

2013-01-01

Objectives We examined the contextual effect of workplace social capital on systolic blood pressure (SBP). Design Cross-sectional. Setting A conglomerate from 58 workplaces in Japan. Participants Of the 5844 workers at a Japanese conglomerate from 58 workplaces, 5368 were recruited. Individuals who received drugs for hypertension (n=531) and who lacked information on any variable (n=167) were excluded from the analyses, leaving 4735 individuals (3281 men and 1454 women) for inclusion. Primary and secondary outcome measures Systolic blood pressure. Results The contextual effect of workplace social capital on SBP was examined using a multilevel regression analysis with a random intercept. Coworker support had a contextual effect at the workplace level (coefficient=−1.97, p=0.043), while a lack of trust for coworkers (coefficient=0.27, p=0.039) and lack of helpfulness from coworkers were associated with SBP (coefficient=0.28, p=0.002). Conclusions The present study suggested that social capital at the workplace level has beneficial effects on SBP. PMID:23386581

A cross-sectional study of workplace social capital and blood pressure: a multilevel analysis at Japanese manufacturing companies.

PubMed

Fujino, Yoshihisa; Kubo, Tatsuhiko; Kunimoto, Masamizu; Tabata, Hidetoshi; Tsuchiya, Takuto; Kadowaki, Koji; Nakamura, Takehiro; Oyama, Ichiro

2013-01-01

We examined the contextual effect of workplace social capital on systolic blood pressure (SBP). Cross-sectional. A conglomerate from 58 workplaces in Japan. Of the 5844 workers at a Japanese conglomerate from 58 workplaces, 5368 were recruited. Individuals who received drugs for hypertension (n=531) and who lacked information on any variable (n=167) were excluded from the analyses, leaving 4735 individuals (3281 men and 1454 women) for inclusion. Systolic blood pressure. The contextual effect of workplace social capital on SBP was examined using a multilevel regression analysis with a random intercept. Coworker support had a contextual effect at the workplace level (coefficient=-1.97, p=0.043), while a lack of trust for coworkers (coefficient=0.27, p=0.039) and lack of helpfulness from coworkers were associated with SBP (coefficient=0.28, p=0.002). The present study suggested that social capital at the workplace level has beneficial effects on SBP.

Study on Material Selection of Reactor Pressure Vessel of SCWR

NASA Astrophysics Data System (ADS)

Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

Adherence to the food-based Japanese dietary guidelines in relation to metabolic risk factors in young Japanese women.

PubMed

Nishimura, Terumi; Murakami, Kentaro; Livingstone, M Barbara E; Sasaki, Satoshi; Uenishi, Kazuhiro

2015-08-28

While Japanese diets have attracted considerable attention because of, for example, the long-life expectancy in Japan, their health benefits have not been examined. In the present study, we cross-sectionally examined whether adherence to the food-based Japanese dietary guidelines is associated with metabolic risk factors in 1083 Japanese women aged 18-22 years. Based on the Japanese Food Guide Spinning Top, adherence to the food-based Japanese dietary guidelines was assessed using dietary information on consumed servings of grain dishes, vegetable dishes, fish and meat dishes, milk and fruits and energy from snacks and alcoholic beverages during the preceding month, which was derived from a comprehensive diet history questionnaire. Higher dietary adherence was associated with higher intakes of protein, carbohydrate, dietary fibre, Na, K and vitamin C, and lower intakes of total and saturated fat. There was also an inverse association between dietary adherence and dietary energy density. After adjustment for potential confounding factors, dietary adherence was inversely associated with waist circumference (P for trend = 0·002). It also showed an inverse association with LDL-cholesterol concentrations (P for trend = 0·04). There was no association with the other metabolic risk factors examined, including BMI, systolic and diastolic blood pressure, total and HDL-cholesterol, TAG, glucose, glycated Hb and insulin concentrations. In conclusion, higher adherence to the food-based Japanese dietary guidelines, which was characterised by favourable dietary intakes of foods and nutrients as well as lower energy density, was associated with lower waist circumference and LDL-cholesterol concentrations in this group of young Japanese women.

Seismic attenuation system for a nuclear reactor

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

Liszkai, Tamas; Cadell, Seth

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vesselmore » are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.« less

Neutron shielding panels for reactor pressure vessels

DOEpatents

Singleton, Norman R [Murrysville, PA

2011-11-22

In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system.

PubMed

Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

2013-01-01

Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ(13)Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate.

Subsize specimen testing of nuclear reactor pressure vessel material

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

Kumar, A.S.; Rosinski, S.T.; Cannon, N.S.

1991-01-01

A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and themore » stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.« less

Subsize specimen testing of nuclear reactor pressure vessel material

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

Kumar, A.S.; Rosinski, S.T.; Cannon, N.S.

1991-12-31

A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and themore » stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.« less

Effect of azilsartan versus candesartan on nocturnal blood pressure variation in Japanese patients with essential hypertension.

PubMed

Rakugi, Hiromi; Kario, Kazuomi; Enya, Kazuaki; Igeta, Masataka; Ikeda, Yoshinori

2013-09-01

Abnormal variations in night-time hypertension such as "non-dipping" type (< 10% decrease in nocturnal systolic blood pressure [SBP] from daytime SBP) are a risk factor for cardiovascular events independent of 24-h BP. As part of a randomized, double-blind study of azilsartan (20-40 mg once daily) and candesartan (8-12 mg once daily) in Japanese patients with essential hypertension, an exploratory analysis was performed using ambulatory BP monitoring (ABPM) at baseline and Week 14. Effects of study drugs on nocturnal BP variations according to patients' nocturnal SBP dipping status were evaluated. ABPM data were available for 273 patients treated with azilsartan and 275 with candesartan. In the dipping group (≥ 10% decrease from daytime SBP), azilsartan produced a greater reduction from baseline in daytime than in night-time SBP (- 14.1 and - 10.9 mmHg, respectively), and the change in daytime SBP was significantly greater with azilsartan than with candesartan (p = 0.0077). In the non-dipping group, azilsartan produced a greater reduction from baseline in night-time than in daytime SBP (- 20.2 and - 9.9 mmHg, respectively), and reductions in both night-time SBP (p = 0.02) and daytime SBP (p = 0.0042) were significantly greater with azilsartan than with candesartan. Once-daily azilsartan improved non-dipping night-time SBP to a greater extent than candesartan in Japanese patients with grade I-II essential hypertension.

Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

NASA Astrophysics Data System (ADS)

Kobak, J. A.; Rollbuhler, R. J.

1981-10-01

A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

NASA Technical Reports Server (NTRS)

Kobak, J. A.; Rollbuhler, R. J.

1981-01-01

A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

PRESSURE SYSTEM CONTROL

DOEpatents

Esselman, W.H.; Kaplan, G.M.

1961-06-20

The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

Breastfeeding leads to lower blood pressure in 7-year-old Japanese children: Tohoku Study of Child Development.

PubMed

Hosaka, Miki; Asayama, Kei; Staessen, Jan A; Ohkubo, Takayoshi; Hayashi, Katsuhisa; Tatsuta, Nozomi; Kurokawa, Naoyuki; Satoh, Michihiro; Hashimoto, Takanao; Hirose, Takuo; Obara, Taku; Metoki, Hirohito; Inoue, Ryusuke; Kikuya, Masahiro; Nakai, Kunihiko; Imai, Yutaka; Satoh, Hiroshi

2013-02-01

This study investigated the association between breastfeeding and both self-measured home blood pressure (HBP) and conventional blood pressure (CBP) in 7-year-old Japanese children. We obtained data pertaining to breastfeeding and blood pressure for 377 mother-offspring pairs from the Tohoku Study of Child Development, which is a prospective birth cohort study. Information on breastfeeding and other factors were obtained from parental questionnaires during the follow-up period. Based on the duration of breastfeeding as a major source of nutrition, mother-offspring pairs were divided into short-term (mean, 5.1 months) and long-term (mean, 11.3 months) breastfeeding groups. At the age of 7 years (84.4±1.8 months), each child's blood pressure was measured. The HBP in the long-term breastfeeding (LBF) group (92.9 mm Hg systolic/55.1 mm Hg diastolic) was significantly lower (P=0.006/0.04) than in the short-term breastfeeding group (94.7/56.4 mm Hg); however, there were no significant differences in the CBP measurements between the short- and LBF groups. Using multiple regression analysis, the duration of breastfeeding (greater than 8 months) was more strongly associated with HBP (P=0.008/0.05) than with CBP (P=0.4/0.9). Furthermore, the adjusted R-squared values for HBP (0.25/0.12) tended to be higher than those for CBP (0.07/0.03). These findings were independent of the birth weight. In conclusion, breastfeeding has a protective effect against elevated blood pressure even in young children, and subtle, but important, differences were precisely detected by self-measurements performed at home.

Officials welcome the arrival of the Japanese Experiment Module

NASA Image and Video Library

2007-04-17

In the Space Station Processing Facility, astronaut Takao Doi (left) and Commander Dominic Gorie pose in front of the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module, or JEM, that recently arrived at Kennedy. Doi and Gorie are crew members for mission STS-123 that will deliver the logistics module to the International Space Station. Earlier, NASA and Japanese Aerospace and Exploration Agency (JAXA) officials welcomed the arrival of the module. The new International Space Station component arrived at Kennedy March 12 to begin preparations for its future launch on mission STS-123. It will serve as an on-orbit storage area for materials, tools and supplies. It can hold up to eight experiment racks and will attach to the top of another larger pressurized module.

Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

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

Corwin, W.R.; Broadhead, B.L.; Suzuki, M.

1997-02-01

There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

NASA Astrophysics Data System (ADS)

Krasikov, E.; Nikolaenko, V.

2017-01-01

Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

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

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Provocative Encounters Reflecting Struggles with Change: Power and Coercion in a Japanese University Situation

ERIC Educational Resources Information Center

Toh, Glenn

2017-01-01

This article examines a case of what Olssen et al. (2004) call "managerial oppression" set in a faculty of international studies of a Japanese university. Japanese universities have, in recent times, been facing the financial pressures of a falling birthrate and dwindling enrolments. To remain solvent, some universities have had to…

Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

PubMed Central

Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

2013-01-01

Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0–100 MPa) and temperature (0–70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate. PMID

Handling Japanese without a Japanese Operating System.

ERIC Educational Resources Information Center

Hatasa, Kazumi; And Others

1992-01-01

The Macintosh HyperCard environment has become a popular platform for Japanese language courseware because of its flexibility and ease of programing. This project created Japanese bitmap font files for the JIS Levels 1 and 2, and writing XFCNs for font manipulation, Japanese kana input, and answer correction. (12 references) (Author/LB)

Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

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

Wang, Jy-An John

2015-08-01

The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 logmore » (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.« less

Japanese Experiment Module (JEM)

NASA Technical Reports Server (NTRS)

2003-01-01

The Japanese Experiment Module (JEM) pressure module is removed from its shipping crate and moved across the floor of the Space Station Processing Facility at Kennedy Space Center (KSC) to a work stand. A research laboratory, the pressurized module is the first element of the JEM, named 'Kibo' (Hope) to arrive at KSC. Japan's primary contribution to the International Space Station, the module will enhance unique research capabilities of the orbiting complex by providing an additional environment in which astronauts will conduct experiments. The JEM also includes an exposed facility or platform for space environment experiments, a robotic manipulator system, and two logistics modules. The various JEM components will be assembled in space over the course of three Shuttle missions.

Dietary Sodium Consumption Predicts Future Blood Pressure and Incident Hypertension in the Japanese Normotensive General Population

PubMed Central

Takase, Hiroyuki; Sugiura, Tomonori; Kimura, Genjiro; Ohte, Nobuyuki; Dohi, Yasuaki

2015-01-01

Background Although there is a close relationship between dietary sodium and hypertension, the concept that persons with relatively high dietary sodium are at increased risk of developing hypertension compared with those with relatively low dietary sodium has not been studied intensively in a cohort. Methods and Results We conducted an observational study to investigate whether dietary sodium intake predicts future blood pressure and the onset of hypertension in the general population. Individual sodium intake was estimated by calculating 24-hour urinary sodium excretion from spot urine in 4523 normotensive participants who visited our hospital for a health checkup. After a baseline examination, they were followed for a median of 1143 days, with the end point being development of hypertension. During the follow-up period, hypertension developed in 1027 participants (22.7%). The risk of developing hypertension was higher in those with higher rather than lower sodium intake (hazard ratio 1.25, 95% CI 1.04 to 1.50). In multivariate Cox proportional hazards regression analysis, baseline sodium intake and the yearly change in sodium intake during the follow-up period (as continuous variables) correlated with the incidence of hypertension. Furthermore, both the yearly increase in sodium intake and baseline sodium intake showed significant correlations with the yearly increase in systolic blood pressure in multivariate regression analysis after adjustment for possible risk factors. Conclusions Both relatively high levels of dietary sodium intake and gradual increases in dietary sodium are associated with future increases in blood pressure and the incidence of hypertension in the Japanese general population. PMID:26224048

Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

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

Le Pape, Yann; Huang, Hai

Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

Overtime work and blood pressure in normotensive Japanese male workers.

PubMed

Nakamura, Koshi; Sakurai, Masaru; Morikawa, Yuko; Miura, Katsuyuki; Ishizaki, Masao; Kido, Teruhiko; Naruse, Yuchi; Suwazono, Yasushi; Nakagawa, Hideaki

2012-09-01

Epidemiological studies have observed conflicting patterns as to whether overtime work increases blood pressure (BP), probably as a consequence of methodological issues. We conducted a prospective cohort study to investigate the relationship between overtime work hours and 1-year changes in BP in 1,235 normotensive Japanese male workers who carried out a variety of jobs in a manufacturing factory. Casual BP measurements were repeated at annual health examinations in 2004-2005, using an automatic manometer. An analysis of covariance that incorporated potential confounding factors including baseline age, body mass index (BMI), and lifestyle factors was used to calculate and compare the means of the 1-year change in systolic (SBP) and diastolic BP (DBP). The participants were grouped according to their average monthly overtime work hours obtained from timecard data between April and September 2004. The multivariate-adjusted mean for 1-year change in DBP in 611 male assembly-line workers was 1.5 mm Hg (95% confidence interval (CI) 0.8-2.2) for <40.0 h/month, 2.3 mm Hg (95% CI 1.3-3.2) for 40.0-79.9 h/month, and 5.3 mm Hg (95% CI 2.7-7.9) for ≥ 80.0 h/month (P for heterogeneity = 0.02). A broadly similar pattern was observed for SBP. In contrast, there was no significant difference in means 1-year change for both SBP and DBP in 315 clerks and 309 engineers/special technicians, grouped according to overtime work hours. Extensive overtime work was associated with increased BP in normotensive male assembly-line workers, but not in clerks and engineers/special technicians.

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

Automatic safety rod for reactors

DOEpatents

Germer, John H.

1988-01-01

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

NASA Astrophysics Data System (ADS)

Krasikov, E.

2015-04-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible.

Effects of Surface Roughness, Oxidation, and Temperature on the Emissivity of Reactor Pressure Vessel Alloys

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

King, J. L.; Jo, H.; Tirawat, R.

Thermal radiation will be an important mode of heat transfer in future high-temperature reactors and in off-normal high-temperature scenarios in present reactors. In this work, spectral directional emissivities of two reactor pressure vessel (RPV) candidate materials were measured at room temperature after exposure to high-temperature air. In the case of SA508 steel, significant increases in emissivity were observed due to oxidation. In the case of Grade 91 steel, only very small increases were observed under the tested conditions. Effects of roughness were also investigated. To study the effects of roughening, unexposed samples of SA508 and Grade 91 steel were roughenedmore » via one of either grinding or shot-peening before being measured. Significant increases were observed only in samples having roughness exceeding the roughness expected of RPV surfaces. While the emissivity increases for SA508 from oxidation were indeed significant, the measured emissivity coefficients were below that of values commonly used in heat transfer models. Based on the observed experimental data, recommendations for emissivity inputs for heat transfer simulations are provided.« less

Fossil fuel furnace reactor

DOEpatents

Parkinson, William J.

1987-01-01

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Exploring the hydraulic fracturing parameter space: a novel high-pressure, high-throughput reactor system for investigating subsurface chemical transformations.

PubMed

Sumner, Andrew J; Plata, Desiree L

2018-02-21

Hydraulic fracturing coupled with horizontal drilling (HDHF) involves the deep-well injection of a fracturing fluid composed of diverse and numerous chemical additives designed to facilitate the release and collection of natural gas from shale plays. Analyses of flowback wastewaters have revealed organic contamination from both geogenic and anthropogenic sources. The additional detections of undisclosed halogenated chemicals suggest unintended in situ transformation of reactive additives, but the formation pathways for these are unclear in subsurface brines. To develop an efficient experimental framework for investigating the complex shale-well parameter space, we have reviewed and synthesized geospatial well data detailing temperature, pressure, pH, and halide ion values as well as industrial chemical disclosure and concentration data. Our findings showed subsurface conditions can reach pressures up to 4500 psi (310 bars) and temperatures up to 95 °C, while at least 588 unique chemicals have been disclosed by industry, including reactive oxidants and acids. Given the extreme conditions necessary to simulate the subsurface, we briefly highlighted existing geochemical reactor systems rated to the necessary pressures and temperatures, identifying throughput as a key limitation. In response, we designed and developed a custom reactor system capable of achieving 5000 psi (345 bars) and 90 °C at low cost with 15 individual reactors that are readily turned over. To demonstrate the system's throughput, we simultaneously tested 12 disclosed HDHF chemicals against a radical initiator compound in simulated subsurface conditions, ruling out a dozen potential transformation pathways in a single experiment. This review outlines the dynamic and diverse parameter range experienced by HDHF chemical additives and provides an optimized framework and novel reactor system for the methodical study of subsurface transformation pathways. Ultimately, enabling such studies will provide

Effect of azilsartan versus candesartan on morning blood pressure surges in Japanese patients with essential hypertension.

PubMed

Rakugi, Hiromi; Kario, Kazuomi; Enya, Kazuaki; Sugiura, Kenkichi; Ikeda, Yoshinori

2014-06-01

Morning blood pressure (BP) surge is reported as a risk factor for cardiovascular events and end-organ damage independent of the 24-h BP level. Controlling morning BP surge is therefore important to help prevent onset of cardiovascular disease. We compared the efficacy of azilsartan and candesartan in controlling morning systolic BP (SBP) surges by analyzing relevant ambulatory BP monitoring data in patients with/without baseline BP surges. As part of a 16-week randomized, double-blind study of azilsartan (20-40 mg once daily) and candesartan (8-12 mg once daily) in Japanese patients with essential hypertension, an exploratory analysis was carried out using ambulatory BP monitoring at baseline and week 14. The effects of study drugs on morning BP surges, including sleep trough surge (early morning SBP minus the lowest night-time SBP) and prewaking surge (early morning SBP minus SBP before awakening), were evaluated. Patients with sleep trough surge of at least 35 mmHg were defined by the presence of a morning BP surge (the 'surge group'). Sleep trough surge and prewaking surge data were available at both baseline and week 14 in 548 patients, 147 of whom (azilsartan 76; candesartan 71) had a baseline morning BP surge. In surge group patients, azilsartan significantly reduced both the sleep trough surge and the prewaking surge at week 14 compared with candesartan (least squares means of the between-group differences -5.8 mmHg, P=0.0395; and -5.7 mmHg, P=0.0228, respectively). Once-daily azilsartan improved sleep trough surge and prewaking surge to a greater extent than candesartan in Japanese patients with grade I-II essential hypertension.

Reactor for tracking catalyst nanoparticles in liquid at high temperature under a high-pressure gas phase with X-ray absorption spectroscopy.

PubMed

Nguyen, Luan; Tao, Franklin Feng

2018-02-01

Structure of catalyst nanoparticles dispersed in liquid phase at high temperature under gas phase of reactant(s) at higher pressure (≥5 bars) is important for fundamental understanding of catalytic reactions performed on these catalyst nanoparticles. Most structural characterizations of a catalyst performing catalysis in liquid at high temperature under gas phase at high pressure were performed in an ex situ condition in terms of characterizations before or after catalysis since, from technical point of view, access to the catalyst nanoparticles during catalysis in liquid phase at high temperature under high pressure reactant gas is challenging. Here we designed a reactor which allows us to perform structural characterization using X-ray absorption spectroscopy including X-ray absorption near edge structure spectroscopy and extended X-ray absorption fine structure spectroscopy to study catalyst nanoparticles under harsh catalysis conditions in terms of liquid up to 350 °C under gas phase with a pressure up to 50 bars. This reactor remains nanoparticles of a catalyst homogeneously dispersed in liquid during catalysis and X-ray absorption spectroscopy characterization.

Heated-Pressure-Ball Monopropellant Rocket Engine

NASA Technical Reports Server (NTRS)

Greene, William D.

2005-01-01

A recent technology disclosure presents a concept for a monopropellant thermal spacecraft thruster that would feature both the simplicity of a typical prior pressure-fed propellant supply system and the smaller mass and relative compactness of a typical prior pump-fed system. The source of heat for this thruster would likely be a nuclear- fission reactor. The propellant would be a cryogenic fluid (a liquefied low-molecular-weight gas) stored in a tank at a low pressure. The propellant would flow from the tank, through a feedline, into three thick-walled spherical tanks, denoted pressure balls, that would be thermally connected to the reactor. Valves upstream and downstream of the pressure balls would be operated in a three-phase cycle in which propellant would flow into one pressure ball while the fluid underwent pressurization through heating in another ball and pressurized propellant was discharged from the remaining ball into the reactor. After flowing through the reactor, wherein it would be further heated, the propellant would be discharged through an exhaust nozzle to generate thrust. A fraction of the pressurized gas from the pressure balls would be diverted to maintain the desired pressure in the tank.

Dietary Sodium Consumption Predicts Future Blood Pressure and Incident Hypertension in the Japanese Normotensive General Population.

PubMed

Takase, Hiroyuki; Sugiura, Tomonori; Kimura, Genjiro; Ohte, Nobuyuki; Dohi, Yasuaki

2015-07-29

Although there is a close relationship between dietary sodium and hypertension, the concept that persons with relatively high dietary sodium are at increased risk of developing hypertension compared with those with relatively low dietary sodium has not been studied intensively in a cohort. We conducted an observational study to investigate whether dietary sodium intake predicts future blood pressure and the onset of hypertension in the general population. Individual sodium intake was estimated by calculating 24-hour urinary sodium excretion from spot urine in 4523 normotensive participants who visited our hospital for a health checkup. After a baseline examination, they were followed for a median of 1143 days, with the end point being development of hypertension. During the follow-up period, hypertension developed in 1027 participants (22.7%). The risk of developing hypertension was higher in those with higher rather than lower sodium intake (hazard ratio 1.25, 95% CI 1.04 to 1.50). In multivariate Cox proportional hazards regression analysis, baseline sodium intake and the yearly change in sodium intake during the follow-up period (as continuous variables) correlated with the incidence of hypertension. Furthermore, both the yearly increase in sodium intake and baseline sodium intake showed significant correlations with the yearly increase in systolic blood pressure in multivariate regression analysis after adjustment for possible risk factors. Both relatively high levels of dietary sodium intake and gradual increases in dietary sodium are associated with future increases in blood pressure and the incidence of hypertension in the Japanese general population. © 2015 The Authors. Published on behalf of the American Heart Association, Inc., by Wiley Blackwell.

Synthesis of Carbon Nanotubes in Thermal Plasma Reactor at Atmospheric Pressure.

PubMed

Szymanski, Lukasz; Kolacinski, Zbigniew; Wiak, Slawomir; Raniszewski, Grzegorz; Pietrzak, Lukasz

2017-02-18

In this paper, a novel approach to the synthesis of the carbon nanotubes (CNTs) in reactors operating at atmospheric pressure is presented. Based on the literature and our own research results, the most effective methods of CNT synthesis are investigated. Then, careful selection of reagents for the synthesis process is shown. Thanks to the performed calculations, an optimum composition of gases and the temperature for successful CNT synthesis in the CVD (chemical vapor deposition) process can be chosen. The results, having practical significance, may lead to an improvement of nanomaterials synthesis technology. The study can be used to produce CNTs for electrical and electronic equipment (i.e., supercapacitors or cooling radiators). There is also a possibility of using them in medicine for cancer diagnostics and therapy.

Automatic safety rod for reactors. [LMFBR

DOEpatents

Germer, J.H.

1982-03-23

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Officials welcome the arrival of the Japanese Experiment Module

NASA Image and Video Library

2007-04-17

In the Space Station Processing Facility, NASA and Japanese Aerospace and Exploration Agency (JAXA) officials welcome the arrival of the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module, or JEM, to the Kennedy Space Center. At the podium is Russ Romanella, director of International Space Station and Spacecraft Processing. Seated at right are Bill Parsons, director of Kennedy Space Center; Dr. Kichiro Imagawa, project manager of the JEM Development Project Team for JAXA; Melanie Saunders, associate manager of the International Space Station Program at Johnson Space Center; and Dominic Gorie, commander on mission STS-123 that will deliver the module to the space station. The new International Space Station component arrived at Kennedy March 12 to begin preparations for its future launch on mission STS-123. It will serve as an on-orbit storage area for materials, tools and supplies. It can hold up to eight experiment racks and will attach to the top of another larger pressurized module.

Officials welcome the arrival of the Japanese Experiment Module

NASA Image and Video Library

2007-04-17

In the Space Station Processing Facility, NASA and Japanese Aerospace and Exploration Agency (JAXA) officials welcome the arrival of the Experiment Logistics Module Pressurized Section for the Japanese Experiment Module, or JEM, to the Kennedy Space Center. At the podium is Bill Parsons, director of Kennedy Space Center. Seated at right are Russ Romanella, director of International Space Station and Spacecraft Processing; Dr. Kichiro Imagawa, project manager of the JEM Development Project Team for JAXA; Melanie Saunders, associate manager of the International Space Station Program at Johnson Space Center; and Dominic Gorie, commander on mission STS-123 that will deliver the module to the space station. The new International Space Station component arrived at Kennedy March 12 to begin preparations for its future launch on mission STS-123. It will serve as an on-orbit storage area for materials, tools and supplies. It can hold up to eight experiment racks and will attach to the top of another larger pressurized module.

Cross-cultural examination of beliefs about the causes of bulimia nervosa among Australian and Japanese females.

PubMed

Dryer, Rachel; Uesaka, Yuri; Manalo, Emmanuel; Tyson, Graham

2015-03-01

To identify similarities and differences in beliefs about the causes of Bulimia Nervosa (BN) held by Asian (Japanese) women and Western (Australian) women, and hence, to examine the applicability of belief models of eating disorders (ED) across different cultures. Four hundred three Japanese and 256 Australian female university students (aged 17-35 years) completed a questionnaire that gauged beliefs about the causes of BN. Among the Australian women, the four-component structure of perceived causes (dieting and eating practices, family dynamics, socio-cultural pressure, and psychological vulnerability) found in Dryer et al. (2012) was replicated. Among the Japanese women, however, a three-component structure (without the psychological vulnerability component) was obtained. The groups also differed in the causal component they most strongly endorsed, that being socio-cultural pressure for the Australian women, and dieting and eating practices for the Japanese women. The Japanese participants were found to endorse three out of the four Western-based causal explanations for BN, but the relative importance they placed on those explanations differed from that of the Australian participants. Further research is needed, particularly to establish whether Japanese women simply fail to see psychological vulnerability as a viable cause of BN, or there are in fact cultural differences in the extent to which such vulnerability causes BN. © 2014 Wiley Periodicals, Inc.

Chemistry experience in the primary heat transport circuits of Kraftwerk Union pressurized water reactors

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

Riess, R.

Chosen for this description of the selected Kraftwerk Union (KWU) pressurized water reactor units were Obrigheim (KWO, 345 MW(e)), Stade (KKS, 662 (MW(e)), Borselle (KCB, 477 MW(e)), and Biblis (KWB-A, 1204 MW(e)). The experience at these plants shows that with a special startup procedure and a proper chemical control of the primary heat transport system that influences general corrosion, selective types of corrosion, corrosion product activity transport and resulting contamination, and radiation-induced decomposition, KWU units have no basic problems.

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

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

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

1997-02-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices ofmore » Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.« less

Effect of azilsartan versus candesartan on morning blood pressure surges in Japanese patients with essential hypertension

PubMed Central

Kario, Kazuomi; Enya, Kazuaki; Sugiura, Kenkichi; Ikeda, Yoshinori

2014-01-01

Morning blood pressure (BP) surge is reported as a risk factor for cardiovascular events and end-organ damage independent of the 24-h BP level. Controlling morning BP surge is therefore important to help prevent onset of cardiovascular disease. We compared the efficacy of azilsartan and candesartan in controlling morning systolic BP (SBP) surges by analyzing relevant ambulatory BP monitoring data in patients with/without baseline BP surges. As part of a 16-week randomized, double-blind study of azilsartan (20–40 mg once daily) and candesartan (8–12 mg once daily) in Japanese patients with essential hypertension, an exploratory analysis was carried out using ambulatory BP monitoring at baseline and week 14. The effects of study drugs on morning BP surges, including sleep trough surge (early morning SBP minus the lowest night-time SBP) and prewaking surge (early morning SBP minus SBP before awakening), were evaluated. Patients with sleep trough surge of at least 35 mmHg were defined by the presence of a morning BP surge (the ‘surge group’). Sleep trough surge and prewaking surge data were available at both baseline and week 14 in 548 patients, 147 of whom (azilsartan 76; candesartan 71) had a baseline morning BP surge. In surge group patients, azilsartan significantly reduced both the sleep trough surge and the prewaking surge at week 14 compared with candesartan (least squares means of the between-group differences −5.8 mmHg, P=0.0395; and −5.7 mmHg, P=0.0228, respectively). Once-daily azilsartan improved sleep trough surge and prewaking surge to a greater extent than candesartan in Japanese patients with grade I–II essential hypertension. PMID:24710336

Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

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

Boucau, Joseph; Mirabella, C.; Nilsson, Lennart

2013-07-01

Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Centermore » for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA

ARD remediation with limestone in a CO2 pressurized reactor

USGS Publications Warehouse

Sibrell, Philip L.; Watten, Barnaby J.; Friedrich, Andrew E.; Vinci, Brian J.

2000-01-01

We evaluated a new process for remediation of acid rock drainage (ARD). The process treats ARD with intermittently fluidized beds of granular limestone maintained within a continuous flow reactor pressurized with CO2. Tests were performed over a thirty day period at the Toby Creek mine drainage treatment plant, Elk County, Pennsylvania in cooperation with the Pennsylvania Department of Environmental Protection. Equipment performance was established at operating pressures of 0, 34, 82, and 117 kPa using an ARD flow of 227 L/min. The ARD had the following characteristics: pH, 3.1; temperature, 10 °C; dissolved oxygen, 6.4 mg/L; acidity, 260 mg/L; total iron, 21 mg/L; aluminum, 22 mg/L; manganese, 7.5 mg/L; and conductivity, 1400 μS/cm. In all cases tested, processed ARD was net alkaline with mean pH and alkalinities of 6.7 and 59 mg/L at a CO2 pressure of 0 kPa, 6.6 and 158 mg/L at 34 kPa, 7.4 and 240 mg/L at 82 kPa, and 7.4 and 290 mg/L at 117 kPa. Processed ARD alkalinities were correlated to the settled bed depth (p<0.001) and CO2 pressure (p<0.001). Iron, aluminum, and manganese removal efficiencies of 96%, 99%, and 5%, respectively, were achieved with filtration following treatment. No indications of metal hydroxide precipitation or armoring of the limestone were observed. The surplus alkalinity established at 82 kPa was successful in treating an equivalent of 1136 L/min (five-fold dilution) of the combined three ARD streams entering the Toby Creek Plant. This side-stream capability provides savings in treatment unit scale as well as flexibility in treatment effect. The capability of the system to handle higher influent acidity was tested by elevating the acidity to 5000 mg/L with sulfuric acid. Net alkaline effluent was produced, indicating applicability of the process to highly acidic ARD.

ENGINEERING TEST REACTOR

DOEpatents

De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

1958-10-21

Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

Do American born Japanese children still grow faster than native Japanese?

PubMed

Kano, K; Chung, C S

1975-09-01

Growth patterns of Japanese schoolchildren in Hawaii, composed of 2,954 boys and 3,213 girls aged between 11 and 17, were compared with those comparable groups of Japanese schoolchildren in Japan based on the data published by the Japanese Ministry of Education. Growth characteristics studied were height, weight, and relative weight index, weight/(height). The Hawaii-Japanese boys were taller at early ages but the difference disappeared by age 16. Native Japanese girls were shorter than Hawaii-Japanese until age 13, but they overtook the latter by age 14, exceeding them in height after age 15. A similar pattern was found in weights of girls but the Hawaii-Japanese boys remained consistently heavier by 5.0 to 9.0 kg than native Japanese. The relative weight measure indicated that the Hawaii boys were more "obese" than native Japanese boys for the growth period studied; whereas the same tendency was maintained until age 15 in girls. These observations indicate a marked degree of convergence of the patterns of physical growth of the two populations, whose differences were unmistakably in favor of American born children in earlier studies. It is concluded that the convergence is due largely to the improved environmental conditions in Japan in recent years.

Detecting pin diversion from pressurized water reactors spent fuel assemblies

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

Ham, Young S.; Sitaraman, Shivakumar

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

Japaneseplex: A forensic SNP assay for identification of Japanese people using Japanese-specific alleles.

PubMed

Yuasa, Isao; Akane, Atsushi; Yamamoto, Toshimichi; Matsusue, Aya; Endoh, Minoru; Nakagawa, Mayumi; Umetsu, Kazuo; Ishikawa, Takaki; Iino, Morio

2018-04-24

It is sometimes necessary to determine whether a forensic biological sample came from a Japanese person. In this study, we developed a 60-locus SNP assay designed for the differentiation of Japanese people from other East Asians using entirely and nearly Japanese-specific alleles. This multiplex assay consisted of 6 independent PCR reactions followed by single nucleotide extension. The average number and standard deviation of Japanese-specific alleles possessed by an individual were 0.81 ± 0.93 in 108 Koreans from Seoul, 8.87 ± 2.89 in 103 Japanese from Tottori, 17.20 ± 3.80 in 88 Japanese from Okinawa, and 0 in 220 Han Chinese from Wuxi and Changsha. The Koreans had 0-4 Japanese-specific alleles per individual, whereas the Japanese had 4-26 Japanese-specific alleles. Almost all Japanese were distinguished from the Koreans and other people by the factorial correspondence and principal component analyses. The Snipper program was also useful to estimate the degree of Japaneseness. The method described here was successfully applied to the differentiation of Japanese from non-Japanese people in forensic cases. This Japanese-specific SNP assay was named Japaneseplex. Copyright © 2018 Elsevier B.V. All rights reserved.

Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies.

PubMed

Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

2013-03-01

Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ∼10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified.

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

DOE PAGES

Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; ...

2015-10-21

Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

Development of Japanese experiment module remote manipulator system

NASA Technical Reports Server (NTRS)

Matsueda, Tatsuo; Kuwao, Fumihiro; Motohasi, Shoichi; Okamura, Ryo

1994-01-01

National Space Development Agency of Japan (NASDA) is developing the Japanese Experiment Module (JEM), as its contribution to the International Space Station. The JEM consists of the pressurized module (PM), the exposed facility (EF), the experiment logistics module pressurized section (ELM-PS), the experiment logistics module exposed section (ELM-ES) and the Remote Manipulator System (RMS). The JEMRMS services for the JEM EF, which is a space experiment platform, consists of the Main Arm (MA), the Small Fine Arm (SFA) and the RMS console. The MA handles the JEM EF payloads, the SFA and the JEM element, such as ELM-ES.

[Understanding the symbolic values of Japanese onomatopoeia: comparison of Japanese and Chinese speakers].

PubMed

Haryu, Etsuko; Zhao, Lihua

2007-10-01

Do non-native speakers of the Japanese language understand the symbolic values of Japanese onomatopoeia matching a voiced/unvoiced consonant with a big/small sound made by a big/small object? In three experiments, participants who were native speakers of Japanese, Japanese-learning Chinese, or Chinese without knowledge of the Japanese language were shown two pictures. One picture was of a small object making a small sound, such as a small vase being broken, and the other was of a big object making a big sound, such as a big vase being broken. Participants were presented with two novel onomatopoetic words with voicing contrasts, e.g.,/dachan/vs./tachan/, and were told that each word corresponded to one of the two pictures. They were then asked to match the words to the corresponding pictures. Chinese without knowledge of Japanese performed only at chance level, whereas Japanese and Japanese-learning Chinese successfully matched a voiced/unvoiced consonant with a big/small object respectively. The results suggest that the key to understanding the symbolic values of voicing contrasts in Japanese onomatopoeia is some basic knowledge that is intrinsic to the Japanese language.

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

1988-01-01

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

The Japanese Mind: Understanding Contemporary Japanese Culture.

ERIC Educational Resources Information Center

Davies, Roger J., Ed.; Ikeno, Osamu, Ed.

This collection of essays offers an overview of contemporary Japanese culture, and can serve as a resource for classes studying Japan. The 28 essays offer an informative, accessible look at the values, attitudes, behavior patterns, and communication styles of modern Japan from the unique perspective of the Japanese people. Filled with examples…

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

REACTOR

DOEpatents

Roman, W.G.

1961-06-27

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

What Is Business Japanese? Designing a Japanese Course for Business Communication.

ERIC Educational Resources Information Center

Koike, Shohei

Experiences in developing "Business Japanese" courses for the undergraduate major in Language and International Trade at Eastern Michigan University are described. In 1987, six new courses in Japanese were proposed so that Japanese could be offered as a language specialty in the program. Issues considered in defining business Japanese…

A Confirmatory Model for Substance Use Among Japanese American and Part-Japanese American Adolescents

PubMed Central

Williams, John Kino Yamaguchi; Else, 'Iwalani R. N.; Goebert, Deborah A.; Nishimura, Stephanie T.; Hishinuma, Earl S.; Andrade, Naleen N.

2013-01-01

Few studies have examined the effect of ethnicity and cultural identity on substance use among Asian and Pacific Islander adolescents. A cross-sequential study conducted in Hawai'i with 144 Japanese American and part-Japanese American adolescents assessed a model integrating Japanese ethnicity, cultural identity, substance use, major life events, and social support. Japanese American adolescents scored higher on the Japanese Culture Scale and on the Peers’ Social Support than the part-Japanese American adolescents. Significant associations for substance use and impairment included culturally intensified events and Japanese cultural identity- behavior subset. Models had good overall fits and suggested that conflict surrounding cultural identity may contribute to substance use. PMID:23480213

Nuclear reactor cavity floor passive heat removal system

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

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Pressurized pyrolysis of rice husk in an inert gas sweeping fixed-bed reactor with a focus on bio-oil deoxygenation.

PubMed

Qian, Yangyang; Zhang, Jie; Wang, Jie

2014-12-01

The pyrolysis of rice husk was conducted in a fixed-bed reactor with a sweeping nitrogen gas to investigate the effects of pressure on the pyrolytic behaviors. The release rates of main gases during the pyrolysis, the distributions of four products (char, bio-oil, water and gas), the elemental compositions of char, bio-oil and gas, and the typical compounds in bio-oil were determined. It was found that the elevation of pressure from 0.1MPa to 5.0MPa facilitated the dehydration and decarboxylation of bio-oil, and the bio-oils obtained under the elevated pressures had significantly less oxygen and higher calorific value than those obtained under atmospheric pressure. The former bio-oils embraced more acetic acid, phenols and guaiacols. The elevation of pressure increased the formation of CH4 partially via the gas-phase reactions. An attempt is made in this study to clarify "the pure pressure effect" and "the combined effect with residence time". Copyright © 2014 Elsevier Ltd. All rights reserved.

Association between Selective Beta-adrenergic Drugs and Blood Pressure Elevation: Data Mining of the Japanese Adverse Drug Event Report (JADER) Database.

PubMed

Ohyama, Katsuhiro; Inoue, Michiko

2016-01-01

Selective beta-adrenergic drugs are used clinically to treat various diseases. Because of imperfect receptor selectivity, beta-adrenergic drugs cause some adverse drug events by stimulating other adrenergic receptors. To examine the association between selective beta-adrenergic drugs and blood pressure elevation, we reviewed the Japanese Adverse Drug Event Reports (JADERs) submitted to the Japan Pharmaceuticals and Medical Devices Agency. We used the Medical Dictionary for Regulatory Activities (MedDRA) Preferred Terms extracted from Standardized MedDRA queries for hypertension to identify events related to blood pressure elevation. Spontaneous adverse event reports from April 2004 through May 2015 in JADERs, a data mining algorithm, and the reporting odds ratio (ROR) were used for quantitative signal detection, and assessed by the case/non-case method. Safety signals are considered significant if the ROR estimates and lower bound of the 95% confidence interval (CI) exceed 1. A total of 2021 reports were included in this study. Among the nine drugs examined, significant signals were found, based on the 95%CI for salbutamol (ROR: 9.94, 95%CI: 3.09-31.93) and mirabegron (ROR: 7.52, 95%CI: 4.89-11.55). The results of this study indicate that some selective beta-adrenergic drugs are associated with blood pressure elevation. Considering the frequency of their indications, attention should be paid to their use in elderly patients to avoid adverse events.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

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

Spencer, Benjamin; Backman, Marie; Hoffman, William

Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components,more » and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.« less

The Japanese aerial attack on Hanford Engineer Works

NASA Astrophysics Data System (ADS)

Clark, Charles W.

The day before the Pearl Harbor attack, December 6, 1941, the University of Chicago Metallurgical Laboratory was given four goals: design a plutonium (Pu) bomb; produce Pu by irradiation of uranium (U); extract Pu from the irradiated U; complete this in time to be militarily significant. A year later the first controlled nuclear chain reaction was attained in Chicago Pile 1 (CP-1). In January 1943, Hanford, WA was chosen as the site of the Pu factory. Neutron irradiation of 238U was to be used to make 239Pu. This was done by a larger version of CP-1, Hanford Reactor B, which went critical in September 1944. By July 1945 it had made enough Pu for two bombs: one used at the Trinity test in July; the other at Nagasaki, Japan in August. I focus on an ironic sidelight to this story: disruption of hydroelectric power to Reactor B by a Japanese fire balloon attack on March 10, 1945. This activated the costly coal-fired emergency backup plant to keep the reactor coolant water flowing, thwarting disaster and vindicating the conservative design of Hanford Engineer Works. Management of the Hanford Engineer Works in World War II, H. Thayer (ASCE Press 1996).

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

Experimental and Numerical Investigation of Pressure Drop in Silicon Carbide Fuel Rod for Application in Pressurized Water Reactors

NASA Astrophysics Data System (ADS)

Abir, Ahmed Musafi

Spacer grids are used in Pressurized Water Reactors (PWRs) fuel assemblies which enhances heat transfer from fuel rods. However, there remain regions of low turbulence in between the spacer grids. To enhance turbulence in these regions surface roughness is applied on the fuel rod walls. Meyer [1] used empirical correlations to predict heat transfer and friction factor for artificially roughened fuel rod bundles at High Performance Light Water Reactors (LWRs). Their applicability was tested by Carrilho at University of South Carolina's (USC) Single Heated Element Loop Tester (SHELT). He attained a heat transfer and friction factor enhancement of 50% and 45% respectively, using Inconel nuclear fuel rods with square transverse ribbed surface. Following him Najeeb conducted a similar study due to three dimensional diamond shaped blocks in turbulent flow. He recorded a maximum heat transfer enhancement of 83%. At present, several types of materials are being used for fuel rod cladding including Zircaloy, Uranium oxide, etc. But researchers are actively searching for new material that can be a more practical alternative. Silicon Carbide (SiC) has been identified as a material of interest for application as fuel rod cladding [2]. The current study deals with the experimental investigation to find out the friction factor increase of a SiC fuel rod with 3D surface roughness. The SiC rod was tested at USC's SHELT loop. The experiment was conducted in turbulent flowing Deionized (DI) water at steady state conditions. Measurements of Flow rate and pressure drop were made. The experimental results were also validated by Computational Fluid Dynamics (CFD) analysis in ANSYS Fluent. To simplify the CFD analysis and to save computational resources the 3D roughness was approximated as a 2D one. The friction factor results of the CFD investigation was found to lie within +/-8% of the experimental results. A CFD model was also run with the energy equation turned on, and a heat

Hydrogasification reactor and method of operating same

DOEpatents

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

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

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

Entropy Production in Chemical Reactors

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián C.

2017-06-01

We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

PubMed

Lindgren, Kristina; Stiller, Krystyna; Efsing, Pål; Thuvander, Mattias

2017-04-01

Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

Homogamy and Intermarriage of Japanese and Japanese Americans with Whites Surrounding World War II

ERIC Educational Resources Information Center

Ono, Hiromi; Berg, Justin

2010-01-01

Although some sociologists have suggested that Japanese Americans quickly assimilated into mainstream America, scholars of Japanese America have highlighted the heightened exclusion that the group experienced. This study tracked historical shifts in the exclusion level of Japanese and Japanese Americans in the United States surrounding World War…

Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

PubMed

Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

2004-08-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

Reactor for exothermic reactions

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

1993-01-01

A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Reactor for exothermic reactions

DOEpatents

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Measurement of /sup 14/C emission rates from a pressurized heavy water reactor

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

Joshi, M.L.; Ramamirtham, B.; Soman, S.D.

Carbon-14 is produced in pressurized heavy water reactors (PHWR), mainly as an activation product in the fuel. It is also produced in the heavy water used as the primary coolant and moderator, and is produced in the air in the annular space between the pressure tube and calandria tubes as well as in the free space in the calandria vault. The production rates in different systems of a PHWR are calculated on the basis of design parameters. During a period of 3 y, /sup 14/C released through the gaseous route has been measured at Rajasthan Atomic Power Station, Kota, India,more » a PHWR unit. These releases have been found to be mainly /sup 14/CO/sub 2/. This reduced form of /sup 14/C is less than 5% of the releases. The normalized releases of /sup 14/C have a geometric mean of 5.17 TBq GWe-1 y-1 and a geometric standard deviation of 1.52. The /sup 14/C present in the form of carbonates in liquid effluents has also been measured and is 0.14% of the gaseous releases.« less

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

Backward Yakudoku: An Attempt to Implement CLT at a Japanese High School

ERIC Educational Resources Information Center

Thompson, Gene; Yanagita, Mayuno

2017-01-01

How can Japanese teachers of English go about introducing more communicative activities suitable for their contexts? This article discusses an attempt by a high school teacher to implement communicative language teaching (CLT) in her classes while responding to institutional pressure to use "yakudoku" (a traditional grammar translation…

The effects of plasma inhomogeneity on the nanoparticle coating in a low pressure plasma reactor

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

Pourali, N.; Foroutan, G.

2015-10-15

A self-consistent model is used to study the surface coating of a collection of charged nanoparticles trapped in the sheath region of a low pressure plasma reactor. The model consists of multi-fluid plasma sheath module, including nanoparticle dynamics, as well as the surface deposition and particle heating modules. The simulation results show that the mean particle radius increases with time and the nanoparticle size distribution is broadened. The mean radius is a linear function of time, while the variance exhibits a quadratic dependence. The broadening in size distribution is attributed to the spatial inhomogeneity of the deposition rate which inmore » turn depends on the plasma inhomogeneity. The spatial inhomogeneity of the ions has strong impact on the broadening of the size distribution, as the ions contribute both in the nanoparticle charging and in direct film deposition. The distribution width also increases with increasing of the pressure, gas temperature, and the ambient temperature gradient.« less

Pressurizer tank upper support

DOEpatents

Baker, T.H.; Ott, H.L.

1994-01-11

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90[degree] intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure. 10 figures.

Pressurizer tank upper support

DOEpatents

Baker, Tod H.; Ott, Howard L.

1994-01-01

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure.

Association between control to target blood pressures and healthy lifestyle factors among Japanese hypertensive patients: longitudinal data analysis from Fukushima Research of Hypertension (FRESH).

PubMed

Yokokawa, Hirohide; Goto, Aya; Sanada, Hironobu; Watanabe, Tsuyoshi; Felder, Robin A; Jose, Pedro A; Yasumura, Seiji

2014-01-01

To determine success rates in controlling target blood pressures longitudinally by measuring several factors, including lifestyle characteristics associated with uncontrolled blood pressures for target treatment goals. This prospective observational cohort study (September 2008-September 2010) collected information on blood pressure control status and healthy lifestyle factors listed in Breslow's seven health practices through medical records and self-administered questionnaires from 884 of the 1264 Japanese hypertensive patients initially registered in the FRESH study. Multivariate analysis adjusted for associated factors was performed to estimate the association between lifestyle change and "uncontrolled blood pressures" at the final follow-up survey. Median age and proportion of men were 73 years and 39.1%, respectively. All survey failure rates were 37.6% among non-elderly patients (<65 years of age) without diabetes mellitus or chronic kidney disease, and 35.0% among patients with these diseases or myocardial infarction. Maintaining a healthy lifestyle was a protective factor against uncontrolled blood pressures in multivariate analysis. Obesity and smoking status were associated with uncontrolled blood pressures, and exercise frequency was borderline significance. The number of participants with healthy responses for these factors remained relatively low during follow up. Our study revealed low rates of controlled blood pressures, especially in non-elderly patients without diabetes mellitus or chronic kidney disease, and patients with these diseases or myocardial infarction. Our data indicate the need to maintain a healthy lifestyle, in particular, ideal body weight and adequate exercise frequency, for better hypertension management according to treatment guidelines. Copyright © 2013 Asian Oceanian Association for the Study of Obesity. Published by Elsevier Ltd. All rights reserved.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

Visualization of Flow in Pressurizer Spray Line Piping and Estimation of Thermal Stress Fluctuation Caused by Swaying of Water Surface

NASA Astrophysics Data System (ADS)

Oumaya, Toru; Nakamura, Akira; Onojima, Daisuke; Takenaka, Nobuyuki

The pressurizer spray line of PWR plants cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation of the plant is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in spray line piping and assess their impact on pipe structure, a flow visualization experiment was conducted. In the experiment, air was used substituted for steam to simulate the gas phase of the pressurizer, and the flow instability causing swaying without condensation was investigated. With a full-scale mock-up made of acrylic, flow under room temperature and atmospheric pressure conditions was visualized, and possible flow regimes were identified based on the results of the experiment. Three representative patterns of swaying of water surface were assumed, and the range of thermal stress fluctuation, when the surface swayed instantaneously, was calculated. With the three patterns of swaying assumed based on the visualization experiment, it was confirmed that the thermal stress amplitude would not exceed the fatigue endurance limit prescribed in the Japanese Design and Construction Code.

The Anglo-Japanese Alliance and Japanese Expansionism 1902-1923.

DTIC Science & Technology

1992-06-05

Alienation 1919-1952. London: Cambridge University Press. 1982. • The Oriains of the Russo-Japanese War. London: Longman Group Limited. 1985. Nitobe ... Inazo . Bushido - The Soul of Japan. Tokyo: Tuttle. 1981. Okamoto, Shumpei. The Japan Oliaarchv and the Russo-Japanese War. New York: Columbia

Pressure-Letdown Machine for a Coal Reactor

NASA Technical Reports Server (NTRS)

Perkins, G. S.; Mabe, W. B.

1986-01-01

Pumps operating in reverse generate power. Conceptual pressure-letdown machine for coal-liquefaction system extracts energy from expansion of product fluid. Mud pumps, originally intended for use in oil drilling, operated in reverse so their motors act as generators. Several pumps operated in alternating phase to obtain multiple stages of letdown from inlet pressure to outlet pressure. About 75 percent of work generates inlet pressure recoverable as electrical energy.

REFLECTOR CONTROL OF A BOILING-WATER REACTOR

DOEpatents

Treshow, M.

1962-05-22

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Rodded shutdown system for a nuclear reactor

DOEpatents

Golden, Martin P.; Govi, Aldo R.

1978-01-01

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

Cardiovascular risk factors in American and Japanese executives. Telecom Health Research Group.

PubMed

Comstock, G W; Suzuki, T; Stone, R W; Crumrine, J L; Johnson, D H; Sakai, Y; Matsuya, T; Sasaki, S

1985-07-01

A standardized cardiovascular risk factor examination was given to executives in the headquarters of the American Telephone and Telegraph Company and the Nippon Telegraph and Telephone Public Corporation. As expected from the national mortality data, evidence of ischaemic heart disease was more common among American than Japanese executives. The frequency of some but not all risk factors was consistent with the observed differences in ischaemic heart disease. Americans were fatter than their Japanese counterparts, obtained a higher proportion of their caloric intake from animal fats, had higher serum cholesterol levels, and more of them felt that their lives were highly stressful. On the other hand, Japanese executives were much more likely to be cigarette smokers and showed a greater increase in blood pressure with age. Serum high-density lipoprotein cholesterol levels and the ratio of saturated to unsaturated fatty acids in the serum were similar in the two groups.

A Conceptual Model of Cultural Predictors of Anxiety among Japanese American and Part-Japanese American Adolescents.

ERIC Educational Resources Information Center

Williams, John Kino Yamaguchi; Goebert, Deborah; Hishinuma, Earl; Miyamoto, Robin; Anzai, Neal; Izutsu, Satoru; Yanagida, Evelyn; Nishimura, Stephanie; Andrade, Naleen; Baker, F. M.

2002-01-01

Develops and assesses a model integrating Japanese ethnicity, cultural identity, and anxiety in Japanese American and part-Japanese American high school seniors. Japanese American adolescents scored higher on the scale and reported fewer anxiety symptoms than part-Japanese American adolescents. The model had a good overall fit, suggesting that…

Nuclear reactor melt-retention structure to mitigate direct containment heating

DOEpatents

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Ceramic membrane reactor with two reactant gases at different pressures

DOEpatents

Balachandran, Uthamalingam; Mieville, Rodney L.

2001-01-01

The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

Degradation of aqueous phenol solutions by coaxial DBD reactor

NASA Astrophysics Data System (ADS)

Dojcinovic, B. P.; Manojlovic, D.; Roglic, G. M.; Obradovic, B. M.; Kuraica, M. M.; Puric, J.

2008-07-01

Solutions of 2-chlorophenol, 4-chlorophenol and 2,6-dichlorophenol in bidistilled and water from the river Danube were treated in plasma reactor. In this reactor, based on coaxial dielectric barrier discharge at atmospheric pressure, plasma is formed over a thin layer of treated water. After one pass through the reactor, starting chlorophenols concentration of 20 mg/l was diminished up to 95 %. Kinetics of the chlorophenols degradation was monitored by High Pressure Liquid Chromatography method (HPLC).

Statistical Machine Translation of Japanese

DTIC Science & Technology

2007-03-01

hiragana and katakana) syllabaries…………………….. 20 3.2 Sample Japanese sentence showing kanji and kana……………………... 21 3.5 Japanese formality example...syllabary. 19 Figure 3.1. Japanese kana syllabaries, hiragana for native Japanese words, word endings, and particles, and katakana for foreign...Figure 3.2. Simple Japanese sentence showing the use of kanji, hiragana , and katakana. Kanji is used for nouns and verb, adjective, and

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Scanning tunneling microscope assembly, reactor, and system

DOEpatents

Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

2014-11-18

An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

Acculturation of Personality: A Three-Culture Study of Japanese, Japanese Americans, and European Americans.

PubMed

Güngör, Derya; Bornstein, Marc H; De Leersnyder, Jozefien; Cote, Linda; Ceulemans, Eva; Mesquita, Batja

2013-07-01

The present study tests the hypothesis that involvement with a new culture instigates changes in personality of immigrants that result in (a) better fit with the norms of the culture of destination and (b) reduced fit with the norms of the culture of origin. Participants were 40 Japanese first-generation immigrants to the United States, 57 Japanese monoculturals, and 60 U.S. monoculturals. All participants completed the Jackson Personality Inventory (JPI) as a measure of the Big Five; immigrants completed the Japanese American Acculturation Scale. Immigrants' fits with the cultures of destination and origin were calculated by correlating Japanese American mothers' patterns of ratings on the Big Five with the average patterns of ratings of European Americans and Japanese on the same personality dimensions. Japanese Americans became more "American" and less "Japanese" in their personality as they reported higher participation in the U.S. culture. The results support the view that personality can be subject to cultural influence.

Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

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

Nad, Shreya; Department of Physics and Astronomy, Michigan State University, East Lansing, Michigan 48824; Gu, Yajun

2015-07-15

The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficienciesmore » (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.« less

Nuclear reactor

DOEpatents

Pennell, William E.; Rowan, William J.

1977-01-01

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Food sources of dietary sodium in the Japanese adult population: the international study of macro-/micronutrients and blood pressure (INTERMAP).

PubMed

Okuda, Nagako; Okayama, Akira; Miura, Katsuyuki; Yoshita, Katsushi; Saito, Shigeyuki; Nakagawa, Hideaki; Sakata, Kiyomi; Miyagawa, Naoko; Chan, Queenie; Elliott, Paul; Ueshima, Hirotsugu; Stamler, Jeremiah

2017-04-01

It is often reported that Na intake levels are higher in Japan than in western countries. Detailed analysis of food intake and its association with Na intake are necessary for supporting further decreases in Na consumption in Japan. We investigated the association between Na and food intake by food group using data from the Japanese participants of the INTERMAP Study. Results from the Japanese participants of the INTERMAP Study who did not use antihypertensive medication and/or consume a reduced Na diet were used (531 men and 518 women, aged 40-59 years), obtained from four 24-h dietary recalls and two 24-h urine collections from each participant. We developed a classification system with 46 food group classifications; food consumption and Na intake from these groups were compared across quartiles of participants determined by 24-h urinary Na excretion per unit of body weight (UNa/BW). Average daily Na intake from Japanese high-Na foods was 2552 mg/day. Participants with a higher UNa/BW consumed a significantly greater amount of high-Na Japanese foods, such as salted fish (P = 0.001) and miso soup (P < 0.001). They also had greater amount of rice (P = 0.001). Participants with lower UNa/BW consumed a significantly greater amount of western foods, such as bread (P < 0.001) and milk and dairy products (P < 0.001). Detailed analyses of various Japanese and western food intakes in addition to Na intake were performed. These results can be used to help draw up effective programs for the reduction in Na intake and prevention of prehypertension/hypertension in the Japanese population.

Americans and Japanese Nonverbal Communication. Linguistic Communications 15 (Papers in Japanese Linguistics 3).

ERIC Educational Resources Information Center

Taylor, Harvey M.

Each culture has its own nonverbal as well as its verbal language. Movements, gestures and sounds have distinct and often conflicting interpretations in different countries. For Americans communicating with Japanese, misunderstandings are of two types: Japanese behavior which is completely new to the American, and Japanese behavior which is…

Detraditionalisation: Japanese Students in the USA.

ERIC Educational Resources Information Center

Ueno, Junko

2001-01-01

Focuses on the identity formation of Japanese students temporarily living in the United States. The students were enrolled in Japanese Saturday school and in American public schools. Student interviews reveal a mixture of Japanese and American characteristics. Suggests Japanese students do not reject either culture--Japanese or American--but that…

Japanese-English language equivalence of the Cognitive Abilities Screening Instrument among Japanese-Americans.

PubMed

Gibbons, Laura E; McCurry, Susan; Rhoads, Kristoffer; Masaki, Kamal; White, Lon; Borenstein, Amy R; Larson, Eric B; Crane, Paul K

2009-02-01

The Cognitive Abilities Screening Instrument (CASI) was designed for use in cross-cultural studies of Japanese and Japanese-American elderly in Japan and the U.S.A. The measurement equivalence in Japanese and English had not been confirmed in prior studies. We analyzed the 40 CASI items for differential item functioning (DIF) related to test language, as well as self-reported proficiency with written Japanese, age, and educational attainment in two large epidemiologic studies of Japanese-American elderly: the Kame Project (n=1708) and the Honolulu-Asia Aging Study (HAAS; n = 3148). DIF was present if the demographic groups differed in the probability of success on an item, after controlling for their underlying cognitive functioning ability. While seven CASI items had DIF related to language of testing in Kame (registration of one item; recall of one item; similes; judgment; repeating a phrase; reading and performing a command; and following a three-step instruction), the impact of DIF on participants' scores was minimal. Mean scores for Japanese and English speakers in Kame changed by <0.1 SD after accounting for DIF related to test language. In HAAS, insufficient numbers of participants were tested in Japanese to assess DIF related to test language. In both studies, DIF related to written Japanese proficiency, age, and educational attainment had minimal impact. To the extent that DIF could be assessed, the CASI appeared to meet the goal of measuring cognitive function equivalently in Japanese and English. Stratified data collection would be needed to confirm this conclusion. DIF assessment should be used in other studies with multiple language groups to confirm that measures function equivalently or, if not, form scores that account for DIF.

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

FOOD IRRADIATION REACTOR

DOEpatents

Leyse, C.F.; Putnam, G.E.

1961-05-01

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Characterization of a novel micro-pressure swirl reactor for removal of chemical oxygen demand and total nitrogen from domestic wastewater at low temperature.

PubMed

Ren, Qingkai; Yu, Yang; Zhu, Suiyi; Bian, Dejun; Huo, Mingxin; Zhou, Dandan; Huo, Hongliang

2017-06-01

A novel micro-pressure swirl reactor (MPSR) was designed and applied to treat domestic wastewater at low temperature by acclimating microbial biomass with steadily decreasing temperature from 15 to 3 °C. Chemical oxygen demand (COD) was constantly removed by 85% and maintained below 50 mg L -1 in the effluent during the process. When the air flow was controlled at 0.2 m 3  h -1 , a swirl circulation was formed in the reactor, which created a dissolved oxygen (DO) gradient with a low DO zone in the center and a high DO zone in the periphery for denitrification and nitrification. 81% of total nitrogen was removed by this reactor, in which ammonium was reduced by over 90%. However, denitrification was less effective because of the presence of low levels of oxygen. The progressively decreasing temperature favored acclimation of psychrophilic bacteria in the reactor, which replaced mesophilic bacteria in the process of treatment.

Apparatus for localizing disturbances in pressurized water reactors (PWR)

DOEpatents

Sykora, Dalibor

1989-01-01

The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig.

Actinide management with commercial fast reactors

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

Ohki, Shigeo

The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

JEM Experiment Logistics Module Pressurized Section

NASA Image and Video Library

2007-04-02

An overhead crane moves the JEM Experiment Logistics Module Pressurized Section above the floor of the Space Station Processing Facility to a scale for weight and center-of-gravity measurements. The module will then be moved to a work stand. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

JEM Experiment Logistics Module Pressurized Section

NASA Image and Video Library

2007-04-02

In the Space Station Processing Facility, an overhead crane moves the JEM Experiment Logistics Module Pressurized Section toward a scale (at left) for weight and center-of-gravity measurements. The module will then be moved to a work stand. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

JEM Experiment Logistics Module Pressurized Section

NASA Image and Video Library

2007-04-02

The JEM Experiment Logistics Module Pressurized Section is lifted from its shipping crate in the Space Station Processing Facility. The module will be moved to a scale for weight and center-of-gravity measurements and then to a work stand. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

JEM Experiment Logistics Module Pressurized Section

NASA Image and Video Library

2007-04-02

In the Space Station Processing Facility, the JEM Experiment Logistics Module Pressurized Section is lowered onto a scale for weight and center-of-gravity measurements. The module will then be moved to a work stand. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Investigation of the DSMC Approach for Ion/neutral Species in Modeling Low Pressure Plasma Reactor

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

Deng Hao; Li, Z.; Levin, D.

2011-05-20

Low pressure plasma reactors are important tools for ionized metal physical vapor deposition (IMPVD), a semiconductor plasma processing technology that is increasingly being applied to deposit Cu seed layers on semiconductor surfaces of trenches and vias with the high aspect ratio (e.g., >5:1). A large fraction of ionized atoms produced by the IMPVD process leads to an anisotropic deposition flux towards the substrate, a feature which is critical for attaining a void-free and uniform fill. Modeling such devices is challenging due to their high plasma density, reactive environment, but low gas pressure. A modular code developed by the Computational Opticalmore » and Discharge Physics Group, the Hybrid Plasma Equipment Model (HPEM), has been successfully applied to the numerical investigations of IMPVD by modeling a hollow cathode magnetron (HCM) device. However, as the development of semiconductor devices progresses towards the lower pressure regime (e.g., <5 mTorr), the breakdown of the continuum assumption limits the application of the fluid model in HPEM and suggests the incorporation of the kinetic method, such as the direct simulation Monte Carlo (DSMC), in the plasma simulation.The DSMC method, which solves the Boltzmann equation of transport, has been successfully applied in modeling micro-fluidic flows in MEMS devices with low Reynolds numbers, a feature shared with the HCM. Modeling of the basic physical and chemical processes for ion/neutral species in plasma have been developed and implemented in DSMC, which include ion particle motion due to the Lorentz force, electron impact reactions, charge exchange reactions, and charge recombination at the surface. The heating of neutrals due to collisions with ions and the heating of ions due to the electrostatic field will be shown to be captured by the DSMC simulations. In this work, DSMC calculations were coupled with the modules from HPEM so that the plasma can be self-consistently solved. Differences in the

Japanese Encephalitis: Frequently Asked Questions

MedlinePlus

... the vaccine, what should I do? What is Japanese encephalitis? Japanese encephalitis (JE) is a potentially severe ... cause inflammation of the brain (encephalitis). Where does Japanese encephalitis occur? JE occurs in Asia and parts ...

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

Node 2 and Japanese Experimental Module (JEM) In Space Station Processing Facility

NASA Technical Reports Server (NTRS)

2003-01-01

Lining the walls of the Space Station Processing Facility at the Kennedy Space Center (KSC) are the launch awaiting U.S. Node 2 (lower left). and the first pressurized module of the Japanese Experimental Module (JEM) (upper right), named 'Kibo' (Hope). Node 2, the 'utility hub' and second of three connectors between International Space Station (ISS) modules, was built in the Torino, Italy facility of Alenia Spazio, an International contractor based in Rome. Japan's major contribution to the station, the JEM, was built by the Space Development Agency of Japan (NASDA) at the Tsukuba Space Center near Tokyo and will expand research capabilities aboard the station. Both were part of an agreement between NASA and the European Space Agency (ESA). The Node 2 will be the next pressurized module installed on the Station. Once the Japanese and European laboratories are attached to it, the resulting roomier Station will expand from the equivalent space of a 3-bedroom house to a 5-bedroom house. The Marshall Space Center in Huntsville, Alabama manages the Node program for NASA.

Significant but weak spousal concordance of metabolic syndrome components in Japanese couples.

PubMed

Okuda, Tomoko; Miyazaki, Tadayoshi; Sakuragi, Sonoko; Moriguchi, Jiro; Tachibana, Hiroshi; Ohashi, Fumiko; Ikeda, Masayuki

2014-03-01

This study was initiated to investigate if spousal concordance in metabolic syndrome (MS) components exist in Japan. In all, 756 couples (mean age: 48.9 and 47.3 years for husbands and wives, respectively) were identified. Each subject was classified as an MS, MS reserves (MSRES) or no risk of MS (NonMS) case after Japanese Ministry of Health, Labour and Welfare (JMHLW) criteria. Criteria of the National Cholesterol Education Program and of the Joint Interim Statement were also applied. With Japanese Ministry of Health, Labor and Welfare (JMHLW) criteria, MS, MSRES and NonMS cases accounted for 11.9, 14.7 and 73.4 % in husbands and 1.6, 3.7 and 94.7 % in wives. Waist circumference (WC), body mass index (BMI), systolic blood pressure (SBP) and hemoglobin A1c (HbA1c) showed significant correlation (p < 0.01). Correlation was also significant (p < 0.05) for mean blood pressure (MBP) and fasting plasma glucose (FPG). When adjusted for age, correlations were significant only for WC, BMI and HbA1c. Furthermore, none of the correlation coefficients were greater than 0.2. Logistic regression analyses did not suggest significant mutual influence in MS status between the couples. Spousal concordance in MS components was detected for WC, BMI, SBP, MBP, FPG and HbA1c, but the correlation was generally weak and modest in Japanese couples.

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (right) looks at the newly arrived Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (right) looks at the newly arrived Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (left) looks at the newly arrived Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (left) looks at the newly arrived Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements

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

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent tomore » the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are

Acculturation of Personality: A Three-Culture Study of Japanese, Japanese Americans, and European Americans

PubMed Central

Güngör, Derya; Bornstein, Marc H.; De Leersnyder, Jozefien; Cote, Linda; Ceulemans, Eva; Mesquita, Batja

2013-01-01

The present study tests the hypothesis that involvement with a new culture instigates changes in personality of immigrants that result in (a) better fit with the norms of the culture of destination and (b) reduced fit with the norms of the culture of origin. Participants were 40 Japanese first-generation immigrants to the United States, 57 Japanese monoculturals, and 60 U.S. monoculturals. All participants completed the Jackson Personality Inventory (JPI) as a measure of the Big Five; immigrants completed the Japanese American Acculturation Scale. Immigrants’ fits with the cultures of destination and origin were calculated by correlating Japanese American mothers’ patterns of ratings on the Big Five with the average patterns of ratings of European Americans and Japanese on the same personality dimensions. Japanese Americans became more “American” and less “Japanese” in their personality as they reported higher participation in the U.S. culture. The results support the view that personality can be subject to cultural influence. PMID:23935211

Imaging Fukushima Daiichi reactors with muons

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

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Dioxin pollution disrupts reproduction in male Japanese field mice.

PubMed

Ishiniwa, Hiroko; Sakai, Mizuki; Tohma, Shimon; Matsuki, Hidenori; Takahashi, Yukio; Kajiwara, Hideo; Sekijima, Tsuneo

2013-11-01

Dioxins cause various adverse effects in animals including teratogenesis, induction of drug metabolizing enzymes, tumor promotion, and endocrine disruption. Above all, endocrine disruption is known to disturb reproduction in adult animals and may, also seriously impact their offspring. However, most previous studies have quantified the species-specific accumulation of dioxins, whereas few studies have addressed the physiological impacts of dioxins on wildlife, such as reduced reproductive function. Here we clarify an effect of endocrine disruption caused by dioxins on the Japanese field mouse, Apodemus speciosus. Japanese field mice collected from various sites polluted with dioxins accumulated high concentrations of dioxins in their livers. Some dioxin congeners, especially, 1,2,3,4,6,7,8-heptachlorodibenzo-p-dioxin, 3,3',4,4',5-pentachloro biphenyl, 1,2,3,4,6,7,8-heptachlorodibenzofuran, and octachlorodibenzo-p-dioxin, which showed high biota-soil accumulation factors, contributed to concentration of dioxins in mouse livers with an increase of accumulation of total dioxins. As for physiological effects on the Japanese field mouse, high levels of cytochrome P450 1A1 (CYP1A1) mRNA, a drug metabolizing enzyme induced by dioxins, were found in the livers of mice captured at polluted sites. Furthermore, at such sites polluted with dioxins, increased CYP1A1 expression coincided with reduced numbers of active spermatozoa in mice. Thus, disruption in gametogenesis observed in these mice suggests that dioxins not only negatively impact reproduction among Japanese field mice, but might also act as a kind of selection pressure in a chemically polluted environment.

Homogamy and Intermarriage of Japanese and Japanese Americans With Whites Surrounding World War II.

PubMed

Ono, Hiromi; Berg, Justin

2010-10-01

Although some sociologists have suggested that Japanese Americans quickly assimilated into mainstream America, scholars of Japanese America have highlighted the heightened exclusion that the group experienced. This study tracked historical shifts in the exclusion level of Japanese and Japanese Americans in the United States surrounding World War II with homogamy and intermarriage with Whites for the prewar (1930-1940) and resettlement (1946-1966) marriage cohorts. The authors applied log-linear models to census microsamples (N = 1,590,416) to estimate the odds ratios of homogamy versus intermarriage. The unadjusted odds ratios of Japanese Americans declined between cohorts and appeared to be consistent with the assimilation hypothesis. Once compositional influences and educational pairing patterns were adjusted, however, the odds ratios increased and supported the heightened exclusion hypothesis.

Flooding Experiments and Modeling for Improved Reactor Safety

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

Solmos, M.; Hogan, K. J.; Vierow, K.

2008-09-14

Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge linemore » can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.« less

New techniques for modeling the reliability of reactor pressure vessels

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

Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.

1986-01-01

In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes several new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel wall thickness, and fluence distributions that vary throughout the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper.more » The effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithms for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RT/sub NDT/. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest toughness with subsequent initiation toughnesses.« less

New techniques for modeling the reliability of reactor pressure vessels

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

Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.

1985-12-01

In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel walls thickness, and fluence distributions that vary through-out the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper. Themore » effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithm for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RTNDT. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest thoughness with subsequent initiation toughnesses. 21 refs.« less

Trust in One’s Physician: The Role of Ethnic Match, Autonomy, Acculturation, and Religiosity Among Japanese and Japanese Americans

PubMed Central

Tarn, Derjung M.; Meredith, Lisa S.; Kagawa-Singer, Marjorie; Matsumura, Shinji; Bito, Seiji; Oye, Robert K.; Liu, Honghu; Kahn, Katherine L.; Fukuhara, Shunichi; Wenger, Neil S.

2005-01-01

PURPOSE Trust is a cornerstone of the physician-patient relationship. We investigated the relation of patient characteristics, religiosity, acculturation, physician ethnicity, and insurance-mandated physician change to levels of trust in Japanese American and Japanese patients. METHODS A self-administered, cross-sectional questionnaire in English and Japanese (completed in the language of their choice) was given to community-based samples of 539 English-speaking Japanese Americans, 340 Japanese-speaking Japanese Americans, and 304 Japanese living in Japan. RESULTS Eighty-seven percent of English-speaking Japanese Americans, 93% of Japanese-speaking Japanese Americans, and 58% of Japanese living in Japan responded to trust items and reported mean trust scores of 83, 80, and 68, respectively, on a scale ranging from 0 to 100. In multivariate analyses, English-speaking and Japanese-speaking Japanese American respondents reported more trust than Japanese respondents living in Japan (P values <.001). Greater religiosity (P <.001), less desire for autonomy (P <.001), and physician-patient relationships of longer duration (P <.001) were related to increased trust. Among Japanese Americans, more acculturated respondents reported more trust (P <.001), and Japanese physicians were trusted more than physicians of another ethnicity. Among respondents prompted to change physicians because of insurance coverage, the 48% who did not want to switch reported less trust in their current physician than in their former physician (mean score of 82 vs 89, P <.001). CONCLUSIONS Religiosity, autonomy preference, and acculturation were strongly related to trust in one’s physician among the Japanese American and Japanese samples studied and may provide avenues to enhance the physician-patient relationship. The strong relationship of trust with patient-physician ethnic match and the loss of trust when patients, in retrospect, report leaving a preferred physician suggest unintended

KENNEDY SPACE CENTER, FLA. - Japanese astronaut Koichi Wakata (left) works with a tray extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

NASA Image and Video Library

2003-09-24

KENNEDY SPACE CENTER, FLA. - Japanese astronaut Koichi Wakata (left) works with a tray extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Japanese astronaut Koichi Wakata looks over the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

NASA Image and Video Library

2003-09-24

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Japanese astronaut Koichi Wakata looks over the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

KENNEDY SPACE CENTER, FLA. - Japanese astronaut Koichi Wakata (right) works with a tray extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

NASA Image and Video Library

2003-09-24

KENNEDY SPACE CENTER, FLA. - Japanese astronaut Koichi Wakata (right) works with a tray extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

Experiences of Japanese aged care: the pursuit of optimal health and cultural engagement.

PubMed

Annear, Michael J; Otani, Junko; Sun, Joanna

2016-11-01

Japan is a super-ageing society that faces pressures on its aged care system from a growing population of older adults. Naturalistic observations were undertaken at eight aged care facilities in central and northern Japan to explore how aged care is configured. Four aspects of contemporary provision were identified that offer potential gains in quality of life and health. The Japanese government mandates that aged care facilities must employ a qualified nutritionist to oversee meal preparation, fostering optimal dietary intake. A concept of life rehabilitation seeks to maximise physical and cognitive performance, with possible longevity gains. Low staff to resident ratios are also mandated by the Japanese government to afford residents high levels of interpersonal care. Finally, Japanese facilities prioritise experiences of seasonality and culture, connecting frail older people to the world beyond their walls. © The Author 2016. Published by Oxford University Press on behalf of the British Geriatrics Society. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Effects of crack tip plastic zone on corrosion fatigue cracking of alloy 690(TT) in pressurized water reactor environments

NASA Astrophysics Data System (ADS)

Xiao, J.; Qiu, S. Y.; Chen, Y.; Fu, Z. H.; Lin, Z. X.; Xu, Q.

2015-01-01

Alloy 690(TT) is widely used for steam generator tubes in pressurized water reactor (PWR), where it is susceptible to corrosion fatigue. In this study, the corrosion fatigue behavior of Alloy 690(TT) in simulated PWR environments was investigated. The microstructure of the plastic zone near the crack tip was investigated and labyrinth structures were observed. The relationship between the crack tip plastic zone and fatigue crack growth rates and the environment factor Fen was illuminated.

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

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

Yun, D.; Taiwo, T. A.; Kim, T. K.

2010-10-01

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluatemore » the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.« less

The modern Japanese color lexicon.

PubMed

Kuriki, Ichiro; Lange, Ryan; Muto, Yumiko; Brown, Angela M; Fukuda, Kazuho; Tokunaga, Rumi; Lindsey, Delwin T; Uchikawa, Keiji; Shioiri, Satoshi

2017-03-01

Despite numerous prior studies, important questions about the Japanese color lexicon persist, particularly about the number of Japanese basic color terms and their deployment across color space. Here, 57 native Japanese speakers provided monolexemic terms for 320 chromatic and 10 achromatic Munsell color samples. Through k-means cluster analysis we revealed 16 statistically distinct Japanese chromatic categories. These included eight chromatic basic color terms (aka/red, ki/yellow, midori/green, ao/blue, pink, orange, cha/brown, and murasaki/purple) plus eight additional terms: mizu ("water")/light blue, hada ("skin tone")/peach, kon ("indigo")/dark blue, matcha ("green tea")/yellow-green, enji/maroon, oudo ("sand or mud")/mustard, yamabuki ("globeflower")/gold, and cream. Of these additional terms, mizu was used by 98% of informants, and emerged as a strong candidate for a 12th Japanese basic color term. Japanese and American English color-naming systems were broadly similar, except for color categories in one language (mizu, kon, teal, lavender, magenta, lime) that had no equivalent in the other. Our analysis revealed two statistically distinct Japanese motifs (or color-naming systems), which differed mainly in the extension of mizu across our color palette. Comparison of the present data with an earlier study by Uchikawa & Boynton (1987) suggests that some changes in the Japanese color lexicon have occurred over the last 30 years.

Homogamy and Intermarriage of Japanese and Japanese Americans With Whites Surrounding World War II

PubMed Central

Ono, Hiromi; Berg, Justin

2010-01-01

Although some sociologists have suggested that Japanese Americans quickly assimilated into mainstream America, scholars of Japanese America have highlighted the heightened exclusion that the group experienced. This study tracked historical shifts in the exclusion level of Japanese and Japanese Americans in the United States surrounding World War II with homogamy and intermarriage with Whites for the prewar (1930–1940) and resettlement (1946–1966) marriage cohorts. The authors applied log-linear models to census microsamples (N = 1,590,416) to estimate the odds ratios of homogamy versus intermarriage. The unadjusted odds ratios of Japanese Americans declined between cohorts and appeared to be consistent with the assimilation hypothesis. Once compositional influences and educational pairing patterns were adjusted, however, the odds ratios increased and supported the heightened exclusion hypothesis. PMID:21116449

Impact of insulin resistance, insulin and adiponectin on kidney stones in the Japanese population.

PubMed

Ando, Ryosuke; Suzuki, Sadao; Nagaya, Teruo; Yamada, Tamaki; Okada, Atsushi; Yasui, Takahiro; Tozawa, Keiichi; Tokudome, Shinkan; Kohri, Kenjiro

2011-02-01

It has been reported that kidney stones are linked to metabolic syndrome (MetS), which is characterized by insulin resistance. The aim of the present study was to examine the association of insulin resistance, insulin and adiponectin with kidney stones in a Japanese population. From February 2007 to March 2008, 1036 (529 men and 507 women) apparently healthy Japanese subjects, aged 35-79 years, were analyzed. Weight, height, waist circumference and blood pressure were measured. Overnight fasting blood was collected to measure insulin and adiponectin levels. Homeostasis model assessment of insulin resistance (HOMA-IR) was calculated to assess insulin resistance. Logistic regression analysis was used to estimate the odds ratio (OR) and 95% confidence intervals for a self-reported history of kidney stones across tertiles of HOMA-IR, insulin and adiponectin. Of the participants, 84 men (15.6%) and 35 women (6.9%) had a history of kidney stones. Age, body mass index, waist circumference, systolic and diastolic blood pressures, HOMA-IR and insulin were significantly higher in women with than in women without kidney stones. There was no difference in adiponectin level between subjects with and without a history of kidney stones in either sex. Furthermore, a significant positive trend was observed in the age-adjusted OR for a history of kidney stones across insulin tertiles (P-value for trend = 0.04) in women. For Japanese women, HOMA-IR and insulin are associated with a history of kidney stones. The findings suggest that MetS components could increase the risk of kidney stones through subclinical hyperinsulinemia and insulin resistance. © 2010 The Japanese Urological Association.

Creep failure of a reactor pressure vessel lower head under severe accident conditions

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

Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y.

A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In ordermore » to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.« less

Modeling the shear rate and pressure drop in a hydrodynamic cavitation reactor with experimental validation based on KI decomposition studies.

PubMed

Badve, Mandar P; Alpar, Tibor; Pandit, Aniruddha B; Gogate, Parag R; Csoka, Levente

2015-01-01

A mathematical model describing the shear rate and pressure variation in a complex flow field created in a hydrodynamic cavitation reactor (stator and rotor assembly) has been depicted in the present study. The design of the reactor is such that the rotor is provided with surface indentations and cavitational events are expected to occur on the surface of the rotor as well as within the indentations. The flow characteristics of the fluid have been investigated on the basis of high accuracy compact difference schemes and Navier-Stokes method. The evolution of streamlining structures during rotation, pressure field and shear rate of a Newtonian fluid flow have been numerically established. The simulation results suggest that the characteristics of shear rate and pressure area are quite different based on the magnitude of the rotation velocity of the rotor. It was observed that area of the high shear zone at the indentation leading edge shrinks with an increase in the rotational speed of the rotor, although the magnitude of the shear rate increases linearly. It is therefore concluded that higher rotational speeds of the rotor, tends to stabilize the flow, which in turn results into less cavitational activity compared to that observed around 2200-2500RPM. Experiments were carried out with initial concentration of KI as 2000ppm. Maximum of 50ppm of iodine liberation was observed at 2200RPM. Experimental as well as simulation results indicate that the maximum cavitational activity can be seen when rotation speed is around 2200-2500RPM. Copyright © 2014 Elsevier B.V. All rights reserved.

324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

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

HUMPHREYS, D C

A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Teammore » counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.« less

Reactor

DOEpatents

Evans, Robert M.

1976-10-05

1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

PUSH-PULL POWER REACTOR

DOEpatents

Froman, D.K.

1959-02-24

Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

Analysis on the Spatial Difference of Bacterial Community Structure in Micro-pressure Air-lift Loop Reactor

NASA Astrophysics Data System (ADS)

Wan, L. G.; Lin, Q.; Bian, D. J.; Ren, Q. K.; Xiao, Y. B.; Lu, W. X.

2018-02-01

In order to reveal the spatial difference of the bacterial community structure in the Micro-pressure Air-lift Loop Reactor, the activated sludge bacterial at five different representative sites in the reactor were studied by denaturing gradient gel electrophoresis (DGGE). The results of DGGE showed that the difference of environmental conditions (such as substrate concentration, dissolved oxygen and PH, etc.) resulted in different diversity and similarity of microbial flora in different spatial locations. The Shannon-Wiener diversity index of the total bacterial samples from five sludge samples varied from 0.92 to 1.28, the biodiversity index was the smallest at point 5, and the biodiversity index was the highest at point 2. The similarity of the flora between the point 2, 3 and 4 was 80% or more, respectively. The similarity of the flora between the point 5 and the other samples was below 70%, and the similarity of point 2 was only 59.2%. Due to the different contribution of different strains to the removal of pollutants, it can give full play to the synergistic effect of bacterial degradation of pollutants, and further improve the efficiency of sewage treatment.

NUCLEAR REACTOR

DOEpatents

Christy, R.F.

1958-07-15

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

CONTROL RODS FOR NUCLEAR REACTOR CORES

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

Bell, F.R.

1961-11-15

A reactor control rod is designed which has increased effectiveness as compared with the width of the aperture in the pressure vessel through which it passes. The control rod carries six fins, three on each side, and two of the fins are fixed while the other, being adjustable, is capable of movement from between the fixed fins to an extended position. Thus, the control rod assembly can be arranged so that the parts within the core form a substantially complete shell around the reactor central axis, while the apertures on the pressure vessel wall are well spaced for strength. (D.L.C.)

Characterization of codon usage pattern and influencing factors in Japanese encephalitis virus.

PubMed

Singh, Niraj K; Tyagi, Anuj; Kaur, Rajinder; Verma, Ramneek; Gupta, Praveen K

2016-08-02

Recently, several outbreaks of Japanese encephalitis (JE), caused by Japanese encephalitis virus (JEV), have been reported and it has become cause of concern across the world. In this study, detailed analysis of JEV codon usage pattern was performed. The relative synonymous codon usage (RSCU) values along with mean effective number of codons (ENC) value of 55.30 indicated the presence of low codon usages bias in JEV. The effect of mutational pressure on codon usage bias was confirmed by significant correlations of A3s, U3s, G3s, C3s, GC3s, ENC values, with overall nucleotide contents (A%, U%, G%, C%, and GC%). The correlation analysis of A3s, U3s, G3s, C3s, GC3s, with axis values of correspondence analysis (CoA) further confirmed the role of mutational pressure. However, the correlation analysis of Gravy values and Aroma values with A3s, U3s, G3s, C3s, and GC3s, indicated the presence of natural selection on codon usage bias in addition to mutational pressure. The natural selection was further confirmed by codon adaptation index (CAI) analysis. Additionally, relative dinucleotide frequencies, geographical distribution, and evolutionary processes also influenced the codon usage pattern to some extent. Copyright © 2016 Elsevier B.V. All rights reserved.

SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

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

Hoffman, William M.; Riley, Matthew E.; Spencer, Benjamin W.

In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vesselsmore » – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which

Cultural Competence in Business Japanese.

ERIC Educational Resources Information Center

Koike, Shohei

Cultural competence in business Japanese requires more than superficial knowledge of business etiquette. One must truly understand why Japanese people think and act differently from their American counterparts. For example, instruction in the use of Japanese taxis must be accompanied by instruction in the concept and implications of seating order…

The evolution of Japanese employer-sponsored retirement plans.

PubMed

Rajnes, David

2007-01-01

make DC plans more attractive to employers in Japan are likely to be implemented in the near future. This article summarizes the Japanese retirement system, with an emphasis on private-sector employees, and the complementary role played by voluntary employer-sponsored retirement plans; describes the financial pressures that faced retirement plan sponsors in the late twentieth century and the factors motivating the reform of Japanese voluntary retirement plans; examines the 2001 legislative changes that have transformed company retirement plans; and concludes with a review of trends and recent developments in employer-sponsored retirement plans since the implementation of the 2001 pension laws.

Statistics in Japanese universities.

PubMed Central

Ito, P K

1979-01-01

The teaching of statistics in the U.S. and Japanese universities is briefly reviewed. It is found that H. Hotelling's articles and subsequent relevant publications on the teaching of statistics have contributed to a considerable extent to the establishment of excellent departments of statistics in U.S. universities and colleges. Today the U.S. may be proud of many well-staffed and well-organized departments of theoretical and applied statistics with excellent undergraduate and graduate programs. On the contrary, no Japanese universities have an independent department of statistics at present, and the teaching of statistics has been spread among a heterogeneous group of departments of application. This was mainly due to the Japanese government regulation concerning the establishment of a university. However, it has recently been revised so that an independent department of statistics may be started in a Japanese university with undergraduate and graduate programs. It is hoped that discussions will be started among those concerned on the question of organization of the teaching of statistics in Japanese universities as soon as possible. PMID:396154

JEM Experiment Logistics Module Pressurized Section

NASA Image and Video Library

2007-04-02

In the Space Station Processing Facility, an overhead crane lifts the JEM Experiment Logistics Module Pressurized Section from its shipping container and moves it toward a scale for weight and center-of-gravity measurements. The module will then be moved to a work stand. The logistics module is one of the components of the Japanese Experiment Module or JEM, also known as Kibo, which means "hope" in Japanese. Kibo comprises six components: two research facilities -- the Pressurized Module and Exposed Facility; a Logistics Module attached to each of them; a Remote Manipulator System; and an Inter-Orbit Communication System unit. Kibo also has a scientific airlock through which experiments are transferred and exposed to the external environment of space. Kibo is Japan's first human space facility and its primary contribution to the station. Kibo will enhance the unique research capabilities of the orbiting complex by providing an additional environment in which astronauts can conduct science experiments. The various components of JEM will be assembled in space over the course of three Space Shuttle missions. The first of those three missions, STS-123, will carry the Experiment Logistics Module Pressurized Section aboard the Space Shuttle Endeavour, targeted for launch in 2007.

Basic English Writers' Japanese-English Wordbook.

ERIC Educational Resources Information Center

Daniels, F. J.

The author of this Japanese-English wordbook suggests that it may be used by Japanese writers of English, by those translating from Japanese into English, and by learners of Japanese, in addition to its main intended uses as an aid to the preparation of teaching material and as a work of reference for teachers. A translator will need to supplement…

High-pressure anaerobic digestion up to 100 bar: influence of initial pressure on production kinetics and specific methane yields.

PubMed

Merkle, Wolfgang; Baer, Katharina; Haag, Nicola Leonard; Zielonka, Simon; Ortloff, Felix; Graf, Frank; Lemmer, Andreas

2017-02-01

To ensure an efficient use of biogas produced by anaerobic digestion, in some cases it would be advisable to upgrade the biogenic gases and inject them into the transnational gas grids. To investigate biogas production under high-pressure conditions up to 100 bar, new pressure batch methane reactors were developed for preliminary lab-scale experiments with a mixture of grass and maize silage hydrolysate. During this investigation, the effects of different initial pressures (1, 50 and 100 bar) on pressure increase, gas production and the specific methane yield using nitrogen as inert gas were determined. Based on the experimental findings increasing initial pressures alter neither significantly, further pressure increases nor pressure increase rates. All supplied organic acids were degraded and no measurable inhibition of the microorganisms was observed. The results show that methane reactors can be operated at operating pressures up to 100 bar without any negative effects on methane production.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

Japanese Nationalism

DTIC Science & Technology

1989-06-01

United States. The chief function of this principle is to cut--it separates all things. It classifies everything into black and white, good and bad . The...content included articles on masturbation , petting, and 99 intercourse. One of Japan’s all time best selling books in recent years, Totto-chan, is a...to the th: every Japanese will be judged by whether he celebrates this or not. That is how people will be determined to be good Japanese or bad 112

[The alteration of Japanese anatomical terminology in the early Showa period and the Japanese language reform campaign].

PubMed

Sawai, Tadashi; Sakai, Tatsuo

2010-03-01

In the second decade of the Showa period, great changes were made in the Japanese anatomical terms. It has been proposed that the presentation of JNA (Jenaer nomina anatomica) was one of the factors leading to the change. The Japanese language reform campaign, however, played an important role. The party kokugoaigo doumei and its successor kokugo kyokai required concise and unified technical terms. The anatomical nomenclature committee of the Japanese Association of Anatomists worked to satisfy this requirement. The committee consulted with nomenclature committees of other medical associations and took account of their opinions. The anatomical nomenclature committee abandoned the literal translation from Latin to Japanese and shaped a succinct Japanese terminology. Modern Japanese anatomical terms are based on this terminology.

Reactor on-off antineutrino measurement with KamLAND

NASA Astrophysics Data System (ADS)

Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, H.; Xu, B. D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T. I.; Fujikawa, B. K.; Han, K.; O'Donnell, T.; Berger, B. E.; Learned, J. G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Detwiler, J. A.; Enomoto, S.; Decowski, M. P.

2013-08-01

The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved sensitivity for other ν¯e signals, in particular ν¯e’s produced in β-decays from U238 and Th232 within the Earth’s interior, whose energy spectrum overlaps with that of reactor ν¯e’s. Including constraints on θ13 from accelerator and short-baseline reactor neutrino experiments, a combined three-flavor analysis of solar and KamLAND data gives fit values for the oscillation parameters of tan⁡2θ12=0.436-0.025+0.029, Δm212=7.53-0.18+0.18×10-5eV2, and sin⁡2θ13=0.023-0.002+0.002. Assuming a chondritic Th/U mass ratio, we obtain 116-27+28 ν¯e events from U238 and Th232, corresponding to a geo ν¯e flux of 3.4-0.8+0.8×106cm-2s-1 at the KamLAND location. We evaluate various bulk silicate Earth composition models using the observed geo ν¯e rate.

Nuclear reactor building

DOEpatents

Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

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

Spencer, Benjamin; Backman, Marie; Chakraborty, Pritam

2015-03-01

Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode Imore » loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.« less

Compliment Responses: Comparing American Learners of Japanese, Native Japanese Speakers, and American Native English Speakers

ERIC Educational Resources Information Center

Tatsumi, Naofumi

2012-01-01

Previous research shows that American learners of Japanese (AJs) tend to differ from native Japanese speakers in their compliment responses (CRs). Yokota (1986) and Shimizu (2009) have reported that AJs tend to respond more negatively than native Japanese speakers. It has also been reported that AJs' CRs tend to lack the use of avoidance or…

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Performance assessment of low pressure nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Gerrish, Harrold P., Jr.; Doughty, Glen E.

1993-01-01

An increase in Isp for nuclear thermal propulsion systems is desirable for reducing the propellant requirements and cost of future applications, such as the Mars Transfer Vehicle. Several previous design studies have suggested that the Isp could be increased substantially with hydrogen dissociation/recombination. Hydrogen molecules (H2), at high temperatures and low pressures, will dissociate to monatomic hydrogen (H). The reverse process (i.e., formation of H2 from H) is exothermic. The exothermic energy in a nozzle increases the kinetic energy and therefore, increases the Isp. The low pressure nuclear thermal propulsion system (LPNTP) system is expected to maximize the hydrogen dissociation/recombination and Isp by operating at high chamber temperatures and low chamber pressures. The process involves hydrogen flow through a high temperature, low pressure fission reactor, and out a nozzle. The high temperature (approximately 3000 K) of the hydrogen in the reactor is limited by the temperature limits of the reactor material. The minimum chamber pressure is about 1 atm because lower pressures decrease the engines thrust to weight ratio below acceptable limits. This study assumes that hydrogen leaves the reactor and enters the nozzle at the 3000 K equilibrium dissociation level. Hydrogen dissociation in the reactor does not affect LPNTP performance like dissociation in traditional chemical propulsion systems, because energy from the reactor resupplies energy lost due to hydrogen dissociation. Recombination takes place in the nozzle due primarily to a drop in temperature as the Mach number increases. However, as the Mach number increases beyond the nozzle throat, the static pressure and density of the flow decreases and minimizes the recombination. The ideal LPNTP Isp at 3000 K and 10 psia is 1160 seconds due to the added energy from fast recombination rates. The actual Isp depends on the finite kinetic reaction rates which affect the amount of monatomic hydrogen

Thermal swing reactor including a multi-flight auger

DOEpatents

Ermanoski, Ivan

2017-03-07

A thermal swing reactor including a multi-flight auger and methods for solar thermochemical reactions are disclosed. The reactor includes a multi-flight auger having different helix portions having different pitch. Embodiments of reactors include at least two distinct reactor portions between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between portions during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and

Alcohol synthesis in a high-temperature slurry reactor

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

Roberts, G.W.; Marquez, M.A.; McCutchen, M.S.

1995-12-31

The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system canmore » be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.« less

Fluidized bed coal combustion reactor

NASA Technical Reports Server (NTRS)

Moynihan, P. I.; Young, D. L. (Inventor)

1981-01-01

A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

Efficacy and Safety of Adjunctive Modafinil Treatment on Residual Excessive Daytime Sleepiness among Nasal Continuous Positive Airway Pressure-Treated Japanese Patients with Obstructive Sleep Apnea Syndrome: A Double-Blind Placebo-Controlled Study

PubMed Central

Inoue, Yuichi; Takasaki, Yuji; Yamashiro, Yoshihiro

2013-01-01

Study Objectives: This double-blind study evaluated the efficacy and safety of modafinil for treating excessive daytime sleepiness in Japanese patients with obstructive sleep apnea syndrome (OSAS). Methods: Patients with residual excessive sleepiness (Epworth Sleepiness Scale [ESS] ≥ 11) on optimal nasal continuous positive airway pressure (nCPAP) therapy (apnea-hypopnea index ≤ 10) were randomized to either 200 mg modafinil (n = 52) or placebo (n = 62) once daily for 4 weeks. Outcomes included baseline-week 4 changes in ESS total score, sleep latency on maintenance of wakefulness test (SL-MWT), nocturnal polysomnography, Pittsburgh Sleep Quality Index (PSQI), and safety. Results: All 114 randomized patients completed the study. Mean change in ESS total score (-6.6 vs -2.4, p < 0.001) and SL-MWT (+2.8 vs -0.4 minutes, p = 0.009) were significantly greater with modafinil than with placebo. ESS total score decreased from > 11 to < 11 at the final assessment in 69.2% of modafinil-treated patients and 30.6% of placebo-treated patients (p < 0.001). Corresponding rates at week 1 were 57.7% and 33.9% (p = 0.014). Changes in nocturnal polysomnography, PSQI, and apnea-hypopnea index from baseline to the final assessment were similar in both groups. Adverse drug reactions occurred in 36.5% and 22.6% of patients in the modafinil and placebo groups, respectively (p = 0.146). Conclusions: Once-daily modafinil was effective and well tolerated for managing residual daytime sleepiness in Japanese OSAS patients with residual excessive daytime sleepiness on optimal nCPAP therapy. Clinical Trial Registration: JapicCTI-No.090777 Citation: Inoue Y; Takasaki Y; Yamashiro Y. Efficacy and safety of adjunctive modafinil treatment on residual excessive daytime sleepiness among nasal continuous positive airway pressure-treated Japanese patients with obstructive sleep apnea syndrome: a double-blind placebo-controlled study. J Clin Sleep Med 2013;9(8):751-757. PMID:23946704

Structural Integrity of Water Reactor Pressure Boundary Components.

DTIC Science & Technology

1980-08-01

Boiler and Pressure Vessel Code , Sec. Ill). Estimates of the upper shelf K level from small-specimen...from Appendix A of Section XI of the ASME Boiler and Pressure Vessel Code [11. Figure 9 shows this same data set, together with earlier data for...0969, NRL Memo- randum Report 4063, Sep. 1979. 11. Section XI - ASME Boiler and Pressure Vessel Code , Rules for Inservice Inspection of Nuclear

Japanese American Identity Dilemma.

ERIC Educational Resources Information Center

Maykovich, Minako K.

The major theme of this book is the label "Quiet American" for the Japanese American. In order to locate Japanese Americans sociologically and psychologically in the structure of American society, various concepts such as "marginal man,""alienation," and "inauthenticity" are examined, specifying these…

Corrigendum to “Atom probe tomography characterization of neutron irradiated surveillance samples from the R.E. Ginna reactor pressure vessel”

DOE PAGES

Edmondson, Philip D.; Miller, Michael K.; Powers, K. A.; ...

2017-03-24

In our recent paper entitled “Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel”, we make reference to a table within the article as providing the average compositions of the precipitates, when in fact the bulk compositions were given. In this correction, we present the average precipitate compositions for the data presented in Ref. [1]. These correct compositions are provided for information and do not alter the conclusions of the original manuscript.

High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

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

Odette, G. Robert

Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences thanmore » have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.« less

Catalytic reactor

DOEpatents

Aaron, Timothy Mark [East Amherst, NY; Shah, Minish Mahendra [East Amherst, NY; Jibb, Richard John [Amherst, NY

2009-03-10

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Numerical Simulations of a 96-rod Polysilicon CVD Reactor

NASA Astrophysics Data System (ADS)

Guoqiang, Tang; Cong, Chen; Yifang, Cai; Bing, Zong; Yanguo, Cai; Tihu, Wang

2018-05-01

With the rapid development of the photovoltaic industry, pressurized Siemens belljar-type polysilicon CVD reactors have been enlarged from 24 rods to 96 rods in less than 10 years aimed at much greater single-reactor productivity. A CFD model of an industry-scale 96-rod CVD reactor was established to study the internal temperature distribution and the flow field of the reactor. Numerical simulations were carried out and compared with actual growth results from a real CVD reactor. Factors affecting polysilicon depositions such as inlet gas injections, flow field, and temperature distribution in the CVD reactor are studied.

Population structure in Japanese rice population

PubMed Central

Yamasaki, Masanori; Ideta, Osamu

2013-01-01

It is essential to elucidate genetic diversity and relationships among even related individuals and populations for plant breeding and genetic analysis. Since Japanese rice breeding has improved agronomic traits such as yield and eating quality, modern Japanese rice cultivars originated from narrow genetic resource and closely related. To resolve the population structure and genetic diversity in Japanese rice population, we used a total of 706 alleles detected by 134 simple sequence repeat markers in a total of 114 cultivars composed of 94 improved varieties and 20 landraces, which are representative and important for Japanese rice breeding. The landraces exhibit greater gene diversity than improved lines, suggesting that landraces can provide additional genetic diversity for future breeding. Model-based Bayesian clustering analysis revealed six subgroups and admixture situation in the cultivars, showing good agreement with pedigree information. This method could be superior to phylogenetic method in classifying a related population. The leading Japanese rice cultivar, Koshihikari is unique due to the specific genome constitution. We defined Japanese rice diverse sets that capture the maximum number of alleles for given sample sizes. These sets are useful for a variety of genetic application in Japanese rice cultivars. PMID:23641181

Non-Native Japanese Listeners' Perception of Vowel Length Contrasts in Japanese and Modern Standard Arabic (MSA)

ERIC Educational Resources Information Center

Tsukada, Kimiko

2012-01-01

This study aimed to compare the perception of short vs. long vowel contrasts in Japanese and Modern Standard Arabic (MSA) by four groups of listeners differing in their linguistic backgrounds: native Arabic (NA), native Japanese (NJ), non-native Japanese (NNJ) and Australian English (OZ) speakers. The NNJ and OZ groups shared the first language…

Body composition and anthropometry in Japanese and Australian Caucasian males and Japanese females.

PubMed

Kagawa, Masaharu; Binns, Colin B; Hills, Andrew P

2007-01-01

The total amount and location of fat deposition are important factors in the development of obesity and the metabolic syndrome. To date there have been no reported studies of ethnic and gender differences in body composition and fat distribution patterns in Japanese and Australian young adults. The aim of this study was to assess body composition of young Japanese and Australian Caucasian adults using whole-body dual energy x-ray absorptiometry (DXA) and anthropometry to examine body fat deposition patterns. Body composition of 45 Japanese males and 42 Australian Caucasian males living in Australia (aged 18-40 years) and 139 Japanese females living in Japan (aged 18-27 years) were measured using whole-body DXA scanning and anthropometry. Differences in relationships between BMI and waist circumference (WC), sum of skinfolds (SigmaSF) and %BF obtained from DXA were assessed using multivariate analyses. Distinct gender and ethnic differences (p<0.05) in bone density and waist circumference were observed but no gender differences in BMI and bone mineral content and no ethnic differences in sum of skinfolds and %BF. Both Japanese males and females showed a greater %BF at given BMI, WC and SigmaSF values (p<0.05). The results indicate differences in relationships between %BF and anthropometric measures in young Japanese compared to Caucasians and the importance of population-specific cut-off points for these indices. These findings also have implications for the development of chronic disease and further research, including studies in other Asian countries, is recommended.

The Japanese containerless experiments

NASA Technical Reports Server (NTRS)

Azuma, Hisao

1990-01-01

There are three sets of Japanese containerless experiments. The first is Drop dynamics research. It consists of acoustic levitation and large amplitude drop oscillation. The second is Optical materials processing in an acoustic levitation furnace. And the third is Electrostatic levitator development by two different Japanese companies.

Intercultural Communication Problems in Japanese Multinationals.

ERIC Educational Resources Information Center

Nishiyama, Kazuo

Many large Japanese-owned multinational corporations have established successful subsidiaries in the United States, but distinct ethnic and cultural differences have caused communication problems between Japanese managers and American laborers and business people. Many top executives of the Japanese subsidiaries are sent to the United States on a…

Shattering Myths: Japanese American Educational Issues.

ERIC Educational Resources Information Center

Yoshiwara, Florence M.

An historical review of the immigration and resettlement patterns, and a demographic profile of Japanese Americans reveals a myth of the "successful minority." Since the founding of the Japanese American Citizens League in 1928, Japanese Americans have defeated alien land laws, discriminatory immigration quotas, anti-miscengenation laws,…

[History of Japanese Committee for Anatomical Nomenclature].

PubMed

Kimura, Kunihiko

2008-12-01

This paper records a history of the Japanese Committee of Anatomical Nomenclature since 1990, as a supplement to the previous report (1991), explains a progressing of the edition of Japanese medical terms by the Japanese Association of Medical Sciences and the Ministry of Education, Sciences and Culture, and points out of some problems on terms in Japanese.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Japanese Media in English.

ERIC Educational Resources Information Center

Tanaka, Sachiko Oda

1995-01-01

Describes the use of English in the media in Japan, focusing on the role and history of English-language newspapers, radio, and television programs, as well as the proliferation of English-language films shown in Japanese cinemas. Discusses the implications of English in the Japanese media. (20 references) (MDM)

Ordinal Expressions in Japanese. Papers in Japanese Linguistics, Vol. 2, No. 1.

ERIC Educational Resources Information Center

Backus, Robert L.

The varied forms and semantic factors of Japanese ordinal expressions are related to one another in a coherent system. In Japanese, the cardinal number form is a numeral compound in construction with a referent. The numeral compound consists of a number and a numeral adjunct. Numeral adjuncts are derived from bound forms, or numeral suffixes, and…

Japanese Interest in “Hotaru” (Fireflies) and “Kabuto-Mushi” (Japanese Rhinoceros Beetles) Corresponds with Seasonality in Visible Abundance

PubMed Central

Takada, Kenta

2012-01-01

Seasonal changes in the popularity of fireflies [usually Genji-fireflies (Luciola cruciata Motschulsky) in Japan] and Japanese rhinoceros beetles [Allomyrina dichotoma (Linne)] were investigated to examine whether contemporary Japanese are interested in visible emergence of these insects as seasonal events. The popularity of fireflies and Japanese rhinoceros beetles was assessed by the Google search volume of their Japanese names, “Hotaru” and “Kabuto-mushi” in Japanese Katakana script using Google Trends. The search volume index for fireflies and Japanese rhinoceros beetles was distributed across seasons with a clear peak in only particular times of each year from 2004 to 2011. In addition, the seasonal peak of popularity for fireflies occurred at the beginning of June, whereas that for Japanese rhinoceros beetles occurred from the middle of July to the beginning of August. Thus seasonal peak of each species coincided with the peak period of the emergence of each adult stage. These findings indicated that the Japanese are interested in these insects primarily during the time when the two species are most visibly abundant. Although untested, this could suggest that fireflies and Japanese rhinoceros beetles are perceived by the general public as indicators or symbols of summer in Japan. PMID:26466535

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

DOEpatents

Ho, Darwin D.; Kulsrud, Russell M.

1991-01-01

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

Nuclear reactor building

DOEpatents

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

77 FR 36581 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-19

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor (US-APWR) will hold a meeting on July 9-10, 2012, Room T-2B3, 11545...

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (center) joins others for a tour. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto (center) joins others for a tour. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module.

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto points to other Space Station elements. Behind him is the Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Executive Director of NASDA Koji Yamamoto points to other Space Station elements. Behind him is the Japanese Experiment Module (JEM)/pressurized module. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of JEM.

First Course in Japanese: Character Workbook.

ERIC Educational Resources Information Center

Niwa, Tamako

This character workbook is an introduction to Japanese writing designed to be used in conjunction with Parts One and Two of this introductory course in Japanese. All the "hiragana", several "katakana", and 88 Japanese characters are introduced in this text. The workbook, consisting of 30 lessons, is divided into three parts.…

Developing Instructional Materials for Business Japanese.

ERIC Educational Resources Information Center

Koike, Shohei

Business Japanese should be the study of Japanese language and culture for business communication and should include values and beliefs and institutional constraints on which the Japanese act as well as business etiquette and terminology. Topics to be covered in instruction will vary depending on the role (seller, buyer, or colleague) played by…

Pressure polymerization of polyester

DOEpatents

Maurer, Charles J.; Shaw, Gordon; Smith, Vicky S.; Buelow, Steven J.; Tumas, William; Contreras, Veronica; Martinez, Ronald J.

2000-08-29

A process is disclosed for the preparation of a polyester polymer or polyester copolymer under superatmospheric pressure conditions in a pipe or tubular reaction under turbannular flow conditions. Reaction material having a glycol equivalents to carboxylic acid equivalents mole ratio of from 1.0:1 to 1.2:1, together with a superatmospheric dense gaseous medium are fed co-currently to the reactor. Dicarboxylic acid and/or diol raw materials may be injected into any of the reaction zones in the process during operation to achieve the overall desired mole ratio balance. The process operates at temperatures of from about 220.degree. C. to about 320.degree. C., with turbannular flow achieved before the polymer product and gas exit the reactor process. The pressure in the reaction zones can be in the range from 15 psia to 2500 psia. A polymer product having a DP of a greater than 40, more preferably at least about 70, is achieved by the transfer of water from the reacting material polymer melt to the gaseous medium in the reactor.

Japanese Quality Control Circles.

ERIC Educational Resources Information Center

Nishiyama, Kazuo

In recent years, United States scholars with an interest in international business and organizational communication have begun to notice the success of Japanese "quality control circles." These are small groups, usually composed of seven to ten workers, who are organized at the production levels within most large Japanese factories. A…

Hydrodynamics of Packed Bed Reactor in Low Gravity

NASA Technical Reports Server (NTRS)

Motil, Brian J.; Nahra, Henry K.; Balakotaiah, Vemuri

2005-01-01

Packed bed reactors are well known for their vast and diverse applications in the chemical industry; from gas absorption, to stripping, to catalytic conversion. Use of this type of reactor in terrestrial applications has been rather extensive because of its simplicity and relative ease of operation. Developing similar reactors for use in microgravity is critical to many space-based advanced life support systems. However, the hydrodynamics of two-phase flow packed bed reactors in this new environment and the effects of one physiochemical process on another has not been adequately assessed. Surface tension or capillary forces play a much greater role which results in a shifting in flow regime transitions and pressure drop. Results from low gravity experiments related to flow regimes and two-phase pressure drop models are presented in this paper along with a description of plans for a flight experiment on the International Space Station (ISS). Understanding the packed bed hydrodynamics and its effects on mass transfer processes in microgravity is crucial for the design of packed bed chemical or biological reactors to be used for water reclamation and other life support processes involving water purification.

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-12-15

A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

NASA Astrophysics Data System (ADS)

van Duysen, J. C.; Meric de Bellefon, G.

2017-02-01

The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

High Throughput Atomic Layer Deposition Processes: High Pressure Operations, New Reactor Designs, and Novel Metal Processing

NASA Astrophysics Data System (ADS)

Mousa, MoatazBellah Mahmoud

Atomic Layer Deposition (ALD) is a vapor phase nano-coating process that deposits very uniform and conformal thin film materials with sub-angstrom level thickness control on various substrates. These unique properties made ALD a platform technology for numerous products and applications. However, most of these applications are limited to the lab scale due to the low process throughput relative to the other deposition techniques, which hinders its industrial adoption. In addition to the low throughput, the process development for certain applications usually faces other obstacles, such as: a required new processing mode (e.g., batch vs continuous) or process conditions (e.g., low temperature), absence of an appropriate reactor design for a specific substrate and sometimes the lack of a suitable chemistry. This dissertation studies different aspects of ALD process development for prospect applications in the semiconductor, textiles, and battery industries, as well as novel organic-inorganic hybrid materials. The investigation of a high pressure, low temperature ALD process for metal oxides deposition using multiple process chemistry revealed the vital importance of the gas velocity over the substrate to achieve fast depositions at these challenging processing conditions. Also in this work, two unique high throughput ALD reactor designs are reported. The first is a continuous roll-to-roll ALD reactor for ultra-fast coatings on porous, flexible substrates with very high surface area. While the second reactor is an ALD delivery head that allows for in loco ALD coatings that can be executed under ambient conditions (even outdoors) on large surfaces while still maintaining very high deposition rates. As a proof of concept, part of a parked automobile window was coated using the ALD delivery head. Another process development shown herein is the improvement achieved in the selective synthesis of organic-inorganic materials using an ALD based process called sequential vapor

Assessing the impact of electrolyte conductivity and viscosity on the reactor cost and pressure drop of redox-active polymer flow batteries

NASA Astrophysics Data System (ADS)

Iyer, Vinay A.; Schuh, Jonathon K.; Montoto, Elena C.; Pavan Nemani, V.; Qian, Shaoyi; Nagarjuna, Gavvalapalli; Rodríguez-López, Joaquín; Ewoldt, Randy H.; Smith, Kyle C.

2017-09-01

Redox-active small molecules, used traditionally in redox flow batteries (RFBs), are susceptible to crossover and require expensive ion exchange membranes (IEMs) to achieve long lifetimes. Redox-active polymer (RAP) solutions show promise as candidate electrolytes to mitigate crossover through size-exclusion, enabling the use of porous separators instead of IEMs. Here, poly(vinylbenzyl ethyl viologen) is studied as a surrogate RAP for RFBs. For oxidized RAPs, ionic conductivity varies weakly between 1.6 and 2.1 S m-1 for RAP concentrations of 0.13-1.27 mol kg-1 (monomeric repeat unit per kg solvent) and 0.32 mol kg-1 LiBF4 with a minor increase upon reduction. In contrast, viscosity varies between 1.8 and 184.0 mPa s over the same concentration range with weakly shear-thinning rheology independent of oxidation state. Techno-economic analysis is used to quantify reactor cost as a function of electrolyte transport properties for RAP concentrations of 0.13-1.27 mol kg-1, assuming a hypothetical 3V cell and facile kinetics. Among these concentrations, reactor cost is minimized over a current density range of 600-1000 A m-2 with minimum reactor cost between 11-17 per kWh, and pumping pressures below 10 kPa. The predicted low reactor cost of RAP RFBs is enabled by sustained ionic mobility in spite of the high viscosity of concentrated RAP solutions.

Generations and Identity: The Japanese American.

ERIC Educational Resources Information Center

Kitano, Harry H. L.

The story of people of Japanese descent in the United States is told in its historic context. The Japanese came to America with cultural values that differed greatly from the mainstream U.S. society. They were also set apart by appearance. Conflict between Japan and the United States exacerbated the problems between the Japanese Americans and the…

The Work Values of Japanese Women.

ERIC Educational Resources Information Center

Engel, John W.

Empirical studies of Japanese work ethics have tended to focus on male workers while neglecting women. In addition, work values in both Japan and the United States appear to be changing. More information is needed on the work values of American and Japanese female workers. A study was conducted to explore the work ethics of Japanese women and to…

Experimental study on the instability of Pressure Balance Injection System (PBIS)

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

Okamoto, Koji; Teshima, Hideyuki; Madarame, Haruki

1996-06-01

The Passive Safety Reactor has been developed to reduce the construction cost and to improve the safety. Japan Atomic Energy Research institute (JAERI) proposed the System-Integrated Pressurized Water Reactor (SPWR) as a Passive Safety Reactor. In the SPWR design, the Pressure Balanced Injection System (PBIS) was introduced for the passive safety concept. The water with boron in a containment vessel were passively injected into the core by the pressure difference between the containment vessel and reactor vessel at a severe accidental condition. However there are few studies on the thermo-hydraulic characteristics of the PBIS. In this study, the thermal hydraulicsmore » of the PBIS are experimentally investigated using the small scale model. The instability of the injected flow was observed in the adiabatic experiment. The instability was caused by the pressure balance between the two vessels. The mechanism of the instability are discussed, resulting in the good agreement with the experimental results. In the steam experiment, another instability was observed, which was caused by the heat balance in the main tank.« less

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

DOEpatents

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

KENNEDY SPACE CENTER, FLA. - An overview of the Space Station Processing Facility shows workstands and ISS elements. The most recent additions are the Japanese Experiment Module (JEM)’s pressurized module and the Italian-built Node 2. The pressurized module is the first element of the JEM, Japan’s primary contribution to the Space Station, to be delivered to KSC. It will enhance the unique research capabilities of the orbiting complex by providing an additional shirt-sleeve environment for astronauts to conduct science experiments. Node 2 will be installed on the end of the U.S. Lab and provides attach locations for the Japanese laboratory, European laboratory, the Centrifuge Accommodation Module and, later, Multipurpose Logistics Modules. It will provide the primary docking location for the Shuttle when a pressurized mating adapter is attached to Node 2. Installation of the module will complete the U.S. Core of the ISS.

NASA Image and Video Library

2003-06-06

KENNEDY SPACE CENTER, FLA. - An overview of the Space Station Processing Facility shows workstands and ISS elements. The most recent additions are the Japanese Experiment Module (JEM)’s pressurized module and the Italian-built Node 2. The pressurized module is the first element of the JEM, Japan’s primary contribution to the Space Station, to be delivered to KSC. It will enhance the unique research capabilities of the orbiting complex by providing an additional shirt-sleeve environment for astronauts to conduct science experiments. Node 2 will be installed on the end of the U.S. Lab and provides attach locations for the Japanese laboratory, European laboratory, the Centrifuge Accommodation Module and, later, Multipurpose Logistics Modules. It will provide the primary docking location for the Shuttle when a pressurized mating adapter is attached to Node 2. Installation of the module will complete the U.S. Core of the ISS.

KENNEDY SPACE CENTER, FLA. - A view of the Space Station Processing Facility shows workstands and ISS elements. The most recent additions are the Japanese Experiment Module (JEM)’s pressurized module and the Italian-built Node 2. The pressurized module is the first element of the JEM, Japan’s primary contribution to the Space Station, to be delivered to KSC. It will enhance the unique research capabilities of the orbiting complex by providing an additional shirt-sleeve environment for astronauts to conduct science experiments. Node 2 will be installed on the end of the U.S. Lab and provides attach locations for the Japanese laboratory, European laboratory, the Centrifuge Accommodation Module and, later, Multipurpose Logistics Modules. It will provide the primary docking location for the Shuttle when a pressurized mating adapter is attached to Node 2. Installation of the module will complete the U.S. Core of the ISS.

NASA Image and Video Library

2003-06-06

KENNEDY SPACE CENTER, FLA. - A view of the Space Station Processing Facility shows workstands and ISS elements. The most recent additions are the Japanese Experiment Module (JEM)’s pressurized module and the Italian-built Node 2. The pressurized module is the first element of the JEM, Japan’s primary contribution to the Space Station, to be delivered to KSC. It will enhance the unique research capabilities of the orbiting complex by providing an additional shirt-sleeve environment for astronauts to conduct science experiments. Node 2 will be installed on the end of the U.S. Lab and provides attach locations for the Japanese laboratory, European laboratory, the Centrifuge Accommodation Module and, later, Multipurpose Logistics Modules. It will provide the primary docking location for the Shuttle when a pressurized mating adapter is attached to Node 2. Installation of the module will complete the U.S. Core of the ISS.

Technological Diversification of Japanese Industry.

ERIC Educational Resources Information Center

Kodama, Fumio

1986-01-01

Describes an approach for measuring industrial technological diversification behavior. Identifies sectoral patterns of Japanese industry as related to diversification behaviors. Delineates the mechanisms and effectiveness of Japanese corporate and government policies relevant to diversification. (ML)

The Challenges of Increasing Capacity and Diversity in Japanese Higher Education through Proactive Recruitment Strategies

ERIC Educational Resources Information Center

Kuwamura, Akira

2009-01-01

There has been fierce competition for a shrinking pool of high school graduates in the higher education market in Japan in recent years. Along came former Prime Minister Fukuda's plan for an intake of 300,000 international students by the year 2020. These have placed Japanese institutions of higher education under further pressure to sustain their…

Modalities of Infant-Mother Interaction in Japanese, Japanese American Immigrant, and European American Dyads

PubMed Central

Bornstein, Marc H.; Cote, Linda R.; Haynes, O. Maurice; Suwalsky, Joan T. D.; Bakeman, Roger

2011-01-01

Cultural variation in relations and moment-to-moment contingencies of infant-mother person-oriented and object-oriented interactions were examined and compared in 118 Japanese, Japanese American immigrant, and European American dyads with 5.5-month-olds. Infant and mother person-oriented behaviors were positively related in all cultural groups, but infant and mother object-oriented behaviors were positively related only among European Americans. In all groups, infant and mother behaviors within each modality were mutually contingent. Culture moderated lead-lag relations: Japanese infants were more likely than their mothers to respond in object-oriented interactions, European American mothers were more likely than their infants to respond in person-oriented interactions. Japanese American dyads behaved more like European American dyads. Interaction, infant effects, and parent socialization findings are set in cultural and accultural models of transactions between young infants and their mothers. PMID:22860874

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Change in Intra-Abdominal Fat Predicts the Risk of Hypertension in Japanese Americans.

PubMed

Sullivan, Catherine A; Kahn, Steven E; Fujimoto, Wilfred Y; Hayashi, Tomoshige; Leonetti, Donna L; Boyko, Edward J

2015-07-01

In Japanese Americans, intra-abdominal fat area measured by computed tomography is positively associated with the prevalence and incidence of hypertension. Evidence in other populations suggests that other fat areas may be protective. We sought to determine whether a change in specific fat depots predicts the development of hypertension. We prospectively followed up 286 subjects (mean age, 49.5 years; 50.4% men) from the Japanese American Community Diabetes Study for 10 years. At baseline, subjects did not have hypertension (defined as blood pressure ≥140/90 mm Hg) and were not taking blood pressure or glucose-lowering medications. Mid-thigh subcutaneous fat area, abdominal subcutaneous fat area, and intra-abdominal fat area were directly measured by computed tomography at baseline and 5 years. Logistic regression was used to estimate odds of incident hypertension over 10 years in relation to a 5-year change in fat area. The relative odds of developing hypertension for a 5-year increase in intra-abdominal fat was 1.74 (95% confidence interval, 1.28-2.37), after adjusting for age, sex, body mass index, baseline intra-abdominal fat, alcohol use, smoking status, and weekly exercise energy expenditure. This relationship remained significant when adjusted for baseline fasting insulin and 2-hour glucose levels or for diabetes mellitus and pre-diabetes mellitus classification. There were no significant associations between baseline and change in thigh or abdominal subcutaneous fat areas and incident hypertension. In conclusion, in this cohort of Japanese Americans, the risk of developing hypertension is related to the accumulation of intra-abdominal fat rather than the accrual of subcutaneous fat in either the thigh or the abdominal areas. © 2015 American Heart Association, Inc.

Effect of methane partial pressure on the performance of a membrane biofilm reactor coupling methane-dependent denitrification and anammox.

PubMed

Cai, Chen; Hu, Shihu; Chen, Xueming; Ni, Bing-Jie; Pu, Jiaoyang; Yuan, Zhiguo

2018-10-15

Complete nitrogen removal has recently been demonstrated by integrating anaerobic ammonium oxidation (anammox) and denitrifying anaerobic methane oxidation (DAMO) processes. In this work, the effect of methane partial pressure on the performance of a membrane biofilm reactor (MBfR) consisting of DAMO and anammox microorganisms was evaluated. The activities of DAMO archaea and DAMO bacteria in the biofilm increased significantly with increased methane partial pressure, from 367 ± 9 and 58 ± 22 mg-N L -1 d -1 to 580 ± 12 and 222 ± 22 mg-N L -1 d -1 , respectively, while the activity of anammox bacteria only increased slightly, when the methane partial pressure was elevated from 0.24 to 1.39 atm in the short-term batch tests. The results were supported by a long-term (seven weeks) continuous test, when the methane partial pressure was dropped from 1.39 to 0.78 atm. The methane utilization efficiency was always above 96% during both short-term and long-term tests. Taken together, nitrogen removal rate (especially the nitrate reduction rate by DAMO archaea) and methane utilization efficiency could be maintained at high levels in a broad range of methane partial pressure (0.24-1.39 atm in this study). In addition, a previously established DAMO/anammox biofilm model was used to analyze the experimental data. The observed impacts of methane partial pressure on biofilm activity were well explained by the modeling results. These results suggest that methane partial pressure can potentially be used as a manipulated variable to control reaction rates, ultimately to maintain high nitrogen removal efficiency, according to nitrogen loading rate. Copyright © 2018 Elsevier B.V. All rights reserved.

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.

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.

Pretending to Be Japanese: Artistic Play in a Japanese-American Church and Family

ERIC Educational Resources Information Center

Goto, Courtney T.

2008-01-01

With high rates of out-marriage and dwindling need for bilingual worship, Japanese-American churches face a critical question: "Why retain the Japanese part of our identity?" This article explores how one layperson (Naomi Takahashi Goto) draws from her experience as an artist, teacher, and mother to help her congregation answer this question.…

New Frontiers for Japanese Youth

ERIC Educational Resources Information Center

Tucker, Frank H.

1974-01-01

Japanese literature, television, movies, and school texts from 1935 to 1955 are analyzed for their influence and contribution to Japanese youths' pioneering spirit and frontiermindedness. "Asian Affairs" is published by the American-Asian Educational Exchange, New York. (DE)

8 CFR 349.1 - Japanese renunciation of nationality.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 8 Aliens and Nationality 1 2010-01-01 2010-01-01 false Japanese renunciation of nationality. 349.1... NATIONALITY § 349.1 Japanese renunciation of nationality. A Japanese who renounced United States nationality... void, shall complete Form N-576, Supplemental Affidavit to be Submitted with Applications of Japanese...

An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

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

Oßwald, Patrick; Köhler, Markus

A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimentalmore » data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.« less

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Lifestyle characteristics assessment of Japanese in Pittsburgh, USA.

PubMed

Hirooka, Nobutaka; Takedai, Teiichi; D'Amico, Frank

2012-04-01

Lifestyle-related chronic diseases such as cancer and cardiovascular disease are the greatest public health concerns. Evidence shows Japanese immigrants to a westernized environment have higher incidence of lifestyle-related diseases. However, little is known about lifestyle characteristics related to chronic diseases for Japanese in a westernized environment. This study is examining the gap in lifestyle by comparing the lifestyle prevalence for Japanese in the US with the Japanese National Data (the National Health and Nutrition Survey in Japan, J-NHANS) as well as the Japan National Health Promotion in the twenty-first Century (HJ21) goals. Japanese adults were surveyed in Pittsburgh, USA, regarding their lifestyle (e.g., diet, exercise, smoking, stress, alcohol, and oral hygiene). The prevalence was compared with J-NHANS and HJ21 goals. Ninety-three responded (response rate; 97.9%). Japanese men (n = 38) and women (n = 55) in Pittsburgh smoke less than Japanese in Japan (P < 0.001 for both genders). Japanese in Pittsburgh perform less physical activity in daily life and have lower prevalence of walking more than 1 h per day (P < 0.001 for both genders). Japanese women in Pittsburgh have significantly higher prevalence of stress than in Japan (P = 0.004). Japanese men in Pittsburgh do not reach HJ21 goal in weight management, BMI, use of medicine or alcohol to sleep, and sleep quality. Japanese women in Pittsburgh do not reach HJ21 goal in weight management and sleep quality. In conclusion, healthy lifestyle promotion including exercise and physical activity intervention for Japanese living in a westernized environment is warranted.

A Powerful Protector of the Japanese People: The History of the Japanese Hospital in Steveston, British Columbia, Canada,18961942.

PubMed

Vandenberg, Helen

2017-01-01

From 1896 to 1942, a Japanese hospital operated in the village of Steveston, British Columbia, Canada. For the first 4 years, Japanese Methodist missionaries utilized a small mission building as a makeshift hospital, until a larger institution was constructed by the local Japanese Fishermen's Association in 1900. The hospital operated until the Japanese internment, after the attack on Pearl Harbor during World War II. This study offers important commentary about the relationships between health, hospitals, and race in British Columbia during a period of increased immigration and economic upheaval. From the unique perspective of Japanese leaders, this study provides new insight about how Japanese populations negotiated hospital care, despite a context of severe racial discrimination. Japanese populations utilized Christianization, fishing expertise, and hospital work to garner more equitable access to opportunities and resources. This study demonstrates that in addition to providing medical treatment, training grounds for health-care workers, and safe refuge for the sick, hospitals played a significant role in confronting broader racialized inequities in Canada's past.

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

KENNEDY SPACE CENTER, FLA. - Executive Director of NASDA Koji Yamamoto (left) and Center Director Roy Bridges Jr. (right) exchange mementos during Mr. Yamamoto’s visit to KSC. Mr. Bridges also holds the logo of the new Japan Aerospace Exploration Agency, a merger of three Japanese aeronautical and space agencies effective Oct.1, 2003. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module. His visit includes a tour of the Columbia Debris Hangar.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - Executive Director of NASDA Koji Yamamoto (left) and Center Director Roy Bridges Jr. (right) exchange mementos during Mr. Yamamoto’s visit to KSC. Mr. Bridges also holds the logo of the new Japan Aerospace Exploration Agency, a merger of three Japanese aeronautical and space agencies effective Oct.1, 2003. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module. His visit includes a tour of the Columbia Debris Hangar.

KENNEDY SPACE CENTER, FLA. - Executive Director of NASDA Koji Yamamoto (left) is welcomed to KSC by Center Director Roy Bridges Jr. (right). On the table between them is the logo of the new Japan Aerospace Exploration Agency, a merger of three Japanese aeronautical and space agencies effective Oct.1, 2003. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module. His visit includes a tour of the Columbia Debris Hangar.

NASA Image and Video Library

2003-06-12

KENNEDY SPACE CENTER, FLA. - Executive Director of NASDA Koji Yamamoto (left) is welcomed to KSC by Center Director Roy Bridges Jr. (right). On the table between them is the logo of the new Japan Aerospace Exploration Agency, a merger of three Japanese aeronautical and space agencies effective Oct.1, 2003. Mr. Yamamoto is at KSC for a welcome ceremony involving the arrival of the newest Space Station module, the Japanese Experiment Module/pressurized module. His visit includes a tour of the Columbia Debris Hangar.

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Japanese astronaut Koichi Wakata (top left) and technicians watch as a tray is extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

NASA Image and Video Library

2003-09-24

KENNEDY SPACE CENTER, FLA. - In the Space Station Processing Facility, Japanese astronaut Koichi Wakata (top left) and technicians watch as a tray is extended from inside the Pressurized Module, or PM, part of the Japanese Experiment Module (JEM). The PM provides a shirt-sleeve environment in which astronauts on the International Space Station can conduct microgravity experiments. There are a total of 23 racks, including 10 experiment racks, inside the PM providing a power supply, communications, air conditioning, hardware cooling, water control and experiment support functions.

Some Design Issues for an Online Japanese Textbook

ERIC Educational Resources Information Center

Nagata, Noriko

2010-01-01

This paper discusses several design issues in the development of a new online Japanese textbook, called "Robo-Sensei: Japanese Curriculum with Automated Feedback". When it is completed, the new online textbook will present a full Japanese curriculum. It extends a previously published online software program, "Robo-Sensei: Personal Japanese Tutor"…

The Semantics and Pragmatics of Japanese Focus Particles

ERIC Educational Resources Information Center

Hasegawa, Akio

2011-01-01

Japanese has a rich set of focus particles, several exclusive and additive particles, and, in addition, contrastive particles. This thesis provides a formal description of the meanings of Japanese focus particles and addresses two general questions: "What kinds concepts do Japanese focus particles express?" and "Why does Japanese have a larger…

Re-Examining Patriotism in Japanese Education: Analysis of Japanese Elementary School Moral Readers

ERIC Educational Resources Information Center

Anzai, Shinobu

2015-01-01

In 1947 the Fundamental Law of Education (FLE) defined the pacifist principles for post-war Japanese education and was revised in 2006 for the first time in nearly 60 years. The revised FLE stipulates the importance of teaching love for country and region and Japanese culture and traditions with special emphasis on moral education. Today, this…

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

NADH Dehydrogenase Subunit-2 237 Leu/Met Polymorphism Modulates the Effects of Coffee Consumption on the Risk of Hypertension in Middle-Aged Japanese Men

PubMed Central

Kokaze, Akatsuki; Ishikawa, Mamoru; Matsunaga, Naomi; Karita, Kanae; Yoshida, Masao; Ohtsu, Tadahiro; Shirasawa, Takako; Sekii, Hideaki; Ito, Taku; Kawamoto, Teruyoshi; Takashima, Yutaka

2009-01-01

Background Habitual coffee consumption has been reported to lower blood pressure in the Japanese population. The NADH dehydrogenase subunit-2 237 leucine/methionine (ND2-237 Leu/Met) polymorphism is associated with longevity and modifies the effects of alcohol consumption on blood pressure in the Japanese population. The objective of this study was to determine whether this polymorphism also modifies the effects of coffee consumption on blood pressure or the risk of hypertension in middle-aged Japanese men. Methods A total of 398 men (mean age ± standard deviation, 53.8 ± 7.8 years) were selected from among individuals visiting the hospital for regular medical check-ups. Hypertension was defined as a systolic blood pressure ≥140 mm Hg, diastolic blood pressure ≥90 mm Hg, or antihypertensive drug treatment. Polymerase chain reaction-restriction fragment length polymorphism using the restriction enzyme AluI was performed to determine ND2-237 Leu/Met genotype. Results In subjects with ND2-237Leu, coffee consumption was significantly and negatively associated with diastolic blood pressure (P = 0.007). The odds ratio (OR) for hypertension was significantly lower in subjects with ND2-237Leu who consumed 2 or 3 cups of coffee per day than in those who consumed less than 1 cup of coffee per day (OR, 0.517; 95% confidence interval [CI], 0.276 to 0.968; P = 0.039). After adjustment, the OR remained significant (OR = 0.399; 95% CI, 0.184 to 0.869; P = 0.020). Moreover, after adjustment, the OR was significantly lower in subjects with ND2-237Leu who consumed more than 4 cups of coffee per day than in those who consumed less than 1 cup of coffee per day (OR, 0.246; 95% CI, 0.062 to 0.975; P = 0.046). However, the association between ND2-237Met genotype and hypertension did not depend on coffee consumption. Conclusions The present results suggest that the ND2-237 Leu/Met polymorphism modulates the effects of coffee consumption on hypertension risk in middle-aged Japanese

Optimization of Pressurized Oxy-Combustion with Flameless Reactor

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

Malavasi, Massimo; Landegger, Gregory

2014-06-30

Pressurized OxyECombustion is one of the most promising technologies for utility-scale power generation plants. Benefits include the ability to burn low rank coal and capture CO 2. By increasing the flue gas pressure during this process, greater efficiencies are derived from increased quantity and quality of thermal energy recovery. UPA with modeling support from MIT and testing and data verification by Georgia Tech’s Research Center designed and built a 100 kW system capable of demonstrating pressurized oxyEcombustion using a flameless combustor. Wyoming PRB coal was run at 15 and 32 bar. Additional tests were not completed but sampled data demonstratedmore » the viability of the technology over a broader range of operating pressures, Modeling results illustrated a flat efficiency curve over 20 bar, with optimum efficiency achieved at 29 bar. This resulted in a 33% (HHV) efficiency, a 5 points increase in efficiency versus atmospheric oxy-combustion, and a competitive cost of electricity plus greater CO 2 avoidance costs then prior study’s presented. UPA’s operation of the bench-scale system provided evidence that key performance targets were achieved: flue gas sampled at the combustor outlet had non-detectable residual fly ashes, and low levels of SO3 and heavy-metal. These results correspond to prior pressurized oxy-combustion testing completed by IteaEEnel.« less

Scores on morningness-eveningness and sleep habits of Korean students, Japanese students, and Japanese workers.

PubMed

Park, Y M; Matsumoto, K; Seo, Y J; Shinkoda, H; Park, K P

1997-08-01

The Morningness-Eveningness Questionnaire and Life Habits Inventory were given to three groups of the same mean age: 533 Korean students, 468 Japanese students, and 311 Japanese workers. The distributions of scores on the questionnaire for these three groups are normal; however the Japanese students' distribution was slightly skewed towards the Evening type. The self-reported waking times and bedtimes for the three groups were late in the order of Morning, Intermediate, and Evening types. It is noteworthy, however, that the Korean students woke earlier than the Japanese students, and the workers always went to bed and woke earlier than the students. For the groups the variations in bedtime, waking time, and length of sleep were large, the sleep latency was long, and mood of the participants upon waking was bad in the order of the Morning. Intermediate, and Evening types. The scores of the Korean students were distributed more highly in the Morning type than were the Japanese students', but the students' sleep habits in both countries were quite similar. The subjects categorized as Evening types had more irregular sleep habits than those of the Morning type. In comparison with the student groups, Japanese workers of the same mean age had higher scores and slightly different sleep habits. The change in sleep habits could be seen as a result of the demands of employment, and the probable basis for difference in scores.

Blood pressure differences between office and home settings among Japanese normotensive subjects and hypertensive patients.

PubMed

Mori, Hisao; Ukai, Hiroshi; Yamamoto, Hareaki; Yuasa, Shouhei; Suzuki, Yoshiro; Chin, Keiichi; Katsumata, Takuma; Umemura, Satoshi

2017-03-01

This study attempted to clarify the differences in blood pressure (BP) between the office (clinic) and home settings in patients with controlled, sustained, masked or white-coat hypertension. The following formula was used: office mean systolic BP (omSBP)-mean morning home SBP (mmhSBP)/office mean diastolic BP (omDBP)-mean morning home DBP (mmhDBP). The paired t-test was used for statistical analysis. The omSBP-mmhSBP/omDBP-mmhDBP calculation yielded the following results: among normotensive subjects, -1.1±11.2/-1.7±8.5 mm Hg (mean SBP and mean DBP were higher at home than in the office; n=451, P=0.038 in SBP, P=0.000 in DBP); in controlled hypertensive patients, -0.42±10.9/-2.2±8.2 mm Hg (n=1362, P=0.160 in SBP, P=0.000 in DBP); among sustained hypertensive patients, 5.6±14.7/0.048±9.9 mm Hg (n=1370, P=0.000 in SBP, P=0.857 in DBP); in masked hypertensive patients, -15.3±12.9/-9.3±9.5 mm Hg (n=1308, both P=0.000); and among white-coat hypertensive patients, 23.7±13.2/8.2±9.1 mm Hg (n=580, both P=0.000). Our results showed a difference of 5 mm Hg in SBP among sustained hypertensive patients, as recommended by the Japanese Society of Hypertension Guidelines for the Management of Hypertension; however, in other hypertensive patient types, the differences in SBP and DBP between office and home measurements differed by >5 mm Hg. Office and home BP measurements should be interpreted cautiously, keeping in mind the clinical setting.

Japanese Characters in Written Japanese.

ERIC Educational Resources Information Center

Buck, James H.

From the sixth to the eighth century A.D., Japan was the recipient of massive cultural infusions from China. This acceptance of the Chinese pattern included, and to a great extent was based on, the acceptance of the Chinese language. The Chinese writing system was applied to Japanese because there was no other model to follow and in spite of the…

Corrosion Behavior of Carbon Steel Coated with Octadecylamine in the Secondary Circuit of a Pressurized Water Reactor

NASA Astrophysics Data System (ADS)

Jäppinen, Essi; Ikäläinen, Tiina; Järvimäki, Sari; Saario, Timo; Sipilä, Konsta; Bojinov, Martin

2017-12-01

Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

Pressure letdown method and device for coal conversion systems

NASA Technical Reports Server (NTRS)

Kendal, J. M.; Walsh, J. V. (Inventor)

1983-01-01

In combination with a reactor for a coal utilization system, a pressure letdown device accepts from a reactor, a polyphase fluid at an entrance pressure and an entrance velocity, and discharges the fluid from the device at a discharge pressure substantially lower than the entrance pressure and at a discharge temperature and a discharge velocity substantially equal to the entrance temperature and entrance velocity. The device is characterized by a series of pressure letdown stages including several symmetrical baffles, disposed in coaxially nested alignment. In each baffle several ports or apertures of uniform dimensions are defined. The number of ports or apertures for each baffle plate is unique with respect to the number of ports or apertures defined in each of the other baffles. The mass rate of flow for each port is a function of the area of the port, the pressure of the fluid as applied to the port, and a common pressure ratio established across the ports.

An overview of Japanese CELSS research activities

NASA Technical Reports Server (NTRS)

Nitta, Keiji

1987-01-01

Development of Controlled Ecological Life Support System (CELSS) technology is inevitable for future long duration stays of human beings in space, for lunar base construction and for manned Mars flight programs. CELSS functions can be divided into 2 categories, Environmental Control and Material Recycling. Temperature, humidity, total atmospheric pressure and partial pressure of oxygen and carbon dioxide, necessary for all living things, are to be controlled by the environment control function. This function can be performed by technologies already developed and used as the Environment Control Life Support System (ECLSS) of Space Shuttle and Space Station. As for material recycling, matured technologies have not yet been established for fully satisfying the specific metabolic requirements of each living thing including human beings. Therefore, research activities for establishing CELSS technology should be focused on material recycling technologies using biological systems such as plants and animals and physico-chemical systems, for example, a gas recycling system, a water purifying and recycling system and a waste management system. Japanese research activities were conducted and will be continued accordingly.

Characteristics of a novel nanosecond DBD microplasma reactor for flow applications

NASA Astrophysics Data System (ADS)

Elkholy, A.; Nijdam, S.; van Veldhuizen, E.; Dam, N.; van Oijen, J.; Ebert, U.; de Goey, L. Philip H.

2018-05-01

We present a novel microplasma flow reactor using a dielectric barrier discharge (DBD) driven by repetitive nanosecond high-voltage pulses. Our DBD-based geometry can generate a non-thermal plasma discharge at atmospheric pressure and below in a regular pattern of micro-channels. This reactor can work continuously up to about 100 min in air, depending on the pulse repetition rate and operating pressure. We here present the geometry and main characteristics of the reactor. Pulse energies of 1.46 and 1.3 μJ per channel at atmospheric pressure and 50 mbar, respectively, have been determined by time-resolved measurements of current and voltage. Time-resolved optical emission spectroscopy measurements have been performed to calculate the relative species concentrations and temperatures (vibrational and rotational) of the discharge. The effects of the operating pressure and flow velocity on the discharge intensity have been investigated. In addition, the effective reduced electric field strength {(E/N)}eff} has been obtained from the intensity ratio of vibronic emission bands of molecular nitrogen at different operating pressures and different locations. The derived {(E/N)}eff} increases gradually from about 550 to 4600 Td when decreasing the pressure from 1 bar to 100 mbar. Below 100 mbar, further pressure reduction results in a significant increase in {(E/N)}eff} up to about 10000 Td at 50 mbar.

Method for automatically scramming a nuclear reactor

DOEpatents

Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

2005-12-27

An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

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

Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

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

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.

1992-03-01

The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences,more » VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the primary acceptance criterion'' in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.« less

Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

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

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.

1992-03-01

The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences,more » VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the ``primary acceptance criterion`` in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.« less

Japanese/Korean Linguistics, Volume 8.

ERIC Educational Resources Information Center

Silva, David J., Ed.

A collection of research in Japanese and Korean linguistics includes: "Repetition, Reformulation, and Definitions: Prosodic Indexes of Elaboration in Japanese" (Mieko Banno); "Projection of Talk Using Language, Intonation, Deictic and Iconic Gestures and Other Body Movements" (Keiko Emmett); "Turn-taking in Japanese…

Geochemical aspects of some Japanese lavas.

NASA Technical Reports Server (NTRS)

Philpotts, J. A.; Martin, W.; Schnetzler, C. C.

1971-01-01

K, Rb, Sr, Ba and rare-earth concentrations in some Japanese lavas have been determined by mass-spectrometric stable-isotope dilution. The samples fall into three rare-earth groups corresponding to tholeiitic, high alumina and alkali basalts. Japanese tholeiites have trace element characteristics similar to those of oceanic ridge tholeiites except for distinctly higher relative concentrations of Ba. Japanese lavas may result from various degrees of partial fusion of amphibole eclogite.

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Phylogeography of Japanese horse chestnut (Aesculus turbinata) in the Japanese Archipelago based on chloroplast DNA haplotypes.

PubMed

Sugahara, Kanako; Kaneko, Yuko; Ito, Satoshi; Yamanaka, Keisuke; Sakio, Hitoshi; Hoshizaki, Kazuhiko; Suzuki, Wajiro; Yamanaka, Norikazu; Setoguchi, Hiroaki

2011-01-01

Japanese horse chestnut (Aesculus turbinata: Hippocastanaceae) is one of the typical woody plants that grow in temperate riparian forests in the Japanese Archipelago. To analyze the phylogeography of this plant in the Japanese Archipelago, we determined cpDNA haplotypes for 337 samples from 55 populations covering the entire distribution range. Based on 1,313 bp of two spacers, we determined ten haplotypes that are distinguished from adjacent haplotypes by one or two steps. Most of the populations had a single haplotype, suggesting low diversity. Spatial analysis of molecular variance suggested three obvious phylogeographic structures in western Japan, where Japanese horse chestnut is scattered and isolated in mountainous areas. Conversely, no clear phylogeographic structure was observed from the northern to the southern limit of this species, including eastern Japan, where this plant is more common. Rare and private haplotypes were also found in southwestern Japan, where Japanese horse chestnuts are distributed sparsely. These findings imply that western Japan might have maintained a relatively large habitat for A. turbinata during the Quaternary climatic oscillations, while northerly regions could not.

Iron catalyst chemistry in modeling a high-pressure carbon monoxide nanotube reactor

NASA Technical Reports Server (NTRS)

Scott, Carl D.; Povitsky, Alexander; Dateo, Christopher; Gokcen, Tahir; Willis, Peter A.; Smalley, Richard E.

2003-01-01

The high-pressure carbon monoxide (HiPco) technique for producing single-wall carbon nanotubes (SWNTs) is analyzed with the use of a chemical reaction model coupled with flow properties calculated along streamlines, calculated by the FLUENT code for pure carbon monoxide. Cold iron pentacarbonyl, diluted in CO at about 30 atmospheres, is injected into a conical mixing zone, where hot CO is also introduced via three jets at 30 degrees with respect to the axis. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Then iron nucleates and forms clusters that catalyze the formation of SWNTs by a disproportionation reaction (Boudouard) of CO on Fe-containing clusters. Alternative nucleation rates are estimated from the theory of hard sphere collision dynamics with an activation energy barrier. The rate coefficient for carbon nanotube growth is estimated from activation energies in the literature. The calculated growth was found be about an order of magnitude greater than measured, regardless of the nucleation rate. A study of cluster formation in an incubation zone prior to injection into the reactor shows that direct dimer formation from Fe atoms is not as important as formation via an exchange reaction of Fe with CO in FeCO.

Iron catalyst chemistry in modeling a high-pressure carbon monoxide nanotube reactor.

PubMed

Scott, Carl D; Povitsky, Alexander; Dateo, Christopher; Gökçen, Tahir; Willis, Peter A; Smalley, Richard E

2003-01-01

The high-pressure carbon monoxide (HiPco) technique for producing single-wall carbon nanotubes (SWNTs) is analyzed with the use of a chemical reaction model coupled with flow properties calculated along streamlines, calculated by the FLUENT code for pure carbon monoxide. Cold iron pentacarbonyl, diluted in CO at about 30 atmospheres, is injected into a conical mixing zone, where hot CO is also introduced via three jets at 30 degrees with respect to the axis. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Then iron nucleates and forms clusters that catalyze the formation of SWNTs by a disproportionation reaction (Boudouard) of CO on Fe-containing clusters. Alternative nucleation rates are estimated from the theory of hard sphere collision dynamics with an activation energy barrier. The rate coefficient for carbon nanotube growth is estimated from activation energies in the literature. The calculated growth was found be about an order of magnitude greater than measured, regardless of the nucleation rate. A study of cluster formation in an incubation zone prior to injection into the reactor shows that direct dimer formation from Fe atoms is not as important as formation via an exchange reaction of Fe with CO in FeCO.

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

Effects of triple combination therapy with azilsartan/amlodipine/hydrochlorothiazide on office/home blood pressure: a randomized-controlled trial in Japanese essential hypertensive patients.

PubMed

Rakugi, Hiromi; Shimizu, Kohei; Sano, Yuhei; Nishiyama, Yuya; Kinugawa, Yoshinobu; Terashio, Souhei

2018-04-01

The efficacy and safety of triple therapy with azilsartan (AZI), amlodipine besylate (AML), and hydrochlorothiazide (HCTZ) compared with dual therapy with AZI/AML or HCTZ monotherapy were evaluated in Japanese essential hypertensive patients in a double-blinded manner. A total of 353 patients with office blood pressure (BP) of at least 150/95 mmHg were randomized to a 10-week treatment with AZI/AML/HCTZ 20/5/12.5 mg, AZI/AML/HCTZ 20/5/6.25 mg, AZI/AML 20/5 mg, HCTZ 12.5 mg, or HCTZ 6.25 mg. The mean change from baseline in office diastolic/systolic BPs at week 10 was -25.9/-41.4, -24.9/-38.6, and -22.4/-34.5 mmHg in the AZI/AML/HCTZ 20/5/12.5 mg, AZI/AML/HCTZ 20/5/6.25 mg, and AZI/AML 20/5 mg groups, respectively. AZI/AML/HCTZ 20/5/12.5 mg led to a significantly greater reduction in diastolic and systolic BP than the dual therapy. In addition, the change in home diastolic BP measured with telemetry devices showed a significant difference between the two triple therapy groups. The incidences of adverse events except dizziness postural were similar among the treatment groups in the triple therapy groups. Triple therapy with AZI/AML/HCTZ 20/5/12.5 mg shows a greater antihypertensive effect than the dual therapy and has acceptable safety profiles for Japanese essential hypertensive patients. It was also observed that home BP measurement by automated telemetry could detect changes in BP that were not detected in office BP measurement, although further investigation is needed.

Efficacy and safety of 10-mg azilsartan compared with 8-mg candesartan cilexetil in Japanese patients with hypertension: a randomized crossover non-inferiority trial.

PubMed

Takahara, Mitsuyoshi; Shiraiwa, Toshihiko; Shindo, Megumi; Arai, Akie; Kusuda, Yuko; Katakami, Naoto; Kaneto, Hideaki; Matsuoka, Taka-Aki; Shimomura, Iichiro

2014-09-01

We investigated whether 10 mg per day of azilsartan, one-half of the normal dosage, would be non-inferior to 8 mg per day of candesartan cilexetil for controlling blood pressure in Japanese patients with hypertension. In this open-label, randomized, crossover trial, 309 hypertensive Japanese adults treated with 8-mg candesartan cilexetil were randomized into two arms and received either 10-mg azilsartan or 8-mg candesartan cilexetil in a crossover manner. The primary efficacy outcome was systolic blood pressure, and the margin of non-inferiority was set to be 2.5 mm Hg. The participants were 67±11 years old, and 180 (58%) were male. The baseline systolic and diastolic blood pressure levels were 127.1±13.2 and 69.7±11.2 mm Hg, respectively. During the study period, the difference in systolic blood pressure between the treatments with 10-mg azilsartan and 8-mg candesartan cilexetil was -1.7 mm Hg, with the two-sided 95% confidence interval (CI) ranged from -3.2 to -0.2 mm Hg. The upper boundary of the 95% CI was below the margin of 2.5 mm Hg, confirming the non-inferiority of 10-mg azilsartan to 8-mg candesartan cilexetil. The difference also reached significance (P=0.037). The corresponding difference in diastolic blood pressure was -1.4 (95% CI: -2.4 to -0.4) mm Hg (P=0.006). Treatment with 10-mg azilsartan was similar to 8-mg candesartan cilexetil in its association with rare adverse events. In conclusion, 10-mg azilsartan was non-inferior to 8-mg candesartan cilexetil for controlling systolic blood pressure in Japanese hypertensive patients already being treated with 8-mg candesartan cilexetil.

Support arrangement for core modules of nuclear reactors

DOEpatents

Bollinger, Lawrence R.

1987-01-01

A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

How the Japanese work.

PubMed

Chambers, D W

1998-01-01

The Japanese do not work harder or even use different approaches so much as they aim for a different result--one that balances process and results and extends the definition of quality beyond the product itself to include cost and convenience to the customer as well. Ten methods of the Japanese kaizen culture of work are presented with applications and contrasts to American dentistry.

Plasma reactor waste management systems

NASA Technical Reports Server (NTRS)

Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

1992-01-01

The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

Westinghouse Small Modular Reactor nuclear steam supply system design

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

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

The wound/burn guidelines - 2: Guidelines for the diagnosis and treatment for pressure ulcers.

PubMed

Tachibana, Takao; Imafuku, Shinichi; Irisawa, Ryokichi; Ohtsuka, Masaki; Kadono, Takafumi; Fujiwara, Hiroshi; Asano, Yoshihide; Abe, Masatoshi; Ishii, Takayuki; Isei, Taiki; Ito, Takaaki; Inoue, Yuji; Ohtsuka, Mikio; Ogawa, Fumihide; Kodera, Masanari; Kawakami, Tamihiro; Kawaguchi, Masakazu; Kukino, Ryuichi; Kono, Takeshi; Sakai, Keisuke; Takahara, Masakazu; Tanioka, Miki; Nakanishi, Takeshi; Nakamura, Yasuhiro; Hashimoto, Akira; Hasegawa, Minoru; Hayashi, Masahiro; Fujimoto, Manabu; Maekawa, Takeo; Matsuo, Koma; Madokoro, Naoki; Yamasaki, Osamu; Yoshino, Yuichiro; Le Pavoux, Andres; Ihn, Hironobu

2016-05-01

The Wound/Burn Guidelines Committee consists of members commissioned by the Board of Directors of the Japanese Dermatological Association (JDA). It held several meetings and evaluations in writing since October 2008, and drafted five guidelines for the diagnosis and treatment including commentaries on wounds in general and the Guidelines for the Diagnosis and Treatment for Pressure Ulcers by taking opinions of the Scientific Committee and Board of Directors of JDA into consideration. © 2016 Japanese Dermatological Association.

Japanese Language School: Aid or Hindrance to the Americanization of Japanese Americans in Hawaii?

ERIC Educational Resources Information Center

Shoho, Alan R.

A study examined the experiences of 60 Japanese immigrants to Hawaii (Niseis), aged 61-80, who attended Japanese-language schools as children. Using a case study oral history approach, the study gathered oral testimonies through semi-structured interviews. Historical documents were also used as primary sources of information about the schools.…

Eating attitudes and body image in ethnic Japanese and Caucasian adolescent girls in the city of São Paulo, Brazil.

PubMed

Sampei, Míriam A; Sigulem, Dirce M; Novo, Neil F; Juliano, Yara; Colugnati, Fernando A B

2009-01-01

Despite investigations into the rapid increase in eating disorders across diverse ethnic groups, conclusions concerning ethnicity and eating disorders are contradictory. The objective of the present study was to investigate eating attitudes in ethnic Japanese and Caucasian adolescents in Brazil. The influence of body mass index (BMI), menarche and social-affective relationships on the development of eating disorders was also assessed. Questionnaires evaluating the incidence of eating disorders and the influence of social-affective relationships were applied to 544 Japanese-Brazilian and Caucasian adolescent girls: 10 to 11-year-old Japanese-Brazilian (n = 122) and Caucasian (n = 176) pre-menarcheal adolescents, and 16 to 17-year-old Japanese-Brazilian (n = 71) and Caucasian (n = 175) post-menarcheal adolescents. Caucasian girls obtained higher scores on the Eating Attitudes Test (EAT-26), showed greater body image dissatisfaction, dieted more often and had more diet models introduced by their mothers and peers than the Japanese-Brazilian girls. CONCLUSION The Caucasian adolescents overall appeared to be more sensitive to aesthetic and social pressures regarding body image than the Japanese adolescents. The high incidence of EAT-26 scores above 20 in the Caucasian pre-menarcheal group indicates that individual body image concerns are developing at an earlier age. Multiple logistic regression revealed several associations between mother-teen interactions and the development of abnormal eating attitudes.

Japanese encephalitis - the prospects for new treatments.

PubMed

Turtle, Lance; Solomon, Tom

2018-04-26

Japanese encephalitis is a mosquito-borne disease that occurs in Asia and is caused by Japanese encephalitis virus (JEV), a member of the genus Flavivirus. Although many flaviviruses can cause encephalitis, JEV causes particularly severe neurological manifestations. The virus causes loss of more disability-adjusted life years than any other arthropod-borne virus owing to the frequent neurological sequelae of the condition. Despite substantial advances in our understanding of Japanese encephalitis from in vitro studies and animal models, studies of pathogenesis and treatment in humans are lagging behind. Few mechanistic studies have been conducted in humans, and only four clinical trials of therapies for Japanese encephalitis have taken place in the past 10 years despite an estimated incidence of 69,000 cases per year. Previous trials for Japanese encephalitis might have been too small to detect important benefits of potential treatments. Many potential treatment targets exist for Japanese encephalitis, and pathogenesis and virological studies have uncovered mechanisms by which these drugs could work. In this Review, we summarize the epidemiology, clinical features, prevention and treatment of Japanese encephalitis and focus on potential new therapeutic strategies, based on repurposing existing compounds that are already suitable for human use and could be trialled without delay. We use our newly improved understanding of Japanese encephalitis pathogenesis to posit potential treatments and outline some of the many challenges that remain in tackling the disease in humans.

Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

NASA Astrophysics Data System (ADS)

Han, Guangdong; Lu, Zhanpeng; Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu

2015-12-01

The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen.

Validating the Japanese translation of the Force and Motion Conceptual Evaluation and comparing performance levels of American and Japanese students

NASA Astrophysics Data System (ADS)

Ishimoto, Michi; Thornton, Ronald K.; Sokoloff, David R.

2014-12-01

This study assesses the Japanese translation of the Force and Motion Conceptual Evaluation (FMCE). Researchers are often interested in comparing the conceptual ideas of students with different cultural backgrounds. The FMCE has been useful in identifying the concepts of English-speaking students from different backgrounds. To identify effectively the conceptual ideas of Japanese students and to compare them to those of their English-speaking counterparts, more work is required. Because of differences between the Japanese and English languages, and between the Japanese and American educational systems, it is important to assess the Japanese translation of the FMCE, a conceptual evaluation originally developed in English for American students. To assess its appropriateness, we examined the performance of a large sample of students on the translated version of the FMCE and then compared the results to those of English-speaking students. The data comprise the pretest results of 1095 students, most of whom were first-year students at a midlevel engineering school between 2003 and 2012. Basic statistics and the classical test theory indices of the translated FMCE indicate that its reliability and discrimination are appropriate to assess Japanese students' concepts about force and motion. In general, the preconcepts of Japanese students assessed with the Japanese translation of the FMCE are quite similar to those of American students assessed with the FMCE, thereby supporting the validity of the translated version. However, our findings do show (1) that only a small percentage of Japanese students grasped Newtonian concepts and (2) that the percentage of Japanese students who used two different concept models together to answer some questions seems to be higher than that of American students.

"A Powerful Protector of the Japanese People": The History of the Japanese Hospital in Steveston, British Columbia, Canada,1896-1942.

PubMed

Vandenberg, Helen

2017-01-01

From 1896 to 1942, a Japanese hospital operated in the village of Steveston, British Columbia, Canada. For the first 4 years, Japanese Methodist missionaries utilized a small mission building as a makeshift hospital, until a larger institution was constructed by the local Japanese Fishermen's Association in 1900. The hospital operated until the Japanese internment, after the attack on Pearl Harbor during World War II. This study offers important commentary about the relationships between health, hospitals, and race in British Columbia during a period of increased immigration and economic upheaval. From the unique perspective of Japanese leaders, this study provides new insight about how Japanese populations negotiated hospital care, despite a context of severe racial discrimination. Japanese populations utilized Christianization, fishing expertise, and hospital work to garner more equitable access to opportunities and resources. This study demonstrates that in addition to providing medical treatment, training grounds for health-care workers, and safe refuge for the sick, hospitals played a significant role in confronting broader racialized inequities in Canada's past.

The Evolution of a Japanese Theory of Conflict Management and Implications for Japanese Foreign Policy

DTIC Science & Technology

2001-12-01

This thesis explores whether there is a uniquely Japanese method of conflict management Given the delicate balance of stability in Northeast Asia...Japanese leadership needs to use conflict management tools to resolve territorial claims with the governments of China, Russia, and South Korea, Given

Cheilitis Glandularis: Two Case Reports of Asian-Japanese Men and Literature Review of Japanese Cases

PubMed Central

Yanagawa, Toru; Yamaguchi, Akira; Harada, Hiroyuki; Yamagata, Kenji; Ishibashi, Naomi; Noguchi, Masayuki; Onizawa, Kojiro; Bukawa, Hiroki

2011-01-01

Cheilitis glandularis (CG) is a rare disorder characterized by swelling of the lip with hyperplasia of the labial salivary glands. CG is most frequently encountered in the lower lip, in middle-aged to older Caucasian men; however Asian cases were rarely reported. In this paper we present two cases of CG in Asian-Japanese men. One was a 23-year-old male with CG of the superficial suppurative type. The other was a 54-year-old male with deep suppurative type. We also reviewed the Japanese cases of CG in the literature and discussed about clinical feature of Japanese CG. PMID:21991474

THERMAL PROPERTIES AND HEATING AND COOLING DURABILITY OF REACTOR SHIELDING CONCRETE (in Japanese)

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

Hosoi, J.; Chujo, K.; Saji, K.

1959-01-01

A study was made of the thermal properties of various concretes made of domestic raw materials for radiation shields of a power reactor and of a high- flux research reactor. The results of measurements of thermal expansion coefficient, specific heat, thermal diffusivity, thermal conductivity, cyclical heating, and cooling durability are described. Relationships between thermal properties and durability are discussed and several photographs of the concretes are given. It is shown that the heating and cooling durability of such a concrete which has a large thermal expansion coefficient or a considerable difference between the thermal expansion of coarse aggregate and themore » one of cement mortar part or aggregates of lower strength is very poor. The decreasing rates of bending strength and dynamical modulus of elasticity and the residual elongation of the concrete tested show interesting relations with the modified thermal stress resistance factor containing a ratio of bending strength and thermal expansion coefficient. The thermal stress resistance factor seems to depend on the conditions of heat transfer on the surface and on heat release in the concrete. (auth)« less

Persistence of Ethnicity: The Japanese of Colorado.

ERIC Educational Resources Information Center

Endo, Russell

This paper presents an overview of the history of Japanese in Colorado. Japanese immigrants first came to Colorado between 1900 and 1910 as railroad laborers. Some became coal miners in southern Colorado; most others became farm laborers. Although the Japanese population during this period was small, communities developed in several locales. The…

Morning and Evening Home Blood Pressure and Risks of Incident Stroke and Coronary Artery Disease in the Japanese General Practice Population: The Japan Morning Surge-Home Blood Pressure Study.

PubMed

Hoshide, Satoshi; Yano, Yuichiro; Haimoto, Hajime; Yamagiwa, Kayo; Uchiba, Kiyoshi; Nagasaka, Shoichiro; Matsui, Yoshio; Nakamura, Akira; Fukutomi, Motoki; Eguchi, Kazuo; Ishikawa, Joji; Kario, Kazuomi

2016-07-01

Our aim is to determine the optimal time schedule for home blood pressure (BP) monitoring that best predicts stroke and coronary artery disease in general practice. The Japan Morning Surge-Home Blood Pressure (J-HOP) study is a nationwide practice-based study that included 4310 Japanese with a history of or risk factors for cardiovascular disease, or both (mean age, 65 years; 79% used antihypertensive medication). Home BP measures were taken twice daily (morning and evening) over 14 days at baseline. During a mean follow-up of 4 years (16 929 person-years), 74 stroke and 77 coronary artery disease events occurred. Morning systolic BP (SBP) improved the discrimination of incident stroke (C statistics, 0.802; 95% confidence interval, 0.692-0.911) beyond traditional risk factors including office SBP (0.756; 0.646-0.866), whereas the changes were smaller with evening SBP (0.764; 0.653-0.874). The addition of evening SBP to the model (including traditional risk factors plus morning SBP) significantly reduced the discrimination of incident stroke (C statistics difference, -0.008; 95% confidence interval: -0.015 to -0.008; P=0.03). The category-free net reclassification improvement (0.3606; 95% confidence interval, 0.1317-0.5896), absolute integrated discrimination improvement (0.015; SE, 0.005), and relative integrated discrimination improvement (58.3%; all P<0.01) with the addition of morning SBP to the model (including traditional risk factors) were greater than those with evening SBP and with combined morning and evening SBP. Neither morning nor evening SBP improved coronary artery disease risk prediction. Morning home SBP itself should be evaluated to ensure best stroke prediction in clinical practice, at least in Japan. This should be confirmed in the different ethnic groups. URL: http://www.umin.ac.jp/ctr/. Unique identifier: UMIN000000894. © 2016 American Heart Association, Inc.

The Effects of Aroma Foot Massage on Blood Pressure and Anxiety in Japanese Community-Dwelling Men and Women: A Crossover Randomized Controlled Trial.

PubMed

Eguchi, Eri; Funakubo, Narumi; Tomooka, Kiyohide; Ohira, Tetsuya; Ogino, Keiki; Tanigawa, Takeshi

2016-01-01

The aim of this study was to investigate the effects of aroma foot massage on blood pressure, anxiety, and health-related quality of life (QOL) in Japanese community-dwelling men and women using a crossover randomized controlled trial. Fifty-seven eligible participants (5 men and 52 women) aged 27 to 72 were randomly divided into 2 intervention groups (group A: n = 29; group B: n = 28) to participate in aroma foot massages 12 times during the 4-week intervention period. Systolic and diastolic blood pressure (SBP and DBP, respectively), heart rate, state anxiety, and health-related QOL were measured at the baseline, 4-week follow-up, and 8-week follow-up. The effects of the aroma foot massage intervention on these factors and the proportion of participants with anxiety were analyzed using a linear mixed-effect model for a crossover design adjusted for participant and period effects. Furthermore, the relationship between the changes in SBP and state anxiety among participants with relieved anxiety was assessed using a linear regression model. Aroma foot massage significantly decreased the mean SBP (p = 0.02), DBP (p = 0.006), and state anxiety (p = 0.003) as well as the proportion of participants with anxiety (p = 0.003). Although it was not statistically significant (p = 0.088), aroma foot massage also increased the score of mental health-related QOL. The change in SBP had a significant and positive correlation with the change in state anxiety (p = 0.01) among participants with relieved anxiety. The self-administered aroma foot massage intervention significantly decreased the mean SBP and DBP as well as the state anxiety score, and tended to increase the mental health-related QOL scores. The results suggest that aroma foot massage may be an easy and effective way to improve mental health and blood pressure. University Hospital Medical Information Network 000014260.

76 FR 18586 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on United...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-04

... as technical reports related to the Gas Turbine Generator design. The Subcommittee will hear... Subcommittee on United States-Advanced Pressurized Water Reactor (US-APWR); Notice of Meeting The ACRS Subcommittee on United States-Advanced Pressurized Water Reactor (US-APWR) will hold a meeting on April 22...

Long working hours and risk for hypertension in Japanese male white collar workers.

PubMed

Nakanishi, N; Yoshida, H; Nagano, K; Kawashimo, H; Nakamura, K; Tatara, K

2001-05-01

To evaluate the association of long working hours with the risk for hypertension. A five year prospective cohort study. Work site in Osaka, Japan. 941 hypertension free Japanese male white collar workers aged 35-54 years were prospectively examined by serial annual health examinations. Men in whom borderline hypertension and hypertension were found during repeated surveys were defined as incidental cases of borderline hypertension and hypertension. 336 and 88 men developed hypertension above the borderline level and definite hypertension during the 3940 and 4531 person years, respectively. After controlling for potential predictors of hypertension, the relative risk for hypertension above the borderline level, compared with those who worked < 8.0 hours per day, was 0.63 (95% confidence intervals (CI): 0.43, 0.91) for those who worked 10.0-10.9 hours per day and 0.48 (95% CI: 0.31, 0.74) for those who worked > or = 11.0 hours per day. The relative risk for definite hypertension, compared with those who worked < 8.0 hours per day, was 0.33 (95% CI: 0.11, 0.95) for those who worked > or = 11.0 hours per day. The multivariate adjusted slopes of diastolic blood pressure (DBP) and mean arterial blood pressure (MABP) during five years of follow up decreased as working hours per day increased. From the multiple regression analyses, working hours per day remained as an independent negative factor for the slopes of systolic blood pressure, DBP, and MABP. These results indicate that long working hours are negatively associated with the risk for hypertension in Japanese male white collar workers.

Inequities in Japanese Urban Schools

ERIC Educational Resources Information Center

Gordon, June A.

2005-01-01

Interviews with Japanese public school educators allow a distinctive view of how the continuing economic decline in Japan has affected educational motivation and decision-making among students and parents. The nature of socioeconomic stratification within Japanese educational opportunity is seen as a continuing situation exacerbated by the costs…

Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

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

Bowman, S.M.

1995-01-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies inmore » the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical

Association Analysis of FOXO3 Longevity Variants With Blood Pressure and Essential Hypertension.

PubMed

Morris, Brian J; Chen, Randi; Donlon, Timothy A; Evans, Daniel S; Tranah, Gregory J; Parimi, Neeta; Ehret, Georg B; Newton-Cheh, Christopher; Seto, Todd; Willcox, D Craig; Masaki, Kamal H; Kamide, Kei; Ryuno, Hirochika; Oguro, Ryosuke; Nakama, Chikako; Kabayama, Mai; Yamamoto, Koichi; Sugimoto, Ken; Ikebe, Kazunori; Masui, Yukie; Arai, Yasumichi; Ishizaki, Tatsuro; Gondo, Yasuyuki; Rakugi, Hiromi; Willcox, Bradley J

2016-11-01

The minor alleles of 3 FOXO3 single nucleotide polymorphisms (SNPs)- rs2802292 , rs2253310 , and rs2802288 -are associated with human longevity. The aim of the present study was to test these SNPs for association with blood pressure (BP) and essential hypertension (EHT). In a primary study involving Americans of Japanese ancestry drawn from the Family Blood Pressure Program II we genotyped 411 female and 432 male subjects aged 40-79 years and tested for statistical association by contingency table analysis and generalized linear models that included logistic regression adjusting for sibling correlation in the data set. Replication of rs2802292 with EHT was attempted in Japanese SONIC study subjects and of each SNP in a meta-analysis of genome-wide association studies of BP in individuals of European ancestry. In Americans of Japanese ancestry, women homozygous for the longevity-associated (minor) allele of each FOXO3 SNP had 6mm Hg lower systolic BP and 3mm Hg lower diastolic BP compared with major allele homozygotes (Bonferroni corrected P < 0.05 and >0.05, respectively). Frequencies of minor allele homozygotes were 3.3-3.9% in women with EHT compared with 9.5-9.6% in normotensive women ( P = 0.03-0.04; haplotype analysis P = 0.0002). No association with BP or EHT was evident in males. An association with EHT was seen for the minor allele of rs2802292 in the Japanese SONIC cohort ( P = 0.03), while in European subjects the minor allele of each SNP was associated with higher systolic and diastolic BP. Longevity-associated FOXO3 variants may be associated with lower BP and EHT in Japanese women.

Low-pressure hydrogen plasmas explored using a global model

NASA Astrophysics Data System (ADS)

Samuell, Cameron M.; Corr, Cormac S.

2016-02-01

Low-pressure hydrogen plasmas have found applications in a variety of technology areas including fusion, neutral beam injection and material processing applications. To better understand these discharges, a global model is developed to predict the behaviour of electrons, ground-state atomic and molecular hydrogen, three positive ion species (H+, \\text{H}2+ , and \\text{H}3+ ), a single negative ion species (H-), and fourteen vibrationally excited states of molecular hydrogen ({{\\text{H}}2}≤ft(\\upsilon =1\\right. -14)). The model is validated by comparison with experimental results from a planar inductively coupled GEC reference cell and subsequently applied to the MAGPIE linear helicon reactor. The MAGPIE reactor is investigated for a range of pressures from 1 to 100 mTorr and powers up to 5 kW. With increasing power between 50 W and 5 kW at 10 mTorr the density of all charged species increases as well as the dissociative fraction while the electron temperature remains almost constant at around 3 eV. For gas pressures from 1-100 mTorr at an input power of 1 kW, the electron density remains almost constant, the electron temperature and dissociative fraction decreases, while \\text{H}3+ density increases in density and also dominates amongst ion species. Across these power and pressure scans, electronegativity remains approximately constant at around 2.5%. The power and pressure determines the dominant ion species in the plasma with \\text{H}3+ observed to dominate at high pressures and low powers whereas H+ tends to be dominant at low pressures and high powers. A sensitivity analysis is used to demonstrate how experimental parameters (power, pressure, reactor wall material, geometry etc) influence individual species’ density as well as the electron temperature. Physical reactor changes including the length, radius and wall recombination coefficient are found to have the largest influence on outputs obtained from the model.

Sexual partner preference in female Japanese macaques.

PubMed

Vasey, Paul L

2002-02-01

Whether animals ever exhibit a preference for same-sex sexual partners is a subject of debate. Japanese macaques represent excellent models for examining issues related to sexual preference in animals because females, in certain populations, routinely engage in both heterosexual and homosexual behavior over the course of their life spans. Multiple lines of evidence indicate that female homosexual behavior in Japanese macaques is a sexual behavior, not a sociosexual one. Additional evidence indicates that female Japanese macaques do not engage in homosexual behavior simply because acceptable male mates are unavailable or unmotivated to copulate. Patterns of sexual partner choice by female Japanese macaques that are the focus of intersexual competition indicate that females of this species choose same-sex sexual partners even when they are simultaneously presented with a motivated, opposite-sex alternative. Thus, in some populations of Japanese macaques, females prefer certain same-sex sexual partners relative to certain male mates, and vice versa. Taken together, this evidence suggests that female Japanese macaques are best characterized as bisexual in orientation, not preferentially homosexual or preferentially heterosexual.

Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA GRC

NASA Technical Reports Server (NTRS)

Reid, Terry V.

2015-01-01

Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA GRC. The TDU consists of three subsystems: the Reactor Simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the Heat Exchanger Manifold (HXM). An Annular Linear Induction Pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now), is referred to as the RxSim subsystem and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the CAD model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras turbulence model), and 2.223 kg/sec (using the k-? turbulence model). The computational error of the predictions for the available mass flow is -0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k-epsilon turbulence model) when compared to measured data.

A Japanese Agenda for Management Development.

ERIC Educational Resources Information Center

Lim, Howard

1982-01-01

Discusses myths about the Japanese management styles; what the West can learn from the Japanese; the concept of nonlinear management; and training modules which teach self-discipline, tolerance, and nonlinear management. (CT)

Japanese neuropathy patients with peripheral myelin protein-22 gene aneuploidy

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

Lebo, R.V.; Li, L.Y.; Flandermeyer, R.R.

1994-09-01

Peripheral myelin protein (PMP-22) gene aneuploidy results in Charcot-Marie-Tooth disease Type 1A (CMT1A) and the Hereditary Neuropathy with Liability to Pressure Palsy (HNPP) in Japanese patients as well as Caucasian Americans. Charcot-Marie-Tooth disease (CMT), the most common genetic neuropathy, results when expression of one of at least seven genes is defective. CMT1A, about half of all CMT mutations, is usually associated with a duplication spanning the peripheral myelin protein-22 gene on distal chromosome band 17p11.2. Autosomal dominant HNPP (hereditary pressure and sensory neuropathy, HPSN) results from a deletion of the CMT1A gene region. Multicolor in situ hybridization with PMP-22 genemore » region probe characterized HNPP deletion reliably and detected all different size duplications reported previously. In summary, 72% of 28 Japanese CMT1 (HMSNI) patients tested had the CMT1A duplication, while none of the CMT2 (HMSNII) or CMT3 (HMSNIII) patients had a duplication. Three cases of HNPP were identified by deletion of the CMT1A gene region on chromosome 17p. HNPP and CMT1A have been reported to result simultaneously from the same unequal recombination event. The lower frequency of HNPP compared to CMT1A suggests that HNPP patients have a lower reproductive fitness than CMT1A patients. This result, along with a CMT1A duplication found in an Asian Indian family, demonstrates the broad geographic distribution and high frequency of PMP-22 gene aneuploidy.« less

Asian Pacific Perspectives: Japanese Americans.

ERIC Educational Resources Information Center

Los Angeles Unified School District, CA.

These instructional materials on Japanese Americans for elementary students were developed through the K.E.Y.S. project (Knowledge of English Yields Success). Information is included on early immigrants, their historical and cultural background, and current problems of Japanese Americans. Resource guides describe the purpose of the unit, how to…

Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

NASA Astrophysics Data System (ADS)

Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

2013-04-01

In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

Academic report on burnout among Japanese nurses.

PubMed

Kitaoka, Kazuyo; Masuda, Shinya

2013-12-01

Japanese nurses have increasingly experienced "burnout" in the past several years. Studies on Japanese nurses are required in order to explore how to prevent nursing burnout. The objectives of this report were to: (i) introduce the concept, definition, and measurement of burnout; (ii) look at an overview of the prevalence, possible causes, and consequences of burnout among Japanese nurses; and (iii) explore how to prevent burnout among nurses. The authors and co-researchers have been studying burnout among Japanese workers for more than 15 years. Therefore, previously performed studies were reviewed and summarized. In Japan, approximately 36% of human services professionals, such as nurses, were burned out compared to 18% of civil servants, and 12% of company employees. It was quite obvious that nurses are prone to burnout. The possible causes and consequences of burnout among Japanese nurses were reviewed. Excessive workloads and interpersonal conflict in the workplace were possible causes of burnout among Japanese nurses. The consequences of nurse burnout are potentially very serious, including medical accidents/errors. Issues to prevent nursing burnout were then reviewed. Enhancement of cognitive coping skills for female nurses and problem-solving skills for male nurses could contribute to prevention of burnout in nurses. The authors' previous study revealed that the new model of the organizational context of burnout developed by Leiter and Maslach could be applied to Japanese. Further examination is needed. This report supports the call to scale up burnout prevention strategy for Japanese nurses. © 2012 The Authors. Japan Journal of Nursing Science © 2012 Japan Academy of Nursing Science.

Nuclear reactor

DOEpatents

Wade, Elman E.

1979-01-01

A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

Japanese Neurosurgeons and Microsurgical Anatomy: A Historical Review

PubMed Central

MATSUSHIMA, Toshio; KAWASHIMA, Masatou; MATSUSHIMA, Ken; WANIBUCHI, Masahiko

2015-01-01

Research in microneurosurgical anatomy has contributed to great advances in neurosurgery in the last 40 years. Many Japanese neurosurgeons have traveled abroad to study microsurgical anatomy and played major roles in advancing and spreading the knowledge of anatomy, overcoming their disadvantage that the cadaver study has been strictly limited inside Japan. In Japan, they initiated an educational system for surgical anatomy that has contributed to the development and standardization of Japanese neurosurgery. For example, the Japanese Society for Microsurgical Anatomy started an annual educational meeting in the middle of 1980s and published its proceedings in Japanese every year for approximately 20 years. These are some of the achievements that bring worldwide credit to Japanese neurosurgeons. Not only should Japanese neurosurgeons improve their educational system but they should also contribute to the international education in this field, particularly in Asia. PMID:25797782

Issei: Japanese Immigrants in Hawaii.

ERIC Educational Resources Information Center

Kimura, Yukiko

Coming to Hawaii before July 1, 1924, when the Japanese Exclusion Act became effective, the experiences of the Issei or first generation are described. Divided into four parts, this book examines the experiences of Japanese immigrants in Hawaii from 1885 through 1970. Part 1, "The Formation and Stabilization of the Issei Community,"…

Pair-List Readings in Korean-Japanese, Chinese-Japanese and English-Japanese Interlanguage

ERIC Educational Resources Information Center

Marsden, Heather

2008-01-01

In English and Chinese, questions with a "wh"-object and a universally quantified subject (e.g. "What did everyone buy?") allow an individual answer ("Everyone bought apples.") and a pair-list answer ("Sam bought apples, Jo bought bananas, Sally bought..."). By contrast, the pair-list answer is reportedly unavailable in Japanese and Korean. This…

Identification of Cha o 3 homolog Cry j 4 from Cryptomeria japonica (Japanese cedar) pollen: Limitation of the present Japanese cedar-specific ASIT.

PubMed

Osada, Toshihiro; Tanaka, Yuki; Yamada, Akira; Sasaki, Eiji; Utsugi, Teruhiro

2018-03-07

About one-third of the Japanese population suffers from Japanese cedar pollinosis, which is frequently accompanied by Japanese cypress pollinosis. Recently, a novel major Japanese cypress pollen allergen, Cha o 3, was discovered. However, whether a Cha o 3 homolog is present in Japanese cedar pollen remains to be determined. Western blot analysis was performed using Cha o 3-specific antiserum. In addition, cloning of the gene encoding Cry j 4 was conducted using total cDNA from the male flower of Japanese cedar trees. Allergen potency and cross-reactivity were investigated using a T-cell proliferation assay, basophil activation test, and ImmunoCAP inhibition assay. A low amount of Cha o 3 homolog protein was detected in Japanese cedar pollen extract. The deduced amino acid sequence of Cry j 4 showed 84% identity to that of Cha o 3. Cross-reactivity between Cry j 4 and Cha o 3 was observed at the T cell and IgE levels. Cry j 4 was discovered as a counterpart allergen of Cha o 3 in Japanese cedar pollen, with a relationship similar to that between Cry j 1-Cha o 1 and Cry j 2-Cha o 2. Our findings also suggest that allergen-specific immunotherapy (ASIT) using Japanese cedar pollen extract does not induce adequate immune tolerance to Cha o 3 due to the low amount of Cry j 4 in Japanese cedar pollen. Therefore, ASIT using Cha o 3 or cypress pollen extract coupled with Japanese cedar pollen extract is required in order to optimally control allergy symptoms during Japanese cypress pollen season. Copyright © 2018 Japanese Society of Allergology. Production and hosting by Elsevier B.V. All rights reserved.

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.

Health monitoring of Japanese payload specialist: Autonomic nervous and cardiovascular responses under reduced gravity condition (L-0)

NASA Technical Reports Server (NTRS)

Sekiguchi, Chiharu

1993-01-01

In addition to health monitoring of the Japanese Payload Specialists (PS) during the flight, this investigation also focuses on the changes of cardiovascular hemodynamics during flight which will be conducted under the science collaboration with the Lower Body Negative Pressure (LBNP) Experiment of NASA. For the Japanese, this is an opportunity to examine firsthand the effects of microgravity of human physiology. We are particularly interested in the adaption process and how it relates to space motion sickness and cardiovascular deconditioning. By comparing data from our own experiment to data collected by others, we hope to understand the processes involved and find ways to avoid these problems for future Japanese astronauts onboard Space Station Freedom and other Japanese space ventures. The primary objective of this experiment is to monitor the health condition of Japanese Payload Specialists to maintain a good health status during and after space flight. The second purpose is to investigate the autonomic nervous system's response to space motion sickness. To achieve this, the function of the autonomic nervous system will be monitored using non-invasive techniques. Data obtained will be employed to evaluate the role of autonomic nervous system in space motion sickness and to predict susceptibility to space motion sickness. The third objective is evaluation of the adaption process of the cardiovascular system to microgravity. By observation of the hemodynamics using an echocardiogram we will gain insight on cardiovascular deconditioning. The last objective is to create a data base for use in the health care of Japanese astronauts by obtaining control data in experiment L-O in the SL-J mission.

Reactor vessel lower head integrity

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

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less